NUREG/CR-5783

BNLrNUREG-52299

Aging Assessment of

the Combustion Engineering

and Babcock & Wilcox

Control Rod Drives

Prprdb

Prepared byE. Grove, W. Gunther

Brookhaven National Laboratory

Prepared for

U.S. Nuclear Regulatory Commission
r

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f:
NUREG/CR-5783

BNL-NUREG-52299

R9

Aging Assessment of

the Combustion Engineering

and Babcock & Wilcox

Control Rod Drives

Manuscript Completed: December 1992

Date Published: January 1993

Prepared by

E. Grove, W. Gunther

S. K. Aggarwal, NRC Program Manager

Brookhaven National Laboratory

Upton, NY 11973

Prepared for

Division of Engineering

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555

NRC FIN A3270
ABSTRACT

The effects of aging upon the Babcock & Wilcox (B&W) and-Combustion Engineering (CE)

control rod drive systems have been evaluated. For this study, the CRD system boundary included the

control rod assemblies, guide tubes, control rod drive mechanism, control system components, rod

position indication components, and cooling system. Detailed operation experience data for 1980 to 1990

was evaluated to identify the predominant failure modes, causes, and effects. The results of this

evaluation, along with an assessment of component material and operating environment, lead to the

conclusion that both the B&W and CE CRD systems are susceptible to age degradation. Failures of the

CRD system have resulted in significant plant effects including power reductions, plant shutdowns,

scrams, and ESF actuations.

Information on current plant system inspection and maintenance practices were obtained from

two B&W plants, and four CE plants through an industry survey. The results of this survey indicate that

some plants have modified the system, replaced components, and established preventive maintenance

programs, some of which effectively address the aging issue, while others do not. The potential

application of some advanced monitoring inspection techniques are discussed.

iii
TABLE OF CONTENTS

Page

ABSTRACT ....................................................... iii

LIST OF FIGURES ....................................................... x

LIST OF TABLES ....................................................... xii

SUMMARY ........................................................ xv

ACKNOWLEDGEMENTS.................................................. xix

1. INTRODUCTION...................................................... 1-1

1.1 Background .............................................. 1-1

1.2 Objectives . .............................................. 1-2

1.3 Analysis Methodology and Report Format .......... ...................... 1-3

2. DESCRIPTION OF THE COMBUSTION ENGINEERING

CONTROL ELEMENT DRIVE SYSTEM . ................................... 2-1

2.1 Introduction . ........................................... 2-1

2.2 Control Element Assemblies ........................................... 2-1

2.3 Control Element Drive Mechanisms ..................................... 2-5

2.3.1 Magnetic Jack CEDM ........................................... 2-7

2.3.2 Rack and Pinion CEDM ......................................... 2-13

2.4 Control Element Assembly Rod Position Indication

System .................................................... 2-13

2.5 Control Element Drive Control System . .................................. 2-16

2.5.1 System Operation ............................................ 2-17

2.5.2 System Description ............................................ 2-19

2.6 CEDM Cooling System ........................................... 2-20

2.7 System 80 ........................................... 2-21

3. DESCRIPTION OF THE BABCOCK & WILCOX CONTROL

ROD DRIVE SYSTEM ................................. 3-1

3.1 Introduction ................................. 3-1

3.2 Control Rod Assemblies ........... ...................... 3-1

3.3 Control Rod Drive Mechanisms ................................. 3-6

3.3.1 Motor Tube ................................. 3-7

3.3.2 Closure Assembly ........... ...................... 3-9

3.3.3 Motor Assembly .......... ....................... 3-10

3.3.4 Rotor Assembly ............................................ 3-11

3.3.5 Thermal Barrier ...... ........................... 3-12

3.3.6 Leadscrew Assembly . ................................. 3-13

3.3.7 Torque Tube and Torque Taker ................................. 3-13

v
TABLE OF CONTENTS (Cont'd)

Page

3.3.8 CRDM Operational Considerations .............................. 3-13

3.4 Control Rod Drive Control System ................................ .. 3-15

3.5 Rod Position Indication ............................... 3-23

3.5.1 Absolute Position Indication System .............................. 3-23

3.5.2 Relative Position Indication System .............................. 3-23

3.6 Control Rod Drive Mechanism Cooling System .......... .. .............. 3-23

4. OPERATING AND ENVIRONMENTAL STRESSES ......................... 4-1

4.1 System Operating Stresses ......................................... 4-1

4.2 Environmental Stresses . ............................................. 4-2

4.3 Effect of Operating and Environmental Stresses on

System Components .............................................. 4-3

4.4 Aging Stressor Tables ............................................. 4-8

5. USNRC AND INDUSTRIAL STUDIES .............. .. ................... 5-1

5.1 USNRC Information Notices ....................................... 5-1

5.1.1 Information Notice No. 85-38 .................................. 5-1

5.1.2 Information Notice No. 85-86 .................................. 5-1

5.1.3 Information Notice No. 86-108 .................................. 5-2

5.1.4 Information Notice No. 88-47 .................................. 5-2

5.2 USNRC Research Programs .................... 5-2

5.2.1 NUREG-0641 ............. ................................. 5-2

5.2.2 NUREG/CR-2410, 3602, 3950 ................................... 5-4

5.2.3 NUREG/CR-4731 ........... ................................ 5-5

5.2.4 NUREG/CP-0091 .. ........................................ ; 5-7

5.2.5 USNRC SRP NUREG-0800 ................................... 5-7

5.2.6 Draft NUREG-1299 ......................................... 5-8

5.3 Industrial Research .............................................. 5-9

5.3.1 EPRI NP-3444 ............................................. 5-9

5.3.2 EPRI-4512 ........... ..................................... 5-10

5.3.3 EPRI NP-5001 ............................................. 5-10

5.3.4 EPRI NP-6232 ............................................. 5-12

5.4 Summary ...................................................... 5-12

vi
TABLE OF CONTENTS (Cont'd)

Page

6. OPERATIONAL EXPERIENCE (1980-1990) ................................ 6-1

6.1 Introduction ................................... 6-1

6.2 Failures Due to Aging ............. ...................... 6-2

6.3 Significant Plant Effects ..................... . 6-4

6.4 Sub-System Failures, Causes, and Effects .6-5

6.4.1 Cables and Connectors . 6-5

6.4.2 Control Rod Drive Control System Failures .6-7

6.4.3 Control Rod Drive Mechanism Failures ............................ 6-9

6.4.3.1 Primary Coolant Leakage .................................. 6-9

6.4.3.2 Stator Failures .6-12

6.4.3.3 Inoperable CRDMs .- 6-13

6.4.4 Rod Position Indication Failures. 6-14

6.4.5 Human Error. 6-16

6.4.6 Unknown Failure Cause .6-17

6.4.7 Miscellaneous Failures. 6-17

6.5 Summary. 6-19

7. EFFECT OF COMPONENT FAILURES ON CRD SYSTEM . .7-1

7.1 Control Rod Assemblies (B&W and CE) .7-1

7.2 Control Rod Drive Mechanisms .7-3

7.2.1 B&W Control Rod Drive Mechanism .7-3

7.2.2 Combustion Engineering Control Element

Drive Mechanisms .7-3

7.3 Control Rod Drive Control Systems .. 7-3

7.3.1 Babcock & Wilcox CRDM Control System ... 7-3

7.3.2 Combustion Engineering Control Rod Drive Control System .7-10

7.4 Rod Position Indication ........................................... 7-13

7.4.1 B&W Rod Position Indication .7-13

7.4.2 CE Rod Position Indication .7-13

7.5 Control Rod Drive Cooling System (B&W and CE) . .7-21

7.6 PRA Insights .7-21

7.6.1 Calvert Cliffs 1 (CE) .7-21

7.6.2 Arkansas 1 (B&W) .7-21

7.6.3 Oconee 3 (B&W). 7-22

vii
TABLE OF CONTENTS (Cont'd)

Page

7.6.4 Crystal River 3 (B&W) ....................................... 7-22

7.6.5 Conclusions . ............................................... 7-22

8. CRD INSPECTION, SURVEILLANCE, MONITORING AND

MAINTENANCE REVIEW ........ ........... 8-1

8.1 Babcock and Wilcox Plants .8-1

8.1.1 Preventive Maintenance and Inspection Practices .8-3

8.2 Combustion Engineering Plants .. 8-8

8.2.1 Preventive Maintenance and Inspection Practices .8-11

8.2.2 CED System Tests. 8-17

8.2.3 General .8-17

8.3 Summary of Survey Results (B&W and CE) .. 8-18

8.3.1 Cable Integrity. 8-18

8.3.2 CRDM Operation. 8-18

8.3.3 Primary Coolant Leakage .8-18

8.3.4 In-Service Inspection .8-19

8.3.5 Power and Control .8-19

8.3.6 Reliability Data Bases. 8-20

9. CONCLUSIONS AND RECOMMENDATIONS ........... .. ................. 9-1

9.1 Conclusions .................................................... 9-1

9.1.1 Combustion Engineering ...................................... 9-1

9.1.2 Babcock & Wilcox ........................................... 9-2

9.2 Recommendations ............................ 9-3

9.2.1 General Recommendations .................................... 9-3

9.2.2 Combustion Engineering ...................................... 9-3

9.2.3 Babcock & Wilcox ....... . .................................... 9-4

9.2.4 Advanced Monitoring Techniques ............................... 9-4

10. REFERENCES. ..................................................... 10-1

viii
TABLE OF CONTENTS (Cont'd)

Page

APPENDICES

A Summary of Licensee Event Reports for Combustion Engineering

Control Element Drive System ...................... A-1

B Summary of Licensee Event Reports for Babcock & Wilcox Control

Rod Drive System ............ B-1

C Industry Survey of Current CEDM Operating Experience, Inspection,

and Maintenance Practices ................... C-1

D Industry Survey of Current CRDM Operating Experience, Inspection,

and Maintenance Practices ................... D-1

ix
LIST OF FIGURES

Page

1.1 CE NPAR System Boundary ............................................ 1-4

1.2 B&W NPAR System Boundary .1-5

2.1 Combustion Engineering Full Length Control Element Assembly .2-2

2.2 Combustion Engineering Magnetic Jack Control Element Drive Mechanism .2-6

2.3 CEA Motor Housing Assembly .2-8

2.4 Motor Assembly .............. ....................................... 2-9

2.5 Upper Pressure Housing ............................................... 2-10

2.6 Extension Shaft Assembly .............................................. 2-11

2.7 Gripper and CEA Spider Hub ........................................... 2-12

2.8 Combustion Engineering Rack and Pinion CEDM ............................ 2-14

2.9 Combustion Engineering Reed Switch Position Transmitters ..................... 2-15

2.10 Combustion Engineering Reed Switch Position Transmitter (RSPT) Typical Circuit .... 2-16

2.11 Typical Pre-System 80 with ACTM Upgrade and Interfaces ...................... 2-18

2.12 Combustion Engineering Typical Power Switch ............................... 2-20

3.1 Babcock & Wilcox Control Rod Assembly .................................. 3-5

3.2 Control Rod Drive Mechanism .......................................... 3-7

3.3 CRDM Rotor Assembly and Leadscrew ...................... .... 3-8

3.4 CRDM Vent Valve Assembly and Closure Parts .............................. 3-8

3.5 Venting Tool . ........................................................ 3-9

3.6 Stator Water Jacket Assembly ........................................... 3-10

3.7 Partial Section Showing Roller Nuts Engaged ................................ 3-11

3.8 Leadscrew Guide Assembly ............................................. 3-12

3.9 Torque Tube Assembly ................................................ 3-14

3.10 Minimum Allowable RC Pressure vs. Temperature for Tripping or Operating CRD .... 3-15

3.11 Control Rod Drive Control System Power Diagram ........................... 3-16

3.12 Rotating Magnetic Field ............................................... 3-18

3.13 Control Rod Drive Regulating Power Supply Programmer ....................... 3-19

3.14 Reed Switches . ........................................................ 3-21

3.15 Absolute Position Indication (API) ........................................ 3-22

3.16 Relative Position Indication ............................................. 3-24

5.1 Combustion Engineering Guide Tube Sleeve ................................ 5-3

6.1 CRD Sub-systems Database Comparison ................................... 6-3

6.2 Percentage of Component Failures Attributable to Aging ....................... 6-3

6.3 Significant Plant Effects Resulting From CRD System Degradation .... ............ 6-4

6.4 Cables and Connectors Failure Cause ..................................... 6-6

6.5 Cables and Connectors Failure Effect ..................................... 6-6

6.6 CRD Control System Failure Causes ...................................... 6-8

6.7 CRD Control System Failure Effect ...................................... 6-8

6.8 Control Rod Drive Mechanisms Failure Cause .............................. 6-10

6.9 Inspection Area for Boric Acid Deposits (B&W Plants) ........................ 6-11

6.10 B&W CRDM Vent Valve .............................................. 6-12

x
LIST OF FIGURES (Cont'd)

Page

6.11 Control Rod Drive Mechanism Failure Effect ............................... 6-13

6.12 Rod Position Indication System Failure Cause ........ ......................... 6-15

6.13 Rod Position Indication System Failure Effect ................... I... ........ 6-15

6.14 CE Control Element Drive System Unknown Failures Effect ..... ............... 6-16

6.15 B&W Control Rod Drive System Source of Loose Parts ........................ 6-18

8.1 CRDM Flexible Hose and Quick Disconnect Couplings ......................... 8-4

8.2 CE Coil Current Trace ................................................ 8-13

xi
LIST OF TABLES

Pag

1.1 Combustion Engineering Plants in NPAR Study ....... . ......... 1-2

1.2 Babcock & Wilcox Plant in NPAR Study .1-2

1.3 CRD Primary Sub-systems .1-3

2.1 Primary Functions of Major CEDM Sub-Systems .2-3

2.2 CEA and CEDM Plant Data .2-5

2.3 Control Element Design Data .2-7

2.4 Magnetic Jack Type CEDM .2-8

2.5 Combustion Engineering Control Element Drive Control Systems .2-17

3.1 Primary Functions of Major CRDM Sub-Systems .3-2

3.2 CRA and CRDM Plant Data .3-4

3.3 Control Rod Assembly Data .3-4

3.4 Axial Power Shaping Rod Assembly Data .3-6

3.5 CRA Grouping by Plant .3-17

4.1 Babcock & Wilcox Control Rod Drive System Potential

Degradation Mechanisms and Failure Modes .4-4

4.2 Combustion Engineering Control Element Drive System

Potential Degradation Mechanisms and Failure Modes .4-5

4.3 Babcock & Wilcox CRD System Stress Summary .4-9

4.4 Combustion Engineering CED System Stress Summary .4-10

5.1 Events Involving Hold Down Springs .5-4

5.2 Cause of Events Involving Control Rod Drive Mechanisms .5-5

5.3 Summary of PWR CRDM Degradation Processes .5-6

5.4 Results of CEDM Wear Tests .5-9

5.5 Cobalt Release from CEDMs by Plant .5-10

5.6 Control Rod Drive Failure Statistics .5-11

5.7 Summary of Combustion Engineering Control

Element Drive Aging Analyzed by NRC and Industry .513

5.8 Summary of Babcock & Wilcox Control Rod Drive

Aging Analyzed by NRC and Industry .5-14

7.1 FMEA-Babcock & Wilcox and Combustion Engineering

Control Rod Assemblies .7-2

7.2 FMEA-Babcock & Wilcox Control Rod Drive Mechanisms .7-4

7.3 FMEA-Combustion Engineering Control Element Drive

Mechanisms. ................................................... 7-6

7.4 FMEA-Babcock & Wilcox CRD Control System .7-8

7.5 FMEA-Combustion Engineering CRD Control

System ...................................................... 7-11

Xii
LIST OF TABLES (Cont'd)

7.6

7.7

7.8

7.9

8.1

8.2

8.3

8.4

8.5

8.6

8.7

8.8

A-l

A-2

A-3

A-4

A-5

A-6

A-7

FMEA-Babcock & Wilcox Absolute Position Indication

System ..........................................

FMEA-Babcock & Wilcox Relative Rod Position

Indication System .

FMEA-Combustion Engineering Reed Switch Position

Indication System .................................

FMEA-Combustion Engineering Pulse Counting Position

Indication System ..................................

Babcock & Wilcox System and Component Description ......

Major Sub-system Modifications - B&W Plants ............

Summary of Preventive Maintenance and Inspection

Practices - B&W Plants .............................

B&W Recommended Annual Stator Preventive Maintenance . .

Combustion Engineering System and Component Description . .

CED System Design Changes and Modifications ...........

CE Recommended Annual CEDM Preventive Maintenance ...

Summary of CE Plant Preventive Maintenance and Inspection

Practices ........................................

CE Cable and Connector LERs .......................

CE Control System LERs ............................

A-2.1 Control Element Drive System (CEDS) ..................

A-2.2 Control Element Drive Mechanism Control System (CEDMCS)

7-14

7-16

7-18

8-1

8-3

8-5

8-7

8-9

8-12

8-13

8-15

. . . . . .

. . .

. . .

CEDM LERs ...

CE Rod Position I

Human Error LET

Unknown Cause .

Miscellaneous LE]

.....................................

IERs .................................

ts ...................................

.....................................

Ks ...................................

A-2

A-2

A-3

A-5

A-6

A-8

Ail

A13

AM5

B-1

B-2

B-3

B-4

B-5

B-6

B-7

B&W Cable and Connector LERs

B&W CRD Control System LERs

CRDM LERs ...............

Rod Position Indication LERs . .

Human Error LERs ..........

Potential Loose Parts LERs ....

Unknown Cause .............

B-2

B-3

34

B-5

B-6

B-7

B-8

xiii
SUMMARY

The Babcock & Wilcox (B&W) and Combustion Engineering (CE) control rod drive (CRD)

systems consist of mechanical and electrical components that position the control rod assemblies in the

core in response to automatic or manual reactivity control signals. Both systems are designed to allow

rapid gravity insertion of the control rods upon removal of the ac power which holds the rods. This study

examines the design, materials, maintenance, and operation of the system to assess the potential for age

degradation.

The boundaries for this CRD system study included the control system, rod position indication

system, control rod drive mechanisms, and the control rod drive cooling systems. The fuel assembly

guide tubes,the upper internal support structures, and the control rods were also included since failure

of these components could also preclude rod insertion.

Both the CE and B&W control rod drive mechanisms are flange mounted on top of the reactor

vessel head. B&W plants use a roller nut/leadscrew design, while the majority of CE plants use the

magnetic jack design. Externally mounted stator coils provide the magnetic field which activates the

roller nuts or magnetic latches resulting in control rod movement. Two CE plants, Palisades and Fort

Calhoun, use the rack and pinion mechanism, in lieu of the magnetic jack type, which uses an electric

motor to drive the rack and pinion mechanisms. Both mechanisms uses similar magnetically actuated

reed switches to indicate the actual position of the rod. The CRDM stators for B&W are water cooled,

while CE uses a forced air cooling system.

A detailed operating experience review of three commercially available databases (Licensee Event

Report Database, Nuclear Plant Reliability Data System, and Nuclear Power Experience), together with

NRC and industry research highlighted age related component degradations and failures which

significantly affected plant operations. These effects included power reductions, reactor shutdowns and

scrams, and Engineered Safety Feature actuation. Neither the B&W nor the CE CRD systems ever

failed to allow for gravity insertion of the control rods. However these component failures and

degradations resulted in increased component stresses and unnecessary thermal and pressure cycleswhich

challenged the other plant systems.

The majority of failures of the CE control element drive system were caused by the degradation

of the control system (61%). Failures of the control rod drive mechanisms accounted for 60% of

reported failures of the B&W control rod drive system. Aging was the direct failure cause for 40% of

the CE power and control system and 55% of the B&W control rod drive mechanism..

The following main failures were highlighted by this operational experience review:

1. Primary coolant leakage resulting from aged flexitallic gaskets (B&W), leaking vent

valves (B&W and CE), degraded rotating seals (CE), and pressure housing cracks (CE).

These occurrences resulted in component degradation due to the corrosive nature of the

boric acid contained in the primary coolant. B&W has upgraded their flexitallic and vent

valve designs in response to these failures.

2. Failures of the CRD control system resulting in dropped control rod assemblies,

primarily due to power supply failures, sluggish gripper operation, and other electronic

component failures (diodes, SCRs, breakers). The CRD systems at five CE plants have

been upgraded (four totally, one partially) to incorporate microprocessors and current

xv
sensors which monitor CRDM mechanical actuation in order to control voltage/current

sequencing during rod movement. These upgrades also monitor for abnormal current

levels and take corrective action. All CE utilities have redundant logic power supplies

for the CRD system. The microprocessor upgrade have eliminated the rod drops

attributed to sluggish gripper operation.

3. Erroneous rod position indication signals have been generated, primarily due to reed

switch failures. Both CE and B&W have improved their systems to permit-continued

operation with some failed reed switches, while B&W has also upgraded the reed switch

design.

4. Degraded electrical cables and connections resulted in inoperable control rods.

Electrical cable aging, primarily due to environmental degradation, was observed at

several plants.

5. Human errors and inadequate system maintenance resulted in dropped rods, power

reductions, reactor shutdowns and ESF actuation.

6. Loose parts in the reactor core were reported, representing a significant hazard to the

insertion of the control rods. The loose parts resulted from the deterioration of internal

core components and handling equipment.

The review of the failures also indicated that a significant amount of reported failures which

either had an unknown or no failure cause defined. This omission represents a lack of proper analysis

of root failure cause and has resulted in similar, repeat failures which might have been prevented if the

actual failure cause was determined.

The operating and environmental stresses for the system, and the potential aging effects from

continued exposure to these stresses were evaluated for the major system components. Detailed Failure

Mode and Effects Analysis were performed for the subsystems. The relative potential of each failure

cause being the result of aging was also assessed. The results obtained closely coincide with the results

seen from the operating experience review.

A survey was made of the surveillance, inspection, monitoring, and maintenance practices of

utilities. Responses were received from two B&W plants, and four CE plants (representing eight units).

The results from this survey, when compared to the operating experience review, highlighted several areas

which indicate the need for increased attention:

1. Operating utilities need to establish a reliability program which includes accessing one

(or more) of the operating experience databases. A more efficient predictive

maintenance program would result in data which would alert utility personnel to failures

at other plants, and allow corrective action to be taken before the component failed.

2. A more aggressive program to identify primary coolant leakage should be established.

Components which have demonstrated a susceptibility to failure which may result in

leakage (seals, gaskets) should be inspected and replaced on a scheduled basis prior to

failure.

xvi
3. The current ISI program applicable to CRDMs requires that only 10% of the peripheral

housings be inspected every ten years. With the continued instances of CRDM housing

defects and failures, modification of this requirement should be considered to also

include interior mechanisms.

4. Commercial advanced system monitoring and -inspection techniques capable of detecting

and trending time related age degradation should be evaluated for use with the CRD

system. These techniques include infrared thermography for electronic components,

motor current signature analysis to detect proper CRDM operation, and Electronic

Characterization and Diagnostics (ECAD) as a possible alternative to meggering for

assessing electrical integrity.

The results of this NPAR study show that aging degradation and failures have occurred in both

B&W and CE Control Rod Drive Systems. These occurrences have not prevented the gravity insertion

of the control rods. However, CRD failures have resulted challenges to other plant safety systems. As

the survey of utility practices indicates, control rod drive aging has been recognized and is being

addressed, to varying degrees, by the utilities' inspection and maintenance programs. However, aging

degradation and failures are still occurring. The results of this study highlight these areas, and provide

recommendations on preventive and predictive maintenance which may reduce aging failures.

xvii
ACKNOWLEDGEMENTS

The authors wish to acknowledge the technical direction of the NRC Program Manager, Mr.

Satish K. Aggarwal. We also would like to express our appreciation to Dr. George Sliter of EPRI and

Mr. Russ Bell of NUMARC for their assistance in obtaining plant responses to the operating experience

questionnaire.

We would. like to express our gratitude to various members of the Engineering Technology

Division of BNL, including Mr. Robert Hall, Mr. John Taylor, Dr. Mano Subudhi, Mr. James Higgins,

Dr. See-Meng Wong, and Mr. Kenneth Sullivan for providing technical assistance, guidance, and for

reviewing this report. We also especially appreciate the efforts of Mr. J. H. Taylor of B&W, Mr. H.

Windsor, and Mr. D. Matteson of ABB-CE, and Mr. H. Wilson of GPU, and staff for their time,

assistance, and interest in this effort.

Finally, we wish to thank Mrs. Sonja Santos for her help in the data presentation, and Ms.

Patricia Ennis for her patience and excellent preparation of this manuscript. We also acknowledge the

Nuclear Safety Library Staff at BNL for their assistance in obtaining the support documentation for this

report.

xix
1. INTRODUCTION

Control rod drive mechanisms and control rod assemblies are essential to the safe and reliable

operation of nuclear power plants. The individual control rods contain neutron absorbing material

(poison) to control fast reactivity transients and to produce a rapid reactor shutdown (scram) upon

demand. Control rod drive control systems and drive mechanisms provide for the insertion or withdrawal

of the control rods from the reactor core, including control of sufficient holding power to maintain the

control rods stationary during reactor operation.

Failure or degradation of these components, due to aging, may significantly affect plant operation

(scrams, power reductions, dropped rods). Any failure which prevents or obstructs the rapid insertion

of the control rods into the core upon demand represents a significant increase in plant risk.

An aging assessment has been performed for Combustion Engineering (CE) and Babcock &

Wilcox (B&W) control rod drive mechanisms and associated sub-systems. The results of this Phase I

assessment are described in the following sections of this NUREG. This program was performed under

the United States Nuclear Regulatory Commissions' (NRC) Nuclear Plant Aging Research (NPAR)

Program.

1.1 Background

Both Combustion Engineering (CE) and Babcock & Wilcox (B&W) control rod drives are

electromechanical devices which are flange mounted on top of the reactor pressure vessel head. The

CE Control Element Drive Mechanisms (CEDM) use magnetically actuated latches which engage a

notched drive shaft to position or hold the attached Control Element Assembly (CEA). In lieu of this

design, two CE plants use a rack and pinion CEDM to perform the same functions. The B&W Control

Rod Drive Mechanism (CRDM) uses a magnetically actuated, rotating roller nut assembly which engages

a threaded leadscrew to position or hold the attached Control Rod Assembly (CRA). All three designs

are fail safe since the removal or loss of electrical power will cause the latches (CE), pinion (CE), or the

roller nuts (B&W) to disengage, allowing the CRA or CEA to insert freely into the core under the

influence of gravity.'

Fifteen plants use the control element drive systems designed by Combustion Engineering (Table

1.1). The years of operation vary from the newest plant, Palo Verde 3 with 4 years, to Palisades with 20

years. All CE plants except two (Palisades and Fort Calhoun) use the magnetic jack type of CEDM.

The total number of CEDMs per plant is a function of reactor size and the number of CEAs, varying

from 37 to 91.

Eight B&W plants (Table 1.2) use the roller nut CRDM. Though B&W has fewer plants than

CE, the plants are older, varying from Davis-Besse with 14 years of operation to Oconee 1 with 18 years.

Typically, B&W plants use 61 CRDMs.

aFor brevity, the general term Control Rod Assembly (CRA) and Control Rod Drive Mechanism

(CRDM) will be used in this report to refer to both B&W and CE designs when discussion is equally

applicable to both designs. When specifically applicable to CE, the terms Control Element Assembly

(CEA) and Control Element Drive Mechanism (CEDM) will be used.

1-1
Table 1.1 Combustion Engineering Plants in NPAR Study

Years of Operation

<5 5-10 10-15 15-20 >20

Palo Verde 3 Palo Verde 1 Arkansas 2 Calvert Cliffs 1 Palisades

Palo Verde 2 St. Lucie 1 Calvert Cliffs 2

San Onofre 2 Fort Calhoun

San Onofre 3 Maine Yankee

St. Lucie 2 Millstone 2

Waterford 3

Table 1.2 Babcock & Wilcox Plants in NPAR Study

Years of Operation

10-15 15-20

Crystal River Arkansas 1

Davis Besse Oconee 1

Oconee 2

Oconee 3

Rancho Seco

Three Mile Island 1

1.2 Objectives

As reactor years of operation increased, a need developed to assess the effects of plant aging on

safety. The Director of the Office of Nuclear Reactor Regulation of the U.S. Nuclear Regulatory

Commission (NRC) identified this need, and the Nuclear Plant Aging Research (NPAR) Program was

developed by the Office of Nuclear Regulatory Research to assess this. The technical and safety issues

of the Program, components and systems to be evaluated, and potential uses of the results, are described

in NUREG-1144.1

The objectives of this Phase I system study are described in NUREG 1144 and the BNL Aging

and Life Extension Assessment Program (ALEAP) Systems Level Plan.2 Specifically, these objectives

are to perform the following:

* a detailed evaluation of operating experience data,

* an analysis of industry operating and maintenance information,

* an identification of failure modes, causes, and effects, and

* a review of design, operating environment, and performance requirements.

1-2
To meet these objectives, which are similar to the Westinghouse CRD aging study completed by

BNL,3 the following tasks were completed for both the B&W and CE CRD systems:

A. The operating experience was reviewed to identify the dominant component failure

modes, effects, and mechanisms.

B. A Failure Modes and Effects Analysis (FMEA) for each main sub-system was completed

to identify the components which affect the functions of the system.

C. A survey of Babcock & Wilcox and Combustion Engineering utilities was performed to

obtain current maintenance, inspection, and surveillance practices. Meetings were held

with B&W and CE to assess the recommended maintenance and operating restrictions.

1.3 Analysis Methodolopiy and Report Format

Reactivity control for CE and B&W reactors is provided by two independent systems. This

function is accomplished by the Chemical and Volume Control System (CVCS) and the Control Element

Drive System (CEDS) in CE reactors, and the Makeup and Purification System and Control Rod Drive

System (CRDS) in B&W reactors.

Both the CVCS and the Makeup and Purification System compensate for long-term reactivity

effects due to coolant temperature changes, xenon concentration, and fuel burnup by controlling the

amount of soluble boron in the reactor coolant. Although important to reactivity control, these systems

are not included in the scope of this study.

The CEDS and CRDS position the movable control rod assemblies within the core to control

the short term reactivity effects. Each system is capable of producing a reactor shutdown through the

rapid, gravity insertion of the control rods. Both the CEDS and CRDS are comprised of five similar,

primary sub-systems (Table 1.3). These five sub-systems encompass the CE and B&W system boundaries

used for this aging study, as shown in Figures 1.1 and 1.2, respectively.

Table 1.3 CRD Primary Sub-systems

Combustion Engineering Babcock & Wilcox

Control Element Drive System Control Rod Drive System

I. a. Control Element Assemblies (CEA) 1. a. Control Rod Assemblies (CRA)

b. Fuel Assembly Guide Tubes b. Fuel Assembly Guide Tubes

c. Upper Guide Structure c. Upper Internal Guide

II. Control Element Drive Mechanism II. Control Rod Drive Mechanism

(CEDM) (CRDM)

III. Control Element Drive Control System III. Control Rod Drive Control System

IV. Control Element Assembly Position IV. Control Rod Assembly Position

Indication Indication

V. Control Element Drive Mechanism V. Control Rod Drive Mechanism

Cooling System Cooling System

1-3
V. CEOM

Cooling

| System\ 11. Control Element|

Drive Mechanisms l

III. Control Element

\ ~~~~~~~~~~~~~~~~~~~~Drive ControlI

IV. CE PositionDrive Systemro

Indication CEDM Nozzle Sse

System l __J

la. Control Element IC. Upper Guide

Assembly Fully Structure

Withdrawn

Reactor Regulating Reactor Trip Plant/Reactor

System Switch Gear Protection

I I System

Reactor Power Motor Generator

Cutback System Set ------ ---- tb. Fuel Assembly

Guide Tubes

Systems Boundary

!_ . -- . . . --- . For NPAR Study

Figure 1.1 CE NPAR System Boundary

The Reactor Protection System (RPS), including the reactor trip breakers, are essential to plant

safety. Reactor trip signals from the RPS cause the breakers to open, removing power from the drive

mechanisms, allowing the control rod assemblies to rapidly insert into the core under the influence of

gravity. This vital system has been addressed in a separate NPAR study.44

To fully understand the effect of system aging, specific information on operating characteristics,

material and design function is presented in Section 2.0 for CE and Section 3.0 for B&W. This

information was obtained from a review of the utilities' Final Safety Analysis Reports, technical reports

and system descriptions.

Section 4.0 evaluates the operational and environmental impacts of the stresses on the

system and its components. The effect of required testing is considered, along with obvious stresses

including mechanical wear, vibration, and electrical stresses. The dominant stresses which affect the

primary components of the system are also presented.

Both the B&W and CE control rod drive systems have been the subject of industry, EPRI, and

NRC studies. Section 5.0 summarizes this work including the four Information Notices generated in

response to significant system and component operating failures.

1-4
-IV. CRD Absolute Position - Indication System

II. Control Rod In. Control Rod Drive i Reactor

l Drive Mechanism Mechanism Power and | Protection

la. Control Rod CoControl System System Assembly HiV.CRDCoolng

v g\3L WaterSystem

I t Ic.Upper BE W Internal |

(-i-I F ee zControl Rod Integrated Control I /t '4 _ v :i Guide Tube ! System

Ilb. Fuel Assembly Manual Control

l |1 4 t ] / Reactor Vessel

Systems Boundary

For NPAR Study

Figure 1.2 B&W NPAR System Boundary

Operating experience for each design, for 1980-1990, is presented in Section 6.0. The information

used to evaluate the operating experience was obtained from a variety of sources, including:

* Nuclear Plant Reliability Data System (NPRDS)

* Licensee Event Reports (LERs)

* Nuclear Plant Experience (NPE)

* Plant Specific Failure Data

* Operating Plant Surveys

* Meetings with B&W and CE System Engineers

This section discusses the primary failure causes and effects for the main system components. The

percentage of failures directly attributed to aging degradation is also presented. Summaries of the

individual LERs are given in Appendix A and B. The systems' susceptibility to human error and

improper maintenance is also evaluated.

The results of the detailed design, operating stressors, previous studies on this subject, and

operating experience reviews are combined into a failure mode and effects analysis for the primary CRD

subsystems (Section 7.0). Each individual FMEA summarizes the main component failures which result

in system or plant effects. The aging potential for each failure cause is evaluated. Based upon plant

1-5
operating experience and engineering judgement, the probability of occurrence for each failure is

qualitatively assessed.

Section 8.0 presents the responses of operating utilities to the BNL industry survey performed

with the assistance of EPRI and NUMARC. Information on system operating experience, inspection,

surveillance, and maintenance practices is presented. The effectiveness of each in mitigating the effects

of aging (discussed in Section 7.0), is assessed and the benefits of advanced monitoring techniques is

presented.

In Section 9.0, the results and conclusions of this Phase I aging assessment are presented.

1-6
2. DESCRIPTION OF THE COMBUSTION ENGINEERING CONTROL ELEMENT DRIVE SYSTEM

2.1 Introduction

The Combustion Engineering control element drive system is comprised of five main sub-systems

as shown in Figure 1.1. These specific sub-systems include:

I. Control ElementAssemblies (CEAs): Each CEA consists of individual absorber rods connected

to a common hub and positioned by fuel assembly guide tubes when inserted in a fuel assembly.

The upper guide structure maintains CEA spacing when withdrawn from the fuel assembly into

the upper plenum region.

II. Control Element Drive Mechanisms (CEDMs): Electro-mechanical devices, which in response

to automatic or manual control signals, inserts, withdraws or holds the CEAs stationary in the

reactor core.

III. Control Element Drive Control System: Provides electrical signals to either hold the CEAs

stationary in the core, or reposition them, to control reactivity during reactor operation.

IV. Control Element Assembly Rod Position Indication System: Actual CEA position is monitored

by magnetically operated reed switches. The plant computer monitors and counts the pulses

supplied to CEDM coils to provide a indirect (demanded) rod position.

V. CEDM Cooling System: A forced air cooling system which maintains the CEDM coil stack

assembly below 3500F.

Table 2.1 summarizes the main components and primary functions for each of the five sub-systems.

2.2 Control Element Assemblies (CEA)

The Control Element Assemblies consist of four, five, or twelve neutron absorber element rods

connected to a spider assembly. The spider assembly (Figure 2.1) geometrically arranges the rods to

ensure engagement with the fuel assembly guide tubes and is coupled to the CEDM drive shaft.

Depending upon specific core design requirements, both full length and part length absorber rods are

used as shown in Table 2.2. The physical dimensions of these two rod types are similar, with the

exception of the poison column length. Table 2.3 summarizes the design data for both of these rods.

The total number of fuel assemblies in the reactor core depends upon plant size, varying from 133

to the new System-80 plants with 241 assemblies. Each fuel assembly consists of fuel rods and Zircaloy-4

guide tubes arranged in a square lattice. The lower ends of the four outer guide tubes are tapered

gradually to form a region of reduced diameter, which in conjunction with the absorber rod, forms a

hydraulic buffer. This reduces the deceleration loads on the CEA at the end of the trip stroke. The

hydraulic damping action is augmented by the spring and plunger arrangement on the spider.

When fully inserted, the CEAs rest on the upper guide structure support plate.

During normal operation, the control elements are withdrawn from the core into the upper plenum

region. In this region, CEAs are enclosed in shrouds which are part of the upper guide structure

assembly. The shrouds maintain the proper positioning of the CEAs and protect them from coolant

cross-flow when withdrawn from the fuel assemblies.

2-1
Top View

Side View

-Identification

Letters for Part-Length CEAs

Numerals for Full-Length CEAs

-Spring

'Poison Rod

Assembly

-Poison Material

132" Full-Length CEAs

I I I I L..i. I 34" Part-Length CEAs

Figure 2.1 Combustion Engineering Full Length Control Element Assembly'

2-2
Table 2.1 Primary Functions of Major CEDM Sub-Systems

Sub-System Component Primary Functions

I. Control Element A. Control Element Rods Provide reactivity control during normal operation.

Assemblies Effect rapid reactor shutdown.

Provide sealed, pressure boundary containing poison material, and sufficient

free volume to accommodate irradiation induced poison swelling and fission gas production.

Provide envelope within each fuel assembly to maintain control element rod spacing

B. Fuel Assembly Guide Tubes and permit insertion of control element rods.

Provide hydraulic snubber action to absorb kinetic energy of the CEA during scram.

Maintain CEA spacing when withdrawn from fuel assembly. C. Upper Guide Structure Protect CEA from coolant cross flow effects in upper plenum.

11. Control Element Drive A. Motor Housing Assembly Serve as primary coolant pressure boundary. Mechanisms Provide free volume to house internal CEDM components (motor assy, driveshaft)

and support coil stack assembly.

B. Motor Assembly Provide linear motion to CEA through proper engagement of latches with drive shaft.

C. Upper Pressure Housing Assembly Provide sufficient free volume to allow for complete withdrawal of drive shaft.

Provide means to vent mechanism after system filling and prior to hydrostatic test or operation.

D. Extension Shaft Assembly Provide circumferentially grooved drive shaft, which when engaged by latches, raises or lowers CEA.

Provide a means of attaching CEA to lower end of extension shaft through

individual gripper fingers.

Provide attachment for permanent magnet to actuate reed switches.

E. Coil Stack Assembly Provide magnetic force necessary to actuate motor assembly mechanical latches for

engaging and driving the CEA extension shaft.

to)
Table 2.1 Primary Functions of Major CEDM Sub-Systems (Cont'd)

[ Sub-System Component Primary Funclons

III. CEDM Control System A. Control Element Drive Control Provide motive power to the CEDM coils or rack and pinion motor.

System and Control Power

Programmer Allow for gravity insertion of CEAs upon removal of CEDM power (by manual or automatic

trip of Reactor Trip Switch Gear)

B. Control Element Drive Mechanism

Control System Provide CEA position indication via pulse counts.

Accept signals from reactor protection and monitoring systems.

C. Rack and Pinion Control System Provide signals to monitoring systems.

IV. CEA Rod Position A. Reed Switch Position Transmitter Provide position signals to monitoring systems and optionally installed protection systems (Core

Indication Assembly Protection Calculator System).

B. Pulse Count Position Indication Provide indirect (demanded) indication of CEA position from Control Element Drive Control

System System through monitoring pulses supplied to CEDM coils.

V. CEDM Cooling System Fans, Cooling Shroud, Instrumentation Provide forced air cooling to CEDM coils to maintain temperature below 350°F.
Table 2.2 CEA and CEDM Plant Data

Number of Full Number Part

Length Control Length Control Number Number

Plant Element Assy. Element Assy. Single CEAs Dual CEAs CEDM Type|

Arkansas 2 73 8 81 0 MJ

Calvert Cliffs 1 77 0 37 20' MJ

Calvert Ciffs 2 77 0 37 20' Mi

Fort Calhoun 49 0 25 12' RP l

Maine Yankee 77 0 37 20' Mi

Millstone 2 73 8 57 12' MJ

Palisades 41 4 45 0 RP

Palo Verde 1 76 13 41 482 . M

Palo Verde 2 76 13 41 482 Mi

Palo Verde 3 76 13 41 482 MJ

St. Luie 1 73 8 57 12' Mi

St. Lucie 2 83 8 91 0 MJ

San Onofre 2 83 8 91 0 Mi

San Onofre 3 83 8 91 0 Mi

Waterford 83 8 91 0 Mi

'Dual CEAs consist of two single CEAs connected to a single extension shaft.

2

System 80 plant, CEAs consist of 12 rods spanning four fuel assemblies.

3

MJ: Magnetic Jack CEDM - RP: Rack and Pinion CEDM

2.3 Control Element Drive Mechanisms

The Control Element Drive Mechanisms position the CEAs within the reactor core to control

reactivity, or effect a rapid reactor shutdown by allowing the CEAs to quickly drop into the core during

a scram. The CEDMs are electromechanical devices which provide controlled linear motion to the

attached CEA in response to operating signals received from the CEDM Control System. Each

mechanism is capable of withdrawing, inserting, holding, or releasing/dropping a CEA from any position

in the core. Typically, CE reactors use the magnetic jack CEDM as shown in Figure 2.2. Table 2.4 lists

the individual component materials for this mechanism. Two plants, in lieu of this type of CEDM, use

a rack-and-pinion mechanism. Section 2.3.2 briefly describes this CEDM.

