ML19298B500

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Amendment 28 to Updated Final Safety Analysis Report, Chapter 13, Conduct of Operations
ML19298B500
Person / Time
Site: Browns Ferry  Tennessee Valley Authority icon.png
Issue date: 10/04/2019
From:
Tennessee Valley Authority
To:
Office of Nuclear Reactor Regulation
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Download: ML19298B500 (184)
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{{#Wiki_filter:BFN-26 CONDUCT OF OPERATIONS TABLE OF CONTENTS 13.0 CONDUCT OF OPERATIONS .................................................................................................. 13.1-1 13.1 Organizational Structure For The Conduct Of Operations ......................................................... 13.1-1 13.1.1 Functions, Responsibilities, and Authorities ............................................................... 13.1-1 13.1.2 Interrelationships With Contractors and Suppliers ..................................................... 13.1-1 13.2 Organization And Responsibility................................................................................................ 13.2-1 13.2.1 Corporate Organization .............................................................................................. 13.2-1 13.2.2 Nuclear Power ........................................................................................................... 13.2-1 13.2.3 Offsite Organizations ................................................................................................. 13.2-1 13.2.4 Onsite Organization ................................................................................................... 13.2-1 13.2.5 Personnel Functions, Responsibilities, and Authorities .............................................. 13.2-1 13.2.6 Qualification Requirements for Nuclear Facility Personnel ........................................ 13.2-2 13.3 Training Programs ..................................................................................................................... 13.3-1 13.3.1 Accredited Training Programs.................................................................................... 13.3-1 13.3.2 General Employee and Fitness for Duty Training Programs ...................................... 13.3-1 13.3.3 Other Training Programs............................................................................................ 13.3-2 13.3.4 References................................................................................................................. 13.3-2 13.4 Preoperational Test Program .................................................................................................... 13.4-1 13.4.1 Objectives .................................................................................................................. 13.4-1 13.4.2 Preoperational Test Schedule Considerations ........................................................... 13.4-2 13.4.3 Construction Test ....................................................................................................... 13.4-3 13.4.4 Summary of Preoperational Test Content .................................................................. 13.4-5 13.5 Startup And Power Test Program .............................................................................................. 13.5-1 13.5.1 Program Description and Objectives .......................................................................... 13.5-1 13.5.2 Discussion of Startup Tests ....................................................................................... 13.5-5 13.5.3 Nuclear System Startup Test Restrictions ................................................................. 13.5-70 13.6 Normal Operations .................................................................................................................... 13.6-1 13.6.1 General ...................................................................................................................... 13.6-1 13.6.2 Normal Operating Instructions ................................................................................... 13.6-1 13.6.3 Emergency Operating Instructions ............................................................................. 13.6-2 13.6.4 Maintenance Instructions ........................................................................................... 13.6-3 13.6.5 Radiological Emergency Plan .................................................................................... 13.6-3 13.6.6 Radiation Control Instructions .................................................................................... 13.6-4 13.6.7 Surveillance Instructions ............................................................................................ 13.6-4 13.6.8 Technical Instructions ................................................................................................ 13.6-4 13.0-i

BFN-26 CONDUCT OF OPERATIONS TABLE OF CONTENTS (Cont'd) 13.6.9 Security Plan Procedures........................................................................................... 13.6-5 13.6.10 Special Test Instructions ............................................................................................ 13.6-5 13.6.11 Modification/Addition Instructions ............................................................................... 13.6-5 13.7 Records (Deleted) ..................................................................................................................... 13.7-1 13.8 Operational Review And Audits ................................................................................................. 13.8-1 13.8.1 General ...................................................................................................................... 13.8-1 13.8.2 Onsite Reviews .......................................................................................................... 13.8-1 13.8.3 Independent Reviews ................................................................................................ 13.8-1 13.8.4 Audit Program ............................................................................................................ 13.8-2 13.8.5 References................................................................................................................. 13.8-2 13.9 Refueling Operations ................................................................................................................. 13.9-1 13.10 Refueling Test Program............................................................................................................. 13.10-1 13.10.1 Program Description and Objectives .......................................................................... 13.10-1 13.10.2 Test Purpose, Description, and Acceptance Criteria .................................................. 13.10-1 13.0-ii

BFN-26 CONDUCT OF OPERATIONS LIST OF TABLES Table Title 13.5-1 Control Rod Drive System Tests (Unit 1) 13.5-2 Major Plant Transients 13.5-3 Stability Tests 13.5-4 Startup Test Program (Unit 1) 13.5-5 Startup Test Program (Unit 2) 13.5-6 Startup Test Program (Unit 3) CONDUCT OF OPERATIONS LIST OF ILLUSTRATIONS Figure Title 13.2-1 (Deleted) 13.2-2 (Deleted) 13.2-3 (Deleted) 13.2-4 (Deleted) 13.2-5 (Deleted) 13.2-6 (Deleted) 13.2-7 (Deleted) 13.2-8 (Deleted) 13.4-1 Typical Preoperational Test Sequence 13.5-1 sht 1 Probable Startup Test Sequence - Unit 1 12.5-1 sht 2 Probable Startup Test Sequence - Unit 1 13.5-2 sht 1 Probable Startup Test Sequence - Units 2 and 3 13.5-2 sht 2 Probable Startup Test Sequence - Units 2 and 3 13.5-2 sht 3 Probable Startup Test Sequence - Units 2 and 3 13.6-1 sht 1 (Deleted) 13.6-1 sht 2 (Deleted) 13.0-iii

BFN-16 13.0 CONDUCT OF OPERATIONS 13.1 ORGANIZATIONAL STRUCTURE FOR THE CONDUCT OF OPERATIONS 13.1.1 Functions, Responsibilities, and Authorities The Browns Ferry Nuclear Plant was designed and constructed and is operated to produce electric power reliably and economically, and with safety to the public and plant personnel. Its three nuclear steam supply systems were supplied by the General Electric Corporation (GE). The TVA Nuclear Engineering Organization served as the plant architect, engineer, and principal contractor for the balance of plant equipment and was responsible for ensuring that the technical requirements of the nuclear steam supply system contracts were met. The TVA Nuclear Construction organization was responsible for constructing the plant in accordance with design specifications supplied by the Nuclear Engineering Organization. TVA Nuclear Power is responsible for the safe operation and maintenance of the plant in compliance with the operating licenses, technical specifications, and other applicable requirements. Nuclear Power is also responsible for preoperational and startup testing programs as discussed in Chapter 13.4 and 13.5. The original design and construction criteria are provided in various appendixes to this FSAR. Appendix D and the Nuclear Quality Assurance Plan, TVA-NQA-PLN89-A, describe the quality assurance plan developed for the design, construction, and operation of Browns Ferry for both TVA and General Electric. Quality assurance of components that were built and supplied by General Electric was the responsibility of General Electric. The responsibility for quality assurance for supplied components and designs is described in Appendix D. Quality assurance programs for plant operations, maintenance, and upgrading are now the responsibility of and are audited by the Nuclear Quality Assurance Organization. 13.1.2 Interrelationships With Contractors and Suppliers General Electric supplied TVA with preoperational and startup test procedures for the nuclear steam supply system. General Electric provided technical direction and guidance during preoperational and startup testing and during initial operation until all plant equipment was fully accepted by TVA. General Electric continues to supply technical direction and guidance as necessary to support continued operation of Browns Ferry. Technical direction and guidance are defined in paragraph 13.5. 13.1-1

BFN-16 General Electric supplied the initial core loads of fully licensable and operable nuclear fuel and committed to supply several reload batches. Nuclear Engineering provided preoperational and startup test-scoping documents and acceptance criteria for the balance of plant systems. Nuclear Construction was responsible for performing the preoperational testing including preparation of a detailed set of test procedures. Nuclear Power assumed responsibility for plant startup testing and operation commencing with fuel loading and for the preparation of detailed startup test procedure operation manuals. During major evolutions such as loading or unloading fuel in the reactor vessel, TVA may rely on General Electric as a source of technical expertise, advise, and/or functional assistance. 13.1-2

BFN-17 13.2 ORGANIZATION AND RESPONSIBILITY 13.2.1 CORPORATE ORGANIZATION The organization of the Tennessee Valley Authority's Nuclear Power Organization is presented in Tennessee Valley Authority Topical Report TVA-NPOD89-A, Nuclear Power Organization Description. 13.2.2 NUCLEAR POWER Nuclear Power (NP) is responsible for the safe design, construction, operation, and modification of TVA nuclear plants; for compliance with TVA policy on safety and quality; and for compliance with regulatory requirements as applicable to all activities. Nuclear Power plans and manages the nuclear energy supply programs to meet the requirements of the TVA power program consistent with safety, environmental, quality and economic objectives. It develops and implements policies, programs and plans for the nuclear power program. 13.2.3 Offsite Organizations The Nuclear Power Organization is presented in Tennessee Valley Authority Topical Report, TVA-NP0D89-A, Nuclear Power Organization Description. Qualification requirements for positions providing corporate technical support, specifying required education and experience are maintained in approved position descriptions on file at the site and central office by the Nuclear Human Resources Organization. Numbers of positions are contained in approved staffing plans also maintained by the Nuclear Human Resources Organization. 13.2.4 Onsite Organization The Browns Ferry Nuclear Plant Organization is presented in Tennessee Valley Authority Topical Report, TVA-NPOD89-A Nuclear Power Organization Description. 13.2.5 Personnel Functions, Responsibilities, and Authorities During normal plant operations, the plant manager is responsible for all plant activities. In the event of absence, incapacitation of personnel, or other emergencies, the plant manager shall delegate in writing the succession to this responsibility in accordance with Technical Specification 5.1.1. 13.2-1

BFN-17 13.2.6 Qualification Requirements for Nuclear Facility Personnel Nuclear Power (NP) personnel at the Browns Ferry Nuclear Plant are required to meet the qualification standards specified by NRC Regulatory Guide 1.8, Revision 2 (which endorses ANSI N18.1-1971 and ANSI/ANS 3.1-1981) with the alternatives outlined in the Nuclear Quality Assurance Plan, TVA-NQA-PLN89-A. Shift Technical Advisors are required to meet the qualifications specified in Technical Specification 5.2.2. The Radiological Control Superintendent shall have five years of experience in applied radiation protection. Minimum qualification requirements are detailed in SPP-1.1. Below are various onsite and offsite positions correlated to ANSI N18.1-1971 and ANSI/ANS 3.1-1981 positions as appropriate. Site positions will meet these requirements at a minimum. Additional qualifications are detailed in SPP-1.1. POSITIONS COVERED BY LICENSING COMMITMENTS NOTE TVA will meet the requirements of Regulatory Guide 1.8, Revision 2 (4/87) for all new personnel qualifying on positions identified in regulatory position C.1 after January 1, 1990. Personnel qualified on these positions prior to this date will still meet the requirements of Regulatory Guide 1.8, Revision 1-R (5/77). As specified in regulatory position C.2, all other positions will meet the requirements of ANSI/ANS N18.1-1971. Source Title from Source TVA Title ANSI N18.1-1971 Plant Manager (or Plant Manager FSAR 13 Assistants) ANSI N18.1-1971 Operations Manager Operations FSAR 13 Superintendent ANSI N18.1-1971 Maintenance Manager Maintenance FSAR 13 Superintendent ANSI N18.1-1971 Technical Manager Systems FSAR 13 Engineering Manager 13.2-2

BFN-17 ANSI N18.1-1971 Supervisors (See Below) FSAR 13 Requiring NRC Licenses ANSI/ANS 3.1-1981 Shift Supervisor Shift Manager USNRC Reg Guide 1.8 (R2) NPP IV FSAR 13/Tech Spec 5.0 ANSI/ANS 3.1-1981 Senior Operator Shift Manager USNRC Reg Guide 1.8 Unit Supervisor (R2) NPP IV FSAR 13/Tech Spec 5.0 ANSI/ANS 3.1-1981 Licensed Operator Nuclear USNRC Reg Guide 1.8 Unit Operator (R2) FSAR 13/Tech Spec 5.0 USNRC Reg Guide 1.8 Shift Technical Unit Supervisor (R2) Advisor Shift Technical Advisor 13.2-3

BFN-17 ANSI N18.1-1971 Supervisors Not Maintenance FSAR 13 Requiring NRC Supervisors License (Mechanical, Electrical, 

               *Note: Position     *Instrumentation);

covered by Site Quality paragraphs 4.3.2 Manager; Site Licensing and 4.4 of ANSI and Industry Affairs N18.1, 1971, do not Manager, Radiological require NRC Control License. *Supervisors; 

                                   *Industrial Safety Manager; General Craft Supervisors over instrument mechanics, machinists, electricians, steamfitters, and boilermakers; Chemistry Manager; Reactor Engineering Supervisor; BOP Systems Manager; NSSS Systems Manager; and Electrical/I&C Systems Manager ANSI N18.1-1971 Reactor Engineering Reactor FSAR 13         and Physics         Engineering Supervisor ANSI N18.1-1971 Instrumentation     Instrument and FSAR 13         and Control         Control/

Electrical Supervisor ANSI N18.1-1971 Radiochemistry Chemistry FSAR 13 Superintendent 13.2-4

BFN-17 ANSI/ANS 3.1-1981 Radiation Protection Radiological USNRC Reg Guide 1.8 Control Superintendent (R2) ANSI N18.1-1971 Technicians Rad Chem Lab FSAR 13 Analysts, Instrument Mechanics, Health Physics Technicians ANSI N18.1-1971 Repairmen Crafts Personnel FSAR 13 (machinists, electricians, steamfitters, boilermakers) ANSI N18.1-1971 Engineer in Charge Site Engineering FSAR 13 Manager ANSI N18.1-1971 Staff Specialists Offsite FSAR 13 Supervisory Personnel 13.2-5

BFN-16 Figure 13.2-1 Deleted by Amendment 11.

BFN-16 Figure 13.2-2 Deleted by Amendment 11.

BFN-16 Figure 13.2-3 Deleted by Amendment 11.

BFN-16 Figure 13.2-4 Deleted by Amendment 11.

BFN-16 Figure 13.2-5 Deleted by Amendment 11.

BFN-16 Figure 13.2-6 Deleted by Amendment 11.

BFN-16 Figure 13.2-7 Deleted by Amendment 11.

BFN-16 Figure 13.2-8 Deleted by Amendment 11.

BFN-24 13.3 TRAINING PROGRAMS 13.3.1 Accredited Training Programs The BFN training programs have been developed in accordance with the Systems Approach to Training as prescribed by the Institute of Nuclear Power Operations (INPO). The National Academy for Nuclear Training, through a formal accreditation process, verifies that BFN training programs meet the established criteria. BFN is a branch of the National Academy and has achieved accreditation of the following programs: nonlicensed operator reactor operator senior reactor operator continuing training for licensed personnel shift manager shift technical advisor instrument and control technician electrical maintenance personnel mechanical maintenance personnel and supervisor health physics technician chemistry technician engineering support personnel The training programs are periodically evaluated and reviewed by management for effectiveness. Revisions are made as appropriate. Records are retained as necessary to support management needs and to provide historical data. 13.3.2 General Employee and Fitness for Duty Training Programs All persons regularly employees at BFN are trained in the following areas commensurate with their job duties: fitness for duty general plant description job related procedures and instructions radiological protection emergency preparedness industrial safety fire protection security quality assurance 13.3-1

BFN-24 13.3.3 Other Training Programs Responsible managers ensure that personnel performing quality-related activities receive indoctrination and training to ensure that adequate proficiency is achieved and maintained. 13.3.4 References

1. 10 CFR Part 55, Operators Licenses
2. 10 CFR Part 50.120, Training and Qualification of Nuclear Power Plant Personnel
3. Deleted 13.3-2

BFN-18 13.4 PREOPERATIONAL TEST PROGRAM This section provides a description of the preoperational test program conducted on Browns Ferry prior to initial startup. It has been retained in the FSAR as an historical record of such testing. 13.4.1 Objectives The purpose of the preoperational test program was four-fold: (1) confirm that construction is complete to the extent that the equipment and systems that have been installed can be put into use during completion of construction on other systems or equipment, (2) adjust and calibrate the equipment to the extent possible in the "cold" plant, (3) assure that all process and safety equipment is operational, and in compliance with license requirements, to the extent possible and necessary to proceed into initial fuel loading and the startup program, and (4) provide baseline data to assist in the evaluation of subsequent periodic tests. An extensive preoperational test program was planned for the reactor unit being started. This program started approximately 11 months before initial fuel loading. The actual duration of each test was short, relative to the entire 11-month program; the longest test, the Reactor Protection System checkout, took approximately 9 weeks. Key systems were sequenced for completion and testing early enough to provide auxiliary services for testing and operation of other systems or for construction activities (e.g., the use of the makeup system for chemical cleaning). This resulted in an early requirement for electrical systems, demineralized water makeup, and cooling water systems. After nuclear fuel is loaded in the reactor, all interconnected auxiliary systems are treated as potentially radioactive. In time, many of these systems become sufficiently radioactive to impose restrictions and time limitations on maintenance work. During normal plant operations, some components and systems cannot be observed for proper performance. To avoid these limitations during the preoperational tests, all of the nuclear steam supply system and its auxiliary systems are normally tested before fuel loading. The preoperational test period is an important phase in the training of the nuclear system operators. Experience and understanding of plant systems and components is gained with a minimum of risk to the equipment or personnel. Minimal restrictions are imposed on either the operators or the testing, except those restrictions required by work in progress on other units. This gives maximum opportunity to evaluate and train individual operators and to troubleshoot plan systems. In 13.4-1

BFN-18 addition, plant equipment and systems are operated for a sufficient period of time to discover and correct any design, manufacturing, or installation errors, and to adjust and calibrate the equipment. Acceptance criteria for each preoperational test will be established by the designer and included in the preoperational test specifications and instructions. These documents will be prepared prior to the start of the preoperational test program. Preoperational test results were reviewed and evaluated to assure compliance with all acceptance criteria. The approval or modification of any test in which the results were affected were reviewed by TVA with technical assistance from General Electric site representatives. 13.4.2 Preoperational Test Schedule Considerations The following key points were considered in developing the sequence and schedule of preoperational tests:

a. Systems are sequenced for early testing and placed in routine operation to provide necessary auxiliary services for other systems. Examples are plant electrical systems, control air and makeup water supply systems.
b. Preoperational testing is coordinated with construction to permit fuel loading as early as possible, without compromising nuclear safety or impeding construction work. As a result, fuel loading occurs while construction work is still in progress on unrelated systems and areas.
c. Stricter controls of plant operations and maintenance work are required following fuel loading. Preoperational testing is performed before fuel loading on all systems that could consequently be exposed to radioactive contamination, to minimize possible contamination problems.
d. Preoperational tests provide an important phase of the plant operators' training program and are scheduled on key systems to permit maximum participation by all operators prior to licensing examinations.
e. Temporary construction power is sometimes required for initial tests at the beginning of the preoperational test program. However, unnecessary use of temporary power and improvised setups is to be avoided because of the possibility of costly errors and inconsistency with the ultimate objective of providing the final installation.

13.4-2

BFN-18

f. Electrical jumpers are used to facilitate preoperational testing in some instances, but their use is minimized and controlled by proper identification of such jumpers, by tags on the equipment jumpered, and by log book records.

All jumpers are removed before fuel loading, unless required by special procedures.

g. When a unit is ready for fuel loading, construction workers are excluded from that portion of the reactor building and the drywell, and strict control is enforced over access to the control room, electrical equipment rooms, and the radioactive waste treatment building. Construction work will proceed on later units under the access limitations of the operating unit (see Appendix F).
h. Tennessee Valley Authority operations personnel operate the plant and equipment (operation of the nuclear system will be under the technical direction of General Electric) during preoperational testing. However, some testing requirements actually precede the preoperational test program. These are categorized as Construction Tests and are performed by TVA construction forces or by subcontractors, with technical direction available from General Electric (see paragraph 13.4.3).
i. Detailed test procedures are prepared by General Electric or Tennessee Valley Authority depending upon system design responsibility. These procedures are specific regarding intent, method, and operating requirements for completing the test and will include detailed blank data sheets to be completed during the test.
j. In general, tests are performed using permanently installed instrumentation for the required data. Where it is not possible to run pumps or similar equipment for a period of time prior to the system preoperational tests, it is necessary to install test thermometers, vibration meters, stroboscopes, or other test instrumentation to ensure safe operation of the equipment. Special instrumentation is specified in the preoperational test procedure.
k. Where the unit being tested shares components or systems with a unit either still under construction or in operation, detailed supplemental preoperational test procedures will define the interactions and control procedures necessary to maintain operating continuity, system integrity, and plant safety without compromising test efficiency.
l. Typical test sequencing for the preoperational test phase is presented in Figure 13.4-1.

13.4.3 Construction Test

a. Initial containment leak rate testing.

13.4-3

BFN-18

b. System hydrostatic test.
c. Chemical cleaning and flushing.
d. Wiring continuity checks.
e. Megger and high potential test.
f. Electrical system tests to and including energizing (e.g., checking grounding, relay checks, checking circuit-breaker operation and controls, phasing check, and energizing of buses).
g. Initial adjustment and "bumping" of motors.
h. Check control and interlock functions of instruments, relays and control devices, except as described in i. below.
i. Calibrate instruments and recheck or set initial trip set points, except as described in paragraph 13.4.2.
j. Clean the control and service air system lines and pneumatically test the systems.
k. Adjustments such as alignment, greasing, and bolt tightening.
l. Surveillance of proper equipment operation during preoperational tests, as required. The primary intent of this item is to cover those instances where measurements such as the above are required to ensure proper operation, but are not obtainable until the entire system is operated during preoperational tests (e.g., measuring motor current and voltage; bearing, lubricating oil, cooling water and seal temperatures; vibration; torque; rpm). These measurements are primarily of importance for protection of equipment, troubleshooting, or supplementing installed instrumentation.
m. Observe the readiness of relief and safety valves.
n. Complete tests of motor-operated valves, including adjusting limit torque switches and limit switches, checking all interlocks and controls, measuring 13.4-4

BFN-18 motor current and operating speed, and checking leaktightness of stem packing and valve seat where required during hydrotests.

o. Complete tests of air-operated valves, including checking all interlocks and controls, adjusting limit switches, measuring operating speed, checking leak tightness of stem packing and valve seat where required during hydrotest, checking leak tightness of pneumatic operators, and checking for proper operation of controllers, pilot solenoids, etc.
p. Reactor vessel and reactor coolant system hydrotest.

13.4.4 Summary of Preoperational Test Content The prerequisites listed below for each preoperational test are typical. Certain portions of each test may be conducted without having completed all prerequisites. 13.4.4.1 Condenser Circulating Water System Interunit Crossties (TVA-1)* 13.4.4.1.1 Prerequisites

a. The Unit 1 condensers, intake and discharge tunnels filled with water and the discharge structure stoplog removed.
b. Limit switches on the condenser outlet valves set to stop the valves in the partially closed position for throttling the pump in order to obtain sufficient head for supplying water to Units 2 and 3 were the condensers for these units in service.
c. Electric power available from the emergency generator system for operation of pumps, valves, control system, and annunciation.
d. Preoperational test for diesel generators (see paragraph 13.4.3.46) should be coordinated with this test.
e. Construction checks on the vacuum priming and cooling water systems completed. These systems shall be operable.
f. The traveling screens should not be operated during this test.

13.4.4.1.2 Test Summary The feature demonstrated by this test is not required until Unit 2 is in operation; therefore, demonstration of this capability is not required for licensing of Unit 1. 13.4-5

BFN-18 However, this test may be performed during preoperational testing of Unit 1 to avoid interference with subsequent operation of Unit 1. A Unit 1 pump represents the worst condition; therefore, repetition of this test for a Unit 2 or 3 pump is not required. If either the Unit 2 of Unit 3 condensers (or both) are unavailable for use during this test, their presence may be simulated.

a. Verify that capability of any one Unit 1 pump and the crosstie piping and valving to supply the minimum flow required by all three units during a system blackout.
b. Throttle the discharge from the condenser until all of the tubes in the tube bank are filled. This operation will require the use of the mechanical vacuum pumps to keep the condenser tubes filled with water.

13.4.4.2 Condensate Storage and Transfer to the NSSS (TVA-2) 13.4.4.2.1 Prerequisites

a. Condensate storage tanks, 24-inch and 20-inch supply headers cleaned and ready for operation.
b. Chemical cleaning and flushing completed on all systems related to this test.
c. Main condensate pumps tested and checklists completed.
d. Condensate booster pumps tested and checklists completed.
e. All valves cross connecting the aluminum and steel headers locked closed.
f. Standpipes installed in condensate storage tanks.
g. Auxiliary boilers cleaned, tested, and ready for operation.
h. Condensate transfer pumps test and checklists completed.
i. HPCI main pump tested and checklist completed.
j. HPCI booster pump tested and checklist completed.
k. At least one condensate filter demineralizer tested and ready for operation.
l. Reactor vessel head removed, dryer and separator removed, and vessel filled with condensate.
m. Permanently installed instrumentation calibrated and checklist completed.

13.4-6

BFN-18

n. All motor-operated valves tested and checklist completed.
o. Strainers installed in pump suction lines as required.

13.4.4.2.2 Test Summary

a. Verify capacity of the system to supply required volume of condensate to fill the balance of reactor vessel, the dryer separator pit, and the reactor well (refueling pool) within about 3 hours.
b. Verify capacity of the system to supply 5000 gpm from the 24-inch header and to return this flow rate to the 20-inch header during the HPCI tests.

(For other testing of the condensate and feedwater system see paragraph 13.4.4.20 and 13.4.4.48.) 13.4.4.3 RHR Service Water System 1 (TVA-3) 13.4.4.3.1 Prerequisites

a. Construction testing and instrumentation calibration completed.
b. Control air supply available.
c. Electrical power available from emergency diesel bus.
d. Electrical AC and DC power available for system control.
e. Electrical AC power available for pump and valve action.
f. System hydrostatic test performed and system flush completed.
g. Raw Cooling Water System operable for unit under test.
h. EECW piping complete.

13.4-7

BFN-18 13.4.4.3.2 Test Summary

a. Verify capability of the RHRSW system to supply an adequate quantity of raw river water to the secondary side of the RHR heat exchangers under normal conditions.
b. Verify that the associated instrumentation and control systems are functioning correctly.
c. For Units 1 and 2 only, verify the capability of the RHRSW system to supply an adequate quantity of raw river water to the secondary side of the RHR heat exchangers under conditions simulating a breach of Wheeler Dam.

13.4.4.4 Emergency Equipment Cooling Water System (TVA-4) 13.4.4.4.1 Prerequisites

a. Construction testing and instrumentation calibration completed.
b. Electrical power available from the emergency diesel bus.
c. Electrical AC and DC power available for the control system.
d. Electrical AC power available for pump and valve operation.
e. Hydrostatic test performed and system flush completed.
f. RCW system operable for unit under test.
g. RHRSW system operable for unit under test.

13.4.4.4.2 Test Summary

a. Verify capability of the system to automatically supply raw river water to the assigned receivers upon initiation by an accident signal or by signals from the assigned receivers.

Demonstrate capability both with each pump individually assigned to EECW and with one pump fully assigned to EECW and one pump on floating backup.

b. Demonstrate operation of the sectionalizing valves.

13.4-8

BFN-18

c. Verify capability of the system to automatically supply the priority receivers of the RCW system upon initiation by a RCW low-pressure signal.
d. Verify ability of the EECW system to block service to certain of the RCW priority receivers in the event of low pressure in the section of header to which the line leading to those receivers is attached.
e. Verify proper operation of the automatic strainers.

13.4.4.5 Heating and Ventilation (H&V): Reactor Building and Control Bay (TVA-5) 13.4.4.5.1 Prerequisites

a. Electrical power available to operate equipment and control systems.
b. Electrical power available from the emergency diesel bus for designated equipment located in the control bay and shutdown board rooms.
c. Correct control air pressure available to operate pneumatically operated damper motors.
d. EECW system or RCW system cooling water available for operating control bay and shutdown room air-conditioning equipment.
e. Mechanical and electrical equipment checked and balancing of air flow quantities for each air supply, return, and exhaust system completed.
f. Standby Gas Treatment System available for service.

13.4.4.5.2 Test Summary for Reactor Building

a. With the building under normal H&V operating conditions demonstrate proper functioning of the control interlocks on the Air-Supply Systems and the Air-Exhaust Systems.
b. Demonstrate automatic actuation, where provided, of standby redundant components (fans and dampers) or trains on simulated failure of the initially operating component or train.
c. With the building under normal H&V operating conditions, demonstrate that the TIP room, the steam and feedwater valve room and rooms containing reactor water cleanup system components are maintained at a negative pressure with respect to the surrounding portions of the reactor zone. (Maintenance of this condition is an operational consideration and serves no essential safety function.)

