TECHNICAL REPORT

TR-3270-9-90

AN OPERATIONAL ASSESSMENT OF THE

BABCOCK & WILCOX AND COMBUSTION ENGINEERING

CONTROL ROD DRIVES

September 1990

Prepared by

E. Grove and W. Gunther

Engineering Technology Division

FIN A-3270

DEPARTMENT OF NUCLEAR ENERGY, BROOKHAVEN NATIONAL LABORATORY

UPTON, NEW YORK 11973

l)1111

( t1 I

Prepared for the U.S. Nuclear Regulatory Commission

Office of Nuclear Regulatory Research

Contract No. DE-AC02-76CH00016

v . . .. .. - . ..
NOTICE

This report was prepared as an account of work sponsored by an agency of the United

StatesGovernment. Neither the United States Govemment nor any a,, ey thereof, or

any of their employees, makes any warranty, expressed or implied, or assumes any

legal liability or responsibility for any third party's use, or the results of such use, of

any information, apparatus, product ofprocess disclosed in this report, or represents

that its use by such third party would not infringe privately owned rights.

The views expressed in this report are not necessarily those of the U.S. Nuclear

Regulatory Commission.
Technical Report

TR-3270-9-90

AN OPERATIONAL ASSESSMENT OF THE

BABCOCK & WILCOX AND COMBUSTION ENGINEERING

CONTROL ROD DRIVES

September 1990

Prepared by

E. Grove and W. Gunther

Engineering Technology Division

Department of Nuclear Energy

Brookhaven National Laboratory

Upton, New York 11973

Prepared for

U.S. Nuclear Regulatory Commission

Washington, DC 20555

FIN A-3270
ABSTRACr

An engineering study describing progress towards completion of a Phase I aging assessment for

the Babcock & Wilcox and Combustion Engineering control rod drive systems has been completed.

This study, along with a system Failure Mode and Effect Analysis and a detailed review of utility

maintenance practices and procedures will complete the Phase I aging assessment. This study is being

performed as part of the Nuclear Plant Aging Research (NPAR) program. The goals of this program

are to assess the impact of aging on plant safety and to develop effective mitigating actions.

Commercially available operating experience databases were reviewed to identify failed

components and resultant plant operating effects for the 1980-1989 time period. Age related failures

for both systems were identified which resulted in significant plant effects including dropped rods,

power reduction and shutdown. System susceptibility to external influences such as maintenance errors

and the operating environment was also shown.

iii
CONTENTS

Page

Abstract ............................. I.............. iii

List of Figures ........................................... vii

List of Tables ........................................... viii

Acknowledgement ........................................... ix

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

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

1.2 Objectives .......................................... 1-3

1.3 System Boundary ..................................... 1-4

1.4 Analysis Methodology and Report Format .................... 1-4

1.5 Future Work .................... .... ............ 1-7

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 .. 2-1

2.3.1 CEDM Design Variations .2-6

2.4 CEA Position Indication System. 2-6

2.5 Control Element Drive Mechanism Control System .2-9

2.6 CEDM Cooling System .......... ....................... 2-9

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 .-. 3-1

3.4 Control Rod Drive Control System .. 3-7

3.5 Control Rod Position Indication .. 3-9

3.6 Control Rod Drive Cooling Water System .. 39

v
CONTENTS (Cont'd)

4. OPERATIONAL EXPERIENCE - B&W AND CE ................... 4-1

4.1 Introduction ........................................ 4-1

4.2 Dominant Component Failures ........................... 4-4

4.2.1 Combustion Engineering ........................... 4-4

4.2.2 Babcock & Wilcox ............................... 4-4

4.3 Failure Effect .................... ................... 4-6

4.3.1 Combustion Engineering ........................... 4-6

4.3.2 Babcock & Wilcox ............................... 4-6

4.4 Specific Component Failures ............................. 4-9

4.4.1 Control Rod Drive Mechanisms ...................... 4-9

4.4.2 Control System Failures ........................... 4-12

4.4.3 Rod Position Indication Failures ...................... 4-14

4.4.4 Human Error .................................. 4.16

4.4.5 Potential for Loose Parts ........................... 4-19

4.4.6 Miscellaneous .................................. 4-20

5. SUMMARY AND CONCLUSION ............. .. ............... 5-1

5.1 Conclusions ......................................... 5-1

5.1.1 Combustion Engineering CEDS ...................... 5-1

5.1.2 Babcock & Wilcox CRDS .......................... 5-2

5.2 Future Work .............. ......................... 5-2

6. REFERENCES ........................................... 6-1

APPENDIX A.SUMMARY OF LICENSEE EVENT REPORTS FOR

COMBUSTION ENGINEERING -

CONTROL ELEMENT DRIVESYSTEM .....................

APPENDIX B. SUMMARY OF LICENSEE EVENT REPORTS FOR

BABCOCK & WILCOX - CONTROL ROD DRIVE SYSTEM ......

A-1

B-1

vi
LIST OFFIGURES

,

No.

1.1

1.2

1.3

1.4

2.1

2.2

2.3

2.4

2.5

3.1

3.2

4.1

4.2

4.3

4.4

4.5

4.6

4.7

4.8

4.9

4.10

4.11

4.12

4.13

4.14

4.15

4.16

4.17

Number of CE plants per age category ...

Number of B&W plants per age category

CE NPAR system boundary ..........

B&W NPAR system boundary ........

CE magnetic jack CEDM .............

Extension shaft assembly ............

CE rack and pinion CEDM ..........

CE reed switch position transmitters ....

CE CEDM power and logic block diagram

B&W roller nut CRDM .............

B&W CRD control system ...........

CE LER frequency comparison ........

B&W LER frequency comparison ......

CED system aging fraction ...........

CRD system aging fraction ...........

Effects of failure on CED system ......

CE significant plant effects ...........

Effect of failure on CRD system .......

B&W significant plant effects .........

CE CEDM failures ................

CRDM failures ..................

CE control system failures ...........

B&W control system failures .........

CE rod position indication failures ......

B&W reed switch comparison .........

B&W rod position indication failures ....

CE CED related human errors ........

Babcock & Wilcox fuel assembly .......

Paue

1-2

1-3

. . .

. . .

. . .

. . .

. . .

. . .

. . .

. . .

. . .

. . .

. . .

. . .

. . .

. . .

. . .

. . .

...

. .

. . .

. . .

. . .

. . .

. . .

1-5

1-6

2-3

2-7

2-8

2-10

2-11

3-4

3-8

4-3

4-3

4-5

4-5

4-7

4-7

4-8

4-9

4-10

4-10

4-13

4-15

4-15

4-17

4-18

4-18

4-21

' vii
LIST OF TABLES

No. Page

1.1 CE Plants in NPAR Study .................................... 1-1

1.2 B&W Plants in NPAR Study .................................. 1-2

1.3 B&W and CE CRD System History .............................. 1-7

2.1 CE Control Element Design Data ............................... 2-2

2.2 CE Magnetic Jack CEDM Typical Materials ........................ 2-4

3.1 B&W Control Rod Assembly Data .............................. 3-2

3.2 B&W Axial Power Shaping Rod Assembly Data ...................... 3-3

viii
ACKNOWLEDGEMENT

The authors wish to thank our NRC technical monitor Satish Aggarwal for his support, input

and guidance.

Acknowledgement is also given to our colleagues in the Engineering Technology Division,

especially Robert Hall, John Boccio, John Taylor, Mano Subudhi and Ken Sullivan. Their technical

input to this continuing effort has been most helpful. The efforts of Helen Todosow and the staff of

the Nuclear Safety Library are also appreciated.

We also wish to acknowledge Donna Storan, Susan Monteleone and Ann Fort for excellent

manuscript preparation.

ix
1. INTRODUCION

Control rods, and the associated drive and control systems, are essential components of nuclear

reactors which insure safe and reliable operation. This engineering study describes these components

and operating experience for both Combustion Engineering (CE) and Babcock & Wilcox (B&W)

reactors. Emphasis is placed on the specific components which may be susceptible to aging related

degradation.

This engineering study will describe the progress made towards the completion of a Phase I

aging assessment for the B&W and CE Control Rod Drive (CRD) Systems.

1.1 Background

Fifteen plants use Combustion Engineering designed control element drive systems (Table 1.1).

The number of plants per age category is shown in Figure 1.1. These plants vary from the newest,

Palo Verde 3 with 2 years of operation, to Palisades with 18 years. All plants utilize a magnetic jack

control element drive mechanism except Palisades and Fort Calhoun, which use a rack and pinion

drive. Likewise, all CE plants use control element rods to regulate reactivity with the exception of

Palisades, which uses a cruciform type of control element. The actual number of control element

drive's used is a function of reactor size, varying from 37 to 91. The logic, control, and rod position

systems are basically the same for all CE reactors.

Table 1.1. Combustion Engineering Plants in NPAR Study

1. Less than 5 years of operation (2)

Palo Verde 2

Palo Verde 3

II. 5 to 10 years of operation (5)

Palo Verde 1

San Onofre 2

San Onofre 3

St. Lucie 2

Waterford 3

III. 10 to 15 years of operation (3)

ANO-2

Calvert Cliffs 2

St. Lucie 1

IV. 15 to 20 years of operation (5)

Calvert Cliffs 1

Fort Calhoun

Maine Yankee

Millstone 2

Palisades

1-1
Number of Plants

. 7 7: /

5-

4-

3-

2-

1-

0-

.. 

__j A

7,

/ 9

<5 5-10 10-15

Plant Age(Years)

16-20

Figure 1.1. Number of CE plants per age category.

Table 1.2 lists the eight plants which use a Babcock & Wilcox designed control rod drive system.

Even though there are fewer plants compared to CE, these plants are older, on the average, (Figure

1.2) varying from Davis Besse with 13 years of operation to Oconee 1 with 17 years. All of the B&W

plants utilize control rods driven by a roller nut/leadscrew drive mechanism. The logic, rod position

information, and control systems for all B&W plants are similar.

Table 1.2. Babcock & Wilcox Plants in NPAR Study

I. Ten to fifteen years of operation (2)

Crystal River

Davis Besse

11. Fifteen to twenty years of operation (6)

Arkansas - 1

Oconee - 1

Oconee - 2

Oconee - 3

Rancho Seco

Three Mile Island - I

1-2
NUMBER OF PLANTS

7

5

4 -

3

2

0

<5 5-10 10-15 15-20

PLANT AGE (YEAPS)

Figure 1.2. Number of B&W plants per' age category.

1.2 ObiectiVes

The main objectives of this engineering study were to:

1) identify aging and service wear effects which could degrade'the system and in 'turn affect

overall plant safety, and

2) evaluate operating experience' including maintenance practices to identify aging of

components.

To complete these objectives, the following tasks were completed for both the B&W and CE systems:

a) The system designs, including drive mechanisms, rod design;'and all applicable reactor

interfaces (the guide structures in the upper plenum and fuel assembly guide tubes) were

reviewed in detail.

b) The operating and maintenance experience for both designs by assessing the operating

information from commercial databases.

*c) Preliminary conclusions regarding the effect of aging on' the CRD system and related

component failures were 'provided.

1-3
1.3 System Boundary

The system boundaries applicable to this study are shown in Figures 1.3 and 1.4. The items

included are:

* control rod drive mechanisms,

* rod control systems,

* power and logic systems,

* rod position indication systems,

* upper internal guide structures,

* fuel assembly guide tubes,

* individual control rods, and

* CRD cooling systems.

The reactor protection system (RPS), including the reactor trip breakers, are important to overall plant

safety. Signals from the RPS' cause the breakers to open, thus removing power from the drive

mechanisms allowing the control rods to insert freely into the core under the influence of gravity.

This vital system was addressed by a separate NPAR study.

1.4 Analysis Methodolory and Report Format

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, as well as the potential use of the results are described in NUREG-1144.(')

The components and systems to be evaluated by the Program are also defined.

This study provides an engineering assessment of the system and operating experience for both

CE and B&W designed CRD systems. Together with an ongoing assessment of individual utility

maintenance and operating practices, this report will be issued as a formal Phase 1 aging assessment,

in accordance with the guidelines of the NPAR Program Plan and Brookhaven National Laboratory's

(BNL) Aging and Life Extension Assessment Program (ALEAP) Systems Level Plan.(2) As described

in these reports, a formal Phase 1 Control Rod Drive system study would include:

a) detailed evaluation of operating experience data,

b) analysis of industrial maintenance and operating information,

c) identification of failure modes, effects, and causes,

d) review of design, operating environment, and performance requirements.