2-5
REED SWITCH ASSY ' BALL SEAL HOUSING

EXTENSION SHAFT ASSY

ASSY

rll - . EC V 1 f

L__

REED SWITCH ASSY

MOTOR ASSYCOIL STACK ASSYDRIVE SHAFT

COOLING SHROUD -

UPPER PRESSURE

HOUSING

-OMEGA SEAL

-MOTOR HOUSING

GRIPPER ASSY

OMEGA SEAL REACTOR VESSEL

NOZZLE

Figure 2.2 Combustion Engineering Magnetic Jack Control Element Drive Mechanism'

2-6
Table 23 Control Element Design Data'

Control Element | Full Length | Part Length ]

Number (Typ) 73 8

No. of Elements Per Assy 5 5

Clad Material Inconel 625 Inconel 625

Clad Thickness (In.) .035 .035

Clad O.D. (In.) .816 .816

Center Element

-Poison Material B4C Inconel 625/

Water/B4 C

*Length (Typ) (In.) 146 75/58/16

Corner Elements

*Poison Material B4C/Ag-In-Cd Inconel 625/

Water/B4C

*Length (In.) 135.5/12.5 75/58/16

Fuel Assembly Guide Tube

* Material Zircaloy-4 Zircaloy-4

-ID (In.) .90 .90

2.3.1 Magnetic Jack CEDM

The CE magnetic jack CEDM has a forty year design life. It is designed to operate without

maintenance for one and one half years, and without component replacement for a minimum of three

years. The main components of the magnetic jack CEDM are:

a) Motor Housing Assembly: The two primary functions of this assembly are to (1) serve as the

primary coolant pressure boundary which houses the internal CEDM components, and (2)

locate and support the gripper coils. This assembly is a Type 403 stainless steel hollow tube

with internally threaded upper and lower end fittings. The lower Inconel end fitting screws

down upon the reactor head nozzle and is omega seal welded. An omega seal is a particular

flange design, which when joined with a mating surface, resembles the greek omega symbol ((1).

The central portion of the motor housing assembly has an external "S" contour, (Figure 2.3),

which allows the gripper coils to be located closer to the internally mounted motor assembly.

The Type 348 stainless steel upper end fitting threads into the upper pressure housing.

b) Motor Assembly: The motor assembly (Figure 2.4) is internally pinned to the motor housing

assembly, and when actuated by the coil stack assembly, provides linear motion to the CEA

2-7
fi,_ , 1. _ .. _M- )

External

"S" Contour

Or" I

'MI'l 1 1 i  .1 , - :: -=WV=02

Figure 2.3 CEA Motor Housing Assembly

Table 2.4 Magnetic Jack Type CEDM'

Component Material

1) Motor Housing Assembly

*Pressure Housing Type 403 Stainless Steel

*End Fittings Nickel-Chromium-Iron Alloy

2) Upper Pressure Housing Type 316 Stainless Steel

*Vent Valve Seal Type 440 Stainless Steel

Ball/Type 316 Stainless Steel Seat

3) Motor Latches, Links, Pins High Cobalt Alloy

4) Motor, Extension Shaft Springs Inconel X-750

5) Motor Magnet Type 410 Stainless Steel

6) Motor Fasteners Type 304 Stainless Steel

7) Extension Shaft Type 304 Stainless Steel

8) Extension Shaft Magnet Alnico No. 5

9) Motor and Extension Shaft Wear Surfaces Chromium Plated

10) Magnet Coils Copper wire insulated with high temp. enamel,

vacuum impregnated with high temp. varnish.

Fiberglass taped and encapsulated with

silicone.

11) Coil Housings Nickel Plated Carbon Steel

2-8
Figure 2.4 Motor Assembly

through two sets of grippers concentrically located around the drive shaft. The upper grippers

position the drive shaft, while the lower grippers transfer the weight of the shaft from the upper

grippers during movement. When the coils are energized, the sliding magnets move, camming

a linkage, which allows the latches to engage the drive shaft.

c) Upper Pressure Housing Assembly: This assembly comprises the upper half of the CEDM

(Figure 2.5). The three main purposes of the assembly are to 1) enclose the extension shaft

assembly, 2) allow for the complete withdrawal of the drive shaft from the core, and 3) allow

for CEDM venting and flushing via a double closure fitting. The upper end of the Type 348

stainless steel assembly contains a double closure fitting for venting and flushing the CEDM

housing. The double closure on the upper end has an internal ball (Type 440 stainless steel)

and seat seal (Type 316 stainless steel) with an internal stem to keep the ball and seat seal

closed. The ball and seat seal subassembly has an omega seal welded end cap over it, which

utilizes an o-ring to prevent seal leakage. Venting and flushing of the housing is accomplished

through the ball seat end fitting. The lower portion of the assembly is threaded, and omega

seal welded to the top of the motor housing.

d) Extension Shaft.Assembly: The three primary functions of this assembly are (1) to mate with

the motor assembly latches to raise or lower the CEA, (2) to couple the CEDM with the CEA

spider hub, and (3) to actuate the reed switches when the CEA is moved (Figure 2.6). This

assembly is a Type 304 stainless steel rod comprised of the following sub-assemblies:

1) expandable collet,

2) plunger,

3) gripper assembly,

4) drive shaft,

5) operating rod,

6) extension sleeve, and

7) magnet assembly.

The expandable collet and plunger connects the operating rod to the spider assembly hub. The

gripper assembly consists of an expandable collet with notched fingers and a spring-loaded

inner-tapered plunger. In the coupled position, the plunger forces the notched fingers radially

outward, locking it into the internal grooves on the CEA hub (Figure 2.7). The permanent

magnet assembly on the top of the drive shaft activates the reed switches to provide position

indication.

2-9
kb .~Steci Ball

I0

liIIII I

! I

i

i

I

11

I

I

II

I

I

Vent Stern

0 Ringt

I otns ing N't

I,

fdwpl-- I- ?r, i

0

Figure 2.5 Upper Pressure Housing

2-10
-TOP

EXTENSION

SHAFT

Plunger \ I

Expandable

Collet

Figure 2.6 Extension Shaft Assembly

2-11
" pelas Shrft dripe

~ttM DMIve Shaft T-eth

i-, 0'_ N

N64-IL:

I4

Figure 2.7 Gripper and CEA Spider Hub

e) Coil StackAssembly: The coil stack assembly consists of five DC coils installed externally to

the pressure housing. When powered, these coils activate the mechanical latches of the motor

assembly, resulting in CEA movement. These coils are functionally identified as the:

1) lift coil,

2) upper gripper coil,

3) pull down coil,

4) load transfer coil, and

5) lower gripper coil.

The pull down coil has been eliminated in the System 80 design, which uses only four coils and

identifies them as:

1) upper lift coil

2) upper gripper coil

3) lower lift coil

4) lower gripper coil

Each coil is fabricated from round copper wire, insulated with high pressure enamel, and

vacuum impregnated with a high temperature varnish. After impregnation, the coil is wrapped

with fiberglass tape, and encapsulated with a silicone compound. A nickel plated carbon-steel

housing encloses the coils.

2-12
23.2 Rack and Pinion CEDM

Two plants (Fort Calhoun and Palisades) use a rack-and-pinion CEDM in lieu of the conventional

magnetic jack CEDM (Figure 2.8.). This type of CEDM has a drive shaft running parallel to the rack,

which drives the pinion gear through a set of bevel gears. An electric motor, operating through a gear

reducer and a magnetic clutch, drives the attached CEA. When the magnetic clutch is de-energized, the

CEA inserts freely into the core. An anti-rotation device is incorporated into the clutch mechanism,

which prevents the CEA from moving upward when it is de-energized. The rack-and- pinion CEDM is

cooled by water, as opposed to forced air.

The main components of this CEDM are:

* Pressure Housing: Consists of lower and upper sections, joined near the drive top by a threaded

autoclave closure. The stainless steel tubular lower housing is welded to an eccentric reducer

and flange piece at the lower end, which mates with the reactor vessel head. The upper portion

of the lower housing forms the autoclave closure with a recessed gasket surface for a spirally

wound gasket. The upper housing contains the flange which mates with the lower housing, a

cavity containing the rotating drive seal, and a tubular housing extension. A flange closure

allows access for attaching and detaching the CEA. The rotating shaft seal is fabricated from

tungsten carbide and Graphitar, and fitted with o-rings to prevent seal leakage. A cooling ring

surrounds the rotating seal, to maintain the temperature below 250'F.

* Rack and Pinion Assembly: This assembly is an integrated unit which fits into the lower

pressure housing, and couples to the motor drive package through the upper pressure housing.

Bevel gears transmit the torque from the vertical drive shaft to the pinion gear. The rack

engages the pinion gear, and is held in proper engagement by the backup roller. A permanent

magnet is attached to the upper end of the rack for reed switch actuation.

* Motor Drive Package: Consists of a fractional horsepower, 120v, single phase 60 Hz motor

which operates the drive. The motor, brake, clutch, position indicator and limit switches are

all mounted on a common frame to maintain position and alignment. The frame is flange

mounted to the upper pressure housing.

* Position Indication: Two independent position indication systems are used. The primary system

consists of a synchro-transmitter geared to the main driveshaft with position readout provided

by synchro-receivers connected to the transmitter. The second system consists of magnetically

actuated reed switches which provide a voltage corresponding to CEA position.

2.4 Control Element Assembly Rod Position Indication System

Position indication for each CEA is provided by two independent systems, the pulse counting and

the reed switch position indication systems. The reed switch position transmitter provides a voltage

indicative of the actual rod position. The Control Element Drive Control System provides pulses

corresponding to step demand requests which are counted by the monitoring system(s).

2-13
A

B

U I-.*-Y_ ~~~~~~Holddown bolts

XFLanes - HSeal weld

:kup position

pcator stn Thermal sleeve

Detail 'A' Detail '1B

Figure 2.8 Combustion Engineering Rack and Pinion CEDM8

2-14
,CABLE CONNECTOR CABLE CONNECTOR\

STEEL GRID

(TYPICAL)

PANEL

F- STAINLESS STEEL

1 I FLEXIBLE CONDUIT

A

CONTROL ELEMENT

DRIVE MECHANISM

VIEW AA

CEDM POWER

CONNECTION

REED SWITCH

0 r<X- \\ POSITION TRANSMITTER 11

REED SWITCH 0 0

POSITION TRANSMITTER 11

Figure 2.9 Combustion Engineering Reed Switch Position Transmitters'

The reed switch position indication system consists of two redundant reed switch position

transmitter (RSPT) assemblies located adjacent to the upper pressure housing (Figure 2.9). Each consists

of one hundred reed switches spaced evenly at 1.5 inch intervals. When activated by the permanent

magnet attached to the CEDM extension shaft, the reed switches supply different resistances and hence

different voltage drop values as shown in Figure 2.10. The voltage output from the amplifier corresponds

to the actual CEA location in the core.

Each RSPT provides a position signal to a monitoring system (such as the plant computer system,

the analog display system, or the core mimic display system) or to an optionally installed protection

system (such as the Core Protection Calculator System). Each RSPT also provides upper and lower

electrical limit and dropped rod signals to the Control Element Control System.

2-15
RI

S3

R2,

R3

S,

REED SWITCH

CONTACT

R5

WHEN NONE OF THE SWITCHES ARE CLOSED. THE INPUT

TO THE AMPLIFIER IS THE VOLTAGE DROP ACROSS

RI + R2 + R3 + R4 + R5

WITH VOLTAGE DROP IS ACROSS

S.CLOSED R1 + R2 + R3

52 CLOSED RI + R2

S3 CLOSED R.

AMP TO CEAC

Figure 2.10 Combustion Engineering Reed Switch Position Transmitter (RSPT) Typical Circuit

Rod position indication via pulse counts is performed by the plant computer. Each step command

to the older Control Element Drive Control System generates a corresponding withdrawal or insertion

pulse. In control systems which have been upgraded with microprocessor controllers, the withdrawal or

insertion pulse is generated only if the step request is successful. The plant computer monitors and

counts the respective pulses and generates a rod position. Some plants may also have a totalizer for

pulse counts for each CEA installed as part of the CEDCS. These totalizers are not normally utilized

for CEA position information.

Since rod position indication via pulse counts is based upon step requests in older CEDCS, as

opposed to actual successful steps with the microprocessor controllers, a stuck rod or a misoperation/nonstep may result in incorrect rod position indication. Rod positions derived from pulse counts are

periodically compared with RSPT rod indication and readjusted. Similarly, rod positions must be

readjusted following a scram or rod slippage.

2.5 Control Element Drive Control System

As shown in Table 2.5, three distinct control element drive control systems depending upon plant

age, are used by CE. The two original plants, Palisades and Fort Calhoun, use the rack-and-pinion

system. The older magnetic jack type CEDM plants use the Control Element Drive System (CEDS),

while the newer plants use the improved Control Element Drive Mechanism Control System (CEDMCS).

The CEDMCS is schematically shown in Figure 2.11.

2-16
Table 2.5 Combustion Engineering Control Element Drive Control Systems

Rack and Pinion Control Element Control Element Drive

Control System Drive System Mechanism Control System

(CEDS) - (CEDMCS)

Fort Calhoun Calvert Cliffs 1&2 Arkansas 2

Palisades Maine Yankee Palo Verde 1,2,3

Millstone 2 San Onofre 2&3

St. Lucie 1 St. Lucie 2

. Waterford 3

2.5.1 System Operation

Each control system consists of the components required to allow for CEA motion and proper

positioning in the core. Signals are received from various plant systems: e.g., automatic CEA motion

signals from the Reactor Regulating System, CEA motion prohibit from the Analog Display System

and/or Reactor Protection System, manual motion signals from the operator, arm and drop signals from

the Reactor Power Cutback System, and sequencing signals from the Plant Monitoring System. These

signals cause the control system to respond by applying sequential power to the CEDM coils resulting

in CEA motion, or to remove CEDM holding power from pre-selected cutback groups. A reactor trip

signal opens the Reactor Trip Switch Gear, which removes power from all of the CEDM coils, resulting

in all of the CEAs being gravity inserted into the core.

The CEDS transmits motion signals from the control panel to the coil power programmers (CPP),

which actuate the stepping cycle to raise or lower the CEA. The CEDMCS combines the control

paneVCPP interface into one integrated system. Several plants have upgraded the CEDMCS by installing

Automatic CEDM Timer Modules (ACTMs). The ACTMs are used to provide on-line monitoring of

CEDM engagement during stepping and to monitor and take corrective actions for inadequate holding

currents in grippers and for high coil currents. Otherwise, both the CEDS and the CEDMCS are

functionally identical.

The CEAs are functionally grouped as shutdown, regulating, and part-length rods. The shutdown

groups are the first to be withdrawn during reactor startup, followed by part-length CEAs. The

regulating rods are the last to be withdrawn to attain criticality. For non-tripped reactor shutdowns, the

regulating rods are inserted first, followed by the part-length groups and then the shutdown groups. In

the event of a reactor trip, all of the CEAs insert.

Four different control modes are utilized in CE reactors, specifically:

1) manual sequential group movement,

2) automatic sequential group movement,

3) manual group movement, and

4) manual, individual CEA movement.

2-17
-

CEOM MG Set 1 CEDM MG Set 2

rt) Y)

TCB2-2 TCB-6(ITB9 - ( CB3-7( 1TCB-3

TCB-S ( ' `TCB-8

TCB It -1Rtr I TCe-4

_ _

Reed Switch ~~~~~~~~~~~~~~ C D ~~Channels Plant

Relay . Monitring

Houssmitts Hsintergace H System

EDM PowSery setlier Steam.

CEDM EI L ~~~~~Switch ',__ , ,CEDMCS <K

_ Assemblies l l l l l l l Control

tel~~~g E311 l l 8~Hosin Rcguain

=rn ~ ~ Time Fi Assemblies~ ~ ~ ~ ~~~~~~~~~yse Control

I Coils I .2 , .

Ia I . .Re uaeterg

Main Power Busbar 1 . I Main Power Busbar 2

1 I I

l t'o

Wo

- BusI

upply r 

--ZeoCossing I'-*-.-

-~UandI

+

~ng

l

.1 ,

.1

.1 .

120Vac (- Plant J

Instrument

Bus I-

|120Vac 1t1 Aux Relays I I | Plant

IInstrument

I I UV and-I I-J Bus

t AuRelays| | Subopevsr

PRelay L

MaInternace Vol

& Suppl gAecitors (SI

i P (2)

PaIe LowI

I I ~~~~Maintenance1 Voltage I

I _(- & Supply D etectors (9)

Protecon

* System * * S

: Reactor

Power |

* Cutback

System i . . . . . .aI Turbine I

.j"_ Control !.-.

I * System ...I

I 'Feedwater : q

-'.+ Control , | System

: , _ _ _ I; __ I* Plant a -. Protection, b

* System . *_ ... _._

Figure 2.11 Typical Pre-Systeni 80 with ACTMI Upgrade and Interfaces
Sequential group movement functions such that when the moving group reaches a programmed low or

high position, the next group begins movement, thus providing overlapping motion of the regulating

groups. Applied successively to all regulating groups, this procedure results in a smooth, continuous

reactivity rate of change. The shutdown CEAs are moved in the manual control mode only, with either

individual or group movement. A selector switch permits withdrawal of only one shutdown group at a

time.

2.5.2 System Description

CEDM power is obtained from redundant motor-generator (MG) sets. The MG set motor is a 480

vac 3-phase, induction motor which receives power from a non-class lE bus. The motor drives a 240 vac,

3 phase, 60 Hz generator.

CEDM power from the Reactor Trip Switch Gear is routed to the control system via two distribution

buses which are tied together at the control system cabinets. Undervoltage devices monitor the CEDM

power, and upon loss of 240vac, provide signals to the Turbine Trip and Feedwater Control Systems.

The main logic and power components include:

* Common Logic Relay Interface: Provides an isolation relay interface for signals to/from systems

external to the control system.

* Common Logic Housing: Synchronizes CEA subgroup and group motion to ensure

compatibility between operational mode and group selection. This provides the system timing

required for single or multi-group operation, including a one step deviation limitation;

transferring raise or lower commands to appropriate subgroup logic housing; and permitting

a low rate of operation for regulating groups in manual sequential mode. The common logic

housing also ensures that only a single subgroup is assigned to a holding bus at one time. Hold

bus subgroup selection relays are mounted elsewhere in the control system.

* Subgroup Logic Housing: Controls the motion and holding of the CEAs by controlling the firing

angle of the Power Switch Assembly SCRs and the actuation sequence of the CEDM coils.

Some control systems have been upgraded by the Automatic CEDM Timer Module (ACTM),

a microprocessor based controller which monitors and controls the current powering the coils.

The current waveforms provide a direct indication-of the movement of the magneticjack. The

ACTM utilizes closed loop control to ensure proper stepping sequence. The ACTM also

monitors for high coil currents and inadequate gripper holding currents, and takes corrective

action. Other plants use a CEA timer card which controls the firing.

* Power Switch Assemblies: Provides coil voltages directly to the interconnected CEDM coils.

Each CEA subgroup has a dedicated Power Switch Assembly. A disconnect switch permits a

CEA subgroup to be removed from the power switch SCR power supply, and to be transferred

to a holding bus for maintenance (Figure 2.12). Coil power programmers (CPP) are used in

the CEDS plants to perform the same function.

* Subgroup Relay Interface: Switches the logic level voltages, representative of the system status,

to lamp and relay voltages for use by the CEDMCS and plant annunciators.

2-19
REACTOR TRIP

CIRCUIT BREAKER

I I

Figure 2.12 Combustion Engineering Typical Power Switch

* Pulse Count Relay Interface: Responds to each CEA step by means of contact closures. Each

CEA pulse controls a separate pulse count relay interface via an opto-isolator, for use by the

pulse count position indication system.

* Undervoltage and Auxiliary Relay Assemblies: Monitors the three phase power inputs to the

control cabinets, and provides a local indication and remote annunciation of an undervoltage

condition.

* Supervisory Panels: Provide for system status display, the reset of Common Logic

Housing latching circuits, and auxiliary circuit control.

For the rack and pinion CEDMs, power is supplied to the fractional horsepower, 120 v single phase 60

Hz motor in lieu of gripper coils.

2.6 CEDM Cooling System

Although not a safety-related system, the CEDM cooling system is required to ensure the continuous

and reliable operation of the gripper coils. Continued operation without forced-air cooling can result

in the overheating of the gripper coils, resulting in dropped CEAs. Typically, the cooling system is a

forced air system, consisting of two to four fans, depending on plant design, which are designed to

maintain the coils at a temperature below 3500F. Redundant, stand-by fans ensure continuous cooling

if the primary pump fails. A sheet-metal cooling shroud assembly located on the top of the reactor vessel

head provides an annulus which directs the air from the fans to the coils. The cooling system is

controlled and monitored remotely from the control room.

2-20
2.7 System 80

The System 80 design is a modification of the standard CE reactor design. New features include

a larger core size, more fuel rods per fuel assemblyand modifications to the guidance method for the

CEAs used for reactor control and shutdown.

The System 80 CEDMs are similar to the previous design CEDMs, with the following features:

* four coils versus five coils (pull down coil eliminated)

* two lift mechanisms and two gripper mechanisms.

The elimination of the pulldown coil necessitated the installation of a coil spring to insure the positive

resetting of the latch assemblies. The drive shaft was also modified to allow the load transfer and

stepping functions to be performed with the same coil.

2-21
3. DESCRIPTION OF THE BABCOCK & WILCOX CONTROL ROD DRIVE SYSTEM

3.1 Introduction

The Babcock & Wilcox control rod drive system is comprised of five main sub-systems (Figure

1.2):

I. Control Rod Assemblies (CRAs): Each individual CRA consists of neutron absorber rods

connected to a common hub. When inserted into a fuel assembly, guide tubes maintain

the proper spacing and position of the rods. Brazement assemblies maintain the CRA

spacing and protect the CRAs from the effects of coolant crossflow when they are

withdrawn from the fuel assemblies into the upper plenum region.

II. Control Rod Drive Mechanisms (CRI)Ms): Electro-mechanical devices which in response

to automatic or manual control signals, insert, withdraw, or hold the CRAs stationary in

the reactor core.

III. Control Rod Drive Power and Control System: Provides electric signals to CRDM stator

coils, resulting in movement of the CRA to control reactivity or to shutdown the reactor.

IV. Control Rod Drive Rod Position Indication: Absolute CRA position indication is provided

by magnetically actuated reed switches. The relative position indication system monitors

pulses provided to the CRDM coils and indicates the demanded position.

V. Control Rod Drive Cooling Mater System: A closed-loop cooling system which supplies

cooling water to the CRDM stator coils.

Table 3.1 summarizes the main components and primary functions for the five sub-systems.

3.2 Control Rod Assemblies

A combination of control rod assemblies (CRAs) and axial power shaping rod assemblies

(APSRAs) control reactivity in B&W reactors. Table 3.2 summarizes the number of control components

used at each plant. Immovable burnable poison rod assemblies (BPRAs) are inserted to control the

reactivity of fresh fuel assemblies, and are removed after the first burnup cycle.

The CRA consists of 16 individual control rods attached to a spider assembly which maintains

the rods in a geometrical pattern to allow for insertion into the fuel assembly guide tube. The bayonet

coupling on the end of the leadscrew mates with the CRA spider hub to connect the two. The typical

control rod assembly is shown in Figure 3.1.

The APSRA controls the axial power shape across the core during the fuel cycle. It resembles

the CRA with the exception of a slight modification to the spider hub to preclude the placing of an

APSRA in a CRA location. The APSRA is also attached to the leadscrew of the CRDM at the spider

hub, and is movable in the core. However, the CRDM is modified to prevent insertion into the core

during a scram.

3-1
Table 3.1 Primary Functions of Major CRDM Sub-Systems

Sub-System Component Primary Functions

I. Control Rod Assemblies A. Control Rods Provide reactivity control during normal operation.

Provide rapid reactor shutdown (scram).

Provide sealed pressure boundary to contain neutron absorber material,

including sufficient free volume to accommodate irradiation induced

poison swelling and off-gas production.

B. Fuel Assembly Guide Tubes Provide envelope within each fuel assembly which maintains CRA

spacing to permit free insertion.

C. Upper Internal Brazement Assemblies Maintain spacing of CRAs when withdrawn from fuel assembly into the

upper plenum region.

Protect CRA from coolant cross flow effects.

II. Control Rod Drive Mechanisms A. Motor Tube Provide primary coolant pressure boundary. Provide sufficient free

volume to permit complete withdrawal of leadscrew. Support and

position motor assembly for proper activation of rotor assembly.

B. Cosurc Assembly Provide means to vent CRDM after system filling prior to operation.

Provide access to remotely couple/decouple leadscrew from CRA.

C. Motor Assembly (stator) Provide magnetic force necessary to engage the roller nuts with the

leadscrew to hold or position the CRA in the core.

D. Rotor Assembly When actuated by motor assembly, roller nuts engage the leadscrew.

When a rotating magnetic field is applied, roller nuts rotate around

leadscrew, resulting in linear CRA motion.

E. Thermal Barrier Restrict circulation of primary coolant into CRDM to control rotor

assembly temperature.

Relieve pressure drop in CRDM following scram.

F. Leadscrew Assembly Provide a threaded leadscrew, when engaged by rotor assembly, results

in linear CRA motion.

Provide means of attaching leadscrew to CRA spider hub.

Provide attachment for permanent magnet which actuates reed switches.

G. Torque Tube Assembly Prevent rotational motion of leadscrew during operation.

Provide snubber assembly which dampens CRDM deceleration load

during scram.
Table 3.1 (Cont'd) Primary Functions of Major CRDM Sub-Systems

Sub-System J Component | Primary Functions

III. CRDM Power and Control A Power Supplies Transform 480 volt, 3 phase plant power to 120 volt. six phase power

System for motor assembly actuation.

B. Programmers Control sequence of power supplied to individual phases of motor

assembly to produce rotating magnetic field.

C Trip Breakers Provide for rapid removal of power to motor assembly for reactor

scram.

D. Programmer Drive Motors (run and jog) Provide means to control speed of CRA insertion or withdrawal.

IV. CRA Rod Position A Absolute Position Indication System Provide direct indication of CRA position through reed switch

Indication actuation.

B. Relative Rod Position Indication Provide indication of demanded CRA position by monitoring and

counting pulses supplied to each motor assembly.

V. CRDM Cooling System Centrifugal Pumps, Heat Exchangers, Surge Closed loop cooling system which provides cooling water to individual

Tank CRDM motor assemblies to prevent thermal overheating.
Table 3.2 CRA and CRDM Plant Data

NumberIoe Number of N

Number of Guide Number of Number of

Fuel Assys Tubes Per CRAs Per APSRAs CRDM

in Plant Fuel Assy. Plant Per Plant Type

Arkansas-1 177 16 60 8 B

Crystal River 177 16 60 8 A

Davis Besse 177 16 53 8 C

Oconee 1 177 16 61 8 A

Oconee 2 177 16 61 8 A

Oconee 3 177 16 61 8 C

Rancho Seco 177 16 61 8 B

Three Mile Island 1 177 16 61 8 A

Table 3.3 Control Rod Assembly Data"1

Standard Rod Extended Life

Number of CRAs 53-61 (Dependant core upon 53-61 (Dependant upon core

design design)

Number of Rods per 16 16

Assembly l l

Control Rod Outer Diameter 0.440 0.441

(in.) l l

Cladding Thickness (in.) 0.021 0.0225

Cladding Material Type 304 SS, Cold-Worked Inconel

End Plug Material Type 304 SS, Annealed Inconel

Spider Material SS, Grade CF3M SS, Grade CF3M

Poison Material 80% Ag, 15% In, 5% Cd 80% Ag, 15% In, 5% Cd

Female Coupling Material Type 304 SS, Annealed Type 304 SS, Annealed

Length of Poison Section (in.) 134 139

Stroke of Control Rod (in.) 139 139

3-4
Top View

Neutron Absorbing

Materiallll

Control Rod

Side View

FIgure 3.1 Babcock & Wilcox Control Rod Assembly'"

Two specific control rod designs are presently used, the standard, and the extended life design.

Both designs use silver-indium-cadmium (Ag-In-Cd) as the poison material. The extended life design

is fabricated from Inconel clad as opposed to the standard, thinner- wall stainless steel tubing, and is also

pre-pressurized with helium to reduce clad stresses. Table 3.3 lists additional design details for the two

types of rods.

Currently, two axial power shaping rod designs, designated gray and black, are used (Table 3.4).

The black APSR utilizes Ag-In-Cd poison, while the gray APSR uses a longer Inconel absorber section,

and is pressurized with helium to reduce differential pressure stresses in the clad. Both use cold worked

Type 304 stainless steel clad, and contain a region above the poison section which is vented to the

primary coolant.

B&W reactor cores consist of 177 fuel assemblies. Each fuel assembly consists of fuel rods

arranged in a 15 x 15 square lattice, with 16 Zircaloy guide tubes. The guide tubes provide a guidance

envelope for the control rods during operation. When the control rods are fully withdrawn from the

core, brazement assemblies in the upper internals maintain proper CRA alignment and protection from

coolant crossflow effects.

3-5
Table 3.4 Axial Power Shaping Rod Assembly Data12

APSRA - Black APSRA - Gray

Number of Axial Power 8 8

Shaping Rod Assemblies

Number of Rods per 16 16

Assembly

Outside Diameter of Axial 0.440 0.440

Power Shaping Rod (in.)

Cladding Thickness (in.) 0.021 0.027

Cladding Material Type 304 SS, Cold-Worked Type 304 SS, Cold-Worked

Plug Material Type 304 SS, Annealed Type 304 SS, Annealed

Poison Material 80% Ag, 15% In, 5% Cd SS, Grade CF3M

Spider Material SS, Grade CF3M Inconel

Female Coupling Material Type 304 SS, Annealed Type 304 SS, Annealed

Length of Poison Section (in.) 36 63

Stoke of Rod (in.) 139 139

3.3 Control Rod Drive Mechanisms

The control rod drive mechanisms (CRDMs), shown in Figure 3.2, are electromechanical devices

consisting of an electrically driven, rotating nut assembly within the primary coolant pressure boundary;

a four pole, six phase stator, and a translating leadscrew which converts the rotary motion of the rollernut assembly to linear travel of the leadscrew and CRA. All electrical components and attachments are

mounted external to the reactor vessel, allowing for maintenance or removal without compromising

system integrity. There are no electrical penetrations through the primary system pressure boundary.

The operation of the roller nut CRDM is illustrated in Figure 3.3. When power is supplied to the

CRDM, the roller nuts engage and rotate about the leadscrew, resulting in vertical CRA motion. When

power is removed, the compression springs cause the two halves of the roller nuts to separate and

disengage the leadscrew, causing the CRA to insert into the core (scram).

B&W designed the CRDM internals for twenty years of operation, based on the following

assumptions:

1) the CRDM would be used exclusively with regulating rods,

2) regulating rods would frequently be re-positioned accumulating a total of 126,000 feet

of leadscrew travel in twenty years, and

3) the plant would experience 500 trips in 20 years (25 trips/year).

3-6
VENT VALVE I

P.L CONNECTORH

STATOR CONNECTO

STATOR THERMO- PRESSURE HlOUSING

COUPLE CONNECTOR

POSITION INDICAO ASSEMBLY

STATOR COOLANT

PIPING

\\ T ROTOR ASSEMBLY

WATER JACKET STTATOR ASSEMBLY

t\W~nRROLLER NUTS

SEGMENT ARU

NG FLANGE

\ LEAOSCREW

CONTROL ROD

COUPLING

Figure 3.2 Control Rod Drive Mechanism'3

Three CRDM types, designated Type A, B, and C are in use as indicated in Table 3.2. The main

components of these CRDMs, the design features, and differences are discussed below.' 5

3.3.1 Motor Tube

The motor tube is a four section weldment, housing the rotor assembly, thermal barrier,

leadscrew, and the torque tube assembly. When bolted to the reactor head, it forms the primary coolant

pressure boundary. The tube wall between the rotor assembly and the stator is constructed of magnetic

material which minimizes the magnetic air gap between the stator and rotor and increases the magnetic

coupling to the rotor assembly. For the Type A and B CRDM's, this area is fabricated from low alloy

steel clad on the ID with Inconel, while on the Type C CRDM, this section is fabricated from unclad

Type 403 stainless steel.

3-7
Roller Nuts Engaged

LEAO SCREW

Power On ~~~~MOTOR

power On X t WINDING

COMPRESS

SPRING

_ i,.1C ROLLER

Rotation

Roller Nuts Dis-engaged

Power off

CRA Insertion

or Withdrawal

CRA Scram

Figure 3.3 CRDM Rotor Assembly and Leadscrew"

Figure 3.4 CRDM Vent Valve Assembly & Closure Parts

3-8
KNURLED LOCKING

SCREW

| | 2 /Y~~~~~~~ENT CAP

PUSH DOWN TO HARDSTOP. ROTATE CLOCKWISE TO HARDSTOP

ENGAGING TOOL STEM

WITH THE VENT VALVE.

a/.~~~~VN HOUSING

INSTALLED /

BACK-UP VALVE

(SHOWN 90- OUT -CLOSURE INSERT

OF ROTATION FOR ASSEMBLY

CLARITY).

IVI CRD VENT VALVE

DISENGAGED POSITION

Figure 3.5 Venting Tool

The upper end of the motor tube supports the absolute position indication assembly, and houses

the withdrawn leadscrew, torque tube, torque taker, and snubbers. It is fabricated from non-magnetic,

stainless steel which is transition welded to the center section. The lower end of the center section is

welded to a non-magnetic stainless steel forging. Double flexitallic gaskets serve as the seal between the

CRDM and the reactor vessel nozzle.

3.3.2 Closure Assembly

The closure assembly (Figure 3.4) is located on the top of the motor tube, and consists of a

removable closure insert assembly and vent plug. Removal of the closure insert assembly permits access

to the torque taker assembly for remote coupling or un-coupling the CRA to the leadscrew. The insert

closure assembly is retained by a closure nut which is threaded to the inside of the motor tube. The

sealing load is supplied either by six jacking screws or by a -hydraulically pre-loaded spring washer

retained by the closure nut.

3-9
Removing the vent plug permits the venting of all non-condensible gases from the CRDM prior

to reactor head removal or coolant fill and heat up. A special venting tool (Figure 3.5) is used for this

task.

3.3.3 Motor Assembly

The motor assembly is mounted over the motor tube and when powered, generates the magnetic

force which actuates the roller nuts and engages the leadscrew. The motor assembly is a synchronous

reluctance unit with a 48 slot, 4 pole slip-on stator arrangement, containing cooling water coils in the

outside casing. The stator is varnish impregnated and may be encapsulated after winding. It is 6 phase

star connected to allow for pulse stepping operation, which advances 15 degrees per step.

Early stator designs used epoxy encapsulation rather than varnish impregnation. Impregnation

improves coil heat transfer and electrical insulation properties. The Type C CRDM requires considerably

less power to operate. Early stator designs (Figure 3.6) had cooling tubes wrapped around the housing.

Machining additional grooves on the casing has increased stator cooling. Stator temperature is monitored

by thermocouples and alarms if the winding temperatures exceed design limits.

THERMOCOUPLE

CONNECTOR

-P.I. ADJUSTMENT SCREW

'POWER CONNECTOR

'POWER CONDUIT CLAMP

-COOLING WATER LINES

-STANDOFF (5-FOR LIFTING1

POWER

Tube Type Jacket (Stator Slot Type) Groove Type Jacket (Tube Type)

Figure 3.6 Stator Water Jacket Assembly

3-10
3.3.4 Rotor Assembly

The rotor assembly engages, holds, and positions the leadscrew when actuated by the motor

(stator) assembly. The major components of the rotor assembly are the rotor tube, segment arms, roller

nuts, pivot pins, segment arm springs, and the bearings (Figure 3.7). The rotor tube is a hollow tube with

bearing journals on each end through which the leadscrew passes and the segment arms are attached.

The lower journal is for the inner race of the thrust bearing, and the top journal is for the inner races

of the synchronizing and radial bearings. The thrust and radial bearings provide radial alignment for the

upper and lower portions of the rotor. The thrust bearing also carries all the axial loads applied to the

rotor. Synchronizing pins on the top of each segment arm engage mating holes in the outer race of the

synchronizing bearing. Radial motion of one segment arm is transmitted through the pin and bearing

arrangement to the other segment arm. Without this synchronizing feature, one segment arm could move

during a trip, while the other segment arm remained engaged with the lead screw.

RETAINING RING

RADIAL BEARING

SPACE R a

SYNCHRONIZING BEARING

SEGMENT ARM

ROTOR TUBE

RETAINING NUT---% -"*-MAX DIAMETER

SPRING

FIgure 3.7 Partial Section Showing Roller Nuts Engaged

3-11
The magnetic stainless steel segment arms are mounted on the rotor tube by four pivot pins

which allow the arms to rotate with, and pivot on, the rotor tube. The upper portion of the segment

arms forms a four pole collapsible rotor. The lower portion of each segment arm has two spindle

mounted roller nuts which mate with the leadscrew thread when the rotor assembly is latched. The

magnetic field established by the stator pivots the upper portion of the segment arms outward, which

moves the lower portion inward, engaging the roller nuts with the leadscrew. When the magnetic force

is removed, or reduced in strength, the segment arm springs force the lower portion outward, disengaging the roller nuts, and inserting the CRA into the core.

3.3.5 Thermal Barrier

The thermal barrier is located in the lower portion of the CRDM where it restricts the circulation of

the primary coolant, and acts as an insulator between the reactor vessel head and the CRDM,

maintaining the rotor assembly between 300 and 350'F during normal operation. To allow unrestricted

travel of the leadscrew, a clearance hole is drilled through the center of the housing. Four ball-check

valves relieve the pressure drop within the drive during a scram. Coolant, which enters the drive during

a scram, returns to the system through this clearance hole when the rod is withdrawn (Figure 3.8).

LEADSCREW THERMAL BARRIER

HOLD DOWN BOLT

BALL (FLOW CHECK)

MOTOR TUBE ' THRUST BEARING

RETA INER

I ~~~~~~~INDEX PIN

ADAPTER RING

LEADSCREW

SUPPORT RETAINING

REACTOR HEAD iNUT NOZZLE

NUT RING

FIgure 3.8 Leadscrew Guide Assembly

3-12
3.3.6 Leadscrew Assembly

The leadscrew assembly is the connecting link between the CRA and the rotor assembly. It

consists of an upper extension, leadscrew, lower extension, and male coupling. The leadscrew and the

torque tube convert the rotational motion of the segment arms into vertical CRA motion. The

leadscrew travels vertically in the torque tube, and is attached to the torque taker by the leadscrew nut.

For the Type A CRDM's, the torque taker contains a key which engages a slot in the torque tube,

preventing leadscrew and torque taker rotation. For Type B and C CRDM's, the torque taker contains

the keyway, and the key is on the torque tube.

The leadscrew thread is a modified ACME with a pitch of .375 inch and a relief angle which

facilitates the disengagement of the roller nuts from the leadscrew. The male bayonet coupling on the

lower end of the leadscrew mates with the hub of the spider assembly connecting the CRA with the

CRDM. The CRA is coupled and uncoupled with a special handling tool inserted through the closure

assembly. To couple a CRA, the leadscrew is lowered until the coupling is positioned inside the spider

hub. Simultaneously, a pin on the upper portion of the leadscrew is positioned within the torque taker.

The handling tool rotates the leadscrew 450, positioning the coupling tabs so that they grapple the CRA.

Tightening the leadscrew nut prevents any movement relative to the torque taker and CRA. The reverse

procedure is used to decouple the CRA.

3.3.7 Torque Tube And Torque Taker

The torque tube is a separate tubular assembly containing either a key or keyway extending the

full length of the leadscrew. The tube assembly is secured against vertical and rotational movement at

the lower end of the closure assembly by a retaining ring, keys, and the insert closure. The lower end

of the torque tube houses a hydraulic snubber assembly which dampens the deceleration load during a

scram. Snubber damping characteristics are determined by the size and position of the holes in the

snubber cylinder wall, and the clearances between the snubber piston and bushing. Practical operating

clearances limit the amount of hydraulic snubbing. At the end of the snubbing stroke, the residual kinetic

energy is absorbed by a buffer spring. The leadscrew contacts the motor-tube closure insert assembly to

stop outward motion.

The torque taker assembly also consists of the permanent magnet which activates the reed

switches, the snubber piston, and positioning key/keyway. This assembly is attached to the top of the

leadscrew and mates with the torque tube providing radial and tangential leadscrew support. (Figure 3.9)

3.3.8 CRDM Operational Considerations

To ensure the proper operation of the CRDM, specific operational limitations have been

established:

a) Cooling Water: A minimum flow of 2 gpm per stator is required. The inlet temperature

should be between 800F (min.) and a maximum of 120'F (max.). If cooling water is lost

or reduced, a high stator temperature alarm will be received at 160'F (max.). A CRDM

must be de-energized when it reaches 180'F. If more than one CRDM attains this level,

the reactor should be tripped. Figure 3.10 shows the minimum allowable operating

pressure and temperature ranges.

3-13
Figure 3.9 Torque Taker Assembly

b) Lubricating Water The primary coolant serves as a hydraulic buffer and lubricant when

the mechanism is tripped. The CRDMs should only be operated in the recommended

temperature and pressure ranges, and when the total concentration of dissolved gas is

less than 100 cc/kg. If the temperature or pressure exceed the limits, the CRDMs should

be driven in and vented. The CRDMs also must be vented whenever the system has

been drained and filled, to prevent the coolant from being displaced by gas.

c) Electrical: Excessive heat generation in the windings, which depends upon current flow

and resistance, may result in stator damage. The control system is designed to limit heat

generation by ensuring that only two phases are energized when the CRDM is stationary.

3-14
1000

a' 800

c 600 Permissible

Operating

Region

U 400

200 | 100 200 300 400 500 600

RC Temperature (Th), F

Figure 3.10 Minimum Allowable RC Pressure vs. Temperature for

Tripping or Operating Control Rod Drives"6

d) Latching. To ensure positive engagement between the roller nuts and the leadscrew, a

latching sequence is required. Once the CRDM is energized, and before rod withdrawal

begins, the CRDM is rotated in the "IN" direction for 15 sec. This rotation corresponds

to one complete mechanical revolution of the roller nuts around the leadscrew. Since

the motion does not place any weight on the roller nuts, the two components will

properly engage.