13.4-9

BFN-18 13.4.4.5.3 Test Summary for Control Bay

a. Demonstrate capability of control room emergency pressuring unit to pressurize the floor at El. 617.0 feet.
b. Demonstrate automatic actuation, where provided, of standby redundant components (fans and dampers) or trains on simulated failure of the initially operating component or train.
c. Demonstrate operation of the emergency cooling units for the battery and board rooms during simulated loss of normal supply or exhaust as well as during conditions simulating those which would result from a fire in the main control room.
d. Demonstrate operation of the emergency battery-and-board-room exhaust fan.
e. Demonstrate operation of the emergency relay-room-cooling crosstie.
f. Demonstrate operation of the main-control-room coolers during simulated loss of normal supply and exhaust.
g. On those air-handling units which are so equipped, demonstrate operation of the direct-expansion cooling coils and associated equipment during simulated failure of the normal cooling coils and associated equipment.
h. Demonstrate capability of operating with the control bay under positive pressure using (1) the normal supply fans, and (2) the standby supply fans which are equipped with HEPA filter assemblies.
i. Demonstrate control bay Pressure-Relief Vent System.
j. Demonstrate ability of the smoke dampers, which are mounted in shutdown board room ventilation supply and exhaust ducts routed between rooms and through control bay wall, to automatically close upon indication of smoke.
k. Verify operation of all annunciators provided to signal the malfunction of Heating, Ventilating, and Air-Conditioning System equipment, or abnormal air temperature conditions.

13.4-10

BFN-18 13.4.4.6 Heating and Ventilation (H&V): Reactor, Offgas Treatment, and Radwaste Evaporator Buildings and Recombiner Room (TVA-19, including Supplements 1through 3) 13.4.4.6.1 Prerequisites

a. Electrical power available to operate equipment and controls.
b. Cooling water available from raw water system.
c. Cooling coils and piping system hydrostatic test and system flush completed.
d. Control air pressure available to operate pneumatic damper motors.
e. Air filters inspected for cleanliness, ducts cleaned and new filters installed where needed.
f. Balancing of air and water flow quantities for each system completed according to TVA DED guide, "Testing and Balancing of Heating, Ventilating, and Air-Conditioning Systems."
g. All instrumentation installed, tested, calibrated, and operational.

13.4.4.6.2 Test Summary

a. Verify the acceptable operation of the ventilation, cooling, and air-conditioning systems, where applicable, for turbine, radwaste, service, offgas treatment, and radwaste evaporator buildings, and offgas recombiner room and dehumidification system. In the service building only the exhaust fan systems equipped with HEPA filters will be tested.

13.4.4.7 Primary Containment Atmospheric Control (TVA-6) 13.4.4.7.1 Prerequisites for Cooling and Ventilation System

a. Construction testing of cooler fan operation, supply and return duct air distribution system, and cooling coil water flow completed.
b. Electric power available from the emergency diesel bus.
c. Control air available to operate pneumatic motor-operated dampers.

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d. Cooling water available to cooling coils from the Reactor Building Closed Cooling Water System.
e. Cooling coils and piping system hydrostatic test and system flush completed.
f. Reactor Building Closed Cooling Water (RBCCW) system operational and its preoperational testing completed (see paragraph 13.4.4.25).

13.4.4.7.2 Test Summary for Cooling and Ventilation System

a. Demonstrate proper operation of the system control logic and components (dampers and fans).
b. Confirm capability of using the system control logic and components to provide additional cooling capacity during a scram.
c. Perform preliminary air flow rate measurements.

13.4.4.7.3 Prerequisites for Inerting System

a. Containment Inerting System including Oxygen Analyzer System.
b. Auxiliary Boiler System.
c. Primary Containment System (to the extent necessary to contain the inerting nitrogen gas).
d. Service and Control Air System.
e. Drywell Control Air System.
f. Drywell Cooling System.

13.4.4.7.4 Test Summary for Inerting System

a. Demonstrate capability of the system to supply nitrogen gas at a sufficient flow rate so that, if continued, the design objectives would be met.
b. Demonstrate automatic control features of the Inerting Control System.

13.4.4.8 Containment Air Dilution (CAD) System (TVA-27) 13.4.4.8.1 Prerequisites

a. Nitrogen supply tank installed and tested.

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BFN-18

b. Piping and valves from nitrogen supply tanks to pressure suppression chamber and drywell of unit under test installed and operational.
c. Instrumentation tested and calibrated as necessary.
d. Adequate quantity of liquid nitrogen in each supply tank.

13.4.4.8.2 Test Summary

a. Demonstrate the ability to continuously supply gaseous nitrogen at 50°F or above and at 6000 scfh over a two-hour period to the drywell or pressure suppression chamber of the unit being tested.
b. Calibrate and verify the performance of the monitoring systems for hydrogen and oxygen.
c. Adjust HCV-84-37 and HCV-84-38 to deliver 6000 scfh of gaseous nitrogen.
d. Verify that valves FCV-64-31 and FCV-64-34 can be opened with Hs-84-35, and that valves FCV-64-29 and FCV-64-32 can be opened with HS-84-36.
e. Verify that the drywell and pressure suppression chamber can be vented properly.

13.4.4.9 CS and RHR Pump Compartment Coolers (TVA-7) 13.4.4.9.1 Prerequisites

a. Construction testing of cooler fan operation, air supply duct distribution system, and cooling coil water flow completed.
b. Electrical power available from the emergency diesel bus.
c. Correct electrical power for operating fan motor and control circuits available.
d. Cooling water available from emergency equipment cooling water systems and/or raw water system.
e. Cooling coil and piping system hydrostatic test and system flush completed.
f. Fan motors lubricated and fan belts checked.
g. RHR and core spray pumps operational.

13.4-13

BFN-18 13.4.4.9.2 Test Summary

a. Demonstrate automatic start of each RHR pump-compartment cooler when its associated RHR pump starts.
b. Demonstrate automatic start of each CS pump-compartment cooler when either associated CS pump starts.
c. Demonstrate automatic start of each pump-compartment cooler when its associated temperature sensor is exposed to temperature greater than approximately 100°F.

13.4.4.10 161-kV Electrical System (TVA-8) 13.4.4.10.1 Prerequisites

a. All construction checks for this system completed and documented.
b. All relay setting sheets completed and signed.

13.4.4.10.2 Test Summary

a. Confirm independence of the 161-kV sources of power to the 4160-V auxiliary power system by the simulation of faults on the Trinity and Athens 161-kV line sections and also lines terminating at the Trinity and Athens substations.

(Where possible, standard transmission line protective relay tests will be used to demonstrate that one 161-kV line and one common transformer group are available for faults on either the Athens or Trinity 161-kV line sections.)

b. Simulate failure of PCB 924 to trip for a line fault to demonstrate action of the backup relay protection to isolate the fault.

13.4-14

BFN-18 13.4.4.11 4.16-kV Electrical System--Normal Auxiliary Power System (TVA-9) 13.4.4.11.1 Prerequisites

a. The 161-kV yard available to energize both common station service transformers and both cooling tower transformers.
b. Main generator links disconnected and unit station service transformer energized from 500-kV yard.
c. All system construction checks completed and documented as outlined in the applicable Browns Ferry Nuclear Plan construction procedures.

13.4.4.11.2 Test Summary The equipment involved in this test will be the common station service transformers, cooling tower transformers, circuit breakers 1412, 1414, 1516, and 1518 (except as they apply to tests performed in preoperational testing of the 161-kV electrical system), the 4-kV unit boards 1A and 1B (2A, 2B, 3A, and 3B will be tested before operation of their respective units), the Unit 1 station service transformer (unit station service transformers 2 and 3 will be tested before operation of their respective units, 4-kV cooling tower switchgear and 4-kV bus tie board. The limit of the test will be the load side of the unit board breakers 1126 and 1132 and the load side of breakers 1642, 1742, 1632, and 1732 to the shutdown bus. No other unit board load is involved in these tests.

a. Demonstrate search and transfer operations of the system in seeking a normal power supply before invoking diesel generator operation.
b. Confirm loading and voltage regulation design objectives under design criteria loading (i.e., under the most degraded conditions).
c. Verify short-circuit and inverse-time protection of all 4-kV circuit breakers supplying power to Class 1 boards.

13.4.4.12 480-V Electrical System--Normal (TVA-10) 13.4.4.12.1 Prerequisites

a. Insulation tests of control and power cables completed for equipment to be tested.
b. Power and control cable checklists completed.

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c. Insulation tests for switchgear and panelboard equipment completed and satisfactory.
d. High potential tests on 5-kV cables completed and satisfactory.
e. Cable installation and documentation completed and satisfactory.
f. Bus installation completed and satisfactory.
g. Inspection and testing of 480-V AC power circuit breakers completed and documented.
h. Switchboard instruments calibrated and protective relays set according to appropriate setting sheets.
i. Power transformer, current transformer, and potential transformers associated with equipment under tests completed and satisfactory.
j. Wiring and continuity checks completed on equipment under test.

13.4.4.12.2 Test Summary With two exceptions the components of this system serve no essential safety functions; therefore, the balance of this system will be tested and documented using accepted power plant practices. The two exceptions are 480-V control bay ventilation board A and 480-V control bay ventilation board B. These two boards are exceptions since they are the only components serving essential safety functions which have ties back to the standby auxiliary power system. The 480-V condensate demineralizer boards are also tied back; but these boards serve no essential safety functions.

a. Demonstrate that for a loss of power to either 4-kV common board A or B, all of 480-V common boards 1 and 3 will remain energized through automatic operation of bus feed and bus tie breakers.
b. Demonstrate that the 480-V control bay vent boards will preferentially seek their 480-V common board supplies in lieu of their 480-V common board supplies in lieu of their 480-V shutdown board supplies.
c. In addition to these automatic transfer tests, the trip setting for the 480-V common board breakers feeding the control bay vent boards will be confirmed 13.4-16

BFN-18 by tests and the test results included in the final documentation. (Acceptance should be based on meeting manufacturer's stated accuracy.)

d. Verify short-circuit and inverse-time protection of all branch circuits from any Class 1 boards.

13.4.4.13 AC Supply--Plant Preferred and I&C (TVA-11) 13.4.4.13.1 Prerequisites All system construction checks completed and documented as outlined in the applicable Browns Ferry construction procedures. 13.4.4.13.2 Test Summary No loads having essential safety functions are supplied from the Plant Preferred System. Therefore, this system will be tested and documented using accepted power plant practices. The Instrument and Control Power System (I&C) is a Class 1E system subject to interruptions for a duration of up to 10 sec during transfer of the 480-V shutdown boards from normal to diesel power. This system will be tested to prove three major design characteristics as described below.

a. Demonstrate independence of the redundant channels.
b. Demonstrate full-load capabilities of the system.
c. Demonstrate current coordination of the system.

This preoperational test will cover the 480-V shutdown board breakers supplying power to the I&C transformers, the Battery Board Distribution System, and the Panel 9-9 Distribution System. It will include the load breakers but will not include tests on any of the I&C loads. The loads having essential safety functions will be tested in specific system preoperational tests. Loads having no essential safety functions will be tested using accepted power plant practices. 13.4.4.14 Plant Communication System (TVA-12) 13.4.4.14.1 Prerequisites

a. Installation verified to agree with latest reference drawings.
b. All system construction tests completed.

13.4.4.14.2 Test Summary This test is limited to demonstrating capability of those portions of the system which are independent of a power source and which have signaling capability. Therefore, 13.4-17

BFN-18 the testing will be conducted only on the Backup Control Center System and on the Health Physics System. The balance of the Plant Communication System serves no essential safety function and will be tested using accepted power plant practices.

a. Verify ability to signal using magneto howlers and to acknowledge using sound-powered equipment.
b. Test backup control center system to determine maximum number of jacks into which handsets can be plugged without overloading system.

13.4.4.15 Evacuation Signal (TVA-16) 13.4.4.15.1 Prerequisites

a. Installation verified to agree with latest reference drawings.
b. All system construction tests completed.

13.4.4.15.2 Test Summary In general, this test is designed to verify that the system functions properly and provides adequate coverage through out the plant. Specific objectives of the test are described below.

a. Demonstrate that the system can be activated at the electrical control room console and the Unit 3 control room console under both normal and abnormal conditions.
b. Verify and document the proper performance of each siren in the system, both quantitatively and qualitatively.
c. Demonstrate that failures within the system, should they occur, are annunciated and are properly identified.
d. Determine that the power distribution network performs properly.

13.4.4.16 Fire Protection System--Water and CO2 (TVA-13) 13.4.4.16.1 Prerequisites

a. Construction tests and instrumentation calibration completed for high-pressure water fire-protection system.
b. Construction tests of CO2 fire-protection system completed for the storage unit and the piping and instrumentation serving the diesel generator building 13.4-18

BFN-18 hazard areas. Flooding and concentration tests for the diesel generator building hazard areas, which are done using accepted power plant practices, shall be recorded as part of these preoperational tests.

c. Electrical power available for controls, CO2 refrigeration unit and fire pump operation.
d. Electrical power available from emergency diesel generator bus to demonstrate that fire pumps may be operated manually in case of an accident signal and loss of normal auxiliary power.

13.4.4.16.2 Test Summary

a. Demonstrate that fire-pump loads are automatically rejected by emergency-equipment logic as long as an accident signal exists coincident with a signal that normal auxiliary power is not available.
b. Demonstrate that individual control switches in the main control room can be used to override the load rejection at any time.
c. Demonstrate automatic start of these pumps under nonaccident conditions.
d. Demonstrate that CO2 alarm horns for each hazard area in the diesel generator building sound when the timer is actuated, and that discharge into the hazard area does not occur until 20 sec after the horns have started sounding.
e. Demonstrate that CO2 system is capable of discharging the correct concentration into each hazard area in the diesel generator building. These hazard areas are the diesel generator rooms, the fuel oil transfer pump room, and the electrical board rooms. The balance of the CO2 Fire Protection System will be tested using accepted power plant practices.

13.4.4.17 Control Air System and Drywell Control Air System (TVA-14) Testing of the Control Air system and Drywell Control Air System is primarily limited to those portions which serve an essential safety function and for which accumulators or receivers have been provided. Each of these portions extends from the device actuated back to the check valve which is used to isolate that portion from the balance of system on loss of system pressure. An accumulator or receiver is located between the device and the check valve. The devices which are served in this manner by the Drywell Control Air System are six of the main steamline relief valves and four inboard main streamline isolation valves. The devices which are served in this manner by the Control Air System are four outboard main steam isolation valves and the pneumatic door seats for the large equipment lock. 13.4-19

BFN-18 13.4.4.17.1 Prerequisites

a. Installation of all air lines completed from the compressors through the devices to be tested.
b. All air lines cleaned and the pressure and leak tests completed on applicable lines.
c. Air compressors, auxiliaries and receivers, and instrumentation and controls checked out and ready for operation.
d. Electrical AC and DC power available as required.

13.4.4.17.2 Test Summary

a. Verify capability of the Control Air System to shut off supply to the Service Air System on low pressure.
b. Measure leakage rate from pneumatic seats, air receiver, and associated piping and check valve with the piping upstream of the check valve depressurized.
c. Demonstrate that the capacity of the air receiver and associated piping is sufficient for full inflation of seals the design number of cycles and that the equivalent of about one week of leakage could occur without makeup and without loss of sealing capability.

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d. Demonstrate that the capacity of each isolation-valve accumulator and associated piping is sufficient to allow the valves to be closed the design number of cycles with the piping upstream of the check valve depressurized.
e. Demonstrate that the capacity of each relief-valve accumulator and associated piping is sufficient to allow the valves to be cycled the design number of cycles with the piping upstream of the check valve depressurized.

13.4.4.18 Offgas System Including SJAE (TVA-15) 13.4.4.18.1 Prerequisites

a. Auxiliary boiler system, offgas system including installation of HEPA filters, and offgas stack dilution system completed, tested, and ready for operation.
b. Turbogenerators and condensers completed and condensers filled to normal operating level.
c. Condensate system and condensate demineralized system completed and ready for operation.

13.4.4.18.2 Test Summary

a. Using steam from the auxiliary boiler, this test will confirm ability of the system to: (1) automatically start the standby SJAE on low condenser vacuum, and (2) automatically isolate the system on high pressure, on high temperature, and on low-low condenser vacuum. It will also determine the steady-state leakage rate through the seals of the condenser. (Incidental to this test the functioning of the Turbine Seal System will be demonstrated.)
b. Demonstrate proper operation of the dilution fan and mechanical vacuum pump control logic.
c. Demonstrate satisfactory efficiency of HEPA filters.

13.4.4.19 Turbine Electrohydraulic Control (EHC) System Including IPR's and Remote Controls (TVA-17) 13.4.4.19.1 Prerequisites

a. Turbine lube oil system completed, tested, and ready for operation.
b. Turbine hydraulic fluids system completed, tested, and ready for operation.

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c. Normal turbine protective systems (hydraulic, mechanical, electric, and electronic) all installed, tested, and ready for zero speed operation. The conventional turbine trips should have been previously checked out, and should produce the required turbine trip through energizing the master trip relay. Since these trips serve no essential safety function, it is enough to demonstrate that a turbine trip from the master trip relay will trip close all turbine valves to initiate a reactor scram.
d. Turbine EHC system installed, properly aligned, and completely checked and tested and ready for zero speed operation.
e. Reactor protection system completed, tested, and determined to be ready for operation through the "Autoscram" channels.
f. Control room annunciation and sequential events systems installed, tested, and ready for operation.
g. Sufficient control room instrumentation in service so as to monitor performance of all necessary turbogenerator auxiliaries.
h. Appropriate electrical supply (power and control, AC and DC) available for items a through f, plus cooling water as required for turbogenerator auxiliaries, heat exchangers, etc.

13.4.4.19.2 Test Summary The conventional turbine tests performed on this system involve features which serve no essential safety functions; therefore, that portion of the testing will be conducted and documented using accepted power plant practices.

a. Verify under simulated conditions the initiation of reactor scram signals on turbine-control-valve closure, and low EHC oil pressure with the turbine first stage pressure greater than 30 percent of the rated value.
b. Verify under simulated conditions the performance of the Bypass-Valve System and its response to control valve closure.
c. Verify under simulated conditions the capability of various turbine trip initiating conditions to trip the turbine. Sequence-of-event printout capability should be available during this portion of the test.

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BFN-18

d. Demonstrate under simulated conditions bypass-valve closure and feedwater turbine trip on insufficient condenser vacuum in addition to main turbine trip.

13.4.4.20 Secondary Containment Leak Rate Test (Including Standby Gas Treatment System--SGTS, Vacuum Relief System, and Primary Containment Purge to the SGTS), (TVA-20) 13.4.4.20.1 Prerequisites

a. Electrical power available to operate equipment and control systems.
b. Electrical power available from emergency diesel boards A and B for designated SGTS equipment.
c. Control air available for pneumatically operated dampers.
d. Air supply, return, exhaust, dampers, and ductwork of the reactor building ventilating system available for service.
e. Doors and seats for personnel and equipment locks in place and operational.
f. Completion of all secondary containment penetrations to the zone or zones to be leak tested.
g. Architectural siding on refueling floor has passed quality leak rate tests.
h. Relief panels at main steam penetrations and between zones installed.
i. SGTS suction piping and ductwork cleaned to remove all foreign debris that might poison the charcoal.
j. Static pressure limiters are installed, tested, and operative.
k. Vacuum relief units installed, tested, calibrated, and operative.
l. SGTS instrumentation installed, tested, calibrated, and operational.
m. SGTS exhaust piping and stack exhaust vent and piping available for use.

13.4.4.20.2 Test Summary

a. Demonstrate that all three of standby redundant components (dampers) or trains start initially. NOTE: The remaining test requirements must be met with the refueling zone alone isolated and with both the refueling zone and the reactor zone isolated.

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b. Demonstrate that any two out of three SGTS blowers can produce the design flow rate.
c. Demonstrate operation and leak tightness of the isolation dampers. The modulating capability of the Vacuum Relief System will be checked. The containment leakage rate will be determined by comparison between the inflow through the SGTS.
d. The operating range of the vacuum relief dampers will be compared with the physical limits.
e. Demonstrate that test conditions do not result in requiring excessive force to open any doors. In addition, demonstrate that the door closers can maintain closure of other doors.
f. Determine the differential pressure between specified locations in any isolated zone and specified locations in adjacent zones or on the building exterior.

13.4.4.21 Leak Test of Drywell to Pressure Suppression Chamber Vacuum Breakers (TVA-24) 13.4.4.21.1 Prerequisites

a. Preoperational Test GE-14A, Integrated Primary Containment Leak Rate Test, completed and primary containment integrity established.

13.4.4.21.2 Test Summary

a. Establish the leakage rate between the drywell and the pressure suppression chamber with the vacuum breakers closed.
b. Detect flow path between the drywell and pressure suppression chamber whose total capacity is equal to or greater than the capacity of a 1-inch diameter orifice.
c. To verify, by superimposing a known flow, the accuracy of the leak rate test.

13.4-24

BFN-18 13.4.4.22 Reactor Building Crane (TVA-21) 13.4.4.22.1 Prerequisites

a. Crane construction testing completed and permanent power available.
b. The crane emergency stop stations along north wall checked for proper operation.

13.4.4.22.2 Test Summary

a. Verify that the 125-ton Reactor Building crane will operate in good mechanical and electrical condition and that it handles its rated capacity.
b. Perform inspection of mechanical and electrical equipment.

13.4.4.23 Miscellaneous Gaseous Radiation Monitoring Systems (TVA-22) This testing includes the following systems:

a. Plant Ventilation Exhaust Radiation Monitoring System.
b. Drywell Leak Detection Air Sampling System.
c. Area Radiation Monitoring System, Air Particulate Monitoring Subsystem.

13.4.4.23.1 Prerequisites

a. Systems listed in Sections 13.4.4.13.a, .b, and .c above completed and construction checks completed.
b. Instrumentation calibrated.
c. Applicable alarm set points calculated.

13.4.4.23.2 Test Summary The following types of testing are required prior to fuel loading:

a. Check continuity and resistance to ground of all signal and power cables.
b. Check response and calibration of all channels with simulated input signals.

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BFN-18

c. Check alarm and trip set point.
d. Check chamber response to bugging sources.
e. Functionally check sampling and recording equipment including failure alarms.

13.4.4.24 Environs Monitoring (TVA-23) 13.4.4.24 Prerequisites Monitors functioning and capable of telemetering data to the logarithmic recorder installed in the Unit 1 control room. 13.4.4.24.2 Test Summary

a. Verify and document capability of local and perimeter radiation monitors to perform their design function.
b. Verify calibration, signal transmission, main control room recording, and malfunction alarm performance.

13.4.4.25 Feedwater System (TVA-25) 13.4.4.25.1 Prerequisities

a. Condensate-feedwater systems completed, tested, and ready for operation.
b. Condensers completed and filled, along with the condensate-feedwater system, to normal operating level.

13.4.4.25.2 Test Summary

a. Demonstrate that condensate booster pumps operating in series with condensate (hotwell) pumps are capable of pressurizing the reactor to approximately 385 psig.
b. Verify that condensate (hotwell) pumps, when operating alone can supply at least 35 psig pressure to the reactor feed pump suction valve to permit operation at low flows for emergency service. For other testing of the condensate and feedwater system see paragraphs 13.4.4.2 and 13.4.4.28.

13.4.4.26.2 Test Summary Demonstrate capability of the system to take water from the pressure suppression pool and pump it to Auxiliary Boiler "A." Water from the upper portion of the upper 13.4-26

BFN-18 drum will be returned to the pressure suppression pool. Auxiliary Boiler "A" shall be isolated from the other boilers and from the heating system during this test. 13.4.4.27 Raw Cooling Water System (TVA-32) 13.4.4.27.1 Prerequisites

a. System construction testing and instrumentation calibration completed.
b. Electrical power available for system control and for system pump operation.
c. Preoperational testing for the EECW System completed.
d. Individual unit condenser-circulating-water pumping system operable.
e. Condenser circulating water diffusers capable of receiving RCW discharge as follows: Unit 1 RCW - Unit 2 CCW, Unit 2 RCW - Unit 3 CCW, Unit 3 RCW - Unit 1 CCW.
f. Vacuum priming system operable.
g. RHR pumps available for operation.

13.4.4.27.2 Test Summary

a. Demonstrate capability of supplying the RHR-pump seal heat exchangers and the RHR-compartment coolers.
b. Demonstrate that one RCW pump per unit can be supplied from the intake conduit with no condenser circulating water pumps operating for that unit.

(This capability is not essential to safe shutdown.) 13.4.4.28 Evaporator for Radioactive Waste System (TVA-34) 13.4.4.28.1 Prerequisites

a. All construction tests, preliminary equipment test, and instrument checks must be completed prior to this test.

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BFN-18

b. Entire evaporator system operable.
c. Auxiliary boiler steam supply and condensate return available.
d. Raw cooling water available.
e. Demineralized water or seal water available.
f. All tanks cleaned.
g. All sumps cleaned and limit switches set.

13.4.4.28.2 Test Summary

a. Verify the functions of all controls and interlocks.
b. Verify the operation of all air operated valves.
c. Verify the operation of all pumps by checking pump operation in recirculation mode and simulating all other operations.
d. Check tanks by draining or filling, recirculating, sampling, and processing to other tanks.
e. Check sumps by filling with water, operating sump pumps to verify functioning of level controls, and checking discharge to proper collection tank in radwaste for no back flow or leakage en route.
f. Check evaporator by demonstrating the ability of the system to remove water and increase the concentration of a sodium sulfate solution to 25 percent by weight.

13.4.4.29 Radioactive Waste Solidification System (TVA-35) 13.4.4.29.1 Prerequisites

a. All construction tests, preliminary equipment test, and instrument checks completed.
b. Entire solidification system operable.
c. Applicable portions of the system completely flushed with demineralized water.

13.4-28

BFN-18 13.4.4.29.2 Test Summary

a. Verify ability of the system to solidify mixtures of Tiger Lock (urea-formaldehyde) and waste solution in the ratio of 1 to 3, respectively.
b. Verify ability of system to solidify mixtures of water, sodium sulfate, and concentrated floor drainwastes of the lowest possible radioactivity level.
c. After solidification check container for free standing water and the homogeneity of the solidified mass.

13.4.4.30 Solid Radwaste System (TVA-18) 13.4.4.30.1 Prerequisites

a. Liquid radwaste preoperational testing complete (see paragraph 13.4.4.54).
b. One phase separator filled with a minimum of settled sludge from both powder demineralizer and solkafloc filters.
c. Spent resin tank contains one charge of depleted head resins.
d. Portable tank available to be used as a simulated waste package. The tank must have an open top to observe dewatering procedure.

13.4.4.30.2 Test Summary

a. Verify ability of the transfer and compacting equipment to take wet solids from phase separator tanks and to prepare these solids for shipment.
b. Verify ability of the bailing equipment to package dry solid wastes.

13.4.4.31 Condensate Demineralizer System Test (GE-17) 13.4.4.31.1 Prerequisites

a. System completed and cleaned.
b. Condenser hotwell and condensate pumps operational.
c. Condensate system cleaned and flushed.

13.4-29

BFN-18 13.4.4.31.2 Test Summary

a. Initiate precoat system and precoat demineralizer with resins.
b. Check instruments and controls.
c. Operate filter demineralizers. Obtain clean pressure drop data for each unit.
d. Check system instrumentation and controls for proper operation.
e. Check operation of filter aid body coat system.
f. Verify that effluent water meets proper water quality specifications.

13.4.4.32 Reactor Building Closed Cooling Water System Test 1GE-15) 13.4.4.32.1 Prerequisites

a. Supply of demineralized water available.
b. Air available for operation of air-operated valves.
c. Chemical injection system operational for injecting corrosion inhibitor.
d. AC and DC control power available.

13.4.4.32.2 Test Summary

a. Fill system with water and inhibitor--check surge tank level indication and alarms.
b. Check operation of reactor building cooling water pumps.
c. Check division of cooling water flow to heat exchangers.
d. Check all system instruments.
e. Check all automatic actions, such as pump start or partial system isolation if necessary.

13.4.4.33 Control Rod Drive Hydraulic System Test (GE-10) 13.4.4.33.1 Prerequisites

a. All CRD piping and wiring installed and connected.

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b. System flushed and cleaned.
c. System filled and demineralized water available in demineralized water reservoir.
d. CRD hydraulic supply pumps operations.
e. Control air available.
f. AC and DC power available.
g. Power available through reactor safety circuit to energize scram valves.