A similar Phase I aging assessment for the Westinghouse Control Rod Drive System was recently

completed by BNL.( )

To meet the objectives as defined in Section 1.2, it was necessary to understand the system's

operating characteristics, materials and design function. Information was obtained from Final Safety

Analysis Reports, technical reports and system descriptions. This data is presented in Section 2.0 for

Combustion Engineering, and in Section 3.0 for Babcock & Wilcox.

1-4
al ' I ,

CONTROL ELEMENT |

DRIVE MECHANISMS

CE POSITION1- 

INDICATION IN-CORE INSTRUMENTATON CONTROL ELEMENT

I DICATIO SYPPOEM .. PLATE DRIVE MECHANISM

SYSTEM . vt j d 4 h ; CEDM NOZZLE NOZZLE .CONTROL SYSTEM

IN-CORE INSTRUMENT INSTRUMENTATION

GUIDIETUBE ~~~~~OZZLE

REACTOR

>, CULAIIWG W J 8;11l~l~llj lf | PROTECTION

ELEiENT SYSTMSYSTEM

ASSEMBLY L| SM PER ySUPUP 

WITHDRAWN STRUCTUREIRECOPLN

I ~~~~~~~~~~REGULATING PROTECTION

SYSTEM SYSTEM

i.SYSTEMS

BOUNDARY

___________ 

FUEL FOR NPAR ~~ASSEMBLY STUDY

Figurc 1.3. CE NPAR system boundary.
CRD POSITION C

INDICATION CONTROL ROD DRIVE REACTOR

SYSTEM CONTROL ROD DRIVE MECHANISM CONTROL '-PR O TECTN

SYSTEM SYSTEM

CONTROL RO.

ASSEMBLY CRD COOLINGWATER SYSTEM

INTEGRATED

11 51 ' , lt. ~~~~~~CONTROL SYSTEM PLENUM ASSEMBLY CT 

CONTROL ROD

\: X r-m 0 + C~~~~~~~UIDE TUBE

MANUAL

CONTROL

CORE ARREL FUEL ASSEMBLY

REACTOR VESSEL

Figure 1.4. B&W NP'AR system boundary.
Three commercial databases were reviewed to obtain information on significant operating and

maintenance events from 1980 to 1989 in order to identify component failures due to aging. These

databases were:

* Nuclear Plant Reliability Data System (NPRDS),

* Nuclear Power Experience (NPE), and

* Sequence Coding and Search System (SCSS).

Detailed information on these three databases is provided in Section 4.0, together with an assessment

of the aging effect for each design, including detailed discussions on the aging effects of the major

components or systems. Specific LER's are presented in the Appendix, sorted into major system or

component affected, LER No., age at failure, and a brief description of cause and effect.

The B&W and CE control rod drives have experienced failures which have been the subject of

several NRC Information Notices. These are shown on Table 1.3, and are discussed in detail in

Section 4.0.

Table 1.3. B&W and CE Control Rod "rive System History

Year Reference Component Agine Concern

1985 IN 85-38 CRDM Loose parts prevent rod movement.

1986 IN 86-108 CEDM Housing Primary coolant leakage causing

and Flanges corrosion and blockage of CEDM

cooling ducts.

1988 IN 88-47 Rod Control Slower than expected rod drop

System times.

1.5 Future Work

As described in Section 1.2, a review of utilities maintenance and inspection practices is ongoing. A detailed survey requesting specific information on types of inspections, their frequency, and

maintenance for major specific components for the control rod drive system has been sent to the

operating utilities via the Equipment Qualification Advisory Group.

The results of the survey will be reviewed, together with the conclusions of this study, to make

specific findings on the assessment of system aging, and methods to mitigate its effect on plant safety.

A detailed Failure Mode and Effects Analysis (FMEA) or an equivalent study will also be performed

to summarize which component failures lead to specific system or plant effects, and to qualitatively

rank the potential safety significance of the event.

1-7
2. DESCRIPTION OF THE COMBUSTION ENGINEERING CONTROL ELEMENT DRIVE

SYSTEM

2.1 Introduction

Reactivity control in a Combustion Engineering (CE) reactor is provided by two independent

systems, the Control Element Drive System (CEDS) and the Chemical and Volume Control System

(CVCS). The CEDS positions the individual control element assemblies (CEA) within the core to

control short term reactivity changes. This system is also capable of producing a''rapid reactor

shutdown through the rapid insertion of the control elements. This section briefly describes the major

components of the CE system which are being evaluated in this study. Additional system details will

be provided in the Phase I NUREG/CR report.

The major components of the control element drive system which are included in this study

are:

* Control Element Assemblies (CEA),

* Control Element Drive Mechanism (CEDM),

* Control Element Drive Mechanism Control System (CEDMCS),

* CEA Position Indication Systems,

* CEDM Cooling System.

2.2 Control Element Assemblies (CEA)

CE reactors utilize a combination of full and part length control elements to control reactivity.

The typical CEA consists of five poison rods attached to a spider structure. The rods are assembled

in a square array with one center rod. The spider assembly allows for coupling to a individual CEDM.

The design data for the typical full and part length control elements is summarized in Table 2.1.

Palisades, in lieu of the rod type control elements, uses cruciform type control blades which

are fabricated from rectangular stainless steel tubes containing Ag-In-Cd poison. Since these cruciform

control blades are located in the channels between fuel assemblies, solid Zircaloy-4 guide bars located

on the perimeter of the fuel assmiblies, are used, instead of guide tubes.

2.3 Control Element Drive Mechanism (CEDM)(5)'

The Control Element Drive Mechanisms (CEDM) vertically position and provide position

indication for the CEA's within the reactor. The typical CEDM is a magnetic jack type design,

however, two plants utilize a rack and pinion type. Each CEDM is capable of withdrawing, inserting,

holding or tripping a CEA from any position in the core.

A typical CEDM is shown in Figure 2.1. Table 2.2 lists the individual component materials.

- 2-1
Table 2.1. Control Element Design Data(4)

Control Element

Number (Typ)

No. of Elements Per Assy.

Clad Material

Clad Thickness (in.)

Clad O.D. (in.)

Center Element

*Poison Material

*Length (in.) (Typ.)

Corner Elements

*Poison Material

* Length (in.)

Fuel Assembly Guide Tube

* Material

*ID (in.)

Full Leneth

73

5

Inconel 625

.035

.816

Part Length

8

5

Inconel 625

.035

.816

B4C Inconel 625/

Water/B4 C

146 75/58/16

B4C/Ag-In-Cd

135.5/12.5

Zircaloy4

.90

Inconel 625/

Water/B4C

75/58/16

Zircaloy-4

.90

2-2
REED SWITCH ASSY ~T j BALL SEAL HOUSING

SHROUD CONDUIT ASSY

EXTENSION SHAFT AS SY S _-MAGNET ASSY

SEISMIC SNUBBER-..

REED SWITCH ASSY

MOTOR ASSY ---- _

COIL STACK ASSY '

DRIVE SHAFT

COOLING SHROUD .

UPPER PRESSURE

HOUSING

.- i l _ ~ OMEGA SEAL

MOTOR HOUSING

E - - , WIRING TROUGHI

I ~~~~UPPER GRIPPER ASSY

II

E

M :7i

I

I fhIAI.fl f Iflfl~fl AICJ

LuVtfCn nlrrcn A L -

OMEGA SEAL

REACTOR VESSEL

NOZZLE

'1 

Figure 2.1. Combustion Engineering magnetic jack control element

drive mechanism.

2-3
Table 2.2. Typical Materials

Magnetic Jack Type CEDM(6)

Comnonent

1) Motor Housing Assembly

* Pressure Housing

* End Fittings

2) Upper Pressure Housing

*Vent Valve Seal

3) Motor Latches, Links, Pins

4) Motor, Extension Shaft Springs

5) Motor Magnet

6) Motor Fasteners

7) Extension Shaft

8) Extension Shaft Magnet

9) Motor and Extension Shaft

Wear Surfaces

10) Magnet Coils

11) Coil Housings

Material

Type 403 Stainless Steel

Nickel-Chromium-Iron Alloy

Type 316 Stainless Steel

Type 440 Stainless Steel

Ball/Type 316 Stainless Steel Seat

High Cobalt Alloy

Inconel X-750

Type 410 Stainless Steel

Type 304 Stainless Steel

Type i04 Stainless Steel

Alnico No.5

Chromium Plated

Copper wire insulated with high temp.

enamel, vacuum impregnated with high temp.

varnish. Fiberglass taped and encapsulated

with silicone.

Nickel Plated Carbon Steel

2-4
The main components of the CEDM are:

a) Upper and lower pressure housings,

b) Motor assembly,

c) Coil stack assembly,

d) Reed switch assembly,

e) Extension shaft assembly.

Normally, each CEA has a dedicated CEDM. However, depending upon plant size and design, the

number of control elements per assembly can vary from 4 to 12. The design life for the CEDM is 40

years; it is designed to operate without maintenance' for one and one-half years and without component

replacement for a minimum of three years.

CEDM Pressure Housings.

The CEDM pressure housings consist of the motor and upper pressure housing assemblies.

The motor assembly pressure housing is attached to the reactor vessel head by a threaded joint

which is seal welded in place. It need not be removed, since any servicing is performed from

the top of the housing.

The upper pressure housing is threaded and welded onto the motor assembly housing. It

encloses the CEDM extension shaft and vent, ad is sealed at the top by a threaded cap and

welded Omega seal.

Motor Assembly.

The motor assembly is an integral unit which fits'into the pressure housing. The major

components are the latch guide tube and the driving and holding latches.

The major CEA stepping is performed by the driving latches, while the holding latches hold

the CEA during repositioning of the drive latch. The holding latches also minimize wear on

the latch and extension he.:fts by performing a load transfer function during CEA movement.

Engagement of the latches with the extension shaft occurs when the appropriate set of magnetic

coils are energized. This moves the sliding magnets, which cam a two bar linkage, causing the

latches to move inward. The driving latches move vertically 3/4 inch, while the holding latches

move vertically 1/16 inch to perform the load transfer.

* Coil Stack Assembly.

The coil stack assembly consists of five large DC magnet coils mounted on the outside of the

motor housing assembly. The coils supply the magnetic force which activates the mechanical

latches and engages the extension shaft. Two separate power supplies service the coils. A

conduit assembly containing the leads for the coil stack assembly is located on the side of the

upper pressure housing assembly.

2-5
* Reed Switch Assembly.

The reed switch assembly transmits indications of individual CEA positions. The reed switches

are actuated by a permanent magnet attached to the top of the extension shaft. El, rical

connectors for the reed switches are provided at the top of the upper pressure housing. Three

additional pairs of reed switches on each CEDM provide indications of the upper and lower

electrical limits as well as a dropped rod.

Extension Shaft Assembly.

The extension shaft assembly, (Figure 2.2), couples the CEDM to the CEA. At the top of the

extension shaft is a housing containing the permanent magnet which activates the reed switches

as well as a coupling for the lifting tool. The lifting tool is used for engaging or disengaging

the CEA from the extension shaft. The center portion of the assembly is the drive shaft which

is a long stainless steel tube with circumferential notches to engage the CEDM. The drive

shaft is pinned to the extension shaft.

2.3.1 CEDM Design Variations

A more recent CEDM design is referred to as the System 80 CEDM. It is similar to the

standard magnetic jack device, with the following exceptions:

1) the elimination of the pulldown coil, and

2) the use of the lift coils to perform both the load transfer and stepping functions.

Two plants use rack and pinion type of CEDMs as opposed to the typical magnetic jack type

(Figure 2.3). This CEDM, has a drive shaft running parallel to the rack which drives the pinion gear

through a set of bevel gears. The trippable CEA is driven by an electric motor operating through a

gear reducer and magnetic clutch, which when de-energized, allows the CEA to drop into the reactor.

The magnetic clutch incorporates an anti-reversing device which prevents the CEA from moving

upward when the clutch is de-energized.

The main components of this CEDM are:

* CEDM Pressure Housing

* Rack and Pinion Assembly

* Motor Drive Package

* Position Readout Equipment

* Extension Shaft Assembly.

2.4 CEA Position Indication System

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

the reed switch position indication systems.

