NUREG- 1123

Rev. 2, Supp. 1

Knowledge and Abilities

Catalog for Nuclear Power

Plant Operators

Boiling Water Reactors

Draft Report for Comment

U.S. Nuclear Regulatory Commission

Office of Nuclear Reactor Regulation

Washington, DC 20555-0001
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NUREG- 1123

Rev. 2, Supp. 1

Knowledge and Abilities

Catalog for Nuclear Power

Plant Operators

Boiling Water Reactors

Draft Report for Comment

Manuscript Completed: April 2007

Date Published: May 2007

Division of Inspection and Regional Support

Office of Nuclear Reactor Regulation

U.S. Nuclear Regulatory Commission

Washington, DC 20555-0001
COMMENTS ON DRAFT REPORT

Any interested party may submit comments on this report for consideration by the NRC staff.

Comments may be accompanied by additional relevant information or supporting data. Please

specify the report number, NUREG-1 123, Revision 2, Draft Supplement 1, in your comments,

and send them within 60 days to the following address:

Chief, Rules, Directives, and Editing Branch

U.S. Nuclear Regulatory Commission

Mail Stop T6-D59

Washington, DC 20555-0001

Comments may also be submitted electronically via the NRC's Public Electronic Reading Room

at http:l/www.nrc.gov/public-involve/doc-commentlform.html.

For any questions about the material in this report, please contact:

David S. Muller

Mail Stop 06F2

U.S. Nuclear Regulatory Commission

Washington, DC 20555-0001

Phone: 301-415-1412

iE-mail: dsm3@nrc.gov
ABSTRACT

The Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Boiling Water Reactors

(BWRs) (NUREG-1 123, Revision 2, Supplement 1) provides the basis for the development of

content-valid licensing examinations for reactor operators (ROs) and senior reactor operators

(SROs). The examinations developed using the BWR Catalog along with the Operator Licensing

Examination Standards for Power Reactors (NUREG-1021, Rev. 8 9, Supplement 1) will sample

the topics listed under Title 10, Code of Federal Regulations, Part 55 (10 CFR 55).

The BWR Catalog contains approximately 7,000 knowledge and ability (K/A) statements for ROs

and SROs at BWRs. The catalog is organized into six major sections: Organization of the

Catalog, Generic Knowledge and Ability Statements, Plant Systems grouped by Safety Functions,

Emergency and Abnormal Plant Evolutions, Components and Theory.

Revision 1 to the BWR Catalog modified the form and content of the original catalog. The K/As

were linked to their applicable 10 CFR 55 item numbers. SRO level K/As were identified by 10

CFR 55.43 item numbers. The plant-wide generic and system generic K/As were combined in one

section with approximately one hundred new K/As. Component Cooling Water and Instrument Air

Systems were added to the Systems Section. Finally, High Containment Hydrogen Concentration

and Plant Fire On Site evolutions were added to the Emergency and Abnormal Plant Evolutions

section.

Revision 2 incorporatesd corrections to the Rev. I catalog that were identified during a pilot testing

program associated with revision of 10 CFR 55 and implementation of NUREG- 1021, Interim

Rev. 8, ."Operator Licensing Examination Standards for Power Reactors." Corrections to the

catalog included:

1. addition of 7 K/As that had been omitted in Rev. 1.

2. correction of typographical errors.

3. addition of importance value modifiers that had been omitted in Rev. 1.

Revision 2, Supplement 11, includes the following changes:

1. total replacement of Section 2, "Generic Knowledges and Abilities"

2. inclusion of one additional Abnormal Plant Evolution (APE), 700000, "Generator Voltage

and Electric Grid Disturbances," in Section 4.2 "Generic Abnormal Plant Evolutions"

Corrections and additions are identified by "redline" marking in the margins.

iii NUREG-1 123, Rev. 2, Supp. I
Draft" Suppement 1 to .Revision2, of -NUREGl- 1123

Record of 'Changes

Instructions::Remove the existih NUREGO--1 123,:Revision 2, pages and insert replacement pages

as noted beloW (corrections and clarificatiois haveb. een4made as indicated).,

'Section Remove Page(s) Re~placement Page(s) /Correction Summary,

Cover page Cover page Cover page / add Supp. I

Abstract iii-biank iui-Kiv/eplaceSection 2, add one additional•APE

______________(700.000),and add Record of Changes

Table •of Conents v-iiv-viii addaSupplement •.,page number changes,

add new APE (700000)for Generator Voltage

and Electric Grid Disturbances to SectionA4.2

mmary of ii-XI~i xiixiil describe the replacemnent of Section 2

SipificntCanges in xv discuss the study basis to reassess jobsfafety

and impoance values

Organization of, 1-1 to 1-6, 1 -1 tol 1-6/ amiend to conform with Revision 9 of

,Catalog NUREo-102 and Section 2 changes

1-9 to 1- 12 1-9 to 1- 12 / add newAPE (700000) Generator

Voltage and Electric Grid Disturbances.

2.0 Generic 2-1 to 2- 15 2-1 to 2-18/ revise< Section 2

'Abilities_________________________

4.2 rGeneric 4.2-1 4.2-1 add new APE (76otooo)for 'G"eneratorF

Abnormal Pl1ant Voltage and Electric Grid Disturbances- Eviolutions :,(APEs)

4.2-43 and blank 4.2-43 to 4.2-45 / add new APE 70.000

____________________________%Generator Voltage and Electric Grid Disturbances

NU EJ 1t2`3,I~ 2,iSupjIl -1iv
TABLE OF CONTENTS

page

ABSTRACT iii

SUPPLEMENT 1,- TO REVISION:20F1 NUREG-1 123 RECORD OF iv

CHANGES

SUMMARY OF SIGNIFICANT CHANGES xi

1 ORGANIZATION OF THE CATALOG

1.1 Introduction 1-1

1.2 Part 55 of Title 10 of the Code of Federal Regulations 1-1

1.3 RO Written Examination 1-1

1.4 SRO Written Examination 1-2

1.5 RO and SRO Operating Test Items 1-3

1.6 Senior Operators Limited to Fuel Handling 1-4

1.7 Organization of the BWR Catalog 1-4

1.8 Generic Knowledge and Abilities 1-5

1.9 Plant Systems 1-6

1.10 Emergency And Abnormal Plant Evolutions 1-10

1.11 Components 1-12

1.12 Theory 1-13

1.13 Importance Ratings 1-13

1.14 Acronyms And Terms 1-14

2 GENERIC KNOWLEDGES AND ABILITIES

2.1 Conduct Of Operations 2-1

2.2 Equipment Control 2-6.

2.3 Radiation Control 2-11t

2.4 Emergency Procedures / Plan 2-13

3 PLANT SYSTEMS

3.1 Reactivity Control 3.1-1

Control Rod Drive Hydraulic System 3.1-3

Control Rod and Drive Mechanism 3.1-7

Reactor Manual Control System 3.1-10

Recirculation Flow Control System 3.1-13

Recirculation System 3.1-17

Rod Control and Information System 3. -22

Standby Liquid Control System 3.1-25

v NUREG-I1123, Rev. 2, Sup.•ji
TABLE OF CONTENTS (Continued) page

3.2 Reactor Water Inventory Control 3.2-1

High Pressure Coolant Injection System 3.2-3

High Pressure Core Spray System 3.2-8

Low Pressure Core Spray System 3.2-12

Reactor Condensate System 3.2-16

Reactor Core Isolation Cooling System 3.2-21

Reactor Feedwater System 3.2-25

Reactor Water Cleanup System 3.2-29

Reactor Water Level Control System 3.2-33

Residual Heat Removal /Low Pressure Coolant Injection: Injection Mode 3.2-38

(Plant Specific)