13.4.4.33.2 Test Summary

a. Calibrate instruments.
b. Check alarms, controls, and interlocks.
c. Obtain pump performance data (e.g., head, flow, suction pressure, bearing, and cooling water temperatures, motor current, and RPM). NOTE: The above portion of the Preoperational Test may be performed much earlier than the remainder of the test because the pumps are used for the flushing listed in item (b) of the prerequisites.
d. Adjust flow control valves.
e. Check operation of proper valves from appropriate selector switches, interlocks, or trip signals, including:
1. Scram valves and scram solenoid pilot valves.
2. Scram backup pilot valves.
3. Scram volume dump and vent valves.
4. Drive selection valves, withdraw and insert control.
f. After drives and rods are installed, adjust individual flow control valves for proper drive speeds.
g. Monitor and record total system performance data with all drives installed, including:

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1. Cooling water flow.
2. Total system flow.
3. Flow returned to reactor.
4. System pressures.
5. Transient response of system during insert and withdraw operations, or following scrams, including checking the scram dump volume level instrumentation and drain and vent valves.

The above control rod drive hydraulic system and control system tests are completed before beginning tests of individual control rod drive mechanisms. All internals are in reactor, including guide tubes and thermal sleeves. Install blades and dummy fuel assemblies. The following tests are required on individual drives:

a. Insertion--continuous and by notch.
b. Withdrawal--continuous and by notch.
c. Stroke timing.
d. Friction measurements.
e. Scram time measurements.
f. Check proper position indication and in/out lights.
g. Repeat those tests in the hydraulic system and manual control system which are required to verify total system performance.
h. Recheck rod control interlocks.
i. Test safety circuit in conjunction with control rod system to verify scram signals and rod withdrawal interlocks from all safety circuit sensors.

13.4.4.34 Primary Containment Leak Rate Measurement Test (GE-14) 13.4.4.34.1 Prerequisites

a. All piping and electrical penetrations of the primary containment in place.
b. Testing described in items 1 through 3, below, need not be performed in sequence; however, the remaining items are performed in sequence, i.e.:

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1. Individual penetration leak rate measurements.
2. Isolation valve operating test.
3. Valve seat leakage measurements.
4. Overpressure test.
5. Design pressure tests (may precede 2 and 3).
6. Combined leak rate measurement.
c. All isolation valves fully operable.
d. Containment spray and core spray systems completed and operable.
e. During the combined leak rate measurement at design pressure, no equipment shall be operating within the containment and no heat sources shall be energized or hot or cold fluids circulating.
f. A complete survey made to locate and remove any instrumentation, light bulbs, etc., which could be damaged by external pressure.

13.4.4.34.2 Test Summary

a. Check individually testable penetrations by applying air pressure and checking with soap bubbles.
b. Stroke all containment isolation valves and leave in closed position.
c. Pressurize to 5 psig and check with soap bubbles all penetration welds made subsequent to design pressure test.

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d. Pressurize to 49 psig and conduct leak rate measurement using absolute method.
e. Repeat the test at 25 psig in order to establish an appropriate relationship between leakage rate and containment pressure.

13.4.4.35 Fuel Pool Cooling and Cleanup System Test (GE-18) 13.4.4.35.1 Prerequisites

a. All piping installed and thoroughly cleaned.
b. Electrical supply complete to this system.
c. Instrumentation installed and calibrated.
d. Fuel pool thoroughly cleaned and filled with demineralized water.
e. Proper air supply available.

13.4.4.35.2 Test Summary

a. Check alarms, controls, and interlocks.
b. Recirculate pool water through heat exchangers, bypassing filter demineralizer.
c. Check operation of filter valves, precoat pumps and filter air pumps.
d. Simulate resin precoating and backwashing operation of filter demineralizer, using demineralized water only (and air as required). After satisfactory simulation and when general cleanliness of fuel pool and refueling floor area warrants, charge resins and place filter demineralizer in routine service. Verify system flow rates from pump headflow characteristics.
e. Check level alarms in pool and surge tanks against actual changes in level.

13.4.4.36 Fuel Handling Equipment Test (GE-11) Equipment covered in this category will be tested with a load equivalent to dummy fuel or blade guide assemblies through dry run simulations of the required operations. This is not one coordinated test of a system, but consists of many separate operations using different pieces of equipment. The equipment is tested on the operating floor, in the fuel storage pool, and both over and in the reactor vessel. 13.4-34

BFN-18 13.4.4.36.1 Prerequisites

a. Refueling platform installed.
b. Fuel preparation machine installed.
c. Fuel racks installed.
d. Electrical Services operational.
e. Control circuits checked and available for service.
f. Pumps and motor properly lubricated, pump shaft packing installed, motors recently meggered and tested to ensure correct rotation, motor ventilation not blocked. All items pertaining to initial start and operation of pumps and motors completed.
g. Refueling platform air complete.

13.4.4.36.2 Test Summary (not necessarily in chronological order)

a. Tests in the storage pool.
1. Install fuel pool gates and fill pool with water. Pressurize seals if necessary. Inspect for leakage.
2. Check fuel preparation machine with simulated dummy fuel assembly.

This also checks auxiliary tools such as channel handling tool and channel bolt wrench.

3. Check fixed overhead lights and movable underwater lights to assure adequate visibility for fuel and blade handling and transfer operations.
4. Check underwater vacuum cleaner.
5. Operate refueling platform over storage pool. Check all equipment on the refueling platform. Transfer fuel assemblies and control blades between storage racks with the grapple. Check all grapple controls and interlocks.

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6. Use jib crane to transport simulated dummy fuel assemblies from storage racks to fuel preparation machine work areas.
b. Tests over reactor vessel.
1. Set service platform assembly on vessel flange. Mount jib crane on service platform and use for installing, removing, or shuffling simulated dummy fuel assemblies, and control blades.
2. Verify procedural methods and tools for:

(a) Removal and replacement of steam dryer assembly. (b) Removal and replacement of shroud head steam separator assembly. (c) Removal and replacement of control rod blades, fuel support pieces, and control rod guide tubes. (d) Removal and replacement of incore flux monitor strings. (e) Removal and replacement of jet pump nozzles and risers. All of the above tests recognize the shielding requirements of doing the job "hot" and attempt to simulate "normal operating" conditions.

3. Transfer simulated dummy fuel assemblies and control blades between the storage pool and the reactor vessel, simulating a refueling operation.
4. Obtain representative values of the time required to do all operations normally in the critical path of a refueling outage.
5. Check installation and removal of shield plugs in the designated peripheral positions.

13.4.4.37 Control Rod Drive Manual Control Tests (GE-9) 13.4.4.37.1 Prerequisites

a. Construction completed on CRD position indicating system.
b. Construction completed on reactor manual control system.

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c. It is not necessary that drives be installed or the hydraulic system functioning.

However, if drives are installed or hydraulic system is operating, the isolation valves to each drive should be closed on the hydraulic control module.

d. The hydraulic control units must be installed and wiring complete and checked.

13.4.4.37.2 Test Summary

a. Verify all rod blocks.
b. Check that rod position information system functions properly.
c. Check for proper timing sequence of CRD control circuit.
d. Verify proper operation of CRD select circuit.

13.4.4.38 Rod Sequence Control System (RSCS)(GE-29) 13.4.4.38.1 Prerequisites

a. Reactor Manual Control System (ge-9) tested and operational.
b. Rod Worth Minimizer System (GE-24) tested and operational.
c. Control Rod Drive Hydraulic System (GE-10) tested and operational.
d. RSCS electrical cables and panel wiring tested.
e. RSCS power supplies adjusted.
f. RSCS self test feature tested and operational.
g. All control rods initially in.

13.4.4.38.2 Test Summary

a. Check power supplies for proper voltage.
b. Check Rod Group push button illuminate properly.

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c. Check the system for group verification, full in bypass tests, fence tests, full out bypass tests, group select blocks, insert blocks, sequence tests and notch control tests.
d. Check the systems for proper push button illumination in the tests.
e. Test the system for annunciation of actual or simulated deviations from the normal sequencing.

13.4.4.39 Reactor Water Cleanup System Test (GE-2) 13.4.4.39.1 Prerequisites

a. AC power from 480-V shutdown board available.
b. AC and DC control power available.
c. Condensate or demineralized water available to supply cleanup pump suction.
d. Control and service air supply available.
e. RBCCW system operational.
f. Condensate storage and supply system operational.
g. Liquid radwaste system functional to receive discharge fluids.
h. Solid radwaste system operational.
i. A supply of ion exchange resins available.
j. Strainers installed upstream of the reactor water cleanup recirculation pump.

The system is flushed, cleaned, and initially checked out while the reactor vessel is empty for the installation of drive mechanisms, by supplying it with demineralized water and routing the discharge either to radwaste or to the hotwell. However, the system cannot be completely checked during the preoperational phase because full temperature and pressure conditions are required in the reactor for "normal" system operation to complete the tests. The filter demineralizer may be operated only when precoated; otherwise, it must be bypassed. 13.4.4.39.2 Test Summary

a. Check operation of pressure control station with simulated pressure input signals.

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b. Check operation of main cleanup pumps by pumping first to the hotwell or radwaste and then to reactor. Do not pump to reactor until filters and demineralizer are fully checked out to prevent injecting pool quality water into the reactor. Check water quality.
c. Check operation of filter demineralizers, holding pumps, precoat equipment and all associated equipment. Perform all required operations, such as precoating, normal operation, standby recirculation, and backwashing. Be sure that system is set up so that filter breakthrough will not dump impurities into the reactor (route to radwaste for initial operation).
d. Simulate pumping of spent resin to radwaste system (pump nonradioactive sludge water generated during preoperational testing program).
e. Check operation of all valve and pump interlocks by simulated signals to appropriate instrumentation.
f. Check calibration, alarm, and/or trip (interlock) set points of all instrumentation.
g. After system is proven to be operational in all modes of operation possible to demonstrate without pressure or temperature in the reactor, charge filter demineralizers. Place the system in normal service when water is in the reactor during later preoperational testing.

13.4.4.40 Offgas Recombiner and Charcoal System (GE-34) 13.4.4.40.1 Prerequisites

a. Required functional testing and quality assurance documentation on all electrical and mechanical components is complete.
b. AC and DC control power is available.
c. Proper air supply available.
d. Preliminary and final cleaning complete by construction.

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e. Helium leak tests completed on all valves and components.
f. Filter media installed in the pre- and after-filters.
g. Recombiner catalyst installed.
h. Required quantity of activated charcoal installed in the absorber vessels.
i. Refrigeration machines placed in service and proper operation of the glycol circulating pumps verified.
j. Vault air-conditioning units placed in service and verification obtained that units can maintain the desired vault temperature, pressure, and relative humidity throughout a given test period.
k. Steam available at 350 lbs/hr and between 180 to 250 psi.
l. Condensate supply available.

The system cannot be completely checked during the preoperational phase because operating temperature and pressure conditions are required in the main condenser for "normal" system operation in order to complete the testing. If steam supply is not available, flow testing of the system must be deferred until nuclear steam is available. The hydrogen analyzers will not be functionally tested until the system is placed in "normal" operation. In all cases where redundant components exist, the "B" component flow path will be isolated and only the "A" flow path will be used. 13.4.4.40.2 Test Summary

a. Calibrate all instruments and check all alarms and set points.
b. Verify that the sensors and associated instrumentation perform the proper function.
c. Verify that all valves operate properly by functionally testing them.
d. Check all components of the system including the pre- and after-filters.
e. Verify that both of the level controllers in the offgas condenser will maintain proper level control.
f. Verify the proper logic on all primary and secondary valves.

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g. Verify that the pre- and after-filters perform efficiently by inplace DOP testing with an air flow equivalent to the nominal rated operating flow.
h. After the system is proven to be operational in all modes of operation the RECHAR system should be lined up for operation in accordance with station operating instructions. Test apparatus should be secured and the system made leak tight and otherwise ready for service.

13.4.4.41 Standby Liquid Control System Test (GE-3) 13.4.4.41.1 Prerequisites

a. Construction completed on entire system including electrical and mechanical.
b. Normal air supply available.
c. Demineralized water supply available.
d. Instrumentation checked out and calibrated.

13.4.4.41.2 Test Summary All portions of this test, except the actual pumping rate into the reactor (item e. below) may be done at any time regardless of the status of the reactor vessel (full or empty, head on or off).

a. Calibrate instruments and check all alarm and set points.
b. Fill standby liquid control solution tank with demineralized water and operate the injection pumps, recirculating to the tank.
c. Check set point of the pump discharge relief valves.
d. Check control circuits for the explosive injection valves thoroughly before connecting to the valves. Use a dummy resistance to simulate the valve during the circuit checkout.
e. Turn the keylock switch to each channel to fire the explosive valve and start the injection pump. Measure pumping rates into the reactor.
f. Check the interlock with the reactor water cleanup system to ensure isolation when the standby liquid control system is actuated.
g. Check operation of the standby liquid control solution temperature controls and air sparger.

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h. Fill test tank with demineralized water and operate the injection pumps in simulated test mode, recirculating to test tank. Stop all leakage from the pump packings.
i. After the system has been demonstrated by the foregoing tests, replace the valve explosive cartridges. Add the required boron chemical to the standby liquid control solution tank. Mix and sample. This should be done very shortly before fuel loading. Recheck that the pump packings are not leaking.

13.4.4.42 Core Spray System Test (GE-12) 13.4.4.42.1 Prerequisites

a. Reactor vessel available to receive water and with vessel head and shroud head removed for observation.
b. Pressure suppression chamber filled with water to operating level.

13.4.4.42.2 Test Summary

a. Calibrate all instrumentation.
b. Check alarms, controls, and interlocks including complete verification of automatic system starting controls.
c. Operate pumps by recirculating to the torus in the test mode. Verify pump and system performance from manufacturer's headflow curves and measured system pressures.
d. Check operation of all motor-operated valves.
e. Initiate system automatically and verify that valves open and that pumps start.
f. Isolate pump suction from torus and route to receive pump supply directly from condensate storage tank. Spray into reactor vessel. Verify proper flow rate and observe spray pattern. Pumping capability from the torus will be verified by recirculating through the test line.

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g. Simulate the accident condition simultaneously with a power failure and observe proper sequential operation of system pumps and valves. This test is run concurrently with the containment cooling system automatic operation test and the diesel generator automatic starting test.

13.4.4.43 HPCI System Test (GE-13) 13.4.4.43.1 Prerequisites

a. Construction completed on entire system.
b. Torus filled to normal water level.
c. Steam from auxiliary boiler available for HPCI turbine.
d. An adequate quantity of water in the condensate storage tank available.
e. All instrumentation checked and calibrated.

13.4.4.43.2 Test Summary

a. Check out the functional capability of all components needed to operate under simulated accident conditions as far as is practical. Final full capacity testing of the HPCI system may be deferred until adequate steam supply is available.
b. Check all components of the system including the turbine, pumps, valves, and associated instrumentation.
c. Verify that the system satisfies its design objective and also determine reference characteristics such as differential pressures and flow rates that can be used as base points for check measurements in subsequent testing of the system.

13.4.4.44 Residual Heat Removal System Test (GE-5) All portions of this system are tested during the preoperational testing of the various modes of the RHRS: Containment Spray and LPCI (see paragraph 13.4.4.45) and Shutdown Cooling (see paragraph 13.4.4.46). 13.4-43

BFN-18 13.4.4.45 Containment Spray and LPCI Test (RHR System) (GE-5) 13.4.4.45.1 Prerequisites

a. Torus filled to normal water level with demineralized water.
b. Instrumentation checked and calibrated.
c. Construction complete on system.
d. Electrical AC and DC power available.

13.4.4.45.2 Test Summary

a. Check alarms, controls, and interlocks.
b. Operate RHR pumps by recirculating in the test mode to torus only (not drywell). Verify proper performance by using installed pressure and flow instrumentation.
c. Check flow rate in LPCI mode by pumping from the torus to the reactor vessel.
d. Flow test several drywell spray nozzles before installation on the spray headers. Inspect all nozzles before installation to verify cleanliness and proper opening size.
e. Blow air through drywell spray headers to verify that all nozzles are installed properly.
f. With containment spray valves closed and locked out of service, initiate system automatically and verify proper pump starting for normal power and emergency power modes of operation.
g. Check the RHR crossconnection between units to demonstrate circulation of torus on reactor vessel water from Unit 1 through the A and C RHR loops of Unit 2.

13.4.4.46 Shutdown Cooling Test (RHR System) (GE-5) 13.4.4.46.1 Prerequisites

a. Construction completed.

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b. Electrical power available.
c. Vessel water at normal level.
d. All instrumentation installed and calibrated.

Test requires water in reactor vessel. System may not be sufficiently complete at the time of reactor vessel hydrotest to do preoperational tests at that time, but performance tests on the pumps may be accomplished at that time. 13.4.4.46.2 Test Summary

a. Check operation of all motor-operated valves.
b. Check interlocks in valve and pump control circuits.
c. Measure system pressures where possible and determine flow rates from pump characteristic curve.
d. Measure closing time of isolation valves.
e. Demonstrate proper functioning of valves in the RPV head spray subsystem.

(Demonstration of flow capability may be deferred until after fuel is loaded and the RPV head is installed.) 13.4.4.47 RCIC System Test (GE-6) 13.4.4.47.1 Prerequisites

a. Construction completed on system.
b. Electrical power available.
c. Steam available from the auxiliary boiler for turbine testing.
d. Instrumentation calibration completed.

13.4.4.47.2 Test Summary

a. Verify the operability of system logic and equipment under no-flow conditions.
b. Verify proper integrated system response to manual and automatic controls using the site boiler to supply steam to the turbine.

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BFN-18 13.4.4.48 Nuclear System Safety and Relief Valve Test (GE-4) 13.4.4.48.1 Prerequisites

a. Electrical power available.
b. All valves properly installed.
c. Discharge the vacuum breaker valves installed and checked.
d. Instrumentation installed and calibrated.
e. Drywell Control Air System operable.

13.4.4.48.2 Test Summary

a. Safety and relief valves will be installed as received from the factory. (Set points were factory adjusted and verified and are indicated on the valves.)
b. Verify proper operation of remote controlled relief valves from main control room.
c. Check automatic blowdown function of the relief valves with a simulated pressure signal.
d. Check automatic initiation of contacts, relays, and logic.

13.4.4.49 Reactor Protection System Test (GE-21) 13.4.4.49.2 Prerequisites

a. All safety system sensors installed and calibrated.
b. All wiring installed and checked for continuity.

13.4.4.49.2 Test Summary

a. Operate RPS MG sets with a resistive load to check capacity and regulation.
b. Energize buses; check controls and power source transfer.

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c. Check operation, pick up and drop out voltages of the protection system relays.
d. Check each safety sensor for operation of proper relay.
e. Using test signals, verify alarm, interlock and scram set points. Check proper operation of level switches by varying water level in the reactor vessel, measuring the actual water level against a suitable reference point such as the vessel flange.
f. Check all positions of the reactor mode switch for proper interlocks and bypass functions.

13.4.4.50 Miscellaneous Neutron and Gamma Radiation Instrument Systems Testing (GE-22 A-E, 23, 25, and 26) This testing includes the following systems:

a. Source range monitoring system.
b. Intermediate range monitoring system.
c. SRM/IRM chamber drives.
d. Average power range monitoring system.
e. Local power range monitoring system.
f. Traversing incore probe system.
g. Area radiation monitoring system.
h. Process liquid and gas monitors.

13.4.4.50.1 Prerequisites

a. Systems (a through h, paragraph 13.4.4.50) completed and construction checks completed.
b. Instrumentation calibrated.
c. Applicable alarm set points calculated.

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BFN-18 13.4.4.50.2 Test Summary The following types of preliminary testing are required (where applicable) prior to fuel loading.

a. Check continuity and resistance to ground of all signal and power cables.
b. Check response and calibration of all channels with simulated input signals.
c. Check alarm and trip set points.
d. Check chamber response to bugging sources.
e. Check all interlocks with the reactor manual control system.
f. Check operation and position indication of all SRM/IRM chamber drives.
g. Using dummy TIP chamber, insert calibration probe in all incore calibration tubes. Verify capability to insert more than one calibration probe in the crosscalibration guide tube. Repeat with all five TIP machines operable.
h. Install all incore, SRM and IRM chambers and verify final system operability.

13.4.4.51 Process Computer System (Rod Worth Minimizer Function) Test (GE-24) 13.4.4.51.1 Prerequisites

a. Computer installation completed.
b. Electrical power available.

13.4.4.51.2 Test Summary After control rod drive system is operational, withdraw control rods in various sequences to expose the rod worth minimizer function of the process computer to simulated operational conditions. These withdrawal patterns should simulate the conditions required for the following operations:

a. Check all programmed normal rod withdrawal sequences for satisfactory performance.
b. Check different short-term sequences within the sequenced rod groups for satisfactory performance.

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c. Attempt improper rod withdrawal at various points in the withdrawal sequence, and verify that the action is blocked.
d. Determine capability to insert drive mechanisms out of sequence to the extent permitted by the rod worth minimizer function. Insertion of two rods out of sequence should be possible.
e. Check all alarms by simulated or actual error conditions:
1. Low power alarm
2. Printing
3. Computer error
4. Input/output error
5. Select error
6. Select block
7. Insert block
8. Withdraw block
9. Bypass block
f. Check all controls
1. Sequence A mode
2. Sequence B mode
3. Shutdown margin mode
4. Scan exit
5. Print log
6. Error clean
g. Check all displays and information printout.
1. Group identification
2. Withdrawal error readout
3. Insertion error readout
4. Printout rod position from scan and from memory for several rod withdrawal patterns 13.4.4.52 Reactor Recirculation System Test (GE-8) 13.4.4.52.1 Prerequisites
a. 4160-V electrical power available
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c. Reactor hydrotest and cleaning completed
d. Water in the vessel during pump tests 13.4.4.52.2 Test Summary This test will determine recirculation loop (recirculation pumps and jet pumps) characteristics to the degree possible with cold water conditions.
a. Operate all recirculation loop valves. Simulate the automatic operation of these valves following a breach of one of the recirculation loops.
b. Calibrate loop instrumentation and check controls and interlocks.
c. Operate recirculation pumps and MG sets at reduced speed.
d. Check flow control transient operation within the range permitted by cold water and atmospheric pressure in reactor. Optimize controller settings for system linearity and response time requirements.

13.4.4.53 Primary Containment Isolation System Test (GE-30) 13.4.4.53.1 Prerequisites

a. Construction completed.
b. Control air available.
c. Reactor protection system operating.
d. Electrical power available.

13.4.4.53.2 Test Summary

a. Check manual and auto logic control of all isolation valves.
b. Check valve stroke time.
c. Check valve control from backup control centers.

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BFN-18 13.4.4.54 Liquid Radioactive Waste Disposal System Test (GE-20) 13.4.4.54.1 Prerequisites

a. Construction completed.
b. Instrumentation installed and calibrated as required.
c. Electrical power available.
d. Proper air supply available.

13.4.4.54.2 Test Summary After fuel is loaded in the reactor, all drains from the reactor, fuel pool, or interconnecting auxiliary systems must be considered to be potentially radioactive. Therefore, most of the Liquid Radioactive Waste Disposal System must be tested and operational before fuel loading. (For testing of Solid Radwaste System see paragraph 13.4.4.30.)

a. Check all controls and interlocks.
b. Recheck all air-operated valves.
c. Pumps and tanks.
1. Clean tanks mechanically.
2. Fill with demineralized water.
3. Check pump operation in recirculation mode whenever possible.
4. Simulate operations associated with the particular tank, such as draining or filling, recirculating, sampling, and processing to a filter demineralizer, another tank, or overboard discharge.
d. Filters (waste collector and floor drain).
1. Perform all required operations such as precoating, normal operation, recirculation, and backwashing.
2. Check operation of filter components only after precoating, using demineralized water only, until system operation is acceptable.

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e. Sumps (Drywell, Reactor, Standby Gas Treatment, Turbine and Radwaste Buildings).
1. Clean out sumps. Check and set level switches.
2. Fill sumps with water.
3. Check operation of sump pumps and proper functioning of level controls, including isolation valves on containment.
4. Verify discharge to proper collection tank in radwaste with no back flow or leakage en route.
5. Check valving to heat exchangers where supplied, and operation of temperature controls.
f. Chemical Waste System.
1. Partially fill chemical waste tank with demineralized water.
2. Demonstrate all pumping operations with demineralized water only; recirculation, sampling, and transfer to floor drain collector tank.
3. Test chemical addition equipment with demineralized water initially, then add chemicals and demonstrate the neutralizing operation.
g. Spent resin system.
1. Simulate transfer of resins from the waste demineralizer to the spent resin tank.
2. Verify resin transfer capability by actual transfer of resins (Perform near end of test program with little or no radioactivity present, or devise means for catching reclaiming resins.)
h. Filter sludge processing.
1. Simulate transfer of sludge from the several filters and filter demineralizers to the respective backwash receiver tanks using water only. Simulate transfer from backwash receiver tanks to phase separator.
2. Repeat sludge transfers with actual filter aid material.

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BFN-18 13.4.4.55 Standby AC Power System Test (GE-31) 13.4.4.55.1 Prerequisites

a. Electric system tests (see paragraphs 13.4.4.10-.11 and .47) completed prior to this test.
b. Vendor representative for the power source (i.e., emergency diesel generator) present for the test.
c. Devices on the control panels such as controls, interlocks, indicators, push buttons, and annunciators (including their associated circuitry) operative, tested, and adjusted before the preoperational test is started.

13.4.4.55.2 Test Summary

a. Perform load tests to verify diesel generator capacity.
b. Check automatic start of the available diesels upon loss of power.
c. Check automatic sequencing of major motors which occurs during accident conditions.
d. Check automatic redistribution of load after failure of one diesel.
e. Perform an integrated test demonstration operation of all plant safety systems subsequent to a loss of offsite power.

13.4.4.56 DC Power System Test (GE-32) 13.4.4.56.1 Prerequisites

a. Complete all construction testing. Complete all controls, polarity, meggering and alarm checks for battery chargers, batteries, and distribution center.
b. Complete battery installation including initial charge and float control.
c. Complete adequate room ventilation and air exhauster system.

13.4.4.56.2 Test Summary

a. Checkout operation of battery chargers.

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b. Check out all system operation actions and alarms.
c. Perform battery service test.

13.4.4.57 Feedwater Control System Test (GE-1) 13.4.4.57.1 Prerequisites

a. System completed.
b. Control air available.
c. Electrical power available.
d. Instrumentation calibrated.

13.4.4.57.2 Test Summary

a. Check proper operation of instruments and controls.
b. Simulate reactor level variations and verify proper response. (For other testing of the condensate and feedwater system see paragraphs 13.4.4.2 and 13.4.4.25.)

13.4.4.58 Unit Preferred Power System Test (GE-33) 13.4.4.58.1 Prerequisites

a. Complete all construction testing for the motor generator sets, transformers and their associated devices.
b. Complete all construction testing for the panelboards and circuits to them (including continuity and phasing checks, dielectric tests and alarms).
c. AC power from 480-V shutdown board available.
d. 250-V DC power available.

13.4.4.58.2 Test Summary

a. Check proper operation of all motor generator sets.

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b. Check proper operation of all automatic transfer operations.
c. Checkout all alarm and indication instrumentation.

13.4.4.59 Process Computer Tests (in addition to rod worth minimizer) (GE-28) 13.4.4.59.1 Prerequisites

a. System complete, power available.
b. All systems that provide input to computer completed and connected.

13.4.4.59.2 Test Summary

a. Perform computer self checks.
b. Verify computer operation.

Computer testing will continue during the startup and power test programs. 13.4.4.60 Primary Containment Atmospheric Monitor (GE-35) 13.4.4.60.1 Prerequisites

a. Required testing and quality assurance documentation on all cables for the primary containment atmospheric control (inerting system) complete.
b. Temporary jumpers marked and recorded.
c. Required testing and quality assurance documentation on all electrical equipment for the primary containment control (inerting system) complete.
d. All instruments have been installed and checked.
e. An inspection has been made of the electrical and mechanical equipment installation and a tentative transfer of the system has been made to DPP.
f. The applicable drawings are maintained by the configuration control group in the master file. The drawings are marked to reflect the latest system configuration. These drawings, with the revisions listed, functionally reflect the operability of the system as it will be preoperationally tested.
g. The control air system must be available for service.
h. All calibration gas supplies are connected and operable.

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i. Construction of system piping complete.

13.4.4.60.2 Test Summary

a. Verify that the log rad monitors operate properly.
b. Verify that the oxygen analyzers operate properly.
c. Verify that the hydrogen analyzers operate properly.