The pulse counting CEA position indication system infers each CEA position by maintaining

a record of the control pulses sent to each CEDM to either raise or lower the CEA. This system is

incorporated into the plant computer which feeds the control board digital displays. One display

provides CEA group information, while the other provides information on individual CEA position.

2-6
;RIPPER X wSSEMBLY

Figure 2.2. Extension shaft assembly.

2-7
A

B

Detail "A" Detail "B"

I

Figure 2.3. Combustion Engineering rack and pinion CEDM.(7

"

2-8
The reed switch CEA position indication system uses a series of magnetically actuated reed

switches to provide electrical signals representing position. Two independent reed switch position

transmitters (RSPT) are provided for each CEA as shown on Figure 2.4. The RSPT provides an

analog position indication signal and three physically separate, discrete, reed switch position signals.

2.5 Control Element Drive Mechanism Control System

The Control Element Drive Mechanism Control System (CEDMCS) receives automatic CEA

motion demand signals from the Reactor Regulating System (RRS), manual motion signals from the

CEDMCS control panel, and Plant Monitoring System (PMS) sequencing signals, which transmit DC

pulses to the CEDM coils causing CEA motion. A reactor trip caused by the. Reactor Protection

System (RPS) causes power to be removed from the CEDMCS by the trip switchgear, which causes

all CEAs to be inserted by gravity.

The CEDMCS has been designed such that no credible single electrical component failure will

result in the drop of more than one CEA/PLCEA subgroup. This design is based upon a subgroup

in which all members are symmetrically located with their CEDM coil power switches physically sharing

the same module, with the following logics dedicated on a subgroup basis:

1) CEDM coil sequencing and timing logic,

2) Individual CEA motion permissive logic,

3) Coil voltage control circuiting,

4) CEDM coil power switch SCR'actuation and control logic.

Figure 2.5 is a block diagram showing the CEAIPLCEA power switch and control logic

relationships. The major design features of this system are:

* redundant auctioneered logic power supplies,

* separate logic for each subgroup, 

* an auxiliary logic power supply to supply power to hold logic upon loss of main logic power

supplies,

* subgroup power switches housed in two separate assemblies, and

* main power bus supply power to 10 subgroups, 8 which contain regulating or shutdown CEAs.

2.6 CEDM Coolins! System

Though not considered a safety related system, the CEDM cooling system is necessary to

ensure the continuous and reliable operation of the CEDM's. Typically, the system is a forced air

cooling system consisting of two to four fans, designed to maintain' the CEDM coils 'at a temperature

below 3501F. The cooling system is controlled and monitored remotely from the control room.

2-9
CABLE CONNECTOR CABLE CONNECTOR'

(TYPICAL) REFUELING DISCONNECT

PANEL

r

A

-~-. -i . STAINLESS STEEL

FLEXIBLE CONDUIT

A

^ -CABLE CONNECTOR

CONTROL ELEMENT

DRIVE MECHANISM

VIEW AA

CEDM POWER.

CONNECTION

REED SWITCH

POSITION TRANSMITTER 11

REED SWITCH

POSITION TRANSMITTER 11

Figure 2.4. Combustion Engineering reed switch position transmitters. 7 )

2-10
CEDMCS CABINET

w .1-

L LOGIC SECTION I I POWER SECTION J

I

C~~~~~~~~~~~~~~~~ Tott ... I I I I I 1 1 I 1 1

REGULATING AND

SHUTDOWN

SUBGROUPS

-< POWER FROM REACTOR

TRIP SWITCHGEAR

_ BUS BAR ENCLOSURE

POWER SWITCH

ENCLOSURE

t-j

Pi. I.-

SUBGROUP

POWER

SWITCH

ASSEMB LY

PLCEA

FLSUBGROUP

POWER

SWITCH

ASSEMBLY

PLCEA

_ . . . . .... . ._

* SUPPLIES POWER TO "HOLD" LOGIC.

UPON LOSS OF NORMAL LOGIC

POWER SUPPLY

111 II"

COIL POWER COIL POWER

PLCEA Pl.CEA

Figure 2.5. CE CEDM power and logic block diagram.(8 )
-------

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

3.1 Introduction

In a Babcock & Wilcox (B&W) reactor, reactivity is controlled by the Control Rod Drive

System (CRDS) and the Makeup and Purification System. The CRDS positions the moveable control

rod assemblies (CRA) in the core to control the relatively fast reactivity effects. This system also

allows the control rod assemblies to be inserted rapidly in response to signals from the RPS to produce

a rapid reactor shutdown. This section briefly describes the major components of the B&W system

which are being evaluated in this study. The major system components are:

* Control Rod Assemblies (CRA),

* Control Rod Drive Mechanism (CRDM),

* Control Rod Drive Control System (CRDCS),

* CRA Rod Position Indication, and

* Control Rod Drive Cooling Water System.

Additional system details will be provided in the Phase I NUREG/CR report.

3.2 Control Rod Assemblies (CRA)

B&W reactors utiliie a combination of control rod assemblies (CRA's) and axial power shaping

rod assemblies (APSRA's) to control reactivity. The typical CRA consists of either sixteen or twentyfour poison rods attached to a spider structure. The spider assembly allows for coupling to a CRDM.

Design data for the typical control rod is summarized in Table 3.1. The extended life design uses

Inconel clad as opposed to the thinner wall stainless steel tubing. It is also pre-pressurized with helium

to reduce clad stresses. Table 3.2 summarizes the typical design data for the axial power shaping rods

(APSR). The recently designed "Gray' APSR utilizes a longer Inconel absorber section and is

pressurized with Helium to reduce differential pressure stresses in the clad.

3.3 Control Rod Drive Mechanisms (CRDM)

The Control Rod Drive Mechanism both vertically position and provides position indication

for the CRA's within the reactor. The CRDM's are sealed, reluctance motor driven screw units, as

shown on Figure 3.1.

The main components of the CRDM are:

a) Motor tube,

b) Motor,

c) Vent Valve,

d) Leadscrew,

e) Rotor Assembly

f) Torque Tube and Torque Taker,

g) Snubber Assembly,

h) Leadscrew Guide,

i) Position Indicators.

3-l
Table 3.1. Control Rod Assembly Data(9 )

ITEM

Number of CRA

Number of Control Rods per Assembly

Outside Diameter of Control Rod, in.

Cladding Thickness, in.

Cladding Material

End Plug Material

Spider Material

Poison Material

Femal Coupling Material

Length of Poison Section, in.

Stroke of Control Rod, in.

DATA

53-61 (Dependant upon core design)

16

0.440

0.021

Type 304 SS, Cold-Worked

Type 304 SS, Annealed

SS, Grade CF3M

80% Ag, 15% In, 5% Cd

Type 304 SS, Annealed

134

139

Control Rod Assembly Data - Extended Life Design

ITEM DATA

Number of CRA

Number of Control Rods per Assembly

Outside Diameter of Control Rod, in.

Cladding Thickness, in.

Cladding Material

End Plug Material

Spider Material

Poison Material

Female Coupling Material

Length of Poison Section, in.

Stroke of Control Rod, in.

53-61 (Dependant upon core design)

16

0.441

0.0225

Inconel

Inconel

SS, Grade CF3M

80% Ag, 15% In., 5% Cd

Type 304 SS, Annealed

139

139

3-2
Table 3.2. Axial Power Shaping Rod Assembly Data - Black(1')

ITEM DATA

Number of Axial Power Shaping Rod Assemblies 8

Number of Axial Power Shaping Rods per Assembly 16

Outside Diameter of Axial Power Shaping Rod, in. 0.440

Cladding Thickness, in. 0.021

Cladding Material Type 304 SS, Cold-Worked

Plug Material Type 304 SS, Annealed

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

Spider Material -SS, Grade CF3M

Female Coupling Material Type 304 SS, Annealed

Length of Poison Section, in. 36

Stoke of Rod, in. 139

Axial Power Shaping Rod Assembly Data - Gray

Number of Axial Power Shaping Rod Assembly 8

Number of Axial Power Shaping Rods Per Assembly 16

Outside Diameter of Axial Power Shaping Rod, in. 0.440

Cladding Thickness, in. 0.027

Cladding Material Type 304 SS, Cold-Worked

Plug Material Type 304 SS, Annealed

Spider Material SS, Grade CF3M

Poison Material Inconel

Female Coupling Material Type 304 SS, Annealed

Length of Poison Section, in. 63

Stroke of Control Rod, in. 139

3-3
,-- 4

/

/.-- s--S--w

- - - aid ~~~~~ ~ ~~- -

!~izz 3 ___igu B_ r CRDM_ __n 

_ ~ ~ ~ ~ ~ ~ ~ ~ _,.. ~ ~ ~ ~ ~ L.. .% 

Figurc 3.1. 13&W roller nut CRDM.(12 )
I . 4i 

The design life of the B&W CRDM is 126,000 feet of leadscrew axial travel and 500 trip cycles.

These values were determined from data taken from operating plants, assuming that the CRDM would

be used with regulating rods for 20 years.(")

Motor Tube

The motor tube is a 3-piece welded assembly which serves'as the main pressure boundary for

the CRDM. The tube wall between the rotor assembly and the stator is constructed of

magnetic material, which presents a small air gap to the motor, increasing the magnetic

coupling to- the rotor assembly. This area of the motor tube is fabricated from low alloy steel

clad on the I.D. with either Inconel or stainless steel.

The upper end of the motor tube serves as the pressure housing for the withdrawn leadscrew.

It is fabricated from stainless steel which is transition welded to the upper end of the low alloy

steel motor section. The lower end of the motor section is welded to a stainless steel forging,

which is flanged to mate with the reactor head control rod nozzle. Double gaskets, separated

by a ported test annulus, serve as the seal between the CRDM and 'the reactor vessel.

* Motor

The motor is a synchronous reluctance unit with a slip-on stator. The stator is a 48 slot, 4

pole arrangement, containing cooling water' coils in the' outside casing. 'The stator is

encapsulated after winding, and is mounted over the motor tube housing. It is 6 phase star

connected for operation in a pulse stepping mode, which advances 15 degrees per step.

Previous stator designs used epoxy rather than varnish impregnated encapsulation. Varnish

impregnation improves heat transfer and electrical insulation properties. The magnetic

properties of the center section have been improved requiring considerable less power to

operate. Cooling was improved by the addition of the machined grooves on the casing.

Thermocouples monitor stator temperature and alarm should the winding temperature exceed

design limits.

* Vent Valve

The upper end of the motor tube is closed by a closure insert assembly consisting of a double

sealed vapor bleed port and vent valve. The insert closure is retained by a closure nut which

is threaded to the inside of the motor tube.' The sealing load for the closure insert is applied

by a hydraulically pre-loaded spring washer that is retained by the closure nut.

Removal of the closure and vent assembly permits access to either couple or release the

leadscrew assembly from the CRA. The vapor bleed port can bleed all noncondensible gases

accumulated in the top of the reactor prior to head removal as well as during reactor coolant

fill and heatup.

3-5
Leadscrew

The leadscrew is connected to the CRA by a bayonet coupling. The thread is a modified

ACME with a pitch of .375 inch and a relief angle which facilitates disengagement of t oiler

nut from the leadscrew.

The coupling/uncoupling is accomplished through the use of a special handling tool. To couple

a CRA, the leadscrew is lowered until the male bayonet coupling is positioned inside the

female coupling of a CRA. At the same time, a pin on the upper portion of the leadscrew

is positioned within the torque taker. The leadscrew is rotated 450 by the handling tool, which

positions the tabs on the bayonet coupling so that it grapples the CRA. The leadscrew nut

on the top of the leadscrew assembly is then tightened in place, preventing any movement

relative to the torque taker and CRA. The reverse procedure is employed to uncouple the

CRA.

* Rotor Assembly

The rotor assembly consists of a ball bearing supported rotor tube which carries, and limits the

travel, of a pair of segment arms. Each of the arms carry a pair of ball bearing supported

roiler nut assemblies which are skewed at the appropriate helix angle so to engage the

leadscrew. Current in the motor stator causes the arms, which pivot in the rotor tube, to

move radially toward the motor tube wall, causing the lower portion of the arms to move

inward to the leadscrew and engage the roller nuts. Four separating springs mounted in the

segment arms keep the rollers disengaged when power is removed from the stator windings.

A second radial bearing mounted to the upper end of the rotor tube has its outer race pinned

to both segment arms, synchronizing their motion during engagement and disengagement.

