ML20138C702

Insp Rept 50-302/86-09 on 860208-0307.Violation Noted: Failure to Have Adequate Procedures for Conducting Preventive Maint on Emergency Diesel Generators & to Have Two Members of Plant Mgt Approve Procedure OP-404 Change

ML20138C702
Person / Time
Site:	Crystal River
Issue date:	03/26/1986
From:	Elrod S, Stetka T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML20138C680	List: ML20138C676 ML20138C683 ML20138C702
References
50-302-86-09, 50-302-86-9, IEB-85-001, IEB-85-1, IEIN-85-071, IEIN-85-71, NUDOCS 8604020574
Download: ML20138C702 (14)

See also: IR 05000302/1986009

Text

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I: * P IEI UNITED STATES

o NUCLEAR REGULATORY COMMISSION

y' , REGION 11

g ,j 101 MARIETTA STREET.N.W.

• * ATI.ANTA, GEORGI A 30323

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Report No.: 50-302/86-09

Licensee: Florida Power Corporation

l 3201 34th Street, South

i St. Petersburg, FL 33733

Docket No.: 50-302 Licensee No.: DPR-72

Facility Name: . Crystal River 3

Inspection Dates: Februaryi - March 7, 1986

Inspec  :

T. F. Stetka, Senior Resident Inspector

3k

Date' Signed

Et,

AccompanyingPe7onn ,/

J. Tedrow, Resident Inspector

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Approved by: // [ j N/[P

Sf W. Elrod, S$ tion Chief /Date' Signed

Division of Redctor Projects

SUMMARY

Scope: This routine inspection involved 199 inspector-hours on site by two

resident inspectors in the areas of plant operations, security, radiological

controls, Licensee Event Reports and Nonconforming Operations Reports, facility

modifications, IE Bulletin and Information Notices, cold weather preparations,

offsite review committee activities, review of the public document room' and

licen.cea action on previous. inspection items. Numerous facility tours were

condut'ed and facility operations observed. Some of these tours and observations

were conoucted on~backshifts.

Results: Two violations were identified: Failure to hav'e adequate procedures

for' conducting preventive maintenance on the Emergency Diesel Generators,

paragraph 5.b.9.c; Failure to have two members of the plant management staff

approve a change to procedure OP-404, paragraph 5.b.10.

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B604020574 860327

PDR ADOCM 05000302

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REPORT DETAILS

1. Persons Contacted

Licensee Employees

• P. Breedlove, Nuclear Records Management Supervisor

• C. Brown, Assistant Nuclear Outage & Modification Manager

• J. Bute, Nuclear Compliance Specialist

• M. Collins, Nuclear Safety & Reliability Superintendent

• J. Cooper, Superintendent Nuclear Sury. & MAR Functional Testing

• D. Fields, Nuclear Quality Engineering Supervisor

• F. Haines, Nuclear Engineer II

• V. Hernandez, Senior Nuclear Quality Assurance Specialist

B. Hickle, Nuclear Chemistry & Radiation Protection Superintendent

J. Lander, Nuclear Outage & Modification Manager

• P. McKee, Nuclear Plant Manager-

• S. Powell, Senior Nuclear Licensing Engineer

• V. Roppel, Nuclear Plant Engineerir.g & Technical Services Manager

• W. Rossfeld, Nuclear Compliance Manager

• P. Skramstad, Nuclear ~ Chemistry / Radiation (Chem / Rad) Protection

Superintendent

'*D. Smith, Nuclear Maintenance Superintendent

• E. Welch, Nuclear Plant Engineering Superintendent

K. Wilson, Manager, Site Nuclear Licensing

• R. Wittman, Nuclear Operations Superintendent

Other personnel contacted included office, operations, engineering,

maintenance, Chem / Rad and corporate personnel.

• Attended exit interview

2. Exit Interview

The inspectors met with licensee representatives (denoted in paragraph 1) at

the conclusion of the inspection on March 7, 1986. During this meeting, the

inspectors summarized the scope and findings of the inspection as they'are .

detailed in this report, with particular emphasis on the Violations and

Inspector Followup-Items (IFI).

The licensee representatives acknowledged the inspectors' comments and did

not identify as proprietary any of the materials provided to or reviewed by

the inspectors during this inspection.

3. Licensee Action on Previous Inspection Items

(0 pen) IFI 302/86-07-02: The licensee is continuing to inspect Rosemount

transmitters in the reactor building and has identified 22 transmitters thus

far that are 'in need of repair. The licensee attributes the cause for the

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loose electrical connections to _ inadequate installation instructions which -

failed to specify adequate tightening of those connections. The licensee is

presently evaluating new methods to ensure that the electrical connections

will be' properly tightened and will revise the installation instructions

accordingly.

