ML20249C453

Final Tech Specs Pages for LARs 256/126

ML20249C453
Person / Time
Site:	Beaver Valley
Issue date:	06/23/1998
From:	DUQUESNE LIGHT CO.
To:
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ML20249C452	List: L-98-128, Forwards Typed,Final TS Pages for LARs 256/126.Summary Descriptions of Editorial Changes Are Also Provided ML20249C453
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NUDOCS 9806300012
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.

ATTACHMENT A-256 Unit NO.1 Technical Specification Pages i

l l'

l l

9906300012 900623 PDR ADOCK 05000334 P

PDR.

ATTACHMENT TO LICENSE AMENDMENT NO.

FACILITY OPERATING LICENSE NO. DPR-66 DOCKET NO. 50-334 l

Replace the following pages of Appendix A, Technical Specifications, with the enclosed pages as indicated.

The revised pages are identified by amendment number and contain vertical lines indicating the areas of change.

Remove Insert I

I 1-2 1-2 1-3 1-3 1-4 1-4 1-5 1-5

DPR-66 INDEX DEFINITIONS i

SECTIO}{

PAGE 1.0 DEFINITIONS Defined Terms.....................................

1-1 Thermal Power.....................................

1-1 Rated Thermal Power...............................

1-1 operational Mode..................................

1-1 Action............................................

1-1 l

Operable - operability............................

1-1 Reportable Event..................................

1-2 Containment Integrity.......................'......

1-2 l

Channel Calibration...............................

1-2 Channel Check.....................................

1-3 l

Channel Functional Test...........................

1-3 1

(

Core Alteration...................................

1-3 Shutdown Margin...................................

1-3 Leakage...........................................

1-3 I

Quadrant Power Tilt Ratio.........................

1-4 1

i Dose Equivalent I-131.............................

1-4 Staggered Test Basis..............................

1-4 Frequency Notation................................

1-5 l

Reactor Trip System Response Time.................

1-5 Engineered Safety Feature Response Time...........

1-5

-Axial Flux Difference.............................

1-5 j

l Physics Tests.....................................

1-5 E-Average Disintegration Energy...................

1-5 l

l BEAVER VALLEY - UNIT 1 I

Amendment No.

'DPR-66 DEFINITIONS REPORTABLE EVENT 1.7 A REPORTABLE EVENT shall be any of those conditions specified in Section 50.73 to 10 CFR Part 50.

CONTAINMENT INTEGRITY 1.8 CONTAINMENT INTEGRITY shall exist when:

1.8.1 All penetrations required to be closed during accident conditions are either:

l a.

Capable of being closed by an OPERABLE containment automatic isolation valve system, or b.

Closed by manual

valves, blind

flanges, or deactivated automatic valves secured in their closed positions, except for valves that are open under administrative control as permitted by Specification 3.6.3.1.

1.8.2 All equipment hatches are closed and sealed, t

1.8.3 Each air lock is in compliance with the requirements of Specification 3.6.1.3, 1.8.4 The containment leakage rates are within the limits of Specification 3.6.1.2, and 1.8.5 The sealing mechanism associated with each penetration (e.g., welds, bellows, or O-rings) is OPERABLE.

CHANNEL CALIBRATION 1.9 A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel output such that it responds with the necessary range and accuracy to known values of the parameter which the channel monitors.

The CHANNEL CALIBRATION shall encompass the entire channel, including the sensor and alarm and/or trip functions, and shall include the CHANNEL FUNCTIONAL TEST.

Calibration of instrument channels with resistance temperature detector (RTD) or thermocouple sensors may consist of an inplace qualitative assessment of sensor behavior and normal calibration of the remaining adjustable devices in the channel.

Whenever a sensing element is replaced, the next raquired CHANNEL CALIBRATION shall include an inplace cross calibration that compares the other sensing elements with the 4

recently installed sensing element.

The CHANNEL CALIBRATION may be

{

performed by any series of sequential,' overlapping or total channel i

steps such that the entire channel is calibrated.

i I

BEAVER VALLEY - UNIT 1 1-2 Amendment No.