13.4.4.61 Leak Detection System Test (GE-19) 13.4.4.61 Prerequisites

a. Electrical checks completed and all buses and MCCs supplying motors, solenoids, and instruments energized.
b. Control air system available.
c. Instruments checked and calibration data sheets completed.
d. Drywell floor drain sump pumps and drywell equipment drain sump pumps checked for alignment and rotation. Sump pumps and motors lubricated.
e. Variable-flow makeup water supply available to drywell sumps.
f. Floor areas which are drained to drywell floor drain sump, cleaned and flushed to drain.
g. Drywell sumps drained and mechanically cleaned prior to operation of sump pumps.
h. Startup suction strainer installed on drywell sump pumps.

13.4.4.61.2 Test Summary

a. Verify proper operation of the sensors associated with detection of leaks from the primary system during normal plant operation.

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b. Verify logic and active components associated with isolation and annunciation of these leaks.

13.4.4.62 Test of Condenser Circulating Water Pump Operating on Diesel Generator Power (TVA Supplement A to GE-31) 13.4.4.62.1 Prerequisites The following preoperational testing completed:

a. Condenser circulating water system interunit crossties (see paragraph 13.4.4.1).
b. 4.16-kV electrical system (normal auxiliary power system) (see paragraph 13.4.4.11).
c. Diesel generator system and emergency power system (see paragraph 13.4.4.55).

13.4.4.62.2 Test Summary

a. Demonstrate that two diesel generators can be operated successfully in parallel with each other.
b. Start and operate one condenser circulating water pump on power supplied by these diesel generators.

13.4-57

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c _____ 52 ~~lRRi.rfrRtJD _; / BROWNS FERRY l'I UCLEAR PLAHT FlNAl SAFETY ANALYSIS REPORT Typical Precpenticnal Test Sequence FIGURE 13.4*1

BFN-22 13.5 STARTUP AND POWER TEST PROGRAM This section presents a general description of the startup testing that was planned for Browns Ferry and has been retained in the FSAR as a historical reference only. This test description is not in conformance with Regulatory Guide 1.68 and should not be used as a model for future test programs. For a description of the retest program see Section 13.10. There are numerous minor discrepancies between this description and the startup testing actually performed. An accurate description of the startup testing performed and the results of the testing are contained in three reports entitled "Summary Report of Startup Tests." The submittal date of each of these reports is given below. Unit 1 September 27, 1974 Unit 2 May 23, 1975 Unit 3 May 9, 1977 This test description is retained in the FSAR only to serve as a readily available general description of the test program. 13.5.1 Program Description and Objectives 13.5.1.1 General The tests comprising the startup and power test program are conducted primarily to show that the overall plant performance is confirmed in terms of the established design criteria, at all times starting with fuel loading. These criteria and the associated tests have either a safety or economic orientation, while often both aspects of the design are being explored. A most important result of the startup test program is that the operator has available to him valuable data upon which the future, normal and safe operation of the plant can be based. The preoperational startup and power test program may be divided into the following discrete and successive test phases: Phase I Preoperational tests (Subsection 13.4) Phase II Fuel loading and shutdown power level tests Phase III Initial heatup to rated temperature and pressure Phase IV Power testing from 25 to 100 percent of rated output Phase V Warranty demonstrations 13.5-1

BFN-22 The tests performed can be broadly classified as Major Plant Transients (Table 13.5-2), Stability Tests (Table 13.5 3), and a residue of tests directed toward demonstrating correct performance of the numerous auxiliary plant systems; clearly, certain tests may be identified with more than one class. Each test is discussed later, but at this juncture the following comments are given by way of outlining the startup and power test program. The various test points as functions of core thermal power and flow are shown for Unit 1 in Table 13.5-4, Unit 2 in Table 13.5-4 and Unit 3 in Table 13.5-6. 13.5.1.2 Fuel Loading and Shutdown Power Level Tests Fuel loading requires the movement of the full core complement of assemblies from the fuel pool to the core with each assembly identified by number before being placed in the correct coordinate position. The procedure controlling this movement is arranged so that shutdown margin and subcritical checks are made at predetermined intervals throughout the loading, thus ensuring safe loading increments. Specially sensitive neutron monitors maintained at the loading face, as loading progresses, serve to provide indication for the shutdown margin measurements and also allow the recording of the core flux level as each assembly is added. A complete check is made of the fully loaded core to ascertain that all assemblies are properly installed, correctly oriented, and occupying their designated positions. At this point in the program, a number of tests are conducted which are best described as initial shutdown power level tests (Phase II). Chemical and radiochemical tests are made to check the quality of the reactor water before fuel is loaded and to establish base and background levels which will be required to facilitate later analysis and instrument calibrations. Plant and site radiation surveys are made at specific locations for later comparison with the values obtained at the subsequent operating power levels. Shutdown margin checks are repeated for the fully loaded core, and, in turn, criticality is achieved with a prescribed rod sequence, data are recorded for each rod withdrawn. The reactor is made critical by means of a prescribed control rod sequence, using the normal Source Range Monitors (SRM) in conjunction with the operational sources to show that adequate response exists for normal operation. Each control rod drive is subjected to scram and friction testing at ambient conditions. Initially the Intermediate Range Monitors (IRM) are set to maximum gain and are verified for overlap with the SRMs. The process computer is checked to see that it is receiving correct values for the available process variables. Vibration characteristics are determined for selected reactor internal components over a range of cold recirculation flows. The water level profile is determined for a range of water levels and core flows. 13.5-2

BFN-22 13.5.1.3 Initial Heatup to Rated Temperature and Pressure Heatup follows satisfactory completion of the Phase II tests, and further checks are made of coolant chemistry together with radiation surveys at the selected plant locations. All control rod drives are scram timed at rated temperature and pressure; selected drives are timed at intermediate reactor pressures and for different accumulator pressures. The control rod sequence is further investigated to obtain rod pattern versus temperature relationships. The process computer checkout continues as more process variables become available for input. The RCIC and HPCI systems will undergo controlled starts at low rector pressure and at rated conditions; the RCIC system is tested in the quick start mode at 1,020 psig. Correlations are obtained between reactor vessel temperatures at several locations and the values of other process variables as heatup continues. The movements of drywell piping systems as a function mainly of expansion are recorded for comparison with design data. The steam separator-dryer test equipment is checked out and initial samples taken. An intermediate APRM calibration is made using coolant temperature rise data during nuclear heatup. 13.5.1.4 Power Testing from 25 to 100 percent of Rated Output The power test phase comprises the following tests, many of which are repeated several times at the different test levels; consequently, reference should be made to Figures 13.5-1 sheets 1 and 2 (Unit 1) and Figures 13.5-2 sheets 1, 2, and 3 (Units 2 and 3) for the probable order of execution for the full series. It must be appreciated that, while a certain basic order of testing is maintained relative to power ascension, there is, nevertheless, considerable flexibility in the test sequence at a particular power level which may be used whenever it becomes operationally expedient. Coolant chemistry tests and radiation surveys are made at each principal test level to preserve a safe and efficient power increase. Selected control rod drives are scram timed at various power levels to provide correlation with the initial data. The effect of control rod movement on other parameters (e.g., electrical output, steam flow and neutron flux level) is examined for different power conditions. Following the first reasonably accurate heat balance (25 percent power) the IRMs are reset. At each major power level (25, 50, 75, and 100 percent) the LPRMs are calibrated; the APRMs are calibrated at each new power level initially and following LPRM calibration. Completion of the process computer checkout is made for all variables, and the various options are compared with backup calculations as soon as significant power levels are available. Further tests of the RCIC and HPCI systems are made with and without injection into the reactor pressure vessel. Collection of data from the system expansion tests is completed for those piping systems which had not previously reached full operating temperatures. The axial and radial power profiles are explored fully by means of Traversing Incore Probe (TIP) system at 13.5-3

BFN-22 representative power levels (25, 50, 75, and 100 percent) during the power ascension. Core performance evaluations are made at all test points above the 10 percent power level and for selected flow transient conditions; the work involves the determination of core thermal power, maximum fuel rod surface heat flux, and the minimum critical heat flux ratio (MCHFR). Overall plant stability in relation to minor perturbations is shown by the following group of tests which are made at all test points: Flux response to rods Pressure regulator setpoint change Water level setpoint change Bypass valve opening For the first of these tests a centrally located control rod is moved and the flux response noted on a selected LPRM chamber. The next two tests require that the changes made should approximate, as closely as possible, a step change in demand, while for the remaining test the bypass valve is opened as quickly as possible. For all of these tests the plant performance is monitored by recording the transient behavior of numerous process variables, the principal one of interest being neutron flux. Other imposed transients are produced by step changes in demanded core flow, dropping a feedwater heater, and failing the operating pressure regulator to permit takeover by the backup regulator. Table 13.5-2 indicates the power and flow levels at which all these stability tests are performed. The category of major plant transients includes full closure of all the main steam isolation valves, fast closure of the turbine generator control valves, fast closure of turbine generator stop valves, loss of the main generator and offsite power, tripping a feedwater pump, and several trips of the recirculation pumps. The plant transient behavior is recorded for each test and the results may be compared with the predicted design performance. Table 13.5-2 shows the operating test conditions for all the proposed major transients. A test is made of the relief valves in which the capacity, leak tightness, and general operability is demonstrated. At all major power levels the jet pump flow instrumentation is calibrated and carryover/carryunder measurements are made to facilitate assessment of the steam separator-dryer performance. The as-built characteristics of the recirculation pump drives are investigated as soon as operating conditions permit full core flow. The local control loop performance (based on the drive motor, fluid coupler, generator, drive pump, jet pumps, and control equipment is checked. The vibration is testing conducted at the cold flow 13.5-4

BFN-22 condition is extended to measurements at several power conditions as the operating power level is raised. 13.5.1.5 Warranty Demonstrations The final test phase consists of a warranty demonstration in which the steaming rate and quality can be shown to comply with contractual obligations. 13.5.2 Discussion of Startup Tests 13.5.2.1 General All tests comprising the startup and power test program are discussed in paragraph 13.5.2.2 with reference to the particular test purpose, brief description, and statement of acceptance criteria, where applicable. In describing the purpose of a test, an attempt is made to identify those operating and safety-oriented characteristics of the plant which are being explored. Where applicable, a definition of the relevant acceptance criteria for the test is given and is designated either "Level 1" or "Level 2." A Level 1 criterion normally relates to the value of a process variable assigned in the design of the plant, component systems, or associated equipment. If a Level 1 criterion is not satisfied, the plant will be placed in a suitable hold-condition until resolution is obtained. Tests compatible with this hold-condition may be continued. Following resolution, applicable tests must be repeated to verify that the requirements of the Level 1 criterion are satisfied. A Level 2 criterion is associated with expectations relating to the performance of systems. If a Level 2 criterion is not satisfied, operating and testing plans would not necessarily be altered. Investigations of the measurements and of the analytical techniques used for the predictions would be started. For transients involving oscillatory response, the criteria are specified in terms of decay ratio (defined as the ratio of successive maximum amplitudes of the same polarity). The decay ratio must be less than unity to meet a Level 1 criterion and less than 0.25 to meet Level 2. 13.5.2.2 Test Purpose, Description and Acceptance Criteria--Unit 1 TEST NUMBER 1 -- CHEMICAL AND RADIOCHEMICAL Purpose The principal objectives of this test are: (a) to maintain control of and knowledge about the quality of the reactor coolant chemistry, and (b) to determine that the 13.5-5

BFN-22 sampling equipment, procedures, and analytic techniques are adequate to supply the data required to demonstrate that the coolant chemistry meets water quality specifications and process requirements. Secondary objectives of the test program include data to evaluate the performance of the fuel, operation of the demineralizers and filters, operation of the offgas system and calibration of certain process instruments. Description Before fuel loading, a complete set of chemical and radiochemical samples will be taken to ensure that all sample stations are functioning properly and to determine initial water quality. Subsequent to fuel loading during reactor heatup and at major power level changes, samples will be taken and measurements will be made to determine the chemical and radiochemical quality of reactor water and reactor feedwater, amount of radiolytic gas in the steam, gaseous activities leaving the air ejectors, decay times in the offgas lines, and performance of filters and demineralizers. Calibrations will be made of monitors in the stack, liquid waste system, and liquid process lines. Criteria Level 1 Water quality must be known and must conform to the water quality specifications at all times. The activity of gaseous and liquid effluents must be known and they must conform to license limitations. Chemical factors defined in the Technical Specifications must be maintained within the limits specified. TEST NUMBER 2 -- RADIATION MEASUREMENTS Purpose The purpose of this test is to determine the background radiation levels in the plant environs prior to operation for base data on activity buildup and to monitor radiation at selected power levels to assure the protection of personnel during plant operation. Description A survey of natural background radiation throughout the plant site will be made before fuel loading. Subsequent to fuel loading, during reactor heatup and at power 13.5-6

BFN-22 levels of 25, 50, and 100 percent of rated power, gamma radiation level measurements and, where appropriate, thermal and fast neutron dose rate measurements will be made at significant locations throughout the plant. All potentially high radiation areas will be surveyed. Criteria Level 1 The radiation doses of plant origin and occupancy times shall be controlled consistent with the guidelines of the standards for protection against radiation outlined in TVA Radiological Control Instructions. TEST NUMBER 3 -- FUEL LOADING Purpose The purpose of this test is to load fuel safely and efficiently to the full core size. Description Before fuel loading, control rods will be installed and tested. A neutron source of approximately 10 neutrons per sec will be installed near the center of the core. At least three neutron detectors calibrated and connected to high flux scram trips will be located to produce acceptable signals during loading. Fuel loading will begin at the center of the core and proceed radially to the fully loaded configuration. The following checks will be performed as each cell is loaded.

1. Subcriticality Check - A control rod surrounded by fuel in the vicinity of the cell to be loaded will be completely withdrawn; the core must remain subcritical.

Then the rod will be reinserted.

2. Control Rod Functional Test - The rod in the cell to be loaded will be completely withdrawn and reinserted.
3. Fuel Loading - Two fuel assemblies will be loaded, the blade guide removed, and the remaining two fuel assemblies loaded to complete the sour-assembly cell.
4. The Subcriticality Check will be repeated.
5. The Control Rod Functional Test will be repeated. This also serves as a Subcriticality Check on the loaded fuel cell.

13.5-7

BFN-22 Shutdown margin demonstrations will be performed periodically during fuel loading. Criteria Level 1 The criteria for successful completion of this test are: (a) the partically-loaded core must be subcritical by at least 0.38 percent k/k with the geometrically strongest rod fully withdrawn, (b) the core is fully loaded, and (c) the full core shutdown margin demonstration has been completed. TEST NUMBER 4 -- FULL CORE SHUTDOWN MARGIN Purpose The purpose of this test is to demonstrate that the reactor will be subcritical throughout the first fuel cycle with any single control rod fully withdrawn. Description This test will be performed in the fully loaded core at ambient temperature in the xenon-free condition. Subcriticality will be demonstrated with the strongest rod fully withdrawn and a series of calibrated rods pulled to a position calculated to be equal to a shutdown margin specified to account for expected reactivity changes during core lifetime. Criteria Level 1 The fully loaded core must be subcritical throughout the fuel cycle with any rod fully withdrawn. TEST NUMBER 5 -- CONTROL ROD DRIVE Purpose The purposes of the Control Rod Drive System test are: (a) to demonstrate that the Control Rod Drive (CRD) system operates properly over the full range of primary coolant temperatures and pressures from ambient to operating, and particularly that thermal expansion of core components does not bind or significantly slow control rod movements, and (b) to determine the initial operating characteristics of the entire CRD system. 13.5-8

BFN-22 Description The CRD tests performed during Phases II through IV of the startup test program are designed as an extension of the tests performed during the preoperational CRD system tests. Thus, after it is verified that all control rod drives operate properly when installed, they are tested periodically during heatup to assure that there is no significant binding caused by thermal expansion of the core components. A list of all CRD tests to be performed during startup testing is given in Table 13.5-1. Criteria Level 1 (a) Each drive speed in either direction (insert or withdraw) must be 3.0 +/- 0.6 in. per sec, indicated by a full 12-foot stroke in 40 to 60 sec. 13.5-9

BFN-22 (b) The average scram insertion time of all operable control rods, based on the deenergization of the scram pilot valve solenoids as time zero, shall be no greater than: Percent Inserted from Average Scram Insertion Fully Withdrawn Times (sec) 5 0.375 20 0.90 50 2.0 90 5.0 (c) The average of the scram insertion times for the three fastest control rods of all groups of four control rods in a two-by-two array shall be no greater than: Percent Inserted from Average Scram Insertion Fully Withdrawn Times (sec) 5 0.398 20 0.954 50 2.120 90 5.3 (d) The maximum scram insertion time for 90 percent insertion of any operable control rod shall not exceed 7.00 seconds. Level 2 (a) With respect to the control rod drive friction tests, if the differential pressure variation exceeds 15 psid for a continuous drive-in, a setting test must be performed, in which case, the differential setting pressure should not be less than 30 psid nor should it vary by more than 10 psid over a full stroke. Lower differential pressures in the setting tests are indicative of excessive friction. (b) With the charging valve closed, the 90 percent scram time should be greater than 1.50 second at ambient (0 psig) reactor pressure. (c) Scram times with normal accumulator charge should fall within prescribed time limits. 13.5-10

BFN-22 TEST NUMBER 6 -- SRM PERFORMANCE AND CONTROL ROD SEQUENCE Purpose The purpose of this test is to demonstrate that the operational sources, SRM instrumentation, and rod withdrawal sequences provide adequate information to achieve criticality and increase power in a safe and efficient manner. The effect of typical rod movements on reactor power will be determined. Description The operational neutron sources will be installed and source range monitor count rate data will be taken during rod withdrawals to critical and compared with stated criteria on signal and signal count-to-noise count ratio. A control rod withdrawal sequence has been calculated which completely specifies control rod withdrawals from the all-rods-in condition to the rated power configuration. Rod patterns will be recorded periodically as the reactor is heated to rated temperature. As each rod group is completed during the power ascension, the electrical power, steam flow, and APRM response will be recorded. Movement of rods in a prescribed sequence is monitored by the Rod Worth Minimizer and Rod Sequence Control System which will prevent acceptable out-of-sequence control rod movements during startup or shutdown. Criteria Satisfaction of the following criteria constitutes adequate source and SRM relationships. Level 1 There must be a neutron signal count-to-noise count ratio of at least 2 to 1 on the required operable SRMs or Fuel Loading Chambers. There must be a minimum count rate of 3 counts/sec. on the required operable SRMs or Fuel Loading Chambers. 13.5-11

BFN-22 TEST NUMBER 10 -- IRM CALIBRATION Purpose The purpose of this test is to adjust the Intermediate Range Monitor system to obtain an optimum overlap with the SRM and APRM systems. Description Initially, the IRM system is set to maximum gain. After the APRM heatup calibration and after the first heat balance calibration of the APRM's, the IRM-APR, overlap will be checked and the IRM gains adjusted if necessary to improve the IRM system overlap between the SRM's and IRM's. Criteria Level 1 (a) Each IRM channel must be adjusted so that overlap with the SRM's and APRM's is assured. (b) The IRM's must produce a scram at 120/125 of full scale. (c) The IRM reading 120/125 of full scale on range 10 will be set equal to or less than 30 percent of rated power. TEST NUMBER 11 -- LPRM CALIBRATION Purpose The purpose of this test is to calibrate the Local Power Range Monitoring system. Description The LPRM channels will be calibrated to make the LPRM readings proportional to the average heat flux in the four corner fuel rods surrounding each chamber at the chamber elevation. The initial calibration factors are obtained from measurements of axial power distribution, precalculated local power distributions, and precalculated radial power distributions. 13.5-12

BFN-22 Criteria Level 1 With the reactor in the rod pattern and at the power level at which the calibration is to be performed, the meter reading of each LPRM chamber will be proportional to the average heat flux in the four adjacent fuel rods at the height of the chamber. TEST NUMBER 12 -- APRM CALIBRATION Purpose The purpose of this test is to present the methods for calibrating the Average Power Range Monitor Channels. Description The APRM's will be initially adjusted to maximum amplifier gain. The LPRM's and APRM's should begin to respond in the region of 1 to 10 percent of rated power when a low power calibration will be performed based on heat balance calculations during reactor heatup. As soon as the reactor power is high enough to obtain more accurate steady state heat balances to determine actual core thermal power, the APRM channels will be calibrated to read percent of core thermal power. Criteria Level 1 (a) The APRM channels must be calibrated to read equal to or greater than the actual core thermal power. (b) Technical Specifications and Fuel Warranty Limits stated in Section B shall not be exceeded. (c) In the startup mode, all APRM channels must produce a scram at less than or equal to 15 percent of rated thermal power. (d) Recalibration of the APRM system will not be necessary for safety considerations if at least two APRM channels per reactor protection system (RPS) trip circuit have readings greater than or equal to core power. Channels will be considered to be reading accurately if they agree with the heat balance to within plus or minus 7 percent of rated power. 13.5-13

BFN-22 TEST NUMBER 13 -- PROCESS COMPUTER Purpose The purpose of this test is to verify the performance of the process computer under operating conditions. Description GE/PAC computer system program verifications and calculational program validations at static and at simulated dynamic input conditions will be preoperationally tested at the computer supplier's site and following delivery to the plant site. Following fuel loading, during plant heatup and the ascension to rated power, the nuclear steam supply system and the balance-of-plant system process variables sensed by the computer as digital or analog signals will become available. In addition, the test is designed to verify that the computer is receiving correct values of sensed process variables and that the results of performance calculations of the nuclear steam supply system and the balance-of-plant are correct. The purpose of this test is also to verify proper operation of all computer functions at rated power operating conditions. Criteria Level 2 Program OD-1 and P-1 will be considered operational when: (1) the MCHFR calculated by an independent method and the process computer either (a) are in the same fuel assembly and do not differ in value by more than 10 percent, or (b) if the two different fuel assemblies are chosen by the two methods, the CHFR calculated by the other method in each assembly agrees with the MCHFR in that assembly by not more than 10 percent, and (2) when the LPRM calibration factors calculated by the independent method and the process computer agree to within 5 percent. The remaining programs will be considered operational upon successful completion of static testing. TEST NUMBER 14 -- RCIC SYSTEM Purpose The purpose of this test is to verify the operation of the Reactor Core Isolation Coolant (RCIC) system at operating reactor pressure conditions. 13.5-14

BFN-22 Description Flow tests of the RCIC System will be performed at reactor pressures between 150 and 1,020 psig. Except for a final demonstration without warmup, the RCIC turbine will be warmed up before a fast start at each test point. This test is designed to verify proper operation of the RCIC system, determine time to reach rated flow and adjust flow controller in RCIC system for proper flow rate. These tests will first be performed with the system in the test mode so that discharge flow will not be routed to the reactor pressure vessel. The final demonstration will be made so that the discharge flow will be routed to the reactor pressure vessel while the reactor is at partial power. Criteria Level 1 The reactor will be allowed to operate at all conditions, including 100 percent power, if the RCIC can deliver rated flow, 600 GPM, in less than or equal to the rated actuation time, 30 sec, against any reactor pressure between 150 and 1,020 psig. TEST NUMBER 15 -- HPCI Purpose The purpose of this test is to verify the proper operation of the High Pressure Coolant Injection (HPCI) system throughout the range of reactor pressure conditions. Description Flow tests of the HPCI System will be performed at reactor pressures between 150 and 1,020 psig. Except for a final demonstration without warmup, the HPCI turbine will be warmed up before a fast start at each test point. The purpose of this test is to verify proper operation of the HPCI system, determine time to reach rated flow, and adjust the flow controller in HPCI system for proper flow rate. These tests will be performed with the system in the Test Mode so that discharge flow will not be routed to the reactor pressure vessel. The final demonstration will be made so that discharge flow will be routed to the reactor pressure vessel while the reactor is at partial power or following a reactor scram from low power. 13.5-15

BFN-22 Criteria Level 1 The time from actuating signal to required flow must be less than 25 sec at any reactor pressure between 150 and 1,020 psig. With pump discharge at any pressure between 1,220 and 150 psig, the flow should be at least 5,000 gpm. The HPCI turbine must not trip off during startup. TEST NUMBER 16 -- SELECTED PROCESS TEMPERATURES Purpose The purposes of this procedure are to establish the minimum reactor recirculation pump speed which will maintain water temperature in bottom head of the reactor vessel within 145°F (63°C) of reactor coolant saturation temperature as determined from reactor pressure and to provide assurance that the measured bottom head drain temperature corresponds to bottom head coolant temperature during normal operations. Description The applicable reactor parameters will be monitored during the initial heatup, the initial cooldown, and after recirculation pump trips in order to determine that adequate mixing of the reactor water is occurring in the lower plenum of the pressure vessel. The adequacy of the bottom-drain-line thermocouple as a measure of bottom reactor vessel temperature will also be determined. Criteria Level 1

1. The reactor recirculation pump shall not be operated unless the coolant temperatures between the dome and the bottom head drain are within 145°F (63°C) of each other.
2. The pump in an idle recirculation loop shall not be started unless the temperature of the coolant within the loop is within 50°F (10°C) of the active loop temperature.