When a six phase rotating magnetic field is applied to the motor stator, the resulting force

produces rotor assembly rotation.

* Torque Tube and Torque Taker

The torque tube is a separate tubular assembly containing a key which extends 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 also acts as the down

motion stop. The leadscrew contacts the motor tube closure insert assembly for the out motion

stop.

The torque taker assembly consists of the permanent magnet for position indication, the

snubber piston, and a positioning keyway. It is attached to the top of the leadscrew and has

a keyway which mates with the key in the torque tube to provide both radial and tangential

leadscrew support.

3-6
Snubber Assembly

The snubber assembly includes a piston which is the lower portion of the torque taker

assembly, a hydra iic snubber ylinder, and a belleville spring assembly which is attach~ to the

lower end of the torque tube. The snubber cylinder is closed at the bottom by the leadscrew

and the snubber bushing. -The snubber cylinder has an active length of 12 inches through which

the leadscrew and CRA can be decelerated during a trip without applying a force greater'than

10g on the control rod. The snubber damping characteristics are determined by the size and

position of holes in the snubber cylinder wall and the water leakage at the snubber piston and

bushing. Practical operating clearances limit the amount of water leakage. Therefore, at the

end of the snubbing stroke, there is kinetic energy from a 5 ft/sec' impact velocity which is

absorbed by the belleville spring assembly; absorption is accomplished by a slight, instantaneous

overtravel past the normal down stop.

Leadscrew Guide Assembly

The leadscrew guide assembly has two functions. First, the' bushing serves as a-thermal barrier

and leadscrew guide, and second it allows coolant to flow into the upper motor guide tube area

during a'trip to replace the volume in the tube formerly occupied by the leadscrew.

As a primary thermal barrier, the bushing allows only a small path for free convection of water

between the mechanism and the closure head nozzle, which governs the fluid temperature in

the mechanism. To obtain acceptable trip times, an additional flowpath'must be provided for

coolant around the leadscrew guide bushing.~ The larger area path is necessary to reduce the

pressure differential required to drive water into the mechanism to equal the leadscrew

displacement. The auxiliary flow paths are closed for small pressure differentials by ball check

valves, which prevent convection flow, -but open fully during- a trip. -

* Rod Position Indication

Two methods of position indication, an absolute and relative position indication system, are

provided. The absolute position indication system consists of a series of magnetically operated

reed switches which produce an analog voltage signal proportional to rod position. The relative

position indication system consist of a small pulse stepping motor connected to a potentiometer

which produces a signal corresponding to'demanided rod position.

3.4 Control Rod Drive Control System

The Control Rod Drive Control System' (CRDCS) provides for the withdrawal or insertion of

the CRA's in response to either automatic signals from the Integrated Control System (ICS) or manual

signals from the operator. A simplified schematic of the CRDCS is shown in Figure 3.2. The CRDCS

also allows the rapid insertion of the CRAs to' produce a shutdown upon a'signal from the Reactor

Protection System. The three'main components of the CRDCS are the control rod drive motor power

supplies, the system logic, and the trip breakers.

The CRDCS power supplies consist of group auxiliary and holding power supplies. The group

and auxiliary power supplies sequentially'energize the-stators to produce a rotating magnetic field used

to position the CRA. The holding power supplies are used to maintain the CRA fully withdrawn,

therefore switching is not required.

3-7
.. V4 S .000*0as00 

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, . "",",", oo." n 211 '>' 

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)-w@ _ 96 suft _ - ...I

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A I 004 o _ . !

N t 0 P . + '.- II

- -~ 4- 4 -LI-- 1.1 .-4 -- -i4+t---

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a4a .- .-.--- * a I I * ~~~~I .

I ~ I. I I, L J~~~~ 

: IIM I -&1 ......

ni ' n hU4i ! n1. . . . I .- 1 li

I1 S ' I' I 1a v1 11 111111 i i 1 1 1

n UU '-4-4 _-:4:-; I--1 ---- I- - -

r

-n UU

l I I I _ I I I ' 1> 1.-h I1 iIi 111111x ii 

i ruuL------ . I . .i 111 . _ - -.~ UU

I . . . 11 11 i 111 0 ff i 1 1 w

60 I LWL L- .

I I

4-4

I I

z~zmI I

F-r:JJ LLT

*l~l~lrlT4Tl~ Iij i fl j11, ICA"ll"

To CA tAtO6S e aO COIO 0. S.ou 8 tO COA. OS *oW 

1T"PI'L CA - -A 5.11) 1 ETTPAM. 005 S1 O aa ( VPC&a 0 *4 00 S

LI

o A Co. * 4 * C.. 60 P a *. * Gaa aa 

f0l0 * * 0 t C11 0 00-0 L

. I I W AI 

0 1 C0U

Figure 3.2. B&W CRD control sysleni.( 13)
. ' 

The system logic encompasses the functions which command CRD motion in either the manual

or the automatic modes, including CRD sequencing, safety and protection features, and manual trip.

The major components of the logic system are:

a) control panel,

b) CRA position indication panels,

c) automatic sequencer which utilizes rod position signals to regulate group position, and

d) relay logic which prohibits out of sequence group movement.

Trip breakers are provided to interrupt power to the control rod drive motors, which allows

the roller nuts to disengage from the leadscrew causing the CRA freely insert into the reactor. A trip

can be initiated by the RPS breakers or by a power interruption to the SCR gating power and DC

holding power circuits. Since parallel power feeds are provided to both the holding and gating power,

interruption of both is required for a trip.

The APSRA drive mechanisms are modified to prevent the roller nuts from disengaging and

tripping upon power removal.

3.5 Control Rod Position Indication

Position indication for each CRA is provided by two separate systems, the absolute and the

relative position indication systems.

The absolute position indication system consists of magnetically operated reed switches mounted

in a fiberglass tube parallel to the motor tube extension. Switch contacts close when the permanent

magnet, mounted on the upper end of the leadscrew extension passes by. The reed switches are

connected to a voltage divider network, which translates varying resistance into position.

The relative position indication system consists of a small, pulse stepping motor which is driven

from the power supply for the rod drive motor. The motor drives a potentiometer which produces

a variable output corresponding to rod position. Though extremely accurate, the system only reflects

rod position from field rotation, and does not provide the correct position indication if the rod is

tripped, dropped or immoveable. The system is reset through the use of a reset pulser after a dropped

rod or reactor trip, since rod motion was not caused by an electrical signal.

3.6 Control Rod Drive Cooling Water System

The Control Rod Drive Cooling Water System is a closed system consisting of two redundant

trains. The major system components are two centrifugal pumps, two heat exchangers, and one surge

tank. This system only functions during normal operations.

The separate cooling loops are made entirely of stainless steel, with the exception of the

copper-nickel cooler tubes. The latter reduces 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 by the CRD stator.

3-9
4. OPERATIONAL EXPERIENCE - B&W AND CE

4.1 Introduction

A review of the operating and failure histories for the B&W and CE control rod dtive systems

indicate that both systems have experienced aging resulting in significant plant effects. This data, for

the period from 1980 through 1989, was obtained primarily from three national sources of nuclear plant

operating experience information:

1) Nuclear Plant Reliability Data System (NPRDS),

2) Nuclear Plant Experience (NPE),

3) Sequence Coding and Search System (SCSS).

The NPRDS is a computerized information retrieval system maintained by the Institute of

Nuclear Power Operations (INPO). Information on performance provided by this system is based upon

failure event reports of key components submitted by the nuclear utilities to INPO. NPRDS provides

access to historical data on engineering failure data reflecting a broad range of operating experience.

The Nuclear Plant Experience (NPE) data base is a commercial technical-publication service

which compiles descriptive summaries of significant events and an indexed reference to all such

occurrences. Though much of the information in this data base is obtained from the Licensee Event

Reports (LER's), NPE also has information from utility operating reports and a wide variety of current

literature.

The Sequence Coding and Search System (SCSS), also commonly referred to as the LER data

base, provides summaries on all LER's. Each entry into this data base has information on the failed

component, the cause of failure (if known), and its effect on plant operation.

Each of the three data bases was searched for operational failure event reports related to the

control rod drive systems. Additional queries were made on control rods, guide tubes and reactor

internals. Although the three data bases provide, to some extent, duplicate information, a review of

all three was required for a thorough search of the operational experiences of the system and its

related components. Assurances were made that the events were not double counted,- through crossreferencing with the appropriate LER number. The specific LER's for both designs are presented in

the Appendices.

Loose parts in the reactor are considered a potential hazard to the CRD system. As such, the

Loose Parts category is included since there were several instances where broken fuel assembly

components and reactor internals could have potentially interfered with or prevented control rod

insertion.

4-1
From the data evaluation, the LER's for CE plants were categorized into the following main

categories:

1) Cables and Conn--tors

2) Coil Power Programmer

3) Coil Power Programmer Power Supplies

4) Control Element Drive Mechanism Control System

5) Control Element Drive Mechanism

6) Rod Position Indication

7) Human Error

8) Unknown Causes

9) Miscellaneous

Similarly, LER's for B&W plants were categorized into the following categories:

1) Cable and Connectors,

2) Control Rod Drive Control System,

3) Control Rod Drive Mechanism,

4) Rod Position Indication,

5) Human Error, and

6) Loose Parts.

A comparison between the Sequence Cooling and NPRDS data bases is presented in Figures

4.1 and 4.2 for the CE and B&W designs respectively. For CE plants, the CEDMCS and CRDM

subsystems account for the highest frequence of o urrences. Similarly, the CRDCS and CRDM

account for the highest frequency for B&W plants.

The differences between the two databases can be explained by the types of events categorized

by each. The SCSS provides LER information only. Events occurring during outages are not always

reported through LER's. NPRDS provides data from maintenance records, industry literature and

LERs also. Therefore, problems found during an outage inspection are likely to be reported in

NPRDS. When reporting to NPRDS, categorization is done by each utility, so some variation between

plants can be expected. Starting in 1984, utilities were only required to report significant events to

INPO, therefore prior occurrences may not have been included in the NPRDS. Utilities typically

generate only one LER which can list multiple failures, but usually list each component failure

separately on NPRDS. Each database provides useful information in assessing the aging of the CRD

systems.

4-2
% OCCURENCE --

COIL CPP CEDMOS CEDM RPI '-HUMAN UNKNOWN MlSC.

POWER PWR ERROR

PRGMMER SUPP EVENT

m NPRDS = SCSS

Figure 4.1. CE LER frequency comparison NPRDS-SCSS.

70%

607.

507.

40%

30%

20%

10%

07

Mab NPRDS 3 SCSS

Figure .2. B&W L frequency . comparIn ' ' .' :' s S 

Figure 4.2. B&W LER frequency comparison NPRDS vs SCSS.

.4-3
4.2 Dominant Component Failures

One primary objective of this study was to assess the impact of CRD aging-related degradation

upon system performance and reliability. To accomplish this, a comparison was made between the

information obtained from the data bases and the NPAR definition of aging related failures, as defined

in NUREG-1144. To be classified as aging related, a failure must satisfy the following criteria:

1. The failure must be the result of cumulative changes with the passage of time, which

if unchecked, could result in loss of function and impairment of safety. Factors causing

aging may include:

a) Natural, internal, chemical, and physical processes which occur during operation;

b) External stresses (i.e. radiation, heat, humidity) caused by either the storage or

operating environments.

2. To eliminate failures related to infant mortality, the component must have been in

service for at least six months.

4.2.1. Combustion Engineering

Using the aging criteria, the LERs for Combustion Engineering plants were reviewed to

determine the fraction which were aging related. As shown on Figure 4.3, 27% of the LER's were

determined to be aging related. Non-aging system failures were described in 20% of the LER's. The

remaining 53% lacked sufficient detailed information regarding failure cause. However, given the

similarities in comparing the unknown cause failure descriptions to similar component failures with

known causes, it is anticipated that a minimum of 50% of these may be attributable to aging. Thus

the 27% identified should be considered as a minimum.

Primary coolant leakage from the rack and pinion CEDM, coil and control system failures

accounted for the majority of the aging failures. Human errors, maintenance errors, and faulty

procedures accounted for most of the non-aging failures, however, many of these did result in

significant plant effects, as will be discussed in section 4.3. These effects demonstrate the control

element drive systems sensitivity to these failures.

4.2.2 Babcock & Wilcox

Utilizing the same aging criteria as described above, the Babcock & Wilcox control rod drive

system LER's were evaluated to identify the degree of aging related failure. As shown on Figure 4.4,

45% of the events were categorized as aging related, with an additional 48% being identified as

potentially aging related.