3.4 Control Rod Drive Control System

The CRDM control system provides electrical signals to the motor assembly to insert, withdraw,

or hold the CRAs in the core in response to automatic signals from the Integrated Control System (ICS)

or manual signals. Reactivity is controlled through the positioning of the eight control rod groups. Four

groups are designated as safety rods, three groups consist of regulating rods, and the eighth group

consists of the axial power shaping rods. The particular number of CRAs per group depends upon the

core cycle design (Table 3.5). The CRAs are arranged into groups at the control system patch panel.

During reactor startup, the safety groups are withdrawn first, enabling the regulatory groups to

be withdrawn next. During a scram, all the groups, except the APSRAs, insert rapidly into the core.

The CRDM control system, schematically shown in Figure 3.11, consists of the control rod drive

motor power supplies, system logic and trip breakers. There are four group power supplies, an auxiliary

power supply and two holding power supplies. The group power supplies may be either a redundant, sixphase, half-wave rectifier or a three phase, full rectifier design. In each half of the group power supply,

silicon controlled rectifiers (SCRs) are used to rectify and switch power. This switching sequentially

energizes first two, then three, then two of the six CRA stator motor windings in stepping motor fashion

to generate a rotating magnetic field (2-3-2-3 sequencing). Switching is achieved by gating the six SCRs

on for the period each winding must be energized. Since each of the six windings use SCRs to supply

power, six gating signals are required.

3-15
480 Vc480 Volt__=

Votg 1280 Vlt___=

@ L S } XFoarmner I- I

M 8 ain Bkr IS Disconneect1 IS:

11 I~ ~~~o Groups 2, 3 & 4 Switches

I2

From Groups 2, 3 & 4

. } | | 1 ' ~~~~~~Power Supplies A

'----------- --t4-i DC Rod Drive

Manual Trip Switch ' . z z Breakers Main

Bus A( gk- . A3 B~sA -. , : . A~o 'A' Phase

' B ...na F. k -' ' Groups 2, 3 & 4

.Bus B 5

Phase'A' Groups

, Vaal { -~ p Elctronc Phase CC Crips Groups Clamping 4 Disc C{ , E nTrip To 'CC' Phase

CC41

CC2| Transfer Delays~~

^ \ 1 1 CCC CRod Drive

Secondary I Sreakers Secondary

480 VoltPowerSuppliesoA

t tt t x ~To Groups G

_ _ = I I Motor Power } S 

'480 Vac . E E r Supplies 0CC |F1o

B

BBCC

(;clamping Contactor

I ~~~~~~~Transfer

11ve IlTI 11 .llT .MRelay

Position Auxillary Power Supply Motor Power

Reset Secondary Side OutputFuses GroA

nal Typical S SiSnalTpicai \

i` 69 Mechs > of 69 Meh etal

I I Is ISC I IS

I ITQ. V From Main Electronic Trip 'E'

Gate Drives

To Main |Main Lamp|

III. 1X1}1X1}1}6I Hold Interlock * Power

aid IJ -Egg GMM Supply § annel A' I - Undercurrent Main Gate A A . Relay

1--U B Encoding Lamps Main

c '1 "Encoder Disc

To Main Gate Drives | Speed

Channels'B'Thru 'CC' M Siee 

Rod Group 5 /B Motor

Power Supply Ph /cc

(Typical of Group Photo __ - Gear In & Out

6,7,8 & Aux DetectorsD Commands Supplies) ~

1 1 1 1 1 Drives - ' B :|Normal |

To Secondary Gate Dv i-, M iSpeed 

Channels'B'Thru 'CC' |

.- , C {l~Encoding Lamps

i1 (Secondary)

channel 'A'l 213A Undercurrent

ISecondary Rela IGate Drive j~ econdar

<'To Sec Lam Poer

( Hold Interlock Supply

aFrom Secondary Electronic Trip 'E'

Gate Drives

T l II-- I J

, , tK 480 Vol:

30 to

_t <Vta 120Volt - Regulator 6 r

Se a XFormer

L---rn/

I _ _ _ __I_ _ _ _I__ _ _I__ _ _I1 _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

.~~~~Aw~~~~~rI~~~~d~~~~y 

J~~~

jQ06W-lUdlyl I I I

Figure 3.11 Control Rod Drive Control System Power Diagram
Table 3.5 CRA Grouping By Plant

Axial Power

Safety Rods Regulating Shaping Rod

Group Rods Group Group

Plant 1 2131 1 6 7 8

Arkansas-l 8 8 8 8 8 8 12 8

Crystal River 8 8 12 8 8 8 8 8

Davis-Besse 4 8 4 4 12 12 9 8

Oconee-1 8 12 9 12 12 4 4 8

Oconee-2 8 12 9 12 12 4 4 8

Oconee-3 8 12 9 12 12 4 4 8

Rancho Seco 4 8 8 8 12 12 9 8

Three Mile Island 1 8 12 9 12 1 4 4 8

The six phases of the stator are designated A, B, C, AA, BB, and CC. The A and AA, B and

BB, C and CC windings are bifilar, with each pair in the same physical location in the stator (wound on

top of each other). When energized, equal but opposite polarity magnetic fields produce two north and

two south poles. This four pole magnetic field pulls the upper portion of the segment arms outward,

causing the roller nuts to engage the leadscrew. When the CRDM is stationary, a maximum of two

phases are energized, preventing heat damage to the stator.

The process of supplying overlapping sequential phase currents results in 12 mechanical steps

for each electrical cycle. Two electrical cycles results in one rotor revolution, moving the leadscrew 0.75

inch. The speed and rotational direction of the magnetic field controls the speed and direction of CRA

travel (Figure 3.12).

The programmer commands the SCR gating sequence in the programmed power supplies to

produce the 2-3-2-3 sequencing. The programmer uses two split-phase drive motors, one which operates

at 60 rpm resulting in a CRA speed of 30 in./min., and the other which runs at 6 rpm resulting in a CRA

speed of 3 in./min.

The drive motors are coupled to an slotted optical disc (Figure 3.13), with redundant light sources

on one side and photo-detectors on the other. As the drive motor rotates the disc, transparent windows

pass in front of the light sources, activating the photodetectors in the desired 2-3-2-3 sequence. The

output of each photodetector drives an optically coupled transistor switch in a 12 v dc gate drive circuit.

The gate drive associated with a particular phase activates the SCRs for that phase. As the SCRs are

gated, they rectify 120v ac and transmit it to the stators as 120 v de. Thus, the 2-3-2-3 sequencing of the

photodetectors results in the sequential energizing of the stator windings from the phases on the main

feeder bus. In addition, some plants have modified the programmer by removing the optical disc and

split phase motors and installed a microprocessor controlled programmer to generate the 2-3-2-3

sequencing signals which are input to the gate drive assemblies.

3-17
Figure 3.12 Rotating Magnetic Field

The seventh photodetector in the set energizes the 3-2 hold circuit. When the CRDM stops with

three phases energized, this cell is activated, energizing a relay resulting in a jog-in signal to the

programmer. When it steps back to the two phase energized position, the photodetector turns off, the

relay is de-energized, and the motion stops.

With no motion command present, the programmer motors are prevented from rotating by

applying 60 v dc instead of 120 v ac. If normal brake power were lost, 24 v dc is supplied through the

direction error circuitry.

3-18
Secondary

Lamp Power

Supply

5 )

larvI-_

DC Power To Winding *A Note: Rod Group 5 Power Supply Shown: Of GowrTou5 Control Rods

Typical Of Group 6. 7.8, and Aux. Of Group 5 Control Rods

Power Supplies

Figure 3.13 Control Rod Drive Regulating Power Supply Programmer

Identical power supplies are used for the regulating groups and for the auxiliary power supply.

Each half of the group power supply is capable of driving 12 (max.) CRDMs. The power supplies have

dual power inputs fed from separate power sources, each capable of carrying the full load.

Because the six-phase holding power supply is used to maintain the safety rods fully withdrawn,

switching is not required. Two holding power supplies are used, each rated to supply power to one

winding of the safety CRDMs.

The auxiliary power supply is used to position the safety rod groups and to control single rods

The safety rods are positioned with the auxiliary power supply, and when in position, are transferred to

3-19
the holding bus. After positioning the safety rods, the auxiliary power supply is available to the regulating

rods through relays, to serve as a single rod repositioner, or as a spare group controller if required. The

auxiliary power supply cannot be used to control more than one group at a time.

System logic encompasses the functions which command manual or automatic CRA motion, CRD

sequencing, safety and protection features, and manual trip functions. The major components of the

logic system are the operators control panel, CRA position indicator panels, automatic sequence and

relay logic.

3.5 Rod Position Indication

Two methods of position indication are used, the relative and absolute position indication

systems. The absolute position indication system monitors the position of the leadscrew through the reed

switches. The relative position indication system monitors the input pulses to the CRDM motor to

provide an indication of the demanded position.

3.5.1 Absolute Position Indication System

The Absolute Position Indication (API) System consists of 72 equally spaced, magnetically

actuated, reed switches mounted in a fiberglass or aluminum housing strapped to the outside of the

motor tube. These switches are actuated by the permanent magnet attached to the torque taker. As the

reed switches sequence, the resistance of the network changes, resulting in a variable voltage output,

corresponding to the actual rod position.

The API system also has two additional, individual reed switch groups which monitor:

- in, and out limits, and

- five zone reference switches at 0%, 25%, 50%, 75%, and 100% travel.

API signals are used to:

1) provide individual position indication on the position indication panel located in the

control room,

2) determine group average position,

3) determine asymmetric rod indication,

4) determine group in and out limits,

5) indicate the first rod from a group in and out,

6) provide electrical signals to the sequence enable circuits,

7) provide electrical signals to the auto and out inhibit circuits, and

8) satisfy the feed and bleed permits.

3-20
'GOLD BUILDUP

(b) LOW DIFFERENTIALSWITCH

IR GAP

-7100111~

L DIFFERENTIAL GAP

(al HIGH DIFFERENTIAL SWITCH

FIgure 3.14 Reed Switches

The current API system uses high differential reed switches (Figure 3.14) which are completely

enclosed in glass, with rhodium plated contact surfaces. During the early 1980's low differential reed

switches were used which incorporated gold plating on top of the rhodium plating. The failure rate and

the erratic operation of these early switches, due to the buildup of contact surface film and the low

closing force required from the decreased gap, necessitated the design change. These switches were also

prone to fluttering in stray magnetic fields. The high-differential switches provided a greater positive

contact due to the higher closing force, thus reducing surface film buildup.

Operating limitations also necessitated the redesign of the API system. The early Type A

indicating circuits used a two channel averaging circuit. As the leadscrew moved, the reed switches closed

in a 2-1-2-1 sequence. A failed open reed switch produced an erroneous dropped rod indication and

initiated an automatic runback to 60% power resulting from the asymmetric fault condition. If any of

the forty-eight reed switches failed, the entire assembly had to be replaced because the failed switch could

not be bypassed.

3-21
V.,

TWO -CHANNEL AVEMGING

TYPE A -R4C

Figure 3.15 Absolute Position Indication (API)

The new circuit design (Type A-R4C), has two parallel sets of voltage divider circuits, comprised

of 36 resistors, each connected in series. A 5 volt dc power supply is connected across the two circuits,

such that 5 volts (at the top end of the circuit) represents the full-out rod position, and 0 volts (at the

bottom end of the circuit), represents the full-in position for the rod. The reed switches for each circuit

are offset such that they are staggered. One end of the 36 reed switches are connected at a junction

between each of the resistors of the two parallel circuits. The other end (output) of the switches are

alternately connected to 4 internal output signal lines. These signals are selectively averaged to form two

output signals. Both circuit designs are schematically shown in Figure 3.15.

The Type A-R4C API is designed such that either 2 or 3 reed switches are closed in the vicinity

of the magnet. Each time a reed switch closes or opens as a result of magnet movement, an analog signal

representing rod position is displayed on the position indicator monitor in the control room. The number

of switches which are closed, and the length of time they are closed during leadscrew travel, are

contributing factors to the accuracy of the API output signal. Another feature of this design is the

isolation switches mounted on the API amplifier card. These switches, when open, prevent the signal

from their respective output channels from being sent to the buffer amplifier.

The In limit, Out limit, 0%, 25%, 50%, 75%, and 100% reference reed switches for the Type AR4C API have the same pickup and dropout sensitivity and accuracy as the original Type A API. The

percentage switches use the same circuitry. The In limit and Out limit circuits for the Type A-R4C API

consist of two reed switches, connected in series, to ensure that those circuits open when the magnet is

moved from either limit position.

3-22
The redundancy introduced into the Type A-R4C circuit compensates for component failure.

The operating history for the Type A API demonstrated that the major component failure was the failure

of the reed switch to function properly. The predominate mode of failure was the inability of the reed

switch to close (pick up) in the presence of the magnetic field produced by the leadscrew magnet. The

reed switch would also occasionally stay closed (not drop out) when the magnetic field was absent.

In the new R4C design, the reed switches are sequentially activated so that position indication

is provided when a single reed switch fails to close without an asymmetric rod condition. Position signal

is lost only when two or more adjacent reed switches fail to close. This results in an asymmetric alarm,

but not sufficient to initiate a runback. The three-channel operation allows the failure to be located,

isolated, and bypassed. The faulty reed switch may be isolated by opening the proper Amplifier Card

isolation switch.

The Type A API has a signal accuracy of +/-2 inches. The Type A-R4C API has a signal

accuracy of +/-2.5 inches when both output channels are operating, and +/-3.5 inches when one output

channel is operating.

3.5.2 Relative Position Indication System

The Relative Position Indication (RPI) (Figure 3.16) provides an indication of the demanded

CRA position. A pulse stepping motor is connected in parallel to the A,C, and BB phases supplying the

CRDM. As the phases are energized, the stepping motor turns, driving a potentiometer which produces

a variable output corresponding to the demanded position. If a rod is dropped, tripped, or stuck, the

stepping motor continues to operate, resulting in an inaccurate position indication. In these instances,

the operator uses the reset pulser to drive the desired rod or group pulse stepping motor with a pulsed

24 V de signal rather than with the normal three phase input. This produces the correct RPI indication

without the motion.

Individual rod RPI is displayed on the position indication panel in the control room. However,

the primary use for the RPI is sequence monitoring. The sequence monitor performs the following:

- checks for overlap between the regulating groups,

- checks for excessive overlap,

- checks for greater than 25% overlap at discrete intervals, and if found, generates a

sequence fault signal.

3.6 Control Rod Drive Mechanism Cooling System

The Control Rod Drive Mechanism Cooling System is a closed system consisting of two

redundant trains. The major system components are two centrifugal pumps, two heat exchangers, and

a surge tank. The pumps provide cooling water to a common supply header which supplies the stator

cooling lines. The coolant discharges to a common discharge header and is cooled by the heat exchanger.

The separate cooling loops are fabricated entirely of stainless steel, with the exception of the

copper-nickel cooler tubes. The latter reduce the concentration of ferrous particles in the cooling water,

which limits the potential for flow blockage resulting from the attraction of the ferrous particles by the

magnetic field produced in the stator.

The cooling water system typically is operational during normal reactor operations only. Forced

air cooling is also supplied tb the top of the reactor vessel head to supplement CRDM cooling.

3-23
LEGEND

NO .NORMALLY Opt"

NC NORMAPLY CLOSED

} RELATIVE POSITION INDICATORS

POSITION INOICATION

PANEL

r------ -- -

I

I

I

-- ---- 1

STEPPI O

I.

I

I

I

I

A

FROM CONTROL c

ROO ORIVE -

MOTOR so

I NC AVERAGE T oOGROUP

AMPLIFIER

(GROUPS S- ONLY)

I I R~~~~~~~~~~FELATIVE POSITION

RESET RELAY V | AMPLIfiEt INDICATION

(ROD SELECTED ON CRO ANlO POSITION

POSITION RESET SWITCH IN _ S - I

jiRAISE OR LOWERI _

IRCLATIVE POSITION INDICATOR ;ROM ASSOLUITE POSlnION AMPLIFIER ABSOLUTE POSITION NOI

L...........NOCATI ELICTLO

Figure 3.16 Relative Position Indication

3-24
4. OPERATING AND ENVIRONMENTAL STRESSES

The B&W and CE CRD systems are subjected to a variety of operating and environmental

stresses, which, over the design life of the components, may lead to age degradation. Common

mechanical stresses include wear, fatigue, vibration, and corrosion. Electrical stresses result from arcing,

power surges, electrical noise and drift. Temperature, radiation, and humidity are common

environmental stresses. Externally induced stresses, such as abnormal operating conditions, improper

or excessive maintenance, testing, and human error may also result in component and system aging.

These stresses, acting in combination tend to produce greater synergistic effects than if they were acting

individually.

The design and the location are primarily responsible for determining which stresses affect

individual system components. The drive mechanisms and rod position indication systems, which are

located on top of the reactor vessel, are subjected to severe operating and environmental conditions. The

components of the power and control systems, in comparison, are located in a more controlled

environment outside of the containment.

Aging failure mechanisms result from the long-term exposure to operating, environmental, and

external stresses. Component degradation results in a decrease in physical properties and functionality,

affecting the component, system, and plant safety. This section describes the individual operating stresses

and the aging effects for the major CRD system components. A qualitative assessment on the probability

of the individual stresses affecting the major components is also provided. Though both the B&W and

CE CRD systems are mechanically different, the actual stresses and aging effects are equivalent.

4.1 System Operatine Stresses

The following stresses affect the B&W and CE control rod drive systems during operation.

These stresses, acting individually or in combination, may significantly degrade the components.

* Mechanical Wear The physical interaction between the system's components produces significant

frictional forces. Over time, these forces may cause material wear, galling, or fretting. Control

rod and guide tube fretting is an example of the wear caused by coolant flow induced vibration.

* - Cyclic Fatigue: Cyclic fatigue results from the application of repeated loads. During the design

life of the CRDM, both high and low cycle fatigue occurs which can initiate cracks resulting in

component failure. Repeated thermal and mechanical loadings result in low cycle fatigue which

affects the material in the plastic region. High-cycle fatigue results from vibration due to highfrequency loading at low amplitudes.

* Debris and Crud: Debris and crud in the reactor coolant may be transported and deposited

throughout the primary system, including the CRDM and guide tubes. This debris may become

trapped in the components, preventing the full insertion of the control rods or result in

immovable drive mechanisms. Crud may also accumulate on the CRA latching mechanisms,

increasing the force required to couple or decouple it from the CRD.

* Reactor Trip: Plant trips result in rapid temperature and pressure excursions. These transients

are capable of inducing stresses on the CRDM, control rods and spider assemblies. Trips also

present a challenge to the other plant safety systems which may contribute to age degradation.

4-1
* Boric Acid Corrosion: Primary coolant leakage in high temperature areas, such as the reactor

vessel head, may cause the boric acid to boil, increasing its acidity and corrosiveness. Boric acid

crystals may accumulate and block the cooling passages for the stator coils, resulting in

overheating, increased thermal stresses, and component failure for the air-cooled CE CEDMs.

* Electrical Surge: Electrical transients, resulting from disturbances in the current supplied to

electrical components, can cause load changes, system faults, and component failures.

* Electrical Noise and Drift: Electrical noise and drift can produce electrical circuit perturbations.

If not detected and corrected in a timely manner, aging degradation or component failure may

occur.

* Electrical Arcing: Electrical arcing, primarily due to the presence of moisture or insulation

degradation, may produce localized stresses, leading to component failure or degradation.

* Vibration: Vibration caused either by CRDM operation or coolant flow, can cause physical

motion of the components. This displacement may eventually cause wear, crack initiation and

growth, galling, and component failure.

* Maintenance: Normal, regularly scheduled maintenance, designed to maintain system operability

may induce stresses on various components. The maintenance performed at each refueling

outage is a typical example. The power and instrumentation connections, which must be removed

at each refueling, may cause connector wear or fatigue resulting in failure.

* Testing: To ensure the operational readiness of the CRD system to perform its safety functions,

regular system testing is required. These tests range from actual drop-time testing of control

rods, to electrical checks for the power and control components, and result in a significant

amount of testing conducted during a components lifetime. The characteristics of the test itself,

such as meggering, may be deleterious. Mechanical stresses induced from making and breaking

electrical connections, can cause it to degrade, affecting the conduction capabilities, which may

lead to eventual component failure.

* Human Error: To maintain the operational readiness of the system, numerous tests and

inspections are performed. Human error in performing these tests may cause significant

mechanical and electrical stresses, accelerating age degradation. The significant effects resulting

from such errors include dropped rods and plant scrams.

4.2 Environmental Stresses

The primary environmental stresses which affect the CRD system and components are

temperature, humidity, and radiation. Operational experience demonstrates that both systems are

susceptible to environmental stresses.

Plant location is the main factor which determines the degree to which system components will

be affected by environmental stresses. The power and control cabinets, located outside the containment,

are generally not exposed to extreme environmental conditions. However, if the electrical cabinet

ambient conditions are not controlled and monitored, localized overheating and component failures may

occur. The CRDM and power cables located in containment, and are exposed to extreme environmental

stresses, which have been demonstrated to cause system failures.

4-2
* Temperature: Temperature is the dominant environmental stress for the CRD system. Cabinet

temperature must be controlled by a forced air system to ensure proper operation of the

modularized power and control components. Cooling system malfunctioning can result in

localized component overheating and failure.

High temperature may lead to material degradation of the pressure housing and cables located

inside containment. Thermal embrittlement of cast Type 304 stainless steel may occur in the

reactor environment. The thermal gradients which occur as a result of reactor trips may also

contribute to low cycle fatigue. CRDM coil and power cable insulation deteriorate in a high

temperature environment, resulting in electrical shorts, dielectric property changes, and

decreases in material strengths. Gaskets, which serve as the primary coolant seal, may also

become brittle and crack at high temperatures.

* Humidity: The containment atmosphere above the reactor pressure vessel head is very humid.

Electrical components located in this area may experience degradations in dielectric properties

of insulating materials, decreasing electrical insulation integrity, resulting in electrical shorts in

the CRDM coils. Moisture may also corrode connector surfaces, interfering with the circuit

current flow and component operation.

* Radiation: For CRD components located outside the containment, radiation is not a severe

degradation mechanism. However, for the CRDM leadscrew, spider mechanism, and power

cables, long-term radiation exposure may present significant stresses. Cables and connectors

experience decreases in dielectric and strength of insulating materials in a high radiation field.

Also, any components fabricated from cast 304 stainless steel are susceptible to irradiation

assisted stress corrosion cracking.

4.3 Effect of Operatine and Environmental Stresses on System Components

As discussed in Sections 2.0 and 3.0, the main sub-systems of the CE and B&W CRD system are

the control rod assemblies, drive mechanisms, rod position, and control systems. Each system is

comprised of individual components, fabricated from a variety of materials, which perform different

functions. The effect of a particular stress upon each is a function of the intensity, frequency, duration

of stress, and material strength.

This section describes the individual operating and environmental stresses which affect the control

rod drive subsystems and components. The potential degradation mechanisms and failure modes caused

by these stresses are summarized in Tables 4.1 and 4.2 for the B&W and CE systems, respectively.

* Control Rod Assemblies (B&W and CE): For both the B&W and CE systems, the control rod

assemblies consist of the spider assembly and the individual absorber rods. The spider assemblies

for each design maintain the control rods in the proper pattern for insertion into the fuel

assembly guide tubes. The primary difference between the B&W and CE control rod is the that

the CE absorber rod is a larger diameter, and consequently, stiffer. During plant operation, the

rods vibrate which can result in cladding wear, and through wall guide tube cracks as experienced

at several CE plants (Section 5.1.1). Control rod cracking may also be caused by other operating

stressors such as IGSCC, thermal and pressure fluctuations, and radiation induced poison

swelling.

4-3
Table 4.1 Babcock & 'Wilcox Control Rod Drive System

Potential Degradation Mechanisms and Failure Modes

Potential Degradation Potential Failure

Subsystem/Component Material Mechanism Mode

I. Control Rod Assembly

a. Control Rods lTpe 304 Stainless Steel Stress Corrosion Clad Cracking

Clad Cracking Poison Wash-Out

Ag-In-Cd Poison Mechanical Wear

b. Spider Grade CF3M Stainless Stress Corrosion Surface Cracks

Steel Cracking Dropped Rod

Mechanical Wear,

Radiation Embrittlement,

Fatigue

c. Fuel Assembly Guide Zircaloy4 Mechanical Wear Tube wall cracking

Tube Tube wall wear

IL Control Rod Drive Mechanism

a. Motor Tube Inconel clad with low Thermal Embrittlement, Housing Crack,

allow steel Corrosion, Fatigue Primary Coolant

Cracking Leaks

b. Rotor Assemblies Stellite Ni-Cr-Fe Alloy, Mechanical Wear, Dropped CRA,

(roller nuts, segment Type 403 Stainless Steel Fatigue, Debris/Crud Immovable CRA

arms, springs) Buildup

c. Leadscrew 17-4 PH Stainless Steel Mechanical Wear, Dropped CRA,

Fatigue, Stress Corrosion Immovable CRA,

Cracking Inoperable Locking

Mech.

d. Stator Coils Copper Wire, Dow Corrosion, Mechanical Dropped Rod,

Corning 997 Varnish, Wear, Insulation Electrical Short,

Kapton, Nomex, Silicone Degradation, Voltage Variation

Rubber, Contamination

e. Vent Valve Stainless Steel, O-rings Corrosion Buildup, Inoperable Valve,

Mechanical Wear, Primary Coolant Leak

Fatigue, Thermal

Embrittlement

III. CRDM Control Elec. Power Supplies, Corrosion, Fatigue, Dropped CRA,

System Semiconductors, SCR's, Mechanical Wear, Spurious CRA

Cables, Connectors, Thermal Degradation, Movement, Inoperable

Circuit Boards Contamination Rods, Electrical Signal

Drift

IV. Rod Position Indication Reed Switches, Stepping, Corrosion, Fatigue, Loss of Position

Systems Motor, Wiring, Cables, Mechanical Wear, Indication,

Connectors, Circuit Thermal Degradation,

Boards, Insulation, Semi- Radiation Degradation, Spurious Position

Conductor Devices, Vibration, Insulation Indication

Electro-Mechanical Degradation,

Components Contamination

4-4
Table 4.2 Combustion Engineering Control Element Drive System

Potential Degradation Mechanisms and Failure Modes

Potential Degradation Potential Failure

Subsystem/Component Material Mechanism Mode

I. Control Element Assembly

a. Control Element Rods Inconel Clad BC and Ag- Mechanical Wear, Crack aad Cracking, Poison

In-Cd Poison Formation Wash Out

b. Spider

Stainless Steel Mechanical Wear, Surface Cracks,

Radiation Embrittlement, Dropped Rod.

Fatigue, Cracking

c. Fuel Assembly Guide Zircaloy-4 Mechanical Wear, Tube Wall Cracking

Tubes Through Wall Crack Tube Wall Wear

II. Control Element Drive

Mechanism

a. Motor and Pressure Type 316 and 403 Stainless Thermal Embrittlement, Housing Crack,

Housing Assemblies Steel Ni-Cr-Fe Alloy Corrosion, Fatigue Primary Coolant

Cracking Leaks

b. Motor Latches, Links, High Cobolt Alloy Mechanical Wear, Dropped CEA,

and Pins Fatigue, Debris/Crud Immovable CEA

Deposition

c. Extension Shaft Type 304 Stainless Steel, Wear, Fatigue, Stress Locking Mech. Oper.

Chromium Plated Corrosion Crack Difficulty

d. Magnet Coils Copper Wire, Varnish, Corrosion, Mechanical Dropped CEA,

Fiberglass tape, Silicone Wear, Insulation Slipped CEA,

Degradation, Electrical Immovable CEA

Shorts

e. Vent Valves Type 440 Stainless Steel Corrosion Buildup, Inoperable Valve,

Type 316 Stainless Steel Mechanical Wear, Primary Coolant Leak

Seat Fatigue, Thermal

Embrittlement

III. CEA Control System Power Switches, Detectors, Corrosion, Fatigue, Dropped CEA,

Relays, Cables, Mechanical Wear, Spurious CEA

Connectors, SCR's Thermal Degradation, Movement, Inoperable

Electrical Noise Drift CEA

IV. Rod Position Indication Reed Switches, PC Boards, Corrosion, Fatigue, Loss of Position

System Cables, Connectors, Mechanical Wear, Indication Signal,

Thermocouples Electrical Thermal Degradation, Incorrect Position

Wiring, Insulation, Semi- Radiation Degradation, Signal, Spurious CEA

Conductor Devices, Vibration, Insulation Movements,

Electro-Mechanical Degradation Erroneous Core

Devices. Penalty Factors

4-5
* SpiderAssemblies (B&Wand CE): The spider assembly is subjected to numerous stresses while

located in the upper internal area of the reactor core. Flow induced vibrations from core coolant

flow cyclic fatigue, variations in water chemistry and thermal stresses from reactor trips are all

potential causes of spider degradation. Another potential source of mechanical stress is system

maintenance. The coupling and uncoupling of the spider assembly from the leadscrew may

contribute to wear of the coupling mechanism. Corrosion and crud from the fuel and other

reactor internals may become lodged in the coupling mechanism, increasing the mechanical force

required to uncouple the CRDM from the control rod assembly.

* Control Rod Drive Mechanisms (B&W and CE): The positioning of the control rod drive

mechanisms on the top of the reactor vessel head expose it to numerous stresses which may

produce aging. These stresses may result in degradation which prevents the CRDM from

performing as designed. The two major pressure housing failures which are possible include a

crack resulting in a small break LOCA, and jamming of the drive mechanism, so it is inoperable

and unable to insert the rods upon demand. The stresses which affect major subcomponents are

identified below. All of the CRDM subcomponents which are exposed to the primary coolant

are fabricated from corrosion-resistant stainless steel or Inconel.

* Motor Housing (B&Wand CE): The pressure housing forms part of the reactor system pressure

boundary between the coolant and the containment. The CE housing is fabricated from Type

403 stainless steel with Inconel end fittings, with the upper pressure housing fabricated from Type

316 stainless steel. The Babcock & Wilcox motor tube is a three piece welded assembly with

either stainless steel or a low alloy steel motor tube wall clad on the ID with an Inconel center

section. The lower end is welded to a stainless steel forging. Any cast stainless steel CRDM

housings which are still in use are susceptible to thermal embrittlement, which can cause a

decrease in the fracture toughness leading to crack formation. These cracks may result in

primary coolant leakage. Transgranular stress corrosion cracking caused by the chlorides and

sulfates contained in the coolant may also induce stress corrosion cracking. Vibration from

normal plant operation may also be transmitted to the pressure housing. Continuous vibration

may result in low cycle fatigue enhancing crack propagation.

* LeadscrewAssembly and Extension Shaft (B& Wand CE): The B&W leadscrew and CE extension

shaft assemblies provide a means for control rod/spider attachment which permits the positioning

of the control rods in the core. The B&W leadscrew uses a modified ACME thread which allows

the roller nut to engage without lifting the screw. CE uses a Type 304 stainless steel drive shaft

with notches machined along the OD. The main stressor arises from the continuous mechanical

interaction with the gripper or roller nut during operation. Because the leadscrew assembly is

exposed to the primary coolant, any crud or debris in the coolant can be transported to the

CRDM internals, which may jam and prevent movement of the leadscrew. Crud may also

accumulate in the spider locking mechanism at the lower end of the leadscrew resulting in a

greater mechanical force required to decouple the leadscrew and the spider. The coupling

mechanism is exposed to a greater radiation fluence than the other system sub-components, due

to its close proximity to the fuel. This cumulative radiation exposure should be monitored or

calculated to ensure that the IASCC threshold of 5 x 10' nvt has not been exceeded (as

described in Section 5.2.2).

* Rotor and Motor Assemblies (B&W and CE): The Combustion Engineering gripper assembly

utilizes a series of mechanical latches (typically three) to raise, lower, or hold the control rod

drive shaft in response to signals from the operator or control system. These latches are

4-6
fabricated from a high cobalt alloy to minimize wear. The B&W rotor assembly consists of a

rotor tube supported by ball bearings, which contains two arms carrying a pair of roller nut

assemblies. Mechanical stress resulting in wear may occur since physical contact is continuously

maintained between these components during normal operations. These components must be

able to move freely. The latches and roller nuts are susceptible to a buildup of crud and debris

which may cause them to jam. While this may not prevent operation, a greater mechanical force

will be required to overcome the interference. Cyclic fatigue is also a stressor due to the

repetitive nature of the latching operation.

* Stator Coils (B&Wand CE): Both designs use dc coils to supply the magnetic force to actuate

the latches or roller nuts. The CE coil stack assembly consists of five coils located concentrically

to the pressure housing. The coils are fabricated from round copper wire, insulated with a high

pressure enamel, and vacuum impregnated. The coil is then wrapped with fiberglass tape and

a silicone compound. The finished coils are contained in a nickel plated steel housing. The

B&W coils consist of six phases in the same physical location in the stator (wound on top of each

other). The windings are either bifilar or monofilar, and six phase star connected. The stator

assembly is surrounded by a cooling water jacket. Thermal stresses from continued operation

or localized overheating may affect the operation of the coils. The forced air and water cooling

system is designed to maintain the coils below approximately 350'F. High ambient containment

temperatures and insulation degradation could result in localized stator temperatures exceeding

this limit resulting in electrical shorts. Power is supplied to the coils through cable connections

on the top of the CRDM. Normal maintenance which requires connecting and disconnecting

these cables during refueling may cause connector wear and damage. This could interfere with

the power supplied to the coils, affecting the magnetic field, resulting in improper latch

operation. Boric acid corrosion from primary coolant leaks or spills is also a major stressor. The

corrosive effects may degrade the insulation, leading to electrical shorts, which could result in

dropped rods.

* Vent Valve (B&Wand CE): The vent valve, located on the top of the pressure housing, is used

primarily to vent the non-condensible gases in the upper portion of the housing. Combustion

Engineering uses a ball-seat type of vent, while B&W typically uses a vent plug and o-ring design.

The mechanical stresses from the repeated vent valve operation have caused mechanical wear

and galling, o-ring deterioration, and primary coolant leakage. The same low-cycle fatigue and

vibration which may affect the pressure housings also affect the vent valve. The vent valve is also

susceptible to improper maintenance induced damage. The high temperature and radiation

environment may cause o-ring embrittlement and failure. Spring relaxation may also result in

insufficient closure force resulting in leakage.

* CRD Control System (B&Wand CE): The control rod drive control system consists primarily of

cabinet mounted electronic components necessary to convert the main power supply to the pulsed

dc power necessary to power the main coils. The logic circuitry controls the sequence of the

power to the coils in order to produce the proper actuation of the latches or roller nuts.

Temperature, humidity, electronic noise and drift, vibration, and human error may significantly

affect this subsystem. Forced air cooling is provided in the cabinets to dissipate the heat

generated by the electronic components. The ambient temperature around the cabinets must

also be controlled. Any increase in temperature, either within the cabinet or outside, may cause

degradation and failure. Electrical noise or drift, due to component aging is also detrimental to

the operation of the system. Vibration from the plant surroundings may be transmitted through

the mountings to the equipment, causing movement of the internal components, which may

4-7
resulting in age degradation. Human error and improper system maintenance may also cause

significant mechanical stresses. Operational effects resulting in component failure, such as

dropped rods or plant scrams, may result from these failures.

* Rod Position Indication Systems (B&Wand CE): Both the B&W and CE CRD systems use two

independent systems to provide rod position indication. The absolute position indication system

utilizes magnetic actuated reed switches, which when closed, provide a voltage corresponding to

rod position. The second system is a relative position indication system, which monitors the

pulses supplied to the CRDs, providing demanded rod position. Because the reed switches for

the absolute position indication system are located in a tube adjacent to the motor tube, they are

subjected to the same operating and environmental stresses as the CRDM. Opening and closing

the switches may cause the contact surface wear, and over time, fatigue degradation. Vibration

transmitted through the housing from CRDM operation may also induce mechanical stresses

affecting switch calibration. Like all electronic equipment, excessive temperature and humidity

may interfere with the proper operation of the equipment. Signals from the reed switches are

transmitted through the cables and connections mounted on top of the housing. Primary coolant

leakage would expose these cables and connectors to the corrosive effects of boric acid which

would interfere with the signals transmitted and degrade the cable insulation. Electrical noise

and other interference can cause the CE and B&W relative position indication systems to

produce anomalous signals. Deterioration of the cable and connectors from vibration, heat,

radiation, and maintenance errors will also degrade the system.

4.4 Asine Stressor Tables

Tables 4.3 and 4.4 provide a qualitative evaluation of the significance of each of the operational

and environment stresses which were discussed in this section and summarized on Tables 4.1 and

4.2. These Tables also highlight the areas where predictive maintenance may be applied. This

aging stressor evaluation is based upon an engineering analysis of the component materials,

design, environmental and operational conditions, and the operating experience for each CRDM.

Generally, a "high" (H) ranking indicates that operating experience supports the assessment, while

a "low" (L) ranking indicates that neither operating experience nor the engineering design

analysis provide evidence to warrant this stressor being an important aging concern. A "medium"

(M) ranking represents a moderate aging effect resulting from a particular operating or

environmental stress.

4-8
Table 4.3 Babcock & Vilcox CRD System Stress Summary

Environmntal

Operating Stresses Stresses Potential Aging Mecianissna

._I I I I I I I .g - ,I I ; I

11

I

W0

-it

I

U,

*0 us!

I

iz

a

I

C

.2

S

I- I

wt

4

.31

C

Ig I

E

a.

E

II0

CA

F5

t;C I

I

E

a1

I

-Ii

-15 .1. I

I

U

aW

Ea

Component

||Con.RodA|L|M|H|L|L|n Cot. L |L |na na M 1M L L 11M L H M L M na L M 1 Rods ,. , L . I 

Spider M H M L L na L L L na na M M L L M L M M|na|

CRDM

Motor Tube M M H | n | L L M na na L M M M L M M M M M on L M

Flexatallic Gasket M L H L H na M L M na n a |n H H L H H na na H H na L na

Coils L M na L H H M M M na H| L M fH IH L H na L H M H na on

Vent Valve H M H L H na H L |1 na na L M M L M H M L M M na L L

Leadscrew M M H L L na L L L na na L H M L M L L L M M na M M

Rotor Assy M M 11 L L na L L L na na L M M L M L L L M M ona L L

TorqueTaker M M |1 L L na M L M |n n L M M L M L L L M M na L L

Snubber M M |H L L na L L L on n L M M L M L |L M M na L L

Leadscrew Guide L M | L L na L L L on n L M M L L L L L M M on L L

||Rod Post. Ind. i Reed l Switch M| M na L M M | L M M M M na L

Cable & Conn. L | H na L H H M M M M M M H |1 M L H na M H H Hf na na

Electronics o| na M na L M H L M M M M M M | na M na j |L I | M M na na

||CRD Control Sys

SCR L M na L n H L M M M L M L M M L na na M H na na ona na

Power Supp. na M |na L na H M M M H M M L M M na na na L H |n J3 na |n

Programmers M M na L na H M L|MH na na | n _ na | n|

Legend: H

M

L

tt

= High probability that stress or aging mechanism will cause component degradation.

= Medium probability that stress or aging mechanism will cause component degradation. = Low probability that stress or aging mechanism will cause component degradation. = Stress or aging mnedanism does not affect component.
Table 4.4 Combustion Engineering CED System Stress Summary

Eaviroamental

Operating Stres Strewss Potential Aging Meelianionn

i 5

11

3

cr

g

b

Ij

U

.17 *1

I-i j

ZN I I.

iU I

I I

-g

a

.5

I I I

C.,

i I s..

0

Component i

ControlEle.Rods M H L L L na na M L L L H MM

Spider - H L|r L L L I na M|M| M |L L M|M| M M |n M

CRDM

Motor Housing L M L M H na L L M n| | L M M M L M M M M M | a L M

Coils M iu L _ H H M M M |ta H L M Hi Hi L H |sa L M M |If nI D

Vent Valve H M H L H na M M M na If L M M L M H M L M M |a L L

Ex:ension Shaft M M H L L na L L L T Ni Da L H M L H L L L M M na M M

Mechanical Latches M M H L L tu L L L na na L M M L H L L L M M Di L L

Omega Seals L L L M H ni H H H | na| L H H| L L H | IUa H | M |a |L |

RPI T _

Reed Switch M M |IU L M H L L L M L M M M M M M a M M M M na L

Cable & Converters L H na L H H M M M L M M H H M L H na M H H H na L

Electtonics || na | M Da na | M | naL H M Min L

Control System I I I I I

Diod & SCR's L M |na L a | M L M M L na na M Ht na na |I L

PowerSupp. n&a M | n na H M M M H M M L M M na na Tna L H 111 na t a L

Power Switch Assy M M H |L M M M L M L M na |na M H lna na |a L

Lerend: H

M

L

= Hligh Probability that stress or aging mechanism will cause component degradation.

= Medium Probability that stress or aging mechanism will cause component degradation.

- Low Probability tlat stress or aging mechanism will cause component degradation

- Stress or aging mechanism does not affect component.
5. USNRC AND INDUSTRIAL STUDIES

Both the B&W and CE control rod drive systems have been the focus of on-going NRC and

industrial research. Much of this work has been in response to specific operational occurrences, such as

failures of the control rod and guide tube, and cobalt reductions required by ALARA commitments.

Recently, emphasis has shifted to component aging and plant life extension concerns. This section

summarizes the programs which have provided pertinent information in attaining the goals of the NPAR

program. In addition, these programs document the system and component failures and degradation

caused by continued exposure to the operating and environmental stresses discussed in Section 4.0.

For the ten year period beginning in 1980, the B&W and CE control rod drive systems have been

addressed by the NRC through regulatory notices and research programs. Four Information Notices

published during this period alerted utilities to operational occurrences resulting in events ranging from

immovable control rods to Technical Specification violations. The annual fuel performance reports

highlight any significant operational occurrences with the control rods and CRD mechanisms. All-of

these documents pertinent to CRDM aging are summarized below.

Draft NUREG-1299 contains the Standard Review Plans (SRPs) addressing plant license renewal.

These SRPs specifically address the aging of the CRDMs and related sub-components, and establish the

requirements for demonstrating their acceptability for license renewal. This NUREG is also summarized

in this section.

5.1 USNRC Information Notices

5.1.1 Information Notice No. 85-38: Loose Parts Obstruct Control Rod Drive Mechanism

This Information Notice alerted all B&W licensees of several incidents which occurred at Davis

Besse, where loose parts lodged in the CRDMs, preventing rod movement. Subsequent investigations

revealed that the loose parts were a broken set screw from a handling tool and four broken locking

springs.