13.5-16

BFN-22 TEST NUMBER 17 -- SYSTEM EXPANSION Purpose The purpose of this test is to verify that the reactor drywell piping system is free and unrestrained in regard to thermal expansion and that suspension components are functioning in the specified manner. Description Observe and record the horizontal and vertical positions of major equipment and piping in the Nuclear Steam Supply System and auxiliary systems to assure components are free to move as designed. Adjust as necessary for freedom of movement. Criteria Level 1 There shall be no evidence of blocking of the displacement of any system component caused by thermal expansion of the system. Hangers shall not be bottomed out or have the spring fully stretched. Level 2 (a) Displacements of instrumented points with special recording devices shall not vary from the calculated values by more than 50 percent or 0.5 ( 1.27) cm) inches, whichever is smaller. Displacements of less than 0.25 (0.64 cm) inch can be neglected, since 50 percent of this value is bordering on the accuracy of measurement. If measured displacements do not meet these criteria, the system designer must be contacted to analyze the data with regard to design stresses; (b) The trace of the instrumented points during the heatup cycle shall fall within a range of 150 percent of the calculated value from the initial cold position in the direction of the calculated value, and 50 percent of the calculated value from the initial position in the opposite direction of the calculated value; (c) Hangers shall be in their operating range (between the hot and cold settings). 13.5-17

BFN-22 TEST NUMBER 18 -- CORE POWER DISTRIBUTION Purpose The purposes of this test are to: (1) confirm the reproducibility of the TIP system readings, (2) determine the core power distribution in three dimensions, and (3) determine core power symmetry. Description A check of the reproducibility of the TIP traces is made twice: (a) the first time the TIP system is used, and (b) again at a later date after the TIP system has been used a number of times and is "broken in." The check is made with the plant at steady-state condition by producing several TIP traces in the same location with each TIP machine. The traces are evaluated to determine the extent of deviations between traces from the same TIP machine. Core Power distribution, including power symmetry, will be obtained during the power ascension program. Axial power traces will be obtained at each of the TIP locations. Several TIP systems have been provided to obtain these traces--a common location can be traversed by each TIP chamber to permit intercalibration. The results of the complete set of TIP traces will be evaluated to determine core power symmetry. Criteria Level 2 In the TIP reproducibility test, the TIP traces should be reproducible within 3.5 percent relative error or 0.15 in. absolute error at each axial position, whichever is greater. TEST NUMBER 19 -- CORE PERFORMANCE Purpose The purpose of this test is to evaluate the core performance parameters of core flow rate, core thermal power level, maximum fuel rod surface heat flux and core minimum critical heat flux ration (MCHFR). Description Core power level, maximum heat flux, recirculation flow rate, hot channel coolant flow, MCHFR, fuel assembly power, and steam qualities will be determined at 13.5-18

BFN-22 existing power levels and assumed over power conditions. Plant and in-core instrumentation, conventional heat balance techniques, and core performance worksheets and nomograms will be used. This will be performed above 10 percent power and at various pumping conditions and can be done independent of the process computer functions. Criteria Level 1 Reactor power, maximum fuel surface heat flux, and MCHFR must satisfy the following limits: (1) Maximum fuel rod surface heat flux shall not exceed 134 W/cm (425,000 Btu/h-ft) during steady-state conditions when evaluated at the operating power level. (2) Minimum CHF ration shall not be less than 1.9 when evaluated at the operating power level. The basis for evaluation of MCHFR shall be "Design Basis for Critical Heat Flux Condition in BWRs," APED-5286, September 1966. (3) Steady-state reactor power shall be limited to values on or below the licensed flow control line (maximum power of 3293 MWt with flow of at least 102.5x10 lb/h). TEST NUMBER 20 -- ELECTRICAL OUTPUT AND PRELIMINARY HEAT RATE TEST Purpose The purpose of this test is to demonstrate that the guaranteed gross electrical output requirements are satisfied without exceeding the reactor power level warranty and to determine a preliminary net plant heat rate value. Description The plant gross electrical output will be measured during sustained operation at a load of at least 1,098.4 MWe. The gross electrical power will be measured at the 13.5-19

BFN-22 generator terminals and corrected to rated conditions of 0.90 power factor, turbine exhaust pressure of 2 inches mercury absolute, and rated generator hydrogen pressure. Data may be taken during this period for an ASME turbine cycle heat rate test as defined in PTC-6 "Interim Test Code for Steam Turbines Operating Predominately Within the Moisture Region with Nuclear Steam Supply." Criteria Level 1 The guaranteed performance calls for a gross output of 1,098,420.0 kWe at a reactor thermal power of 3293.0 MW and a net plant heat rate of 10,459 Btu/kWh. The power factor shall be 0.9 and the generator hydrogen pressure shall be 75.0 psig. TEST NUMBER 21 -- FLUX RESPONSE TO RODS Purpose The purpose of this test is to demonstrate stability in the power-reactivity feedback loop with increasing reactor power and determine the effect of control rod movement on reactor stability. Description Rod movement tests will be made at chosen power levels to demonstrate that the transient response of the reactor to a reactivity perturbation is stable for the full range of reactor power. A centrally located rod will be moved, and the neutron flux signal from a nearby LPRM chamber will be measured and evaluated to determine the dynamic effects of rod movement. Criteria Level 1 The decay ratio must be less than 1.0 for each process variable that exhibits oscillatory response to control rod movement. Level 2 The decay ratio is expected to be less than or equal to 0.25 for each process variable that exhibits oscillatory response to control rod movement when the plant is operating above the lower limit setting of the Master Flow Controller. 13.5-20

BFN-22 TEST NUMBER 22 -- PRESSURE REGULATOR Purpose The purposes of this test are to: (a) determine the reactor and pressure control system responses to pressure regulator setpoint changes, (b) demonstrate the stability of the reactivity-void feedback loop to pressure perturbations, (c) demonstrate the control characteristics of the bypass and control valves, (d) demonstrate the takeover capabilities of the backup pressure regulator, and (e) to optimize the pressure regulator settings to give the best combination of fast response and small overshoot. Description The pressure setpoint will be decreased rapidly and then increased rapidly in 0.5 psi steps and the response of the system will be measured. The backup regulator will be tested by increasing the operating pressure regulator setpoint rapidly until the backup regulator takes over control. The load reference setpoint will be reduced, and the test repeated with the bypass valve in control. The response of the system will be measured and evaluated and regulator settings will be optimized. Criteria Level 1 The decay ratio must be less than 1.0 for each process variable that exhibits oscillatory response to pressure regulator changes. Level 2 In all tests except the simulated failure of the operating pressure regulator, the decay ratio is expected to be less than or equal to 0.25 for each process variable that exhibits oscillatory response to pressure regulator changes when the plant is operating above the lower limit setting of the Master Flow Controller. During the simulated failure of the operating pressure regulator, the backup regulator is expected to control the transient such that the reactor does not scram. 13.5-21

BFN-22 TEST NUMBER 23 -- FEEDWATER SYSTEM Purpose The purposes of this test are to: (a) demonstrate acceptable reactor water level control, (b) evaluate and adjust feedwater controls, (c) demonstrate general reactor response to inlet subcooling changes on reactor power and pressure, (d) demonstrate individual feedwater pump response. Description Reactor water level setpoint changes of approximately 6 in. will be used to evaluate and acceptably adjust the feedwater control system settings for all power and feedwater control system settings for all power and feedwater pump modes. Operate each pump through its flow range to verify acceptable feedwater pump linearity. Response time on each feedwater pump will be verified by changing the flow by 10 percent and measuring the turbine speed and flow response times. Criteria Level 1 The decay ratio must be less than 1.0 for each process variable that exhibits oscillatory response to feedwater system changes. Level 2 The decay ratio is expected to be less than or equal to 0.25 for each process variable that exhibits oscillatory response to feedwater system setpoint changes when the plant is operating above the lower limit of the Master Flow Controller. System response for large transients should not be unexplainably worse than preanalysis. TEST NUMBER 24 -- BYPASS VALVES Purpose The purposes of this test are to: (a) demonstrate the ability of the pressure regulator to minimize the reactor pressure disturbance during an abrupt change in reactor steam flow, and (b) demonstrate that a bypass valve can be tested for proper functioning at rated power without causing a high flux scram. 13.5-22

BFN-22 Description One of the turbine bypass valves will be tripped open by a test switch. The pressure transient will be measured and evaluated to aid in making final adjustments to the pressure regulator. Criteria Level 1 The decay ratio must be less than 1.0 for each process variable that exhibits oscillatory response to bypass valve changes. Level 2 The decay ratio is expected to be less than or equal to 0.25 for each process variable that exhibits oscillatory response to bypass valve changes when the plant is operating above the lower limit setting of the Master Flow Controller. The maximum pressure decrease at the turbine inlet should be less than 50 psig to avoid approaching the low steamline pressure isolation or cause excessive water level swell in the reactor. TEST NUMBER 25 -- MAIN STEAM ISOLATION VALVES Purpose The purposes of this test are to: (a) functionally check the main steamline isolation valves (MSIV) for proper operation at selected power levels, (b) determine reactor transient behavior during and following simultaneous full closure of all MSIV and following closure of one valve, and (c) determine isolation valve closure time. Description Fast full closure of each MSIV will be performed at hot standby and selected power levels to determine the maximum power conditions at which individual valve full closure tests can be performed without a reactor scram. Functional checks (10 percent closure) of each isolation valve will be performed at selected power levels above the maximum power condition for individual MSIV full closure determined above. A test of simultaneous full closure of all MSIV's will be performed at about 100 percent of rated thermal power and proper operation of the relief valves and the RCIC will be shown. Reactor process variable will be monitored to determine the transient behavior of the system during and following each isolation test. MSIV delay and movement times will be determined. Proper seating of the MSIV's will be demonstrated. 13.5-23

BFN-22 Criteria Level 1 MSIV stroke time will be between 3 and 5 sec, exclusive of electrical delay time. Reactor pressure shall be maintained below 1230 psig (the setpoint of the first safety valve) during the transient following closure of all valves. During full closure of individual valves, scram should not occur. Level 2 The maximum reactor pressure should be about 1200 psig, 30 psi below the first safety valve setpoint following closure of all valves. This is a margin of safety for safety valve weeping. During full closure of individual valves, pressure must be 20 psi below scram, neutron flux must be 10 percent below scram, and steam flow in individual lines must be below the trip point. TEST NUMBER 26 -- RELIEF VALVES Purpose The purposes of this test are to: (a) verify the proper operation of the dual purpose relief safety valves, (b) determine their capacity, and (c) verify proper reseating following operation. Description The main steam relief valves will each be opened manually so that at any time only one is open. Capacity of each relief valve will be determined by the amount the bypass or control valves close to maintain reactor pressure. Proper reseating of each relief valve will be verified by observation of temperatures in the relief valve discharge piping. Criteria Level 1 The combined capacity of the relief valves will be demonstrated to be at least 61 percent (FSAR, Sections 14.5.1.2, 14.1.5.3, 14.1.5.4, and 14.1.5.6) of their design 13.5-24

BFN-22 capacity (11 x 800,000=8.8 x 10 lb/hr at 1,100 psig) considering the relief valve with the largest measured capacity to be inoperable. Level 2 Each relief valve is expected to have a capacity of at least 800,000 lb/h at a pressure setting of 1100 psig. Relief valve leakage must be low enough that the temperature measured by the thermocouples in the discharge side of the valves falls to within 10°F of the temperature recorded before the valve was opened. TEST NUMBER 27 -- TURBINE STOP AND CONTROL VALVE TRIPS Purpose The purposes of this test are to (a) determine the response of the reactor system to a turbine stop or control valve trip and (b) evaluate the response at the bypass, relief valve and reactor protection systems. The parametric responses of particular interest are the peak values and the rate of change of both reactor power and reactor steam dome pressure. Description The stop of control valves will be tripped closed at selected reactor power levels and neutron flux, feedwater flow and temperature, vessel water level and pressure will be monitored. Responses of selected control valves, stop valves, relief valves, and bypass valves will be recorded. Criteria Level 1 The safety valves should not open; therefore, reactor pressure should not rise above 1230 psig (the setpoint of the first safety valve). Reactor scram must limit the severity of the neutron flux and simulated heat flux transients within thermal limitations. The turbine stop valves must close before the control valves for the turbine stop valve trip. The turbine control valves must close before the stop valves for the turbine control valve fast closure. 13.5-25

BFN-22 Level 2 Reactor pressure should not rise so close to the safety valve setting that weeping or leakage occurs; therefore, the pressure should not rise above 1200 psig, which is 30 psi below the setpoint of the first safety valve. The measurement of simulated heat flux must not be significantly greater than preanalysis. Trip scram must meet Reactor Protection System specification. The pressure regulator must regain control before a low pressure reactor operation occurs. TEST NUMBER 29 -- FLOW CONTROL Purpose The purposes of this test are to: (a) determine the plant response to changes in the recirculation flow, (b) adjust all control elements, and (c) demonstrate the plant load following capability in all flow control modes (local manual, master manual, and automatic). Description Various process variables will be recorded while step changes are introduced into the recirculation flow control system (increased and decreased) at chosen points on the 50, 75, and 100 percent load lines. Up to 30 percent/min. change in recirculation flow will be made from all flow conditions down to the lower limit of the Master Flow Controller and return. Load following capability will be demonstrated in all flow control modes. Criteria Level 1 The decay ratio must be less than 1.0 for each process variable that exhibits oscillatory response to flow control changes. Level 2 The decay ratio is expected to be less than or equal to 0.25 for each process variable that exhibits oscillatory response to flow control changes when the plant is operating above the lower limit setting of the Master Flow Controller, and also at test point No. l. Scram must not occur. Automatic flow control range must be at least 80 to 100 percent power along the full power load line. Load response to a 20 percent load demand step must be at least 0.5 percent/sec. 13.5-26

BFN-22 TEST NUMBER 30 -- RECIRCULATION SYSTEM Purpose The purpose of this test is to investigate the performance of the recirculation system including transient responses and steady-state conditions following recirculation pump trips at selected power levels and calibration of the jet pump core flow measurement system. Description Two recirculation pumps will be tripped at power levels of 50, 75, and 100 percent of rated power. Single pump trips will be performed at 50 and 100 percent of rated power. The single pump trips will be initiated by opening the generator field breaker. Two pump trips will be initiated by tripping the M-G set drive motors. Reactor pressure, steam and feedwater flow, set pump P, and neutron flux will be recorded during the transient and at steady-state conditions. MCHFR evaluations will be made for conditions encountered during the transient. The jet pump instrumentation will be calibrated to read total core flow. Criteria Level 1 Not applicable Level 2 For each pump trip test, the minimum transient MCHFR (based on operating data divided by the minimum transient MCHFR evaluated from design values) is expected to be equal to or greater than 1.0. Flow instrumentation has been calibrated such that the Reactor Jet Pump Total Flow Recorder provides correct flow indication. TEST NUMBER 31 -- LOSS OF TURBINE-GENERATOR AND OFFSITE POWER Purpose The purpose of this test is to demonstrate proper performance of the reactor and the plant electrical equipment and systems during the loss of auxiliary power transient. 13.5-27

BFN-22 Description The loss of auxiliary power test will be performed at 25 percent of rated power. The proper response of reactor plant equipment, automatic switching equipment, and the proper sequencing of the diesel generator load will be checked. Appropriate reactor parameters will be recorded during the resultant transient. Criteria Level 1 All test pressure transients must have maximum pressure values below 1230 psig, which is the setpoint of the first safety valve. All safety systems, such as the Reactor Protection System, the diesel generator, RCIC and HPCI, must function properly without manual assistance. Level 2 Normal reactor cooling systems should be able to maintain adequate torus water temperature, adequate drywell cooling and prevent actuation of the automatic depressurization system. The maximum reactor pressure should be 30 psi below the first safety valve setpoint. This is a margin of safety for safety valve weeping. TEST NUMBER 32 -- RECIRCULATION M-G SET SPEED CONTROL Purpose The purposes of this test is: (a) to demonstrate that the recirculation speed control system can satisfactorily perform its function by comparing transient test results against system criteria. 13.5-28

BFN-22 Description Make several small step changes in speed at various pump speeds and record appropriate recirculation loop transient signals. Demonstrate performance over the full speed range with small speed demand step tests. Criteria Level 1 The decay ratio must be less than 1.0 for each process variable that exhibits oscillatory response to recirculation M-G set speed changes. Level 2 When the unit is operating above the lower limit setting of the master manual limiter, the decay ratio should be less than or equal to 0.25 for each process variable that exhibits oscillatory response to recirculation M-G set speed changes. TEST NUMBER 35 -- RECIRCULATION AND JET PUMP SYSTEM CALIBRATION Purpose The purpose of this test is to obtain a complete integrated calibration of the installed jet pump and recirculation system instrumentation with the reactor shutdown. Description This test involves applying, simultaneously to an entire loop, a closely controlled pressure to obtain an integrated calibration check of the system instrumentation. Actual calibration of the jet pump flow instrumentation will be completed during hot pressurized operation by comparison of the single and double tapped pressure drops as a function of flow. Criteria Not applicable. 13.5-29

BFN-22 TEST NUMBER 36 -- EQUALIZER OPEN Purpose The purposes of this test are (a) to explore the allowable operating range and performance of the recirculation system under conditions of one pump operation with the equalizer line valves open, and (b) to develop operating procedures for one pump equalizer open operation. Description The reactor recirculation system consists of the reactor vessel and two piping loops. Each loop contains a recirculation pump, suction and discharge isolation valves, and ten parallel jet pumps situated in the reactor downcomer. An equalizer line with two valves connects the loops. Under normal one and two pump operation the equalizer valves have been kept closed. This test will explore one pump operation with the equalizer open. Initial equalizer valve opening will be made at a high pump speed by rapid jogging until the inactive loop jet pumps go from reverse to forward flow. Successive valve openings and pump speed increases will be scheduled to avoid pump loop P, pump speed and pump motor current limits. When the valve is full open, or when limits are reached the available operating region will be explored and appropriate data will be obtained. The test will be concluded by rapidly closing the equalizer valve while recording the transient. Criteria Level 1 Operations shall be conducted so that the following conditions are met: (1) At pump flows greater than 115 percent of the design value, the pump loop shall at all times be greater than 71 psi. (2) At pump flows greater than 115 percent of the design value, operation with pump loop P's in the range of 71 to 80 psi shall be minimized, and shall not exceed 30 minutes total. Operating time with partially open equalizer under conditions of reverse jet pump flow shall be minimized. If the equalizer valve sticks under this condition the following action will be taken: (1) In the absence of vibration monitoring, the remaining equalizer valve shall be immediately closed or the pump tripped. (2) If vibration is being monitored, the vibration engineer will decide whether the situation warrants closure of the remaining equalizer valve or tripping of the pump. Pump motor and MG set temperatures and currents shall not be allowed to exceed vendor specifications. If these limits are reached, corrective actions include 13.5-30

BFN-22 reduction of pump speed to the minimum head limit, equalizer closure and pump trip. During steady-state operation, reactor power shall be kept below 68 percent of rated with one pump operating and closed equalizer, and below 80 percent with open equalizer. Level 2 Operation should be free of repeated running into limits. If control system instabilities preclude acceptable operation, the test should be terminated by closure of the equalizer or tripping of the pump. TEST NUMBER 39 -- WATER LEVEL VERIFICATION IN REACTOR VESSEL Purpose The purpose of this test is to verify the calibration of YARWAY and GEMAC level instrumentation under varying conditions. Description Reactor water level is monitored by four level instrument systems: YARWAY wide and narrow range, and, GEMAC wide and narrow range instruments. This test is divided into two parts. The first part involves measuring the YARWAY reference by temperature to verify it agrees with the temperature correction factor used in calibration. The second portion of the test verified the ability of the feedwater control system to regulate reactor water level at two power levels: 50 percent flow/50 percent power and 100 percent flow/100 percent power. Criteria Not applicable. TEST NUMBER 70 Purpose Demonstrate the operability of the reactor water cleanup system under actual reactor operating temperature and pressure. Description 13.5-31

BFN-22 This system provides a continuous purifying treatment of the reactor water by removing various impurities to maintain reactor water at specification quality. The system contains two recirculation pumps, regenerative heat exchangers, nonregenerative heat exchangers, and a cleanup demineralizer system. The modes of operation are normal, startup, blowdown, refueling, and hot standby. Criteria Level 1 Reactor water quality must be maintained according to specifications in fuel warranty documents. Level 2 The temperature at the tube outlet of the nonregenerative heat exchangers shall not reach 140°F (60°C) in any cleanup system operating mode. TEST NUMBER 71 -- RESIDUAL HEAT REMOVAL SYSTEM Purpose The purpose of this test is to demonstrate the ability of the Residual Heat Removal (RHR) system to: (a) remove residual and decay heat from the nuclear system so that refueling and nuclear system servicing can be performed, and (b) remove heat from the pressure suppression pool water. Description The RHR system is a closed loop system of piping, water pumps, and heat exchangers, the purpose of which is to remove post-power-operation energy from the reactor under both operational and accident conditions. 13.5-32

BFN-22 During a reactor cooldown, after operation at normal temperature and pressure, the ability of the Shutdown Cooling Subsystem to remove enough of the residual heat (decay and sensible heat) from the reactor primary system will be demonstrated. The reactor water will be cooled by pumping water from one of the recirculation loops, through the RHR system heat exchanger(s) and back to the reactor vessel by way of a (the) recirculation loop(s). It will also be possible to divert part of the flow to a spray nozzle in the reactor vessel head to condense steam generated from the hot walls of the vessel while it is being flooded, thereby maintaining saturated pressure and temperature conditions. The Suppression Pool Cooling subsystem cools the pressure suppression pool to limit the water temperature such that the temperature immediately after a blowdown does not exceed 170°F when reactor pressure is above 135 psig. During this mode of operation, water is pumped from the pressure suppression pool, through the RHR system heat exchangers and back to the pressure suppression pool. No Suppression Pool Cooling test is necessary if the heat exchanger capability is established by the Shutdown Cooling test, and the flow capability of the RHR system in the "Suppression Pool Cooling" mode has previously been established. Criteria Level 1 The heat removal capability of each RHR heat exchanger in the "Shutdown Cooling" mode or the "Suppression Pool Cooling" mode shall be 18.7 by 10 Btu/hr ( 4.69 by 10 Kcal/hr) or greater. TEST NUMBER 72 -- DRYWELL ATMOSPHERE COOLING SYSTEM Purpose The purpose of this test is to verify the ability of the Drywell Atmosphere Cooling System to maintain design conditions in the drywell during operating conditions. Description The Drywell Atmosphere Cooling System will be placed in operation and its ability to maintain the following temperature in the drywell with 8 of the 10 fans in operation will be demonstrated. 13.5-33

BFN-22 During Operation: Average temperature throughout drywell -- 135°F Maximum around the recirculating pump motors -- 128°F Maximum all other areas -- 150°F Maintain a uniform circumferential temperature of the refueling bellows/bulkhead assembly within 10°F Within 8 to 20 hours following shutdown, all areas in the drywell beneath the vessel-to-drywell bulkhead shall be within 10°F of Reactor Building Closed Cooling Water inlet temperature. Criteria Level 2 The heat removal capability of the drywell cooler shall be approximately 4.8 x 10 Btu/hr. The drywell cooling system shall have a standby capability of 25 percent of above. The drywell cooling system shall maintain temperatures in the drywell below the design valves given in the description during normal operation. TEST NUMBER 73 -- COOLING WATER SYSTEMS Purpose The purpose of this test is to verify that the performance of the Reactor Building Closed Cooling Water (RBCCW) and the raw cooling water systems is adequate with the reactor at rated condition. Description With the reactor at 1,000 psig (70 kg/cm) following the initial heatup, data will be obtained to verify that the flow rates in the RBCCW heat exchangers are adequate and properly balanced, and that the heat exchangers outlet temperatures are balanced and within design values. Flow rate adjustments will be made as necessary to achieve satisfactory system performance. The test will be repeated at selected power levels to verify continued satisfactory performance with higher plant heat loads. 13.5-34

BFN-22 Criteria Level 2 3.2.1 Verification that the system performance meets the cooling requirements constitutes satisfactory completion of this test. 3.2.2 The RBCCW was designed to transfer a maximum heat load of 31.3 x 10 Btu/hr (7.9 x 10 Kcal/hr) in order to limit equipment inlet water temperature to 100°F (38°C) assuming a service (raw cooling) water inlet temperature of 90°F (32°C). TEST NUMBER 90 -- VIBRATION MEASUREMENTS Purpose Determine the vibration characteristics of selected reactor internals and recirculation loops induced by cold recirculation flow and by hot, two-phase flows. Description Vibratory responses will be recorded at various recirculation flow rates at temperatures below 150°F using strain gages on in-core guide tubes, control rod stub tube, shroud support legs, and jet pump riser braces; accelerometers on the recirculation loops and displacement gages on the shroud, steam separator and jet pumps. Portable vibration sensor surveys will be made on the recirculation loops and differential pressure measurements will be made across the core plates, shroud head and shroud wall. At hot, two-phase flow conditions, similar measurements will be made on the in-core guide tubes, shroud, jet pump riser and shroud head. The results of vibration measurements made at other BWR installations will be considered in the final selection of components to be tested. Where possible, vibration measurements will be made as Preoperational Test. Criteria Level 1 The criteria by which the results of the vibration tests will be judged involve complex, precalculated relationships among spatial locations, vibrational amplitudes, and vibrational frequencies as related to stress and limited by ASME Code, Section III. 13.5-35

BFN-22 TEST NUMBER 92 -- STEAM SEPARATOR-DRYER Purpose Determine carryunder and carryover characteristics of the steam separator-dryer. Description Samples will be taken from the inlet and outlet of the steam dryers, and the inlet at the steamline at various power levels at chosen water levels and recirculation flow rates. The amount of carryunder will be estimated from these samples and carryover will be determined from Na-24 activities in samples taken from the outlet of the steam dryers. Criteria Level 2 Water carryover from the dryer shall be no greater than 0.002 weight fraction. The design value of steam carryunder to the jet pumps in 0.01 weight fraction. 13.5.2.3 Test Purpose, Description and Acceptance Criteria -- Units Two and Three TEST NUMBER 1 -- CHEMICAL AND RADIOCHEMICAL Purpose The principal objectives of this test are (a) to secure information on the chemistry and radiochemistry of the reactor coolant, and (b) to determine that the sampling equipment, procedures and analytic techniques are adequate to supply the data required to demonstrate that the chemistry of all parts of the entire reactor system meet specifications and process requirements. Specific objectives of the test program include evaluation of fuel performance, evaluations of demineralizer operations by direct and indirect methods, measurements of filter performance, confirmation of condenser integrity, demonstration of proper steam separator-dryer operation, measurement and calibration of the off-gas system, and calibration of certain process instrumentation. Data for these purposes is secured from a variety of sources: plant operating records, regular routine coolant analysis, radio-chemical measurements of specific nuclides, and special chemical tests. 13.5-36

BFN-22 Description Prior to fuel loading a complete set of chemical and radio-chemical samples will be taken to ensure that all sample stations are functioning properly and to determine initial concentrations. Subsequent to fuel loading during reactor heatup and at each major power level change, samples will be taken and measurements will be made to determine the chemical and radiochemical quality of reactor water and reactor feedwater, amount of radiolytic gas in the steam, gaseous activities leaving the air ejectors, decay times in the off-gas lines, and performance of filters and demineralizers. Calibrations will be made of monitors in the stack, liquid waste system and liquid process lines. Criteria Level 1 Chemical factors defined in the Technical Specifications must be maintained within the limits specified. The activity of gaseous and liquid effluents must conform to license limitations. Level 2 Water quality must be known at all time and should remain within the guidelines of the Water Quality Specifications. TEST NUMBER 2 -- RADIATION MEASUREMENTS Purpose The purposes of this test are (a) to determine the background radiation levels in the plant environs prior to operation for base data on activity buildup and (b) to monitor radiation at selected power levels to assure the protection of personnel during plant operation. Description A survey of natural background radiation throughout the plant site will be made prior to fuel loading. Subsequent to fuel loading, during reactor heatup and at power levels of 25 percent, 50 percent, 75 percent, and 100 percent of rated power, gamma radiation level measurements and where appropriate, thermal and fast neutron dose rate measurements will be made at significant locations throughout the plant. All potentially high radiation areas will be surveyed. 13.5-37

BFN-22 Criteria Level 1 The radiation doses of plant origin and the occupancy times of personnel in radiation zones shall be controlled consistent with the guidelines of the standards for protection against radiation outlined in 10 CFR 20 AEC General Design Criteria. TEST NUMBER 3 -- FUEL LOADING Purpose The purpose of this test is to load fuel safely and efficiently to the full core size. Description Prior to fuel loading, control rods and neutron sources and detectors will be installed and tested. Fuel loading will commence with the loading of four fuel assemblies around the central neutron source. Fuel loading will be accomplished by loading complete control coils that sequentially complete each face of an ever-increasing square core loading in a counterclockwise direction. Control rod drive functional tests are performed during the last week before fuel loading. Criteria Level 1 The partially loaded core must be subcritical by at least 0.38 percent k/k with the analytically strongest rod fully withdrawn. TEST NUMBER 4 -- FULL CORE SHUTDOWN MARGIN Purpose The purpose of this test is to demonstrate that the reactor will be subcritical throughout the first full cycle with any single control rod fully withdrawn. 13.5-38

BFN-22 Description This test will be performed in the fully loaded core at ambient temperature in the xenon-free condition. The shutdown margin will be measured by withdrawing the analytically strongest rod or the equivalent (another rod plus an added reactivity) and one or more additional rods which have been calibrated by calculation until criticality is reached. Criteria Level 1 The shutdown margin of the fully loaded core with the analytically strongest rod withdrawn must be at least 0.38 percent k/k (plus an additional margin for exposure to be determined later). Level 2 Criticality should occur within 1.0 percent k/k of the predicted rod configuration. TEST NUMBER 5 -- CONTROL ROD DRIVE SYSTEM Purpose The purposes of the Control Rod Drive System test are (a) to demonstrate that the Control Rod Drive (CRD) System operates properly over the full range of primary coolant temperatures and pressures from ambient to operating, and (b) to determine the initial operating characteristics of the entire CRD system. Description The CRD tests performed during Phases II through IV of the startup test program are designed as an extension of the tests performed during the preoperational CRD system tests. Thus, after it is verified that all control rod drives operate properly when installed, they are tested periodically during heatup to assure that there is no significant binding caused by thermal expansion of the core components. A list of all control rod drive tests to be performed during startup testing is given below. 13.5-39

BFN-22 CONTROL ROD DRIVE SYSTEM TESTS Reactor Pressure with Core Loaded Test Accumulator Preop psig (kg/cm) Description Pressure Tests 0 600 800 Rated (42.2) (58.2) Position Indication all all Normal Times Insert/ Withdrawn all all 4* Coupling all all*** Friction all 4* Scram Normal all all 4* 4* all Scram Minimum 4* Scram Zero 4* Scram (Scram Normal 4(Full 4* Discharge core Volume scram) High Level) Scram Normal 4** 

  *Value refers to the four slowest CRDs as determined from the normal accumulator pressure scram test at ambient reactor pressure.

Throughout the procedure, "the four slowest CRDs" implies the four slowest compatible with rod worth minimizer and CRD sequence requirements. 

  **Scram times of the four slowest CRDs will be determined at 25 percent and 100 percent of rated power during planned reactor scrams.
  ***Establish initially that this check is normal operating procedures.