3.3 Reactor Pressure Control 3.3-1

Automatic Depressurization System 3.3-3

Main and Reheat Steam System 3.3-6

Reactor/Turbine Pressure Regulating System 3.3-11

Relief/Safety Valves 3.3-18

3. 4 Heat Removal From Reactor Core 3.4-1

High Pressure Coolant Injection System 3.4-3

High Pressure Core Spray System 3.4-8

Isolation (Emergency) Condenser 3.4-12

Low Pressure Core Spray System 3.4-16

Main and Reheat Steam System 3.4-20

Main Turbine Generator and Auxiliary Systems 3.4-25

Reactor Core Isolation Cooling System 3.4-29

Recirculation System 3.4-33

Residual Heat Removal /Low Pressure Coolant Injection: Injection Mode 3.4-38

Shutdown Cooling System (RHR Shutdown Cooling Mode) 3.4-42

3.5 Containment Integrity 3.5-1

Primary Containment System and Auxiliaries 3.5-3

Primary Containment Isolation System /Nuclear Steam Supply Shut-Off 3.5-8

Reactor Vessel Internals 3.5-12

RHR/LPCI: Torus/Suppression Pool Cooling Mode 3.5-15

RHR/LPCI: Containment Spray System Mode 3.5-19

RHR/LPCI: Torus/Suppression Pool Spray Mode 3.5-23

Secondary Containment 3.5-28

3.6 Electrical 3.6-1

A.C. Electrical Distribution 3.6-3

D.C. Electrical Distribution 3.6-6

Emergency Generators (Diese/Jet) 3.6-9

NUREG-I1123, Rev. 2
TABLE OF CONTENTS (Continued) page

Uninterruptable Power Supply (A.C. /D.C.) 3.6-13

3.7 Instrumentation 3.7-1

Average Power Range Monitor/Local Power Range Monitor System 3.7-3

Intermediate Range Monitor System 3.7-7

Nuclear Boiler Instrumentation 3.7-10

Radiation Monitoring System 3.7-15

Reactor Protection System 3.7-19

Rod Block Monitor System 3.7-24

Rod Control and Information System 3.7-27

Rod Position Information System 3.7-30

Rod Sequence Control System (Plant Specific) 3.7-33

Rod Worth Minimizer System (Plant Specific) 3.7-35

Source Range Monitor System 3.7-38 Traversing In-Core Probe 3.7-41

3.8 Plant Service Systems 3.8-1

Fire Protection System 3.8-3

Fuel Handling. 3.8-7

Instrument Air System 3.8-10

Component Cooling Water System 3.8-13

3.9 Radioactivity Release 3.9-1

Main Steam Isolation Valve Leakage Control System 3.9-3

Offgas System 3.9-7

Plant Ventilation Systems 3.9-11

Radiation Monitoring System 3.9-14

Radwaste 3.9-18

Reactor Vessel Internals 3.9-21

Fuel Pool Cooling and Clean-up 3.9-24

Standby Gas Treatment System 3.9-29

Control Room Heating, Ventilation and Air Conditioning 3.9-33

4 EMERGENCY AND ABNORMAL PLANT EVOLUTIONS

4.1 Generic Emergency Plant Evolutions 4.1-1

High Drywell Pressure 4.1-3

High Reactor Pressure 4.1-5

Suppression Pool High Water Temperature 4.1-7

High Containment Temperature 4.1-8

High Drywell Temperature 4.1-9

High Suppression Pool Water Level 4.1-11

vii NUREG- 1123, Rev. 2
TABLE OF CONTENTS (Continued) page

Low Suppression Pool Water Level 4.1-12

Reactor Low Water Level 4.1-14

High Secondary Containment Area Temperature 4.1-16

High Secondary Containment Area Radiation Levels 4.1-18

Secondary Containment Ventilation High Radiation 4.1-20

Secondary Containment High Differential Pressure 4.1-22

Secondary Containment High Sump / Area Water Level 4.1-23

SCRAM Condition Present and Reactor Power Above APRM Downscale 4.1-24

or unknown

High Off-Site Release Rate 4.1-26

High Containment Hydrogen Concentration. 4.1-28

4.2 Generic Abnormal Plant Evolutions 4.2-1

Partial or Complete Loss of Forced Core Flow Circulation 4.2-3

Loss of Main Condenser Vacuum 4.2-5

Partial or Complete Loss of A.C. Power 4.2-7

Partial or Complete Loss of D.C. Power 4.2-9

Main Turbine Generator Trip 4.2-10

SCRAM 4.2-12

High Reactor Pressure 4.2-14

High Reactor Water Level 4.2-16

Low Reactor Water Level 4.2-18

High Drywell Pressure 4.2-19

High Containment Temperature (Mark III Containment Only) 4.2-21

High Drywell Temperature 4.2-22

High Suppression Pool Temperature 4.2-23

Inadvertent Reactivity Addition 4.2-24

Incomplete SCRAM 4.2-26

Control Room Abandonment 4.2-28

High Off-Site Release Rate 4.2-29

Partial or Complete Loss of Component Cooling Water 4.2-31

Partial or Complete Loss of Instrument Air 4.2-33

Inadvertent Containment Isolation 4.2-35

Loss of Shutdown Cooling 4.2-37

Loss of Control Rod Drive Pumps 4.2-39

Refueling Accidents 4.2-40

Plant Fire On Site 4.2-42

Generator Voltage and Electric Grid Disturbances 4.2-44

NUREG-I 123, Rev. 2, Supp., Ii viii
SUMMARY OF SIGNIFICANT CHANGES

The changes described in paragraphs 1 through 6, below, were incorporated in Revision 1

of the catalog in August, 1995. Paragraph 7 describes changes that are incorporated with Revision 2 of the catalog. Paragraph 8 describes Supplement toRevisi0on2, ;1 which

incdudesa replacemIent of Section 2, "Generic Knowledges andAbilities,"'and a new

abnormal' pl "ant. evolution (APE).

1 ORGANIZATION OF THE BWR CATALOG

1. 1 The 10 CFR 55 items listed.

The content of the written examinations and operating licensing tests is dictated by Sections

55.41, 55.43, and 55.45 of Title 10 of the Code of Federal Regulations (10 CFR). The

thirty four (34) items listed in 10 CFR 55 are listed in the catalog to reduce the need for

cross referencing.

1.2 Knowledge and Abilities linked to 10 CFR 55

The linkage of K/As to the 10 CFR 55.41, 43 and 45 requirements was done to help ensure

that the examinations include a representative sample from among the applicable items.

Throughout the catalog, 10 CFR 55 section references are shown in parentheses following

the appropriate K/A statement, such as (CFR: 41 .x / 43.x / 45.x).

1.3 Senior Reactor Operator (SRO) K/As identified

NUREG- 1021, Rev. 8, "Operator Licensing Examination Standards for Power Reactors,"

Section ES-40 1, requiresd at least 25% of the site-specific written examination for SROs to

evaluate K/As required for the higher license level. The original catalogs did not explicitly

identify the K/As that represented the higher license level. Differences in RO and SRO

importance ratings were sometimes used, but, the rating differences were not linked to the 10

CFR 55.43 SRO items. In this catalog revision, SRO license level K/As were linked to the

items associated with the 10 CFR 55.43 SRO items. This is intended to remove subjectivity

from selection of higher license level K/As.

1.4 Senior Reactor Operator Limited to Fuel Handling (LSRO) examination

specifications added.

NUREG- 1021, Rev. 8 , Section 701 refersred to the K/A catalog. In an effort to assure

consistency between the Examination Standards and the catalog, a brief discussion of the use

of the catalog for LSRO examinations was included.

xi NUREG- 1123, Rev. 2, Supp.. I
1.5 An updated catalog organization was implemented.

1. ORGANIZATION OF BWR CATALOG

I. Added 10 CFR information

2 Generic Knowledge and Abilities

1. Combined old plant-wide and system generic K/As in one section.

2. Eliminated duplicate generic K/As.

3. Organized the section into 4 topic areas.

4. Added about 100 new generic K/As.

3 Plant Systems

1. Moved System Generic K/As to new generic section

2. Added Instrument Air System

3. Added Component Cooling Water System

4 Emergency & Abnormal Plant Evolution Section

1. Moved Generic K/As to Section 2

2. Added High Containment Hydrogen Concentration

3. Added Plant Fire On Site

5 Components

No change

6 Theory Section

No change

NUREG-1 123, Rev. 2 xii
4.4 Emergency and Abnormal Evolution K/As were linked to 10 CFR 55 item numbers

at the stem statement level.

5 COMPONENTS

5.1 All component K/As were linked to 10 CFR 55 item numbers.

6 THEORY

6.1 Reactor Theory and Thermodynamics theory K/As were linked to 10 CFR, 55 item

numbers.

7 REVISION 2 CHANGES

7.1 Approximately 7 K/As that had been omitted in Rev. I were added.

7.2 Typographical errors were corrected.

7.3 Importance value modifiers that had been omitted in Rev. I were added.

Corrections and additions are identified by "redline" marking in the margins.

8 , CHANGES IN SUTPPLEMENT 1 TO REVISION-2-

8.1 Regardinig replacement of Seto ,te rsuie Water Reactor ~Owners G1roup,

with the endorsement of the U.S. NuclearRegulatory Commission (NRC), first

intitdu athe iompgoilassssci intite astudy to ess~ess tejob relevance and imotne-to-safety ratings of 129

K/As in Section 2 of NUREG-1 122. The stud~y was subsequentfly expanded to

include the Boiling Watver ReactorOwners Grouip toreassess its counterpart Section

2 of NUREG-1 1123.

8.2 Regarding the additional APE 700000, "Generator Voltage and ElectricnGrid

Disturbances," the NR§ requestedthat additional K/As related to degraded voltage

beincorporated and validated during the study's secon phase to address lessons

learnied fr~oma major blackout in August 2003.~

X .v NUREG-1 123, Rev. 2,Sipi
Intentionally blank
* 1 ORGANIZATION OF THE

CATALOG

1.1 INTRODUCTION

The Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Boiling Water

Reactors (BWR) NUREG-1 123, Revision 2, Supplement I provides the basis for

development of content-valid written and operating licensing examinations for reactor

operators (ROs) and senior reactor operators (SROs). The Catalog is designed to ensure

equitable and consistent examinations.

1.2 PART 55 OF TITLE 10 OF THE CODE OF FEDERAL REGULATIONS

The catalog is used in conjunction with NUREG- 1021, Revision 9, Suppleme I ,

"Operator Licensing Examination Standards for Power Reactors." NUREG- 1021 provides

policy and guidance and establishes the procedures and practices for examining licensees and

applicants for RO and SRO licenses pursuant to Part 55 of Title 10 of the Code of Federal

Regulations (10 CFR 55). All knowledge and abilities (K/As) in this catalog are directly

linked by item number to I 0 CFR 55.

1.3 RO WRITTEN EXAMINATION

.The items for RO written examinations are specified in 10 CFR 55.41 (b). The RO written

examination questions should be generated from a representative sample of K/As derived

from among the 10 CFR 55.41 (b) items listed below:

(1) Fundamentals of reactor theory, including fission process, neutron multiplication,

soui'ce effects, control rod effects, criticality indications, reactivity coefficients, and

poison effects.

(2) General design features of the core, including core structure, fuel elements, control

rods, core instrumentation, and coolant flow.