An inadequate assembly procedure was the cause of the broken locking springs. Prior to these

occurrences, a technician was required to verify the correct positioning of these springs by "feel' using

a long-handled tool. If the spring was not positioned properly, it would hit the inside of the torque tube

cap and snap when the leadscrew was fully withdrawn. The brittle, intergranular nature of the fracture

surface, as well as the gouges on the cap, confirmed this failure mechanism.

Corrective action at Davis Besse consisted of examining all the CRDMs for additional broken

springs, and visually verifying that the locking springs were properly placed.

5.1.2 Information Notice No. 85-86: Lightning Strikes at Nuclear Power Generating Stations

This Information Notice alerted PWR owners of the potential for reactor trips and instrument

damage resulting from lightning strikes. Solid state circuitry was particularly susceptible to lightning

induced line surges.

ANO-2 (CE) experienced a trip from 100% power on a low DNBR signal resulting from a

lightning strike transient induced in two of the core protection channels. Subsequent investigations

revealed no additional equipment or instrumentation damage.

5-1
A similar occurrence at Zion (Westinghouse) resulted in RCCA power system fuse failures,

disabling several DC power supplies, resulting in control rod drops.

5.1.3 Information Notice No. 86-108: Degradation of Reactor Coolant System Pressure Boundary

Resulting from Boric Acid Corrosion

This Information Notice, including two supplements, documented the deleterious effects of boric

acid corrosion from primary coolant leakage. Though the leakages occurred at a high-pressure injection

nozzle and a instrument tube seal, similar results would occur from CRDM pressure housing leaks.

Boric acid will rapidly corrode ferritic (carbon) steel components, and is most pronounced on

metal surfaces which are cool enough to remain wetted. Corrosion rates in excess of one inch per year

have been experienced in plants and laboratory tests where low quality steam from borated reactor

coolant impinged upon a surface, keeping it wet. If the metal is hot, then the surface remains dry, and

this loss of electrolyte will slow the corrosion rate. If the leakage occurs in hot surroundings, such as the

reactor pressure vessel head, then the boric acid solution boils, increasing its acidity and corrosiveness.

The evaporation of the water also causes boric acid crystals to accumulate, which could block the CRDM

cooling passages, as occurred at Turkey Point (Westinghouse).

Research conducted by Westinghouse indicated the corrosion rates may have been

underestimated. In one series of tests, an Inconel CRDM head weld with typical crevice geometry was

exposed to dripping 15% boric acid at 2100F. Extensive general corrosion of the steel occurred (400

mils/month), with no preferential attack to the crevice or the Inconel.

Such leakage could also degrade non-metallic components, such as the insulation on cable, stator

coils, and other electrical components.

5.1.4 Information Notice No. 8847: Slower-Than-Expected Rod-Drop Times

This Information Notice documented a violation of Technical Specification control rod drop

times, resulting from a test methodology change at ANO-2. The new test interrupted the power to all

the rod drive mechanisms simultaneously, using the reactor trip breakers. Previously, power was

interrupted individually to each CEDM. Simultaneous interruption of power is a more realistic

representation of the desired safety function, but leads to a slower dissipation of the stored energy in the

holding coils. This resulted in a maximum of 3.2 seconds drop time for all the rods, which exceeded the

maximum 3.0 seconds Technical Specification requirement.

A review of all of the FSAR accident analysis, incorporating the increased drop times was

required to ensure conformance with all of the original licensing basis.

5.2 USNRC Research Programs

5.2.1 Control Rod Guide Tube Wear in Operating Reactors, NUREG-0641, April 198017

This NUREG, published early in 1980, documented the discovery of guide tube cracks at

Millstone 2 after the first cycle of operation. The wear was attributed to flow induced vibration of the

Inconel clad control rods rubbing against the Zircaloy guide tubes. On-site inspection revealed that the

maximum wear corresponded to the location of the control rod tips when the rod was in the full out

position.

5-2
-Slots

Holm

I | LI No Gap

Zircaloy Guide Tube

Expanded Section

(not to scale)

Figure 5.1 Combustion Engineering Guide Tube Sleeve

To prevent this type of wear, chrome-plated stainless steel sleeves (Figure 5.1) were installed in

the upper portion of the guide tubes and fuel assembly modifications were proposed which would modify

the coolant flow characteristics. The insert was held in place by of a mechanical crimp, which produced

a bulge in both the sleeve and guide tube. This crimp prevented axial motion of the sleeve in the cold

condition. Differential thermal expansion between the stainless steel sleeve and the Zircaloy guide tube

provided additional resistance against axial motion at operating temperatures.

Eddy current inspections after one cycle of operation with the sleeves revealed no detectable

wear. CEA examination confirmed that the rod tips experienced no appreciable wear from contact with

the harder surface of the stainless steel sleeve inserts. To assure the continued adequacy of the crimps,

Combustion Engineering developed a pull test which was performed for each crimp.

5-3
This NUREG also evaluated the Babcock & Wilcox guide tube design for similar problems.

From their evaluation, the NRC concluded that B&W plants were not susceptible, based primarily upon

three main considerations:

1) the lack of indicated through wall cracks based upon the results of air tests,

2) design differences between CE and B&W, particularly the greater flexibility of the B&W

rods, and the use of stainless steel cladding rather than Inconel, and

3) design similarities between B&W and Westinghouse, which also did not experience guide

tube wear.

5.2.2 Fuel Performance Annual Reports for 1980-1988, NUREG/CR-2410, 3602, 3950 Vols. 1-61%19J°

These annual NRC reports on LWR fuel performance, began in 1978. They include information

on continuing research and development projects, trending data, and significant performance and

operational problems. Information on control rods and control rod drives are also included in these

reports.

Broken fuel assembly springs were first identified at three B&W reactors in 1980. NRC, utility

and vendor analysis concluded that most failures were due to fatigue initiated cracking, followed by stress

corrosion crack propagation, which eventually caused the spring to fail. These failures were not

considered a significant safety hazard, however, the potential for loss of positive hold-down force, loose

parts, and interference with CRA movement were identified. Since the single failure discovered at

Crystal River did not also appear to be stress corrosion cracking related, further site inspections were

required of the licensees. These inspections identified 64 other spring failures (Table 5.1); no failures

were reported in 1984, 1985 and 1986.

Table 5.1 Events Involving Hold-Down Springs

Year Plant | Number of Broken Spring(s)

1988 Arkansas-I 4

Oconee-2 4

Oconee-3 5

1987 Crystal River-3 9

Oconee-3 3

1983 Oconee-1 4

1982 Davis Besse-1 1

Oconee-1 1

Oconee-2 4

Oconee-3 2

1981 Arkansas-I 1

1980 Crystal River-3 1

Davis Besse-1 20

Oconee-1 S

54
The 1986 report first identified the potential for irradiation assisted stress corrosion cracking

(IASCC) for core structural materials, including the control rods and their drive mechanisms. Reactor

components with a cumulative fluence above the JASCC fluence threshold of S x 10° nvt may be

susceptible to IASCC. Because many CRDMs remain in service for up to 40 years, their total fluence

may exceed this limit, particularly the leadscrew and coupling mechanisms.

These annual reports highlighted the deleterious effect of boric acid corrosion, and its potential

effect upon reactor operation. The 1987 report documented the failure of a reactor vessel head o-ring,

causing boric acid build-up on the CEDMs, which eventually led to overheating of the upper control

element gripper coils.

Numerous CRD operational, maintenance, and procedural deficiencies were described in these

reports. Table 5.2 presents some of the major CRDM failure events and their particular cause. There

were no events reported in 1982-1984, 1987 and 1989. The guide tube wear problem identified by

Combustion Engineering, the R&D programs instituted, and the design of the guide tube sleeve were

also discussed in the annual reports.

Table 5.2 Cause of Events Involving Control Rod Drive Mechanisms

Cause of Event

Screws SpringsI

Overheating Cracking

Improperly ofRod Of CRD

Year Plant Loose Broken Stated Broken Gripper Coils Components

1988 Millstone-2 X(a)

Palisades X(b)

1986 Crystal River-3(c)

Palisades

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ~~~X(d)

1985 Davis Besse-1 X X X

1981 Davis Besse-1 X

(a) Event caused by boric add buildup on CRD mechanism coils.

(b) Eleven more cracked housings found.

(c) Cause of event not determined.

(d) Contaminant resulted in transgranular stress corrosion cracking of three control rod drive seal housings.

5.2.3 Residual Life Assessment of Major Light Water Reactor Components -- Overview,

NUREG/CR4731, November, 198921

This NUREG assessed the aging of selected, major light-water-reactor components and

structures, including reactor internals and PWR control rod drive mechanisms. The stressors, potential

degradation sites and mechanisms, failure modes, and inspection requirements were qualitatively

discussed.

The important CRD stressors included operational thermal transients, CRD stepping,

temperature, radiation, and corrosion. The primary degradation sites subject to fatigue damage include

the pressure housing, seal welds, and the control rod drive coupling mechanism. Typically, the fatigue

5-5
usage factor for these components was low, 0.1 as compared to an allowable factor of 1.0. Mechanical

wear is a concern for the CRD internals, such as the latches, drive rods, and roller nuts. The actual wear

on these components may take the form of bearing spalling, or rubbing on roller nuts, latches, and other

mating parts.

The past occurrences of boric acid leakage, and its deleterious effects were discussed. Laboratory

tests indicated that the corrosion rates for both low alloy and carbon steels may be greater than

previously estimated when exposed to the primary coolant, therefore, the need for adequate leakage

monitoring procedures was highlighted.

Three potential failure modes for CRDMs which would affect safety were presented, including

rupture of the CRDM pressure housing, pressure boundary leakage, and failure to insert the control rods

upon demand. The probability of pressure boundary leakage is higher than rupture, which would have

the same effect as a small break LOCAL Seal degradation was a common cause of leakage. Mechanical

binding or interference, particularly when caused by loose reactor internals, was also discussed.

Various in-service inspection requirements were presented, particularly those required by the

ASME Boiler and Pressure Vessel Code, Section XI. Some common non-destructive inspection methods

used by nuclear plants to meet these requirements were also presented.

Based upon information presented, the critical CRDM sub-components with respect to plant

aging, are listed in Table 5.3. Stressors, degradation and failure mechanisms are also provided for each.

aging:

The following general recommendations were made on the detection and mitigation of CRD

a) increased inspection requirements for all the CRDM pressure housings and welds,

b) improved primary leakage detection methods, and

c) periodic inspections of CRD mechanisms to assure the validity and accuracy of the

vendor life estimates.

Table 5.3 Summary of PWR CRDM Degradation Process

Inservice

Inspection

Degradation Surveillance

Site Stressor Degradation Mechanisms Potential Failure Modes Methods

Pressure Thermal stress, high Thermal embrittlement Cracking, leading to Volumetric or

Housing temperature water Low-cycle fatigue leakage Surface

Latch Assembly Loose parts impacting, metal Fretting, wear, spalling Binding, stuck rods None

to metal contact

Coil Stack Moisture, temperature, Insulation breakdown, Dropped rods None

radiation electrical shorting

Drive Rod Rubbing, impacting Wear, low-cycle fatigue Uncoupling of control None

l _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ .__ _ _ _ _ _ _ _ assem bly

External Boric acid (if leak is present) Boric acid corrosion Leaks None

Components

5-6
5.2.4 Proceedings of the USNRC Fifteenth Water Reactor Safety Information Meeting,

NUREG/CP-0091 Vol. 3, October 198722

This NUREG contains the conference proceedings of the 15th Water Reactor Safety Meeting

held on October 26-29, 1987 at.Gaithersburg, MD. One paper presented at this conference, "Technical

Safety Issues Related to Residual Life Assessment of Major LWR Components and Structures," by V.

Shah and P. MacDonald of INEL assessed aging degradation of selected LWR components, including

PWR CRDMs.

The four main conclusions are similar to those reported above for NUREG/CR-4731:

1) Only 10% of the peripheral CRDM pressure housing welds are required to be inspected.

Because interior CRDM welds have cracked and leaked, the inspection requirements

should be revised to include all of the CRDM pressure housings.

2) The actual degree ofthermal embrittlement for pressure housings fabricated from cast

stainless steel needs to be monitored.

3) Representative CRDMs should be removed from service periodically and inspected for

wear. When returned to service, they should be rotated and placed in a different core

location.

4) Cumulative travel of the leadscrew should be monitored and compared to life test results,

and the need for replacement determined.

5.2.5 USNRC Standard Review Plan, NUREG-0800, Rev. 1, July 1981'

This document consists of individual Standard Review Plans (SRPs) which provide guidance to

the NRC staff for the review of utilities operating license submittal and amendments. Generally, these

'SRPs define the minimum structural, material, and operating requirements for the systems during normal

and upset conditions. System and component degradation due to aging is not specifically addressed, nor

are any inspections required to assess the effect of aging upon system and component operability.

Four sections of NUREG-0800 are specifically applicable to the design and operability of the

CRDMs:

1) Section 3.9.4 Control Rod Drive Systems -- This section defines the minimum design

requirements for the actual mechanism portion of the CRDMs. Specific references are

provided to industrial codes and standards to which the design must adhere. Loading

combinations are defined to insure operability under all normal and upset conditions.

Life cycle tests for all new designs are also required to insure operability for the design

life.

2) Section 3.9.5 Reactor Pressure Vessel Internals -- This section provides the same

requirements as Section 3.9.4 for the reactor internals, including the individual control

rods and guide tubes. Minimum design requirements are provided to insure that these

components will not fail or degrade in a manner which would not allow the CRDMs to

function as designed to control reactivity excursions or to provide for a rapid shutdown.

5-7
3) Section 4.5.1 Control Rod Drive Structural Materials -- This section defines the

material requirements for the CRDM, including material cleaning and cleanliness

controls. Materials and material conditions which are susceptible to SCC type failures

are not acceptable for the CRDMs. Materials used in the design of the CRDM are

required to conform to Reg. Guide 1.85 "Code Case Applicability for ASME Section III

Material." Adherence to this guide provides additional assurances that the CRDM

materials will not be susceptible to SCC failures for the life of the CRDM.

4) Section 4.6 Functional Design of Control Rod Drive System -- This chapter defines the

requirements necessary to insure the operability of the CRDM and supporting systems

for all normal and upset conditions. Specific emphasis is placed on the adequacy of the

CRD cooling system. Electrical systems and instrumentation are also verified operable

under all modes of operation, so their failure will not prevent the CRDM from

operating. The total CRD system must be designed so that no single failure will render

the system inoperable.

5.2.6 Standard Review Plan for the Review of License Renewal Applications for Nuclear Power

Plants--Draft Report, NUREG-129924

This draft SRP provides guidance to utility and NRC personnel in preparing and reviewing of

safety evaluations applicable to nuclear plant life extension. The main emphasis is to insure that aging

will not adversely affect the component performance for the term of the license renewal. This is

accomplished by the following measures:

1) identifying the systems, structures, and components whose functions and performance

must be assured,

2) insuring that significant age-related degradation has been identified and evaluated,

typically by performing an Integrated Plant Assessment (IPA),

3) verifying that an adequate management program for age-related degradation has been

implemented to ensure that none of the current licensing bases have been compromised.

This program should specifically address surveillance, maintenance, trending,

replacement, refurbishment, and operating life assessment, and

4) identifying and providing specific details on how aging will be assessed in any structure,

system, or component not specifically addressed by such a program, or why significant

age-related degradation is not anticipated.

The PWR Control Rod Drive System is specifically addressed by this Draft NUREG.

Degradation process information, similar to Table 5.3 (Section 5.2.3) is also provided. The importance

of considering age-related effects for system sub-components is also discussed. Guidance is provided for

the following CRDM subcomponents.

a) piping,

b) cables and wiring,

c) relays, circuit breakers, and switchgear,

d) electrical motors,

e) sensors,

5-8
f) electronic components, and

g) electronic devices.

5.3 Industrial Research

Component and system research, primarily sponsored by EPRI, dealt mainly with the

performance of the control rods and the CRD mechanisms. More recently, the PWR plant life extension

project at Surry addressed the aging of various systems and components, including the control rods and

drive mechanisms. These reports are summarized below.

5.3.1 Wear Measurements of Nuclear Power Plant Components, EPRI NP-3444, May 1984 2

This report presented the results of an EPRI sponsored program, performed jointly by Battelle

and Combustion Engineering, to obtain wear data for some major reactor components in support of the

EPRI cobalt reduction program. Two CEDMs, one from Millstone-2 and a System 80 design, were

included in this study. The Millstone-2 CEDM was representative of similar mechanisms installed at

Maine Yankee, San Onofre 2 and 3, Arkansas 2, and St. Lucie 2. The primary differences between the

two design types were that the Millstone CEDM had an extra gripper assembly which served as an antiejection device, and an improved latch-pin-capture method which minimized rotational wear by using

square headed pins.

Both CEDMs were subjected to accelerated life tests in an autoclave simulating the typical PWR

operating environment. The Millstone 2 type CEDM was tested before this program for an equivalent

of 83 years of design operation, or 6.31 x 104 linear meters of drive-shaft travel. The System 80 CEDM

was tested for an equivalent of 40 years of design operation, or 1.02 x 104 linear meters of drive shaft

travel.

The accelerated life tests on both types of CEDMs resulted in similar wear patterns, although

the Millstone 2 CEDM exhibited more wear. The majority of material loss occurred at the leading edge

of the latch teeth. Profilometry measurements on the three upper and lower System 80 grippers resulted

in a total weight loss of 1.5 g, which agreed very closely with the 1.7 g obtained from component

weighing. The Millstone 2 CEDM resulted in a weight loss of 4.0 g for one of the grippers. Since three

grippers are normally used, a total weight loss of 12.0 g can be expected, which is 38% less than the 19.6

g loss obtained from weight measurements. (Table 5.4).

Table 5.4 Results of CEDM Wear Tests

Measured Weight

Equiv. Test Measured Weight Change fromn

CEDM Drive Shaft Travel Usage (jrs) Change (g) Profiloneter (g)

Millstone 2 6.3 x 104 83 19.6 12

System 80 J 1.0 x lo, 40 J 1.7 J 1.5

The cobalt wear rate for the Millstone 2 CEDM was calculated to be 1.7 x 104 gmi/m of drive

shaft travel, which is 75% greater than the System 80 rate of 9.7 x 10-5 gm/m (Table 5.5). This wear rate,

when multiplied by actual plant operating data, was significantly less (2.5 gm/year) than the 51 gm/year

previously calculated.

5-9
Table 5.5 Cobalt Release from CEDMs by Plant

Drivesbaft Cobalt Rekase Total Cobalt

Number of TraveVCEDM Total Travel, Rate, Released,

Plant CEDMs meters/CEDM/yr meters/tr g/meter g/Plant-yr

Maine Yankee 65 1.5 x 102(1) 9.8 x 103 1.7 x 104 2

Arkansas

Nudear One #2 81 2.5 x 102(2) 2.1 x 104 1.4 x 104(3) 3

System 80

(Design Basis) 89 2.6 x 102 2.3 x 10' 9.7 x 10' 2

(1) Average Value

(2) Average Value

(3) The anti-ejection gripper was not used for this plant

The report concluded that the wear tests performed on the two representative types of CEDMs

showed in no deleterious results which would affect the stated 40-year lifetime.

5.3.2 Lifetime of PWR Silver-Indium-Cadmium Control Rods, EPRI-4512, March 198626

This report documents work by Westinghouse to determine the lifetime of their Ag-In-Cd control

rods. Many of the results are applicable only to Westinghouse control rods because of the design

differences of the CRA and the upper internal structures. However, one applicable conclusion

established that there is a clad-cracking threshold of 2-2.5 x 10O nvt total fluence threshold for stainless

steel clad rods. Because the control rods are replaced after approximately ten years, aging is not a

concern. However, it is important to track the total rod exposure history, and to visually examine the

rods to ensure there is no fretting or cracking. A severely cracked rod could fail and prevent the CRDM

from either inserting or withdrawing the CRA.

5.3.3 Characterization of the Performance of Major LWR Components, EPRI NP-5001, Jan. 198727

This report presented the results of an EPRI study performed by the S.M. Stoller Corp. to

characterize historical LWR experience on major component failures and modifications. The information

presented included:

a) failure rate,

b) repair time,

c) fraction of the repair time which resulted in forced outage hours,

d) the shadowing factor,

e) the performance losses caused by reduced load operation,

f) failure frequency as a function of unit age and failure type, and

g) overall average performance loss due to the failure.

Babcock & Wilcox control rod drives and holdown springs, and Combustion Engineering control

element drive mechanisms were included in this study.

5-10
Compared to other PWR designs, the B&W units have experienced the most widespread trouble

with the CRDMs, primarily due to electrical shorts originating in the stator winding endturns. B&W

determined that the major contributors to these failures were the following:

1) epoxy breakdown due to its incompatibility with the wire,

2) moisture,

3) bifilar design (side-by-side phasing), and

4) fabrication defects.

Typically, these shorts caused the CRDM power-supply fuses to open, resulting in dropped, inoperable

control rods, and subsequently, a reactor trip. The CRDM was then either repaired, replaced, or

operations resumed at a reduced power. The failure statistics are summarized in Table 5.6.

Table 5.6 Control Rod Drive Failure Statistics

| EFPH per ___________ Frequency | Outage Repair Time |CF Loss (%) % Forced

B&W CRDM Stator

Failures .60 116 116 .79 95

CE Major CRDM

Failures (wlo Palisades) .15 161 161 .28 95

Palisades Major CRDM

Failures 1.69 154 154 2.97 98

B&W Fuel Assembly

Hold-Down Spring .005 2210 .04 100

The Combustion Engineering magnetic jack CEDM, compared to the rack-and-pinion type, has

performed well. The only generic problem reported was the failure of the 15 v dc power supply for the

logic system which controlled rod stepping. This failure generally caused dropped rods, resulting in some

lost effective-full-power-hours. In response to this recurring problem, CE installed redundant power

supplies in some units. There were ten more serious failures due to stuck rods, gripper coil failures,

CEDM ventilation, and CEDM shaft failures resulting in more than 100 lost effective full power hours.

Smaller losses due to clutch coil failures, and the installation of rod block circuitry were also included.

The failures with the rack-and-pinion type CEDM at Palisades were more significant, resulting

in 98% of all lost effective full power hours for CEDM-related problems. The rack-and-pinion CEDM

requires a pressure boundary seal against a rotating drive shaft, and most of the problems resulted from

failures of this seal. However, problems with the clutch, brake assembly, and other components also

contributed to the outage time at Palisades. Failure statistics for CE-type CEDMs are summarized in

Table 5.6.

Failures of the B&W fuel assembly holdown spring were also discussed in this report. These

failures presented a potential loose parts problem which could interfere with control rod insertion, as

discussed in Section 5.2.2. The Inconel X-750 spring material, along with its coarse grain size, were

major failure contributors. The crack initiation was due to high-cycle, low-amplitude fatigue stresses

resulting from flow-induced vibrations. Either the same mechanism, or stress corrosion, caused the

5-11
cracks to propagate. The coarse grain structure of the spring accounted for the increased sensitization

to fatigue and IGSCC cracking.

5.3.4 PWR Pilot Plant Life Extension at Surry Unit 1: Phase 2, EPRI NP-6232, March, 19890

This report describes the scope of work and principal results from Surry 1 plant life extension

study. Although Surry 1 is a Westinghouse-designed PWR, the aging issues addressed for the CRDMs

are applicable to the B&W and CE systems.

Early work in this project highlighted the need to obtain actual data on temperature and

radiation, so that the operating life of the CRDMs and cables in containment could be accurately

assessed. This data is not normally monitored, so a degree of uncertainty is factored into the component

life assessment calculations. This work identified three main issues in CRDM aging:

1) Type 304 stainless steel castings, with a ferrite content of 20%, are vulnerable to thermal

embrittlement; to accurately assess this CRDM surface temperatures must be known.

2) The CRDMs are subject to thermal and low cycle fatigue failure. Fatigue calculations must use

conservative values when postulating the effect of certain transient events. Actual CRD

operating temperatures during these operational transients would verify the accuracy of the

calculations, giving a more accurate assessment of component life.

3) CRDM coils are susceptible to high temperature degradation. Typically, coil temperatures are

calculated from resistance and current data. Actual coil temperature data is essential to

accurately determine the life of the coils. This knowledge would enable a plant to have a

program to replace coils which have experienced high temperature prior to failure and possible

rod drop.

Predicting the effect of temperature and radiation on cabling located in containment is important

to the operation of the CRDM. Organic polymer cable materials are subject to aging due to thermal,

oxidation, radiation, electrical stress, and moisture intrusion leading to insulation degradation. EQ

calculations on the cabling typically assumed conservative temperatures (1250F) and radiation exposure

(37R). Actual exposure data would determine the validity of these assumptions, and allow replacement

of cabling which have exceeded failure thresholds.

5.4 Summary

Age-related operating failures of the CE and B&W control rod drive systems has resulted in

numerous NRC and industry studies. Tables 5.7 and 5.8 summarize the major components analyzed by

these studies for CE and B&W, respectively. Instances of primary coolant leakage were common to both

designs due to pressure housing cracks (CE) and seal degradation (B&W). The majority of the studies

recommended increased leakage monitoring and inservice inspections as a means to identify age

degradation.

The current Standard Review Plant (SRP) and the draft SRP for license renewal was also

reviewed. The latter specifically addresses aging for the control rod drives and related sub-systems. Each

plant will be required to develop an aging management program which identifies and evaluates

component and system aging.

5-12
Table 5.7 Summary of Combustion Engineering Control Element Drive

Aging Analyzed by NRC and Industry

Sub-System Component Failure J Recommended Resolution(s)

1. a. Control Element Assy Control Element Rod Clad Cracks Rod Exposure Tracking, Increased

Visual Inspection

b. Fuel Assembly Guide Guide Tube Through Wall Cracks Guide Tube Sleeving, Increase Rod and

Tubes Guide Tube Inspections

c. Upper Guide Structure _

11. Control Element Drive Primary Coolant Leakage Caused by Improved Leakage Monitoring,

Mechanisms Seal Failures and Housing Cracks, Improved Fabrication Procedures;

Resulting in Corrosion and Gripper Coil Increased Inservice Inspections

Overheating

III. CEDM Control System Lightning Induced Surges Causing Solid Additional Circuit Protection

State Circuitry Failures

Failed 15 v Power Supplies Install Redundant Power Supplies

Slower than Expected Rod Drop Times Review of all FSAR Analysis to Ensure

Continued Compliance

IV. CEA Position Cabling Failure Due to Environmental Monitor Actual Exposure Data to

Indication System Degradation Allow Replacement when Failure

Thresholds are Reached

V. CEDM Cooling System Boric Acid Obstructing Coil Cooling Improved Leakage Monitoring

Passages

5-13
Table 5.8 Summary of Babcock & Wilcox Control Rod Drive

Aging Analyzed by NRC and Industry

| Sub-System Component Failure | Recommended Resolution(s)

1. a. Control Rod Assy. Control Rod Clad Cracking Rod Exposure Tracking; Increased

Visual Inspection

b. Fuel Assembly Guide Loose Parts Lodging in Guide Tubes Core Component Design Modification

Tubes

c. Upper Internal Guide

11. Control Rod Drive Broken Lock Springs Binding CRDM Improved Visual Inspection Techniques

Mechanisms

Primary Coolant Leakage Caused by Improved Gasket Design; Improved

Housing Cracks and Seal Aging Leakage Monitoring; Increased

Inservice Inspection

Roller NutlLeadscrew Wear Periodic Wear Measurements

III. CRDM Control System Lightning Induced Surges Causing Solid Additional Circuit Protection

State Circuitry Failures

IV. Control Rod Assy Position Cabling Failure Due to Environmental Monitor Actual Exposure Data to

Indication System Degradation Allow Replacement when Failure

Thresholds are Reached

V. CRDM Stator Cooling System System Not Evaluated NA

5-14
6. OPERATIONAL EXPERIENCE (1980 - 1990)

6.1 Introduction

The primary objective of this study was to assess the impact of aging on the CRD system. To

accomplish this, a comparison was made between the failures documented by the data bases, and the

NPAR definition of aging failures. As defined in NUREG-1144, the following criteria must be satisfied

for failures to be classified as aging related:

- The failure must be the result of cumulative changes with the passage of time, which if

unchecked, could result in the loss of function and impairment of safety. Failures

causing aging include:

a) natural, internal, chemical, and physical processes which occur during operation,

b) external stresses (radiation, heat, humidity) caused either by storage or operating

environments.

In addition, to eliminate failures due to "infant mortality", the component must have been in service for

at least six months.

A review of operating and failure history for the B&W and CE control rod drive systems suggest

that both designs have experienced age degradation, resulting in significant plant effects. This data, for

1980-1990, was obtained from three national sources of nuclear plant operating experience information1) Nuclear Plant Reliability Data System (NPRDS),

2) Sequence Coding and Search System (SCSS), and

3) Nuclear Plant Experience (NPE).

The NPRDS is a computerized information retrieval system maintained by the Institute of

Nuclear Power Operations (INPO). Performance information provided by this system is based upon

failure event reports of key components submitted by nuclear utilities. NPRDS provides access to

historical engineering failure data reflecting a broad range of operating experience.

The Sequence Coding and Search System (SCSS), also known as the LER data base, provides

summaries for each LER. The entries provide information on the failed components mentioned in each

LER, the root cause of the failure (if known), and the effect upon plant operation.

The Nuclear Plant Experience (NPE) data base is a commercial technical publication which

compiles descriptive summaries of significant events and an indexed reference to all such occurrences.

Much of the information in this data base is obtained from the LERs. However, information is also

contained from utility operating reports and current technical literature.

Each of the three data bases were searched for control rod drive system (CE and B&W) failure

events. Additional queries were made for the control rods, fuel assembly guide tubes, and upper plenum

guide structures. A review of the failures revealed that all of the information identified by the NPE data

base was also included in the SCSS and NPRDS data bases. Because of this duplication, the NPE was

6-1
not included in the review of control rod drive operating experience. To a lesser degree, there was

duplication in the NPRDS and SCSS data. Both contained component failures and degradations which

occurred during reactor operations, and resulted in trips, power reductions or ESF actuation. However,

items discovered during outages that did not result in operational effects, typically were identified only

in the NPRDS data base.

Based upon the results of these searches, the failures of the control rod drive system were divided

into the following seven component and sub-system categories:

1) Cables and Connectors,

2) CRD Control System,

3) CRD Mechanisms,

4) Rod Position Indication,

5) Human Error,

6) Unknown Failure Causes, and

7) Miscellaneous.

These categories encompass four of the five major subsystems for the CE and B&W control rod

drive systems as defined in Section 1.0. There were no failures of the cooling system. The other

categories (cables and connectors, human error, unknown failure cause, and miscellaneous) were chosen

following the review. Control rod and guide tube failures were categorized as miscellaneous, because

there were only four LERs during the ten years. These few failures, however, were very significant since

broken control rods, poison pellet loss, and immovable CEAs resulted.

Figure 6.1 shows the failure frequencies, as reported to the SCSS and NPRDS data bases, for

the B&W and CE CRD system for each category. The actual number of failures reported in each data

base is also presented. For CE, the control system accounted for the majority of the failures while for

B&W, degradation of the control system and the control rod drive mechanisms were the leading causes

of failure.

6.2 Failures Due to Agin2

As shown in Figure 6.2, the percentage of failures attributable to aging degradation varies

between the individual component and sub-system. The control rod drive mechanisms for both CE and

B&W experienced the greatest percentage of aging failures, primarily due to seal degradation and

housing cracks. This high rate of failure corresponds with the high likelihood of aging when exposed to

the operating stressors, defined in Section 4.4, and the research performed on this problem by the NRC

and industry, summarized in Section 5.0.

A large percentage of the failures have been classified as potentially aging, primarily due to the

lack of a complete root cause failure analysis. Electrical connection degradation, failures of individual,

modularized power and control components, and reed switch failures generally did not have failure causes

reported. Small, modularized power supplies, for example, were regularly replaced with spare power

6-2
% of Occurrence

% of Occurrence

Cables t

and C

Connectors S

8320- | 2030 2 _ - _ 40 10 10

20

10 ~~~~10 - -

'ED ~~~~~~~~~~~~0

y CEDM RPI Human Unknown Misc. Cables CRD CRDM RPI Human Unknown Misc. ontrol Error lystem and Control Error

NPRDS M SCSS Connectors System =NPRDS ~SCSS

Combustion Engineering Babcock & Wilcox

Figure 6.1 CRD Sub-systems Database Comparison

100%

02

10

80% - - - 117 ~~~~~~~~~~~~~~~~~~31

se 60% -

6 ~~~~~~~~~~~~~~~~~40

40% 2-

10

20% - -

12~~~~~~~~~~~~~~~~

2

0%

blos CED CEDM Rod Cables CRD CRDM Rod

id Control Position and Control Position

*ctors System indication Connectors System Indication

Aging = Non-Aging 0 Potential Aging Aging M Non-Aging ED Potential Aging

Combustion Engineering Babcock & Wilcox

Figure 6.2 Percentage of Component Failures Attributable to Aging

6-3

Cat

* 5ra

Conni
REDUCED LOAD 42

UNINT AUTO ESF ACT

UNINT AUTO SCRAM 1 3

ATO SCRAM 12

MANUAL SHUTDOWN IS

MANUAL SCRAM 1 AUTO ESF ACT 5

Combustion Engineering

RDCED LOAD 5

AUTO SCRAM a

MANUAL ESF ACT 1 (Tech. Spec. Viol.,

MANUAL SCRAM 3 DrOpped CRA)

SCSS

Babcock & Wilcox

Figure 6.3 Significant Plant Effects Resulting from CRD System Degradation

supplies, with no failure cause determined. The design of these small power supplies led to replacement

rather than repair. Though engineering judgement dictates that many of these failures were due to aging,

no assumptions were made in this study as to possible causes.

6.3 Significant Plant Effects

Component and control rod drive sub-system failures also resulted in significant plant effects, in

addition to affecting the availability and operability of the system (Figure 6.3). These effects

included reduced plant loads, reactor shutdowns, manual and automatic scrams, and actuation of

Engineered Safety Features (ESF).

The operating system review for the CE and B&W control rod drive systems indicated that

failures never resulted in prevention of gravity insertion of the rods upon demand via opening of the

reactor trip switchgear. However, component degradation and failures resulted in increased component

stresses, and unnecessary thermal and pressure cycles which challenged the operation of other safety

systems. Consequently, these occurrences may represent a significant increase in plant risk.

For the ten years evaluated (1980-1990), CE plants reported 75 significant plant effects compared

with 18 for B&W plants; these numbers correspond to an average of 5 significant effects per CE plant,

and 2.5 per B&W plant. When compared to the cumulative number of operating reactor years over this

period (119 reactor years for CE, and 70 reactor years for B&W), failures of the control rod drive system

accounted for less than one significant effect per operating year.

6-4
Most significant plant effects resulted from slipped or dropped control rods. As per Standard

Technical Specifications for both B&W and CE, operators were required to reduce plant power to 70%

and ensure adequate remaining shutdown margin if the time to recover and re-align dropped rods

exceeded one hour. Multiple dropped rods resulted in plant scrams and ESF actuation due to the rapid,

unplanned reactivity transient. The actual root cause for the dropped rods varied from control system

failures to component overheating caused by inadequate maintenance and human error.

Other failures of the CRD system for both designs, while not resulting in significant effects, did

affect plant operations. A loss of one of the rod position indication systems resulted in a loss of

redundancy, and required plant operators to verify rod position every four hours while the system was

inoperable.

Specific system failures, and the effect's, will be discussed for the individual component and subsystems identified in Section 6.1

6.4 Sub-System Failures, Causes and Effects

6.4.1 Cables and Connectors

Though not a specific sub-system, the CRD system utilizes electrical cables and connectors

extensively. Inside the containment, electrical cables provide power to the stators, and transmit rod

position information from the reed switches. The CRD control systems, remotely located outside of

containment, consist primarily of modularized electrical components which also use cables and connectors

extensively.

Both B&W and CE have experienced system degradation due to aging failures of the cables and

connectors (Figure 6.2). Because cables and connectors are replaceable, the root cause for many failures

has not been determined and has been classified as potential aging in this study.

Loose electrical connections were the prime failure cause for the Combustion Engineering CED

system. Primarily affected were the plug in type of modularized electrical components. Broken CEDM

connectors and poor or broken electrical contacts were also noted, as shown on Figure 6.4. RPI cable

degradation, located between containment and the refueling disconnect panel, was a problem at Millstone

2. Brittle and cracked CEDM electrical cabling was found.during an NRC Maintenance Team Inspection

at Palisades; because of this aging failure, the cabling was replaced for all 45 CEDMs.'

Similarly, degraded electrical connections was the primary failure cause for the cables and

connectors used in the Babcock & Wilcox CRD system. Davis Besse reported three occurrences of

excessive electrical noise from faulty penetration modules. Electrical contact and cable degradation was

also noted.

A contributor to cable and connector degradation may be the limited space available on top of

the reactor vessel head. As described in Section 4.3, the power cables must be disconnected from all the

individual drive mechanisms prior to reactor head removal. Due both to the physical size and amount

of cabling, there is no convenient way to safely store these cables and frequently, they are draped over

shielding blocks so as not to interfere with refueling operations. Discussions with several utilities

indicated that regardless of the amount of care exercised, it was not uncommon for the cables and

connectors to be damaged by other equipment and personnel. If the damage was not noted at the time

6-5
Failure Cause

Failure

Loose C

Cable D

Cont

Broken

Cause -_

Connection Loose Connection

Connection

eadtltn Cable Degradation

Megradation 4 _2

tact Failure m w Dirty Contacts f

mConnector m Penetration Module

0% 10% 20% 30% 40% 50% 60% 0 10 20 30 40 50 60 70

% of Occurence

_ NPRDS ,SCSS

Combustion Engineering Babcock & Wilcox

Figure 6.4 Cables and Connectors Failure Cause

tact Faldwe Ef eot

ble CEA 1 Immovable CRA _

lductlon Power Reduction

ped CEA Dropped CRA I

tion Ind. Erroneous Poet. Ind

ctor Trip Reactor Trip

'e Conn. Delayed Start Up

Failure Et

Immova

Power Rt

Drop;

* No Pod

Rea(

Wooeng Phan

0 10 20 30 40 60 80 70 80 0 10 20 30 40 50 80

% of Occurenre S of Occurence

_ NPROS , _C = NP 8 CtOs

Combustion Engineering Babcock & Wilcox

Figure 6.5 Cables and Connectors Failure Effect

6-6
of occurrence, and if it was severe enough, it would be discovered during pre-operational system

inspections, which may delay start-up for repair or replacement.

Similar CE and B&W failure effects were noted for the cable and connector degradation. (Figure

6.5). The primary effect was immovable and dropped rods due to loss of power or poor electrical

contact. Power reduction, start up delay, or reactor trips also occurred, often as a result of the dropped

or immovable rods.

6.4.2 Control Rod Drive Control System Failures

The CE and B&W control systems are primarily responsible for control rod movement in

response to automatic or manual control signals. These systems are designed to provide power to the

CRD mechanism coils, in the proper sequence, to actuate control rod movement in response to these

signals.

The control systems consist primarily of modularized electrical components in centrally located

control cabinets outside the primary containment. Degradation and failure of components in this subsystem accounted for the greatest number of CRD system failures. Aging was responsible for 40% of

the system failures, while an additional 55% of the failures were potentially due to aging (Figure 6.2).

Aging degradation was the cause of 21% of the B&W CRDM system failures, while an additional 67%

of the reported incidents were potentially due to aging. The relatively high incidence of potentially aging

related failures is primarily due to the modularized system design where electronic components are easily

replaced.

Figure 6.6 shows the various components which contributed to the failure of the CE control

system, with power supply failures accounting for the majority of the incidents. Numerous LERs were

generated for the failure of the 15v power supplies which were a replacement design for the originally

installed equipment. Because of the high failure rate of the replacement power supplies, the original

design was re-installed by the utility, and the system design revised to incorporate redundant power

supplies. No specific cause was given for the failure of the replacement power supply. A common

symptom of power supply degradation was output drift, which was typically discovered during post outage

maintenance testing. If the drift was beyond the allowable range, and could not be repaired, the power

supply was generally replaced. Responses to the survey (Section 8.0) indicate that power supplies are

included on the maintenance program and are checked, re-calibrated, and cleaned every outage.

Failures of the current sensor and automatic CEDM timer module (ACTM) also accounted for

significant degradation of the control system. Current sensors detect the applied current to the gripper

coils, and provide feedback signals to the control system. The ACTMs continually monitor and adjust

the voltages applied to the stator coils, based upon the current and wave shape supplied to them.

Failures of these components resulted in dropped or slipped CEAS due to incorrect gripper actuation

and timing. As described in Section 2.5, the original CE power system design did not use the ACTM,

so any such malfunction would not be detected before a CEA slipped. Numerous instances of sluggish

gripper operation were reported before the root cause of the problem was traced to the gripper actuation.

The CEDMCS system was improved to incorporate gripper monitoring to preclude these problems.

Failures of the power switch and optical isolator, (a sub component of the power switch

assembly), also occurred. These failures interfered with, and prevented voltage from being supplied to

the CEDM coils and were normally discovered while obtaining voltage traces prior to control rod

exercising. Aging has been listed as the failure cause for the recent optical isolator failures.

6-7
Failure Caue Failure Cauee

POWER SUPPLY

ACTM

POWER SWITCH MODULE

CURRENT SENSOR

FUSE

MOTION INHIBIT CIRC.

COIL DRIVER ACT.

MISC. COMPONENTS

OPTICAL ISOLATOR

CEA TIMER CARD

OPER. AMPLIFIER

CONTROL MODULE

CEA ENABLE CARD

GRIPPER SENSORS

=14

ii

i '

0 6 10 16 20 26 30 36

% Of Gocrrfc

NtS 00 M cgs

Power Supply

Gate Drive

Fusee

Programmer Board

Motor Programmer

Translstor

Relay

Transfer SwItch

Diode

Logic Failure

0 5 10 16 20 26 30 36

% of Occurence

_ NPRDS i 8CeS

Combustion Engineering Babcock & Wilcox

Figure 6.6 CRD Control System Failure Causes

Failure Effect Failure Effect

lnmvabbe Rod

Delayed Start Up

Dropped Rod

Power Reduction

Reactor Trip

?i "I 0:

IIIg

Immovablo CRA

Delayed Start Up

Dropped CRA

Power Reduction as

Reactor Trip

Misaligned CRA Group

Erroneous Pot. Ind.