NOTE: Single CRD scrams should be performed with the charging valve closed (do not ride the charging pump head). 13.5-40

BFN-22 Criteria Level 1 Each CRD must have a normal withdraw speed less than or equal to 3.6 inches per second (9.14 cm/sec), indicated by a full 12-foot stroke in greater than or equal to 40 seconds. The mean scram time of all operable CRDs must not exceed the following times: (Scram time is measured from the time the pilot scram valve solenoids are deenergized.) Scram Time Scram Time (Seconds) (Seconds) Vessel Dome Vessel Dome Pressure Pressure Percent 950 psig 950 psig Inserted (66.9 kg/cm) (66.9 kg/sm) 5 0.375 0.475 20 0.90 1.100 50 2.0 2.0 90 3.5 3.5 The mean scram time of the three fastest CRD's in a two-by-two array must not exceed the following times: (Scram time is measured from the time the pilot scram valve solenoids are deenergized.) Scram Time Scram Time (Seconds) (Seconds) Vessel Dome Vessel Dome Pressure Pressure Percent 950 psig 950 psig Inserted (66.9 kg/cm) (66.9 kg/sm) 5 0.398 0.504 20 0.954 1.166 50 2.120 2.120 90 3.800 3.800 13.5-41

BFN-22 Level 2 Each CRD must have a normal insert or withdrawn speed of 3.0 +/- 0.6 inches per second (7.62 sma,b 1.52 cm/sec), indicated by a full 12-foot stroke in 40 to 60 seconds. With respect to the control rod drive friction tests, if the differential pressure variation exceeds 15 psid (1 kg/cm) for a continuous drive in, a settling test must be performed, in which case, the differential setting pressure should not be less than 30 psid (2.1 kg/cm) nor should it vary by more than 10 psid (0.7 kg/cm) over a full stroke. Scram times with normal accumulator charge should fall within the time limits indicated on Figure 5.3-1 of the Startup Test Instructions. TEST NUMBER 6 -- SRM PERFORMANCE AND CONTROL ROD SEQUENCE Purpose The purpose of this test is to demonstrate that the operational sources, SRM instrumentation, and rod withdrawal sequences provide adequate information to achieve criticality and increase power in a safe and efficient manner. The effect of typical rod movements on reactor power will be determined. Description The operational neutron sources will be installed and source range monitor count-rate data will be taken during rod withdrawals to critical and compared with stated criteria on signal and signal count-to-noise count ratio. A withdrawal sequence has been calculated which completely specifies control rod withdrawals from the all-rods-in condition to the rated power configuration. Critical rod patterns will be recorded periodically as the reactor is heated to rated temperature. Movement of rods in a prescribed sequence is monitored by the Rod Worth Minimizer and the Rod Sequence Control System, which will prevent out-of-sequence withdrawal and insertions. As the withdrawal of each rod group is completed during the power ascension, the electrical power, steam flow, control valve position, and APRM response will be recorded. 13.5-42

BFN-22 Data will be obtained to verify the relationship between core power and first stage turbine pressure to ensure that the RSCS properly fulfills its intended function up to the required power level. Criteria There must be a neutron signal count-to-noise count ratio of at least 2 to 1 on the required operable SRM's or Fuel Loading Chambers. There must be a minimum count rate of 3 counts/second on the required operable SRM's or Fuel Loading Chambers. The IRM's must be on scale before the SRM's exceed the rod block set point. The Rod Sequence Control System shall be operable as specified in the Technical Specification. TEST NUMBER 9 -- WATER LEVEL MEASUREMENT Purpose To verify the calibration and agreement of the GEMAC and YARWAY water level instrumentation under various conditions. Description The test is divided into two parts. The first part will be done at rated temperature and pressure and steady-state conditions and will verify that the reference leg temperature of the YARWAY instrument is the value assumed during initial calibration. If not, the instrument will be recalibrated using the measured value. After the (re)calibration, the GEMAC and YARWAY indications should be in reasonable agreement. The second part of the test consists of determining the agreement of the water level instrumentation at two core flow rates and various heights. Criteria Not applicable. TEST NUMBER 10 -- IRM PERFORMANCE Purpose The purpose of this test is to adjust the Intermediate Range Monitor System to obtain an optimum overlap with the SRM and APRM systems. 13.5-43

BFN-22 Description Initially the IRM system is set to maximum gain. After the APRM calibration, the IRM gains will be adjusted to optimize the IRM overlap with the SRM's and APRM's. Criteria Level 1 Each IRM channel must be adjusted to that overlap with the SRM's and APRM's is assured. The IRM's must produce a scram at 96 percent of full scale. TEST NUMBER 11 -- LPRM CALIBRATION Purpose The purpose of this test is to calibrate the Local Power Range Monitoring System. Description The LPRM channels will be calibrated to make the LPRM readings proportional to the neutron flux in the narrow-narrow water gap at the chamber elevation. Calibration factors will be obtained through the use of either an off-line or a process computer calculation that relates the LPRM reading to average fuel assembly power at the chamber height. Criteria Level 1 The meter readings of each LPRM chamber will be proportional to the neutron flux in the narrow-narrow water gap at the height of the chamber. TEST NUMBER 12 -- APRM CALIBRATION Purpose The purpose of this test is to calibrate the Average Power Range Monitor System. Description A heat balance will generally be made each shift and after each major power level change. Each APRM channel reading will be adjusted to be consistent with the core thermal power as determined from the heat balance. During heatup a preliminary 13.5-44

BFN-22 calibration will be made by adjusting the APRM amplified gains so that the APRM readings agree with the results of a constant heatup rate heat balance. The APRM's should be recalibrated in the power range by a heat balance as soon as adequate feedwater indication is available. Criteria Level 1 The APRM channels must be calibrated to read equal to or greater than the actual core thermal power. Technical Specification and Fuel Warranty Limits on APRM scram and Rod Block shall not be exceeded. In the startup mode, all APRM channels must produce a scram at less than or equal to 15 percent of rated thermal power. Recalibration of the APRM system will not be necessary from safety considerations if at least two APRM channels per RPS trip circuit have readings greater than or equal to core power. Level 2 If the above criteria are satisfied then the APRM channels will be considered to be reading accurately if they do not read greater than the actual core thermal power by more than 7 percent of rated power. TEST NUMBER 13 -- PROCESS COMPUTER Purpose The purpose of this test is to verify the performance of the process computer under plant operating conditions. Description GE/PAC Computer system program verifications and calculational program validations at static and at simulated dynamic input conditions will be preoperationally tested at the computer supplier's site and following delivery to the plant site. Following fuel loading, during plant heatup and the accession to rated power, the nuclear steam supply system and the balance-of-plant system process variables sensed by the computer as digital or analog signals will become available. Verify that the computer is receiving correct values of sensed process variables and that the results of performance calculations of the nuclear steam supply system and 13.5-45

BFN-22 the balance-of-plant are correct. At steady-state power conditions the Dynamic System Test Case will be performed. Criteria Level 1 Not applicable. Level 2 Programs OD-1 P1 and OD-6 will be considered operational when:

1. The MCPR calculated by BUCLE and the process computer either:
a. Are in the same fuel assembly and do not differ in value by more than 2 percent, or
b. For the case in which the MCPR calculated by the process computer is in a different assembly than that calculated by BUCLE, for each assembly the MCPR and CPR calculated by the two methods shall agree within 2 percent.
2. The maximum LHGR calculated by BUCLE and the process computer either:
a. Are in the same fuel assembly and do not differ in value by more than 2 percent, or
b. For the case in which the maximum LHGR calculated by the process computer is in a different assembly than that calculated by BUCLE, for each assembly the maximum LHGR and LHGR calculated by the two methods shall agree within 2 percent.
3. The MAPLHGR calculated by BUCLE and the process computer either:
a. Are in the same fuel assembly and do not differ in value by more than 2 percent, or
b. For the case in which the MAPLHGR calculated by the process computer is in a different assembly than that calculated by BUCLE, for each assembly the MAPLHGR and APLHGR calculated by the two methods shall agree within 2 percent.

13.5-46

BFN-22 TEST NUMBER 14 -- RCIC SYSTEM Purpose The purpose of this test is to verify the proper operation of the Reactor Core Isolation Cooling (RCIC) system over its expected operating pressure range. Description The RCIC system test consists of two parts: injection to the condensate storage tank and injection to the reactor vessel. The CST injections consist of controlled and quick starts at reactor pressures ranging from 150 psig (10.5 kg/cm) to rated, with corresponding pump discharge pressures throttled between 250 psig (17.6 kg/cm) and 1220 psig (85.8 kg/cm). During this part of the testing, proper operation of the system will be verified and adjustments made as required to meet this criteria. The reactor vessel injection will consist of a cold quick start of the system with all flow routed to the reactor vessel at 25 percent power. 13.5-47

BFN-22 Criteria Level 1 The time from actuating signal to required flow must be less than 30 seconds at any reactor pressure between 150 psig (10.5 kg/cm) and rated. With pump discharge at any pressure between 150 psig (10.5 kg/cm) and 1220 psig (85.8 kg/cm), the required flow is 600 grm. (The limit of 1220 psig includes a conservatively high value of 100 psi for line losses. The measured value may be used if available. The RCIC turbine must not trip off during startup. Level 2 The turbine gland seal condenser system shall be capable of preventing steam leakage to the atmosphere. The delta P switch for the RCIC steam supply line high flow isolation trip shall be adjusted to actuate at 300 percent of the maximum required steady state flow. TEST NUMBER 15 -- HPCI SYSTEM Purpose The purpose of this test is to verify the proper operation of the High Pressure Coolant Injection (HPCI) System over its expected operating pressure range. Description The HPCI system test consists of two parts: injection to the condensate storage tank and injection to the reactor vessel. The CST injections consist of controlled and quick starts at three reactor pressures ranging from 150 psig (10.5 kg/cm) to rated, with corresponding pump discharge pressures throttled between 250 psig (17.6 kg/cm) and 1220 psig (85.8 kg/cm). During this part of the testing, proper operation of the system will be verified and adjustments made as required to meet the criteria. The reactor vessel injection will consist of a cold quick start of the system with all flow routed to the reactor vessel at 50 percent power. Criteria Level 1 The time from actuating signal to required flow must be less than 25 seconds at any reactor pressure between 150 psig (10.5 kg/cm) and rated. 13.5-48

BFN-22 With pump discharge at any pressure between 150 psig (10.5 kg/cm) and 1220 psig (85.8 kg/cm) the flow should be at least 5000 gpm. (The limit of 1220 psig includes a conservatively high value of 100 psi for line losses. The measured value may be used if available.) The HPCI turbine must not trip off during startup. Level 2 The turbine gland seal condenser system shall be capable of preventing steam leakage to the atmosphere. The delta P switch for the HPCI steam supply line high flow isolation trip shall be adjusted to actuate at 225 percent of the maximum required steady state flow. TEST NUMBER 16 -- SELECTED PROCESS TEMPERATURES Purpose The purposes of this procedure are to establish the proper setting of the low speed limiter for the recirculation pumps and to provide assurance that the measured bottom head drain temperature corresponds to bottom head coolant temperature during normal operations. Description During initial heatup while at hot standby conditions, the bottom drain line temperature and applicable reactor parameters are monitored as the recirculation pump speed is slowly lowered from 30 percent of maximum pump speed to either minimum stable speed or 20 percent of maximum pump speed, whichever is the greater. The parameters above are recorded during pump trips as well. Utilizing this data the first purpose as stated above is satisfied. The second purpose is satisfied by comparing recirculation loop coolant temperature with bottom drain line temperature when core flow is 100 percent. Criteria Level 1 The reactor recirculation pump shall not be operated unless the coolant temperatures between the upper and lower regions of the reactor vessel are within 145°F (80°C). 13.5-49

BFN-22 Level 2 The bottom head coolant temperature as measured by the bottom drain line thermocouple should be within 50°F (28°C) of reactor coolant saturation temperature. TEST NUMBER 17 -- SYSTEM EXPANSION Purpose The purpose of this test is to verify that the reactor drywell piping system is free and unrestrained in regard to thermal expansion and that suspension components are functioning in the specified manner. The test also provides data for calculation of stress levels in nozzles and weldments. Description Observe and record the horizontal and vertical positions of major equipment and piping in the Nuclear Steam Supply System and auxiliary systems to ensure components are free to move as designed. Adjust as necessary for freedom of movement. Criteria Level 1 There shall be no evidence of blocking of the displacement of any system component caused by thermal expansion of the system. Hangers shall not be bottomed out or have the spring fully stretched. The shock suppressor pistons must be within their operating range. Electrical cables shall not be fully stretched. Level 2 At the steady-state condition the displacements of instrumented points with displacement measuring devices shall not vary from the calculated values by more than 50 percent or 0.5 inch ( 1.27 cm), whichever is smaller. Displacements of less than 0.25 inch (0.64 cm) can be neglected since 50 percent of this value is bordering on the accuracy of measurement. If measured displacements do not meet these criteria, the piping design engineer must be contacted to analyze the data with regard to design stresses. 13.5-50

BFN-22 During the heatup cycle the trace of the instrumented points shall fall within a range of 150 percent of the calculated value from the initial cold position in the direction of the calculated value, and 50 percent of the calculated value from the initial position in the opposite direction of the calculated value. Hangers will be in their operation range (between the hot and cold settings). TEST NUMBER 18 -- CORE POWER DISTRIBUTION Purpose The purposes of this test are to (a) confirm the reproducibility of the TIP system readings, (b) determine the core power distribution in three dimensions, and (c) determine core power symmetry. Description Core power distribution data will be obtained during the power ascension program. Axial power traces will be obtained at each TIP location at the 50-percent power level. At least one and possibly more sets of TIP data will be taken above the 75-percent power level. These data sets will be used to determine overall TIP uncertainty including random noise and geometrical uncertainties. TIP data taken and used in this test will be submitted to NRC with the summary test report. TIP data will be obtained at the power levels discussed above with the reactor operating with a symmetric rod pattern and at steady-state conditions. The total TIP uncertainty for the test will be calculated by averaging the total TIP uncertainty determined from each set of TIP data taken. The total TIP uncertainty is made up of random noise and geometric components. The random noise uncertainty is obtained from four traces from each of the five TIP machines. The standard deviation due to random noise is calculated from the individual deviations of nodal power at each nodal level 5 through 22. The total random noise deviation is the average of the standard deviations for each nodal 13.5-51

BFN-22 level. The geometric uncertainty is determined for each complete set by dividing, for each symmetric TIP pair, the nodal BASE value in the upper left half of the core by its counterpart in the lower right half. The average and standard deviations of these ratios are calculated. The geometric standard deviation is calculated by first dividing the deviation of the ratios by 2 and then statistically subtracting the random noise uncertainty. The total TIP uncertainty is calculated by statistically adding the random noise and geometrical uncertainties. Because the TIP/LPRM instrument tubes are arranged symmetrically about the core diagonal, only half-core symmetry about this diagonal can be measured. A meaningful level criteria cannot be applied to this measurement. Any assymetry, as measured by the TIP system, will be accounted for on the calculation of MCPR using the GETAB method (or later the GEXL as it is incorporated). The thermal limits of the four assemblies surrounding an LPRM string are calculated based on the LPRM/TIP data from that string, independent of the data from its symmetric counterpart. This TIP/LPRM data is also reflected to its three pseudo locations with the appropriate corrections and uncertainties applied for the differences in bundle configurations and fuel types. Criteria Level 1 The total TIP uncertainty (including random noise and geometrical uncertainties) shall be less than 7.8 percent. This total TIP uncertainty will be obtained by averaging the total uncertainty for all data sets obtained. A minimum of two data is sufficient for the determination of total TIP uncertainty. However, if the first two data sets do not meet the above criteria, testing may be continued and up to six data sets obtained and compared with the criteria. If the 7.8 percent total TIP uncertainty criteria has not been met by the six sets of data, testing may continue and additional data sets may be obtained provided (a) the MCPR limit is adjusted to reflect the TIP uncertainty determined by the six data sets, (b) the NRC is informed of the adjusted MCPR limit, (c) the data generated from the six sets of data is transmitted to the NRC, and (d) TVA's intentions for continuing to test and expand the data base are provided to NRC. If the total TIP uncertainty is reduced by taking additional sets of data to expand the data base, the MCPR limit will be adjusted accordingly until the 7.8 percent total TIP uncertainty is met. At this time, the MCPR limit will be returned to its original value. Level 2 Not applicable 13.5-52

BFN-22 TEST NUMBER 19 -- CORE PERFORMANCE (Unit 2 Only) Purpose The purpose of this test is to evaluate the core performance parameters of core flow rate, core thermal power level, maximum fuel rod surface heat flux, core minimum critical heat flux ratio (MCHFR), Minimum Bundle Power Ratio (MBPR) and Maximum Average Planar Linear Heat Generation Rate (MAPLHGR). Description Core power level, maximum heat flux, recirculation flow rate, hot channel coolant flow, minimum critical heat flux ratio, fuel assembly power and steam qualities will be determined at existing power levels. Plant and Incore instrumentation, conventional heat balance techniques and core performance worksheets and nomograms will be used. This will be performed above 10 percent power and at various pumping conditions and can be done independent of the process computer functions. Criteria Level 1 The maximum fuel rod heat flux during steady-state conditions shall not exceed the design allowable heat flux of 135 2/cm. The Minimum Bundle Power Ratio (MBPR) shall be maintained greater than or equal to 1.0. The Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) shall not exceed the limits of the Plant Technical Specification. MCHFR shall be maintained at or above the flow dependent Minimum Fuel Warranty MCHFR Limit (Line "B," Figure 19.3-2, of the Startup Test Instructions). Steady-state reactor power shall be limited to 3293 mWt and values on or below the design flow control line (defined at 3440 MWt with core flow of at least 102.5 x 10 lb/hr). TEST NUMBER 19 -- CORE PERFORMANCE (Brown Ferry Unit 3 Only) Purpose The purposes of this test are (a) to evaluate the core thermal power and (b) to evaluate the following core performance parameters: (1) maximum linear heat 13.5-53

BFN-22 generation rate (MLHGR), (2) minimum critical power ratio (MCPR), and (3) maximum average planar heat generation rate (MAPLHGR). Description The core performance evaluation is intended to determine the principal thermal and hydraulic parameters associated with core behavior. These parameters are core flow, core thermal power, MLHGR, MCPR, and MAPLHGR. These core parameters will be evaluated by manual calculations, the process computer, or the off-line computer program BUCLE. If the process computer is used as a primary means to obtain these parameters, it must be proved that it agrees with BUCLE within 2 percent on all thermal parameters (see test number 13) or the results must be corrected to do so. If the BUCLE and process computer results do not agree within 2 percent for any thermal parameter, the parameter calculated by the process computer will be corrected by a multiplication factor to bring it within the 2-percent criteria. Criteria Level 1 The maximum linear heat generation rate (LHGR) of any rod during steady-state conditions shall not exceed the limit specified by the technical specifications. Steady-state reactor power shall be limited to 3,293 MWt and values on or below the design flow control line (defined as 3,440 MWt with core flow of at least 102.5 x 10 lb/hr). The minimum critical power ratio (MCPR) shall not exceed the limits specified by the technical specifications. The maximum average planar linear heat generation rate (MAPLHGR) shall not exceed the limits of the technical specifications. TEST NUMBER 20 -- ELECTRICAL OUTPUT AND HEAT RATE (Browns Ferry Unit 2 Only) Purpose The purpose of this test is to demonstrate that the plant net electrical output and net heat rate requirements are satisfied. 13.5-54

BFN-22 Description The plant gross electrical output and net heat rate will be measured during sustained operation at rated conditions. The gross electrical output is defined as the gross electrical output measured at the generator terminals and must be maintained for 300 hours. The net plant heat rate is defined as the thermal output from the reactor less the thermal content in the feedwater supplied to the reactor all divided by the net electrical output. All corrections for losses and auxiliary loads will be agreed to prior to the start of the test. The 2-hour net plant heat rate test would normally be done concurrently with the net plant electrical output test although this is not necessary. Criteria Level 1 The guaranteed performance calls for a gross output of 1,098,420 kWe at a reactor thermal power of 3293.0 MW and a net plant heat rate of 10,359 Btu/kWh. The power factor shall be 0.9 and the generator hydrogen pressure shall be 75.0 psig. TEST NUMBER 20 -- STEAM PRODUCTION (Browns Ferry Unit 3 Only) Purpose The purpose of this test is to demonstrate that the Nuclear Steam Supply System is providing steam sufficient to satisfy all appropriate warranties. Description Operate continuously for 300 hours at rated reactor conditions. When it is determined that all plant conditions are stabilized, the steam production rate will be measured during two 2-hour periods at conditions prescribed in the Nuclear Steam Generating System warranty. Criteria Level 1 The NSSS parameters as determined by using normal operating procedures shall be within the appropriate license restrictions. The Nuclear Steam Supply System must produce 13,422,000 lbs/hr of steam of not less than 99.7 percent of quality and 985 psia pressure at the second isolation valve. 13.5-55

BFN-22 This output is contingent upon the feedwater flow being 13,372,000 lbs/hr at 378°F, and CRD flow being 50,000 lbs/hr at 80°F. TEST NUMBER 21 -- FLUX RESPONSE TO RODS Purpose The purpose of this test is to demonstrate the stability of the core local power reactivity feedback mechanism with regard to small perturbations in reactivity caused by rod movement. Description Rod movement tests will be made at chosen power levels to prove that the transient response of the reactor to a reactivity perturbation is sufficiently stable over the full range of reactor power and flow conditions. The signal from a nearby LPRM will be recorded and evaluated to determine the local core dynamic effects of the rod movement. The control rod chosen should be strong enough to produce a 5 percent local power change and should be near the channel with most limiting thermal conditions. Criteria Level 1 The decay ratio must be less than 1.0 for each process variable that exhibits oscillatory response to control rod movement. Level 2 The decay ratio is expected to be less than or equal to 0.25 for each process variable that exhibits oscillatory response to control rod movement when the plant is operating above the lower limit setting of the Master Flow Controller. TEST NUMBER 22 -- PRESSURE REGULATOR Purpose The purposes of this test are to (a) determine the optimum settings for the pressure control loop by analysis of the transients induced in the reactor pressure control system by means of the pressure regulators, (b) to demonstrate the takeover capability of the backup pressure regulator upon failure of the controlling pressure regulator and to set spacing between the set points at an appropriate value, and (c) 13.5-56

BFN-22 to demonstrate smooth pressure control transition between the control valves and bypass valves when reactor steam generation exceeds steam used by the turbine. Description The pressure set point will be decreased rapidly and then increased rapidly by about 10 psi (0.7 kg/cm) and the response of the system will be measured in each case. It is desirable to accomplish the set point change in less than 1 second. At applicable test conditions the load reference set point will be set so that the transient is handled by control valves, bypass valves and both. The backup regulator will be tested by simulating a failure of the operating pressure regulator so that the backup regulator takes over control. The response of the system will be measured and evaluated and regulator settings will be optimized. Criteria Level 1 The decay ratio must be less than 1.0 for each process variable that exhibits oscillatory response to pressure regulator changes. Level 2 In all tests except the simulated failure of the operating pressure regulator, the decay ratio is expected to be less than or equal to 0.25 for each process variable that exhibits oscillatory response to pressure regulator changes when the plant is operating above the lower limit setting of the Master Flow Controller. Pressure control, deadband, delay, etc., if any shall produce variations in steam flow to the turbine no larger than the values of rated flow specified in the following table, as measured by gross generated electrical power. Percent of Full Power Percent of Rated Flow 90 - 100 0.5 70 - 90 1.5 to 0.5 70 and below 1.5 Optimum gain values for the pressure control loop shall be determined to give the fastest return from the transient conditions to the steady-state condition within the limits of the above criteria. During the simulated failure of the controlling pressure regulator, if the set point of the backup pressure regulator is optimumly set, the backup regulator shall control the transient such that the reactor does not scram. Following a 10 psi (0.7 kg/cm) pressure set-point adjustment, the time between the set-point change and the occurrence of the pressure peak shall be 10 seconds or less. 13.5-57

BFN-22 TEST NUMBER 23 -- FEEDWATER SYSTEM Purpose The purposes of this test are (a) to adjust the feedwater control system for acceptable reactor water level control, (b) to demonstrate stable reactor response to subcooling changes, (c) to demonstrate the capability of the automatic core flow runback feature to prevent low water level scram following the trip of one feedwater pump. Description Reactor water level setpoint changes of approximately 3 to 5 inches (7.5 to 12.5 cm) will be used to evaluate and adjust the feedwater control system settings for all power and feedwater pump modes. The level set-point changes will also demonstrate core stability to subcooling changes. One of the three operating feedwater pumps will be tripped and the automatic flow runback circuit will act to drop power to within the capacity of the remaining pump. Criteria Level 1 The decay ratio must be less than 1.0 for each process variable that exhibits oscillatory response to feedwater system changes. Level 2 The decay ratio is expected to be less than or equal to 0.25 for each process variable that exhibits oscillatory response to feedwater system changes when the plant is operating above the lower limit of the Master Flow Controller. Following a 3-inch level set-point adjustment in three-element control, the time from the set-point step change until the water level peak occurs shall be less than 35 seconds without excessive feedwater swings (changes in feedwater flow greater than 25 percent of rated flow). The automatic core flow runback feature will prevent a scram from low water level following a trip of one of the operating feedwater pumps. With the condensate system operating normally, the maximum turbine speed limit shall prevent pump damage due to cavitation. 13.5-58

BFN-22 TEST NUMBER 24 -- BYPASS VALVES Purpose The purposes of this test are (a) to demonstrate the ability of the pressure regulator to minimize the reactor pressure disturbance during an abrupt change in reactor steam flow and (b) to demonstrate that a bypass valve can be tested for proper functioning at rated power without causing a high flux scram. Description One of the turbine bypass valves will be tripped open and closed (after the opening disturbance has settled out) by a test switch; an opening time and closing time of less than 3 seconds is desirable. The pressure transient will be measured and evaluated to aid in making final adjustments to the pressure regulator. Criteria Level 1 The decay ratio must be less than 1.0 for each process variable that exhibits oscillatory response to bypass valve changes. Level 2 The decay ratio is expected to be less than or equal to 0.25 for each process variable that exhibits oscillatory response to bypass valve changes when the plant is operating above the lower limit setting of the Master Flow Controller. 13.5-59

BFN-22 The maximum pressure decrease at the turbine inlet during valve opening shall not exceed 50 psi to avoid approaching low steam line pressure isolation. The regulator shall limit the pressure disturbance during valve reclosure so that a margin of at least 5 percent shall be maintained below flux scram. Steam pressure should reach a steady state within 25 seconds after a bypass valve has been opened or closed. TEST NUMBER 25 -- MAIN STEAM ISOLATION VALVES Purpose The purposes of this test are (a) to functionally check the main steam line isolation valves (MSIVs) for proper operation at selected power levels, (b) to determine reactor transient behavior during and following simultaneous full closure of all MSIV's and following full closure of one valve, (c) to determine isolation valve closure time, and (d) to determine maximum power at which a single valve closure can be made without scram. Description During hot standby at rated pressure, both slow and fast single valve closure will be performed. A test of the simultaneous full closure of all MSIVs will be performed at about 100 percent of rated thermal power. Correct performance of the RCIC and relief valves will be shown. Reactor process variables will be monitored to determine the transient behavior of the system during and following the Main Steam Line (MSL) isolation. The maximum power conditions at which individual valve full closures tests can be performed without a reactor scram is to be established, and one individual valve full closure test will be performed on the 100 percent power load line to check ability to perform surveillance tests on this load line. Criteria MSIV closure time must be greater than 3 and less than 5 seconds. The initial transient rise in vessel dome pressure occurring within 20 seconds of the main steam isolation valve trip initiation shall not be greater than 150 psi, and the transient rise in simulated heat flux shall not exceed 10 percent. Level 2 The initial transient peak in vessel dome pressure occurring within 20 seconds following initiation of the MSIV closure and the transient peak in simulated surface 13.5-60

BFN-22 heat flux shall not be more limiting than the predicted transients in the Transient Analysis Design Report (100 psi and no heat flux increase). TEST NUMBER 26 -- RELIEF VALVES (Browns Ferry Unit 2 Only) Purpose The purposes of this test are (a) to verify the proper operation of the primary system relief valves, (b) to determine their capacity and response characteristics and (c) to verify their proper seating following operation. Description The main steam relief valves will each be opened using the "manual" control mode so that at any time only one is open. During heatup at 250 psig (17.5 kg/cm), each valve will be opened and closed to demonstrate proper functioning. Capacity of each relief valve will be determined at rated pressure by the amount of bypass or control valve closure required to maintain reactor pressure. Proper reseating of each relief valve will be verified by observation of temperatures in the relief valve discharge piping. At selected test conditions each valve will be manually actuated, and at least one valve will be timed. Additional timing data will be obtained in conjunction with those transient tests which result in automatic relief valve opening. Criteria Level 1 Each relief valve shall have a capacity of at least 800,000 lb/hr at an inlet pressure of 1112 psig. Level 2 Relief valve leakage shall be low enough that the temperature measured by the thermocouples in the discharge side of the valves returns to within 10°F (5.6°C) of the temperature recorded before the valve was opened. TEST NUMBER 26 -- RELIEF VALVES (Browns Ferry Unit 3 Only) Purpose The purposes of this test are (a) to verify the proper operation of the primary system relief valves, (b) to determine their capacity and response characteristics, and (c) to verify their proper seating following operation. 13.5-61