(3) Mechanical components and design features of reactor primary system.

(4) Secondary coolant and auxiliary systems that affect the facility.

1-1 NUREG-1 123, Rev. 2, Supp. II
(5) Facility operating characteristics during steady state and transient conditions, including

coolant chemistry, causes and effects of temperature, pressure and reactivity changes,

effects of load changes, and operating limitations and reasons for these operating

characteristics.

(6) Design, components, and function of reactivity control mechanisms and instrumentation.

(7) Design, components, and function of control and safety systems, including

instrumentation, signals, interlocks, failure modes, and automatic and manual features.

(8) Components, capacity, and functions of emergency systems.

(9) Shielding, isolation, and containment design features, including access limitations.

(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.

(11) Purpose and operation of radiation monitoring systems, including alarms and survey

equipment..;

(12) Radiological safety principles and procedures.

(13) Procedures and equipment available for handling and disposal of radioactive materials

and effluents.

(14) Principals of heat transfer, thermodynamics and fluid mechanics.

The RO written examination is administered in two sections, a generic fundamentals

examination (GFE) section and a site-specific examination. The GFE covers those knowledge

items that do not vary significantly among reactors of the same type (refer to NUREG- 1021,

ES-205). The GFE covers components, reactor theory, and thermodynamics knowledge. The

component knowledge items are derived from 10 CFR 55.41 (b) items 3 and 7. Reactor theory

,knowledge items are derived from 10 CFR 55.41 (b) 1. Thermodynamic knowledge items are

derived from 10 CFR 55.41 (b) 14.

The site-specific RO written examination covers K/As that vary among reactors of the same

type. The guidance for preparation of RO written examination is presented in NUREG-102 1,

ES-40 1. The RO examination includes a balanced mix of generic K/As, plant systems K/As,

and emergency/abnormal evolution K/As. The K/As associated with the RO site-specific written

examinations are derived from 10 CFR 55.41 (b) items 2 through 13.

1.4 SRO WRITTEN EXAMINATION

The items for SRO written examinations are presented in 10 CFR 55.43(b). The guidance for

NUREG-1123, Rev. 2 1-2
preparation of the SRO written examination is presented in NUREG- 1021, ES-40 1. The

examination for SRO• shottld includes at-least-25 questions based on the higher license level

K/As from the 7 items listed under 10 CFR 55.43(b), and No more than- 75 questions i.pereen+

(75•)•-of, -SR-•• he , As may-be-derived from the 10 CFR 55.41 (b) RO K/As. The 7 SRO

items listed under 10 CFR 55.43(b) include:

(1) Conditions and limitations in the facility license.

(2) Facility operating limitations in the technical specifications and their bases.

(3) Facility licensee procedures required to obtain authority for design and operating changes

in the facility.

(4) Radiation hazards that may arise during normal and abnormal situations, including

maintenance activities and various contamination conditions.

(5) Assessment of facility conditions and selection of appropriate procedures during normal,

abnormal, and emergency situations.

(6) Procedures and limitations involved in initial core loading, alterations in core

configuration, control rod programming, and determination of various internal and

external effects on core reactivity.

(7) Fuel handling faciiti esand procedures.

1.5 RO AND SRO OPERATING TEST ITEMS

The items for operating tests for ROs and SROs are presented in 10 CFR 55.45(a). The

guidance for preparation of the operating tests is presented in NUREG- 102 1, ES-30 1. The

operating test should include a representative selection of K/As derived from 13 items listed in

10 CFR 55.45(a). The 13 items listed in 10 CFR 55.45(a) are:

(1) Perform pre-startup procedures for the facility, including operating of those controls

associated with plant equipment that could affect reactivity.

(2) Manipulate the console controls as required to operate the facility between shutdown and

designated power levels.

(3) Identify annunciators and condition-indicating signals and perform appropriate remedial

actions where appropriate.

(4) Identify the instrumentation systems and the significance of facility instrument readings.

(5) Observe and safely control the operating behavior characteristics of the facility.

1-3 NUREG-1 123, Rev. 2, Supp.!i
(6) Perform control manipulations required to obtain desired operating results during

normal, abnormal, and emergency situations.

(7) Safely operate the facility's heat removal systems, including primary coolant, emergency

coolant, and decay heat removal systems, and identify the relations of proper. operation of

these systems to the operation of the facility.

(8) Safety operate the facility's auxiliary and emergency systems, including operation of those

controls associated with plant equipment that could affect reactivity or the release of

radioactive materials to the environment

(9) Demonstrate or describe the use and function of the facility's radiation monitoring

systems, including fixed radiation monitors and alarms, portable survey instruments, and

personnel monitoring equipment.

(10) Demonstrate a knowledge of significant radiation hazards, including permissible levels in

excess of those authorized, and ability to perform other procedures to reduce excessive

levels of radiation and to guard against personnel exposure.

(11) Demonstrate knowledge of the emergency plan for the facility, including, as appropriate,

the operator's or senior operator's responsibility to decide when the plan should be

executed and the duties under the plan assigned.

,(12) Demonstrate the knowledge and ability as appropriate to the assigned position to assume

the responsibilities associated with the safe operation of the facility.

(13) Demonstrate the applicant's ability to function within the control room team as

appropriate to the assigned position, in such a way that the facility licensee's procedures

are adhered to and that the limitations in its license and amendments are not violated.

1.6 SENIOR OPERATORS LIMITED TO FUEL HANDLING

The specifications for examinations for Senior Operators Limited to Fuel Handling (LSRO) are

provided in Examination Standard, NUREG 1021, Section ES-70 1. The LSRO examination

process includes both a written examination and an operating test. This examination and test

include, but are not limited to, items associated with 10 CPR 55.43(b) items 5 through 7, and

10 CFR 55.45(a) items 5 and 6.

1.7 ORGANIZATION OF THE BWR CATALOG

The Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Boiling Water

Reactors is organized into 6 major sections. Knowledge and ability statements (K/As) are

grouped according to the major section to which they pertain. This organization is shown

schematically below.

NUREG- 1123, Rev. 2 1-4
1 ORGANIZATION OF THE BWR CATALOG

2 GENERIC KNOWLEDGE AND ABILITIES (120)

Conduct of Operations K/As

Equipment Control K/As

Radiation Control K/As

Emergency Procedures / Plan K/As

3 PLANT SYSTEMS (56) WITHIN 9 SAFETY FUNCTIONS

Knowledge Categories (KI - K6)

Ability Categories (Al - A4)

4 EMERGENCY & ABNORMAL PLANT EVOLUTIONS (4 1)

Knowledge Categories (E/A K 1 - E/A K3)

Ability Categories (E/A A I - E/A - A2)

5 COMPONENTS

Component Knowledge Categories (8)

6 THEORY

Reactor Theory Knowledge Categories (8)

Thermodynamics Knowledge Categories (10)

1.8 GENERIC KNOWLEDGE AND ABILITIES

Generic knowledge and abilities are generally administrative knowledges and abilities with broad

application across systems and operations. They are listed in Section 2 of the catalog. The four

(4) categories of generic K/As are listed below:

I Conduct of Operations K/As

2 Equipment Control K/As

3 Radiation Control K/As

4 Emergency Procedures /Plan K/As

The generic K/As for "Conduct of Operations," are used to evaluate the applicant's

knowledge of the daily operation of the facility. The types of information covered under this

category may include for example, shift turnover or temporary modification procedures.

The generic K/As for "Equipment Control "are used to evaluate the administrative requirements

associated with the management and control of plant systems and equipment. Examples of the

types of information evaluated under this topic include maintenance and temporary modifications

of systems.

The generic K/As for "Radiation Control" are used to evaluate the applicant's knowledge and

abilities with respect to radiation hazards and protection (personnel and public). Examples of

1-5 NUREG- 1123, Rev. 2, S86"i'
the types of information that should be evaluated under this topic are knowledge of significant

radiation hazards or radiation work permits.

The generic K/As for "Emergency Procedures / Plan" are used to evaluate the applicant's

general knowledge of emergency operations. The K/As are designed to evaluate knowledge of

the emergency procedures use. The emergency plan K/As may be used to evaluate the

applicant's knowledge of the plan, including, as appropriate, the RO's or SRO's responsibility to

decide whether it should be executed and the duties assigned under the plan.

1.9 PLANT SYSTEMS

1.9.1 Plant System Organization by Safety Function

Nine (9) major safety functions must be maintained to ensure safe BWR nuclear power plant

operation. The safety functions groups are:

1. Reactivity Control

2. Reactor Water Inventory Control

3. Reactor Pressure Control

4. Heat Removal From the Core

5. Containment Integrity

6. Electrical

7. Instrumentation

* 8. Plant Service Systems

9. Radioactivity Release.

Fifty six (56) plant systems have been included in the BWR Catalog based on their relationship

and importance to safety functions. Table I contains a list of these plant systems, arranged

within safety function. It should be noted that some plant systems contribute to more than one

safety function.

Each plant system has a 6-digit code number. The words "Plant Specific" following the system

title, indicates that the knowledge or ability does not apply to all plants.

See Section 3 of the BWR catalog for the delineation of K/As for the plant systems.

Table 1

Plant Systems by Safety Functions

Safety Function 1: Reactivity Control

201001 Control Rod Drive Hydraulic System

201003 Control Rod and Drive Mechanism

201002 Reactor Manual Control System

NUREG-! 123, Rev. 2 1-6
1.9.2 Knowledge and Ability Stem Statements for Plant Systems

The information delineated within each plant system is organized into 6 different types of

knowledge and 4 different types of ability. If there are no knowledge or ability statements

following a stem statement there is no applicable K/A.