No Oper. Signal

C .;.~~~~~~16

4

0 6 t0 16 20 26 30 36

1. of Occurence

NPRDS M 8CS8

No Signlf. Effect

Other r "I°, 'aa, I

0 10 20 30 40 60 80 70

% of Occurence

_ NPRD8 E CSS

Unknown

Combustion Engineering Babcock & Wilcox

Figure 6.7 CRD Control System Failure Effect

6-8
While control system breaker failures were not commonly reported, St Lucie 2 experienced a

significant manual trip due to this failure. Although the root cause was not determined, subsequent

testing indicated that the breaker dropped at a current of 30 amps, 25% less than the designed 40 amps.

As a result of a follow up NRC maintenance team inspection, all of the sub system molded case circuit

breakers (MCCBs) were added to the preventive maintenance program to ensure continued reliability

for the 40 year design life.'4

The B&W control system also experienced age-related failure and degradation, similar to CE,

as shown on Figure 6.6. Failed diodes in the gate drive assembly, primarily due to aging, accounted for

25% of the NPRDS failures for the CRD system. The gate drives control the SCR sequencing which

power the individual stator phases resulting in rod motion.

Motor control circuitry fuse failures were commonly reported in both data bases. Davis-Besse

experienced three failures of transfer switch module fuse failures. These switches transfer a fully

positioned rod group to the hold bus. Utility investigation into the failure cause was inconclusive; there

was no evidence of either excessive heat or current being applied to the fuse, fuse holder, or associated

wiring.

Power supply failures, due to age degradation, were also reported. These failures were commonly

discovered during normal plant preventive maintenance and testing, and replaced. Similar to the CE

experience, replacement power supplies were often installed without determining the exact failure cause.

Motor programmer failures, and control system logic failures due to faulty logic cards were also observed.

Two occurrences of programmer-board failure due to overheating demonstrated the system's

susceptibility to external stresses and inadequate maintenance as discussed in Section 4.1. The root cause

of the overheating was concrete dust contamination from nearby construction. The dust entered the

electrical cabinets, covering the internal component boards. The cabinets were cleaned, the failed

components replaced, and the system returned to normal service.

Dropped or immovable control rod assemblies were the common effects of B&W and CE control

system failures (Figure 6.7). Depending upon the number of assemblies involved, and the time required

to recover and reposition them, these occurrences frequently led to subsequent power reductions, reactor

scrams, and start-up delays.

6.4.3 Control Rod Drive Mechanism Failures

The CE and B&W control rod drive mechanisms are flange mounted on top of the reactor vessel

head. In this location, they are subjected to a variety of stresses (mechanical, electrical and thermal)

resulting in age related degradation and failures. (Figure 6.2). Compared with other sub-systems, the

CRDMs have experienced the greatest percentage of age-related failures. Though the actual

mechanisms differ in the operational design, both have experienced similar seal degradation, and stator

failure due to overheating and moisture intrusion (Figure 6.8).

6.4.3.1 Primary Coolant Leakage

The frequent occurrences of seal failures due to aging is documented by fifty-two NPRDs entries

from CE plants and forty seven from B&W plants. The majority of CE occurrences were confined to

two plants which use the rack and pinion mechanism. This CEDM, as described in Section 2.3.2

6-9
Failure Caude

Failure Cause

Stator Failure

Seal Degradation

Lack of Venting

Mach. Degradation

Bent Extension Shaft

Housing Cracks

Unknown

Debris

0 10 20 30 40 60 60 70

% of Failure

NPRDS EM SC88

Stator Moisture

Gaskets

Stator Short

Vent Valves

Broken Leaf Spring

! f i, , . ... .

_.1,. I I . I . . . . -

1

, i ,3 t . ' . !

Cooling Water Leak I '., .;

Thermocouple l

Leadscrew Lock Nut

. *LLLL,,1 0 10 20 30 40 60 80 To 80

% of Ocourence

_ NPR M aces

Combustion Engineering Babcock & Wilcox

Figure 6.8 Control Rod Drive Mechanisms Failure Cause

utilizes both rotating seals and o-rings. Degradation of the rotating seals has been the prime cause for

the leaks. Since most CE plants use the magnetic jack design, primary coolant leakage has not been a

common occurrence. As described in Section 2.3.1, this CEDM design uses a welded omega-seal at the

reactor vessel flange, as opposed to a gasket seal. It is better in resisting leaks, but removal of the

CEDM is more difficult and time consuming.

Failures of the spiral wound asbestos/stainless steel gaskets between the motor tube and the

flange (Figure 6.9) were the main source of leakage for B&W plants. The root cause for the leaks were

inadequate QC, normal wear and age deterioration. B&W redesigned the flexitallic gasket following a

extensive R&D program which evaluated various materials and configurations. The new spiral wound

graphite/stainless steel gasket is being installed by plants when leakage from the original seals is noted.

Seal degradation resulting in primary coolant leakage, particularly in a high temperature area

such as the top of the reactor head, will cause the boric acid in the coolant to boil and concentrate,

increasing its corrosiveness and acidity. Left uncorrected, the boric acid crystals may accumulate and

block cooling passages, resulting in overheating and winding insulation degradation. Millstone 2 replaced

twenty-three failed gripper coils because boric acid precipitation blocked the cooling passages resulting

in overheating and dropped CEAs.

Boric acid corrosion, as described in Section 4.0, is a significant potential degradation mechanism,

which if uncorrected, could result in control rod drive failure. Utilities were alerted to this problem by

Information Notice 86-108 (Section 5.1). This Notice, and subsequent revisions, documented leakage

from CRDM seals, which was detected during an outage, but judged acceptable for continued operation

6-10
.CRDM Motor

Tube

Bolt

- Problem

Flexitallic

Gasket

Leaking

Split Nut

Ring

Hex He

Bolt

Deposit On i

CRDM/Nozzle

Flange

CRDM Mech

Housing Detail A-A

Deposit On

RV Insulation

U

Deposit On

RV Nozzle

ADeposit On

RV Head

Figure 6.9 Inspection Areas for Boric Acid Deposits (B&W Plants)

6-11
Figure 6.10 B&W CRDM Vent Valve

and left uncorrected. Subsequent inspections, however, revealed severe corrosion on the cooling shroud

and other components.

Coolant leaks from other components were also noted. Combustion Engineering identified

several leaking CEDM pressure housings at Palisades, which when removed and inspected, revealed

circumferential cracks on the seal housings. Similar cracks were found on the remaining two housings

from the same fabrication lot. Additional detailed analysis concluded that the cracks were most likely

due to contaminant induced stress corrosion cracking. However, other housing cracks identified in the

remaining CEDM housings, and at Fort Calhoun, indicate that the root cause of the failures was lack

of, or inadequate, venting, creating conditions conducive to transgranular stress corrosion cracking.

B&W plants also have experienced coolant leakage from the CRD vent valve (Figure 6.10). This

valve is opened to bleed and vent the CRDMs following each refueling. Mechanical wear, resulting in

a decrease in the torque on the jacking screws, and o-ring degradation, were the root causes for the leaks.

B&W has addressed this problem by redesigning the valve. The bleeding process has also been

simplified, decreasing the spill potential, and the jacking screws have been replaced with a hydraulic

sealing mechanism.

The corrosion rate of low alloy and carbon steel caused by boric acid may be greater than

previously estimated, as discussed in Section 5.0. This highlights the need for continued monitoring of

leakage, and its rapid correction. Utilities should consider a scheduled replacement of these components,

as opposed to waiting till leakage appears.

6.43.2 Stator Failures

Stator failures, due to electrical degradation and overheating, were also reported. The stators

were particularly susceptible to moisture failures due to their location on the outside of the CRD

6-12
Failure Effect Failure Etffct

Dropped CEA

Reactor Trip

Primary Coolant Leak

Power Reduction

Immovable CEA

Delayed Start Up

Int. CEOM Corroslon

No Signit. Eflect

I10

imimimi~~4

Dropped CRA

Reactor Trip

Primary Coolant Leak -

p.