CRDM stator failures and primary coolant leakage were the dominant aging related failures,

while control system failures accounted for many of the potentially aging failures. Prior to redesigning

the rod position indication system in 1985, B&W plants experienced relatively frequent system failures.

This frequency has been decreased through the use of more reliable reed switches and redundant

circuitry, as described in Section 4.4.

No trend was seen to demonstrate system susceptibility to human related errors. However, the

control systems susceptibility to environmental degradation due to heat and dust contamination was

seen.

4-4
POTENTIAL AGING*

53%

NON-AGING

20%

SCSS DATA

* See discussion Section 4.2.1

Figure 4.3. CED system aging fraction.

AGING

45%

NON-AGING

NPRDS DATABASE

POTENTIAL AGING

48%

Figure 4.4. CRD system aging fraction.

4-5
4.3 Failure Effect

4.3.1 Combustion Engineering

Failures of individual components of the control element drive system not only affected the

CED system, but also caused significant plant effects. As shown on Figure 4.5, CED system failure

resulted in degraded operations 47% of the time, and a loss of redundancy another 25%. Perturbations

to plant operations were defined as degraded operations. Loss of redundancy events were comprised

primarily of rod positions indication system failures. Since CE plants utilize redundant systems to

provide position indication, the loss of one does not have an immediate effect upon plant operations,

other than requiring plant operators to verify rod position more frequently. However, from a risk and

reliability standpoint, loss of redundancy may be significant. The no effect category included procedural

and human errors which resulted in no plant operational effect.

As indicated on Figure 4.6, the most common plant effect was reduced load, primarily due to

efforts necessary to recover from a dropped or slipped CEA. As per Standard Plant Technical

Specifications,('4 ) power must be reduced to 70% if the time required to re-align a dropped or slipped

CEA exceeds one hour.

For the ten year time period reviewed, automatic scrams occurred ten times. Eight of these

scrams occurred on a low DNBR signal due to dropped CEA's, faulty CEAC, or control system

overheating. Malfunctioning CEAC memory cards accounted for the remaining two scrams. As a

result of these CEAC malfunctions, slight overcooling of the RCS occurred due to the combined effects

of steam generator blowdown, main steam line drain valves remaining open, the boric acid concentrator

being in service, the main feedwater pump turbine remaining in operation, and the low level of decay

heat generated at that particular point in the fuel cycle.

Multiple rod events, due either to equipment failure or maintenance errors accounted for the

manual plant shutdowns. A mismatch between reactor power and turbine load, due primarily to a

dropped CEA, caused a rapid increase in steam generator pressure with a level shrink, leading the

operators to manually scram the reactor on one occasion.

Engineered Safety Feature (ESF) actuation occurred seven times over the time period, five of

which were automatic. The majority of the automatic actuations were the results of low DNBR signals

due to dropped CEA's, which actuated the emergency feedwater system. One event caused a rapid

RCS cooldown initiated by excessive steam demand due to feedwater valves which incorrectly remained

open. One unintentional ESF actuation was due to a multiple CEA drop due to a maintenance error,

while the other was due to operator failure when the Safety Injection Actuation Signal (SIAS) was not

blocked during a reactor shutdown.

4.3.2 Babcock & Wilcox

Failures of the Control Rod Drive System for B&W plants, as shown in Figure 4.7, resulted

in degraded plant operations 50% of the time, and loss of redundancy an additional 27%. As described

above for CE, malfunctions and failures accounted for the majority of the redundancy losses.

4-6
Jv,

LOSS OF REDUNDANCY

25% A

NO EFFECT

&. 28%

DEGRADED OPERATION .., 47%

SCSS DATA

Figure 4.5. Effects of failure on CED system SCSS data.

0ar:Ietn I A ncuv -V I aU4 

i__~~~~~~~~..

_ t~~::::: 

UNINTEAXIT8N 4LF ACT 2

UN ro SCRAM 10 INTEN TION,

AUTO S-CRJ

MANUIAL SH

MANUAL SCRAM 1 AUTO ESF ACT 6

Figure 4.6. CE significant plant effects 1980-1989.

4-7
S /1 i W S /l t R/I / l

w / s \ 5 / s \ S

\ /

I

\ / mn

; OF REDUNDANCY

27%

DEGRADED OPERATION

50%

EFFECT

23%

SCSS DATABASE

Figure 4.7. Effect of failure on CRD system SCSS database.

CRD system failures also caused significant perturbations on plant operations as shown on

Figure 4.8. Control Rod assembly slippage or complete dropping were the primary cause of reduced

load incidents. A loose power connection twice caused rod drops resulting in a reactor tilt exceeding

the Technical Specification limits. A failed fuse in a control rod power supply accounted for another

rod drop occurrence.

Three automatic reactor scrams were reported during the period. In one instance, a failed

control system logic card prevented reactor operators from inserting the axial power shaping rods in

response to a core unbalance, which subsequently led to a trip. An Integrated Control System (ICS)

failure caused by concrete dust contamination of the control cabinet from nearby construction, caused

another reactor trip. The third reactor trip was actually caused by low flow to the main feedwater

pumps, however, the event was initiated by a rod drop.

The manual actuation of the ESF was initiated by a dropped rod and subsequent reactor trip

due to low reactor system pressure. The feedwater block valves failed to close causing a high steam

generator level. The emergency feedwater system was manually activated, and the main feedwater

pumps manually tripped.

During the time period, two plant manual scrams were also reported. Multiple dropped rods

due to a failed power supply initiated one event. The other manual scram was caused by a loose

solder joint on a control rod sequencing card resulting in a power loss to the control rods.

4-8
REDUCED LOAD 6

AUTO SCRAM 3

MANUAL ESF ACT 1 N iiFi THER 3

MANUAL SCRAM 2

Figure'4.8. B&W significant plant effects' 1980-1989.

The three plant events classified as other occurred while the plant was either at zero power

or refueling. An immovable control rod due to a fractured CRDM leaf spring halted a plant start up,

while broken fuel assembly holddown springs found during a refueling outage inspection accounted for

the remaining two events. Both of these occurrences highlight potential loose parts which may have

interfered with the movement of the control rods. The third zero power occurrence was initiated by

a failed fuse in the absolute position indication system, which delayed start up.

It is' important to note that many other control rod system'failures, for both designs, also

affected plant operations. For example, loss of a rod position indication system required plant

operators to verify rod position more frequently. Dropped rods required operators to verify adequate

shutdown margin, and to undertake measures to recover the misplaced rod. However, since the time

required to complete this in some instances was less than one hour, the event was not classified as

significant.

4.4 Specific Component Failures

4.4.1 Control Rod Drive Mechanisms

A review of the reportable'failures indicate's that CEDM failures occurred frequently. For the

period, nineteen LER's were written documenting CEDM failures at CE plants. The specific failure

causes are shown on Figure 4.9. A similar review of the NPRDS data base is presented in Figure 4.10.

Since many of the failures resulting in primary coolant leakage are found during outages, LER's are

not necessarily written, since it did not affect the safe operation of the plant.

.4-9
FAIL TO VENT 

SEAL HOUSING 3

MISC. 2

INADEQUATE COOLING 2

COIL DEGRADATION 6

SCSS DATABASE

Figure 4.9. CE CEDM failures 1980-1989.

70

60

50

40

30

20

10

0

% FAILURE

-in

, _'

STATORS INOP. CRDM PCS LEAKAGE

_CE BW

NPRDS DATABASE

Figure 4.10. CRDM failures 1980-1989.

4-10
Several instances of dropped rods due to the elimination of CEDM venting before startup were

also reported. This operation was eliminated as'an unnecessary step since there was no apparent

reason for performing it. However, since no other cause for the dr6pped rods was found, venting was

returned to the start-up prcced..es, ad the problem subsided.

a. Primary Coolant Leakage

Primary coolant leakage in a high-temperature area, such as the reactor vessel head, will cause

the boric acid solution to boil and concentrate, increasing its acidity and corrosiveness. Additionally,

the boric acid crystals could -accumulate and block the cooling passages. If this condition'is left

uncorrected, the stator coils could overheat and fail, or the winding insulations could be degraded

causing an electrical short.

Nuclear utilities were first alerted to the potential problems of primary coolant leakage in

Information Notice No.86-108, Degradation of Reactor Coolant System Pressure Boundary Resulting

From Boric Acid Corrosion". This IN detailed leakage from a CRDM seal which was' first observed

during an outage. The leakage was judged acceptable for continued operation and left uncorrected.

However, during a subsequent shutdown and inspection, it was found that boric acid crystals had

severely corroded the CRDM cooling shroud.

Both B&W and CE plants have experienced primary coolant leakage as seen'from Figure 4.10.

The NPRDS data base listed fifty entries from CE plants and thirty seven from' B&W plants. The

majority of leaks from CE plants were from the seals of the rack and pinion type CEDM, while B&W

experienced leaks between the motor tube and nozile flange. The causes of the leaks, as determined

by the utilities, were poor seal quality due primarily to inadequate QC, normal wear and deterioration.

B&W also experienced leakage from the double sealed'CRDM vent valve located on the top

of the motor tube. The sealing load is applied by a' hydraulically' pre-16aded spring washer. Boron

deposits were noted on this assembly during visual inspections, necessitating an increase in the torque.

Another significant CE occurrence identified leaking CEDM pressure housings, which when

removed and inspected, revealed circumferential crack indications on the seal housing. Similar

indications were also found on the remaining two housings from the same fabrication lot. Further

detailed analyses concluded that the faults were likely due to contaminant induced stress corrosion

cracking. Eleven housings exhibited similar faults two years later. The exact contaminant was 'never

conclusively identified, but it was thought to have been introduced during fabrication.

b. Stator Failure

Numerous instances of dropped rods due to stator failure for both B&W and CE plants'were

also reported. Coil failure due to overheating and electrical degradation were the common causes.

Not all stator defects were found during operation as many were found during meggering tests

performed at shutdowns and outages. Dropped rods, which appeared to have been' caused by failed

power supply fuses, were frequently caused by shorted stators.

Babcock & Wilcox identified four'major causes 'for stator failure: epoxy breakdown due to wire

incompatibility, moisture intrusion, bifilar design 'and fabrication defectsls).'

4-11
Overheating due to blocked cooling air passages was also observed. Millstone 2 replaced

twenty three upper gripper coils which' shorted due to overheating caused by boric acid precipitates

blocking air paths. This type of problem did not occur in B&W plants primarily due to the use of

water for stator cooling.

c. Loose Parts

Instances of inoperable CRDM's caused by broken internals were also reported to NPRDS.

B&W utilities were'alerted to the. problem by Information Notice No. 85-38, "Loose Parts Obstruct

Control Rod Drive Mechanism." Davis Besse reported two incidents, four years apart, where internal

CRDM pieces broke and became jammed, thus preventing the CRDM from moving.

In 1981, while recovering from a reactor trip, operators discovered that one rod would not

withdraw. Upon subsequent disassembly and inspection of the CRDM, it was found that the leaf

spring anti-rotational device of the leadscrew nut assembly had fractured into several fragments which

became stuck between the buffer spring and the leadscrew, preventing the leadscrew from rising. Since

no other similar occurrences had been reported, it was concluded that the failure was an isolated event.

However, while performing CRD drop time tests at Davis Besse in 1985, a CRDM failed to

insert in the time required by the Technical Specification.("') Examination revealed that a setscrew

fragment had jammed, thus preventing disengagement of the latching mechanism from the leadscrew.

Further inspection also revealed a broken leaf spring on the top of the leadscrew.

The setscrew which jammed the leadscrew broke off a CRDM maintenance tool. The spring

failure was a brittle intergranular failure caused by mechanical interference with the torque tube cap

in the top of the CRDM housing. Improper seating caused the springs to extend further than

designed, and thus, struck the torque tube cap when the rod was raised. Further inspections revealed

four additional CRDM's with similar problems.

Maintenance procedures were changed in an attempt to preclude a reoccurrence. CRDM tools

were inspected after use to ensure that they had no broken parts, and visual techniques were employed

to ensure properly positioned leaf springs. Prior to this, no requirement existed to ensure proper

seating.