The applicable 10 CFR 55.41 / 43 / and 45 item numbers are included with each stem

statement. In most cases the K/As associated with the stem statements can be used for both the

written examination and the operating test. See Table 2 below:

Table 2

Knowledge and Ability Stem Statements for Plant Systems

K I. Knowledge of the physical connections and/or cause-effect relationships

between (SYSTEM) and the following:

(CFR: 41.2 to 41.9/45.7 to 45.8)

K2. Knowledge of electrical power supplies to the following:

(CFR: 41.7)

K3. Knowledge of the effect that a loss or malfunction of the (SYSTEM) will

have on the following:

(CFR: 41.7 /45.4)

K4. Knowledge of (SYSTEM) design feature(s) and or interlock(s) which provide

for the following:

(CFR: 41.7)

K5. Knowledge of the operational implications of the following concepts as

they apply to the (SYSTEM):

(CFR: 41.5 / 45.3)

K6 Knowledge of the effect that a loss or malfunction of the following will

have on the (SYSTEM):

(CFR: 41.7 / 45.7)

Al. Ability to predict and/or monitor changes in parameters associated with

operating the (SYSTEM), controls including:

(CFR: 41.5 / 45.5)

A2. Ability to (a) predict the impacts of the following on the (SYSTEM) and

(b) based on those predictions, use procedures to correct, control, or

mitigate the consequences of those abnormal operation:

(CFR: 41.5 / 45.6)

1-9 NUREG-I 123, Rev. 2
A3. Ability to monitor automatic operations of the (SYSTEM) including:

(CFR: 41.7 / 45.7)

. A4. Ability to manually operate and/or monitor in the control room:

(CFR: 41.7 / 45.5 to 45.8)

1.10 EMERGENCY & ABNORMAL PLANT EVOLUTIONS

Section 4 of the BWR catalog contains emergency plant evolutions (EPEs) and abnormal plant

evolutions (APEs). An emergency plant evolution is any condition, event or symptom which

leads to entry into the plant-specific emergency operating procedures (EOPs). An abnormal

plant evolution is any degraded condition, event or symptom not directly leading to an EOP

entry condition, but, nontheless, adversely affecting a safety function. The listing of EPEs and

APEs was developed to include those integrative situations crossing several plant systems

and/or safety functions.

The emergency plant evolution strategies described in Revision 4 to the Boiling Water Reactor

Owners Group Emergency Procedures Guidelines, cover five broad areas:

1) Reactor Pressure Vessel Control

2) Reactor Pressure Vessel Control With SCRAM Condition Present and Reactor

Power Above APRM Downscale or Unknown.

3) Primary Containment Control

4) Secondary Containment Control

5) Radioactivity Release Control.

If the operator controls the five broad areas of emergency plant evolutions listed above, the plant

safety functions will be safely maintained. Table 3, below, contains a numerical list of the 16

emergency plant evolutions and the 25 abnormal plant evolutions covered by this catalog.

Table 3

Emergency and Abnormal Plant Evolutions

EPEs

295024 High Drywell Pressure

295025 High Reactor Pressure

295026 Suppression Pool High Water Temperature

295027 High Containment Temperature (Mark III Containment Only)

295028 High Drywell Temperature

295029 High Suppression Pool Water Level

295030 Low Suppression Pool Water Level

295031 Reactor Low Water Level

295032 High Secondary Containment Area Temperature

NUREG-I1123, Rev. 2 !-10
295033 High Secondary Containment Area Radiation Levels

295034 Secondary Containment Ventilation High Radiation

295035 Secondary Containment High Differential Pressure

295036 Secondary Containment High Sump / Area Water Level

295037 SCRAM Condition Present and Reactor Power Above APRM Downscale

or unknown

295038 High Off-Site Release Rate

500000 High Containment Hydrogen Concentration.

APEs

295001 Partial or Complete Loss of Forced Core Flow Circulation

295002 Loss of Main Condenser Vacuum

295003 Partial or Complete Loss of A.C. Power

295004 Partial or Complete Loss of D.C. Power

295005 Main Turbine Trip

295006 SCRAM

295007 High Reactor Pressure

295008 High Reactor Water Level

295009 Low Reactor Water Level

295010 High Drywell Pressure

295011 High Containment Temperature (Mark III Containment Only)

295012 High Drywell Temperature

295013 High Suppression Pool Water Temperature

295014 Inadvertent Reactivity Addition

295015 Incomplete SCRAM

295016 Control Room Abandonment

295017 High Off-Site Release Rate

295018 Partial or Complete Loss of Component Cooling Water

295019 Partial or Complete Loss of Instrument Air

295020 Inadvertent Containment Isolation

295021 Loss of Shutdown Cooling

295022 Loss of Control Rod Drive Pumps

295023 Refueling Accidents

600000 Plant Fire On Site

176060- Generator Voltage and Electric Grid Disturbances

1. 10.1 Knowledge and Ability Stem Statements for Emergency and Abnormal

Plant Evolutions

The information delineated within each emergency or abnormal evolution is organized into three

(3)different types of knowledge and two (2) different types of ability. If there are no knowledge

or ability statements following a stem statement there is no applicable K/A.

The applicable 10 CFR 55.41, 43, and 45 item numbers are included with each stem

statement. In most cases the K/As associated with the stem statements can be used for both

1-11 NUREG-1 123, Rev. 2, I
the written and operating examinations. See Table 4, below:

Table 4

Knowledge and Ability Stem Statements for

Emergency and Abnormal Plant Evolutions

E/AKI Knowledge of the operational applications of the following concepts as they apply

to the (EMERGENCY OR ABNORMAL PLANT EVOLUTION):

(CFR: 41.8 to 41.10)

E/AK2 Knowledge of the interrelations between (EMERGENCY OR ABNORMAL

PLANT EVOLUTION) and the following:

(CFR: 41.7 / 45.8)

E/AK3 Knowledge of the reasons for the following responses as they apply to

(EMERGENCY OR ABNORMAL PLANT EVOLUTION):

(CFR: 41.5 / 45.6)

E/AA1 Ability to operate and / or monitor the following as they apply to (EMERGENCY

AND ABNORMAL PLANT EVOLUTION)..

(CFR: 41.7 /45.6)

E/AA2 Ability to determine and interpret the following as they apply to

(EMERGENCY AND ABNORMAL PLANT EVOLUTION):

(CFR: 41.10 / 43.5 / 45.13)

1. 11 COMPONENTS

Basic components such as valves and pumps are found in many systems. NUREG-1 021,

section ES-205, "General Fundamentals Examination," lists 8 categories of components. The

following eight (8) categories of Components, for which additional knowledge statements are

necessary are listed below and delineated in Section 5 of the BWR catalog.

The component knowledge statements are more detailed than those provided in the system

listing, yet at the same time they are generic to the component types. Each component has a

unique six (6) digit code number and 10 CFR 55.41 (b) item number. See Table 5, below.

Table 5

Components

291001 Valves (CFR: 41.3)

291002 Sensors and Detectors ( CFR: 41.7)

291003 Controllers and positioners (CFR: 41.7)

291004 Pumps(CFR: 41.3)

NUREG-I1123, Rev. 2 1-12
2.0 GENERIC KNOWLEDGES AND ABILITIES

2.0. 1' Technical Require ments Manual(T•) -For the pur•ose of this catalog,

.K/Asthat reference Technical Specific'ations (TS) may include the Technical

Requi're ments Man'ual, Where applicable.,

2.0.2 K/A ClarifyinigEkamples -K/As that include t4he words "such as" list

suggested topical areas as exampliesand are notinte•nded to-be all inclusive.

2.1 Conduct of Operations

2.•1 Knowledge of conduct of operations requirements.

(CER: 4l1.10 /4 5.13)

IMPORTANCE RO 3.8 SRO 4.2

2.1.2 Kniowledge of operator responsibilities during all modes of pln

operation. (CFR: 41.10/45.13)

IMPOQRTAN.~CE RO 4.1 SR4.

.(CFR: 41-10 / 45-13) IMPORTANCE RO 3.7 'ý5 R}39 O 3.9:

2ý.1.A Kniowledge. of individual licensed operator responsibilities related to shift.

staffing, such as medicalrequirements, "nosou g" operationmain•enance

of active license statuis, I OCFR55, etc.

(CER: 41.101/ 43.2)

IMPORTANCE RO 3.3 SRO 3.8

2.1.5 Ability to use procedures related to shift staffing, such as minimum crew

complement, overtimnelimitations, etc.,

(CFR: 41.10/43.5 / 45.1V2)

IMPORTANCE RO 2.9* SRO 3.4

21 bility to manage the coto rooaiiewseInts

ý(CFR: 41.10 f43.5 / 45. 12/1 5. .13)

IMPORTANCE RO 38* SQ 4.8

2".1. AAb'ihity~ to evaluzate. plant performance-and ýmake operational judgments based on. operating ~characteristics, ~reActo bhair A ndisrmn

(CFR: 41.5 /43.5 /45.12 / 45.13)

IMPORTANCE RO 4.4 SRO 4.7

2-1 2-1NURE.G-I1,J23, Rev. 2, Supp. 1
2.1 Conduct of Operations (continued)

2.1i.8 Ability to coordinate personnel activities outside "the control room.