I l

DPR-66 DEFINITIONS CHANNEL CHECK 1.10 A CHANNEL CHECK shall be the qualitative assessment of channel behavior during operation by observation.

This determination shall include, where possible, comparison of the channel indication and/or status with other indications and/or status derived from independent instrument channels measuring the same parameter.

CHANNEL FUNCTIONAL TEST 1.11 A

CHANNEL FUNCTIONAL TEST shall be the injection of a

simulated signal into the channel as close to the primary sensor as practicable to verify OPERABILITY including alarm and/or trip functions.

CORE ALTERATION 1.12 CORE ALTERATION shall be the movement or manipulation of any component within the reactor pressure vessel with the vessel head removed and fuel in the vessel.

Suspension of CORE ALTERATIONS shall not preclude completion of movement of a

component to a

safe conservative position.

SHUTDOWN MARGIN 1.13 SHUTDOWN MARGIN shall be the instantaneous amount of I

reactivity by which the reactor is or would be subcritical from its present condition assuming all full length rod cluster assemblies (shutdown and control) are fully inserted except for the single rod cluster assembly of highest reactivity worth which is assumed to be fully withdrawn.

LEAKAGE 1.14 LEAKAGE shall be:

a. Identified LEAKAGE 1.

LEAKAGE, such as that from pump seals or valve packing (except reactor coolant pump seal water injection or leakoff), that is captured and conducted to collection systems or a sump or collecting tank; 2.

LEAKAGE into the containment ' atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be Pressure Boundary LEAKAGE, or BEAVER VALLEY - UNIT 1 1-3 Amendment No.

1

' DPR-66 DEFINITIONS 3.

Reactor Coclant System LEAKAGE through a

steam j

generator to the secondary system.

(

b. Unidentified LEAKAGE Unidentified LEAKAGE shall be all LEAKAGE (except reactor coolant pump seal water injection or leakoff) that is not Identified LEAKAGE.

I

c. Pressure Boundary LEAKAGE I

Pressure Boundary LEAKAGE shall be LEAKAGE (except steam generator tube LEAKAGE) through a nonisolable fault in a Reactor Coolant System component body, pipe wall or vessel wall.

1.15 THROUGH 1.17 (DELETED)

OUADRANT POWER TILT RATIO (OPTR) 1.18 QPTR shall be the ratio of the maximum upper excore detector calibrated output to the average of the upper excore detector calibrated outputs, or the ratio of the maximum lower excore detector i

l calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater.

DOSE EOUIVALENT I-131 i

1.19 DOSE EQUIVALENT I-131 shall be that concentration of I-131

)

(microcuries/ gram) that alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present.

The DOSE EQUIVALENT I-131 is calculated with 1

l the following equation:

= C _131 +

+

+

+

C -131

.E.

I D

z 170 6

1000 34 i

Where "C"

is the concentration, in microcuries/ gram of the iodine l

isotopes.

This equation is based on dose conversion factors derived from ICRP-30.

l l

STAGGERED TEST BASIS 1.20 A STAGGERED TEST BASIS shall consist of:

l I

l

a. A test schedule for n systems, subsystems, trains or other designated components obtained by dividing the specified test interval into n equal subintervals; I

l BEAVER VALLEY - UNIT 1 1-4 Amendment No.

'DPR-66 DEFINITIONS

b. The testing of one (1) system, subsystem, train or other

, designated component at the beginning of each subinterval.

FREOUENCY NOTATION 1.21 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.2.

REACTOR TRIP SYSTEM RESPONSE TIME 1.22 The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its trip setpoint at the channel sensor until loss of stationary gripper coil voltage.

ENGINEERED SAFETY FEATURE RESPONSE TIME 1.23 The ENGINFERED SAFETY FEATURE RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ESF actuation setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e.,

the valves travel to their required positions, pump discharge pressures reach their required values, etc.).

Times shall include diesel generator starting and sequence loading delays where applicable.

{

AXIAL FLUX DIFFERENCE 1.24 AXIAL FLUX DIFFERENCE shall be the difference in normalized flux signals between the top and bottom halves of a two section excore neutron detector.