4.4.2 Control System Failures

a. Combustion Engineering

Control system failures accounted for the greatest occurrence of failures in CE plants. The

result of the failures normally were dropped, slipped, or immovable CEA's. The control system was

also susceptible to exterior stresses as well, since several instances of component overheating due to

inadequate cooling system operation were reported. The specific component failures frequency are

shown in Figure 4.11.

Failures of the coil power programmer were common, as sluggish upper gripper operation

accounted for a large number of slipped or dropped CEA's. Corrective action consisted of adjusting

the gripper timing sequences and increasing the coil voltages. Some plants have replaced the existing

magnetic coils, with larger System 80 coils, designed to produce stronger magnetic fields resulting in

improved latching capability.

4-12
I:

CEDMOS 29%

COIL PWR. PROG. 35%

CPP POWER SUPPLY 37%

SCSS DATABASE

Figure 4.11. CE control system failures 1980-1989.

Another related problem was caused by a failed upper gripper coil DC current sensor.

Investigation revealed that the failed sensor sent a false signal indicating that the gripper was energized

when, in fact, it was de-energized. The CEA timer, sensing that 2 CEA coils were energized and

capable of holding the CEA, de-energized one of the coils, allowing the CEA to drop. The root

cause 'of the sensor failure was control cabinet and component damage caused by overheating in the

CEDMCS room.

Failed 1SV power supplies, often replacements for the originally installed model, accounted for

a large portion of reportable occurrences resulting in numerous dropped or slipped CEA's. Due to

the high incidence of failures, the original power supply model was reinstalled, and the design modified

to incorporate redundant power supplies. It can be assumed that some of the LER's with unknown

causes listed in Appendix A-8 also were caused by these same failures. However, since definitive

failure causes were not reported, no correlation was drawn.

One unit reported a manual plant trip resulting from multiple dropped rods caused by the

failure of a CEA subgroup breaker. The root cause for the breaker trip was not known, however

testing indicated that the breaker tripped at a current of approximately 30 amps, 25% less than the

40 amp continuous design load. A NRC maintenance team inspection reviewed this event, and

recommended that all the subgroup molded case circuit breakers be tested during the next outage, and

then placed in the PM program for periodic testing. The basis for this recommendation was that

molded case circuit breakers should be tested periodically to assure their reliability for the 40 year

design life.

A slipped CEA due to an intermittent ground on the lower lift coil caused by degraded

insulation on the coil lead wire was also reported. The ground occurred immediately following the

4-13
voltage increase associated with energizing the lower lift coil, and resulted from coil movement which

brought the damaged lead near the coil stack housing assembly. These grounds were first identified

in 1985. Based upon successful meggering and continuity checks, the coils were judged acceptable for

use. A similar ground problem occurred during the first cycle at Palo Verde Unit 2. Examination

of the coil revealed substanual damage to the coil lead wire due to a fabrication deficiency. Two

factors contributed to the defective lead wire becoming a pathway for a short circuit to ground namely:

1) the motion of the lower lift coil leads, and

2) the orientation of the coil within the housing.

The remaining control system failures were due to overheating, breaker failures and electrical

grounding of the coils.

The relative susceptibility of the control systems to the outside environment can be seen in the

fact that LER's were generated on four separate occasions where the failure cause was overheating

due to inadequate ventilation. The end effect of these events were dropped or immovable CEA's.

b. Babcock & Wilcox

As shown on Figure 4.2, control system failures accounted for approximately 25% of the

reported events. However, with the exception of blown fuses, no other component accounted for any

significant majority as shown on Figure 4.12.

Failed fuses accounted for three LER's and six NPRDS entries. Three separate instances of

dropped rods caused by failure of the transfer switch module fuse were reported by Davis Besse. The

transfer switch is used when a fully positioned rod group is moved to a holding bus. Failure of this

switch may result in dropped rods. Investigation on the cause of fuse failures were inconclusive, and

there was no evidence of either excessive heating or current on either the fuse, fuse holder, or

associated wiring. Another fuse failure in the motor programmer also caused a rod drop. Again, no

reason was found to explain the failure as no evidence of overheating or overcurrent was seen.

The susceptibility of the B&W control system to external factors was also demonstrated. On

two occasions, CRD control system programmer board failures resulted in dropped rods. Investigations

into the failure cause concluded that concrete dust from work performed in the vicinity of the control

cabinets caused the boards to fail. Cabinet circuit board contacts were cleaned, faulty programmer

boards replaced, and the system then performed as designed.

4.4.3 Rod Position Indication Failures

a. Combustion Engineering

CE plants utilize two systems to provide rod position indication, the reed switch and the pulse

counting systems. For the period reviewed, component failures of each accounted for a significant

portion of the reportable events, as shown on Figure 4.13.

4-14
% OCCURENCE

80

60

40

20

0 - I :CNTRL. SYS. FAILURES'

SS M NPRDS

FUSE FAILURES

- S aFigure 4.12. B&W control system failures 1980-1989. u . . I SYS

REED SWITCH

31%

PLNT. COMP. FLR

15% -

POWER SUPPLY

i 8%

DDPS

13%

CEAC

33%

Figure 4.13. CE rod position indication failures 1980-1989.

4-15
Control Element Assembly Calculator (CEAC) failures accounted for the highest frequency

of failures. The CEAC is an integral part of the position indication system, and its loss could generate

spurious rod position signals leading to a reactor trip. The typical CEAC failure was related to circuit

board malfunctions, however, each report did not provide sufficient detail in each instance to determine

exact failure cause. It is interesting to note that one CEAC failure was credited to natural enu uf fife

failure during the first cycle of operation.

Reed switch failures were the other main failure cause. Since each CEA has independent reed

switch position transmitters, the loss of one does not necessarily impair the function of the CEA. It

does however, represent the loss of redundancy.

Plant computer failures affected the operation of the pulse counting system. No dominant

failure cause was evident, however, as causes ranged from software errors to circuit board failures.

b. Babcock & Wilcox

A review of the LER's generated by B&W plants documenting rod position indication system

failures indicate that the majority were prior to 1985. Prior to this time, a single reed switch failure

would render the system inoperable. The system was redesigned to provide for redundant circuitry to

allow f continual operations with failed reed switches.

Also, a new reed switch design, designated R4C was utilized. These new reed switches were

more reliable and allowed the system to continue to operate with one failed reed switch. These

switches were completely encased in glass, with rhodium contact points. Previous designs used a low

differential reed switch with gold plating on top of the rhodium plating as shown in Figure 4.14. The

change was necessitated by the failure rate and erratic operation of the previous design. The previous

design had a tendency to build up a surface film on the contacts, which in conjunction with a small

closing force resulting from the small gap, led to switch failures. The R4C reed switch provides a

more positive contact because of the higher closing force, thus reducing surface film buildup.

As shown on Figure 4.15, no other major component failure caused the system to fail, and the

occurrence of system failures since 1985 has decreased markedly.

4.4.4 Human Error

a. Combustion Engineering

The control element drive system was susceptible to component failures and environmental

degradation as well as human errors. As shown on Figure 4.16, human errors could be divided into

either maintenance errors or procedural errors. While it may be argued that errors of this type are

not aging related, they do stress the plant systems, particularly when scrams, power reductions and ESF

actuations result.

4-16
ps 

-GOLD BUILDUP

GAP R-HRHODIUM PLATING

.I- . DIFFERENTIAL GAP

(b) LOW DIFFERENTIAL SWITCH

(a) HIGH DIFFERENTIALSWITCH . .

Figure 4.14. Babcock Wilcox reed switch comparison.

4-17
REEn SW

(0tD

SCSS. DAI

. I.R L

, ~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~. RELAYS

10%

UPPLIES

"ABASE

POWER S

20

Figure 4.15. B&W rod position indication failures 1980-1989.

MAINTANANCE ERRORS

69% _

PROCEDURAL ERRORS

41% SCSS DATABASE

.

Figure 4.16. CE CED related human errors 1980-1989.

4-18
The majority of maintenance' errors resulted in dropped rods' and power reductions. Other

events were caused by procedural errors which either did not account for proper position verification

of the CEA's, or caused a portion of the system to be out of service. While it did not result in any

particular system effect, in rare instances' a loss of redundancy resulted.'

One occurrence of human error resulted in a violation of applicable Tech. Specs. due to the

cooling coil water shutoff valve limit switch power supply not having the required backup overcurrent

protection. The applicable control wiring diagrams erroneously showed a 120V AC cable connection

on the line side of the fuse which should have provided the required'overcurrent protection. To

correct this problem, the cable was moved from the line side of the fuse to the load side. Again, while

not an aging issue, the installed system did not match the documentation, which could have significantly

impaired the cooling system, which in turn could have exposed the CED system to higher than normal

temperatures.

b. Babcock & Wilcox

The review of reportable events for B&W plants for the 1980-1989 time period indicated no

susceptibility of the control rod drive system to human or procedural errors.

4.4.5 Potential for Loose Parts

a. Combustion Engineering

The CE Control Element Drive System sustained no significant events related to reactor

internal degradation. One event documented a guide tube deformation discovered during an outage

inspection. The probable cause of the'deformation was mechanical interference with the top nozzle.

At the time of the damage, the fuel assembly was in a non-control element assembly location, and

therefore did not present a system problem.

b. Babcock & Wilcox

While not directly related to the CRD system, B&W cores have experienced several reactor

internal component degradations which posed apotential hazard to CRD insertion if debris either

became lodged in the fuel assembly guide tubes or CRDM internals. Broken fuel assembly holddown

springs, thermal shield bolts and reactor coolant pump fragments were examples of the debris source.

Debris from a damaged reactor coolant pump at Oconne 3 was carried by the coolant into the

core where it became lodged in the fuel assemblies. Though confined mostly to the 'lower end fitting

and first spacer grid, these fragments may have posed a threat to CR insertion if they blocked a guide

tube or wedged in a CRDM. ' ' 

Broken thermal shield support bolts caused primarily by intergranular stress corrosion cracking

also posed a loose parts problem. Oconee 1, Oconee 2, Rancho Seco and Crystal River all reported

similar problems. After detailed metallographic studies, it was concluded that the high degree of cold

reduction induced during fabrication produced grain structure changes in an area of high stress. To

rectify the problem, the lower thermal shield was redesigned the material of the bolts was changed

from A286 to Inconel X750 and locking clips were attached to the bolts.

4-19
*1 ~ Broken fuel assembly holddown springs were identified at Crystal River, Davis Besse and in

Oconee 1 and 3. The Mark B fuel assembly design, as shown on Figure 4.17, utilized one large helical

coil spring, which is positioned on top of the upper end fitting and held in place by a spring retainer.

Inspection of this spring revealed cracks, conclude' that it was caused by low stress high cycle tigue

and stress corrosion cracking. Of all the broken spring identified, none actually interfered with control

rod insertion or withdrawal.

Analyses by B&W concluded that no loose parts would have been generated from the failure.

However, since the control rods are close to the spring when they enter the guide tubes, any part that

may have broken off and become displaced could have interfered with CRA movement.

4.4.6 Miscellaneous

a. Combustion Engineering

Two significant effects were reported during the time period, which did not result in component

failure, but did affect the systems operation.

During implementation of a change in the rod drop time testing procedure, it was discovered

that slow.r than anticipated drop times resulted. Previously, power was interrupted to each mechanism

individually, and the rod drop times met the requirement of the Technical Specification. The new

method interrupted power to all of the rods simultaneously by means of the reactor trip breakers.

This simultaneous power interruption to all of the drive mechanisms led to a slower dissipation of the

stored energy in the holding coils, thus increasing rod drop times.

In response to this, Information Notice 88-47 was issued documenting the slow rod drop times

which violated CE Technical Specification requirements.( 8 While no operational safety problem

resulted, this increase in drop time made it necessary to review all of the accident analysis in the

FSAR to ensure continued conformance to all licensing basis. From the viewpoint of risk, this event

was similar to the loss of redundancy, since the system would not have responded in the required 3

seconds.

The second event documents dropped rods caused by a shorted coil power programmer. The

event was not hardware related, but was caused by water seepage from an overflowing toliet in the

control room. The water seeped into the cable spreading room, where it caused the coil power

programmer to short.

The problem was resolved, and the equipment, cables, and cabinets dried with no further

difficulty. This event highlights the importance of'remaining cognizant of the potential effects to the

system from xterior'events. Moisture, high temperature, or high humidity are stresses, which, if

uncorrected or undetected, could produce aging related failures.