r2

Power Reduction I

Immovable CRA

Delayed Start Up

Lose of Cool Water

Spurious Alarms I

~~~~~~~~~~~~~~~~~~~~~~~~~~~iI

' i i i i i

0 10 20 30 40 50 60

% of Occurence

- NP11 m SC"

Unknown _ t

I i, i, i i I

0 10 20 30 40 60 eo 70 80

% of Occurence

_ NPRDS E 8CS8

Combustion Engineering Babcock & Wilcox

Figure 6.11 Control Rod Drive Mechanism Failure Effect

pressure housing, below the vent valves. This location exposes them both to coolant spills during venting

and a humid operating environment.

B&W identified four root causes for the stator failures; epoxy breakdown due to wire

incompatibility, moisture intrusion, bifilar design, and fabrication defects. Moisture intrusion into the

stators was due to degradation of the 0-rings which seal the stator housing against the CRD housing, and

leakage from the stator cooling water system. Typically this problem was discovered during the preoperational 500 v meggering which tests the integrity of the electrical systems. A resistance reading less

than 200 megohms was usually traced to stator moisture. To correct this, the stator is removed and

energized to drive off the entrapped moisture, and retested before being returned to service.

6.43.3 Inoperable CRDMs

Inoperable CRDMs due to broken internals were reported to the NPRDS data base. As

described in Section 5.1, this problem was documented by the NRC in Information Notice 85-38. This

Notice documented two separate occurrences at Davis Besse, four years apart, where internal CRDM

pieces fractured and jammed the leadscrew preventing CRDM movement.

The first event, in 1981, was discovered when a CRA would not withdraw following a reactor trip.

Dis-assembly and inspection revealed that the leaf spring anti-rotational device on the leadscrew nut

assembly had fractured into several fragments. These fragments became stuck between the buffer spring

and the leadscrew, preventing the leadscrew from moving. Since no other similar instances had been

reported, it was concluded that this was an isolated occurrence. However, while performing rod drop

tests in 1985, another CRA failed to insert within the time specified by the Technical Specification.

6-13
Examination revealed that a set-screw fragment from a CRDM handling tool had jammed the leadscrew

again, preventing disengagement of the latching mechanism. A broken leaf-spring on the top of the

leadscrew was also discovered. The brittle, inter-granular spring failure was caused by mechanical

interference between the torque tube cap and the top of the CRDM housing. Improper seating of the

spring caused it to extend further than designed and to strike the torque tube cap when raised.

Inspections of the remaining CRDMs discovered four additional CRDMs with improperly seated leaf

springs.

Maintenance procedures were modified to include visual inspections of the CRDM handling.

tools, before and after use, to detect broken or missing parts. Visual techniques were instituted to verify

the proper positioning of the leaf springs, since prior to these occurrences there was no requirement to

verify this.

The effect of CRDM failure was dependant upon the specific failure type (Figure 6.11). Stator

shorts occurring during operation typically resulted in dropped or slipped rods, and depending upon the

power level at the time of the incident, and the time required to recover the rod, led to power decreases,

and reactors scrams. Failures discovered during normal maintenance testing during outages delayed

reactor startup for repairs.

6.4.4 Rod Position Indication Failures

Both B&W and CE use similar redundant systems for control rod position indication. The

absolute position indication system consists of magnetically operated reed switches for actual rod position

indication, and the relative position indication system provides an indication of the demanded rod

position. Each design has shown susceptibility to age degradation and failures (Figure 6.2).

The location of the position indication systems in the plant exposes each to different stresses

leading to age degradation, as discussed in Section 4.3. The reed switches, which are housed in a tube

alongside the CRD housing, are exposed to similar mechanical, electrical, and environmental stresses as

the CRDM. The relative position indication system, which counts pulses resulting from step demands,

is located outside of containment, in a controlled environment.

As shown in Figure 6.12, reed switch failures were the main cause, of system failure. Typically,

failures occurred during operation, and were not immediately repairable. The failures were electrically

isolated, minimizing the operational effect. The CE design incorporates redundant reed switch position

transmitters and can remain operational with reed switch failures. B&W redesigned the circuitry from

a two-channel design, which became inoperable when a reed switch failed, to a four-channel design which

remains functional with a failure. Since the replacement of the old two-channel RPI circuitry, the B&W

RPI sub-system has performed well. Not all B&W plants have a four-channel design, however.

Control Element Assembly Calculator (CEAC) failure occurrences have led to power reductions

and reactor scrams at CE plants. The safety-related CEACs monitor the CEA position and any

misalignment, and provide signals to the Core Protection Calculations (CPC). Based on this input, the

CPC provides penalty factors to the automatic protection and control system ensuring safe reactor

operation. Failed reed switches, which were not immediately recognized, have generated overlyconservative penalty factors, and subsequent power reduction and scrams.

Plant computer malfunctions, due to software problems and electrical component failures typically

led to relative position indication failures for CE. Only two instances of RPI failures at B&W plants

6-14
Failure Cause

Reed Switch

Failure Cause

RSPT

Reed Switch

Limit Switch

Power Supplies

Plant Computer

Control Circuits

CEAC

Power Supplies

Relays

RPI Motor

J~h~Ih 4.

t l0 20 40 e0

% Of Occurence

= NPRO8 8C8s

B0 100 0 10 20 30 40 60 60 70 80

% of Occurence

_ NPRDO M 8C88

Combustion Engineering Babcock & Wilcox

Figure 6.12 Rod Position Indication Failure Cause

Failure Effect

-_ _ 1 '

I Spurious Post Ind

Failure Effect

Erroneous Post. Ind.

No Position Ind.

Delayed 8tart Up

Inoperable CEAC

Power Reduction

II

it

5

a

No. Post. Indication

Delayed Start Up

Reactor Trip

Spurious RPI Alarms

I

i i

I

i

I:

a

0 20 40 so 80 100

% of Occurence

NPROS S 8C88

0% 10% 20% 80% 40%

% Of Occurrence

_ NPROG S 8C88

60% 80%

Combustion Engineering Babcock & Wilcox

Figure 6.13 Rod Position Indication Failure Effect

6-15
Dropped CEA

72%

Spurious

Position Ind.

3%

ower Reduction

22%

Immovable CEA

3%

SCSS

Figure 6.14 CE Control Element Drive System Unknown Failures Effect

were noted. At Crystal River, one RPI failure was due to a relay failure, and the other was due to low

voltage.

The common effects of these failures were erroneous or loss of position indication signals,

resulting in power reductions or a start-up delay to permit maintenance (Figure 6.13). Plant Technical

Specifications require redundant position indication signals. When one system fails, power is reduced,

and the operators verify the proper position every four hours until the redundancy is available.31,32 If

discovered before operation, the system will be repaired before start-up.

6.4.5 Human Error

Though not an effect of component aging, the CE Control Element Drive System demonstrated

a susceptibility to human error as well as component failures due to aging and environmental

degradation. Maintenance errors and refueling operations were the prime causes of these incidents, and

most of these events resulted in dropped rods and subsequent power reductions. Procedural errors

resulting in violations to the plant Technical Specifications were also noted. These occurrences can be

eliminated through increased cognizance and understanding of the requirements contained in the

individual Technical Specification.

As a result of these human errors, system component damage occurred. An incorrect

maintenance procedure at a CE unit permitted high voltage to be applied to the gripper coils causing

failure. Several refueling incidents, resulting in bent extension shafts, were noted when the upper

internals were being reinserted into the reactor vessel following refueling. These resulted in plant start

up delays while the extension shafts were replaced. Section 6.4.7 discusses this particular degradation

mechanism, and highlights the potential damage that could be caused to the extension shafts during

refueling. The B&W CRDM is removed as a unit with the reactor head, precluding similar damage.

6-16
6.4.6 Unknown Failure Cause

As documented in Appendix B, a significant number of LERs documented CE Control Element

Drive System failures due to unknown causes. As Figure 6.14 shows, these failures resulted in significant

plant effects including dropped or immovable CEAs, spurious position indication, and reactor power

reductions. Upon review of these failures, it was noted that many were similar to occurrences with

known causes. However, as discussed in Section 6.1, no conclusions were made regarding the causes of

these failures. Numerous incidents of dropped and slipped rods affected CE plants early in the 1980s.

This problem has been ascribed to sluggish gripper operation with the older CEA Timer, which did not

provide for feedback to confirm gripper operation. This allowed the weight of the CEA to be transferred

to another gripper which was not correctly engaged, causing the rods to slip.

The significant plant effects which occurred as a result of these failures highlight the need for a

thorough root cause assessment to preclude similar events from occurring.

6.4.7 Miscellaneous Failures

The operating database review identified several other significant system failures. These events

were not categorized into particular sub-system failures, but did pose operational problems for the

system.

While implementing a revision to the rod drop procedure, Arkansas-2 discovered slower-thananticipated drop times. Before this change, power was interrupted to each CEDM individually, which

historically produced acceptable results. The new procedure was revised to reflect actual system

operation, and simultaneously tripped all the CEAs by interrupting power to the reactor trip breakers.

This led to 'a slower dissipation of the stored energy in the holding coils, increasing the total drop time.

In response to this event, the NRC issued Information Notice No. 88-47, documenting the

occurrence and the potential consequences to plant safety. The increase in drop time necessitated a

review of all of the FSAR accident analysis to ensure continued conformance to operating restrictions.

Calvert Cliffs reported a rod drop incident caused by a shorted coil power programmer. The root

cause of this incident was not the CPP failure, but rather water seepage from an overflowing control

room toilet. The water entered the cable spreading room, resulting in several shorted electrical

components.

The CEA which failed at Maine Yankee was a five finger design, with the center rod containing

B4C in the lower portion. Due to the radiation induced swelling of the B4C, the outer rods did not

contain full length B4C. The root cause of the center rod failure was B4C swelling due to the total fastfluence exposure. The swelling caused the clad stress to increase eventually leading to propagation of

a crack around the circumference of the rod. This resulted in a broken rod, and the poison pellets fell

into the fuel assembly guide tube, preventing CEA insertion.

The design life of the CEA is specified as 10 years (13-14 EFPY). The failed CEA exceeded

these limits, and was judged acceptable due to its specific exposure history. This failure highlights the

need for tracking exposure of the rods, as recommended in Section 5.0, to prevent this failure from

occurring.

6-17
FA Holddown Spring

Unknown

10%

Thermal

Shield Bolts Reactor Coolant

30% Pumps

10%

SCSS

Figure 6.15 B&W Control Rod Drive System Source of Loose Parts

Arkansas experienced a similar stuck CEA, due to foreign material wedging between the control

element rod and the guide tube. Loose parts in the core are a potential significant threat to CEA

movement, and care must be taken by operating plants to monitor for loose parts.

Babcock & Wilcox reactors reported several instances of reactor internals degradation (Figure

6.15). The sources of debris included broken fuel assembly hold down springs, thermal shield bolts, and

reactor coolant pump fragments. The debris and loose parts from these failures could have lodged in

the fuel assembly guide tubes or CRDM internals, preventing CRA insertion.

Debris from a damaged reactor coolant pump at Oconee-3 was carried by the coolant to the core

where it lodged in the fuel assemblies. Although these fragments were confined mostly to the lower end

fittings and the first spacer grid, they could have interfered with CRA insertion.

Broken thermal shield support bolts, primarily due to intergranular stress corrosion cracking,

posed similar problems at Oconee-1, Oconee-2, Rancho Seco, and Crystal River. Following detailed

metallographic studies, the lower thermal shield was redesigned, and the bolt material changed from

A286 to Inconel X750. Locking clips were also attached to the bolts to retain them in the event of

failure.

Broken fuel assembly hold down springs were identified at Crystal River, Davis Besse, Oconee-1,

and Oconee-3. These plants used the Mark-B fuel assembly design, which consists of one large, helical

coil spring, positioned on top of the upper end fitting, and held in place by a spring retainer. These

cracks were discovered during regular fuel assembly inspections. Low stress, high-cycle fatigue and stress

corrosion cracking were the root causes of the failure. Analysis by B&W concluded that these spring

failures would not result in a displaced fracture. However, due to the close proximity of the spring to

the control rods and the guide tubes, any piece which may have broken off, could easily have lodged in

the guide tube and interfered with CRA movement or caused fretting of the control rod. Instances of

broken hold-down springs continue to be reported, as discussed in Section 5.2. Sixty-four failures were

identified from 1980 through 1988.

6-18
6.5 Summary

The detailed operating experience review discussed in this Section identified the primary aging

failure causes, and resultant effects, for both CE and B&W control rod drive systems. These results

agree closely to that which would be anticipated when the system is exposed to the operating and

environmental stressors discussed in Section 4.0. A close correlation was also observed between these

findings and the focus of the NRC and industrial research summarized in Section 5.0.

Significant plant effects resulted from the aging degradation and failure of the control rod drive

subsystems. Dropped or slipped rods were the most common effect of power and control subsystem

failures. These resulted in power reductions or shutdowns while the rods were re-aligned. System

failures which resulted in multiple rod drops led to reactor shutdowns and ESF actuation on several

occasions.

Primary coolant leakage, from motor tube housing cracks and seal degradation, was another

common failure. The boric acid in the coolant severely corroded nearby components and blocked cooling

air passages, resulting in coil failures.

The review of both the CE and B&W operating experience indicated that the control rod drive

system failure occurrences have not prevented gravity insertion of control rods. Component failures and

degradation did result in increased component stresses and unnecessary thermal and pressure cycles,

which challenged the operation of other plant safety systems. This review also highlighted the importance

of performing root failure cause analysis. Numerous failures were reported to the data bases with no

failure cause given. Multiple failures of the same components occurred before the cause of failure was

determined. These failures, and the resulting plant operating effects may have been avoided if the failure

cause was determined initially.

6-19
7. EFFECT OF COMPONENT FAILURES ON CRD SYSTEM

The control rod drive systems for both CE and B&W plants consist of the five major subsystems

described in Sections 2 and 3. Component failures in any one of these may result in degradation and

operational failure in any of the other subsystems. To determine the effect of these failures, an individual

Failure Mode and Effect Analysis (FMEA) was performed for each subsystem. Each FMEA included

the following items:

a) Failure Mode: The basic manner(s) which a component may fail or cease to perform as

designed. Failure modes for common components (e.g., electrical components, switches,

motors) were obtained from industry reliability standards. For other CRD specific

components (e.g., control rods, CRDM) engineering judgement was used to define the

failure modes."334

b) Failure Cause: The particular type of degradation mechanisms which may cause

component failure. These stressors (operating and environmental) were discussed in

Section 4.0.

c) Failure Effect: The effect on the CRD system caused by the component failure.

d) Aging: A subjective, engineering assessment indicating whether the failure cause is

directly attributable or potentially susceptible to time-related effects or environmental

degradation.

e) Probability of Occurrence: A subjective assessment of the likelihood of occurrence for

each failure cause. The operational experience described in Sections 5.0 and 6.0 was

used to determine the probability, and was assigned either a high, medium, or low rating.

7.1 Control Rod Assemblies (B&W and CE)

Table 7.1 shows the FMEA for the Babcock & Wilcox and Combustion Engineering control rod

assemblies including the fuel assembly guide tubes and spider assemblies. Though the two designs differ

in the number of control rods, the failure modes and effects for the two designs are similar.

A failed CRA does not necessarily compromise safety; however, it does affect plant operation.

An inoperable CRA requires that the plant operators ensure sufficient shutdown margin, or repair the

problem within the time limitations defined by the Technical Specifications. If a dropped rod cannot be

re-aligned with its group within one hour, then a power reduction or plant shutdown is required.

The CE control element rods have a larger diameter and are stiffer than the B&W control rods,

which may increase the likelihood of wear between the control rods and the guide tubes due to flowinduced vibration.

Poison swelling is another possible control rod failure cause. This could cause clad cracking, and

coolant intrusion into the rod resulting in poison depletion (washout). In the extreme case of a total

circumferential-through-wall crack, the lower portion of the rod may break off, allowing the poison pellets

to fall out. This would result in decreased shutdown capability and may cause the rods to jam if the

debris fell into the guide tubes.

7-1
Table 7.1 FMEA - Babcock & Wilcox and Combustion Engineering Control Rod Assemblies

Failure Failure Failure

Component Mode Causes Effect A g Probability Occu ce

Control Rod Circumferential Through-Wall High Rod Pressure Loss of Lower Control Yes Low (B&W), Medium (CE)

Crack IGSCC/IASCC Element Rod End Cap Yes Medium (B&W), Medium

(CE)

Poison Swelling Poison Pellet Loss Yes Low (B&W), Medium (CE)

Flow Induced Vibration Immovable CRA Yes Low (B&W), Medium (CE)

Excessive Bow Fabrication Error Immovable CRA No Low (B&W, CE)

Irradiation Induced Bow Increased Drag Load Yes Low (B&W, CE)

Clad Cracking Poison Swelling Poison Washout Yes Low (B&W), Medium (CE)

Flow Induced Vibration Decreased Clad Strength Yes Medium (B&W), High (CE)

IGSCCAIASCC Yes Medium (B&W), Medium

.__ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _(C E )

Fuel Assembly Guide Tube Bowed or Twisted Guide Tube Fabrication Induced Increased Drag Load No Low (B&W, CE)

Irradiation Induced Fuel Mechanical Wear Yes Low (B&W, CE)

Assembly Bow

Guide Tube Cracking Flow Induced Vibration Immovable CRA Yes Low (B&W), High (CE)

Increased Drag Load

Debris Induced Fretting Decreased Structural Yes Low (B&W, CE)

Strength

Spider Assembly Broken Arm IGSCC Dropped Rod Yes Low (B&W, CE)

Vibration Immovable Rod Yes Low (B&W, CE)

Fabrication No Low (B&W, CE)

Inability to Couple/Uncouple Debris/Crud Buildup on Male Increased Force Required to No Low (B&W, CE)

From Leadscrew Bayonet Coupling Couple/Uncouple CRA
7.2 Control Rod Drive Mechanisms

7.2.1 B&W Control Rod Drive Mechanisms 35

The CRDMs used by B&W consist of a roller nut assembly and a non-rotating leadscrew

contained in the motor tube. The CRDMs are flange mounted on top of the reactor vessel head. Table

7.2 shows the FMEA for the major CRDM components.

The main effects of CRDM failure are primary coolant leakage and dropped/slipped CRAs. As

with CRA failures, a CRDM failure resulting in a dropped or immovable CRA does not necessarily

compromise plant safety, but does affect normal plant operations. Large leaks may represent a small

break LOCA and potentially cause the failure and degradation of other components and systems.

Leaking vent valves and failed spiral wound flexitallic gaskets are typical causes for primary coolant

leakage.

Many CRDM failure mechanisms are age related. As reflected by the low probability of

occurrence, aging degradation of other components has not been widely evident, with the exception of

primary coolant leakage.

7.2.2 Combustion Engineering Control Element Drive Mechanisms

The CE control element drive mechanisms are flange mounted on top of the reactor vessel head,

and consist of a series of magnetically operated latches which engage the CEA extension shaft to effect

CEA holding and movement. Table 7.3 presents the FMEA for the magnetic jack type of CEDM. The

rack and pinion type of CEDM, with the exception of latch and coil failures, has similar failure modes

and effects. As described in Section 6.0, the rack and pinion CEDMs have been primarily susceptible

to rotating seal failures resulting in primary coolant leakage.

Several recent core designs used by CE incorporate a 12 element CEA, which control reactivity

for four fuel assemblies. Due to the effect upon more than one fuel assembly, failure of these CEDMs

may have a greater operational and safety impact.

Primary coolant leakage and dropped CEAs were common effects of system failure. Coolant

leakage was also a potential failure cause for several other subcomponents, particularly the stators. All

of the identified system failure modes, with the exception of those caused by fabrication, human, or

maintenance errors, are aging related. Based upon recent Combustion Engineering operating experience,

these failure modes have a high probability of occurrence.

7.3 Control Rod Drive Control Systems

7.3.1 Babcock & Wilcox CRDM Control System

Table 7.4 provides an FMEA for the major components of the B&W CRDM Control System.

These components include the system logic necessary to control rod grouping, motion and position, and

the power supplies to drive the CRDMs.

7-3
Table 7.2 FMEA - Babcock & Wilcox Control Rod Drive Mechanisms

. 1 Probabiliy

Component Failure Mode Failure Causes Failure Effect Aging of

_______________________ I _______________________ I O ccurrence

Motor Tube Cracked Housing SCC Primary Coolant Yes Low _______________________ ~~Leakage_ _ _ _ _ _ _ _ _ _ _

Weld Cracks Vibration Primary Coolant Yes Low

SCC Leakage Yes Low

Fabrication No Low

Vent Valve O-Ring Embrittlement Thermal Embrittlement Primary Coolant Yes Medium

Assembly Radiation Embrittlement Leakage Yes Medium

Mechanical Wear Operation Induced Primary Coolant Yes Medium

Human Error Leakage No Medium

Torque Taker Mechanical Wear Fatigue Immovable Yes Low

Assembly Debris CRA Yes Low

Maintenance No Low

Human Error No Low

Leadscrew Roller Nut Failure Human Error Immovable No Medium

Mechanical Wear CRA Yes Low

Mechanical Wear Vibration Between Roller Slipped Yes Low

Nuts/Leadscrew CRA

Bowed Leadscrew Maintenance Inoperable No Low

Fabrication Induced CRDM No Low

l_________________ Binding Debris/Crud Inoperable CRDM Yes Low

Rotor Assembly Segment Arm Springs Radiation Induced Immovable Yes Low

Fail Due to Loss of Mechanical Failure CRA Yes Low

l__________________ Compression Fatigue Yes Low

Segment Arm Assembly Fatigue Immovable Yes Low

Pivot Pin Failure Mechanical Wear CRA Yes Low

SCC Yes Low

Roller Nut' Failure Loss of Coolant Immovable CRA Yes Low

Mechanical Wear Dropped/Slipped Yes Low

Debris/Crud CRA Yes Medium

Vibration Yes Low

Bearing Failure Mechanical Wear Yes Low

Vibration Immovable Yes Low

Debris/Crud CRA Yes Medium

'Thermal Overheating Yes Low

Thermal Barrier Failure SCC Thermal Degradation Yes Low

Vibration of CRDM Internals Yes Low

Crud/Debris Yes Medium

Flexitallic Gaskets Mechanical Wear Embrittlement Primary Yes High

Thermal Overheating Wear Coolant Yes High

Material Degradation Corrosion Leakage Yes High

Maintenance No High

Hold Down Bolts Loss of Torque Mechanical Wear Primary Yes Low

Galling Coolant Yes Medium

Vibration Leakage Yes Medium

Human Error No Low

74
Table 7.2 FMEA - Babcock & Wilcox Control Rod Drive Mechanisms (Cont'd.)

Probability

Component Failure Mode Failure Causes Failure Effect Aging of

Occurrence

Buffer Spring Loss of Spring Force SCC High Kinetic Yes Low

Mechanical Wear Energy Force Yes Low

Fatigue At End of Scram Yes Low

CRDM Cooling Inadequate Cooling Pump Failure Stator Coil Failure Yes Low

System Power Loss Yes Low

Heat Exchanger Failure Dropped/Slipped Yes Low

Degraded Water CRA Yes Low

Chemistry

Control System Failure Inoperable CRDM Yes Low

Maintenance Error No Medium

Boric Acid Buildup Yes High

Stator Coils Thermal Overheating Coolant Pump Failure Dropped CRA Yes Low

Loss of Forced Air Yes Low

Cooling Immovable CRA Low

Blocked Cooling Passages Yes

Electrical Short Corrosion Immovable CRA Yes Medium

Electrical Insulation Dropped CRA Yes Medium

Degradation

Fabrication Error Erroneous Rod Yes Medium

Moisture Position Indication Yes Medium

Mechanical Wear Yes Low

Loss of Power Connector Corrosion Dropped CRA Yes High

Human Error No Medium

Mechanical Wear Immovable CRA Yes High

Fatigue Yes High

Control Systems Failure Yes High

7-5
Table 7.3 FMEA - Combustion Engineering Control Element Drive Mechanisms

Probabily

Comsponent Failure Mode Failure Causes Failure Ellect 1 Aging of

I___ ____ ____ ____ _ _Occurrence

Motor Housing Circumferential SCC Primary Yes Medium

Cracks Vibration Coolant Yes Medium

Weld Cracks Fabrication Error Leakage No Medium

Upper Pressure 0-Ring Thermal Yes Low

Housing Assy. with Embrittlement Embrittlement

Vent Valve Housing Cracks SCC Primary Yes Low

Weld Cracks Radiation Coolant Yes Medium

Mechanical Wear Embrittlement Leakage

Human Error No Medium

Fabrication Error No Low

Galling Yes Medium

Coil Stack Assembly Overheating Blocked Cooling Dropped/Slipped Yes High

Passages CEA

CEDM Cooling Inoperable CEA Yes Low

System Failure

Electrical Short Corrosion Dropped/Slipped Yes Medium

Electrical Insulation CEA Yes Medium

Degradation

Fabrication Error No Low

Moisture Inoperable CEA Yes Medium

Mechanical Wear Yes Medium

Loss of Electrical Connector Corrosion Yes Medium

Power Cable Degradation Dropped/Slipped Yes Medium

Mechanical Wear CEA Yes Medium

Fatigue Inoperable CEA Yes Medium

Power Supply Yes High

Failure

Human Error No Low

Extension Shaft Mechanical Wear Mechanical Slipped/Dropped Yes Low

Vibration Between CEA

Gripper and Shaft

Bowed Leadscrew Maintenance Inoperable No Low

Fabrication CEA No Low

Binding Debris/Crud Slipped CEA Yes Medium

Latches, Links, Pins Latches Fail to Mechanical Wear Slipped/Dropped Yes Low

Engage Fatigue CEA Yes Low

Crud/Debris Inoperable Yes Low

Coil Failure CEDM Yes Medium

Latches Fail to Coil Failure Inoperable Yes Low

Disengage Crud/Debris CEDM Yes Low

No Insertion

Latches Move Out Coil Failure Slipped/Dropped Yes Medium

of Sequence Control System CEA Yes High

Failure

Crud Debris Yes Low

7-6
Table 7.3 FMEA - Combustion Engineering Control Element Drive Mechanisms (Cont'd.)

Probability

Component Failure Mode Failure C Failure Elfectgig of

Occurrnce

Latches, Links, Pins Transfer Latches Coil Failure Slipped/Dropped Yes Medium

(Cont'd) Fail to Operate Control System CEA Yes High

Failure

Crud/Debris Yes Low

Broken Latches Yes Low

CEDM Cooling Inadequate Cooling Fan Failure Coil Failure Yes Low

System Power Loss Dropped/lipped Yes Low

Heat Exchanger CEA Yes Low

Failure

Control System Inoperable Yes Low

Failure CEDM

Shroud Failure Yes Low

Boric Acid Buildup Yes High

7-7
Table 7.4 FMEA - Babcock & Wilcox CRD Control System

Component | Failure Mode | Failure Causes | Failure Effect I A I Probability of Occurence

Transformer Fails to Operate Due to: Electrical Insulation No power supplied to Yes Low

Open Circuit Degradation CRD system resulting in

rod drop

Shorted Turns Thermal Overheating Yes Low

Redundancy Loss

Electrical Shorts Yes Low

Component damage due

Electrical Drift to high voltage supply Yes Low

Power Supplies Power Supply: Electronic Component Loss of Redundancy Yes Low

(24, +1-15, 5 VDC) Fails to Operate Overheating

Dropped CRAs.|

Output Drift Electrical Power Surge Yes Low

Immovable CRA

Electrical Drift Yes High

Fuse Failure Yes Low

Programmer Motor Fails to: Electrical Power Supply Immovable CRA Yes Medium

Operate Upon Demand Surge

Dropped CRA

Fails to Stop on Demand Mechanical Binding Yes Low

Incorrect

Operates at Wrong Speed Thermal Overheating insertion/withdrawal rate Yes Low

resulting in power

Loss of Power excursion and reactor trip Yes Low

Incorrect direction of

operation

Stator damage due to

incorrect sequencing of

SCRs

o0
Table 7.4 FMEA - Babcock & Wilcox CRD Control System (Cont'd.)

Component | Failure Mode Failure Causes Failure Effect | Aging Probability of Occurrence

Encoding Lamp No Power Supplied to Lamps Loss of Primary/Secondary Immovable CRA Yes Low

Lamp Power Supplies

Decreased Lamp Output Dropped/Slipped CRA

Undercurrent Relay Failure due to incorrect SCR Yes Low

Fails to Operate activation

Thermal Overheating Yes Low

No power supplied to

stator phases

Redundancy Loss

Photo-Detector Detector Fails: Loss of Electrical Power Phase continuously Yes Low

Open energized resulting in

Electrical Power Surge immovable CRA Yes Low

Closed

Degraded Electrical Phase not energized Yes Low

Connections resulting in dropped or

immovable CRA

Cable Degradation Yes LOw

Stator damage due to

continous power supplied

Gate Driver Fails to Operate on Demand Electronic Component Failure to gate on SCR Yes Medium

Overheating band resulting in no

CRA motion

Electrical Noise Yes Medium

Incorrect sequencing

Electrical Power Surge resulting in immovable, Yes Medium

l _________________________ ._______________ .______ slipped C R A

Transfer Relays Fails: Electrical Contact Unable to transfer group Yes Low

Open Degradation or single CRA to/from

auxilliary regulating

Closed Power Supply Failure power supply Yes Low

Intermittent Operation Contact Mechanical Wear Dropped rod/group Yes Low

during transfer operation

between power supplies
Two transformers convert three-phase 480-volt power to six-phase 120-volts for the CRDMs, and

provide redundant sources of power. Failure of one transformer will result in a loss of redundancy but

should have no additional effect on system operation. This subsystem also contains several additional

power supplies which provide power to specific components. Redundant supplies are typically provided

for these components to allow continued operation when one fails.

Two programmer motors are used to regulate CRA insertion speed. One motor provides for

normal insertion, and the other provides for the slow (or jogging) insertion of the CRAs. These motors

control the rotation of the optical disc, which via optical detectors, actuate the SCRs in a controlled

pattern resulting in CRA insertion, withdrawal, or holding. Failures of these components may result in

immovable CRDMs or dropped CRAs. Stator damage may also occur due to incorrect energization of

the stator phases.

A common failure cause for all of the system components was overheating. The electronic

components are typically mounted in electrical cabinets which are forced air cooled. This system is

degraded by dirty or clogged filters or power loss, resulting in its inability to dissipate the heat generated.

If not corrected quickly, component failure may occur.

The operating experience for the system has been good, as reflected by the low failure

probability.

7.3.2 Combustion Engineering Control Rod Drive Control System

The CE CRD Control System provides the logic and controls which provide power to the

CEDMs to effect individual rod and rod group motion and positioning. The fail-safe system design

allows for CEA insertion in the event of a power loss to the CEDMs. Degradation of the control

components typically lead to slipped, dropped, or immovable CEAs. The FMEA for this system is

provided in Table 7.5.

The control system logic synchronizes and controls CEA motion and positioning by supplying

power to the CEDM coils via phase angle firing of SCRs. Degradation and failure of these components

are primarily due to electrical stresses. The effect of these failures may include droppedCEAs due to

improper gripper operation, and coil damage due to improper energization.

The Automatic CEDM Timer Module (ACTM) is an upgrade replacement for the CEA Timer

Module in the CEDMCS. The ACTM is a microprocessor based controller which controls and monitors

the currents powering each of the coils via non-intrusive current sensors. The current waveforms provide

a direct indication of the mechanical movement of the magnetic jack. The ACTM utilizes closed loop

control to ensure proper stepping sequences, and has successfully eliminated sluggish gripper occurrences.

The ACTM also monitors for high coil currents and inadequate gripper holding currents and takes

corrective action. The step pulses generated by the ACTM (for counting by the Plant Monitoring

Computer) are based on actual successful steps and not on motion demand signals. The step count may

be more accurate with ACTMs. ACTM failure does not typically lead to dropped or immovable CEAs.

Failed subgroup, pulse count, undervoltage, and auxiliary relays may result in improper system

operation, including erroneous pulse count position indication and alarms. Relay failures may also result

in loss of control/indication signals from/to the Reactor Regulating, Turbine Control, Reactor Power

Cutback, and Plant/Reactor Protection Systems.

7-10
Table 7.5 FMEA - Combustion Engineering CRD Control System

Component | Failure Mode Failure Causes Failure Effect Aging Probability of Occurnce

Power Supplies Fails to Operate Fuse Failure Dropped/slipped CEA Yes Medium

when transferred to a hold

Output Drift Power Surge bus Yes Medium

Electrical Noise Upper Gripper Coil Yes Medium

DamageThermal Overheating Yes Medium

Common Logic Relay Fails: Thermal Overheating Loss of signals to/from Yes Low

Interface Open external reactor control

Electrical Power Surge system Yes Low

Closed

Electrical Noise Yes Low

Intermittent Operation

Degraded Electrical Yes Low

Connection

Logic Housing Fails to Operate Loss of Power Out of synchronous CEA, Yes Medium

subgroup and group motion

Control System (ACTM) Intermittent Operation Power Spike Yes Low

Electrical Noise Yes Low

Degraded Electrical Yes Low

Connection

Power Switch Assemblies Fails to Operate Mechanical Fatigue Improper voltage to Yes Medium

CEDM coils

Fails to De-energize Power Loss Yes Medium

Dropped/slipped CRA

Sporadic Operation Thermal Overheating during transfer group to Yes Medium

hold bus

Electrical Noise Yes Medium
Table 7.5 FMEA - Combustion Engineering CRD Control System (Cont'd.)

Component Failure Mode | Failure Causes | Failure Effect | Aging [ Probability of Occumne

Relays (Channel, Subgroup, Fails: Mechanical Fatigue Loss of signal to pulse Yes Low

Pulse Count, Undervoltage, Open count relay interface

and Auxiliary) Mechanical Wear resulting in no PI Yes Low

Closed

Power Loss Redundancy Loss Yes Low

Erratic Operation

Power Spike Erroneous PI Alarms Yes Low

Thermal Overheating No Undervoltage Yes Low

Indication

Electrical Noise Yes Low

Loss of control/indication

signals from/to: Reactor

Regulating System, Turbine

Control System, Reactor

Power Cutback System,

Plant/Reactor Protection

Systems

Supervisory Panels Fails to: Power Failure No system status display in Yes Low

Operate Control Room

Power Spike Yes Low

Operates Intermittently Loss of Redundancy

Thermal Overheating Yes Low
7.4 Rod Position Indication

7.4.1 B&W Rod Position Indication

Babcock and Wilcox plants use two separate systems to provide continuous redundant rod

position indication signals. Tables 7.6 and 7.7 show the FMEA for the Absolute Position Indication

System and the Relative Position Indication Systems, respectively. The common failure effects for each

system are the loss of position indication, and erroneous or spurious rod position indication signals.

The Absolute Position Indication System (API) consists primarily of reed switches mounted in

a housing adjacent to the CRDM pressure tube. In this location, the same stresses which act upon the

CRDM housing also affect the API. Mechanical stresses (fatigue, vibration, wear), electrical stresses

(degraded connections), and environmental (radiation, temperature, and humidity) are the common

failure causes. API component failures result in loss of individual rod or group position indication,

spurious alarms (asymmetric rods, group average, no sequence enable), and no feed and bleed permit.

The relative position indication system counts the pulses supplied to the CRD stator which

indicates the demanded rod position. Failure of the system to count these pulses results in the stepping

motor not operating, resulting in erroneous relative rod and group position indications. When this

occurs, the reset relay must be closed to align the reset pulser with the stepping motor. The pulser then

supplies dc pulses to the stepping motor, repositioning the RPI output to correspond with actual rod

position.

The main failure causes which lead to system degradation are typically electrical in nature

(component overheating, degraded electrical connections). The primary effect of RPI failure is a

sequence fault caused by incorrect RPI input to the sequence monitor, which checks for greater than 25%

overlap at discrete intervals, and results in a sequence operation fault if the rods are out of sequence.

7.4.2 CE Rod Position Indication

Combustion Engineering also uses two redundant methods to provide rod and group CEA

position indications. The reed switch position indication system uses a series of magnetically actuated

reed switches to provide actual rod position indication. The pulse counting position indication system

monitors the pulses generated, in older plants, upon motion demand signals, or in plants with the ACTM

upgrade, upon successful motion completion. FMEA's for both of these systems are provided in Tables

7.8 and 7.9.

Due to the location adjacent to the CEDM, reed switch position transmitters are susceptible to

mechanical, electrical, and environmental stresses. The predominant effect of component failure is the

loss of, or erroneous input to the rod position indication system. In plants with a Core Protection

Calculator System, this may result in spurious automatic reactor trips. These failures may also result in

erroneous penalty factors generated by the CPC.

Failure of the pulse counting position indication system is typically due to failure of the

monitoring system which counts the pulses provided to it. The effect of RPI failure depends upon

whether the failure occurs during times of rod motion. Failures during rod motion result in a spurious

CEA deviation or out-of-sequence alarm. The primary effect of failures occurring during no rod motion

is a spurious CEA deviation.

7-13
Table 7.6 FMEA - Babcock & Wilcox Absolute Position Indication System

Probability |

.Failure Failure Failure of

ConIponent Mode Causes Effect Aging Occurrence |

Reed Switches Switch Fails: No Rod Position Indication

Open Boric Acid Corrosion Spurious Rod Position Yes Low

Closed Fatigue Indication Yes Low

Partially Open Upon Signal Mechanical Wear Erratic Rod Position Yes Low

Partially Closed Upon Vibration Indication Yes Low

Signal Thermal Overheating Spurious High/Low Yes Low

Radiation Degradation Asymmetric Rod Yes Low

Moisture Intrusion Indication/Alarm Yes Low

Electrical Connection Spurious High/Low Group Yes Medium

Degradation Average Alarm/Indication

Loss of Redundancy

Resistors Out of Tolerance Resistance: Overheating Spurious Rod Position Yes Medium

Indication

Resistor Fails Open Material Degradation No Rod Position Indication Yes Low

Vibration Erratic Rod Position Yes Low

Resistor Fails Closed Indication

Electrical Contact Degradation Spurious Group Average Yes Medium

Indication

Boric Acid Corrosion Spurious Asymmetric Rod Yes Medium

Indication

Group Average Amplifier Fails Component Overheating Incorrect High/Low Rod and Yes Low

Amplifier High Group Average Indication

Electrical Noise No Group Average Indication Yes Medium

Shorted Erratic Rod and Group Yes Medium

Power Surge Average Indications

Low Spurious Rod Group Average Medium

Maintenance Testing and Rod Alarms No

No Sequence Enable Function Low

Human Error No

_4
Table 7.6 FMEA - Babcock & Wilcox Absolute Position Indication System (Cont'd)

1 1 ~~~~~~~~~~~~~~~~~~~

Failure Failure Failure of

Component Mode Causes Effect Aging Occurrence

Limit Switches Switch Fails: No Sequence Enable Function

Open Same as Reed No Automatic or Out Inhibit

Switch Signals Same as Same as

Closed Failure No Feed and Bleed Permit Reed Reed Switch

Signal Switch Failure

Fails to Fully Open or Close Upon No CRA Full In or Full Out Failure

Demand Signal

Spurious Full Out/In Signal

Loss of Redundancy .-

Position Indicator Fails on: Electrical Connection Erroneous High/Low Position Yes Low

High Indication Degradation Indication

Low Indication Mechanical Wear No Rod Pbsition Indication Yes Low

Erratic Rod Position

No Position Indication Maintenance Error Indication No Low

Change on Demand

Position Select Switch Switch Fails: Mechanical Wear Unable to Select Desired Yes Low

Open Position Indication System

LOW

Closed Electrical Contact Degradation Spurious Rod Fault Signals Yes

Electrical Power Surge Erroneous Rod Position Yes Low

Maintenance Error No Low

I-.
Table 7.7 FMEA - Babcock & Wilcox Relative Rod Position Indication System

-j

L

ON

Component Failure Mode Faiur Causes Failure Effect Aging Prabablity of Occurrece|

Position Reset Switch Switch Fails: Mechanical Wear No powerlerratic power Yes Low

Open supplied to reset pulsar

Thermal Overheating resulting in inability to Yes Low

Closed reset RPI output to actual

Electrical Contact rod position Yes Low

Fails to Operate Upon Degradation

Demand Incorrect directional signal

Electrical Power Surge sent to reset pulsar Yes Low

resulting in incorrect RPI

Maintenance Error motion No Low

Reset Pulsar Fails to Operate Upon: Thermal Degradation No DC pulses sent to Yes Low

Demand stepping motor to correct

Electrical Short RPI output to correspond Yes Low

Operates on No Signal with API

Electrical Noise Yes Low

Spurious DC pulses

Electrical Power Surge supplied to stepping motor Yes Low

resulting in incorrect RPI

Degraded Electrical Yes Low

Connection

Fails Position Reset Switch Yes Low

Stepping Motor Fails to Position: Mechanical Wear Failure to drive Yes Medium

potentiometer resulting in

Fails to Operate Upon Bearing Failure no change in RPI Yes Medium

Demand

Thermal Overheating Spurious or erratic signals Yes Medium

Fails to Position Correctly to potentiometer resulting

Electrical Contact in erroneous RPI Yes Medium

Degradation

Incorrect direction of

Electrical Power Surge operation resulting in Yes Low

erroneous potentiometer

Maintenance Error output No Low
Table 7.7

FMEA - Babcock & Wilcox Relative Rod Position Indication System (Cont'd.)

Component | Failure Mode | Failure Causes Failure Effect I Aging i Prabablity of Occurrence

Potentiometer Component Fails: Overheating No signal sent to amplifier Yes Low

Open resulting in no PI

Degraded Electrical Yes Low

Closed Connection Erratic signals to amplifier

resulting in erroneous PI

Output Drift Electrical Power Surge Yes Low

Reset Relay Fails to: Mechanical Wear Incorrect PI Yes Medium

Open

Contact Contamination Incorrect group average Yes Medium

Close signal

Overheating Yes Medium

Spurious Operation Stepping motor inoperable

Electrical Surge with reset pulsar signal Yes Low

l . '
Table 7.8 FMEA - Combustion Engineering Reed Switch Position Indication System

Component Failure Mode Failure Causes Failure Effect Aging Probability of Occurence

Reed Switch Switch Fails: Electrical Connection No CEA position indication Yes Medium

Open Degradation to plant monitoring system

Closed Boric Acid Corrosion Erratic CEA position Yes Low

indication signal to plant

Fails to Fully Close Upon Moisture Intrusion monitoring system Yes Low

Signal

Mechanical Vibration No UEL or LEL indicaiton Yes Low

Fails to Fully Open Upon

Signal Mechanical Wear Incorrect CEA Yes Medium

misalignment faults

Fatigue Yes Low

Spurious reactor trips

Radiation Degradation Yes Low

Thermal Degradation Yes Low

Fixed Resistor Resistor Fails: Overheating Erratic CEA position Yes Low

Open indication

Electrical Contact Yes Medium

Closed Degradation Spurious Rod Drop Alarms

Resistor Out of Tolerance Mechanical Vibration Spurious CEA Deviation Yes Low

Alarm

Boric Acid Corrosion Yes Low

Amplifier Component Fails: Thermal Overheating Spurious signal supplied to Yes Low

High CEAC

Electrical Degradation Yes Low

Low No PI signal to CEAC

Electrical Noise Yes Low

Shorted

Electrical Power Spikes Yes Low
Table 7.8 FMEA - Combustion Engineering Reed Switch Position Indication System (Cont'd)

Comnponent Failure Mode Failure Causes Failure Effect Aging Probability or Occurrence

Cable Open Circuit Thermal Overheating Erratic signal supplied to Yes Medium

RSPT

Electrical Short Radiation Degradation Yes Medium

No signal supplied to RSPT

High Resistance Moisture Yes High

Boric Acid Corrosion Yes High

Testing Error No Medium

CRT Display Failure to Display Power Failure No CEA Position Indication Yes Low

Power Spike Yes Low
Table 7.9 FMEA - Combustion Engineering Pulse Counting Position Indication System

Component Failure Mode | Failure Causes | Failure Effect | Aging Probability of Occurrence

Plant Monitoring Systems Fails to Operate Loss of Electrical Power Spurious CEA Deviation Yes Medium

Operation Electrical Noise Alarm out-of-sequence Yes Medium

Electrical Power Surge Alarms Yes Medium

Pulse Count Relay Interface Relays Fail: Mechanical Fatigue *Loss of indication pulses Yes Low

Open Contamination sent to CEDM Yes Low

Closed Overheating -Incorrect indication of Yes Low

Operate Intermittently Electrical Connection CEDM pulses Yes Low

Degradation -Redundancy loss Yes Low

Electrical Power Surge Yes

CRT Display Failure to Display Power Failure Redundancy Loss Yes LOW

Power Spike No Position Indication Yes Low

Display

04
7.5 Control Rod Drive Cooling System (B&W and CE)

FMEAs for the B&W and CE control rod drive cooling systems have been incorporated into the

CRDM FMEA [Table 7.2 (B&W) and Table 7.3 (CE)I.

7.6 PRA Insi~hts

In addition to an FMEA, probabilistic risk assessments (PRAs) may also be used to determine

the effects of component failure. However, where an FMEA typically will assess the immediate effect

of component failure on the sub-system or system, a PRA will extend this to the overall plant level.

Several B&W and CE PRAs were reviewed to determine the specific plant effect due to these failures.

The four specific plant PRAs which were reviewed were Calvert Cliffs 1, Arkansas 1, Oconee 3, and

Crystal River 3.

Though CRD system failure was not specifically addressed in any of the PRAs, it was included

in the Reactor Protection System (RPS) analysis. In all instances, CRD failures were not significant

contributors themselves to plant risk, due to the fail-safe design of the control rods. However, if a CRD

system failure causes a plant trip, this challenges safety systems, and would be reflected in the PRA as

an increase in initiating event frequency and ultimately in plant risk.

7.6.1 Calvert Cliffs 1 (CE)3'

A system fault tree model was developed in the PRA for the RPS system. The configuration

boundary encompassed NSSS parameter measurement channels, bistable trip units, protective system

logic, CEDM power trip parts and RPS testing system. The analysis determined that a reactor scram

from full power would be successful if only one CEDM power supply bus de-energized, resulting in the

insertion of half of the CEAs. The top event for the RPS fault tree was a failure to scram if less than

41 CEAs entered the core when required. Contributing events to this failure were:

1) CEDM hold latches fail to release,

2) CEDM hold coils fail to de-energize, and

3) reactor core mechanical disruption prevents the insertion of greater than 31 CEAs.

The RPS fault tree developed for the Calvert Cliffs 1 PRA was not used in the quantification and

accident sequence evaluation due to the low failure estimates. Reliability estimates based upon operating

experience provided a more realistic estimate of RPS failure.

7.6.2 Arkansas 1 (B&W)37

The RPS fault tree in the ANO-1 PRA was developed to signify failure to achieve a satisfactory

reactor trip when demanded. Failures which contributed to this included RPS logic relay failures, trip

breaker failures, and mechanical failures of the control rods which prevented their insertion in the core.

The fault tree development began at the RPS trip breakers and relays, and was developed back

to the reactor trip parameters (i.e., high reactor coolant temperature, low reactor coolant pressure). The

combination of breaker/breaker and breaker/relay failures formed the top intermediate events. Cable

faults which represent shorts to power, were low probability events and were included for completeness.

Control rod drive mechanism mechanical faults were included as one of the RPS fault tree primary

events, with an event probability of zero.

7-21
The probability of RPS failure was calculated to be 4.2 x 106, and no RPS cut-set was greater

than 1 x 10.6.

7.6.3 Oconee 3 (B&W)3

A fault tree analysis of the RPS was not performed in the Oconee 3 PRA. Estimates of RPS

unavailabilityused to assess the frequencyof AnticipatedTransient Without Scram (ATWS) and transient

sequences were obtained from NRC and EPRI publications. These estimates, which were based upon

statistical evaluations of the data on failure and demand, indicate a wide uncertainty band of 10'- to 10.6

per demand. The approach in the Oconee 3 PRA for RPS unavailability was to use the NRC value of

3 x 10-5 per demand at 50 percent confidence. This value was a median estimate and a conservative

assumption was used as a mean value in the evaluation of ATWS sequences. The use of this RPS

unavailability value indicated that ATWS sequences were not significant contributors to the overall plant

risk. Therefore, a more refined estimate with a detailed RPS fault tree model was not performed.

Inadvertent CRA withdrawal and drop events were determined to have no effect on the other

reactor systems. Thus, these events were treated as reactor trip events, and event frequencies were

judged to be very low. The consequences of these events were limited to brief RCS overpressure

conditions.

7.6.4 Crystal River 3 (B&W)"'

A detailed RPS fault tree was used in the Crystal River 3 PRA to identify events which

contributed to a failure to insert the control and safety rods when required. The top event of the fault

tree was the failure to insert six or more control rod groups. In this model, failure to remove power to

the CRDM motors constituted RPS failure. Core description which inhibited CRA insertion, and stuck

rods, were included as contributors to reactor trip failures.

The dominant contribution to RPS unavailability was due to test faults of the reactor trip

modules. Faults in the CRD power train primary (ac power) breakers, and secondary (DC power)

breakers contributed approximately 35% of the total RPS unavailability. Other contributors were core

description which inhibited rod insertion, an insufficient number of CRAs dropping into the core, and

common mode failures in RPS instrumentation. These contributions were negligible since failure

probabilities were assessed to be less than 10'.

7.6.5 Conclusions

From the review of the four PRA studies of the B&W and CE control rod drives, it was

concluded that these systems were not modelled as broadly and as detailed as other important emergency

systems, such as emergency core cooling systems. The less detailed fault tree modelling was primarily

related to fundamental reactor design reasons. Generally, the RPS is a highly reliable system which is

designed to allow gravity insertion of the control rods by gravity upon demand. Also, in assessing

initiating event frequency for accident sequence evaluations, operating experience indicated that failures

of the control rod drive system were rare. This fact was primarily responsible for not performing a

detailed RPS fault tree model. The RPS fault trees for the PRAs evaluated were constructed so that the

overall risk assessment of the plant was complete.

7-22
-8. CRD Inspection, Surveillance, Monitoring, and Maintenance Review

Upon completion of the operating experience reviews for the B&W and CE designed control rod

drives, as described in Section 6.0, BNL prepared detailed questionnaires for each design. The

questionnaires concentrated on areas indicating increased failure or repair rates and which were most

critical to the continued operation of the CRD system. Specific information regarding the plants

corrective and preventive maintenance programs was requested. The questionnaires were sent to the

plants and returned to EPRI which then forwarded them to NUMARC, which then sent the responses

to BNL. Two of the eight operating B&W units, and four CE utilities responded, representing eight of

fifteen total operating CE units. Plant operating and maintenance procedures also were received from

several utilities. Meetings were held with B&W and CE to obtain further information on CRD design,

operating, and maintenance experience. The questionnaires which were sent to CE and B&W utilities

are included in Appendix C and D, respectively.

8.1 Babcock and Wilcox Plants

A system design comparison between the two survey respondents, designated as plants A and B, is

shown in Table 8.1. Both plants use similar control rod assemblies and part-length axial power shaping

rod assemblies. Each plant has accumulated approximately ten years of operating time on the CRDMs.

Plant A uses the original Type A design, while Plant B utilizes the improved Type C. The major design

improvements of the Type C CRDM include:

a) decreased operating power because of improved magnetic properties of the center section

of the motor housing,

b) increased stator assembly cooling due to cooling water grooves machined into the

casing rather than being wrapped on the outside,

c) varnish impregnated encapsulated stator's instead of epoxy encapsulated, and

d) belleville springs, instead of helical springs in the buffer assembly.

Table 8.1 Babcock & Wilcox System and Component Description

Plant A Plant B

Number of CRDMs 69 61

CRDM Type A C

Accumulated CRDM Operating Time 10 yrs. 10 yrs.

CRA Description 61 full length CRAs. Ag-ln-Cd poison 53 full length CRAs. Ag-ln-Cd poison

material. material.

Representative Reactor-Vessel Head 150TF 130-150TF

Ambient Temperature

Representative Stator Temperature 115F 90-105TF

Reed Switch Type Straight Axial Lead R4C

8-1
The lower average temperature for the Type C mechanism reflects the improved cooling

characteristics and the decreased power requirements. Heat generation in the stator depends upon

current flow and the resistance. Both units operate below the maximum design temperature of 180'F.

Operating procedures require that a mechanism which reaches 1800F be de-energized, and the reactor

tripped if more than one CRDM reaches this maximum temperature. Plants A and B use thermocouples

mounted in the top filler bushing to monitor stator temperature. The plant computer at each unit

monitors the temperature during operation, and provides an alarm to the operator if the temperature

is excessive.

Neither plant reported any major operational difficulties with the CRD system. There was no

abnormal interference or crud buildup for the control rods or the guide tubes. Neither plant performs

any regularly scheduled maintenance or inspections on the CRDMs. Plant A reported a one-time special

inspection on the guide tubes and the leaf springs on the torque tube, in response to operational

problems experienced at another plant.

As discussed in Section 6.0, the operational performance history of the B&W rod position

indication system has been good, and this is supported by the survey responses. Neither plant has

experienced any major problems or failures with the RPI system. Plant A replaced all the absolute

position indication (API) system cabling on the reactor head, including some to the reactor building

penetrations. Plant B reported two reed switch failures, and five relative position indication gearbox and

stepping motor failures due to binding.

Each plant reported isolated failures associated with the control and power systems. Plant B has

replaced six programmers and command modules, two sequence monitors, and three logic related cards.

Plant A reported isolated, non-related module failures.

Both plants have made, or are making significant improvements in system and components, as

shown in Table 8.2. Primary coolant leakage, has resulted in numerous incident reports to NPRDS, and

an NRC Information Notice alerting all PWR utilities to reported occurrences of such leakage and its

potential deleterious effects. Most leaks have been caused by the aging and deterioration of the spiral

wound asbestos/stainless steel gaskets, the primary seal between the reactor head and the motor tube.

Both plants retrofit these gaskets when leakage is discovered. The original asbestos filled, spiral-wound