(Closed) Unresolved Item 302/85-11-02: The licensee has revised procedure

CP-115, In-Plant Equipment Clearance and Switching Orders, to clarify the

use of equipment clearances to allow valve alignments different from that

required by procedure. The licensee will use equipment clearances to remove

degraded systems or components from service for repairs or testing. System

valve lineup changes -for other purposes will be made by approved changes to

the appropriate procedure. The inspector has reviewed this policy and

considers it ' appropriate to avoid future confusion in this area by plant

personnel.

(Closed) IFI 302/85-44-02: Review of this item by the licensee indicates

that . positive position indication does exist for both valves. It was

determined that if more than one valve position indication light was

illuminated, it indicates that the valves are not in the same position.

. Review ~ of the valve position indicator logic diagram by the inspector

confirms the licensee's conclusions. Action on this item is considered to

be complete.

(Closed) IFI 302/84-30-05: The licensee has issued a new procedure, entitled

" Florida Power Corporation, Crystal River Coal Plant, 50 2 Emergency

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Procedure", that requires immediate notification of the nuclear plant (CR-3)

Shift Supervisor if a leak in the sulfur dioxide (50 ) system

2 has occurred.

In addition, coal plant personnel have replaced the gasket on the 502 tank ,

with an improved gasket material to provide a higher integrity seal.

4. Unresolved Items

Unresolved items were not' identified in this inspection period.

5. Review of Plant Operations

The plant remained in the cold shutdown condition (Mode 5) for the duration

of this inspection period.

a. Shift Logs-and Facility Records

The inspector reviewed records and discussed various entries with

operations personnel to verify compliance with the Technical

Specifications (TS) and the licensee's administrative procedures.

The following records were reviewed:

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Shift Supervisor's Log; Reactor Operator's Log; Outage Shift Manager's

Log; Shift' Relief Checklist; Auxiliary Building Operator's Log; Active

Clearance Log; Daily Operating Surveillance Log; Short Term

Instructions (STI); and Selected Chemistry / Radiation Protection Logs.

In addition to these record reviews, the inspector independently

verified clearance order tagouts.

No violations or deviations were identified.

b .- Facility Tours and Observations

Throughout the inspection period, facility tours were conducted to

observe operations and maintenance activities in progress; some of

these observations were conducted during backshifts. . Also, during this

inspection period, licensee meetings were attended by the inspector to

observe planning and management activities.

The facility tours and observations encompassed the following areas:

security perimeter fence; central alarm station; control room;

emergency diesel- generator room; auxiliary building; intermediate

building; reactor building; battery rooms; and electrical switchgear

rooms.

During these tours, the following observations were made:

(1) Monitoring Instrumentation - The following . parameters were

observed to verify compliance with the TS for the current

operational mode:

Equipment operating status; area atmospheric and liquid

radiation monitors; electrical system lineup; reactor

operating parameters; and auxiliary equipment operating

parameters.

No violations or deviations were identified.

(2) Safety Systems Walkdown - The . inspector conducted a walkdown of

the Decay Heat Removal (DHR) system to verify that the lineup was

in accordance with license requirements for system operability and

that the system drawing and procedure correctly reflect "as-built"

plant conditions.

The inspector made the following observations:

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two differential pressure transmitters, OH1-dPT3 and

DH1-dPT4, appear to have switched identification labels from

that specified on system drawing F0-302-641; and

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the vacuum breaker on top of the Borated Water Storage Tank

was operating even though there were no evolutions in -

progress to cause this operation.

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The inspector discussed these matters with licensee personnel who

verified .the inspector's observations. The licensee investigated

the reason for the vacuum breaker operation and found the setpoint

for operation of this device to be set too low. A work request

(WR), #77371, was initiated to correct the problem. The. inspector

will verify implementation of the licensee's corrective action

during subsequent routine inspection activities.-

(3) Shift Staffing - The inspe,' tor verified that operating shif t

staffing was in accordance 'with TS requirements and that control

room ~ operations were being conducted in an orderly and

professional manner. In addition, the inspector observed shift

turnovers on various occasions to verify the continuity of plant

status, operational problems, and other pertinent plant

informavion during'these turnovers.

No violations or deviations were identified.

(4) Plant Housekeeping Conditions - Storage of material and components

and cleanliness conditions of various areas throughout the

facility were observed to determine whether safety and/or fire

hazards existed.

No violations or deviations were identified.

(5) Radiation Areas - Radiation Control Areas (RCAs) were observed to

verify proper identification and i nplementation. These

observations included selected licensee-conducted surveys, review

of step-off pad conditions, disposal of contaminated clothing, and

area ' posting. Area postings were independently verified for

accuracy through the use of - an NRC radiation monitoring

instrument. The inspector also reviewed selected radiation work

permits and observed the use of protective clothing, respirators,

and personnel monitoring devices to assure that the licensee's

radiation monitoring policies were being followed.

On February 24, while reviewing the Outage Shif t Manager's (OSM)

Log, the inspector noted that decontamination efforts in the

reactor building on February 22 had caused excessive airborne

activity. The NRC Region II Office dispatched a team to the site

to investigate the occurrence.