PHYSICS TESTS 1.25 PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related i

instrumentation and 1) described in Chapter 13.0 of the FSAR,

2) authorized under the provisions of 10 CFR 50.59, or

3) otherwise approved by the Commission.

E - AVERAGE DISINTEGRATION ENERGY 1.26 5 shall be the average sum (weighted in proportion to the j

concentration of each radionuclides in the reactor coolant at the time of sampling) of the average beta and gamma energies per disintegration (in MeV) for isotopes, other than iodines, with half lives greater than 15 minutes, making up at least 95% of the total non-iodine activity in the coolant.

l BEAVER VALLEY - UNIT 1 1-5 Amendment No.

ATTACHMENT A-126 Unit No. 2 Technical Specification Pages i

)

i ATTACHMENT TO LICENSE AMENDMENT NO.

FACILITY OPERATING LICENSE NO. NPF-73 l

DOCKET NO. 50-412 Replace the following pages of Appendix A, Technical Specifications, with the enclosed pages as indicated.

The revised pages are identified by amendment number and contain vertical lines indicating the areas of change.

Remove Insert I

I II II 1-2 1-2 1-3 1-3 1-4 1-4

'l 1-5 1-5 1

1-6 1-6 l

1-7 1-7 l

l 1

I l

L

NPF-73 INDEX DEFINITIONS l

SECTION PAGE 1.0 DEFINITIONS 1.1 DEFINED TERMS.....................................

1-1 1.2 THERMAL POWER...................

1-1 1.3 RATED THERMAL POWER...............................

1-1 1.4 OPERATIONAL MODE..................................

1-1 1.5 ACTION............................................

1-1 1.6 OPERABLE - OPERABILITY............................

1-1

)

1

/

l j

1.7 REPORTABLE EVENT..................................

1-1 l

1.8 CONTAINMENT INTEGRITY.............................

1-1 1.9 CHANNEL CALIBRATION...............................

1-2 1.10 CHANNEL CHECK.....................................

1-2 1.11 CHANNEL FUNCTIONAL TEST...........................

1-2 s

1.12 CORE ALTERATION...................................

1-3 l

1.13 SHUTDOWN MARGIN...................................

1-3 1.14 LEAKAGE...........................................

1-3 l

1.15 DELETED 1.16 DELETED 1.17

. DELETED t-1.18 QUADRANT POWER TILT RATIO.........................

1-4 l

1.19 DOSE EQUIVALENT I-131.............................

1-4 1.20 STAGGERED TEST BASIS..............................

1-4 l.

1.21 FREQUENCY NOTATION................................

1-4 1.22 REACTOR TRIP SYSTEM RESPONSE TIME.................

1-4 I

1.23' ENGINEERED SAFETY FEATURE RESPONSE TIME..........

1-4 i

I 1

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BEAVER VALLEY - UNIT 2 I

Amendment No.

NPF-73 INDEX DEFINITIONS SECTION PAGE 1.24 AXIAL FLUX DIFFERENCE.............................

1-5 l

1.25 PHYSICS TESTS.....................................

1-5 1.26 E-AVERAGE DISINTEGRATION ENERGY....................

1-5 1.27 SOURCE CHECK......................................

1-5 1.28 PROCESS CONTROL PROGRAM...........................

1-5 1.29 DELETED 1.30 OFFSITE DOSE CALCULATION MANUAL (ODCM)............

1-5 j

l 1.31 GASEOUS RADWASTE TREATMENT SYSTEM.................

1-6

)

1.32 VENTILATION EXHAUST TREATMENT SYSTEM..............

1-6 1.33 PURGE - PURGING...................................

1-6 1.34 VENTING...........................................

1-6

)

I 1.35 MAJOR CHANGES.....................................

1-6 i

1.36 MEMBER (S) OF THE PUBLIC...........................

1-7 1.37 CORE OPERATING LIMITS REPORT......................

1-7 l

TABLE 1.1 OPERATIONAL MODES............................

1-8 TABLE 1.2 FREQUENCY NOTATION...........................

1-9 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS i

l l ll l

EE.CTION PAGE 2.1 SAFETY LIMITS 2.~1.1 REACTOR CORE................................