design is being replaced with a graphite-filled stainless steel spiral wound design. Plant A has upgraded

approximately 20% of the mechanisms, while Plant B has retrofitted 36%.

The vent valve assembly was another source of primary coolant leakage. This valve is used to

bleed off non-condensible gases before restart, after removal of the reactor vessel head. Component

wear and improper assembly were the primary causes of leakage. Babcock and Wilcox has redesigned

the vent valve assembly, replacing it with a quick vent type. This change has resulted in quicker and

simpler venting, reducing instances of reactor coolant spillage. Plant B has retrofitted all of the CRDMs

to the quick-vent design. Plant A continues to use the original design, but has instituted a maintenance

program to clean, inspect and lubricate the vent valves. Several spills, not caused by vent valve operation,

were reported during fill and vent operations, resulting in damage to the stators and the rod position

indicating systems.

Both plants have incorporated minor improvements into the CRDM cooling system. Both plants

have installed flexible braided hoses with quick disconnect connections on the stator supply and return

lines (Figure 8.1). Each of these changes minimizes the time needed to connect and dis-connect the

cooling lines, thus reducing radiation exposure to maintenance personnel and cooling water spillage.

8-2
Table 8.2 Major Sub-System Modifications -- B&W Plants

PlantA PlantB ]

Flange Seals

Original Equipment Asbestos Filled Spiral Wound Asbestos Filled Spiral Wound

New Design Graphite Filled Stainless Steel Spiral Graphite Filled Stainless Steel Spiral

Wound Wound

Number of New Seals Installed 14 22

Vent Valve Original Design Quick Vent Type

Cooling System Flexible Braided Hoses with Quick Flexible Hose with Quick Disconnect

Disconnect Connections Installed on Connections Installed on Each Stator

each Stator Supply and Return Lines Supply and Return Lines

Cables and Connectors Replaced EPR Insulated Cables and Replaced Some Power Connector Pin

Connectors with Silicone Insulated Support Inserts from Neoprene to

Cables and Connectors. Cables Covered Silicone.

with Stainless Steel Braided Jacket

Instead of Fiberglass Braid.

Control System Modifications Asymmetric Rod Runback Modified to a Direction Error Circuitry Modified.

Dropped Rod Runback. Gate Drive Rod Stop Pushbotton Added to Inhibit

Units Modified. all Rod Motion.

Miscellaneous Out-of-Containment Added Test Panels to Facilitate Rod

Modification Drop Time Testing, and API

Calibration. Added Shunt Trip Feature

to DC Breakers.

However,these changes did not eliminate maintenance errors entirely, Plant B reported incomplete

coupling attachment to three stators, which were energized for three days before the error was

discovered. Subsequent testing on the affected stators revealed no apparent problems; however,

replacement is scheduled for the next refueling outage.

Embrittlement of the original neoprene insulated cables and connectors caused both plants to

replace them with silicone-insulated ones. Plant A has also replaced all of the fiberglass braided position

indication and temperature control cables with stainless steel braided-jacket cables.

CRD control system modifications have been instituted at both units. Plant A changed the

asymmetric rod runback alarm to a dropped rod runback. The gate drive units were also modified to

improve reliability. Auxiliary test panels were installed to facilitate testing the control rod drop and for

API system calibration and troubleshooting. Plant B modified the rod direction error circuitry to

improve reliability, and added a rod stop button to inhibit rod motion. Both units monitor the various

logic power supplies during operation to alert operators to malfunctions or failures.

8.1.1 Preventive Maintenance and Inspection Practices

Table 8.3 summaries information on preventive and corrective maintenance programs, obtained

from the questionnaire.

8-3
, / NUT ASSEMBLY

PORT CCNNECTCR

QUICK DISCONNECT

BODY ASSY

FLEXIBLE

METAL HOFSE QUICK DISCONNECT

STEM ASSY

PORT CONNECTOR

NUT ASSEMBLY

V\ ~ COOLING AATER HEACER

EXISTING MALE ELBOW

EXISTING CRCM STATOR COOLING

WATER TUBING AND NUT

Figure 8.1 CRDM Flexible Hose and Quick Disconnect Couplings

The plants' PM programs do not specifically address the mechanical components of the CRD

mechanism, with the exception of the 10 year inspection on motor tube welds as required by the ASME

Boiler and Pressure Vessel Code,Section Xl. Both plants non-destructively inspect 10% of the welds on

the periphery motor tubes, using dye penetrant.

An advantage of the Babcock & Wilcox design, compared to other PWR mechanical latch

designs, is that the entire CRD, including the leadscrew, may be lifted with the reactor vessel head during

refueling. After being decoupled from the CRA, the lcadscrew is fully withdrawn and mechanically held

in place to prevent insertion when power is removed from the stator. This permits a visual inspection

8-4
Table 83 Summary of Preventive Maintenance and Inspection Practices - B&W Plants

Plant A Plant B

Roller Nuts None None

Leadsc-ew None None

Male Bayonet Coupling None None

Motor Tube

NDE Dye Penetrant -- 3 drives every 10 Dye Penetrant -- 4 Periphery

years assemblies every 10 years.

Hyrdostatic Yes, at refueling Yes, 10 years.

Miscellaneous Video leak insp. of CRD flange

connection; Bolt and Nut Insp.

Leaf Spring Inspection on Torque

Tube

Flange Seals None Yes -- All are inspected

Gasket PM

Vent Valves Opened at Refuel Opened at Refuel

Hydrostatic Tested None None

Leak check at system pressure

Stators

Physical Insp. None None

Electrical Yes -- 18 months Yes -- 18 months

Megger -- Stator Stator Coil

Resistance for phases Resistance

500 v Megger

Cooling System Visual leak check hydro (10 yrs.) Check heat transfer from each

system

Cables and Connectors

Insulation Degradation Yes (18 months) Megger Yes (18 months) Megger

Connectors Yes (18 months) Yes (18 months) Megger

Moisture Seals Yes (18 months) None

Position Indication Systems

Physical Insp. Yes (18 months) -- Visual Insp. -- None

Cabinet Cleaning

Electrical Yes -- Calibration Test of Cabinet Yes -- RPI & API Calibration to

Equipment rod position

8-5
Table 8.3 Summary of Preventive Maintenance and Inspection Practices - B&W Plants (Cont'd)

| . Plant A | Plant B

Outside of Containment

Cables & Connectors Replace individual stator fuses. 500 v megger power cables and

Megger power cables connectors

Power Supplies Clean, inspect, tighten electrical Check and adjust output voltage

connections

Control & Logic Cabinets Clean, inspect, calibrate Clean, Inspect, Calibrate

Other CRD trip breakers CRD Trip Breaker

CRDM System Tests

Control Rod Position Tech. Spec. Required Tech. Spec. Required

Verification

Drop Time Tech. Spec. Required Tech. Spec. Required

Control Rod Exercising Bi-weekly Monthly Rod Exercising

Procedure

of the male bayonet coupling, if required. While not a trivial inspection due to the high radiation levels,

it can be remotely performed if necessary.

Both plants are in the process of upgrading the spiral wound asbestos/stainless steel flexitallic

seals with the improved graphite-filled seal on an "as needed" basis when flange leakage is identified.

Plant A performs a video inspection for leaks, while Plant B has incorporated the gasket inspection into

a PM procedure during refueling. This inspection for boric acid leakage can be performed with the head

in place, or with the head removed and stored on the refuel floor. One B&W plant has machined nine

viewing ports, one foot in diameter, around the CRDM shroud to monitor for boric acid leakage and ISI

inspections.

Table 8.4 shows the B&W recommended annual preventive maintenance.' The majority of

preventive maintenance inspections performed at both plants are electrical. The CRD stators are

functionally checked during each refueling by performing a 500 v megger on the stators and cables, and

by checking the resistance of each phase winding. If the phase-to-neutral readings are out of

specification, individual phase-to-phase readings are taken (15 for each stator). Similarly, 500v meggering

is used to test the integrity of the CRD cabling insulation and connectors during each refueling outage.

Plant A performs additional run, latch, and dropout current acceptance tests for each new stator.

The primary cause of megger failures has been stator moisture. The stator assembly is sealed

around the motor tube by two o-rings. O-ring degradation allows moisture to accumulate in the stator,

which is prevented from leaking out by the lower o-ring. When this condition is identified, the stator is

removed from service and powered for the time required to drive the moisture out. To minimize the

potential for moisture intrusion, these 0-rings should be replaced every time a stator is removed.

8-6
Table 8.4 B&W Recommended Annual Stator Preventive Maintenance

A. Electrical Tests

1. DC Resistance

2. Insulation Resistance

3. Thermocouple Resistance

B. Functional Tests

1. Minimum Run Current

2. Latching and Unlatching Current

Electrical tests are also performed on the relative position indication and the absolute position indication

systems, at each refueling, in both plants. The API is calibrated by adjusting the amplifiers to agree with

actual rod position,and the RPI signals are verified correct using the PI reset pulser to operate the RPI

motor to simulate various rod positions. Plant A also inspects the API sensor connections visually and

physically.

The B&W designed CRD system uses redundant 5,15,and 24 v DC power supplies. Preventive

maintenance practices verify and adjust, if necessary, the output of each power supply. An additional

maintenance procedure identifies faulty gate drives and SCR/diode failures in the power supplies which

ensure the proper trip function of the gate drive units and SCRs. Each power supply is cleaned during

this inspection. Both plants monitor the power supplies which alert the operators to malfunctions.

Plant A includes the control and logic circuitry in their PM program. Each module is calibrated

according to specific plant procedures, and a complete cleaning process is performed during refueling

outages, including a visual inspection of cabinet mounting bolts and welds, and verification that all fuses

are securely mounted and not open. Wires and cables are inspected for loose connections, broken wires,

harness chafing, and for any discoloration caused by overheating. The cabinet components are vacuumed,

the cabinet cooling system inspected, and the filters removed and cleaned. Plant B did not specifically

describe any PM procedure. Though excessively high ambient temperatures have caused logic and

control module failures, neither plant monitors the cabinet temperatures. System performance is

monitored, however; Plant A uses indicating lights which illuminate upon failure detection, and Plant B

uses a data acquisition system to monitor important control signals.

Each plant uses a reliability and trending program for the CRDMs. Plant B monitors three

indicators: component failure, stator isolation fuse temperature, and flange gasket replacement. Plant

A did not provide any specific program details. Each plant reported tracking system performance at

other plants; Plant A by reviewing industrial newsletters and Plant B by reviewing transient assessment

reports. Neither plant specifically mentioned using the LER or the NPRDS databases.

Each plant ranked three parameters, inspections, or tests which they found to be most important

in ensuring operational readiness. Plant A listed rod drop time testing, bi-weekly exercising, and cabinet

cleaning. Plant B relied on component failure trending, thermography of stator isolator fuses and.gasket

replacements.

8-7
8.2 Combustion Engineerine Plants

Four CE utilities representing eight units responded to the questionnaire. The four respondents,

designated as Plants C, D, E, and F have one, two, two, and three units, respectively. Generally, the

utilities with multiple units did not distinguish between the plants when responding to the survey. Table

8.5 presents the CEDM system design details for each plant which responded to the survey.

Combustion Engineering plants use either the magnetic latch or the rack-and-pinion type

CEDMs. All four respondents use the magnetic latch design. The operating time accumulated on these

CEDMs range from three to eighteen years.

All the plants use single and dual CEAs for reactivity control. Typically, the dual CEAs have

four control element rods per assembly and straddle two or four fuel assemblies, depending upon the

core design. The single CEAs are dedicated to one fuel assembly. Plants C and D use similar control

element rods consisting of a combination of B4C poison with Ag-In-Cd in the rod tips. Plant F uses all

B4C rods; however, due to irradiation induced swelling, the diameter of the pellets in the lower portion

of the rod are reduced and encased in felt metal. The felt metal contracts under irradiation, limiting the

amount of clad strain and swelling. Plant E did not provide information on the design of their control

elements.

With the exception of Plant C, the other three plants use part-length control elements. Plants

D and F use a three-zone design consisting of 134C in the top zone, a hollow, water-filled center zone,

and a solid Inconel tip.

Each plant uses different methods to attach the individual fingers to the spider. Plant C typically

uses dowel pins and welding; however, two of the corner rods screw into the spider. Plant D uses colletconnected hubs, and Plant F uses threaded and lock nut connections. The actual method used to connect

the rods to the spider does not affect system aging. However, the use of a nut permits individual rod

replacement in the event of rod failure.

Plant C reported irradiation-induced swelling of the B4C pellets in the tips of the control rods.

This swelling was measured by eddy current and has also caused wear near the tips due to flow-induced

vibration against the upper guide structure. Plant E also reported similar guide tube wear.

No unusual wear or indications of crud buildup on either the drive shaft or latches was reported.

Three plants reported damaged extension shafts, which occurred during refueling operations. Unlike the

B&W design, the drive shafts for the CEDM remain connected to the CEAs during reactor vessel head

removal. When completed, the drive shaft extensions are then disconnected from the CEAs and locked

in place for removal with the vessel's upper guide structure, which is stored underwater. Upon

completion of refueling, the upper guide structure is replaced and the drive shaft extensions are

reconnected to the CEAs. The head is partially lowered until the drive shaft extensions are engaged by

the CEDMs, and then lowered until sealed. At this step, the extension shafts may be damaged, because

the operators must ensure that the drive shafts are properly positioned within the CEDMs before

continuing to lower the head.

The refueling procedure is significantly modified for the System 80 plants. The head removal

process is the same; however, instead of disconnecting the extension shafts from the CEAs, both are

withdrawn into a lift rig, and latched to the work platform. After the upper guide structure is removed

8-8
Table 8.5 Combustion Engineering System and Component Description

Plant C Plant D Plant E Plant F

What CEDM is used at your plant? Magnetic Jack Magnetic Jack Magnetic Jack System 80 Magnetic Jack System 80

Number of CEDMs in the plant. 61 91 65 89

Operating time accumulated on 18 years approx. 7-8 years 12 years (Unit 1) 3 units with a total of 12 years

CEDMs. 10 years (Unit 2)

Full Length CEA (FLCEA) 80 FLCEAs, Poison unit 83 assemblies Poison with Ag- 57 per unit; 20 dual CEA and 75 FLCEAs

Description Ag-In-Cd Tip In-Cd Tips 37 single CEA BC poison with felt metal and

reduced dia. B4C in lower rod

Part Length CEA (PLCEA) No PLCEAs 8 PLCEAs, BC Top, hollow 8 per unit. 13, Inconel on top of B4C in

Description water filled middle, and solid lower rod portion

Inconel laser portion l

CEA Rod to Spider Connection 79 are dowel pinned and

Method welded, 2 have corner rods Collect connected hub . Threaded and lock nut in place

that screw into spider

Reactor Head Ambient Air Not Measured 120TF approx. 150TF (est) 11HOF

Temperature

CEDM to RV Seal Omega Omega Omega Omega

00

so
and stored, the extension shafts are uncoupled from the CEAs. The lift rig and the extension shafts are

then moved to a separate storage area, which allows the CEAs to be removed or shuffled within the

upper guide structure using special tools and the CEA change platform. Upon completion, the upper

guide structure is re-inserted, the drive shaft extensions and CEAs recoupled and lowered into position,

and the reactor head replaced.

None of the four plants reported major problems with motor tube welds. Plant E had to cut two

welds to install additional reactor monitoring instrumentation. Plant F had to cut one motor tube to

determine the cause for one CEDM which could not be fully inserted during rod-drop testing. The cause

of the problem was a dislodged ball bearing which fell from the multi-stud tensioner, and became wedged

between the control element and the guide tube.

Three of the plants (C, D, and F) reported instances of vent valve leakage. The vent valve is

located on the top of the motor tube and is used to vent the reactor following refueling. Seal welding

was the primary fix for the leaks at Plant C, while the, remaining leaks were repaired by replacing the

balls and the o-rings. Plants D and F also reported vent valve leakage which required seal welding.

All four plants have either replaced some or all of their original coil stacks. Plants C and D

replaced the upper and lower gripper coils. Plants D and E retrofitted their CED system with heavier

duty, System-80 coils.

Plant F reported a multiple CEA drop which was caused by damaged insulation on the coil leads

due to a fabrication deficiency. In addition to lower lift coil motion during CEA movement, this problem

resulted in an intermittent ground which introduced noise into the CEDMCS circuitry; interfering with

the holding voltage for another CEA, causing it to slip also. A program was initiated to inspect all of

the insulation, replace any defective coils, and to sleeve all of the coil leads. Additional investigations

concluded that the lower lift coil motion could be minimized by coupling the lower latch coil. This design

is being revised at the plant to reverse the coil polarity on each of the CEDMs.

The cables and connectors at all four plants have not required any major repairs. Plant C

reported cleaning connector plugs to improve resistance readings and reed switch connections. Plant E

reported isolated, loose pin connections on some connectors, and isolated cable insulation damage.

Each plant reported failures with the rod position indication system, particularly with the reed

switch position transmitters. Plant F reported that one of the RSPTs lagged the other RSPT during

insertion and withdrawal. The problem was corrected by re-tightening the input terminations for the

RSPTs. Plant E also reported that some cable jackets were cut by the metal identification tags.

There was not a particularly high incidence of power-supply failures. Plant D reported only

minor failures, but Plant F reported several power-supply failures over the last three months, prompting

the utility to evaluate and develop a replacement cycle.

Similarly, all plants reported minor operational difficulties with the CED control system, due

primarily to the modular construction of a majority of the components. Plant D reported regular optical

isolator failures (ten to fifteen per year). Similarly, Plant F reported approximately 45 failures in the

three units. Examples of these failed components include CEA timersphase synchronous cards, optical

isolators, coil driver cards, and CEA enable and sequencer cards. Plant E reported failures mostly in the

cabinet mounted 15 and 28v dc power supplies, and individual and group program module failures.

8-10
Each responding plant has made system improvements to the CED system (Table 8.6). Two

plants have retrofitted their original CEDM system with System-80 coils.

In an effort to extend coil life, Plant D reduced the holding voltage from 35v to 25v. Plant F

undertook a program to sleeve all CEDM coil leads, and to reverse coil polarity to minimize the motion

of the lower lift coils.

Combustion Engineering plants use forced air for CEDM cooling. The operating experience for

the system has been good. Plant C reported one instance of loss of cooling, but due to system

redundancy, no degradation occurred. All plants monitor the inlet air temperature, and an alarm is

annunciated in the control room if the setpoint is exceeded. Plant C maintains the temperature between

90 and 120'F, with an alarm setpoint at 130'F. Plant F has installed dampers to prevent backflow

through the standby fans, ensuring maximum flow to the CEDMs.

Plants C and E reported changes to the rod position indication system. Plant C modified the RPI

system so that all signals are multiplexed into a microprocessor which displays position information for

all rod groups, individual rods in a group, the group mean, and rod height deviation. Plant E modified

the output of the reed switches to provide more precise rod position information. The output was

changed from 1-5V dc at 2-inch increments to 5-1OV dc for 1.5-inch increments.

The CED control system has been modified at each plant. Plants C and E installed redundant

12 v power supplies for the coil power programmers. Plant C also installed a hold bus which permits the

coil power programmers to be bypassed during maintenance. Plant D installed automatic timer modules,

replacing the original timer cards. Plant F, in response to the multiple dropped CEA event discussed

above, updated the ground fault detectors in the CEDM M-G Control System. A new under voltage

(UV) detection system was also added, replacing the old system which was not adjustable and susceptible

to drifting with age.

8.2.1 Preventive Maintenance and Inspection Practices

Table 8.7 summaries the Combustion Engineering recommended CEDM maintenance. Specific

recommendations are provided for the gripper coils, vent valves, and the drive shafts.

Following each refueling, coil traces are obtained for each individual CEDM. These traces, verify

the proper actuation of the coils, and the current level applied to the coils (Figure 8.2). CE recommends

that these traces be visually compared to the traces obtained from the previous cycle for each CEDM.

Any changes should be evaluated and dispositioned by CE. Coil resistance is also obtained to ensure that

there are no electrical shorts. A visual inspection is also performed if any indication of boric acid leakage

is found.

The o-rings and the stainless steel ball are also replaced each time the vent valve is opened to

ensure that no worn or damaged parts are re-installed, resulting in primary coolant leakage. A visual

inspection is also made on the drive shafts, during refueling, following the removal of the reactor vessel

head and upper internals. Although this inspection is performed from above the components, CE states

that any significant area of wear would be visible.

8-11
Table 8.6 CED System Design Changes and Modifications

Plant C Plant D Plant E Plant F

Coil Stack Assembly No Change Reduced holding voltage Retrofitted System 80 Coils Coil polarity reversed to

from 35 v to 25 v, minimize lower lift coil

Retrofitted System 80 Coils motion. Sleeved all CEDM

coil leads.

CEDM Cooling System No Change No Change No Change Dampers installed on top of

fan stack to prevent

backflow through standby

fan.

Cable and Connectors No Change No Change Threads modified on ..

connectors from five thread

to ACME

Rod Position Indication RPI multiplexed into No Change Reed switch output changed No Change

System microprocessor which from 1-5 vdc step outputs at

displays PC for all rod 2 in. inc. to 5-10 vdc at 1.5 in

groups, ind. rods, and ind. increments

rods in a group about mean.

System also identifies rod.

height direction

CEDM Control System Redundant power supply and ACIM replaced original Changed no. of CEAs New undervoltage detector

monitoring system installed timer cards between Groups 4 & 5, system installed.

for each coil power therefore had to reverse logic Ground fault detector

programmer. Hold bus for autometer and sequ. improvement installed in

installed to bypass CPP group control. CEDM M-G control system

during maintenance activities. Installed dual 12 v power

supplies for coil power

programmer.

00
Table 8.7 CE Recommended Annual CEDM Preventive Maintenance

I. Gripper Coils

a) Coil Operating Traces

b) Coil Resistance

c) Visual Inspection (if the CEDM Cooling

System is degraded)

II. Vent Valve

a) Replace O-Rings

b) Replace Stainless Steel Sealing Ball

III. Drive Shafts

a) Visual Inspection (when upper reactor internals are

removed)

Applied

Actuation Time

Figure 8.2 - CE Coil Current Trace

8-13
Table 8.8 summarizes the preventive maintenance and inspection practices for the CED system

used at each plant. Plant B reported eddy current inspections of the control element rods following every

third cycle. This is normally not performed, but was instituted to monitor wear on the CEA rod tip due

to flow-induced vibration against the guide tubes in the upper guide structures, as well as cladding strain

induced by B4C swelling. As discussed in Section 6.0, Maine Yankee experienced a failed CEA center

finger end cap, which allowed the B4C pellets to spill into the guide tube, preventing the CEA from fully

inserting. Subsequent inspections revealed two additional fingers had missing end caps, and six others

with cracks. CE uses several different rod designs; plants with the old design, which did not account for

the degree of B4C swelling, may be susceptible to similar type of failures. None of the other three plants

reported similar inspections; however, Plant E reported inspecting the guide tubes during each refueling,

but few details were provided about this inspection.

No inspections are performed on the CED latches, and only Plants C and D made cursory visual

inspections on the CED drive shafts during refuelling. A detailed inspection is difficult to make since

the lower portions of the drive shafts are moved with the upper internals. Therefore, only gross defects

would be visible during an inspection.

Each plant inspects pressure housing welds as required by the ASME Boiler and Pressure Vessel

Code, Section XI. Weld integrity is inspected by ultrasonic and dye-penetrant techniques. These

inspections are required every ten years on 10% of the periphery assemblies. Hydrostatic integrity is also

inspected at each refuelling and during the ten year inspection. The Omega seal used to seal the flange

to the vessel head is also subjected to a remote visual inspection. The CEDM is only removed if there

are indications of primary coolant leakage. There are no programs at the plants which track and trend

seal degradation.

The vent valve, located on the top of the CEDM assembly, is used to vent the non-condensible

gases following refueling or coolant level decrease. Plant C visually checks the integrity of this subassembly after each opening, while Plant D includes this inspection in their ten year inspection program.

Plants E and F provided no specific details for their program.

The majority of inspections performed on the CEDMs are electrical. Only Plant E physically

inspects the coil stack by performing a visual inspection for any discoloration caused by overheating.

Electrical inspections and testing consist of a 500v megger to test the coils' insulation and a

resistance check on the coils to ground. Typically, these checks are made from termination points in the

CEDM control cabinets outside the containment. The cables and connectors are also meggered to check

the insulation. The connector and connector seals are inspected at the same time. Plants C and D

visually inspect the cables and connectors. None of the plants directly monitor the coil stack

temperature. However, by measuring the coils' resistance, a conversion to temperature is obtainable, but

these calculations are not normally performed.

The preventive maintenance program at each plant includes the CEDM cooling system. Plant

C annually sounds the foundation bolts, and inspects and repairs as necessary the fan blades, duct work

and housing. Plant F cleans and inspects the cooling coils, visually inspects all of the fans, and lubricates

the fans' motor bearings.

The rod position indication system consists of the pulse-counting and reed switch position

indication systems. None of the four plants reported any physical inspections; however, all performed

electrical tests. Plant C tests the functioning of the rod drop, lower limit, and zero position switches.

8-14
Table 8.8 Summary of CE Plants Preventive Maintenance and Inspection Practices

Plant C Plant D Plant E Plant F

Control Elements, Guide Tubes, Eddy Current Insp. on CEA rods Not normally inspected Guide tubes inspected each No Inspection

CEA Shrouds every 3rd cyde. refueling

Visual inspection of CEA

shrouds and upper guide

structure.

CED Latches Not Inspected Not Inspected Not Inspected Not Inspected

CED Drive Shaft Cursory visual inspection during Cursory visual inspection Not Inspected Not Inspected

refueling l

CED Pressure Housing Seal Not Inspected Yes as per ASME Section XI No Response Yes, every time a weld is made

Welds requirements using both LP exam

ultrasonic and dye penetrant.

Hydrostatic Yes, following refueling and 10 Yes, 10 year intervals No, welds are inspected at Yes, visually every time a weld

year inspection system pressure is made

Seal Inspection Not Inspected Yes, as per Section XI No Yes, remote visual exam for

l_____________________________ requirements boric acid leakage

Vent Valve Hydrostatic following every Yes, every 10 years No answer No

refueling l

Coil Stack Assembly Physical No No Yes, when installed No

Inspection

Electrical Inspection Coil resistance and coil-to- 500 v megger and resistance coils VOM resistance test No Response

ground resistance is measured on each coil every refueling

for each coil using 500 v megger

CED Cooling System Cooling fans are included in General inspection each refuel, Inspection are part of STP's Cooling coils inspection and

annual PM lube and filter cleaned. CEDM fan motors

lubed

Cables and Connectors

Inspection for Insulation Yes, visually Yes, 500 v megger Yes, visual and meggering No

Degradation or Wear

Connector Pin Condition Yes, during coil stock resistance No Yes, each refueling No

test

00
Table 8.8 Summary of CE Plants Preventive Maintenance and Inspection Practices (Cont'd)

[ ______________Plant C | Plant D [ Plant E PlantF ]

Connector Watertight Seal Yes, during coil stick resistance No No No

Inspection testing l

Rod Position Indication System No No No No

Physical Inspection

Electrical Tests Yes, voltage traces are taken for No Yes, following each refuel, RSPT Yes, RSPT are functionally

coil sticks. Lower limit and zero traces and pulse counting tested. CEA position isolation

position switches are exercised. indicators are tested amplifier is also calibrated every

18 months. Also, the RSP'Ps

CEAs are cold functional tested and CEA pulse counting post

and exercised monthly. ind. is verified within 5-2 inches.

Cables, Connectors, and No Inspection No Inspection After each refuel visual and No specific tests are written,

Termination Outside resistance readiness tests however, termination are

Containment inspected during 18 month

breaker PM

Power Supplies Functionally tested each refuel Logic power supplies Power supply recalibration each Power supplies are inspected,

recalibrated each refuel refuel tested, and recalibrated every 18

months.

CEDMCS UV relays are tested

every refuel. CEA position

amplifier are calibrated every 18

months.

Viscorder traces of the CEA

coil voltage are taken to verify

proper operation of CED

system.
Plant E tests the individual rods which trace both position indication systems. Similarly, Plant F

functionally tests the RSPTs, and calibrates the CEA position isolation amplifier. Both systems are

verified to be accurate within 5.2 inches of each other.

Each plant performs preventive maintenance on the CED system power supplies, by testing,

inspecting, and re-calibrating the power supplies during refueling outages. Each plant monitors the

functioning of the power supplies during operation, and provides an alarm in the control room upon low

voltage. Plant C also noted that a replacement program is in place, each refueling outage, to improve

the reliability of the power supplies.

Similarly, the plants all have a program to maintain the logic and control cabinets. Each plant

cleans the cabinets and monitors the operability of the system during the functional and surveillance CEA

testing. Plants C and D monitor the temperature of the cabinets, and provide high-temperature alarms

in the control room. The average cabinet operating temperature varies between 70-80'F, with 1000F

being the alarm setpoint.

8.2.2 CED System Tests.

As required by plant Technical Specifications, each plant performs functional and surveillance

tests throughout the cycle to ensure the operability of the CEA system. All of the plants perform the

required position verification test. All plants compare the responses of the reed switch and pulse

counting position indication systems. Plants D and F perform this test every shift, while Plant C performs

it monthly.

Following refueling, each plant is also required to perform rod-drop time tests to ensure that the

CEDMs are operable and can insert in less than 3 seconds. Each plant monitors the CEA breakers and

the RSPTs to ensure compliance.

During a typical reactor operating cycle, the CEAs do not move often and therefore, stay in the

same approximate location in the upper internals. Due to the coolant crossflow effects in this region of

the core, vibration between the control elements and the guide tube may cause fretting and subsequent

wear, particularly when the rods are stationary. To alleviate this potential problem, Plants C, D, and F

perform a CEA exercising program monthly. Plant C specifies that the CEAs move a minimum of eight

steps, Plant D moves the rod three steps in alternate directions, and Plant F moves each rod four inches.

Plant E did not discuss their specific program.

8.2.3 General

Plant F is the only plant which reported a reliability and trend analysis program for the CEDMs.

The NPRDS database was used to monitor component failures at other plants.

Finally, each respondent listed the three most critical tests, parameters, or inspections that assure

operational readiness. Plant C listed monthly CEA exercising, refueling drop time checks, and position

information and coil resistance readings. Plant D responded with drop tests, monthly exercising, and

refueling calibrations. Plant E chose connector-to-cable resistance, venting, and rod testing. Plant F

considered monthly CEA exercising, coil voltage adjustments, and rod drop time testing the most

important.

8-17
8.3 Summary of Survey Results - (B&W and CE)

The responses to the questionnaire provided valuable information on the operating history and

maintenance programs required for the control rod drive mechanisms. The preventive maintenance

programs at the plants demonstrates that the system requires both maintenance and surveillance to

ensure it is operational, and that the system will allow gravity insertion of the control rods upon removal

of the holding power by opening reactor trip switchgear.

8.3.1 Cable Integrity

The use of meggering to check the electrical integrity of both cables and components is very

common in the nuclear industry. This test is primarily a go/no-go type of test which cannot detect

insulation degradation over time.4" Failure of meggering may not always indicate component

degradation, since the test is sensitive to environmental conditions, such as humidity. Since the area

inside containment tends to be humid, a utility may energize a cable or component for a period in an

attempt to dry it, and then retest, before replacing the component.

A potential alternative to meggering is to use an on-line monitoring system. The Electronic

Characterization and Diagnostics (ECAD) System,42 developed by ECAD Services, is an example of a

commercial system. ECAD has been used at San Onofre to demonstrate the integrity and operability

of various electrical circuits by measuring standard electrical characteristics. From this data, ECAD

verified the operational condition of a circuit, located failed areas, and established reference baseline data

for subsequent retesting to determine cable degradation. Such a system, when used to monitor the

CRD's, could detect and highlight general problem areas which may be indicative of aging, and permit

corrective maintenance to be performed before a failure occurs. A similar on-line data analyzing system

has been developed by Mitsubishi, and has been used at several Japanese plants.

8.3.2 CRDM Operation

Large scale CRD failures which prevented control rod insertion have not occurred. However,

no current inspection program monitors the condition of the CRD internals, which could indicate age

degradation. Motor current signature analysis (MCSA)43 is a non-intrusive, non-destructive, advanced

monitoring method which could provide this information. Originally developed for motor-operated

valves, MCSA senses variations in the electrical current supplied to a device. When analyzed either in

the time or frequency domains, and trended over time, these variations provide accurate indications of

age-related degradation. Examples include bearing wear, motor RPM, and roller nut/leadscrew meshing.

Such information would be essential in determining the amount of aging degradation and allow for

replacement before any major failure which could affect both the functioning of the drive and plant

safety.

8.3.3 Primary Coolant Leakage

Primary coolant leakage and its deleterious effects, have been the subject of several operating

experience reports, including a NRC Information Notice. The Babcock & Wilcox units which responded

to the survey are retrofitting the originally installed spiral wound asbestos/stainless steal gasket with a

graphite-filled one. Both units have replaced some of the gaskets, and have stated that the remainder

will be replaced on an "as needed" basis when leakage is observed. Typically, no inspection is performed

on these gaskets since the CRD mechanisms are not normally removed. Due to the importance of these

gaskets, the remaining old ones should be replaced as soon as practical, even if no leakage is observed.

8-18
An on-going periodic inspection program should be instituted to monitor the integrity of the replacement

gaskets to allow the development of a life time estimate. With this data, gaskets could be replaced on

a scheduled basis, and not upon failure.

Babcock & Wilcox's redesign of the ball type vent valve has simplified the venting process, and

decreased coolant leakage and personnel exposure. Plants still using the original vent valve should

evaluate the redesign for use at their facility. All of the vent valves could be replaced, or in lieu of

repairs, individual valves could be replaced when they leak.

The results of the survey have shown the need for sound maintenance practices and procedures

is not alleviated by component redesign and simplification. This point was highlighted at the plant which

experienced lack of stator cooling resulting from a failure to connect the cooling lines after refueling.

This event occurred even though the plant modified the cooling system and simplified the process by

installing quick disconnect couplings.

None of the Combustion Engineering respondents alluded to any type of inspection program for

the Omega seals, other than checking for boric acid leakage. A scheduled, periodic inspection program

for these seals is needed that would allow the seals' condition to be monitored over time, and would

identify deterioration so that seals of a similar age could be repaired before failure.

83.4 In-Service Inspection

Each plant inspects the integrity of the welds on the CRD housing every ten years as required

by the ASME Boiler and Pressure Vessel Code, Section XI. However, this is a partial inspection,

covering only 10% of the periphery housings. The other mechanisms also should be subjected to the

same type of inspection. Present inspection methods do not easily lend itself to this type of inspection.

The area on the top of the reactor head is cramped, and the mechanisms do not have much space

between them. A gasket inspection program, as discussed above, would also allow the welds on the

housing to be inspected. From this data, lifetime estimates could be developed which would allow

decisions to be made about which other mechanisms may have to be removed for inspection.

83.5 CRD Control System

Excessively high temperatures have been identified as the root cause of failure of several CRD

control modules. Failures in these complex systems may directly affect plant availability. Repairs are

difficult, because there is no well-defined diagnostic method. Only two of the responding CE plants

monitor cabinet ambient temperature during operation. In the event of a slipped or dropped rod, an

adequate shutdown margin must be demonstrated to continue full power operation. An on-line

monitoring system, capable of assessing the operational status of the system and diagnosing malfunctions,

would be of great assistance. Such a system could range from simple temperature and power monitors

to a state-of-the-art expert system. CE is currently modifying the ACTM at one plant to provide

historical data on ACTM and CEDM occurrences which could be used for preventive maintenance

purposes also.

Similarly, overheating of the stator coils has been identified as the cause of system failures. Both

of the B&W plants directly monitor stator temperature with thermocouples. The CE plants without

ACTMs, rely on an indirect determination using the resistance of the coils to calculate temperature. A

direct measuring system is more advantageous. An alarm can be provided in the control room which

alerts the operator to high temperatures, allowing action to be taken, or power to that coil reduced while

8-19
the problem is investigated. Temperature data is also important in determining aging changes in the

coils, and could be used to determine the life of the coils. The ACTM upgrade provides for coil current

monitoring via non-intrusive current sensors. Excessive coil currents are indicative of coil heating and

aging effects.

The complexity of the control system lends itself to the use of an on-line expert system, which

would be valuable for operational monitoring and troubleshooting. The expert system would evaluate

fault signals and determine the exact location of the system failure. An expert system uses two reasoning

mechanisms,system design and operation knowledge, to accomplish this. Thus, by evaluating the causeand-effect relationships learned, the expert system produces a diagnosis graph for normal operation.

System malfunctions are represented by deviations from this graph. The exact cause of the failure may

be identified to the smallest system component, depending upon the complexity of the system's design.

A monitoring system, supplemented with a preventive maintenance program, could increase the

reliability of the system. Infrared thermography scans are an example of a simple inspection, which when

performed on regularly, can provide valuable information on the condition of the equipment. Scans of

electrical modules, connectorsand cables for signs of overheating would allow corrective maintenance

to be performed before they fail. Commercial equipment permits recording of the field scans for

comparison to an original baseline scan. This would allow an evaluation of the characteristics of the

component as a function of aging. From this, component life may be determined, supporting a scheduled

replacement of major modules before failure. Software is also available to convert the video output to

a digital format, so the information can be analyzed and processed on a PC.

8.3.6 Reliability Data Bases

A important component of any maintenance program is a reliability data base. Such a database

should include component information both from plant specific and industry-wide data. One of the CE

respondents, and both of the B&W plants, reported having such a program in place. Commercial

databases, such as the NPRDS maintained by INPO, provide detailed data on component operation and

failure, including age at failure. The use of such a program would assist a single utility to obtain this data

quickly from other similar plants. The plant then could evaluate occurrences at other plants, and take

appropriate action before similar failures at their plant.

8-20
9. CONCLUSIONS AND RECOMMENDATIONS

The control rod drive systems for Combustion Engineering and Babcock & Wilcox provide

reactor control functions by positioning the control rod assemblies in response to manual and automatic

control signals. The control rod drive control systems control the short term reactivity effects during

normal operation, and allow a rapid reactor shutdown through the gravity insertion of the control rods

upon removal of .CRDM power via opening of the reactor trip switch gear, which is part of the

reactor/plant protection system.

A Phase I aging evaluation has been completed for both system designs. The major components

of each system, and the primary operating and environmental stresses which cause age degradation and

failure, were discussed. A Failure Mode and Effect Analysis (FMEA) for each major sub-system was also

completed. The CE and B&W CRD systems have been the subject of several NRC and EPRI studies.

These studies and the four NRC Information Notices documenting system degradation and failure were

reviewed, and the results summarized.

The review of the 1980-1990 operating experience indicates that each system has experienced age

related failures and degradations, resulting in significant plant effects, including dropped rods, power

reductions, and plant scrams. The systems are also susceptible to human errors and inadequate

maintenance. The control rod drive system has never failed to shut the reactor down upon demand.

System failures have resulted in increased component stresses, and in thermal and pressure cycles, which

challenge the operation of the other plant safety systems.

A survey of utilities was conducted to obtain additional CRD operating experience, and

information on their maintenance, surveillance and inspection activities that mitigate aging. Meetings

were also held with CE and B&W to determine recommended maintenance, inspection, surveillance, and

operating restrictions.

9.1 Conclusions

9.1.1 Combustion Engineering

The CE Control Element Drive System is being used in fifteen plants. Thirteen of the plants use

the magnetic latch CEDM, while Palisades and Fort Calhoun use the rack-and-pinion design. Based

upon the review of the operating data bases, the following conclusions were reached.

1) Degradation and failures of the CE control rod drive control system accounted for the

majority of system failure occurrences. The primary result of control system degradation

was dropped or slipped rods due to the improper or sluggish gripper operation. In

response to these, CE has upgraded the control system in five units to incorporate

microprocessors and current sensors which monitor the mechanical actuation of the drive

mechanism in order to control the voltage/current sequencing during rod movement.

The upgrades also monitor for abnormally high or low currents, and take corrective

action. All CE units also have redundant logic power supplies.

2) The majority of primary coolant leakage occurrences resulted from the failure of the

rack-and-pinion rotating seals. Only-the two oldest CE plants (Palisades and Fort

Calhoun) use this type of CEDM. (Section 6.4.3.1)

9-1
3) Failures of the reed switches and Reed Switch Position Transmitter (RSPT) typically

resulted in loss of position indication. These failures also represented a loss of

redundancy, since two (and in some plants-three) independent rod position indication

signals are required by Technical Specifications. (Section 6.4.4)

4) The CED System was susceptible to degradation and failures caused by human error.

Inadequate and improperly performed maintenance procedures resulted in leadscrew and

gripper coil failures. When these failures occur, utility management review the

importance of proper system maintenance, and also highlight potential plant effects, with

the operating staffs. (Section 6.4.5)

5) The CED system was susceptible to environmental degradation. Several occurrences of

electronic component overheating caused by the loss of forced air cooling of the

electrical cabinets were reported. Plants should consider installing a monitoring system

which would alert the operators to system degradation and initiate maintenance before

system failure. (Section 6.4.5)

6) A significant number of LERs documenting system degradation and failure had unknown

causes. This was representative of inadequate analysis of root cause of failure. Instances

of repeat failures were noted before the problem cause was determined and corrected.

This placed unnecessary stress upon the CRD system and components. (Section 6.4.6)

9.1.2 Babcock & Wilcox

The review of the B&W CRD operating experience with the B&W Control Rod Drive system

also demonstrated susceptibility to age related degradation and failures. From this study, the following

conclusions were reached.

1) Numerous incidents of primary coolant leakage occurred as a result of age degradation

and failure of the CRDM flange spiral wound, asbestos filled gaskets. B&W has

redesigned the gasket to eliminate future leakage. (Section 6.4.3.1)

2) The early two-channel B&W rod position indication system became inoperable upon a

reed switch failure. The new redundant four-channel design remains operational with

several failed reed switches. The reed switch was also redesigned after numerous

instances of spurious operation. These changes have improved the reliability of the

subsystem. (Section 6.4.4)

3) Failure of the gate drives, which actuate the SCRs, accounted for the majority of control

system failures. Aging of the diodes was the primary failure cause. (Section 6.4.2)

4) Failure of the CRDM vent valve assembly also accounted for primary coolant leaks. The

hydraulic seal and quick vent redesigns have decreased the failure occurrences. (Section

6.4.3.1)

5) Several incidents of loose parts in the core have been reported. Though not age related,

continued operation with loose parts should be avoided due to the potential that they

may become lodged either in the CRDM internals or guide tubes, which could prevent

the full insertion of the CRA. (Section 6.4.7)

9-2
6) The B&W CRDM internals currently are licensed for a 20-year lifetime, and several

plants are close to this limit. This limit is based upon several conservative estimates,

including 126,000 feet of leadscrew travel, 500 trips, and exclusive use with regulating

rods. Plant data indicates that a typical CRDM leadscrew has travelled only a fraction

of the allowable limit, the total number of plant scrams is considerably less, and the

CRDMs are used interchangeably between regulating and safety rods. B&W is planning

to remove several CRDMs with approx. 15 years of service from Oconee and Crystal

River and to measure component wear. These results should be evaluated before

considering any increase in life. (Section 3.3)

9.2 Recommendations

Based upon the results of this Phase I aging study, and the results obtained from the survey of

operating utilities, the following general and design-specific recommendations are made. The general

ones are applicable to both CE and B&W plants, while the specific ones are applicable to either CE or

B&W plants.

These recommendations are intended to highlight the critical areas which have been susceptible

to aging degradation and failure. If the recommendations are implemented, utilities will be able to detect

and mitigate age related degradation more effectively.

9.2.1 General Recommendations

1) Plants should continue to access one or more of the operating data bases to gain

information on CRDM system performance and failures experienced by other utilities.

Often, similar failures are experienced at other plants, and knowing the failure cause and

corrective action taken may be useful for other plants. Relying solely upon vendorsupplied information may not provide sufficient information, in time, to preclude similar

failures from re-occurring.

2) Emphasis should continue to be placed upon detecting the root cause of failures.

Numerous instances of similar failures, at the same and different plants, were reported

with unknown failure cause. A thorough, initial analysis of root cause failure would

decrease the potential of repeated failures.

3) The lifetime of both the B&W and CE control rod drive mechanisms is based upon

conservative estimates for total number of steps and leadscrew travel. Utilities should

monitor these parameters for each control rod drive to ensure that the limits are not

exceeded. This action will also provide useful information if the control rod drive

mechanisms are to be considered for use beyond their present specified lifetimes.

9.2.2 Combustion Engineering

1) Because the gripper coils are susceptible to overheating, the ambient temperature of the

area above the reactor head should be monitored to provide an early indication of

degraded cooling, and ensure that maintenance is performed before the coils fail.

9-3
2) Based upon operating experience, all CEDM flange seals should be inspected each

refuelling for indications of boric acid leakage. Any indication of leakage should be

located and repaired prior to return to power.

3) Installation of gripper coil thermocouples (instead of the indirect temperature methods

currently used in some plants) would provide a direct, early indication of potential coil

failure due to overheating. Alternatives such as the non-intrusive, coil current

monitoring performed by the microprocessor based ACTM, would also provide the same

benefit.

4) Because the electrical components contained in cabinets are susceptible to failures from

overheating, the ambient temperature inside of the electrical cabinets could be

monitored. An alarm should be annunciated if the temperature exceeds the design

setpoint.

5) The thermal embrittlement of any cast stainless steel pressure housings should be

monitored. Inspection of the cast pieces should be included in the ten year inspections

performed by the plant.

9.2.3 Babcock & Wilcox

1) Because the high occurrences of spiral wound, asbestos filled gasket leakages, all the

original designed gaskets should be considered for replacement. The replacement should

be performed when primary coolant leakage is evident. The remaining CRDM flexitallic

gaskets should be changed on scheduled basis over several outages to minimize plant

schedule impact. In addition, the replacement gasket performance should be monitored

to ensure that there are no age related degradation mechanisms uniquely associated with

the hew material.

2) Vent valve inspection should be performed at each outage. The valves should be

replaced if primary coolant leakage is evident. The use of new designs does not preclude

the need for proper component maintenance.

3) Electrical cabinet PM should continue, particularly cleaning the electronic components

and the filters. Numerous instances of dropped or slipped CRAs due to thermal

overheating were reported.

4) The o-rings which seat the stator coils around the motor housing should be replaced each

time the stator coils are removed. Deteriorated o-rings may allow moisture to intrude

on the coils, causing electrical shorts.

5) Plant operators should remain cognizant of the potential that loose parts may play in

causing inoperable CRDMs. Control rod handling tools should be inspected before and

after use to ensure no broken parts.

9.2.4 Advanced Monitoring Techniques

Both CE and B&W plants should evaluate the use of commercial advanced monitoring

techniques. The results of this Phase I study indicated that most control rod drive maintenance is

9 i4
performed in response to failures, rather than being preventive. These techniques can provide early

indications of aging degradation, trendable for the life of the component. These techniques are noninvasive, which could decrease system failures and the resultant significant plant effects. Each application

should be evaluated for its own cost benefit.

1) Motor current signature analysis may be considered to detect mechanical degradation

including wear, bearing and seal failures. This is a non-invasive procedure which

monitors the current supplied to the CRD. The results may be stored for trending.

Several utilities have incorporated this technique with various degrees of success.

2) Infra-red thermography may be used to inspect for component overheating, particularly

for cabinet mounted equipment.

3) Most plants commonly use meggering to check for electrical degradations. Meggering

is not capable of detecting certain types of degradation. ECAD (Electronic

Characterization and Diagnostics), or other, non-invasive techniques have been used at

several plants to determine the electrical integrity of the CRD system. The results may

allow for trending analysis to be performed.

9-5
10. REFERENCES

1. Vora, J.P., "Nuclear Plant Aging Research (NPAR) Program Plan," NUREG-1144, Rev. 2, June

1991.

2. Fullwood, R., et al., "Aging and Life Extension Assessment Program (ALEAP) Systems Level

Plan," BNL Report A-3270-12-86, December 1986.

3. Gunther, W., and Sullivan, K., "Aging Assessment of the Westinghouse PWR Control Rod Drive

System," NUREG/CR-5555, BNL-NUREG-52232, March 1991.

4. Final Safety Analysis Report, Maine Yankee, Docket No. 50-309, January 1984.

5. Final Safety Analysis Report, Arkansas Nuclear One, Unit 2, Docket No. 50-368, August 1992.

6. Final Safety Analysis Report, Waterford Steam Electric Station, Docket No. 50-382,

December 1991.