BFN-22 Description The main steam relief valves will each be opened using the "manual" control mode so that at any time only one is open. During heatup at 250 psig (17.5 kg/cm), each valve will be opened and closed to demonstrate proper functioning. Capacity of each relief valve will be determined at rated pressure by the amount of bypass or control valve closure required to maintain reactor pressure. Proper reseating of each relief valve will be verified by observation of temperatures in the relief valve discharge piping. At selected test conditions each valve will be manually actuated, and at least one valve will be timed. Criteria Level 1 The sum total of capacities from 11 relief valves shall be equal to or greater than 8.83 x 10 lb/hr 2 percent corrected for an inlet pressure of 1112 psig. Level 2 Relief valve leakage shall be low enough that the temperature measured by the thermocouples in the discharge side of the valves returns to within 10°F (5.6°C) of the temperature recorded before the valve was opened. Each individual relief valve shall have a minimum capacity of 720,000 corrected for an inlet pressure of 1112 psig. TEST NUMBER 27 -- TURBINE TRIP AND GENERATOR LOAD REJECTION Purpose The purpose of this test is to demonstrate the response of the reactor and its control systems to protective trips in the turbine and generator. Description The turbine stop valves will be tripped at selected reactor power levels and the main generator breaker will be tripped in such a way that a load imbalance trip occurs. Several reactor and turbine operating parameters will be monitored to evaluate the response of the bypass valves, relief valves, and reactor protection system (RPS). 13.5-62

BFN-22 Additionally, the peak values and change rates of reactor steam pressure and heat flux will be determined. The effect of recirculation pump overspeed, if any, will be checked during the generator load rejection. The ability to ride through a load rejection within bypass capacity without a scram will be demonstrated. Criteria Level 1 The initial transient rise in vessel dome pressure occurring within 10 seconds of the turbine/generator trip initiation shall not be greater than 150 psi and the transient rise in simulated heat flux shall not exceed 10 percent. The turbine stop valves must begin to close before the control valves for the turbine trip. The turbine control valves must begin to close before the stop valves during the generator load rejection. Following fast closure of the turbine stop and control valves, a reactor scram shall occur if the turbine first stage pressure is greater than 154 psig. Feedwater systems must prevent flooding of the steamline following the transients. Level 2 The initial transient rise in vessel dome pressure occurring within 10 seconds of the turbine/generator trip initiation and the transient rise in simulated surface heat flux shall not be more limiting than the predicted transient presented in the Transient Analysis Design Report (100 psi and no heat flux increase). The pressure regulator must prevent a low-pressure reactor isolation. The feedwater controller must prevent a low-level initiation of the HPCI and MSIV's as long as feedwater remains available. The load rejection within bypass capacity must not cause a scram. The trip scram function for higher power levels must meet RPS specifications. For the case of turbine trip at 75-percent power, the measured transient parameters will be compared with the predicted values. If any parameter is significantly different from the predicted values, the test will be repeated at 100-percent power. 13.5-63

BFN-22 TEST NUMBER 29 -- FLOW CONTROL (Browns Ferry Unit 2 Only) Purpose The purposes of this test are (a) to determine the plant response to changes in the recirculation flow, (b) to optimize the setpoints of the Master Flow Controller and Transient Pressure Set-Point Adjuster, and (c) to demonstrate the plant load following capability in Master Manual, and Automatic Flow Control modes. Description Various process variables will be recorded while load changes (increase and decrease) are introduced into the recirculation flow control system at chosen points on the 50 percent, 75 percent, and 100 percent load lines. The master flow controller and transient pressure set-point adjuster will be set to achieve acceptable performance over the entire auto flow control range. Ramp changes will be made with the concurrence of the customer at rates within the range of 10 percent-30 percent power per minute. Load following capability will be demonstrated in the automatic and master manual flow control modes. Criteria Level 1 The decay ratio must be less than 1.0 for each process variable that exhibits oscillatory response to flow control changes. Level 2 The decay ratio is expected to be less than or equal to 0.25 for each process variable that exhibits oscillatory response to flow control changes when the plant is operating above the lower limit setting of the Master Flow Controller. Following a 10 percent step change in load demand on the 100 percent load line, the reactor must not scram and the load change must be achieved within 40 seconds. The automatic flow control range must be at least 80 percent to 100 percent power along the 100 percent load line. The Master Flow Controller output limiters shall be set accordingly. The load change resulting from a maximum ramp increase in load reference from 80 percent to 100 percent load must be achieved within 1 minute without reactor scram. 13.5-64

BFN-22 Steady-state limit cycles, if any, shall produce turbine steam flow variation no larger than 0.5 percent of rated flow as measured by the gross generator electrical power output. TEST NUMBER 30 -- RECIRCULATION SYSTEM (Browns Ferry Unit 2 Only) Purpose The purposes of this test are (a) to evaluate the recirculation flow and power level transients following trips of one or both of the recirculation pumps, (b) to obtain recirculation system performance data, and (c) to verify that no recirculation system cavitation will occur on the operable region of the power-flow map. Description Single and both recirculation pumps will be tripped at various power levels. Two pump trips will be initiated by tripping the MG set drive motors. One single pump trip at 50 percent power will be initiated by opening the generator field breaker. The remaining single pump trips are to be initiated by tripping the MG set drive motor. Reactor operating parameters will be recorded during the transient and at steady-state conditions. MCHFR evaluations will be made for conditions encountered during the transient. With the recirculation pumps operating at the speed corresponding to rated flow at rated power, power will be reduced by inserting rods to 23 percent power where the recirculation pumps will automatically run back to 20 percent speed. A check will be made to determine if recirculation or jet pump cavitation occurs. Criteria Level 1 MCHFR shall be greater than 1.0 during the pump trip transient. Level 2 For each pump trip test, the minimum transient MCHFR based on operating data divided by the corresponding minimum transient MCHFR evaluated from design values is expected to be equal to or greater than 1.0. 13.5-65

BFN-22 TEST NUMBER 30 -- RECIRCULATION SYSTEM (Browns Ferry Unit 3 Only) Purpose The purposes of this test are (a) to evaluate the recirculation flow and power level transients following trips of one or both of the recirculation pumps, (b) to obtain recirculation system performance data, and (c) to verify that no recirculation system cavitation will occur on the operable region of the power-flow map. Description Single and both recirculation pumps will be tripped at various power levels. Two pump trips will be initiated by tripping the MG set drive motors. One single pump trip at 50 percent power will be initiated by opening the generator field breaker. The remaining single pump trips are to be initiated by tripping the MG set drive motor. Reactor operating parameters will be recorded during the transient and at steady-state conditions. Criteria Level 1 Not applicable. Level 2 The power and flow coastdowns are expected to agree with precalculated power and flow coastdown rates. The plant shall not scram as a result of a high level turbine trip. TEST NUMBER 31 -- LOSS OF TURBINE-GENERATOR AND OFFSITE POWER Purpose The purpose of this test is to determine the reactor transient performance during the loss of the main generator and all offsite power, and to demonstrate acceptable performance of the station electrical supply system. Description The loss of auxiliary power test will be performed at 20 to 30 percent of rated power. The proper response of reactor plant equipment, automatic switching equipment, 13.5-66

BFN-22 and the proper sequencing of the diesel generator load will be checked. Appropriate reactor parameters will be recorded during the resultant transient. Criteria Level 1 The initial transient rise in vessel dome pressure occurring within 10 seconds of turbine/generator trip action when initiated simultaneously with loss of offsite power when performed at 25-percent power shall not exceed 150 psi and the simulated heat flux rise shall not exceed 10 percent. All safety systems, such as the RPS, its diesel generators, and the RCIC and HPCI, must function properly without manual assistance. Level 2 The initial transient rise in vessel dome pressure occurring within 10 seconds of turbine/generator trip shall not be greater than 75 psi, and there shall be no significant increase in simulated heat flux. Normal reactor cooling water systems should be able to maintain adequate pressure suppression pool water temperature, adequate drywell cooling, and prevent actuation of the autodepressurization system. TEST NUMBER 32 -- RECIRCULATION MG SET SPEED CONTROL (Browns Ferry Unit 2 Only) Purpose The purposes of this test are (a) to determine the speed control characteristics of the MG sets in the recirculation control system, (b) to obtain acceptable speed control system performance, and (c) to determine the maximum allowable pump speed. Description During the initial startup testing, data will be collected to optimize the loop gains. The cams in the scoop tube positioner feedback loops will be programmed to reduce the effect of abrupt nonlinearities in the coupler characteristics. The time response of the individual recirculation pump speed loops will be optimized by adjusting the gains of the speed controllers. The response of the speed loops will then be checked by step changes in speed demand at all test conditions. 13.5-67

BFN-22 Criteria Level 1 The decay ratio must be less than 1.0 for each process variable that exhibits oscillatory response to recirculation MG set speed changes. Level 2 The decay ratio should be less than or equal to 0.25 for each process variable that exhibits oscillatory response to recirculation MG set speed changes over the entire range from 20 percent to 100 percent speed. Following a 10 percent step change in speed demand from any speed in the speed control range, the time from the step demand until the generator speed peak occurs shall be greater than 10 but less than 25 seconds. Steady-state limit cycles, if any, shall cause turbine steam flow variations no larger than 0.5 percent of rated flow as measured by the gross generator electrical power output. TEST NUMBER 32 -- RECIRCULATION SPEED CONTROL AND LOAD FOLLOWING (Browns Ferry Unit 3 Only) Purpose The purposes of this test are (a) to determine correct gain for optimum performance of individual recirculation loops, (b) to determine that the recirculation loops are correctly set up for desired speed range and for acceptable variations in loop gain, (c) to demonstrate plant response to changes in recirculation flow. Description During the initial startup testing, data will be collected for programming the CAM's in the scoop tube positioner feedback loops to reduce the effect of abrupt nonlinearities in the coupler characteristics. The time response of the individual recirculation pump speed loops will be optimized for smooth control while staying within the PCIOMR. The response of the speed loops will then be checked by step changes in speed demand at all test conditions. The master flow controller and transient pressure setpoint adjuster will be set to achieve acceptable performance over the auto flow control range. 13.5-68

BFN-22 Criteria Level 1 The decay ratio must be less than 1.0 for each process variable that exhibits oscillatory response to flow control changes. Level 2 The decay ratio should be less than 0.25 for any process variable that exhibits oscillatory response to 10 percent speed change inputs in local or master manual modes. Steady-state limit cycles, if any exist, must not cause turbine steam flow to vary in excess of 0.5 percent rated flow as measured by the gross generator electrical power output. TEST NUMBER 33 -- MAIN TURBINE STOP VALVE SURVEILLANCE TEST Purpose The purpose of this test is to demonstrate acceptable procedure for daily turbine stop valve surveillance tests at a power level as high as possible without producing reactor scram. Description Individual main turbine stop valves must be closed daily during plant operation as required for plant surveillance testing. At several test points the response of the reactor will be recorded, and the maximum possible power level for performance of this test along with the 100 percent power flow control line will be established. Each stop valve closure is manually initiated and reset. Rate of valve stroking and timing of the close-open sequence will be chosen to minimize the disturbance introduced. Criteria Level 1 Not applicable 13.5-69

BFN-22 Criteria Level 1 The release of radioactive gaseous and particulate effluents must not exceed the limits specified in the site technical specifications. There shall be no loss of flow of dilution steam to the noncondensing stage when the steam jet air ejectors are pumping. Level 2 The system flow, pressure, temperature, and relative humidity shall comply with design specifications. The catalytic recombiner, the hydrogen analyzer, the activated carbon beds, and the filters shall be working properly during operation. 13.5.3 Nuclear System Startup Test Restrictions All operations and tests must comply with the warranty limitations specified by GE, and primarily with the safety limitations and limiting conditions for operations specified by licensing authorities. Restrictions are minimized because the prime objective of the startup program is to demonstrate that the plant is capable of operating safely and satisfactorily up to rated power. Any restrictions are detailed in the written startup procedures and in supporting instructions. An example is that during initial fuel loading operations, a special test neutron source is installed near the first fuel loading location such that sufficient neutron flux is obtained to provide positive indications of neutron multiplication. The source is enclosed in a rod-like holder and is installed in the core during initial criticality measurements. The source is removed when no longer needed for initial tests, and is replaced with the normal operational neutron sources, which provide ample neutron flux for nuclear instrument readings to meet the criteria on noise, signal-to-noise ratio and response to changes in core reactivity. Several more-sensitive neutron detectors may be submerged in the core near the loading location and surrounding the test source to provide additional information during initial fuel loading. These detectors will be removed following completion of the open vessel reactivity testing. During the startup program the power level will not exceed the upper power level for the test plateau (flow control line) at which testing is being performed. The upper power level values for the test plateaus are stated in Tables 13.5-4, 5 and 6. These upper power level values for testing at the 50 percent, 75 percent, and 100 percent flow control lines are 60 percent, 85 percent, and 100 percent power respectively. 13.5-70

BFN-22 TEST NUMBER 75 -- REACTOR SHUTDOWN FROM OUTSIDE THE MAIN CONTROL ROOM Purpose The purpose of this test is to demonstrate that the plant is designed and constructed with adequate instruments and controls to permit safe reactor shutdown from outside the main control room and maintain it in a safe condition, that the minimum number of personnel required by the technical specifications is adequate without affecting the safe, continuous operation of the other units, and that the plant emergency operating instruction is adequate. Description With the unit operating at greater than 10-percent generator output, the reactor will be scrammed by closing the MSIV's from the backup control station. Operators will man their backup control stations as described in the emergency operating instruction. The RCIC system will be operated from the backup controls to supply water to the reactor vessel. The suppression pool cooling system shall be placed in operation using the backup controls. An extra licensed operator will remain in the main control room to assure that the test is terminated and control returned to normal if any unexpected conditions occur. The test will be terminated when it is assured that the reactor can be maintained in a safe hot standby condition from the backup controls. Criteria Level 1 Not applicable. Level 2 Reactor scram initiated from outside the control room must occur. Reactor water level must be maintained greater than 490 inches above vessel 0 and less than the high level turbine trip point. The RHR and RHRSW pumps and control valves shall be operable from the backup controls to initiate suppression pool cooling. The minimum number of shift personnel as specified in the technical specifications is adequate for shutdown from outside the main control room. 13.5-71

BFN-22 Level 2 The temperature at the tube side outlet of the nonregenerative heat exchanges shall not exceed 130°F in any mode. The pump available NPSH will be 13 feet or greater during the hot standby mode defined in the process diagrams. The cooling water supplied to the nonregenerative heat exchanges shall be within the flow and outlet temperature limits indicated in the process diagrams. (This is applicable to "normal" and "blowdown" modes.) TEST NUMBER 71 -- RESIDUAL HEAT REMOVAL SYSTEM Purpose The purpose of this test is to demonstrate the ability of the Residual Heat Removal (RHR) System to remove residual and decay heat from the nuclear system so that refueling and nuclear system servicing can be performed. Description With the reactor at 100 psig (3.5 kg/cm ) or less, the shutdown cooling mode of the RHR system will be demonstrated. The suppression pool cooling mode will also be demonstrated unless its functionability is shown elsewhere. Criteria Level 1 Not applicable. Level 2 The heat removal capability of each RHR heat exchanger in the shutdown cooling mode shall be at least 187 x 10 Btu/hr when the inlet flows and temperatures are as indicated on the process diagrams. 13.5-72

BFN-22 TEST NUMBER 72 -- DRYWELL ATMOSPHERE COOLING SYSTEM Purpose The purpose of this test is to verify the ability of the Drywell Atmosphere Cooling System to maintain design conditions in the drywell during operating conditions and post scram conditions. Description During heatup and power operation, data will be taken to ascertain that the drywell atmospheric conditions are within design limits. Criteria Level 1 Not applicable. Level 2 The heat removal capability of the drywell coolers shall be approximately 5.19 x 10 Btu/hr with eight fans and coils in operation. The drywell cooling system shall have a standby capability of 25 percent of the above heat removal capability. The drywell cooling system shall maintain temperatures in the drywell below the following design values during normal operation. Average temperature throughout the drywell - 150°F Maximum around the recirculating pump motors - 135°F Maximum above the bulkhead - 200°F Maximum all other areas - 180°F Maintain a uniform circumferential temperature of the refueling bellows/bulkhead within 25°F point-to-point variation. Within 10 hours following shutdown the average temperature throughout the drywell will be within 15°F of the reactor building closed cooling water inlet temperature. 13.5-73

BFN-22 TEST NUMBER 73 -- COOLING WATER SYSTEMS Purpose The purpose of this test is to verify that the performance of the Reactor Building Closed Cooling Water (RBCCW) and the raw cooling water systems is adequate with the reactor at rated temperature. Description With the reactor at rated pressure, following initial heatup, data will be obtained to verify that the flow rates in the RBCCW heat exchangers are adequate and properly balanced, and that the heat exchanger outlet temperatures are balanced within design values. Flow rate adjustments will be made as necessary to achieve satisfactory system performance. The test will be repeated at selected power levels to verify continued satisfactory performance with higher plant heat loads. Criteria Level 1 Not applicable. Level 2 Verification that the system performance meets the cooling requirements constitutes satisfactory completion of this test. The RBCCW system was designed to transfer a maximum heat load of 31.3 x 10 Btu/hr in order to limit equipment inlet water temperature to 100°F, assuming a service (raw cooling) water inlet temperature of 90°F. TEST NUMBER 74 -- OFF GAS SYSTEM* Purpose The purposes of this test are to verify the proper operation of the Off Gas System over its expected operating parameters and to determine the performance of the activated carbon adsorbers. Description The pressure, temperature, relative humidity, system flow, and percentage of radiolytic hydrogen in the off gas are periodically monitored during startup and at 13.5-74

BFN-22 steadystate conditions. Provided that measurable and sufficient fission gases and fission gas daughter products are present in the off gas, decontamination factors across the pre- and post-filters and several charcoal beds are determined. The performance of the catalytic recombiner will be compared with the Catalytic Recombiner Guaranteed Performance Curve. Criteria Level 1 The release of radioactive gaseous and particulate effluents must not exceed the limits specified in the site technical specifications. There shall be no loss of flow of dilution steam to the noncondensing stage when the steam jet air ejectors are pumping. Level 2 The system flow, pressure, temperature, and relative humidity shall comply with design specifications. The catalytic recombiner, the hydrogen analyzer, the activated carbon beds, and the filters shall be working properly during operation.

  • Applies to Unit 3 and, subsequent to equipment installation, to Units 1 and 2.

13.5-75

BFN-16 Table 13.5-1 CONTROL ROD DRIVE SYSTEM TESTS (Unit 1) Reactor Pressure,psig Preop (With Core Loaded) Test Description Tests 0 600 800 1000 Position Indication All All Normal Insert/Withdraw Times All All 4* Coupling All All Friction All 4* Scram Times (Normal Accumulated Pressure) All All 4* 4* All Scram Times (Minimum Accumulator Pressure) 4* Scram Times (Zero Accumulator Pressure) 4* Scram Times (Scram Discharge Volume High Level) All Scram Times, Rated Power (Normal Accumulator Pressure) 4** 

*Value refers to the four slowest drives as determined from the normal accumulator pressure scram test at ambient reactor pressure which are compatible with the Rod Worth Minimizer and withdrawal sequence requirements.
    • Scram times of the four slowest rods will be determined at 25, 50, and 100 percent of rated power during planned reactor scrams at these power levels.

Table 13.5-2 MAJOR PLANT TRANSIENTS TEST CONDITION Nominal Power, Percent of Rated 25 50 75 80 100 Nominal Core Flow, Percent of Rated TEST TITLE 36 100 100 70 100 Feedwater Pump Trip A C Main Steam Isolation Valves (One Valve) C A C Main Steam Isolation Valves (All Valves) C Turbine-Generator Stop Valve Fast Close B A Turbine-Generator Contorl Valve Fast Close A A A Recirculation Pump Trip (One Pump) C B C Recirculation Pump Trip (Two Pumps) C C C Loss of Generator and Offsite Power C A - Unit 1 B - Units 2, 3 C - Units 1, 2, 3

BFN-16 Table 13.5-3 STABILITY TESTS TEST CONDITION Nominal Power, Percent of Rated 25 40 50 25 47 60 75 37 65 80 100 50 97 Nominal Core Flow, Percent of Rated TEST TITLE 47 70 100 NC 48 70 100 NC 48 70 100 NC 109 Flux Response to Rods C A C A A C A A A C A Pressure Regulator Setpoint C A C A A C A C C C A Pressure REgulator Backup Regulator C C C C Feedwater System Setpoint C A C A A C A C C C A B Feedwater System Drop Heater A Bypass Valve C A C B A A C A C C C C B Flow Control C C C C C C C C C NC= Natural Circulation A - unit 1 B - units 2, 3 C - units 1, 2, 3

BFN-16 TABLE 13.5-4 STARTUP TEST PROGRAM (UNIT 1) OPEN 50% Flow Line 75% Flow Line 100% Flow Line Power1 .% VESS HEAT 15-35 30-50 40-60 15-35 37-57 50-70 65-85 27-47 55-75 70-90 95-100 40-60 20% Flow2.% OR UP 48 70 104 NC 48 70 102 NC 48 70 100 NC 105% TEST CONDITION (See Figure 13.5-2) COLD 1 3 2 2A 5A 5 4 4A 7A 7 6 6A 8 TEST 1 Chemical & Radiochemical X X X X X X 2 Radiation Measurements X X X X X 3 Fuel Loading X 4 Full Core Shutdown Margin X 5 Control Rod Drive System X X X X X 6 SRM Performance & Control Rod Sequence X X X X X 7 Not Applicable 8 Not Applicable 9 Not Applicable 10 IRM Calibration X X X 11 LPRM Calibration X X X X 12 APRM Calibration X X X X X X 13 Process Computer X X X X 14 RCIC System X M 15 HPCI System X M 16 Reactor Vessel Temperature X X X X 17 System Expansion X X X 18 Core Power Distribution X X X X X 19 Core Performance X X X X X X X X X X X X X 20 Elec. Output & Heat Rate X 21 Flux Response to Rods M M M M M M M X M M M X 22 Press. Reg.: Setpoint Changes M M M M M M M X M M M X

Backup Regulator M M M M 23 FW System: FW Pump Trip M M
Water Level Stpt. Chg. MA M MA M M M MA X MA MA MA X
Heater Loss M*

24 Bypass Valves MA MA MA M MA MA MA X MA MA MA X 25 Main Steam Iso. Valves: Each Vlv. X M,SP M,SP M,SP

Full Iso. M,SE 26 Relief Valves X M M,SE M 27 Turbine : Stop Valve Trip M,SE
Control Valve Trip M,SP M,SE A,SE 29 Flow Control LMA MA MA MA MA MA MA MA MA 30 Recirc. System: One Pump Trip M M
Two Pump Trip M M M
Flow Calibration X X X X X X 31 Loss of T-G & Offsite Power M,SE LSE 32 Recirc. Loop Control X X X X 35 Recirc. And Jet Pump Calibration X 36 Equalizer Open M M 39 Water Level verification in Reactor X X X Vessel 70 Reactor Water Cleanup System X 71 Residual Heat Removal System**

72 Drywell Atmosphere Cooling System X X 73 Cooling Water Systems X X 90 Vibration** X X X X X X 92 Steam Separator and Dryer X X XSP X XSP X*,X XSP 93 Not Applicable 1Power is in percent of rated power, 3293Mwt. M=Master Manual Control Mode SP=Scram Possibility 2Flow is in percent of rated flow, 102.5x105 lbs/hr A=Automatic Control Mode NC=Natural Circulation

  • = 90% rated power L=Local Manual Control Mode SE=Scram Expected
    • = Actual test condition to be determined X=Test independent of flow control mode.

BFN-16 TABLE 13.5-5 STARTUP TEST PROGRAM (UNIT 2) OPEN 50% Flow Control Line 75% Flow Control Line 100% Flow Control Line Power, %1 VESSEL 15-35 30-50 40-60 25 53 37-57 50-70 65-85 37 55-75 70-90 95-100 50 95-100 Flow, %2 OR HEAT 47 70 104 NC 116 48 70 102 NC 48 70 100 NC 109 COLD UP 1 2D 2E 2A 2F 3C 3D 3E 3A 4C 4D 4E 4A 4F TEST CONDITION (See Figure 2) TEST 1 Chemical & Radiochemical X X X X X X 2 Radiation Measurements X X X X X X 3 Fuel Loading X 4 Full Core Shutdown Margin X 5 CRD X X X X X 6 SRM Perf. & Control Rod Seq. X X X 9 Water Level Measurement X X X 10 IRM Calibration X X X 11 LPRM Calibration X X X X 12 APRM Calibration X X X X X 13 Process Computer X X X6 X 14 RCIC X M 15 HPCI X X X 16 Selected Process Temperatures X X X X 17 System Expansion X X X 18 Core Power Distribution X X X X 19 Core Performance X X X X X X X X X X X X X X 20 Steam Production X 21 Flux Response to Rods M M M M 22 Press. Reg.: Setpoint Changes M M M M M M

Backup Regulator M M M M 23 FW System: FW Pump Trip M
Water Level Stpt. Chg. M,A M,A M,A M,A M,A M,A M 24 Bypass Valves M,A M,A X M,A MA, M,A M,A M M 9

25 Main Steam Iso. Valves: Each Vlv. X M,SP M,SP

Full Iso. M,SE 26 Relief Valve: Capacity M 3 9
Actuation X M M 5

27 Turbine Trip and M,SE 5 Generator Load Rejection M,SP M,SF 29 Flow Control A A A A A A A A M,A M,A 7 8 7 7 7 7 30 Recirc. System : Trip One Pump M ,M M ,M M ,M 7 Trip Both Pumps M M M Sys. Performance X X X X X X X X 12 Non-Cavit. Verif. X 5 31 Loss of T-G & Offsite Power M,SE 32 Recirc. MG Set Speed Control L,M M M M M M M M M M 11 33 Turbine Stop Valve Surv. Test M M M M,SP 13 34 Vibration Measurements 4 X X X X X X X X X X X X X X 35 Recirc. System Flow Calibration X X X 70 Reactor Water Cleanup System X 71 Residual Heat Removal System** X 72 Drywell Atmosphere Cooling System 10 X X 73 Cooling Water Systems 10 X X 74 Offgas System* X X X X 1 7 2 Percent of rated power, 3293 Mwt 6 8Included only to meet Test 34 Requirements L = Local Manual Flow Control mode 3Percent of rated flow, 102.5 x 10 lb/hr 9Trip the Generator Field Breaker M = Master Manual Flow Control Mode 4Also obtain data with Tests 25, Full Iso & Test 27 Heat up tests of MSIVs & Relief Valves are A = Automatic flow Control Mode 5Obtain data with test 30 10 to check operation only X =Test Independent of Flow Control Mode Perform Test 5, timing of 4 slowest control rods 11Unit II only (cf. Page 36) SP = Scram Possibility 6 in conjunction with these scrams 12Determine Maximum power without scram SE = Scram Expected Perform the Dynamic System Test Case 13From Test Condition 2E to 5 NC = Natural Circulation Not required if 50% power testing will be done within about 2 months

  • Applies to unit 3 and, subsequent to equipment installation, to unit 2.