2-1 2.1.2 REACTOR COOLANT SYSTEM PRESSURE.............

2-1 l

2.2 LIMITING SAFETY SYSTEM SETTINGS 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS...........................

2-3 l-BEAVER VALLEY - UNIT 2 II Amendment No.

lt-

l NPF-73 DEFINITIONS i

CONTAINMENT INTEGRITY (Continued) b.

Closed by manual

valves, blind

flanges, or deactivated automatic valves secured in their closed positions, except for valves that are open under administrative control as permitted by Specification 3.6.3.1.

1.8.2 All equipment hatches are closed and sealed, 1.8.3 Each air lock is in compliance with the requirements of Specification 3.6.1.3, 1.8.4 The containment leakage rates are within the limits of Specification 3.6.1.2, and 1.8.5 The sealing mechanism associated with each penetration (e.g., welds, bellows, or 0-rings) is OPERABLE.

CHANNEL CALIBRATLQH 1.9 A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel output such that it responds with the necessary range and accuracy to known values of the parameter which the channel monitors.

The CHANNEL CALIBRATION shall encompass the entire channel, including the sensor and alarm and/or trip functions, and shall include the CHANNEL FUNCTIONAL TEST.

Calibration of instrument channels with resistance temperature detector (RTD) or thermocouple sensors may consist of an inplace qualitative assessment of sensor behavior and normal calibration of the remaining adjustable devices in the channel.

Whenever a sensing element is replaced, the next required CHANNEL CALIBRATION shall include an inplace cross calibration that compares the other sensing elements with the recently installed sensing element.

The CHANNEL CALIBRATION may be performed by any series of sequential, overlapping, or total channel steps such that the entire channel is calibrated.

CHANNEL CHECK 1.10 A CHANNEL CHECK shall be the qualitative assessment of channel behavior during operation by observation.

This determination shall include, where possible, comparison of the channel indication and/or status with other indications and/or status derived from independent instrument channels measuring the same parameter.

CHANNEL FUNCTIONAL TEST 1.11 A

CHANNEL FUNCTIONAL TEST shall be the injection of a

simulated signal into the channel as close to the primary sensor as practicable to verify OPERABILITY including alarm and/or trip functions.

BEAVER VALLEY - UNIT 2 1-2 Amendment No.

l NPF-73 DEFINITIONS CORE ALTERATION 1.12' CORE ALTERATION shall be the movement or manipulation of any component within the reactor pressure vessel with the vessel head removed and fuel in the vessel.

Suspension of CORE ALTERATIONS shall not preclude completion of movement of a

component to a

safe conservative position.

SHUTDOWN MARGIN 1.13 SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which the reactor is or would be subcritical from its present condition assuming all full length rod cluster assemblies (shutdown and control) are fully inserted except for the single rod cluster assembly of highest reactivity worth which is assumed to be fully withdrawn.

LEAKAGE 1.14 LEAKAGE shall be:

a. Identified LEAKAGE 1.

LEAKAGE, such as that from pump seals or valve packing (except reactor coolant pump seal water injection or leakoff), that is captured and conducted to collection systems or a sump or collecting tank; 2.

LEAKAGE into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be Pressure Boundary LEAKAGE, or 3.

Reactor Coolant System LEAKAGE through a

steam generator to the secondary system.

b. Unidentified LEAKAGE Unidentified LEAKAGE shall be all LEAKAGE (except reactor coolant pump seal water injection or leakoff) that is not Identified LEAKAGE.

c. Pressure Boundary LEAKAGE Pressure Boundary LEAKAGE shall be LEAKAGE (except steam generator tube LEAKAGE) through a nonisolable fault in a Reactor Coolant System component body, pipe wall or vessel wall.

BEAVER VALLEY - UNIT 2 1-3 Amendment No.

l.

' NPF-73 DEFINITIONS 1.15 THROUGH 1.17 (DELETED)

OUAdRANT POWER TILT RATIO (OPTR) 1.18 QPTR shall be the ratio of the maximum upper excore detector calibrated output to the average of the upper excore detector calibrated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater.