(CFR: 4110 / 45.5 /45.12 /45.13)

IMPORTANCE RO 3.4 SRO 4. 1

2.1.9 Ability to direct personnel activities inside the control room.

(CER: 41.10/45.5 /45.12 /45.13)

IMPORTANCE RO 2.9* SRO 4.5

2.1.10 Moved to 2.2.38

2.1i.11 Moved to 2.2.39

2.1.12 Moved to 2.2.40,

2.1.13 Knowledge of facility requirements for controlling vitapmcontrolled access.

(CFR: 41.10 / 43.5 / 45.9 / 45. 10)

IMPORTANCE RO 2.5 SRO 3.2

2.1.14 Knowledge of criteria or conditions that require plant-wide

announcements, such as- pump starts, reactor trips, mode changes, etc.

(CFR: 41.10 /43.5 /45.12)

IMPORTANCE RO 3.1 SRO 3.1

2.11.5 Knowledge of administrative requirements for temporary management

directives, such as standing orders, night orders, opera.ions memos, etc.

(CFR: 41.10/•/45.12)

IMPORTANCE RO 2.7 SRO 3.4

2.1 .16 Deleted

2.1.17 Ability to make accurate, clear, and concise verbal reports.

(CFR: 41.10 /45-12 /45.13)

IMPORTANCE RO 3.9 SRO 4.0

2.1.18 Ability to make accurate, clear, and concise logs, records, status boards,

and reports.

(CFR: 41.10 /45.12 /45.13)

IMPORTANCE RO 3.6 SRO 3.8

2.1.19 Ability to use plant computers to evaluate system or component status.

(CFR: 41.10 /4 5.12)

IMPORTANCE RO 3.9 SRO 3.8

NUREG-1123, Rev. 2, Supp. I 2-2
2:.1 Conduct of Operations (continued)

2. 1.20 Ability to interpret and executte ýprocedure's'tepsý.

(CR : -4-1, -10 /43.5 /45'.12

IMPORTANCE RO 4.6 SRO 4.6,

2.1.21 Ability to verify the controlled ~procedure cop-y.

(CFR:41.10 /45.10 /45.13) -

IMPORTANCE RO 3.5* SRO 3.60

2.1.22 Revised and moved to 2.2.35

2.1.23 Ability to perform specific system and inte~ae patpocedures durinig

all modes of plant operation.

(CFR: 41.10 / 43.5 /45..2/ 45.6)

IMPORTANCE RO 4.3 SRO 4.4

2.1.24 Moved to 2.2.41

2.1.25 Ability to interpret reference maeilsc s s ýyvs a~!ý,

etc.

(CER: 4l .101 /43.5/ 45~.12)

IMP-ORTANCE RO 3.9 SRO 4.2.

2.1.26, Knowledge of industrial safetyr procedures (such as -rot-ati ng equipmen~t,

elcrcl ihtmeatrhg rsue. i chl orine, oxy~gen andi

hydrogen).

(CFR: 41.10/45.~1,2

IMPORTANCE RO:3.4 S~RO 3.

1.F27, Kno~wledgeof system purpose and/or func~tion.~

(CFR 4 1.7)

IMPORTANCE R0 3.9 1 SRO 4.0

2.1.28 Knowledge of the purpose and muction.,.ofiiiajor-sys ,teniom coi s nd

controls.

(CFR: 41.7)

IMPORTANCE RO 4.1 SRO 4.1

2.1.29 Knowledge of how to cond'uct system lineups,§suchkas, vales,.breaker,

switches, etc.

(CFR: 41:1 1/ 45.1 /45.'12)

IMPORTANCE RO 4.1 SRO 4.0

2-3 N2t-3 NURG-'I 1I23, R"". 2, S -up p.
2.4 Conduct of Operations_(continued)

2.1.30 Ability to locate and operatecomponents, including local controls.

(CER: 4.1.7/ 45.7)

IMPORTANCE RO 4.4 SRO 4.0

, 2.1.31 Ability to locate control room switches, controls, and indications, and to

determine that they correctly reflect the desired plant lineup.

(CER: 41.10 /45'12)

IMPORTANCE RO 4.6 SRO 4.3.

:2. •2 •Ability to explain and apply system. limits and precautwons,

(CFR: .•01/43.2• 45.12)

IMPORTANCE R 3.8 SR~O4.0

21.33 Moved to 2.2.42'

2.1.34 Knowledge of primary and secondary plant chemistry limits.

(CER: 41.10 f43.~5 /45.12)

IMPORTANCE RO 2.7 SRO 3.5

27.1.35 Knowledge of the fuel-handlinig responsibilities. of.SRýs

IMPORTANCE RO2.2 SRO 3.9

2.1 .36 Knowledge of pro•edures and limitations involved in core alterations.

(CFR 41.10~/43.6 /45.7)

IMPORTANCE RO 3.0 SRO 4.1

2.1.37 Knowledge of proceduresgudn g l or l limitatiosssciated.w.•• a ass d wth

reactivity management.

(CER:) 4.143.~6/f45.6)

IMPORTANCE RQ 4.3 SRO 4.6

2•!. 0 8 K now ledge ofthe :station's !reg uirem ents for verba Icom Im unicauions w ehen:::..........................

implementing procedures.,

(CER: 41.10 / 4 5.13)

IMPORTANCE RO 3.7* SRO 3.8

2?:T... knowledge of conservative deion makng pactces.

(CER: 41.10 / 43.~5 / 45.12)

IMPORTANCE RO 3.6 SRO 4.3

2:1.40 Knowledge of refueling administrative requirements.

(CFR: 41.10 43.5/45:13)

IMPORTANCE F0O2.8 SRO 39

NUREG-I 1123, Rev. 2,~Sp~I2 ýý4
2.11 Conduct of Operations'(continued)

.2 1.1.41 Knowledge -,of .t .h e ref Iuelin Igprocess--:1

(CER: 41.12/-41.10/43)6/5.13)

IMORTTANCE R0 2.8 SRO 3.7,

2.14.42 Kno-wledge oif new and spent fuel movement-procedures.

(CER: 41.10 / 43.7 / 45.13)

IMPORTANCE RO 2.5 SRO 3.4

2.143 Ability t~o use procedures todetermine It Ihe e'etects on reactivity of plant

chaniges, such as reactor coolant system temperature, secondary plant,

fuel depletion, etc.

(CFR: 41.10 /43.6 /45.6)

IMPORTANCE RO 4.1 SRO 4.3

241.44 Knowledge of RO duties~ in the control room duringi fuel ha~ndling such as

respondinig to alarms from the fuel handling area, communication with the

fuel storage ~facility, systems operated from the control rominspport of

jqli~ng perations, rand suppiorting~ inistrumenitation.;

(CFR: 14.10 /43.7 /45.12)

IMPORTAN4CE RO 3.9 SRO 3.8

2. .45- Ability to i dent ify and interpret diverse ndkA tions to val idate thle

response of another indicator.

(CFR: 41.7 / 43.5 / 45.4)

IMPORTANCE RO 4.3 SRO 4.3

M; 2-5 NUREG-1 123, Rev. 2, Supp. 1
2.2 Equipment Control

2'2.12 Ability-to perform pre-4startup procedures for the facility, including

operating those controls associated with plant equipment that could affect

reactivity. ... C R:l 41.°51i/41'.1'i0`/`"43.J5.-,43.16/4 5 I )

IMPORTANCE RO 4.5 SRO 4A4

222Ability to manipulate the console controls las required to opera-te the:

facility between shutdown and designated power levels.

(CFR: 41.6 /41.7 /45.2)

IMPORTANCE RO 4.6 SRO ~4.1

2.2.3 (multi-unit license) Knowledge of the design, procedural, and operational

differenc~es between units.

(CFR: 41.5 41.6 / 41.7 /41 .10 /45-1J2)

IMPORTANCE RQ318 SRO3~

2.2.4 (mul~ti-unit license) Ability to ~explain the variations i cont'rol~ board

laytouts, systems, intuettoadpoeua actions between ui~ts

(CFR:<46>41.7 / 41.10/45.1 /45. 1 3)

IMPORTANCE RO 3.6 SRO 3.6

2~5' Knowledge of the process for makin design or ove ,rating changes to the

facility.

(CFR: 41.10 / 43.3 / 45.13)

IMPORTANCE RO 2.2 SR~O 3.2

2.2.6, Knowledge of the process for making changes to procedures.

(CFR: 41.10 / 43.3 / 45.13)

IMPORTANCE RO 3.0 SRO 3.6.

2.2.7, Knowledge of the process for conducti-ng speciJ~aor infrequent tests%.,

(CFR: 41.10 43.3 /45.13)

IMPORTANCE RO29. SRQ 3.6

2.8-. Deleted

2.2~. D el"et"e d!

2.2.10 Deleted

NUREG-I f23', Rv. 2, Su~pp. I 2,'ý'6
2.2 Equipment Control (continued)

2.2.11 Knowledge of the process rfor ,controlling temporary design changes.

(cFR:A1 I10 / 43.3 / 45. 13)

IMPORTANCE RO 2.3 SRO 3.3

2.2.12 Knowledge of surveillance procedures.

(CER: 41.10 /45.13)

IMPORTANCE R0 3.7 SR4.

2.2.13 Knowledge of taggingand dclearance procedures.

(CFR: 41.10 /45.13)

,IMPORTANCE RQ 4.1 'SRO 4.3

2.2. 14 Knowledge of the process iforuconrollhng tequipiiimconfiguration or

status.

(CFR: 41.10 43.3/45.13)

IMPORTANCE lRO 3.9 SRO 4.3

2.2.15 Ahility to determine the excted plant configurationuing design and.

configuration control documentation,~ such as drw~ing, line-.ups, tagots

etc.