7. Dufrane, K.F., et al., "Wear Measurements of Nuclear Power Plant Components," EPRI-NP-3444,

May 1984.

8. Final Safety Analysis Report, Fort Calhoun, Docket No. 50-285, November 1991.

9. Final Safety Analysis Report, Arkansas Unit 2, Docket No. 50-368, August 1992.

10. Final Safety Analysis Report, Arkansas Nuclear One, Unit One, Docket No.50-313, August 1992.

11. Final Safety Analysis Report, Crystal River 3, Docket No. 50-302, January 1987.

12. Final Safety Analysis Report, Arkansas Unit One, Docket No. 50-313, August 1992.

13. Control Rod Drive Mechanism - Instruction Manual, Rancho Seco Nuclear Generating Station,

Unit 1, July 1974.

14. Lewis, Harold W., "The Safety of Fission Reactors," Scientific American, March 1980, Vol. 242,

Number 3.

15. Williams, J.T., et al., "Control Rod Drive Mechanism Test Program," B&W Report No. BAW10029A, Rev. 3., August 1976.

16. Three Mile Island One Operations Plant Manual, Control Rod Drive System, March 1989.

17. Riggs, R., "Control Rod Guide Tube Wear in Operating Reactors," NUREG-0641, April 1980.

18. Bailey, W.J., et al., "Fuel Performance Annual Report for 1980," NUREG/CR-2410, December

1981.

19. Bailey, W.J., et al., "Fuel Performance Annual Report for 1982," NUREG/CR-3602, March 1984.

10-1
20. Bailey, WJ., et al., "Fuel Performance Annual Report for 1983 Through 1988," NUREG/CR3950, Volumes 1-6, 1983-1988.

21. Shah, V.N., et al., "Residual Life Assessment of Major Light Water Reactor Components -

Overview," NUREG/CR-4731, November 1989.

22. Weiss, A.J., et al., "Fifteenth Water Reactor Safety Information Meeting - Proceedings,"

NUREG/CP-0091, February 1988.

23. USNRC Standard Review Plan - Rev. 1, NUREG-0800, July 1981.

24. USNRC Standard Review Plan for the Review of License Renewal Applications for Nuclear

Power Plants," Draft NUREG-1299, November 1990.

25. Dufrane, K.F., et al., "Wear Measurements of Nuclear Power Plant Components," EPRI NP-3444,

May 1984.

26. Sipush, P., et al., "Lifetime of PWR Silver-Indium Cadmium Control Rods," EPRI NP-4512,

March 1986.

27. Koppe, R.H., et al., "Characterization of the Performance of Major LWR Components," EPRI

NP-5001, January 1987.

28. "PWR Pilot Plant Life Extension Study at Surry Unit 1: Phase 2," EPRI NP-6232-M, March

1989.

29. USNRC Maintenance Team Inspection Report 50-255/88-20 on 880919-23,1011-14 and 1020 and

Notice of Violation, January 10, 1989.

30. USNRC Maintenance Team Inspection Report 50-335/89-24 and 50-389/89-24 on 89 1002-06,

1016-20, 1030-1103 and 1129, January 10, 1990.

31. "Standard Technical Specifications - Babcock and Wilcox Plants," Draft NUREG-1430, January

1991.

32. "Standard Technical Specifications - Combustion Engineering Plants," Draft NUREG-1432,

January 1991.

33. "Guide to the Collection and Presentation of Electrical, Electronic, Sensing Component, and

Mechanical Equipment Reliability Data for Nuclear Power Generating Stations," IEEE Standard

500-84, 1983.

34. Fuqua, Norman B., Reliability Engineering for Electronic Design, New York: Marcel Dekker,

Inc., 1987.

35. Montague, D.F., et al., "A Common Cause Failure Analysis of the Rodded Scram System of the

Arkansas Nuclear One - Unit 1 Plant," ORNL/TM-10045, October 1986.

10-2
36. Payne, A.C., "Interim Reliability Evaluation Program: Analysis of the Calvert Cliffs Unit 1

Nuclear Power Plant," NUREG/CR-3511, May 1984.

37. Kolb, G.J., "Interim Reliability Evaluation Program: Analysis of the Arkansas Nuclear One -

Unit 1 Nuclear Power Plant," NUREG/CR-2787, August 1982.

38. "Oconee PRA - A Probabilistic Risk Assessment of Oconee Unit 3," NSAC/60, June 1984.

39. Garcia, A.A., et al., "Crystal River 3 Safety Study," NUREG/CR-2515, March 1982.

40. Control Rod Drive Mechanism - Instruction Manual, Rancho Seco Nuclear Generating Station,

Unit 1, July 1974.

41. Subudhi, M., et al., "Improving Motor Reliability in Nuclear Power Plants, Vol. 1: Performance

Evaluation and Maintenance Practices," NUREG/CR-4939, November 1987.

42. Pentecost, L., et al., "Condition Monitoring Utilizing the ECAD Automated Test System at San

Onofre Nuclear Generating Station," Presented at the 17th Inter-ram Conference for the Electric

Power Industry, Hershey, PA, June 1990.

43. Kryter, R.C., et al., "Condition Monitoring of Machinery Using Motor Current Signature

Analysis," Sound and Vibration, September 1989.

44. Meyer, L.C., "Nuclear Plant-Aging Research on Reactor Protection Systems," NUREG/CR-4740,

January 1988.

10-3
Appendix A

Summary of Licensee Event Reports for Combustion Engineering

Control Element Drive System

1980 - 1990

Table A-1

Table A-2

Table A-2.1

Table A-2.2

Table A-3

Table A-4

Table A-5

Table A-6

Table A-7

Cable and Connector LERs

CED Control System LERs

Control Element Drive System (CEDS)

Control Element Drive Mechanism Control System (CEDMCS)

CEDM LERs

Rod Position Indication LERs

Human Error LERs

LERs with Unknown Causes

Miscellaneous LERs

A-1
Table A-1. Cable and Connector LERs

| | Age at

Plant LER No. Failure Failure Description

l ~~~~~~~~~~~~~~(years)

1 St. Lucie 1 335/80-005 4 Loose lead CPC lSv power supply caused CEA to drop

twice.

2 St. Lucie 1 335/80-003 4 Cable problem between containment and refueling

disconnect panel caused position indication to be lost

twice.

3 St Lucie 1 335/81-026 5 Cable problem between containment and refueling

disconnect panel caused position indication to be lost.

4 Millstone 2 336/80-010 5 Erratic position indication caused by intermittent open

circuit in cable or connector from reed switch.

5 Millstone 2 336/80-028 5 Faulty jumper cable between refuel disconnect panel

and reed switch caused dropped CEA and power

reduction.

6 San Onofre 2 361183-124 1 Faulty connector in CEAC position circuit gave

erroneous position indication twice.

7 Waterford 3 38V86-013 1 Loose cable connection caused fluctuating position

indication signal and subsequent reactor trip.

8 Arkansas 2 368/81-010 3 CEAC inoperable due to a loose cable connection.

9 Arkansas 2 368/82-027 4 CEA had erroneous position signals due to faulty RSPT

cable connection.

10 San Onofre 2 361183-098 1 CEAC inopearable due to loose screw connection on

position transmitter input.

11 Waterford 3 382/90-002 5 Two CEAs dropped into core during transfer to hold

bus. Damaged connector between CEDM and

CEDMCS. Reactor trip.

12 San Onofre 2 36V83-096 1 Dirty contacts on CEA timer card caused CEA to slip

and power reduction.

13 Millstone 2 336/81-038 7 CEA could not be withdrawn due to loose connection

between timer module and power switch.

14 Millstone 2 336/82-025 8 CEA could not be withdrawn due to loose connection

between timer moduel and power switch.

15 Arkansas 2 368/81-031 4 CEA dropped on 2 separate occasions due to poor

contact on coil driver card causing power loss and

sluggish upper gripper movement.

16 San Onofre 2 361183-090 1 Slipped CEA due to poor connection on CEDMCS

power switch.

A-2
Table A-2.1 Control Element Drive System (CEDS)

[ I I ~~~~~~~~~~Age at Plant LER No. Failure Failure Description

I I ~~~~~~~~ ~~ _ ~~ _ ~~(years) _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

l1 Calvert Cliffs 1 317/82-045 12 CEA dropped due to erratic upper gripper latch action.

2 Calvert Cliffs 2 318/87-008 11 CEA dropped due to failure of upper gripper power

switch module.

3 Calvert Cliffs 2 318/80-010 4 CEA dropped, power reduced due to faulty timer

l module.

4 St. Lucie 1 335/80-007 4 CEA dropped, power reduced due to failed timer

module.

5 Millstone 2 336/80-013 6 CEA dropped due to failed timer module, power

reduced.

6 Calvert Cliffs 2 318/82-026 6 CEA dropped twice. Timer and upper gripper power

switch replaced. Reactor power reduced.

7 Calvert Cliffs 1 317/81-071 7 Control module failure caused continuous insert signal

resulting in CEA misalignment. Power reduction.

8 Calvert Cliffs 1 317/82-045 8 Erratic upper gripper latch action, dropped CEA,

power reduction. Increased HV power supply.

9 St. Lucie 1 335/80-002 4 Failed 1Sv power supply caused dropped CEA and

power reduction. (Powermate SU-UNI-30A-BV)

10 St. Lucie 1 335/80-023 4 2 CEAs dropped. Same as above.

11 St. Lucie 1 335/80-032 4 CEA dropped twice, timer module and power supply

replaced. Reactor power reduced.

12 St. Lucie 1 335/80-034 4 Failed ISv power supply.

13 St. Lucie 1 335/80-035 4 Failed 1Sv power supply.

14 St. Lucie 1 335180-036 4 Failed 15v power supply.

15 St. Lucie 1 335/80-010 4 Failed 1Sv power supply.

16 St. Lucie 1 335/80-043 4 Failed 1Sv power supply.

17 St. Lucie 1 335/80-045 4 Failed ISv power supply.

18 St. Lucie 1 335/80-046 4 Failed lSv power supply. All power supplies replaced

with original Powermate UNI-88.

19 St. Lucie 1 335/80-048 4 Failed lSv power supply. All power supplies replaced

with original Powermate UNI-88.

A-3
Table A-2.1 Control Element Drive System (CEDS) (Cont'd)

[I I I ~~~~~~~~~~Age at

Plant | LERNo. Failure Failure Description

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ (y e ars) _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

20 St.Lucie 1 335/80-049 4 Failed lSv power supply. All power supplies replaced

with original Powermate UNI-88.

21 St. Lucie 1 335180-050 4 While changing power supplies, voltage spike led to 2

CEAs dropping. Reactor manually tripped.

22 St. Lucie 1 335/80-051 4 Fuse blew in alternate power supply line, CEA

dropped.

23 St Lucie 1 335/80-052 4 Dropped CEA. Actual cause unknown but probably

due to either power supply failure, inadequate

ventilation or power supply mounting.

24 St. Lucie 1 335/81-020 5 Dropped CEA, reduced power, replaced power supply.

25 Millstone 2 336/82-041 8 CEA dropped twice, power reduction, failed 1Sv dc

l________ _ _power supply (Lambda Elec LCD-A-22).

26 Maine Yankee 309/84-001 12 Failed power supply caused dropped rod and reactor

shutdown.

27 St. Lucie 1 335/82-056 6 CEA motion inhibit circuit for CEA out of sequence

deviation scanner failed due to loose relay.

28 St Lucie 1 335/82-055 6 CEA motion inhibit circuit for CEA out of sequence

deviation scanner failed due to loose relay.

29 St. Lucie 1 335/81-030 .5 CEA motion inhibit circuit for PDIL circuit failed due

to failed relay.

30 Millstone 2 336/82-027 8 During CEA surveillance testing, CEA motion inhibit

for all CEA groups became inoperable due to failed

logic chip.

31 Millstone 2 336/82-030 8 CEA motion inhibit interlock did not function due to

l ______________________ faulty operational amplifier.

32 Calvert Cliffs 1 317/81-066 7 While troubleshooting CEDMCS, a logic module which

was inserted caused spurious signals which caused rod

drop due to failed off switch on the control panel.

33 Calvert Cliffs 2 318/83-019 7 Circuit burn of newly installed circuit boards caused

PDIL function to be inoperable.

34 Calvert Cliffs 2 318/83-027 7 Failure of PDIL auctioneering card output semiconductor rendered CEA motion inhibit inoperable.

35 St. Lucie 1 835/90-008 14 Dropped rod on 4 occasions caused by power losses

from fuse not locked in place in the 12v dc logic

circuit. Reactor shutdown.

A4
Table A-2.2 Control Element Drive Mechanism Control System (CEDMCS)

Age at

Plant LER No. Failure Failure Description

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ~(years) _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

I San Onofre 2 361/86-018 4 CEA slipped due to inherent design deficiency in software

which control DC lo CEA coNl.

2 San Onofre 3 36Z184-003 1 Sluggish gripper operation caused CEA to slip and

subsequent reactor scram.

3 Waterford 3 382/86-001 1 Faulty timing module caused CEA drop and reactor trip.

4 San Onofre 2 361/85-031 3 Missing lug nutcaused abnormal energization of power coils,

causing CEA subgroup to drop and reactor trip.

5 San Onofre 2 361/83-114 1 Dropped CEAs due to defective coil driver actuation card.

6 San Onofre 2 361/83-054 1 CEA dropped during surveillance testing due to slow gripper

operation.

7 Arkansas 2 368/82-004 4 CEA dropped due to sluggish upper gripper.

8 Arkansas 2 368/83-040 5 Dropped CEA due to blown fuse.

9 Arkansas 2 368/84-024 6 Dropped CEA due to failure of SCR, power supply fuses,

opto-isolator cards or coil driver cards. Subsequent reactor

scram.

10 San Onofre 3 362185-020 2 Dropped CEA due to blown fuse in the hold bus logic

circuit. Subsequent reactor trip.

11 Waterford 3 382/89-017 4 Unable to withdraw CEA due to control circuitry problems.

Reactor trip.

12 St. Lucie 2 389/89-007 6 Breaker tripped at less than rated current resulting in

dropped CEAs. Power decrease.

13 Palo Verde 3 530/90-004 3 Failure of microchip on optical isolator card resulted in

CEA drop. Reactor shutdown.

14 Palo Verde 3 530/90-006 3 Failure of coil driver actuating logic card caused slipped

CEA. Reactor shutdown.

A-5
Table A-3 CEDM

I I ~~~Age at

Plant ILER No. F alure Failure Description

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ I _ _ _ _ _ _ _ _ _ ~ (y ears) _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

1 Palisades 255/81-049 10 Plant shutdown due to excessive coolant leakage past

_____________________ |_______ improperly installed CEDM seal housing gasket.

2 Palisades 255184-024 13 Failed CEDM seal housing. Plant brought to cold

shutdown.

3 Palisades 255185-006 14 CEA inoperable due to failed motor brake drive

_ package.

4 Palisades 255/86-040 15 14 CEDM seal housings showed cracks due to

contaminant induced stress corrosion cracking.

5 Calvert Cliffs 2 318/80-040 4 Dropped CEA due to omission of CEA venting during

primary system fill.

6 Calvert Cliffs 2 318/80-041 4 Dropped CEA due to omission of CEA venting during

primary system fill. Reduced power.

7 Calvert Cliffs 2 318/80-048 4 2 dropped CEAs due to omission of CEA venting.

8 Calvert Cliffs 2 318/80-056 4 CEA dropped due to omission of CEA venting, power

reduced.

9 Calvert Cliffs 2 318/80-057 4 CEA dropped due to omission of CEA venting during

| primary system fill.

10 Calvert Cliffs 2 318/81-054 5 CEA dropped due to omission of CEA venting during

primary system fill. Reactor power reduced.

11 Millstone 2 336/881008 14 Overheating of the upper gripper coils due to

degradation of CEDM cooling system due to air flow

blockage by boric acid, deposition caused 2 CEAs to

drop. Plant shutdown.

12 Millstone 2 336/884009 14 Dropped CEA and reactor shutdown due to overheating

of upper gripper coil due to lack of cooling caused by

boric acid deposition.

13 San Onofre 2 361183-102 1 Malfunction of upper gripper coil stack caused dropped

CEAX

14 Arkansas 2 368/84-026 6 Shorted upper gripper coil caused CEA to drop.

Reactor trip.

15 Waterford 3 3821864023 1 Failed lower gripper sensor caused dropped CEA and

_____ ____ ____ ____ ___ subsequent reactor trip.

16 St. Lucie 2 389/85-010 2 Failed upper gripper coil led to dropped CEA and

reactor shutdown.

17 Palo Verde 1 528/88-020 3 Lower lift coil ground caused dropped CEA.

A-6
Table A-3. CEDM (Cont'd)

Age at~

Plant LER No. Failure Failure Description 1 ~~~~~~~~~~~~(years)

18 Palo Verde 1 528/88.026 3 Lower lift coil ground caused dropped CEA.

19 Palisades 255/80-020 9 Dirty control relay armature and holddown contactor

interlock stuck resulting in slipped CEA.

20 San Onofre 2 361190.019 7 Degraded CEDM electrical condition. Resulted in

inadvertent ESF actuation and reactor trip.

A-7
Table A4. Rod Position Indication

.~~~ _ Age at

Plant LER No. Failure Failure Description

(years)I

1 Palisades 255/80-031 9 During control rod interlock testing, group of control

rods withdrew from core due to loss of primary and

._____ ______ ______ ________secondary data loggers.

2 Palisades 255/88-025 17 Oscillating power supply caused spurious alarms, which

led operators not to recognize an out of sequence

alarm as valid.

3 Calvert Cliffs 1 317/83-08 9 Failed RSPTs gave erratic position information on 2

CEAs (Electro Mechanics N9027-1).

4 Calvert Cliffs 1 317/83-026 9 Failed RSPT gave erratic rod position information

(Electro Mechanics N9027-1).

5 Calvert Cliffs 1 317/83-036 9 Shorted RSPT overloaded power supply causing loss of

all reed switch position indication. Power reduced.

6 Calvert Cliffs 2 318/82-019 6 Failed RSPT gave intermittent position indication.

7 Calvert Cliffs 2 318/82-022 6 Failed RSPT gave erroneous position indication.

8 Calvert Cliffs 2 318/83-065 7 2 failed RSPTs produced erroneous rod position

information (Electro Mechanics N9027-1).

9 Calvert Cliffs 2 318/83-069 7 Lost all reed switch position information due to failed

metroscope power supply.

10 Calvert Cliffs 2 318/83-075 7 Failed RSPT caused spurious CEA Motion Inhibit

Alarms.

11 St. Lucie 1 335/80-022 4 Due to programming error, DDPS malfunctioned,

causing loss of all backup CEA position indication.

12 St. Lucie 1 335/80-059 4 Due to software error, pulse counting CEA

information lost when DDPS failed.

13 St. Lucie 1 335/81-007 5 DDPS system failure resulting in loss of CEA backup

position indicating system.

14 St. Lucie 1 335/81-002 5 DDPS system failure resulting in loss of CEA backup

position indicating system.

15 St. Lucie 1 335/82-049 6 Pulse counting function for a CEA was deleted by

plant computer.

16 St Lucie 1 835/82-044 6 Pulse counting function for 1 CEA was deleted by

plant computer.

17 Millstone 2 336/80-008 6 Pulse counting position indication system inoperable

due to faulty analog input driver card.

18 Millstone 2 336/81-009 7 C(rcuit card failure in plant computer caused pulse

counting indication system to be inoperable.

A-8
Table A-4 Rod Position Indication (Cont'd.)

I ~~~~Age at

Plant LER No. Failure Failure Description

19 Millstone 2 336/81-037 7 Blown resistor in plant computer 36v power supply

caused pulse counting position indicating system to

become inoperable.

20 San Onofre 2 36V83-011 1 Erroneous rod position indication caused by faulty

isolation amplifier card.

21 San Onofre 2 361/83-087 1 Faulty reed switch produced erroneous rod position

indication.

22 San Onofre 2 361/86-027 4 Loose solder joint on RSPT produced wrong position

indications leading to reactor trip.

23 Arkansas 2 368/80-053 2 Software problem caused failure of CEAC.

24 Arkansas 2 368/80-058 2 Software problem caused failure of CEAC

25 Arkansas 2 368/80-080 2 Data link input card on the optical isolator and failed

test circuit module card caused CEAC failure.

26 Arkansas 2 368/82-005 4 Failure of high level MUX gate card caused CEAC to

show incorrect CEA positions.

27 Arkansas 2 368/83-029 5 Failure of high level MUX card caused CEAC failure.

28 Arkansas 2 368/85-018 7 Erroneous CEA position signals caused by failed field

effect transistors in high level MUX card for CEAC

caused reactor trip.

29 Arkansas 2 368/82-009 4 CEAC failed during excore instrumentation test

30 Arkansas 2 368/82-040 4 CEAC failed, possibly as a result of lighting storm.

31 Waterford 3 382186-009 1 Reactor trip caused by reed switch failure.

32 St. Lucie 2 389/83-047 7 Pulse counting CEA position system inoperable due to

loss of computer.

33 San Onofre 3 362184-024 1 Intermittent failure on computer board caused CEAC

malfunction and reactor trip.

34 Palo Verde 1 528/89-004 4 CEAC inoperable due to failed processor board caused

reactor scram.

35 San Onofre 2 361/84-043 2 Failed power supply in analog CEA position indication

system, caused reactor scram and ESF actuation.

36 San Onofre 2 361/84-O09 2 Faulty CEAC caused spurious position signals which

caused reactor trip.

37 Calvert Cliffs 1 317/81-081 7 Primary CEA position indication lights and analog

I___ ___ __ __ __ system malfunctioned.

A-9
Table A4 Rod Position Indication (Cont'd.)

Plant | LER No. Failure Failure Description

38 San Onofre 2 361183-041 1 CEAC inoperable due to faulty isolation amplifier card

due to overheating in the cabinet.

39 Arkansas 2 368/90-014 12 Failed RSPT transmitted erroneous PC signal to

: CEAC causing reactor trip.

A-10
Table A-5. Human Error

Age at,

Plant LER No. Failure Failure Description

1 Calvert Cliffs 2 318/80-015 4 While changing backup power supply, the prime iSv

CEA power supply was accidently grounded by

electrician causing CEA to drop.

2 Calvert Cliffs 2 318/82-018 6 Electricians mistakenly started work on Unit-2 CEAs

instead of Unit-i, causing CEA drop.

3 Calvert Cliffs 2 318/83-071 7 PDIL was rendered inoperable due to incorrect

setpoints out of tolerance due to personnel error.

4 Calvert Cliffs 2 318/87-008 11 Upper gripper power switch for incorrect CEA was

removed for maintenance, causing dropped CEA and

manual shutdown.

5 St. Lucie 1 335/80-033 4 Technician shorted 1Sv power supply while taking

measurements, causing rod drop and power reduction.

6 St. Lucie 1 335/80-040 4 While performing maintenance on drive control system,

CEA dropped.

7 San Onofre 2 361/88-031 6 Rod drop test procedure did not correctly account for

delay time during rod drop. Some rods may have

exceeded requirements.

8 San Onofre 3 362183-086 1 Technician interrupted -power supply voltage to 23

CEA's during SU test, rendering CEAC inoperable.

9 Arkansas 2 368/85/015 7 While troubleshooting CEAC, electronic transient

caused erroneous PI signal, causing DNBR trip.

Procedures modified.

10 Arkansas 2 368/88-009 10 Rod drop test procedure did not correctly account for

delay time during rod drop.

11 Waterford 3 382/85-051 1 While performing surveillance on CEAC, a incorrect

constant was entered, which caused reactor trip.

12 Waterford 3 382187-012 2 Inadequate procedure allowed both CEACs to be

inoperable, resulting in reactor trip.

13 St. Lucie 2 389/85.006 2 Technician pulled wrong circuit card during

troubleshooting causing 2 dropped CEAs.

14 Palo Verde 2 529/87-003 1 Procedural inadequacy did not require position

verification for all CEA groups upon loss of CEAC.

15 Calvert Cliffs 2 318/87-008 11 Technicians removed power switch module for wrong

CEA causing rod drop and reactor scram.

16 St. Lucie 1 335/86-005 10 During SD, problems were encountered with DDPS,

which required reloading with a magnetic tape

containing incorrect sensitivity factors. The technical

manual did not require verification of sensitivity

factors.

A-11
Table A-S Human Error (Cont'd.)

l | Age at

Plant LER No. Failure Failure Description

l ~~~~~~~~~~~~~~(years)

17 Arkansas 2 368/80-057 2 CEAC inputs to CPC were inoperable due to

personnel error, during maintenance.

18 St. Lucie 2 389/83-074 7 Pulse counting CEA position indicating system

inoperable due to programming error.

19 St. Lucie 1 335/80-038 4 Failed timing module which overheated due to

ventilation fan turned off for maintenance.

20 Waterford 3 383/86-002 1 Dropped CEA due to cabinet cooler switch being

turned off, causing overheating of circuits. Reactor

trip.

A-12
Table A-6. Unknown Cause

I ~~~~Age at

| | Plant | LER No. Failure Failure Description

1 Calvert Cliffs 1 317/82-011 8 CEA dropped, power reduced.

2 Calvert Cliffs 2 318/80-007 4 CEA would not drive down electrically. During

troubleshooting of control circuitry, rod became

.__________ __________________ operable.

3 St. Lucie 1 335/82-051 6 While at full power, CEA dropped for unknown

reason. I

4 St. Lucie 1 335/82-069 6 While at full power, CEA dropped for unknown

reasons.

S St. Lucie 1 335/83-006 7 CEA slipped during CEA exercising, no cause found.

6 St. Lucie 1 335/83-007 7 CEA slipped during normal full power operation. No

cause found.

7 St. Lucie 1 335/82-061 6 During rod positioning to minimize guide tube wear,

rod dropped for unknown cause.

8 SL Lucie 1 335/85-005 9 While at full power, rod dropped for unknown reason.

9 San Onofre 2 361183-155 1 CEAC declared inoperable due to spurious CEA

position indication. No reason found.

10 Ft. Calhoun 285/82-005 9 CEA dropped while at full power for unknown cause.

Power reduction.

11 Millstone 2 336/80-040 6 CEA dropped for unknown reasons while at full power.

12 Millstone 2 336/81-038 7 During routine CEA movement, rod dropped for

. ~~~~~unknown reasons.

13 Millstone 2 336/83-004 9 During routine surveillance, CEA dropped for

_ ~~~~~~~~~~~~~~unknown reasons. Power reduction.

14 Millstone 2 336/83-015 9 While at full power, rod dropped for unknown reasons.

Power reduced.

15 Arkansas 2 368/84-013 6 Reactor trip from full power due to CEA drop for

unknown cause.

16 St. Lucie I 335/81-027 5 CEA dropped for unknown reason.

17 St. Lucie 1 335/81-034 5 CEA dropped during normal CEA exercise for

unknown reason.

18 St. Lucie 1 335/82-028 6 During normal CEA periodic CEA exercise, CEA

dropped for unknown reason;

19 Calvert Cliffs 1 317/80-012 6 CEA dropped while performing routine surveillance

test.

A-13
Table A-6 Unknown Cause (Cont'd.)

J I j~~~~~~~~~Age at

Plant LER No. Falure Failure Description

____ _______________ ~ ~~~~years)

20 Calvert Cliffs 1 317/80-006 6 CEA dropped while performing routine surveillance

test.

21 Calvert Cliffs 1 317/81-039 7 While performing routine maintenance test, CEA

dropped for unknown reason. Power reduction.

22 Calvert Cliffs 1 317/82-036 8 During start up tests, rod dropped for unknown reason.

23 Calvert Cliffs 2 318/80-00 4 CEA dropped during performance of routine test for

unknown reason.

24 Calvert Cliffs 2 318/83-076 7 During routine surveillance, tests over 30 day period, 2

CEAs have dropped into core with subsequent power

reduction. No cause found.

25 St. Lucie 2 389/87-005 4 While at 100%o power, 2 CEAs dropped into core for

no apparent reason. Reactor shutdown.

A-14
Table A-7. Miscellaneous

l |Plant LER No. Failure Failure Description

____ ________________ _________(years)

- Millstone 2 336/83-026 9 Top nozzle damage caused guide tube deformation.

2 Calvert aiffs 2 318/83-060 7 Water from overflowing control room toilet seeped

into cable spreading room where it shorted out coil

power programmer components.

3 Arkansas 2 368/90-005 10 CEA stuck due to foreign material wedged between

CEA finger and ID of FA guide tube. Reactor

shutdown.

4 Maine Yankee 309/904-04 18 Failure of CEA center finger and caps, caused B4C

pellets to fail into GT preventing CEA full insertion.

Subsequent testing showed 2 additional CEA missing

center finger end caps and 6 with cracks.

A-15
Appendix B

Summary of Licensee Event Report for Babcock & Wilcox

Control Rod Drive System

1980 - 1990

Table B-1 Cable and Connector LER's

Table B-2 CRD Control System LERs

Table B-3 CRDM LERs

Table B-4 Rod Position Indication LERs

Table B-5 Human Error LERs

Table B-6 Potential Loose Parts LERs

Table B-7 Unknown Failure Cause LERs

B-1
Table B-i. Cable and Conector LER's

l | Age at

Plant LER No. Failure Failure Description

l____________ 1 _____________ (years)

1 Oconee 1 269/80-027 Power reduction resulted from dropped rod. C-phase on stator

opened due to loose connector.

2 Oconee 1 269/87-010 14 Control power lost to control rods due to loose solder joint on

a control rod sequencing card. Resultant reactor trip.

3 Davis Besse 346/81-012 4 Electrical noise from faulty penetration module caused faulty

l________ position indication.

4 Davis Besse 346/81-019 4 Faulty rod position indication caused by faulty API pentration

module.

5 Davis Besse 346/81-061 4 Erratic API signals due to faulty penetration module.

B-2
Table B-2. CRD Control System LER's

Age at j Plant LER No. Failure Failure Description

______________ ~~~~(years)

1 Arkansas 1 313/83-024 9 Control rod group misalignment due to out movement

restriction resulting in pole slippage.

2 Arkansas 1 313/88-003 14 Control rod group dropped due to malfunction of power

sequencer programmer. Reactor trip.

3 Davis Besse 346/80-023 3 Failed 24 VDC power supply in programmer controller in

group SCR supply cabinet caused improper rod movement.

Reactor trip.

4 Davis Besse 346/82-011 5 Control rod drop due to blown fuse in transfer switch

module. Power reduction.

5 Davis Besse 346/83-014 6 Rod drop due to blown fuse in transfer switch module.

6 Davis Besse 346/83-054 6 RPI inoperable due to failed phase of motor programmer.

7 Davis Besse 346/83-068 6 Rod drop due to failed fuse in motor programmer. Power

reduction.

8 Davis Besse 346/83-071 6 Faulty motor power return SCR gate drive circuit caused

RPI failure. Power reduction.

9 Davis Besse 346/84-001 7 Faulty logic card rendered APSRAs inoperable. Reactor

power reduction.

10 Davis Besse 346/83-062 6 Reactor trip due to failed programmer board in CRDCS

cabinets. Excessive dust in cabinets from concrete work.

11 Oconee 3 287/90-001 16 Dropped CR group due to power supply failure. Reactor

_ ___________ trip.

12 Crystal River 302/89-091 12 Two CR groups simultaneous withdraw due to failed realy

in CR transfer logic.

13 Ocnnee 3 287/90-003 16 Dropped CR group due to failed solid state programmer.

Reactor trip.

B-3
Table B-3. CRDM LER's

1 1 Age at

Plant LER No. Failure Failure Description

(years)

1 Arkansas 1 313/82-020 8 Dropped rod due to stator failure. Reactor trip.

2 Davis Besse 346/81-038 4 Unable to withdraw rod due to fractured leaf spring on antirotational device. Reactor scram.

3 Davis Besse 346/85-006 8 Failed leaf spring setscrew prevented disengagement of lead

screw. Failed TS drop time requirement.

B-4
Table B4. Rod Position Indication LER's

Age at

Plant LER No. Failure Failure Description

l________________ ________________ (years)

1 Crystal River 302/82-035 5 Reed switch failure rendered API system inoperable.

2 Crystal River 302/83-006 6 API system inoperable due to failed reed switch.

3 Crystal River 302/83-061 6 API system inoperable due to failed reed switch.

4 Crystal River 302/85-023 8 Low voltage failure rendered RPI system inoperable. Reactor

trip.

S Crystal River 302/86-011 9 Relay failure in RPI circuitry rendered system inoperable.

6 Davis Besse 346/80-004 3 Lost rod position indication due to failed reed switch caused

l________ by excessive high temperature.

7 Davis Besse 346/80-013 3 API malfunction caused by blown fuse in power supply.

l_______ Delayed SU.

8 Davis Besse 346/80-015 3 API system inoperable due to reed switch failure caused by

excessive high temperature.

9 Davis Besse 346/80-025 3 Excessive high temperatures caused reed switch failure. API

I inoperable.

10 Davis Besse 346/80-015 3 Reed switch failure caused by excessive high temperature. API

_ _______________ _______________ _________ inoperable.

B-5
Table B-S. Human Error LER's

Failure Description

Rod drop due to maintenance error while performing work on

CRDCS. Power reduction.

B-6
Table B-6. Potential Loose Parts LER's

- T - ~~~~~~Age atT

Plant LER No. Failure Failure Description

l _ l__ 1 _ _ _ _ (years) ll

1 Oconee 1 269/80-015 7 Broken fuel assembly holddown springs.

2 Oconee 1 269/81-011 8 Broken lower thermal shield bolts.

3 Oconee 1 269/83-013 10 Broken fuel assembly holddown springs.

4 Oconee 2 270/82-002 9 Failed lower thermal shield bolts.

S Oconee 3 287/82-007 8 Broken fuel assembly holddown springs.

6 Oconee 3 287182-008 8 Failed bolts for core barrel thermal shield.

7 Oconee 3 287/87-001 13 Debris from failed reactor coolant pump lodged in fuel

assembly.

8 Crystal River 302/80-019 3 Failed fuel assembly holddown springs.

9 Davis Besse 346/80-040 3 Twenty broken fuel assembly holddown springs.

10 Davis Besse 346/88-015 11 Debris found in reactor vessel prior to refuel.

B-7
Table B-7. Unknown Cause

11=1 - ~~~~~~~~Age at Plant LER No. Failure Failure Description 1=1 ___ ___ J ~~~~~~(years)______

1 J Oconee 2 270/89-007 16 Rod dropped for unknown reason. Unit operated in

L ______________ I I I unanalyzed condition in violation of TS. Unit shutdown.

B-8
APPENDIX C

Industry Survey of Current CEDM Operating Experience,

Inspection, and Maintenance Practices

C-1
NPAR QUESTIONNAIRE

Combustion Engineering Control Element Drives

Background and Objective

This questionnaire requests information from utility engineers concerning maintenance and

operating experience with Combustion Engineering control element drive systems (both rack and pinion

and magnetic jack designs). Details on inspections and preventive maintenance, as well as for repairs

and modifications, will be of assistance to the NRC Nuclear Plant Aging Research (NPAR) program.

A goal of the program is to identify inspection and maintenance methods that assure detection of aging

effects prior to loss of safety functions.

Questionnaire Organization

The questionnaire is divided into three main sections. The first section contains detailed

operating and experience questions for CED components both inside and outside of containment. The

second section addresses general system tests. The final section contains general type questions, and

requests that respondents provide information copies of any relevant system descriptions, maintenance

schedules, procedures or calibrations which would supplement questionnaire responses.

C-2
Table Of Contents

I. CEDM System Inspections, Maintenance, Modifications

A. CED System Components Inside Containment

1. Control Element Assemblies

2. Latches and Drive Shaft

3. Pressure Housings

a. Seal Welds

b. Vent Valve

c. Other Pressure Housing Features

4. Coil Stack Assembly

a. Coil Stack Components

b. CEDM Cooling System

5. Cables and Connectors

6. Position Indication System

7. Other

B. CED System Components Outside Containments

1. Cables, Connectors, Terminations

2. Power Supplies

3. Control and Logic Cabinets

4. Other

II. CEDM System Tests

A. Control Element Position Verification Test

B. Control Element Drop Time Test

C. Other Tests

III. General

A. Control Element Exercising

B. Reliability and Trend Analysis

C. Overall Ranking

D. Supporting Documents

C-3
NPAR QUESTIONNAIRE

CE Control Element Drives

Plant Name:

Plant Contact: Phone No.: __ _

Please provide the requested information. If additional sheets are required, please clearly identify

appropriate section.

What CEDM Model No. is installed at your plant? (If more than one, please list)

How many such CEDM's are installed at your plant?

How long has it been since such CEDM's were initially installed?

How much operating time has accumulated for these CEDM's (i.e. hours,

months/years of operation)?

Which CEDM design is used at your plant?

Rack and Pinion Magnetic Jack

I. CEDM SYSTEM INSPECTIONS. MAINTENANCE. MODIFICATIONS

A. CED System Components Inside Containment

1. Control Element Assemblies

Please briefly describe the full length CEA's used in your plant (number, poison material, unique design

features, corner rod/center rod differences).

Please briefly describe the part length CEA's used in your plant (number, poison material, unique design

features).

C4
Please summarize any indications of unusual CEA wear, crud buildup or any other indication of

interference with guide tubes, upper guide structure assembly, or CEA shrouds. When and how were

the indications discovered? Where were the indications?

What inspections are performed on control element, guide tubes, CEA shrouds, etc. How many

components are inspected and with what frequency (each refueling outage, etc.)?

How are your plants control elements attached to the spider?(by a nut, or threaded into spider arm)

2. Latches and Drive Shaft

Are inspections/preventive maintenance performed on:

Latches (wear, crud buildup, etc.)? Yes_ No_

If yes, how often?

Describe inspection or PM:

Drive Shaft (wear, crud buildup, etc.)? Yes_ No

If yes, how often?_

Describe inspection or PM:

Summarize corrective maintenance or replacements of latches or drive shafts (component, when, reason).

If used, are the CEDM's for the part length CEA's modified to prevent insertion

during a scram? Yes_ No_

3. Pressure Housing

a. Seal Welds

In response to the following questions, please consider both welds individually.

1. Motor Assembly Housing to Reactor Vessel Head

2. Upper Pressure Housing to Motor Assembly Housing

C-5
Are the following inspections performed on the seal welds?

NDE? Yes-No Frequency

Description of tests:

Hydrostatic? Yes_ No Frequency

Please describe any other inspections or tests which are performed (indicate which weld and frequency).

How many seal welds have been cut and rewelded?

Which Frequency

Reason

Has the design of the seal weld been modified? Yes No_

Which

Describe modification:

What type of seal is used (Omega, etc.)?

Which Has the seal design changed?Yes _ No_

If yes, please describe

Does your plant have a seal inspection program? Yes NoFrequency

b. Vent Valve

C-6
Are the vent valves opened and closed for any maintenance operations?

Yes _ No _ Frequency

Reason:

Are the vent valves hydrostatically tested? Yes_ No_

Frequency

Has any vent valve leakage (or indications of) been noted?

Yes _ No _ How many

Describe (severity, effects):

Summarize what corrective maintenance or replacement of valves has been performed at your plant:

Has the vent valve design changed? Yes_ No_

Please describe:

c. Other Pressure Housing Features

Please briefly discuss any other inspections, maintenance, replacements or modifications not addressed

above:

4. Coil Stack Assembly

a. Coil Stack Components

Are the following inspections or tests performed on the coils?

Physical inspection? Yes-No Frequency

Features examined:

C-7
Electrical tests? Yes_ No _ Frequency

Describe test procedure and equipment:

Summarize corrective maintenance or replacements of coils or assembly wiring (component, when,

reason):

Has the coil stack assembly (or individual coils) design or operating parameters (voltage, current) been

modified? Yes_ No_

When

Please describe:

b. CEDM Cooling System

Is the performance of the cooling system monitored during operation?

Yes _ No_

Describe (parameters, alarms):

Are inspections, test, or preventive maintenance performed on the cooling system? Yes_ No_

Please describe:

Has the cooling system been modified? Yes_ No_

Describe (include effect of change on coil cooling):

C-8
Are the coils and coil stack assembly temperatures monitored? Yes_ No_

How

Are the temperatures recorded? Yes_ No_

What are representative coil temperaturese

Location

(estimate if not available)

Have there been any incidents of loss of coil cooling? Yes_ No_

Describe (number, duration, effect on coils and CEDM).

5. Cables and Connectors

Are the following inspections, tests or preventive maintenance performed?

Insulation degradation or wear? Yes_ No

Frequency

Describe (visual, electrical test):

Connector pin condition (corrosion, loose)? Yes_ No_

Frequency

Connector watertight seal? Yes_ No_

Frequency

Other tests (description, frequency)

Please summarize corrective maintenance or replacement of connectors or cables (component, when,

reason):

If replaced please indicate original and replacement model No. and manufacturer:

Original:

Replacement:

Have cables or connectors been modified? Yes_ No_

C-9
* Describe:

What is the reactor head area ambient temperature during operation?

(Actual or estimated)

6. Position Indication System'

Are the following tests or inspections performed on the

#1: Pulse Counting CEA Position Indication System

#2: Reed Switch CEA Position Indication System

Physical inspection? Yes_ No _ Which system?_

.

Frequency

Describe (feature examined, inspection method):

Electrical tests? Yes _ No_ Which system?_

Frequency

Describe (procedure, test equipment):

Other tests or inspections? Yes No_

Whiclhystem Frequency_

Describe (tests, equipment used):

Please provide summary of corrective maintenance or replacements (component, system, when, reason):

Has the design of the RPI systems been modified? Yes_ No_

System:

C-10
Describe:

Has there been any unusual calibration or system drift problems?

Yes No System

Describe:

7. Other

Please summarize any inspections/preventive maintenance of CEDM components inside containment that

were not addressed above (component, frequency, description)

Please summarize any modifications or corrective maintenance of CEDM components inside containment

which were not addressed above (components, when, reason):

Please summarize any incidents or occurrences which affected CED system performance (i.e. spills,

impacts, work on other systems, etc.):

B. CED System Components Outside Containment

1. Cables, Connectors, Terminations

Please describe inspection and preventive maintenance (frequency, activity):

C-11
Please summarize corrective maintenance performed (type of repair, component):

2. Power Supplies

Please describe inspection, periodic component testing, and preventive maintenance (Frequency, which

power supply):

Are the power supplies monitored during operation?

Yes_ No_

Describe:

Please summarize corrective maintenance performed (type or repair, number):

3. Control and Logic Cabinets

Describe Inspection, periodic component testing and preventive maintenance.

Description (component, activity, frequency):

Are the components monitored during operation? Yes_ No_

Describe:

Please summarize corrective maintenance performed (number, type of repair):

C-12
Is the ambient temperature of the cabinets monitored?

Yes No Method

What is the average cabinet ambient temperature?__

Is this measured or estimated?_

Has there been any significant logic or control changes? Yes_ No

Describe (what, when, components changed)

4. Other

Please discuss briefly any inspection, periodic component testing, and preventive maintenance of CEDM

components outside of containment that were not described above (component, frequency, description

of activity):

Please briefly discuss any modifications to the original CEDM system located outside of containment

which were not discussed above (component, when):

Please summarize other CEDM system parameters which are monitored during operation not discussed

above:

II. CEDM SYSTEM TESTS

A. Control Element Position Verification Test

Is this test performed? Yes_ No_ Frequency

Is this a Tech. Spec. requirement? Yes_ No_

What parameters are monitored?

B. Control Element Drop Time Test

Is this test performed? Yes_ No_ Frequeni I

C-13
Is this a Tech. Spec. requirement? Yes_ No_

What parameters are monitored?

C. Other tests

What tests, other than those above, are routinely performed? (test, frequency, parameters monitored)

What tests are performed on the CEDM pressure housing to ensure integrity? (test, frequency)

III. General

A. Control Element Exercising

During operation, are individual control elements periodically exercised if they have not been moved in

normal operation for some period? Yes_ No_

Frequency

Procedure

B. Reliability and Trend Analysis

Is a reliability and trend analysis in place for the CEDM system?

Yes_ No_

What key parameters are evaluated?

Does your program track other plants' experience? Yes No_

What is the information source used for your trend analysis?

C. Overall Ranking

In your opinion, what three parameters monitored, inspections performed, or tests conducted are most

important to ensure operational readiness of the CRD system?

1.

2.

3.

C-14
D. Supporting Documents

1. Please provide an information copy if the following to assist in understanding the CEDM operation

and maintenance.

Calibration and functional test procedures for rod position and CED control systems. Attached?

Yes _ No_

Control Element drop time test procedure.

Attached? Yes_ No_

CEDM inspection procedure.

Attached? Yes_ No_

CED System Description.

Attached? Yes_ No_

2. Please attach any computer printouts which describe preventive maintenance practices and schedules,

component calibration intervals, and/or corrective maintenance that has been performed, if possible.

3. Please provide any additional information which may help in understanding the responses to this

questionnaire, or which you feel is pertinent to the study of CED system aging.

C-15
Appendix D

Industry Survey of Current CRDM

Operating Experience, Inspection, and

Maintenance Practices

D-1
Background and Objective

This questionnaire requests information from utility engineers concerning maintenance and

operating experience with Babcock & Wilcox control rod drive systems. Details on inspections and

preventive maintenance, as well as for repairs and modifications, will be of assistance to the NRC Nuclear

Plant Aging Research (NPAR) program. A goal of the program is to identify inspection and

maintenance methods that assure detection of aging effects prior to loss of safety functions.

Questionnaire Organization

The questionnaire is divided into three main sections. The first section contains detailed

operating and experience questions for CRD components both inside and outside of containment. The

second section addresses general system tests. The final section contains general type questions, and

requests that respondents provide information copies of any relevant system descriptions, maintenance

schedules, procedures or calibrations which would supplement questionnaire responses.

D-2
Table Of Contents

I. CRDM System Inspections, Maintenance, Modifications

A. CRD System Components Inside Containment

1. Control Rod Assemblies

2. Roller Nuts and Leadscrew

3. Motor Tube

a. Welds

b. Flange Seals

c. Vent Valve

d. Other Motor Tube Features

4. Coil Stack Assembly

a. Coil Stack Components

b. CRDM Cooling System

5. Cables and Connectors

6. Position Indication System

7. Other

B. CRD System Components Outside Containments

1. Cables, Connectors, Terminations

2. Power Supplies

3. Control and Logic Cabinets

4. Other

II. CRDM System Tests

A. Control Rod Position Verification Test

B. Control Rod Drop Time Test

C. Other Tests

III. General

A. Control Rod Exercising

B. Reliability and Trend Analysis

C. Overall Ranking

D. Supporting Documents

D-3
NPAR QUESTIONNAIRE

B&W Control Rod Drives

Plant Name:

Plant Contact: Phone No.:(_)

Please provide the requested information. If additional sheets are required, please clearly identify

appropriate section.

What CRDM (Type A, B, or C) is installed at your plant? (If more than one, please

list)

How many such CRDM's are installed at your plant?

How long has it been since such CRDM's were initially installed?

How much operating time has accumulated for these CRDM's (i.e. hours,

months/years of operation)?

I. CRDM SYSTEM INSPECTIONS. MAINTENANCE. MODIFICATIONS

A. CRD System Components Inside Containment

1. Control Rod Assemblies

Please briefly describe the control rod assemblies used in your plant (number, poison material etc.).

Please briefly describe the axial power shaping rod assemblies used in your plant (number, poison

material, etc.).

D-4
Please summarize any indications of unusual CRA wear, crud buildup or any other indication of

interference with fuel assembly guide tubes, CRA guide tube assemblies. When and how were the

indications discovered? Where were the indications?

What inspections are performed on control rods, axial power shaping rods, guide tubes, CRA guide tubes,

etc. How many components are inspected and with what frequency (each refueling outage, etc.)?

2. Roller Nuts and Leadscrew

Are inspections/preventive maintenance performed on:

Roller Nuts (wear, crud buildup, etc.)? Yes_ No_

If yes, how often?_

Describe inspection or PM:

Leadscrew (wear, crud buildup, etc.)? Yes_ No_

If yes, how often?_

Describe inspection or PM:

Male Bayonet Coupling (wear, etc.)? Yes_ No_

Describe inspection and any significant findings:

How often

Summarize corrective maintenance or replacements of roller nuts or lead screw (component, when,

reason).

D-5
3. Motor Tube

a. Welds

Are the following inspections performed on the motor tube welds?

NDE? Yes_ No _ Frequency

Description of tests:

Hydrostatic? Yes_ No Frequency

Please describe any other inspections or tests which are performed (indicate which weld and

frequency).

How many motor tube welds have been cut and rewelded?_ _

Which Frequency_

Reason:

b. Flange Seals

What type of gaskets are used to seal the CRDM to the reactor head? (size, number, material, etc.)

How often does your plant replace these gaskets?

Are gasket inspections part of a normal PM procedure? Yes_ No_

Howmanyareinspected?

Has the design of these seals changed? Yes No__ How

D-6
c. Vent Valve

Are the vent valves opened and closed for any maintenance operations?

Yes No Frequency_

Reason:

Are the vent valves hydrostatically tested? Yes No

Frequency

Has any vent valve leakage (or indications of) been noted?.

Yes _ No _ How many

Describe (severity, effects):

Summarize what corrective maintenance or replacement of valves has been performed at your plant:

Has the vent valve design changed? Yes_ No

Please describe:

d. Other Motor Tube Features

Please briefly discuss any other inspections, maintenance, replacements or modifications not addressed

above:

4. Stator Assembly

a. Stator Components

Are the following inspections or tests performed on

Physical inspection? Yes_ No_ Frequency

the stator coils?

D-7
Features examined:

Electrical tests? Yes_ No _ Frequency_ _

Describe test procedure and equipment:

Summarize corrective maintenance or replacements of coils or assembly wiring (component, when,

reason):

Has the stator assembly (or individual coils) design or operating parameters (voltage, current) been

modified? Yes_ No

When

Please describe:

b. CRDM Cooling System

Is the performance of the cooling system monitored during operation?

Yes_ No__

Describe (parameters, alarms):

Are inspections, test, or preventive maintenance performed on the cooling system? Yes _ NoPlease describe:

Has the cooling system been modified? Yes_ No

D-8
Describe (include effect of change on coil cooling):

Are the stator assembly temperatures monitored? Yes_ No___

How

Are the temperatures recorded? Yes_ No

What are representative coil temperatures?_

Location

(estimate if not available)

Have there been any incidents of loss of coil cooling? Yes_ No_

Describe (number, duration, effect on coils and CRDM).

5. Cables and Connectors

Are the following inspections, tests or preventive maintenance performed?

Insulation degradation or wear? Yes_ No_

Frequency

Describe (visual, electrical test):

Connector pin condition (corrosion, loose)? Yes_ No_

Frequency

Connector watertight seal? Yes _ No

Frequency

D-9
Other tests (description, frequency)

Please summarize corrective maintenance or replacement of connectors or cables (component, when,

reason):

If replaced please indicate original and replacement model No. and manufacturer:

If replaced please indicate original and replacement model No. and manufacturer:

Original:

Replacement:

Have cables or connectors been modified? Yes_ No_

Describe:

What is the reactor head area ambient temperature during operation?

(Actual or estimated)

6. Position Indication System

Are the following tests or inspections performed on the

#1: Relative Position Indication System

#2: Absolute Position Indication System

Physical inspection? Yes_ No Which system,

Frequency

Describe (feature examined, inspection method):

Electrical tests? Yes_ No_

Frequency

Which system?

D-10
Describe (procedure, test equipment):

Other tests or inspections? Yes_ No_

Which system Frequency

Describe (tests, equipment used):

Please provide summary of corrective maintenance or replacements (component, system, when, reason):

Has the design of the reed switch system been modified? Yes_ No_

Describe:

Can the absolute system operate with failed reed switches? Yes_ No_

What type of reed switches does your plant use? (i.e. R4C)

Has there been any unusual calibration or system drift problems?

Yes _ No_ System

Describe:

7. Other

Please summarize any inspections/preventive maintenance of CRDM components inside containment that

were not addressed above (component, frequency, description)

D-11
Please summarize any modifications or corrective maintenance of CRDM components inside containment

which were not addressed above (components, when, reason):

Please summarize any significant occurrences with the CRA coupling tool (design changes, regular

maintenance, inspections, operating difficulties, etc.):

Please summarize any incidents or occurrences which affected CRD system performance (i.e. spills,

impacts, work on other systems, etc.):

B. CRD System Components Outside Containment

1. Cables, Connectors, Terminations

Please describe inspection and preventive maintenance (frequency, activity):

Please summarize corrective maintenance performed (type of repair, component):

2. Power Supplies

Please describe inspection, periodic component testing, and preventive maintenance (Frequency, which

power supply):

D-12
Are the power supplies monitored during operation?

Yes _ No Describe:

Please summarize corrective maintenance performed (type or repair, number):

3. Control and Logic Cabinets

Describe Inspection, periodic component testing and preventive maintenance.

Description (component, activity, frequency):

.

Are the components monitored during operation? Yes_ No_

Describe:

Please summarize corrective maintenance performed (number, type of repair):

Is the ambient temperature of the cabinets monitored?

Yes No Method

What is the average cabinet ambient temperature?_

Is this measured or estimated?_

Has there been any significant logic or control changes? Yes_ No_

Describe (what, when, components changed)

D-13
4. Other

Please discuss briefly any inspection, periodic component testing, and preventive maintenance of CRDM

components outside of containment that were not described above (component, frequency, description

of activity):

Please briefly discuss any modifications to the original CRDM system located outside of containment

which were not discussed above (component, when):

Please summarize other CRDM system parameters which are monitored during operation not discussed

above:

II. CRDM SYSTEM TESTS

A. Control Rod Position Verification Test

Is this test performed? Yes_ No _ Frequenc

Is this a Tech. Spec. requirement? Yes_ No_

What parameters are monitored?_-

B. Control Rod Drop Time Test

Is this test performed? Yes_ No _ Frequenc

Is this a Tech. Spec. requirement? Yes_ No_

What parameters are monitored?

C. Other tests

What tests, other than those above, are routinely performed? (test, frequency, parameters monitored)

What tests are performed on the CRDM motor tube to ensure integrity? (test, frequency)

D-14
III. General

A. Control Rod Exercising

During operation, are individual control rods periodically exercised if they have not been moved in

normal operation for some period? Yes_ No_

Frequency

Procedure

B. Reliability and Trend Analysis

Is a reliability and trend analysis in place for the CRDM system?

Yes *No

What key parameters are evaluated?

Does your program track other plants' experience? Yes_ No

What is the information source used for your trend analysis?

C. Overall Ranking

In your opinion, what three parameters monitored, inspections performed, or tests conducted are most

important to ensure operational readiness of the CRD system?

1.

2.

3.

D. Supporting Documents

1. Please provide an information copy if the following to assist in understanding the CRDM operation

and maintenance.

Calibration and functional test procedures for rod position and CRD control systems. Attached?

Yes _ No_

Control Rod drop time test procedure.

Attached? Yes_ No_

CRDM inspection procedure.

Attached? Yes_ No_

CRD System Description.

Attached? Yes_ No_

D-15
2. Please attach any computer printouts which describe preventive maintenance practices and schedules,

component calibration intervals, and/or corrective maintenance that has been performed, if possible.

3. Please provide any additional information which may help in understanding the responses to this

questionnaire, or which you feel is pertinent to the study of CRD system aging.

D-16
NRC FORM 335 U.S. NUCLEAR REGULATORY COMMISSION 1. REPORT NUMBER

(2-89N (Assigned by NFIC. Add Vol., Supp., Rev.,

NRCM 1102, and Addendum Numbers, It any.)

3201,3202 BIBLIOGRAPHIC DATA SHEET NUREG/CR-5783

(See instructions on the reverse) BNL-NUREG-52299

2. TITLE AND SUBTITLE

Aging Assessment of the Combustion Engineering and 3 DATE REPORTPUBLISHED

Babcock & Wilcox Control Rod Drives MONTH R YEAR

January 1993

4. FIN OR GRANT NUMBER

A3270

5. AUTHOR(S) 6. TYPE OF REPORT

E. Grove, W. Gunther Technical

7. PERIOD COVERED (inclusiveDates)

8. P E RFO RM ING O R GAN I ZAT I O -NAM E AN D AD D RESS [if NRtC, provide Division, Office or Region, U.S. Aluclear Regulatory Commission, and mailing address; if con tractfor, provide name and mailing address.)

Brookhaven National Laboratory

Upton, NY 11973

9. SPONSOR ING ORGANIZATION - NAME AND ADDA ESS (if NRC, type "Same as above": if contractor. provide NRC Division. Office or Region, U.S. Nut/ear Regulatory Commission.

9. SPONSO RI NG OR GAN I ZATION -NAM E AN D ADD RESS (If NRC, type 'iSame as abow-, if contractorprovide NRC Division, Of fice or Region, U.S. Nuclear Regulatory Commission,

and mailing address.)

Division of Engineering

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555

10. SUPPLEMENTARY NOTES

11. ABSTRACT (20oo words or les)

The effects of aging upon the Babcock & Wilcox (B&W) and Combustion Engineering (CE)

control rod drive systems have been evaluated. For this study, the CRD system boundary included the

control rod assemblies, guide tubes, control rod drive mechanism, control system components, rod

position indication components, and cooling system. Detailed operation experience data for 1980 to 1990

was evaluated to identify the predominant failure modes, causes, and effects. The results of this

evaluation, along with an assessment of component material and operating environment, lead to the

conclusion that both the B&W and CE CRD systems are susceptible to age degradation. Failures of the

CRD system have resulted in significant plant effects including power reductions, plant shutdowns,

scrams, and ESF actuations.

Information on current plant system inspection and maintenance practices were obtained from

two B&W plants, and four CE plants through an industry survey. The results of this survey indicate that

some plants have modified the system, replaced components, and established preventive maintenance

programs, some of which effectively address the aging issue, while others do not. The potential

application of some advanced monitoring inspection techniques are discussed.

12. KEY WORDS/DESCRI PTORS (LiSt words or phrases that will assist researchers in locating the report,' 13. AVAI LABI LITY STATEMENT

Control Rod Drives - Aging, PWR Type Reactors - Aging, Reactor Unlimited

Control Systems - Control Rod Drives, Reactor Components - failures, 14.SECURITY CLASSIFICATION

Reactor Safety, failure mode analysis, thermal degradation, wear, (ThisPagel -

service life, scram, Reactor Maintenance, Reactor Monitoring Systems Unclassified

inspection (This Report)

Unclassified

15. NUMBER OF PAGES

16. PRICE

NRC FORM 335 (2-89)
Federal Recycling Program
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UNITED STATES

NUCLEAR REGULATORY COMMISSION

WASHINGTON, D.C. 20555-0001

SPECIAL FOURTH-CLASS RATE

POSTAGE AND FEES PAID

USNRC

PERMIT NO. G-67

OFFICIAL BUSINESS

PENALTY FOR PRIVATE USE, $300