I DOSE EQUIVALENT I-131 1.19 DOSE EQUIVALENT I-131 shall be that concentration of I-131 i

(pci/ gram) which alone would produce the same thyroid dose as the quantity and isotJpic mixture of I-131, I-132, I-133, I-134, and I-135 actually present.

The thyroid dose conversion factors used for this calculation shall be those listed in Regulatory Guide 1.109, 1977 or TID 14844.

STAGGERED TEST BASIS 1.20 A STAGGERED TEST BASIS shall consist of:

a. A test schedule for n systems, subsystems, trains or other designated components obtained by dividing the specified test interval into n equal subintervals;

b. The testing of one (1) system, subsystem, train or other designated component at the beginning of each subinterval.

FREOUENCY__ NOTATION 1.21 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.2.

REACTOR TRIP SYSTEM RESPONSE TIME 1.22 The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its trip setpoint at the channel sensor until loss of stationary gripper coil voltage.

ENGINEERED SAFETY FEATURE RESPONSE TIME 1.23 The ENGINEERED SAFETY FEATURE RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ESF actuation setpoint at the channel sensor until the ESF equipment is capable of performing its safety.tunction (i.e.,

the valves travel to their required positions, pump discharge pressures reach their required

values, etc.).

Times shall include diesel generator starting and sequence loading delays where applicable.

BEAVER VALLEY - UNIT 2 1-4 Amendment No.

NPF-73 DEFINITIONS AXIAL FLUX DIFFERENCE 1.24' AXIAL FLUX DIFFERENCE shall be the difference in normalized flux signals between the top and bottom halves of a two-section excore neutron detector.

PHYSICS TESTS 1.25 PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation and 1) described in Chapter 14.0 of the FSAR,

2) authorized under the provisions of 10 CFR 50.59, or

3) otherwise approved by the commission.

E - AVERAGE DISINTEGRATION ENERGY 1.26 E shall be the average sum (weighted in proportion to the concentration of each radionuclides in the reactor coolant at the time of sampling) of the average beta and gamma energies per disintegration (in MeV) for isotopes, other than iodines, with half lives greater than 15 minutes, making up at least 95% of the total non-iodine activity in the coolant.

SOURCE CHECK 1.27 A SOURCE CHECK shall be the qualitative assessment of channel

{

response when the channel sensor is exposed to a radioactive source.

PROCESS CONTROL PROGRAM 1.28 The PROCESS CONTROL PROGRAM (PCP) shall contain the current j

formulas, sampling, analyses, test, and determinations to be made to ensure that processing and packaging of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure compliance with 10 CFR Parts 20, 61, and 71, State regulations, burial ground requirements, and other requirements governing the disposal of solid l

radioactive waste.

1.29 DELETED l

OFFSITE DOSE CALCULATION MANUAL (ODCML l

l 1.30 The OFFSITE DOSE CALCULATION MANUAL (ODCM) shall contain the l

methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring Alarm / Trip Setpoints, and in the conduct of the Environmental Radiological Monitoring Program.

The ODCM shall also contain (1) the Radioactive

{

BEAVER VALLEY - UNIT 2 1-5 Amendment No.

l

NPF-73 DEFINITIONS OFFSITE DOSE CALCULATION MANUAL (ODCM) (Continued)

Effl'uent Controls and Radiological Environmental Monitoring Programs required by Section 6.8.6 and (2) descriptions of the information that should be included in the Annual Radiological Environmental Operating and-Annual Radioactive Effluent Release Reports required by Specifications 6.9.1.10 and 6.9.1.11.

GASEOUS RADWASTE TREATMENT SYSTEM 1.31 A GASEOUS RADWASTE TREATMENT SYSTEM is any system designed and installed to reduce radioactive gaseous effluents by collecting Primary Coolant System offgases from the primary system and providing l

for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment.

VENTILATION EXHAUST TREATMENT SYSTEM 1.32 VENTILATION EXHAUST TREATMENT SYSTEM is any system designed and installed to reduce gaseous radiciodine or radioactive material in particulate form in effluents by passing ventilation or vent exhaust gases through charcoal adsorbers and/or HEPA filters for the purpose of removing iodines or particulate from the gaseous exhaust stream prior to the release to the environment (such a system is not considered to have any effect on noble gas effluents).