4-20
i U cm fPTwo

TOP VIEW sowtls-~JU11!S g mcEFR GRID- use~~HI

ISS SECIM- ,

LOWER too

FITTING

INS131WAT T1J

Figure 4.17. Babcock & Wilcox fuel assembly.(t 7 )

cm

4-21
5. SUMMARY AND CONCLUSIONS

The Control Rod Drive System for both Combustion Engineering and Babcock & Wilcox

reactors provide both reactor crntrol 2nd plant safety functions. Both systems position the individual

control rod assemblies to control short term reactivity changes, as well as effecting a rapid reactor

shutdown through 'the rapid insertion of the control assemblies. Age related failures for both systems

have occurred, as seen from a review of the operating experience. As a result of these component

failures, loss of system redundancy, dropped rods, and reactor scrams have occurred. The susceptibility

of the system to external influences, such as human error and the operating environment were also

evident.

5.1 Conclusions

5.1.1 Combustion Engineering CEDS

Based upon the system review and the specific operating event reports presented in this study,

the CE control element drive system has experienced age related failures. A minimum of 27% of the

failures were classified as aging related with several specific subassemblies identified as being the main

contributors:

* CED control system: electrical component failures especially power supplies and timing

modules. Component. failures due to overheating also occurred.,

CEDM housing failures: primary coolant leakage principally from the rack and pinion seal

failures have been reported. 'Housing cracks due to contaminant induced stress corrosion

cracking have also occurred. Though likely induced by fabrication, these highlight the

susceptibility of the housings to this type of failure mechanism.

* CEDM coils: coil failures due to electrical and insulation degradation have occurred, caused

primarily by high temperature or corrosion.

* Rod Position'Indication: failures'of the reed switches and the Control Element Assembly

Calculator causing redundancy loss.

Component failures resulted in degraded operation approximately 47% of the time, with a loss

of redundancy in an'additional 25% of the events. The effects on reactor operation due to these

failures was significant. Manual and automatic scrams, ESF actuation,-dropped or slipped CEA's, and

loss of redundancy occurred. In addition to challenging the CED system, stresses were also placed on

other plant systems such as Reactor Protection, Reactor Regulating and other power systems.

The system was also susceptible to' external factors. Examples of'dropped rods and reactor

scrams caused by human error were presented. Transients caused by maintenance on the CED system

have occurred. Seemingly minor operational and procedural changes, such as the elimination of CED

venting resulted in adverse effects.

'Variations in operating environmental -conditions also 'resulted in aging-related failures.

Because of the location of the CEDM coils on the reactor vessel head, continuous forced air cooling

must be provided. ' Lack of cooling has led to overheating of the coils. Cooling must also be provided

5-1
I

for the electrical components in the control and power cabinets. There were numerous examples of

failed components due to insufficient or no cabinet cooling.

A very serious problem is primary coolant leakage from the CEDM. The primary camle of

such leakage has been seal failure. Boric acid in the primary coolant is very corrosive at high

temperatures. Left uncorrected, this leakage has caused corrosion problems for vessel penetrations

and CEDM cooling systems resulting in the generation of an Information Notice by the Nuclear

Regulatory Commission.

5.1.2 Babcock & Wilcox CRDS

The Babcock & Wilcox Control Rod Drive System has also experienced age-related failures

as seen through the review of the operating event reports. A minimum of 45% of the failures

reviewed were aging-related.

The sub-assemblies which were most susceptible to failures were:

* CRDM Coils: stator degradation leading to electrical shorts.

* CRDM Motor Tube: flange leakage due primarily to gasket aging has led to numerous

instances of primary coolant leakage. Failures of the CRDM internals also have jammed the

leadscrew, leading t inoperable control rod assemblies. Similar occurrences have also resulted

in slower-than-expected rod drop times violating the Technical Specifications.

* Control System: failures resulting from dropped rods have occurred due to fuse failures and

environmental degradation in the control cabinets.

* Rod Position Indication System: early failures with the reed switches and the inability of the

analog position indication system to monitor rod position with a failed reed switch resulted in

major improvements in' system design improvements. A more reliable design consisting of

improved reed switches and circuit improvements decreased the occurrences of erroneous rod

position signals. The present system is capable of functioning with a failed reed switch.

System failures resulted in degraded operations 50% of the time, with an additional 27% loss

of redundancy, and significant operational effects have resulted from these failures. Dropped or

slipped rods, reactor power decreases and scrams have all resulted. Operational occurrences such as

these stress not only the CRD system, but other plant systems such as the reactor protection and other

power regulating systems.

System susceptibility to external influences was also demonstrated. Dropped rods due to

excessive temperatures in the area of the reactor vessel head and the accumulation of concrete dust

in control system cabinets have occurred.

5.2 Future Work

The control rods and associated drive systems are essential components ensuring the safe and

reliable operation of a nuclear plant. Based upon the detailed system and operating data presented,

it was shown that both the B&W and CE control rod drive systems experienced age related failures.

Each designs major sub-components that were most susceptible to these failures were identified. These

5-2
failures resulted primarily in component degradation, the loss of system redundancy, reactor power

reductions and scrams.

Although the datmbases have rt provided complete failure information for every eve-, they

have provided sufficient information for this preliminary assessment. Future work will be designed to

augment these findings,

As described in the Introduction, additional information and analysis are required before a

formal Phase I aging assessment can be completed. A detailed Failure Mode and Effect Analysis, or

equivalent, must be performed to qualitatively evaluate the system for component interfaces and their

safety impact. Utility inspection and maintenance procedures must also be assessed for their

effectiveness in identifying component degradation before failure. This is essential if the impact of the

failures upon plant safety is to be minimal.

Detailed questionnaires have been prepared requesting specific utility operating and

maintenance practices and procedures. These have been forwarded to the EPRI Equipment

Qualification Advisory Group through NUMARC. The responses will be evaluated and an assessment

of the effectiveness of the results in identifying and mitigating aging will be performed. Similarly,

preventive maintenance programs will be assessed, with special attention being given to the

subcomponents most susceptible to aging degradation.

5-3
6. REFERENCES

1. Vora, J.P., "Nuclear Plant Aging Research (NPAR) Program Plan," NUREG-1144, Revision

1, September 1987.

2. Fullwood, R., et al., "Aging and Life 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, September 1990.

4. Final Safety Analysis Report, Arkansas Nuclear One,-Unit 2, Docket No. 50-368.

5. Ibid.

6. Final Safety Analysis Report, Waterford Steam Electric Station, Docket No. 50-382.

7. Final Safety Analysis Report, Fort Calhoun, Docket No. 50-285.

8. FSAR, ANO-2, Docket No. 50-368.

9. Final Safety Analysis Report, Arkansas Nuclear One, Unit One, Docket No. 50-313.

10. Ibid.

11. Williams, J., et al., "Control Rod Drive Mechanism Test Program," BAW-10029A, Revision 3,

August 1976.

12. Final Safety Analysis Report, Rancho Seco 1, Docket No. 50-312.

13. FSAR, ANO-1, Docket N3. 50-313.

14. U.S. Nuclear Regulatory Commission, "Standard Technical Specifications for Combustion

Engineering Pressurized vWater Reactors," NUREG-0212, Revision 2, Fall 1980.

15. Koppe, R., et al., "Characterization of the Performance of Major LWR Components," EPRI

NP-5001, January 1987.

16. U.S. Nuclear Regulatory Commission, "Standard Technical Specifications for Babcock & Wilcox

Pressurized Water Reactors," NUREG-0103, Revision 4, 1980.

17. Riggs, R., "Control Rod Guide Tube Wear in Operating Reactors," NUREG-0641, April 1980.

18. USNRC, CE Standard Technical Specification, NUREG-0212, Revision 2, Fall 1980.

6-1
Appendix A

Summary of Licensee Event Reports for Combustion Engineering

Control Element Drive System

1980 - 1989

Table A-1 Cable and Connector LER's -

Table A-2 Coil Power Progranmer LER's

Table A-3 Coil Power Programmer Power Supplies LER's

Table A-4 CEDMCS LER's

Table A-5 CEDM LER's

Table A-6 Rod Position Indication LER's

Table A-7 Human Error LER's

Table A-8 LER's with Unknown Causes

Table A-9 Miscellaneous LER's

A-1
Table A-1. Cable and Connector LERs

Plant | LER No. Failure | Failure Description l___ _(years) __ __ __ __ __ _ }

1 St. Lucie 1 335/80-005 4 Loose Lead CPC 15V 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 ____ ____ ___ ____ _ __ ____ ____ ____ ___ ost twice.

3 St. Lucie 1 335/81-026 5 Cable problem between contaimient and refueling

disconnect panel caused position indication to be

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ ___ ____ __ ____ ___ lo s t .

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 2 361/83-124 1 Faulty connector in CEAC position circuit gave erroneous position indication twice.

7 Waterford 3 382/86-013. 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 .____________________ ____________ _______ ~ ..SPT cable connection.

10 San Onofre 2 361/83-098 1 CEAC inopearable due to loose screw connection on position transmitter input.

A-2
Table A-2. Coil Power Programmer

Plant I ~~~Age atFalr Plant LER No. Faililure Description

1 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-grlpper power

- switch module. - -

3 Calvert Cliffs 2 318/80-010 4 CEA dropped, power reduced due'to faulty timer

module.

4 St. Lucie 1 335/80-007 4 CEA dropped, power reduced due to failed timer

.____________________ . _._:_- _ module. 

5 Hillstone 2 336/80-013 6 CEA dropped due to failed timer module, power

reduced.- -- -

6 San Onofre 2 361/83-096 1 Dirty contacts on CEA timer card caused CEA to

slip and power reduction.

7 San Onofre 2 361/86-018 4 CEA'sLipped due to inherent design deficiency in

software which control DC to'CEA coils. -

8 San Onofre 3 362/84-003 1 Sluggish gripper operation caused CEA to slip'and

subsequent reactor scram. - -

9 Waterford 3 382/86-001 1 Faulty timing module caused CEA drop'and reactor

_______________ ~~~~~trip. --

10 Millstone 2 336/81-038 7 CEA could not be withdrawn due to loose

connection between timer module-and power switch.

11 Millstone 2 336/82-025 8 CEA'could not be withdrawn due to loose

connection between timer module and power switch.

12 San Onofre 2 361/85-031 3 Missing lug nut caused abnormal energization of

power coils, causing CEA subgroup to drop and

L ' reactor trip.

13 Calvert Cliffs 2 -318/82-026 6 -CEA dropped twice. -Timer and upper gripper power

, - _ _switch replaced. Reactor power reduced.

14 Calvert Cliffs 1 317/81-071 7 - Control module failure caused continuous insert

signal resulting in CEA misalignment. Power -

reduction.

15 San Onofre 2 361/83-014 1 Dropped CEA's due to defective coil driver

- . . ' actuation card.

16 San'Onofre 2 361/83-054- 1 '' CEA dropped during surveillance testing due to

- - slow gripper operation. -

17 Arkansas 2 368/81-031 4 - CEA dropped on 2 separate occasions due to poor -

-. w .contact on coil-driver card causing power loss

: - ! * - -and sluggish'upper gripper movement. 

18 Arkansas 2 368/82-004 4 CEA dropped due to sluggish upper gripper.

A-3
Table A-3. Coil Power Programmer

Power Supplies

Age at

Plant LER N. Failure Failure Description

(years)l

1 Calvert Cliffs 1 317/82-045 8 Erratic upper gripper latch action, dropped CEA,

_________ power reduction. Increased HV power supply.

2 St. Lucie 1 335/80-002 4 Failed 15V power supply caused dropped CEA and

power reduction. (Powermate SU-UNI-30A-BV)

3 St. Lucie 1 335/80-023 4 2 CEA's dropped. Same as above.

4 St. Lucie 1 335/80-032 4 CEA dropped twice, timer module and power supply

replaced. Reactor power reduced.

5 St. Lucie 1 335/80-034 4 Failed 15V power supply.

6 St. Lucie 33580-035 4 Failed 15V power supply.

7 St. Lucie 1 335/80-036 4 Failed 15V power supply.

8 St. Lucie 1 335/80-010 4 Failed 15V power supply.

9 St. Lucie 1 335/80-043 4 Failed 5V power supply.

10 St. Lucie 1 335/80-045 4 Failed 15V power supply.

9l St. Lucie 1 335/80-046 4 Failed 15V power supply. Alt power supplies

10_____________________ S t . ______ replaced with origlnl Powermate ulNI-88.