(CFR 4 1.10/ 43.3/45*113)

.IMP~ORTANCE R 3.9 SRO 4.3

2.2.16 Delee~d

.2.:J7 Knowledge of the proc ssLor managig maintenance, activities during

powersk assessments, wrk roithon and

coordination with the trasmiionsystem or

(CFR: 41.10 / 43.5 I/ 45.13)

IMPORTANCE 'RO 2.6 SRO 3.8

2.2.18 Knowledge oftthe process for managingagmaint . . ".......vities duringg..en.an•. 

shutdown operations, such as risk assessrnents, woik ,prioritization, etc.

IMPORTANCE R0 2.6 S0R3.9.

2.2.19 Knowledge of maintenance work order"requrements.

(CFR: 41.10 43.5 / 45.13)

IMPORTANCE RO 2.3 R34

.2-7, 2-7 NIJREOLI 123~ R&v. 2, Supp. I
2.2 Equipmenit-Conro 61(coninued)

21.2.20 Knowledge of the process for man'aging ýtroubleshooting activities.';

(CER: 41.10/43.5 /45.13)

IMPORTANCE RO 2.6 SRO 3.*8

2.2.21 Knowledge of pre- and post-maintenance operability requirements.

(CER: 41.10 /43.2)

IMPORTANCE RO 2.9 SRO 4.1

22.2 2 'knowledge of limiting conditions for operations and safety limits.

(CFR: 41.5 / 43.2 / 45.2)

IMPORTANCE RO 4.0 SRO 4.7

2.2.23, Ability to track Technical Specification limiting conditions for operations.

(CFR: 41-10/43.2 /45.13)

IMPORTANCE RO 3.1 SRO 4.6

2.2.24 Moved to 2.2.36

2.2.25 Knowledge of the bases in Technical, Specifications for limiting conditions

for operations and safety limits.

(CFR: 41 .5 /41 .7 /43.2)

IMPORTANCE RO 3.2 SRO 4.2

2.2.26 Moved to 2.1.40

2.2.27 Moved to 2.1.41

2.2.28 Moved to 2.1.42

2.2.29 Moved to 2.1.35

2.2.30 Moved to 2.1.44

2.2.31 Revised and moved to 2.1.36

-2.2.32 Deleted

2.2.33 Deleted

2.2.34 Revised and moved to 2.1.43

NUJREG- 1123, Rev. 2, Supp. I2- 2-8
2.2 Equipment Control (continued)

2.2.35 Ability to determine Technical, "Specification Mode of Operation.,

(CFR: 41.7 /•4 1.10/43.2/ 45.13)

IMPORTANCE RO 3.6 SRO 4.5

2 2.2.36 Ability to analyze the effect of maintenance activities, such as degraded

power sources, on the status of limiting conditions for operations.

(CFR: 41.10/ 43.2a/ 45.13)

IMPORTANCE RO 3.1 SRO 4.2

2.2.37 Ability to determine operability and/or availability of safety related

equipment.

(CFR: 41.7 / 43.5/45.12)

IMPORTANCE R0 3.6 SRO 4.6

2.2.38 Knowledge of conditions and limitations in the facility license.

(CFR: 41.7 /41-10 /43.1 / 45.13)

IMPORTANCE _R0 3.6 SRO 4.5

2.2.39 Knowledge of less than or equal to one hour Technical Specification

action statements for systems.

(CER: 41.7 /41.10 /43.2 / 45.13)

IMPORTANCE R0 3.9 SRO 4.5

2.2.40 Ability to apply Technical Specifications for a system.

(CFR: 41.10 / 4 3.2 / 4 3.5 / 4 5.3)

IMPORTANCE RO 3.4 SRO 4.7

2.2.4 1 Ability to obtain and interpret station electrical and mechanical drawings.

(CER: 41.10 / 45.12 / 45.13)

IMPORTANCE RQ 3.5 SRO 3.9

2.2.42 Ability to recognize system parameters that are entry-level conditions for

Technical Specifications.

(CER: 41.7 /41.10 /43.2 /43.3 / 45.3)

.IMPORTANCE RQ 3.9 SRO 4.6

2.2.43 Knowledge of the process used to track inoperable alarms.

(CFR: 41.10 /43.5 /45.13)

IMPORTANCE RO 3.0 SRO 3.3

2ý9 NUREG- 1123, Rev. 2, Supp. 1
2.2 Equipment Control (continued)

2.2.44 Ability to interpret control Iroom indications: to .verii the status and .operation of asystem, and understand hoW operator actions and

directives affect plant and system conditions..

(CFR: 41.5/43.5/ /45.12)

'IMPORTANCE RQ 4.2 SRO 4.4

EQI123, 1 Re.2,Spp-l 212-10
2.3 Radiation Control

2.3.ý11 Deleted

2.3.2 Deleted

2.3.3 Deleted

234 Knowledge of ýradiation exposur limitsýu'nder normalo mrec

conditions.

(CFR: 41.12/43.4/45.10)

IMPORTANCE RO 3.2 SRO 31.7

ý2.3.5 Ability to use radiation m~onitoring systems, such as fixed radiation

monitors and alarms, portable survey instrumients, personnel monitoring

equi~pment, etc.

(CFlW 4i1.11 / /41.127/43.4 /45.9)

IMPORTANCE RO 2'.9 SRO 2.9

2.3.6 Aityto approve release permiits.

(CFR: 41 .13 / 43.4 / 45.10)

IMPORTANCE RO 2.0 SRO 3.8

2.3.7 Ability to comply with radiation work perm~it requirementsduring normal

or abnormal conditions.

IMPORTANCE RO 3.5 SRO 3.6

2.3.8 Deleted

2.3.9 Deleted

2.3.10 Deletedl

2.3.11 Ability to control radiation releases.'

(GFR: 4 1. 11/43.4 /45.1 ,0)

IMPORTANCE RO~ 3.8 SRO 4.3

2.3.12 Knowledge~ of radiological safety principles pertaining to licensed

operator dtesuch as containiment entry requiremients, fuel handling

responsibilities, access to locked high-radiation areas, alignjing filters,

etc.

(CFR:'4112 / 45.9 /45. 10))

IMPRTA E R-O 3.2 SRO-3.7

2-11. 2-li, NUREG-1 12j3, Rev. 2, Supp
2.3 Radiation Control (continued)

2.3.,13 Knowledge of radiological safety procedures pertaining to licensed

operator duties, :such as response to-radiation monitor alarms,.

containment entry requirements, fuel handling responsibilities, aiccessto

llocked ;high-radiation areas, aligning filters, .etc.

(CER: 41.12/43.4 /45.9/45.10)

IMPORTANCE RO33A4 SRo 3.

2..4Knowledge of radiation or contamination haards thtmyaie during

normal, abnormal, or emergency coniiditions -,or -activities.

(CER: 41.12 /43.4 /45.10)

IMPORTANCE RO 3.4 SRO 3.8

2.3.15 Knowledge of radiation monitoring systems, such as fixed radiation

monitors and alarms, portable survey instruments, personnel monitoring

equipment, etc.

(CFR: 4L12 /43.4 / 45.9)

'IPRAC" -RO 2.9. SR 3.11

NIJREO-212," Rev. 2, Supp. I 2-12 21
2.4 Emergency Procedures / Plan

2.4.z Knowledge of EEOP entry Conditions and immediate action steps.

(CFR: 41.10 / 43.5/45.1 3)

IMPORTANCE RO 4.6 SRO 4.8

2.4..2: Knowledge of system set points, interlocks and automatic actions

associated with EOP entry conditions.

(CFR: 41.7 /45.7 / 45.8)

Note: The issue of setpoints and automatic safety features is not specifically

covered in the systems sections.

IMPORTANCE RO 4.5 SRO 4.6

2.4.3 Ability to identify post-accident instrumentation.

(CFR: 41.6 /45.4.)

IMPORTANCE RO 3.7 SRO 3.9

2.4.4 Ability to recognize abnormal indications for system operating

parameters that are entry-level conditions for emergency and abnormal

operating procedures.

(CFR: 41.10 /43.2 / 45.6)

IMPORTANCE RO 4.5 SRO 4.7

2.4.5 Knowledge of the organization of the operating procedures network for

normal, abnormal, anod emergency evolutions.

(CFR: 41.10 /43.5 /45.13)

IMPORTANCE RQ 3.7 SRO 4.3

2.4.6 Knowledge of EOP mitigation strategies.,

(CFR: 41.10/4 3..5/ 45.13)

IMPORTANCE RO 3.7 SRO 4.7

2.4.7 Deleted

2.4.8 Knowledge of how abnormal operating procedures a~re used in conjunction

with EOPs.

(CFR:41.10/43.5 /45.13)

IMPORTANCE RO 3.8 SRO 4.5

ý2.14.9 Knowledge of low power/shutdown implications in accident (e.g., loss of

coolant accident or loss of residual heat removal) mitigation strategies.

(CFR: 41.10 /4 3.5 /45.13)

IMPORTANCE RO 3.8 SRO 4.2

2-13 2-13NUREG- 1 123., Rev. 2, Supp. I
2.4 Emergency Procedures / Plan (continued)

2.4.10 Deleted

2.4.1 1 Knowledge of abnormal condition procedures.

(CFR: 41.10 / 43.5 / 45.13)

IMPORTANCE RO 4.0 SRO 4.2

2.4.12 'Knowledge of general operating crew resp.1onsblte during emergency'

operations.