Engineered Safety Feature (ESF) atmospheric cleanup systems are not considered to be VENTILATION EXHAUST TREATMENT SYSTEM components.

PURGE-PURGING 1.33 PURGE or PURGING is the controlled process of discharging air or gas from a

confinement to maintain temperature,

pressure, humidity, concentration or other operating conditions, in such a manner that replacement air or gas is required to purify the confinement.

VENTING 1.34 VENTING is the controlled process of discharging air or gas from a confinement to maintain temperature,

pressure, humidity, concentration or other operating conditions, in such a manner that replacement air or gas is not provided or required during VENTING.

Vent, used in system names, does not imply a VENTING process.

MAJOR CHANGES 1.35 MAJOR CHANGES to radioactive waste systems (liquid, gaseous and solid),

as addressed in the PROCESS CONTROL PROGRAM, shall include the following:

BEAVER VALLEY - UNIT 2 1-6 Amendment No.

NPF-73 DEFINITIONS MAJOR CHANGES (Continued) 1)' MAJOR CHANGES in process equipment, components, structures, l

and effluent monitoring instrumentation from those described in the Final Safety Analysis Report (FSAR) or the Hazards Summary Report and evaluated in the staff's Safety Evaluation Report (SER)

(e.g.,

deletion of evaporators and installation of demineralizers; use of fluidized bed calciner/ incineration in place of cement solidification systems);

2) MAJOR CHANGES in the design of radwaste treatment systems l

(liquid,

gaseous, and solid) that could significantly increase the quantitlen or activity of effluents released or volumes of solid waste stored or shipped offsite from those previously considered in the FSAR and SER (e.g.,

use of asphalt system in place of cement);

3) Changes in system design which may invalidate the accident analysis as described in the SER (e.g.,

changes in tank capacity that would alter the curies released); and

4) Changes in system design that could potentially result in a significant increase in occupational exposure of operating personnel (e.g.,

use of temporary equipment without adequate shielding provisions).

MEMBER (S) OF THE PUBLIC 1.36 MEMBER (S) OF THE PUBLIC shall include all persons who are not occupationally associated with the plant.

This category does not include employees of the utility, its contractors, or its vendors.

Also excluded from this category are persons who enter the site to service equipment or to make deliveries and persons who traverse portions of the site as the consequence of a public highway, railway, or waterway located within the confines of the site boundary.

This category does include persons who use portions of the site for recreational, occupational, or other purposes not associated with the plant.

CORE OPERATING LIMITS REPORT 1.37 The CORE OPERATING LIMITS REPORT (COLR) is the unit-specific document that provides core operating limits for the current operating reload cycle.

These cycle-specific core operating limits shall be determined for each reload cycle in accordance with Specification 6.9.1.12.

Plant operation within these operating limits is addressed in individual specifications.

BEAVER VALLEY - UNIT 2 1-7 Amendment No.

t

l ATTACHMENT B l

l l

Change Summary i

License Amendment Unit Reauest Tvoed Pages Editorial Channes 1

256 I

The index was changed to revise page numbers.

i The page numbers were changed due to the Channel Calibration definition change which shifted definitions to following pages.

1-2 Added a comma in the second sentence of the Channel Calibration definition.

1-3, 1-4, 1-5 These pages were added due to the Channel Calibration definition change which shifted definitions to following pages. Definition 1.14.a.3 capitalized the first letter of" Coolant l

System."

2 126 I, II The index was changed to revise page numbers.

The page munbers were changed due to the Channel Calibration definition change which shifted definitions to following pages.

I l-2 Added a comma in the second sentence of the Channel Calibration definition.

l-3, 1-4, 1-5, These pages were added due to the Channel 1-6,1-7 Calibration definition change which shifted definitions to following pages.

Defm' ition 1.31 capitalized the first letter of

" Primary Coolant System."

Definition 1.35.1) and 2) capitalized " Major Changes."

L-- ----- ---------------