12 St. Lucie 1 335/80-048 4 Failed 15V power supply. All power supplies

replaced with original Powermate UNI-88.

13 St.Lucie 1 335/80-049 4 Failed 15V power supply. All power supplies

replaced with original Powermate UI-88.

14 St. Lucie 1 335/80-00 4 While changing power supplies, voltage spike led

___________________ _______ to 2 CEA's dropping. Reactor manually tripped.

15 St. Lucie 1 335/80-051 4 Fuse blew in alternate power supply line, CEA

dropped.

16 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.

17 St. Lucie 1 335/81-020 5 Dropped CEA, reduced power, replaced power

_ . _________ .___ supply

18 Millstone 2 336/82-041 8 CEA dropped twice, power reduction, failed 15V

____________ ________ DC power supply (Lambda Eec LCD-A-22).

19 Maine Yankee 309/84-001 12 Failed power supply caused dropped rod and

reactor shutdown.

V.

A-4
Table A-4. CEDMCS

Age at

Plant LER et% Failure Failure Description

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _(y ea rs ) _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

1 Arkansas 2 368/83-040 5 Dropped CEA due to a bLown fuse.

2 St. Lucie 1' 335/80-038 4 Failed timing module which overheated due to.

-_________ _ ventilation fan turned off for maintenance.

3 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.

4 St. Lucie 1 335/82-056. 6 CEA motion inhibit circuit for CEA out of

sequence deviation scanner failed due to loose

________ ___ ___________ _______reLay .

5 St. Lucie 1 335/82-055 6 CEA motion inhibit circuit for CEA out of

sequence deviation scanner failed due to loose -_________ .relay. _ 

6 St. Lucie 1 335/81-030 5 CEA motion inhibit circuit for PDIL circuit

faile& due to failed relay.

7 MiLLstone 2 336/82-027 8 During CEA surveillance testing, CEA motion

inhibit for all CEA groups became inoperable due

to failed Logic chip.

8 Millstone 2 _ 336/82-030 8 CEA motion inhibit interlock did not function

due to faulty operational ampLifier.

-9 San Onofre 3 362/85-020. 2 -Dropped CEA due to blown fuse in the hold bus

__________ . logic circuit. Subsequent reactor trip.

10 Waterford 3 383/86-002 1 Dropped CEA due to cabinet cooler switch being

turned off, causing overheating of circuits.

Reactor trip.

11 San Onofre 2 361/83-090 1 Slipped CEA due to poor connection on CEDMCS

- power switch.

12 Calvert Cliffs 1 317/81-066 .7 While troubleshooting CEDMCS, a ogic-module'

which was inserted caused spurious signals which

____ caused rod drop due to a failed off switch on

the control panel.

13 Waterford 3- 382/89-017 4 Unable to withdraw CEA due to control circuitry

l______ _problems. Reactor trip.

-14 Calvert Cliffs 2 .318/83-019 7 Circuit burn in of newly installed circuit

boards caused PDIL function to be inoperable.

15.. Calvert Cliffs 2 .318/83-027 7 Failure of PDIL uctioneering card output semiconductor rendered CEA motion inhibit

_ ______________________ -____________ _________ inoperable.

A-5
Table A-5. CEDM

Age at

Plant LER No. Failure Failure Description

.________ _ .(years)

1 Palisades 255/81-049 10 Plant shutdown due to excessive coolant Leakage

past improperly installed CEDN seal housing

gasket.

2 Palisades 255/84-024 13 Failed CEOM seal housing. Plant brought to cold

shutdown.

3 Palisades 255/85-006 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 renting during

primary system fill. Reduced power.

7 Calvert Cliffs 2 318/80-048 4 2 dropped CEA's 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 2 336/88/008 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

CEA's to drop. Plant shutdown.

12 Millstone 2 336/88-009 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 2 361/83-102 1 Malfunction of upper gripper coil stack caused

dropped CEA.

14 Arkansas 2 368/84-026 6 Shorted upper gripper coil caused CEA to drop.

Reactor trip.

15 Waterford 3 382/86-023 1 Failed ower 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.

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.

A-6
Table A-6. Rod Position Indication

Plant' T .ER No. Faieua Failure Description

_ _ _ _ _ _ _ _ _ _ _ _ _ j _ _ _ _ _ _ _ j (years) ] 

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 1 317/83-008 9 Failed RSPT's gave erratic position Information

_ ____________________ on 2 CEA's (Electro MechanicsN9027-1).

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 RSPT's 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 3i8/83-075 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 a. v/81-00;- 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 - Circuit card failure in plant computer-caused

pulse counting indication system to be

inoperable. - -

19 Millstone 2 336/81-037 7 Blown resistor in plant computer 36V power

supply caused pulse counting position indicating

system to become inoperable.

A-7
Table A-6 (Cont'd.)

LER Age at

Plant LER No. Failure Failure Description

l ~~~~~~~~~~~~(years)l

20 San Onofre 2 361/83-011 1 Erroneous rod position indication caused by

faulty isolation amplifier card.

21 San Onofre 2 361/83-087 Faulty reed switch produced erroneous rod

1 position indication.

22 San Onofre 2 361/86-027 4 Loose solder joint on RSPT produced wrong

position ndications eading to reactor trip.

23 San Onofre 2 361/89-019 7 Misalignment of RSPT associated with full in

lights caused Lights not to coe on when CEA

fully inserted.

24 Arkansas 2 368/80-053 2 Software problem caused failure of CEAC.

25 Arkansas 2 368/80-058 2 Software problem caused failure of CEAC.

26 Arkansas 2 368/80-080 2 Data link input card on the optical isolator and

failed test circuit module card caused CEAC

failure.

27 Arkansas 2 368/82-005 4 FaiLure of high level HJX gate card caused CEAC

to show incorrect CEA positions.

28 Arkansas 2 368/83-029 5 Failure of high level MX card caused CEAC

_ ~~~~~~~~~~~~~~~failure.

29 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.

30 Arkansas 2 368/82-009 4 CEAC failed during excore instrumentation test.

31 Arkansas 2 368/82-040 4 CEAC failed, possibly as a result of Lighting

______________ storm.

32 Uaterford 3 382/86-009 1 Reactor trip caused by reed switch failure.

33 St. Lucie 2 389/83-047 7 Pulse counting CEA position system inoperabledue to Loss of computer.

34 San Onofre 3 362/84-024 1 Intermittent failure on computer board caused

CEAC malfunction and reactor trip.

35 Palo Verde 1 528/89-004 4 CEAC inoperable due to failed processor board

caused reactor scram.

36 San Onofre 2 361/84-043 2 Failed power supply in analog CEA position

indication system, caused reactor scram and ESF

actuation.

37 San Onofre 2 361/84-019 2 Faulty CEAC caused spurious position signals

_ ______________________ _____________ which caused reactor trip.

38 CaLvert Cliffs 1 317/81-081 7 Primary CEA position indication Lights and

analog system malfunctioned.

A-8
Table A-6 (ont'd.)

Plant | LER Failure Failure Description

Plant I-years)

39 San Onofre 2 361/83-041 1 CEAC inoperable due to faulty isolation

. [ J amplifier card due to overheating in the

.cabinet. - ' ' . ' ' ' I 

, i-

. . I

A-9
Table A-7. Human Error

1 j ~~Age at

Plant | LER No. Failure Failure Description l____________ (years) i

1 Calvert Cliffs 2 318/80-015 4 WhiLe changing backup power supply, the prime

15V 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

CEA's instead of unit 1, 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 15V 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 3 362/83-086 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 P 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 3 382/87-012 2 Inadequate procedure allowed both CEAC's 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 CEA's.

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 SO, 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.

17 Arkansas 2 368/80-057 2 CEAC inputs to CPC were inoperable due to

I______________I______ I____________ I_________ personnel error, during maintenance.

A-10
. I . s 

Table A-7 (Cont'd.)

I I ~A ge at' If

Plant | -:R No. Failure | Failure Description I (years) 

18 St. Lucie 2 - 389/83-074 - ; 7 Pulse counting CEA position indicating system

I I______I_ jinoperable due to programming error.

I 

A-11
Table A-8. Unknown Cause

Age at

Plant LER No. Failure Failure Description

____ ___ ___ ____ __ _ _ ____ ___ ___ (years)

I 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. _______ _____ _ 

4 St. Lucie 1 335/82-069 6 While at full power, CEA dropped for unknown

reasons.

5 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 St. Lucie 1 335/85-005 9 While at full power, rod dropped for unknown

reason.

9 San Onofre 2 361/83-155 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.

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.

V

A- 12
Table A-8 (Cont d.)

I ~~Age at 

Plant LER . FaiLure Failure Description

(years)-

22 Calvert Cliffs 1 317/82-036 8 During start up tests, rod dropped for no reason. -

23 Calvert Cliffs 2 318/80-009 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 CEA's have dropped into core with _ ____________________ ____________ subsequent power reduction. No cause found.

25 St. Lucie 2 389/87-005 4 While at 100X power, 2 CEA's dropped into core I _____________________ ____________ _________ for no apparent reason. Reactor shutdown.

A-13
Table A-9. Miscellaneous

Plant LER No. Fafture Failure Description

l_____________________ ____________ (years)

1 Millstone 2 336/83-026 9 Top nozzle damage caused guide tube deformation.

2 CaLvert CLiffs 2 318/83-060 7 Water from overflowing control room toilet

seeped nto cable spreading room where t

shorted out coil power programme components.

-

A- 14
Appendix B

Summary of Licensee Event Reports for Babcock & Wilcox

Control Rod Drive System

1980 - 1989

Table B-1 Cable and Connector LER's

Table B-2 CRD Control System LER's

Table B-3 CRDM LER's

Table B-4 Rod Position Indication LER's

Table B-5 Human Error LER's

Table B-6 Potential Loose Parts LER's

B-1
Table B-i. Cable and Conector LER's

Age at

Plant LER No. Failure Failure Description

1 Oconee 1 269/80-027 Power reduction resulted from dropped rod. C phase on stator

opened due to loose connector.

2 Oconee 1 269/87010 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

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
T C

Table B2. RD Control System LER's

Age at

Plant T.ER No. Failure Failure Description

1 Arkansas 1 - 313/8342 - Control rod group misalignment due to out movement

restriction resulting in pole slippage.

2 Arkansas 1 31318840 14 : Control rod group dropped due to malfuncion of power

sequencer programmer. Reactor trip.

3 Davis Besse 346/804232 3 Failed 24 VDC power supply in programmer controller in

group SCR supply cabinet caused improper rod movement.

Reactor trip.

4 Davis Besse 346/82011 5 Control rod drop due to blown fuse in transfer switch

module. Power reduction.

S Davis Besse 346/83014 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 346/83068 6 Rod drop d We 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 APSRA's inoperable. Reactor

._______________ II 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.

B-3
Table B-3. CRDM LERs

1Age atj

Plant LEP "o. Failure Failure Description

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 IS drop time requirement

-

B14
Table B-4. Rod Position'Indication LER's

I ~~~Agenat

Plant LFR l7- Failure Failure Description

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ~(years) _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

1 Crystal River 302182-035 5 Reed switch'failure rendered API system inoperable.

2 Crystal River 302/83-06 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 302186-011 9 Relay failure in RPI circuitry rendered system inoperable.

6 Davis Besse 346/80-004 3 Lost rod position indicator due to failed reed switch caused by

excessive high temperature.

7 Davis Besse 346/80-013 3 API malfunction caused by b!awn fuse in power supply.

_____ _____ _ ___ ____ Delayed SU.

8 Davis Besse 346/80-015 3 API system inoperable due to reed switch failure causcd by

.____________ ________ excessive high temperature.

9 Davis Besse 346/-025 3 Excessive high temperatures caused reed switch failure. API

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 LERs

Age at

Plant LER No. Failure Failure Description

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ (years)

Davis Besse |346/88-029 11 Rod drop due to maintenance error while performing work on I___I CRDCS. Power reduction.

B-6
Table B-6. Potential Loose Parts LER's

I ~~~Age at

Plant | .ER No. Failure Failure Description

1 Oconee 1 269/8"015 7 Broken fuel assembly holddown springs.

2 Oconee 1 269181-011 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.

5 Oconee 3 287/82-007 8 Broken fuel assembly holddown springs.

6 Oconee 3 287/82-08 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 346188-015 11 Debris found in reactor vessel prior to refuel.

B-7