(CFR: 41.10 /45.12)

IMPORTAN~CE RO 4.0 SRO 4.3

24.,13 Knowledge of crew roles and responsibilities during EOP usage.

(CFR: 41.10/ 45.12)

IMPORTANCE RO 4.0 SRO 4.6

2.4.14 Knowledge of general guidelines for EOP usage-

(CER: 41.10 / 45.13)

IMPORTANCE RO 3.8 SRO 4.5

2.4.15 Revised and moved to 2.1.38

2.4.16 Knowledge of EOP implementation hierarchy and coordination with other

support procedures or guidelines such as, operating procedures,

abnormal operating procedures, and severe accident management,

guidelines.

(CFR:4 1.10/ 43.5 / 45.13)

IMPORTANCE R0 3.5 SRO4.4

2.4.17 Knowledge of EOP terms and definitions.

(CER: 41.10 /45.13)

IMPORTANCE RO 3.9 SRO 4.3

2.4.18 Knowledge of the specific bases for EOPs.,

I)MPORTANCE RO 3.3 SRO 4.0

2.4.19 Knowledge of EOP layout, symbols, and icons.

(CFR: 41.10 /45.13)

IMPORTANCE RO 3.4 SRO 4.1

2.4.20 Knowledge of the operational implications of EOP warnings, cautions, and

notes.

(CFR: 41.10 /43.5 /45.13)

IMPORTANCE RO 3.8 SRO43

NUREG- 1123, Rev. 2, Sup. I4 2-1t4
. Emergency Procedures /Plan i(continued)

2.4.21 Knowledge of the parameters and logic used to; assess the status of safety

functions, such as reactivity control, core cooling :and heat removal,

reactor coolant system integrity,containment conditions, radioactivity

release control, etc.

(CFR: 41.-7 / 43.5 / 45.12)

IMPORTANCE RO 4.0 SRO 4.6

2.4.22 Knowledge of the bases for prioritizing safety functions during

abnormaVemergency operations.

(CFR: 41.7 /41 .10 / 43.5 / 45.12)

IMPORTANCE RO 3.6 SRO 4.4

24A ••. Knowledge of the bases for prioritizing emergency procedure

implementation during emergency operations.

(CFR: 41.10/43. / 45.13)

IMPORTANCE RO 3.4 SRO 4.4

2.4.24 Deleted

2.4.2 5 Knowledge of fire protection procedures.

(CFR: 41.10 / 43.5 / 45.13)

IMPORTANCE RO 3.3 SRO 3.7

2.4.26 Knowledge of facility protection requirements, including fire brigade and

portable fire fighting equipment usage.

(CFR: 41.10 /4 3.5 / 45.12)

IMPORTANCE RO.3.1 SRO 3.16

2.4.2 7 Knowledge of "'fire in the plant" procedures.

(CFR: 41.10O/43.5 /45.13)

IMPORTANCE R0 3.4 SRO 3.9

2.4.28 knowledge of procedures relating to a security event.

(CFR: 4110 /43.5 /45.13)

IMPORTA•CE R 03.2 SRO 4.1

2..9Knowledge of the emergency plan'.

(CFR: 41.10Q/43.5 /45.1 1)

IMPORTANCE RO 3.1 SRO 4.4

2-:115 2-5N.URIEG-1 12.3, R~ev. 2, Supp. I
2 .4 Emergency ýProcedures,/Y Plan (continued)

2..3 Kowlege of events relatedto, system operation/status that must, be

reported to -Internal ýorganizations or external agnis suh as the'State,

thje ýNRC, othtransm ission. system operator.ý

(CER: 41".10/43.5 /45'.-11)

IMPORTANCE RO 2.7 SR 4.1'

2.4.3 1 Knowledge of annunciator alarms, indications, or response procedure "s.

(CFR: 41.10 / 45.3)

IMPORTANCE RO 4.2 SRO 4.1-

2.4.32 Knowledge of operator response to loss of all annunciators.

IMPORTANCE RO 3.6, SRO 4.0

ý2.4.33 Moved to. 2.2.43

2.4.34; Knowledge of RO tasks performed outside the ma~in control room during

an emnergency and the resultant operational effects.

(CFR:41.10 /43.5 /45.13) -

IMPORTANCE R0 4.2 SO.

.2.4'. 35 Knowledge of local auxiliary operator tasks'durgan e rgncy and the

resultant operational effects.

(CFR: 4 1.10 /43.5 /45.13)

IMPORTANCE RO 3.8 SRO 4.0

2ý.4.36 Deleted

2.4.37 Knowledge of the lines of authority during implementation of the

emergency plan.

(CFR: 41.10 / 45.13)

IMPORTANCE RO 3.0 SRO 4.1

2.4.38 Ability to take actions called for in theJfacitityeerec plan, including

supporting or acting as emergency coordin itorifrqired.

(CFR:41.10/43.5/45.11)

IMPORTANCE I RO 2.4 SRO 4.4

2ý.4.39 Knowledge of RO responsibilities in emergency "plan impleenation.

(CFR: 41.10 /45.11)

IMPORTANCE RO 3.9 SRO 3.8

NUREG- 1123, Rev. 2, Supp. I2-6 Z-1i6
2.4 Emenrgmency Procedures /,Plan (continued)

2.4.4 0 Kno'wledgreý of SRO re Ispo ,ns Iibil .it Iie sý in eme .r gen c y plan. 1AI H-jmp. 01lementation.

(CFRI:4.10,/ 43.5 /45.,1-1)

IMPORT;ANCE RO, 2.7 -SRO;4.5

2.4.41 Knowledge of th emltergenc action level thresholds and 'classifications.

(CFR:44 1.010/43.5j £45.1 1)

IMPORTANCE RO 2.9 SRO4.6,

2 .4.42: knowledge of emergency resonse facilities.

(CFR: 41.10 /45.11)

IMPORTANCE R0 2.6 SRO 3.8

2.4.43 Knowledge of emergency comnmunicationis systems andte~chniques.

~(GF~R:41.10 / 45.13)

IMPORTAN~CE RO 3.2 SRO 3.8

2.4.441 Knowledge of emergency plan protective action recommendations.

(CFR: 41.10 0/ 4.2 /43.5 / 45.1 1)

IMPORTANCE RO 2.4 SRO 4.4

,2.4.45, Ability to prioritize and interpret the significance of -each annunciator or,

alarm.

(CER: 41.10 43.5/45.3 /45.12)

IMPORTANCE RO 4.1 SRO 4.3

A~.4 bility to verify that the alarms are conisistent wihtepat odtos

IMPORTANCE RO 4.2 SRO 4.2

2.4 .47 Ability to dia~gnose and recognize trends in an accurate and timelY

manr tlzn the appropriate control room referenceinmterial":

(CFR:~ 41~.10 /43.5 / 45.12)

IMPOR~TANCE RO 4.2 SRO 4.2

2.4.48 Revised and moved to 2.2.44

2..9Ability to performn without reference to procedures those actions that

require immediate operation of system components and controls.

(CER: 41-10 / 43.2 / 45.6)

IMPORTANCE RO 4.6 SRO 4.4

2-:17 2-17NURG:I1-j23; Rv2, 'V Supp. I'
2.4 Emergency•Procedures / Plan (continued)

2.4.i-50 Abilityto ~Yverifywsystem alarm, setpoints -and oPerate controls identified in

the alarm response manual.

(CR 41.10/43.5/-45.3)

tMPORTANCE RO 4.2 SRO 4-.0

NýURG-1l 1,23,, Rev. 2, Supp. I 2-18 21
4.2 Abnormal Plant Evolutions page

295001 Partial or Complete Loss of Forced Core Flow Circulation 4.2-3

295002 Loss of Main Condenser Vacuum 4.2-5

295003 Partial or Complete Loss of A.C. Power 4.2-7

295004 Partial or Complete Loss of D.C. Power 4.2-9

295005 Main Turbine Trip 4.2-10

295006 SCRAM 4.2-12

295007 High Reactor Pressure 4.2-14

295008 High Reactor Water Level 4.2-16

295009 Low Reactor Water Level 4.2-18

295010 High Drywell Pressure 4.2-19

295011 High Containment Temperature (Mark III Containment Only) 4.2-21

295012 High Drywell Temperature 4.2-22

295013 High Suppression Pool Water Temperature 4.2-23

295014 Inadvertent Reactivity Addition 4.2-24

295015 Incomplete SCRAM 4.2-26

295016 Control Room Abandonment 4.2-28

295017 High Off-Site Release Rate 4.2-29

295018 Partial or Complete Loss of Component Cooling Water 4.2-31

295019 Partial or Complete Loss of Instrument Air 4.2-33

295020 Inadvertent Containment Isolation 4.2-35

295021 Loss of Shutdown Cooling 4.2-37

295022 Loss of Control Rod Drive Pumps 4.2-39

295023 Refueling Accidents 4.2-40

600000 Plant Fire On Site 4.2-42

7,0,0,000: Generator Voltag andEeti Grid Disturbances 424

4.2-I NRE- 123 Re. UPP'. I
Intentionally Blank
APE: 600000 Plant Fire On Site

IMPORTANCE

K/A NO. ABILITY RO SRO

AAI Ability to operate and / or monitor the following as

they apply to PLANT FIRE ON SITE:

AA1.01 Respirator air pack ............................................... 3.0 2.9

AA1.02 Re-installation of fire detector ........................................ 1.8 1.9

AA1.03 Bypass of fire zone detector ......................................... 1.8 2.2

AA1.04 Bypass of heat detector ............................................ 1.9 2.2