BFN-16 TABLE 13.5-6 STARTUP TEST PROGRAM (UNIT 3) OPEN 50% Flow Control Line 75% Flow Control Line 100% Flow Control Line Power, %1 VESSEL 15-35 30-50 40-60 25 37-57 50-70 65-85 37 55-75 70-90 95-100 50 Flow, %2 OR HEAT 47 70 104 NC 48 70 102 NC 48 70 100 NC COLD UP 1 2D 2E 2A 3C 3D 3E 3A 4C 4D 4E 4A TEST CONDITION (See Figure 2) TEST 1 Chemical & Radiochemical X X X X X X 2 Radiation Measurement X X X X X X 3 Fuel Loading X 4 Full Core Shutdown Margin X 5 CRD X X X10 X 6 SRM Perf. & Control Rod Seq. X X X 9 Water Level Measurement X X X 10 IRM Performance X X X 11 LPRM Calibration X X X X 12 APRM Calibration X X X X X 13 Process Computer X X X6 X 14 RCIC X X 15 HPCI X X 16 Selected Process Temperature X X 17 System Expansion X X X 18 Core Power Distribution X X X X 19 Core Performance X X X X X X X X X X X 20 Steam Production X 21 Flux Response to Rods X X X X 22 Press. Reg.: Setpoint Changes X X X X X X

Backup Regulator X X X X 23 FW System: FW Pump Trip X
Water Level Stpt. Chg. X X X X X X 24 Bypass Valves X X X X X X X X 25 Main Steam Iso. Valves: Each Vlv. X9 X,SP X,SP
Full Iso. X,SP 26 Relief Valve: Capacity X
Actuations3 X9 X X 27 Turbine Trip and X,SE5 Generator Load Rejection X,SP X,SE5 30 Recirc. System : Trip One Pump X7,X8 X7,X7 X7,X7 Trip Both Pumps X X X Sys. Performance X X X X X Non-Cavit. Verif. X12 31 Loss of T-G & Offsite Power X,SE5 32 Recirc. MG Set Speed Control L,M L,M L,M L,M L,M L,M L,M L,M L,M 33 Turbine Stop Valve Surv. Test X X X X,SP11 34 Vibration Measurements 4 X13 X X X X X X X X X X 35 Recirc. System Flow Calibration X X X 70 Reactor Water Cleanup System X 71 Residual Heat Removal System X 72 Drywell Atmospheric Cooling System X X 73 Cooling Water Systems X X 74 Offgas System X X X X 75 Reactor Shutdown From Outside Control X14 Room 1Percent of rated power, 3293 MWt 7 Included only to meet Test 34 Requirements L = Local Manual Flow Control Mode 2Percent of rated flow, 102.5 x 10 6 lb/hr 8 Trip the Generator Field Breaker M = Master Manual Flow Control Mode 3Also obtain data with Tests 25, Full Iso & Test 27 9 Heat up tests of MSIVs & Relief Valves are A = Automatic Flow Control Mode 4Obtain data with Test 30 to check operation only x =Test Indepentent of Flow Control Mode 5Perform Test 5, timing of 4 slowest control rods 10 RSCS cleared ( 40%) SP = Scram Possibility 11 Determine maximum power without scram in conjunction with these scrams SE = Scram Expected 6Perform the Dynamic System Test Case 12 From Test Condition 2E to 5 NC = Natural Circulation 13 Not required if 50% power testing will be done in about 2 months 14 At greater than 10% generator output

Chemistry C!JD Final Water Proce ss Computer ,._ END 

                    ;'Te,e._st"s' - - -
  • END Scrrun Shutdo."11l I,evel Cold Tests Marg in Profil ibrntior, Tests Shutdmm
                                                                          '                                                                               6 CR.D Scram           Increase to r

Tests 30% Power System Expansion Heat Bal APRM/lRM Ca.lib Heat Bs.1 RCIC Xenon Stab . Perio APRM Cal c Open1.tion Flot.1 Calib MCHFR I'\ Press Bypass ll:eg n ve Relief TC Control Valve Control Rod Recover to 4o;t Paver 

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BFN-27 13.6 Normal Operations 13.6.1 General Day-to-day operations are carried out by the various plant organizations. Each organization, within its assigned area of responsibility, operates with some degree of independence and freedom from close supervision, yet their actions are closely coordinated to best achieve the common purpose. The Site Vice President, BFN, issues procedures governing employee actions and establishing standards for plant activities. Additionally, standard NPG administrative procedures are issued by the Vice President Nuclear Support which are applicable to all TVA Nuclear plants. Managers of principal organizations issue instructions governing activities under their cognizance. The plant manager issues instructions which contain administrative restrictions and station requirements established to ensure safe operation of the plant within the limits set by the facility licenses and technical specifications. They provide that plant activities will be conducted in a manner to protect the general public, plant personnel, and equipment. A formalized system of written procedures is employed conforming to the requirements of the Nuclear Quality Assurance Plan (TVA-NQA-PLN-89-A). Procedures covering plant activities which might adversely affect safety are put into effect only after being reviewed and approved by appropriate members of the plant staff as specified in the NQAP. These activities include operation, maintenance, testing and modifications. The Plant Operations Review Committee (PORC), is responsible for reviewing proposed changes as outlined in the NQAP. The plant manager has the responsibility to ensure that safety related procedures prepared by his staff or submitted to him by other organizations receive required reviews and approvals before authorizations are issued. There is, in addition to planned changes in the plant and procedures, the area of accidental or gradual changes in plant equipment characteristics or conditions. Each supervisor and employee has the responsibility to be continually alert for such changes and for reporting them upon detection. The periodic inspection of plant equipment and the continuing review and analysis of operating data from plant logs, instruments, and tests provide regular sources of information on plant conditions. 13.6.2 Normal Operating Instructions Normal Operating Instructions include those procedures for individual system operation, including precautions, preoperational requirements, startup, and 13.6-1

BFN-27 shutdown; integrated operations for major plant activities, including startup, shutdown, and refueling; general equipment operation, such as valves and electrical breakers; responses to annunciators; anticipated equipment or system abnormalities; layup of systems; and instructions for operation during special circumstances to compensate for changed plant conditions (usually used only once for those special conditions). 13.6.3 Emergency Operating Instructions As a result of the accident at the Three Mile Island Nuclear Plant, the NRC required the upgrade of Emergency Operating Instructions (EOIs) from event based procedures to symptom-based procedures. Symptom-based procedures are procedures which do not require the operator to diagnose the event in order to be able to correctly respond to the event. Instead, the operator responds to symptoms that result from the initiating event(s) and attempts to control these symptoms within established limits. Browns Ferry EOIs are symptom-based procedures and are based on the Boiling Water Reactor Owners Group (BWROG) Emergency Procedure Guidelines (EPGs). The EOIs specify actions necessary to address any event including less than design basis events, design basis events, and beyond design basis events (e.g., multiple equipment failures) and specifies limits to assure continued safe operation of the plant. The operator monitors the general state of the plant. If any of the entry conditions for the EOIs are met, the operator takes actions to control reactor power, reactor pressure, reactor water level, primary containment pressure and hydrogen concentration, drywell temperature, pressure suppression pool temperature, pressure suppression pool water level, and secondary containment temperatures, water levels, and radiation levels. The EOIs are used until either directed out by the EOIs or when the operator concludes that an emergency condition no longer exists. All operating personnel through training and experience learn to recognize and evaluate impending failures or malfunctions and to initiate proper corrective actions. The EOIs are used to train the operating personnel and make them aware of the accidents or situations that could occur, and the proper course of action. Specialized Fire Safe Shutdown (FSS) procedures directing activities to bring a unit to a safe and stable condition following a fire (if required) detail which equipment can be relied upon for performing the required steps, dependent on the location of the fire. 13.6-2

BFN-27 13.6.4 Maintenance Instructions The plant maintenance program is designed to safely and efficiently provide maintenance and repair to keep the plant in good operating order. Maintenance is initiated through the maintenance/work request and a preventive maintenance program. Safe working conditions are assured by the use of TVA's Hold Order, Clearance, and Radiation Work Permit procedures. Complex maintenance operations require step-by-step performance and therefore are detailed in written instructions. These instructions, covering mechanical, electrical, and instrumentation maintenance will provide information to assure proper coordination of operating and maintenance personnel as well as step-by-step procedures for items such as removal and installation of control rod drives. 13.6.5 Radiological Emergency Plan 13.6.5.1 General The Radiological Emergency Plan contains the precautionary planning, delegation of authority and responsibility, and implementing procedures to protect the public, plant employees, and equipment in case of unusual incidents. The plan contains information for control of emergency conditions modeled after those contained in NUREG-0654, Rev. 1 and Regulatory Guide 1.101, Rev. 3 as appropriate. 13.6.5.2 Radiological Emergency Plan The Radiological Emergency Plan provides that the Shift Manager on duty is responsible for placing the plan in effect based on conditions listed in the implementing procedures. Upon declaration of an emergency, the Shift Manager provides for notification of the designated people, and assumes the duties of site emergency director until relieved by the plant manager or an alternate. The emergency organization is as described in the Radiological Emergency Plan, and staffed by pre-selected, experienced personnel. The entire plant facilities and personnel, as well as other TVA organizations, are at the site emergency director's disposal. Adequate qualified personnel are available to staff the emergency organization around the clock until the emergency is over. Advance plans and arrangements have been made in conjunction with state and local authorities, where applicable, for warning the local populace of an emergency and possible evacuation, evacuating the area around the plant site, preventing entry of the public to affected areas, medical care of injured or exposed personnel, surveying affected areas for radioactivity, and restricting use of water supplies and foods. The plant emergency organization will be controlled from the Technical Support Center. This center is adequately shielded to ensure it is occupiable throughout an 13.6-3

BFN-27 incident. Communications consist of a dedicated emergency telephone system and the TVA radio system. The Technical Support Center contains area maps, plant drawings, copies of the emergency plan and necessary emergency supplies and equipment. The overall TVA emergency organization is controlled from the Central Emergency Control Center in Chattanooga. Additional emergency supplies and equipment are stored in various plant locations. One vehicle is fully equipped for area radiological surveillance duty. When an emergency notification is received, all personnel with emergency duty assignments report to the director and assume preassigned duties. All other personnel will normally remain at their work stations, unless accountability is required. They then assemble in a designated area and wait for further instructions. Radiation surveillance teams and other emergency personnel will survey affected areas and report to the director who will evaluate the situation and take appropriate actions as outlined in the emergency plan. Complete details are available in the Radiological Emergency Plan and Emergency Plan Implementing Procedures. 13.6.6 Radiation Control Instructions These procedures provide guidance for the protection of employees and the public from nuclear radiation and contamination. Applicable requirements of the NRC, as published in Title 10, Code of Federal Regulations, Part 20 and 30, and regulations of the Department of Transportation have been incorporated. 13.6.7 Surveillance Instructions These instructions cover periodic tests and inspections required by the technical specifications to assure proper operations, and to prove the adequacy and availability of critical systems and equipment. Formalized schedules and check sheets are used to ascertain that all critical equipment and safeguards systems will satisfy their design intent. Test schedules and records will be maintained so as to provide an orderly test and surveillance program. 13.6.8 Technical Instructions Instructions concerning analytical techniques, calculations, and test procedures will be prepared as required. Examples are chemical control procedures (chemical 13.6-4

BFN-27 instructions) which provide plant personnel with instruction on the types and frequency of chemical and radiochemical analysis, steps to be taken to maintain conditions within established limits, reactor calculations procedures, and test procedures on CSSC equipment beyond that required by technical specifications. The plant computer normally will perform the calculations and assessments on the reactor core and primary system to assure that the reactor is operated safely and efficiently within limits specified by the technical specifications. The reactor calculation procedures outline the backup calculations necessary to assure overall compliance when the computer is unavailable. Fuel Accountability Procedures delineating the requirements, responsibilities, and methods of nuclear material control from the time new fuel is received until such time as it is shipped from the plant as spent fuel are contained within the technical procedures. They provide detailed steps for physical safeguards, inventory, accounting, and for preparing reports to the Nuclear Regulatory Commission. 13.6.9 Security Plan Procedures Procedures for implementing the plant Physical Security Plan are prepared as required. Included are procedures regarding badging, access control, search requirements, posting of personnel, keys and locks, lighting, and reporting requirements. 13.6.10 Special Test Instructions Instructions for special activities which are normally performed one time are not covered by existing plant instructions. These are prepared as required. 13.6.11 Modification/Addition Instructions Routine continuing requirements for fabrication, installation, and checkout of materials and components concerning plant modifications are placed in modification/addition instructions. Typically included are instructions regarding supports, bolting, coatings, cabling, and electrical terminations. 13.6-5

BFN-17 Figures 13.6-1 sheets 1 and 2 (Deleted by Amendment 17)

BFN-17 13.7 RECORDS (Deleted) 13.7-1

BFN-23 13.8 OPERATIONAL REVIEW AND AUDITS 13.8.1 GENERAL Reviews and audits are performed to ensure that activities which affect the safety related functions of the plant's structures, systems, and components are carried out without undue risk to the health and safety of the public including plant employees and that the plant is operated under a well formulated and detailed administrative controls and quality assurance program. Review and audit organizations and their responsibilities are described in the Nuclear Quality Assurance Plan (TVA-NQA-PLN-89-A). 13/9/2 Onsite Reviews 13.8.2.1 PLANT OPERATIONS REVIEW COMMITTEE (PORC) The PORC is one of the two onsite organizations involved in the reviews of safety related activities at the plant. A detailed description of the functions, composition, meeting frequency, quorum, responsibilities, authority, and records requirements of the PORC are contained in the Nuclear Quality Assurance Plan (Reference 2). 13.8.2.2 Site Quality Assurance The site quality assurance organization has specific responsibilities in the area of review of plant operations. These responsibilities are described in the Nuclear Quality Assurance Plan (TVA-NQA-PLN-89-A). Site Quality Assurance is responsible for assisting site management in meeting quality assurance and quality control requirements, pursuing the necessary corrective actions when quality assurance problems occur up to and including the initiation of "Stop Work" action, verifying that site instructions contain applicable quality assurance requirements, reviewing complete work document packages for compliance with applicable requirements on a selective basis, developing and implementing a quality control inspection program covering material receipt, construction modifications, maintenance, and testing activities, surveying quality assurance program implementation and making recommendations to site management regarding identified deficiencies, performing trend and root cause analysis of quality problems, evaluating disposition of major quality issues, interfacing with the line management on quality improvement initiatives, and developing quality assurance and startup readiness assessment plans. 13.8.3 Independent Reviews Independent review of operational activities is the function of the TVA Nuclear Safety Review Board (NSRB). A detailed description of the function, composition, 13.8-1

BFN-23 qualifications, and review responsibilities of the NSRB are contained in the Nuclear Quality Assurance Plan. The function of the NSRB encompass plant operations, the scientific and engineering disciplines, radiological safety, and quality assurance. The NSRB reviews 10 CFR 50.59 evaluations, changes to the Technical Specifications and Operating License, violations having nuclear significance, significant operating abnormalities or deviations, reportable events, and indications of unanticipated deficiencies in the design or operation of the plants structures, systems, or components. Additionally, the NSRB performs oversight of those quality assurance audit activities identified in the Nuclear Quality Assurance Plan. The chairman and members of the NSRB are appointed by the Chief Nuclear Officer (CNO). It advises the CNO on the nuclear safety significance of their reviews, the quality assurance audits, and on the adequacy and implementation of the TVA nuclear safety policies and programs. 13.8.4 Audit Program The Nuclear Quality Assurance organization is responsible for performing comprehensive planned and periodic audits of the plant and site organizations, other TVA organizations which support the nuclear program, and contractors/suppliers to determine and assess the adequacy and effectiveness of the QA program. A detailed description of this audit program including areas to be audited, frequency of audits, and disposition of the audit findings is contained in the Nuclear Quality Assurance Plan (TVA-NQA-PLN-89-A). 13.8.5 References

1. Deleted.
2. "Nuclear Quality Assurance Plan." TVA-NQA-PLN-89-A
3. Deleted.

13.8-2

BFN-19 13.9 Refueling Operations Detailed refueling procedures will be used to ensure a safe and orderly refueling. The procedures will specify or make reference to other system operation documents that specify periodic shutdown margin checks, detailed channeling and fuel handling techniques, and other precautionary steps to assure that the facility license and technical specifications are not violated. BFN has chosen to comply with the criticality requirements specified in 10 CFR 50.68(b). Appropriate restrictions are provided in plant procedures which prohibit the handling at any one time of more fuel assemblies than have been determined to be safely subcritical under the most adverse moderation conditions feasible by unborated water. When fuel is being inserted, removed, or rearranged in the core or when control rods are being installed, removed, or manipulated, licensed operators will be in the control room and on the refueling floor supervising the operations. Technical personnel will provide guidance where necessary and will verify that all fuel has the proper orientation and is in the correct location. An essential part of plant nuclear materials control and of refueling outage requirements is to have complete knowledge of the identity, location, composition, and condition of all fuel and other core components. The location of each control rod in the core will be recorded by serial number. Each fuel assembly is identified by a serial number on the handle. A permanent file of NRC material transfer reports will be maintained onsite. Documentation for each fuel assembly will have assembly type, unit and batch number, serial number, date received, as-built uranium weight, as-built U-235 weight, net weight, and other applicable data. The fuel transfer forms and documentation are lifetime records. In addition, there are records for the reactor and the spent fuel storage pool. All instructions for removing, rearranging, or adding fuel to the core are performed from detailed procedures. An independent check will be made after the core is fully loaded to ascertain that all fuel assemblies have been loaded correctly. During the reactor refuelings the fuel of highest burnup in general will be removed from the core, some fuel will be rearranged, and new fuel will be loaded into the core. The loading patterns for all refuelings are selected to provide an optimum power distribution to satisfy plant safety and economic considerations. Other refueling operations will include the replacement of control rods and in-core monitors, channeling operations, fuel sipping when necessary, and the inspection of selected portions of the reactor vessel and primary system. 13.9-1

BFN-19 Refueling operations will be similar for all three units. 13.9-2

BFN-26 13.10 REFUELING TEST PROGRAM This section presents a general description of the testing currently performed following each refueling outage at Browns Ferry. The tests described in this section are similar to those in Section 13.5, Startup and Power Test Program. Some of the initial startup tests need only be performed once in a plant's lifetime and are therefore not repeated. Due to the plant surveillance program, improved maintenance method, and greater operational experience, several of the tests were deleted or modified. Currently tests described in Sections 13.10.2.1 through 13.10.2.6 are performed each cycle. Other tests are often performed since they provide an efficient method of controlling and scheduling activities during the startup. These include tests described in Sections 13.10.2.7 through 13.10.2.10. 13.10.1 Program Description and Objectives The tests comprising the Refueling Test Program are conducted to demonstrate that overall plant or system performance is acceptable based on established design criteria and/or is consistent with previous operating history. The Refueling Test Program may be broadly divided into three phases: Phase I : Open Vessel, Phase II : Heat-up to 55% power, Phase III : 55% Power to 100% power Some individual tests are more restrictive and specify the power level at which they must be performed while others are equally valid for more than one phase. 13.10.2 Test Purpose, Description, and Acceptance Criteria 13.10.2.1 Fuel Loading Purpose This test safely and efficiently loads fuel to full core size. Description Two types of core reload are possible. One is a complete core unload, where all fuel bundles are removed from the core during unloading and the other is an in-core shuffle where spent fuel is expelled, reusable fuel is relocated and fresh fuel is loaded. After loading, a core verification is performed to ensure proper loading configuration. 13.10-1

BFN-26 Criteria The core must be altered to exactly reflect the final design configuration while maintaining subcriticality. 13.10.2.2 Full Core Shutdown Margin Purpose This test verifies that the reactor will remain subcritical throughout the cycle with any single control rod withdrawn and all other control rods fully inserted. Description After core loading is complete, the shutdown margin (SDM) is demonstrated to be at least 0.38 percent K/K throughout the upcoming cycle with the analytically determined strongest rod withdrawn. SDM is calculated by evaluating core reactivity after achieving criticality. The actual amount of rods withdrawn are compared to the predicted amount to verify that 0.38 percent K/K or better margin is available. An open vessel SDM demonstration can be performed by pulling rods equivalent to 0.38 percent K/K (plus the worth of the strongest rod) and proving subcriticality. Acceptance Criteria Level 1: The SDM of the fully loaded core must be calculated to remain at least 0.38 percent K/K, with the analytically determined strongest rod fully withdrawn, at any point in the upcoming cycle. Level 2: Criticality should occur within 1 percent K of the predicted critical rod configuration. 13.10.2.3 Control Rod Drive System Purpose This test demonstrates that the CRD system is operating properly and is capable of meeting its normal and emergency functions. Description This test is separated into the testing done at zero psig and the testing done at greater than 800 psig. Following maintenance on a control rod, that rod is 13.10-2

BFN-26 functionally checked by stroking it to full length. While the rod is being stroked, a visual check of the RPIS position indication is made; and, upon reaching the end of travel the coupling is checked for over travel. After fuel loading is finished, the functional/position indication check and the coupling check are performed for each rod. Additionally, friction testing is performed as a diagnostic tool on suspected problem rods. Sometime after reaching rated pressure but before 40 percent power, all CRD's are scram timed and the LPRM hookup to the process computer is verified by observing LPRM response to control rod motion. Acceptance Criteria Level 1

1. Each CRD must have a normal withdraw speed of less than or equal to 3.6 inches per second, indicated by a full 12-foot stroke time of no less than 40 seconds.
2. The control rod scram insertion times must be within the limiting conditions for operations specified in Technical Specifications.

Level 2

1. Each CRD must have a nominal insert or withdraw speed of 3.0 .6 inches per second, indicated by a full 12-foot stroke time between 40 and 60 seconds.
2. With response to the control rod drive friction tests, if the differential pressure variation exceeds 15 psid for a continuous drive in, a settling test may be performed, in which case, the differential settling pressure should not be less than 30 psid nor should it vary by more than 10 psid over a full stroke.
3. Proper LPRM connections shall be verified and adjustments made to provide proper inputs to the process computer.

13.10.2.4 Process Computer Purpose This test verifies the capability of the process computer to monitor plant conditions and to evaluate core performance parameters. Description Phase I 13.10-3

BFN-28 Following completion of the Cycle NSSS data installation the NSSS data will be verified for accuracy and proper location in the computer memory. Phase II

1. After reaching 10 percent power but before reaching 23 percent power, the core thermal power calculated by the process computer will be compared to a detailed manual heat balance.
2. The thermal limits for minimum critical power ratio (MCPR), maximum average planar linear heat generation rate (MAPLHGR), and maximum linear heat generation rate (LHGR) are compared to a qualified backup method.

Phase III The testing done for Phase II.2 is repeated after calibrating the LPRMs. Criteria Level 1 Not applicable. Level 2 Core Monitoring Software System (CMSS) core performance monitoring is considered operational when:

1. The location and value of the MCPR as calculated by the process computer are in the same location and within 2 percent of the MCPR as determined by any offline computer system qualified; or,
2. If the MCPR, as determined by the process computer, is in a different location than that determined by any offline computer system qualified, the values calculated for CPR by the two methods shall agree to within 2 percent for each fuel assembly.

13.10.2.5 Core Power Distributions Purpose This test confirms the reproducibility of the TIP system readings, determines the core power distribution and checks the core power symmetry. 13.10-4

BFN-26 Description The core must be in an octant symmetric rod pattern to perform these tests. For Global Nuclear Fuel Analyzed Cores: At least two full TIP sets will be run in order to measure the TIP uncertainty. The data from these TIP sets will be compared statistically to determine total TIP uncertainty. At the beginning of each operating cycle, TIP set data is used to calibrate the LPRMs prior to reaching 100 percent power (performed separately from Refueling Test Program). This activity also satisfies Technical Specifications Surveillance Requirement 3.3.1.1.7. In addition, TIP data and CMSS data taken after TIP sets will be analyzed to determine TIP asymmetry and core power symmetry. One of the TIP sets must be taken at > 75 percent power level and it is recommended that neither TIP set be taken below 50 percent power. For Framatome-ANP Analyzed Cores: The radial bundle power uncertainty is a key component of the MCPR Safety Limit methodology. A TIP asymmetry test is performed at startup to confirm that the actual integral TIP response uncertainties are consistent with expected internal TIP uncertainties. The test performed is the Chi-Squared test for Symmetric Integral TIP Measurements described in EMF-2508, POWERPLEX-III Core Monitoring Software System - Operational Comparison Test Procedure. Failure of the test does not necessarily indicate a lack of consistency but is cause for further investigation and may require a repeat of the test. Acceptance Criteria Level 1 Not applicable. Level 2 For GNF Analyzed Cores:

1. The total TIP uncertainty shall be less than 6.0 percent. This total TIP uncertainty will be obtained by averaging the total uncertainty for all data sets obtained. A minimum of two data sets is sufficient for determination of the total TIP uncertainty. The 6.0 percent represents the limiting uncertainty for which the present MCPR safety limit is valid. If this 6.0 percent uncertainty is exceeded, a detailed analysis will be made and possibly additional data sets will be taken.

13.10-5

BFN-26

2. The gross check of the TIP signal symmetry should yield a maximum deviation between symmetrically located pairs of less than 25 percent. If this criterion cannot be met, the cause of the asymmetry should be investigated.

For FANP Analyzed Cores:

1. The acceptance criteria are dependent upon the available number of symmetric TIP pairs. For a full set of TIPs, there are 19 pairs. The Chi-squared value for 19 pairs of TIPs should be less than 36.19.
2. For tests with fewer symmetric TIP pairs equivalent criteria are defined in EMF-2508.

13.10.2.6 APRM Calibration Purpose This test calibrates the Average Power Range Monitor System. The test is only applicable below the power level of Technical Specification Surveillance Requirement 3.3.1.1.2. Description Before there is sufficient feedwater flow to obtain an accurate heat balance, the APRMs are calibrated to a core thermal power determined either by a constant heatup rate balance or by a bypass valve comparison. APRMs will be calibrated using a heat balance after feedwater flow becomes reliable. Criteria Level 1 (Units 2 and 3) At least three of the four APRMs must be calibrated to read greater than or equal to the actual thermal power. (Unit 1 only) At least two APRMs in each RPS channel must be calibrated to read greater than or equal to the actual thermal power. 13.10-6

BFN-26 Level 2 If the level 1 criterion is satisfied, then the APRM channels are considered to be reading accurately if they do not read more than 7 percent greater than the actual core thermal power. 13.10.2.7 Pressure Regulation Purpose This test demonstrates: (a) smooth pressure control during transients induced in pressure by the pressure control system, and (b) smooth pressure control transition between control valves and bypass valves. These tests will be performed in both Reactor Pressure control and Header Pressure control to ensure expected control response in either mode. Description A pressure setpoint bias will be introduced (both increased and decreased) to produce step changes in pressure and the response of the system will be measured. These tests will be performed.

1) Off-line to test bypass valve ability to control the transient.
2) On-line with control valve position limit set so that control valves alone with control the transients.
3) On-line with control valve position limit set to test bypass valve capability to handle the excess transients.

Criteria Level 1 The decay ratio must be less than 1.0 for each process variable that exhibits oscillatory response to pressure control system changes. Level 2

1. In all tests, the decay ratio is expected to be 0.25 for each process variable that exhibits oscillatory response to pressure control system changes when the plant is operating above the lower limit of the master flow controller.

13.10-7

BFN-26

2. Pressure control deadband, delay, etc., shall be small enough that steady state limit cycles, if any, shall produce variations in turbine steam flow no larger than those specified in the following table.

Percent of Full Power Percent of Rated Flow 90 - 100 0.5 70 - 90 1.5 to 0.5 70 and below 1.5

3. Following each pressure setpoint bias change (2 to 10 psi), the time between the setpoint bias change and the occurrence of the pressure peak shall be 10 seconds or less. The peak neutron flux and peak vessel pressure should remain below scram setting by 7.5 percent and 10 psi, respectively for pressure setpoint changes 5 psid.

13.10.2.8 Feedwater System Purpose This test demonstrates that the components of the feedwater control system satisfactorily control reactor water level. Description To provide an observable feedwater system and reactor response, two methods are used to initiate test disturbances. The first is to change the input level setpoint about 4 to 6 inches. The second is to change one pump flow approximately 10 percent in the manual mode. The remaining pumps are in automatic mode to control reactor water level. Both one element and three element modes of level control will be tested. Tests are to be performed between 55 percent and 70 percent of rated power. Acceptance Criteria Level 1 The decay ratio must be less than 1 for each process variable that exhibits oscillatory response to feedwater system changes. 13.10-8

BFN-26 Level 2

1. The decay ratio of an oscillatory control loop mode must be 0.25 for each process variable that exhibits oscillatory response to feedwater system changes where the unit is operating above the lower limit setting of the master flow controller.
2. The transient response of each feedwater pump to a 10 percent flow demand input change, as measured by the turbine speed and flow recorder outputs shall be as follows:
a. Time to 10 percent of demand should be 1.1 seconds and must be less than or equal to 2.2 seconds.
b. Time from 10 percent to 90 percent of demand should be 1.9 seconds and must be less than or equal to 2.5 seconds.
c. Settling time to within 5 percent of the final value should be 14 seconds.
d. Peak overshoot should be equal to or less than 15 percent of demand.

13.10.2.9 Recirculation Motor-Generator (M-G) Set Control Purpose This test demonstrates that the recirculation speed control system can satisfactorily perform its function by comparing transient test results against system criteria. Description Pre-heatup tests are performed to test individual components and make other preparations for the tests at power. Once the unit has reached the 100 percent rated core flow point, several small step changes to M-G set speed will be made and applicable data recorded. Acceptance Criteria Level 1 The decay ratio must be less than 1.0 for each process variable that exhibits oscillatory response to recirculation system changes. 13.10-9

BFN-26 Level 2 When the unit is operating above the lower limit of the master manual limiter, the decay ratio of any oscillatory control loop mode must be 0.25 for each process variable that exhibits oscillatory response to recirculation system changes. 13.10.2.10 Drywell Atmosphere Cooling System Purpose This test verifies the ability of the drywell atmosphere cooling system to maintain design conditions in the drywell during operating conditions. Description The drywell atmosphere cooling system will be placed in operation and its ability to maintain normal operating temperatures inside the drywell is verified. For this test, 8 of 10 fans (and associated coils) are on, thereby demonstrating 25 percent standby heat removal capability. Acceptance Criteria Level 1 Not applicable. Level 2

1. Deleted.
2. Deleted.
3. The drywell cooling system shall maintain the bulk volumetric average temperature in the drywell below design values during normal operation.

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