AA1.05 Plant and control room ventilation systems ............................... 3.0 3.1

AA 1.06 Fire alarm ...................................................... 3.0 3.0

AA1.07 Fire alarm reset panel ......................... ................... 2.3 2.4

AA1.08 Fire fighting equipment used on each class of fire ........................... 2.6 2.9

AA1.09 Plant fire zone panel (including detector location) .......................... 2.5 2.7

AA2 Ability to determine and interpret the following as they

apply to PLANT FIRE ON SITE:

AA2.01 Gas treatment system .... ...... ................................... 2.2 2.3

AA2.02 Damper position .................................................... 2.8 2.9

AA2.03 Fire alarm .................................................. 2.8 3.2

AA2.04 The fire's extent of potential operational damage

to plant equipment ................................................ 2.8 3.1

AA2.05 Ventilation alignment necessary to secure

affected area .................................................... 2.9 3.0

AA2.06 Need for pressurizing control room

(recirculating mode) ......... .................. I............... 2.5 2.8

AA2.07 Whether malfunction is due to common-mode electrical

failures ..................... .................................. 2.6 3.0

AA2.08 Limits of affected area

AA2.09 That a failed fire alarm detector exists .................................. 2.4 2.8

AA2. I0 Time limit of long-term-breathing air system for

control room .................................................... 2.9 3.1

AA2.1 I Time limit for use of respirators ...................................... 2.9 3.0

AA2.12 Location of vital equipment within fire zone ............................ 3.1 3.5

AA2.13 Need for emergency plant shutdown ................................ 3.2 3.8

AA2.14 Equipment that will be affected by fire suppression

activities in each zone ............................................... 3.0 3.6

AA2.15 Requirements for establishing a fire watch ............................... 2.3 3.5

AA2.1 6 Vital equipment and control systems to be maintained

and operated during a fire........................................... 3.0 3.5

AA2.17 Systems that may be affected by the fire ................................. 3.1 3.6

4.2-43 NUREG-! 123, Rev. 2
A PE: 700000 Generator Voltae and Electric Grid-Disturbances

K!/A NO. KNOWLEDGE RO SRO

AKi. Knowledge of the operational implication's of 'the

following conceptsas they appiytolGENERATOR VOLTAGE AND

ELECTRIC GRID DISTURBANCES:

(C:41.4 41 .5, 4 1.7, 4 11 0/ 4 5 8)

MK1.O Definition of terms: volts, watts, amps, VARs, power-factor ..... ....... . .... 3.3 35

AK .02 Over-excitation . ... <. .. .... .,.,.. . . .. . . . . 3.3 3:4

AKUnerexton ...... 33 3.4

AK2.1 'knowledge of the interrelations between-GENERATORi OLTAGEeAN

ELECTRIC GRID DISTURBANCES and the following:

(cFR: 41.4, 41.5,41.7, 41JO//45.8)

AK2.01 M otors .. . . .I. . .. . . . . . . .. . . . . . ...... ~ 3.1 3.2

AK2.02 Breakers, relays .. . . . . . . . .. . . .. . . . . . '.3 . . . . . . 3 1 3.3,

AK 2.03 Sensors, detectors, indicators ....... '...'... ... ... ... .............. . 3.0 3.1

AK2.04 Controllers, positioners .. . . . . . . . . . . . . . . . . . . . . .3.0 3.0

AK2.05 Pumps .. . . . . . . . . . . . . .. . . . . . . . . . . . . . 3.1 3.2

AK2.06 Reactor power .. . . .. . . .. . . .. . . .. . . .. . . .. . . . 39 4.0

AK2.07 Turbina/generator control .... ...... .. .. .. . . . . . ........ .3.6 3.7,

AK.kno'wledlgeof the reasons for the' folowi 'ng r eslionses

as they applyto GENERATOR yVOLTAGE'AND ELECTRiCdGRI

DIST~URBAN$CES:

(CFR: 41I.4, 41 .5, 41.7, 41.1 /4 / 5.8)

AK3.01 Reactor and turbine trip criteria . ... . . . . .. .< i~3.9 . . . . ... ..... 4.2ý

AK3.02 Actions contained in abnormal operating procedu~re for voltage and~,

grid disturban ces .............. ... ...... .................... .. .. .. .3.6 3.9

ABILITY

A.Ai.' Ability -to' -operate and/or. monitor the foll owing as

th~ey apply to GENERATOR VOLTAGE AND ELECTRIC GRIPD

DISTURBANCES:

(CFR: 4 1.5 and 41.10 / 45.5, 45.7, and45'.8)

A l .01 Grid frequency and voltage .. . . . . . . . . . .. . . . . . . . . .. 3.6 3.7

AA 1.02 Tuirbine/generator controls . . . . . . . . . . . . . . . . ... .. .3.8 31.7

AA 1.03 Voltage regulator. controls ... . . . ... ....... ... 3.8, 3.7

AA [.04 Reactor, controls ... .. . .. . ... .. .. .. .. .-4.1 4.1

AAI .05 Engineered safety, features..... ........ .. 4 .... .. 3... 9 ~4.0

NUEAE- 123, Rev. 2, Supp.. I4 4'.27,44
APE: 700000 Generator Voltage and'Electric Grid Disturbances

IMPORTANCEý

KVA NO. AB.ILITY RO SRO

AA2. Ability-to determine and/or interpretthe following.as

they apply to GENERATOR VOLTAGE AND ELECTRIC GRID

DISTURBANCES:

ý(CFR: 41.5 and 43.5 /45.5, 45.7, and 45.8)

AA2.0 I Operating point on the generator capability 'curve .............. ............ ý3.5 3 -ýý6

AA2.02 Voltage outside the generaor capabilt cuv . ......................... 3.5 3.6

AA2.03 Generator current outside the capability curve 36....

AA2.04 V.,ARs outside capabilitycurve . ... .. . 3.6 3.6

AA2.05 Operational status of ofisite circuit .. .. . .. . .. . . .. . . .. . . . 3.'2" -3.8

AA2.06 Generator frequency limitations .. . . . . . . . . . . . . . . . . . . 3.4 3,5

AA2.07 Operational status of engineered safety features .............. 3.6 4.0

AA2.08 Criteria to trip the turbine or reactor .. . . . . . . . . . . . . . . .4.3 4.4

AA2.09 ~Operational status of emergency diesel generators.. ............ 39 43

AA2. 10 Generator overheating and th~e re~qwre actions ....... I..........3.6 3.8

4ý 4.2.45 NUREG-1 123, Rev. 2, Supp.1
- U

NRC FORM 335

(9.2004)

NRCMD 3.7

U.S. NUCLEAR REGULATORY COMMISSION

BIBLIOGRAPHIC DATA SHEET

(See instructions on the reverse)

1. REPORT NUMBER

(Assigned by NRC, Add Vol., Supp., Rev.,

and Addendum Numbers, if any.)

NUREG-1 123, Revision 2, Draft

Supplement 1

9

2. TITLE AND SUBTITLE

Knowledge and Abilities Catalog for Nuclear Power Plant Operators:

Boiling Water Reactors

Draft Report for Comment

3. DATE REPORT PUBLISHED

MONTH YEAR

May 2007

4. FIN OR GRANT NUMBER

5. AUTHOR(S) 6. TYPE OF REPORT

D. Muller Technical

7. PERIOD COVERED (Inclusive Dates)

8. PERFORMING ORGANIZATION - NAME AND ADDRESS (If NRC, provide Division, Office or Region, U.S. Nuclear Regulatory Commission, and mailing address; if contractor,

provide name and mailing address.)

Division of Inspection and Regional Support

Office of Nuclear Reactor Regulation

U.S. Nuclear Regulatory Commission

Washington, DC 20555-0001

9. SPONSORING ORGANIZATION - NAM E AND ADDRESS (If NRC, type "Same as above"," if contractor, provide NRC Division, Office or Region, U.S. Nuclear Regulatory Commission,

and mailing address.)

Same as above

LEMENTARY NOTES

11. ABSTRACT (200 words or less)

This document provides the basis for development of content-valid licensing examinations for reactor operators and senior

reactor operators. The examinations developed using this document will cover those topics listed under 10 CFR 55, "Operators'

Licenses." The BWR catalog contains approximately 7,000 knowledge and ability (K/A) statements for reactor operators and

senior reactor operators. Each K/A statement has been rated for its importance to safety.

Draft Supplement 1 to Revision 2 of NUREG-1123 proposes to reword and reorganize Section 2, "Generic Knowledge and

Abilities," and add a new K/A topic to Section 4, "Emergency/Abnormal Plant Evolutions," to address generator voltage and

electric grid disturbances. Other proposed changes are minor and typographic in nature.

The NRC seeks comments on Draft Supplement 1 to Revision 2 of NUREG-1123. Comments must be provided within 60 days

from the date of publication. Comments received after 60 days will be considered if practicable to do so, but only those

comments received on or before the due date can be assured for consideration.

T

12. KEY WORDS/DESCRIPTORS (List words or phrases that will assist researchers in locating the report.)

operator licensing

examinations

reactor operator

senior reactor operator

training

ask a analysis oiling water reactors

13. AVAILABILITY STATEMENT

unlimited

14. SECURITY CLASSIFICATION

(This Page)

unclassified

(This Report)

unclassified

15. NUMBER OF PAGES

16. PRICE I a

NRC FORM 335 (9.2004) 

PRINTED ON RECYCLED PAPER

NRC FORM 335 (9-2004) PRINTED ON RECYCLED PAPER