Details on this investigation and the results will be delineated

in NRC Inspection Report 50-302/86-11.

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(6) Security Control - Security controls were observed to verify that

security barriers were intact, guard forces were on duty, and

access 'to the protected area (PA) was controlled in accordance

with the facility security plan. Personnel within the PA were

observed to verify proper display of badges and that personnel

requiring escort were properly escorted. Personnel within vital-

areas were observed to ensure proper authorization for the area.

No violations or deviations were identified.

(7) Fire Protection - Fire protection activities and equipment were ,

observed to verify that fire brigade staffing was appropriate and

that fire alarms, extinguishing equipment, actuating controls,

fire fighting equipment, emergency equipment, and fire barriers

were operable.

No violations or deviations were identified.

(8) Surveillance - Surveillance tests were observed to verify that

, approved procedures were being used; qualified personnel were

4 conducting the tests; tests ' were adequate tu verify -equipment

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operability; calibrated test equipment was utilized; and TS

, requirements were followed. .

The following tests were observed and/or data reviewed:

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SP-130, Engineered Safeguards Monthly Functional Tests;

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SP-132, Engineered Safeguards Channel Calibration;

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SP-210, ASME Class 3 Hydrostatic Testing;

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SP-335, Radiation Monitoring Instrumentation Functional

Test;

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SP-350, Turbine-Driven Emergency Feedwater Pump 38 Over-

speed Trip Test;

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SP-354, Emergency Diesel Fuel Oil Quality & Diesel

Generator Monthly Test;

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SP-523, Station Batteries Service Test;

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SP-605, Emergency D'Nsel Generator Engine Inspection /

Maintenantc; and

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SP-904, Calibration of 4160 Volt ES Bus Degraded Grid

Relays.

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During observation of the maintenance and testing on the Emergency

Diesel Generator (EDG), the inspector noted that the local diesel

engine tachometers were _ tagged as out of calibration. .These

tachometers were subsequently calibrated prior to EDG post-

maintenance testing. Durin'g the post-maintenance acceptance runs

conducted in accordance with procedure SP-354 (A and B), the

tachometers were again found to be out of calibration. The

licensee has had a continuing problem maintaining these

tachometers in calibration and is presently investigating the

possibility of replacing these tachometers with ones of a

different design.

Inspector. Followup Item (332/86-09-01): Review the licensee's

activities to replace the local tachometers on EDGs "A" and "B".

(9) Maintenance Activities - The inspector observed maintenance

activities to verify that correct equipment clearances were in

effect; work requests and fire prevent. ion work permits, as

required, were issued and being followed; quality control

personnel were available for inspection activities as required;

and TS requirements were being followed.

Maintenance was observed and work packages were reviewed for the

following maintenance activities

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Replacement of the speed changer motor on the governor for

the "B" Emergency Diesel Generator (EDG-1B);

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Periodic maintenance and inspection of EDG-1B, including main

bearing replacement and post-maintenance testing in

accordance with procedure SP-605;

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Troubleshooting and replacement of electrical cable for the

EDG-1B overspeed trip relay in accordance with procedure

MP-531;

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Rebuilding of the "B" Decay Heat Pump (DHP-18) rotating

assembly in accordance with procedure MP-131;

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Oil slinger ring and bearing replacement (due to bearing

failure), in the turbine of Emergency Feedwater Pump 2

(EFP-2) in accordance with procedure MP-162;

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Periodic electrical checks on the generator of EDG-1A in

accordance with procedure PM-123;

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Rerlacement of a " SLUR" relay for EDG-1B and post-maintenance

ttsting in accordance with procedure SP-904;

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Replacement of impeller capscrews on the "B" Reactor Coolant

Pump in accordance with Modification Approval Record (MAR)

86-02-10-02 and engineering instructions; and

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Perio'dic maintenance of the "B" Reactor Coolant Pump motor in

accordance with proc.cdure MP-172.

As a result of these observations the following items were

identified:

(a) As part of the troubleshooting to correct the turbine bearing

failure on 'EFP-2, an oil plug was installed on the bearing.

The inspector discussed the purpose of this plug with

licensee representatives who stated that the plug

installation was intended to increase the amount of oil

- residing between the shaft and the bearing thus providing

more lubrication to prevent excessive temperatures and

bearing wear during operation. Licensee representatives

stated that this " experiment" was still being evaluated to

determine its effectiveness in preventing bearing failure. A

decision as to whether to remove the plug or document its

installation with a MAR will be made at the completion of

this evaluation.

During observation of the post-maintenance and overspeed trip

tests of EFP-2, which were conducted by recirculating the

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discharge water of EFP-2 back to the Condensate Storage Tank

through a small recirculatio'n line (11/2" diameter), the

inspector noted that the test was conducted at a pump speed

of approximately 115% of rated speed which created a

discharge pressure of approximately 2,000 psig. The licensee

informed the inspector that 'due to this high pressure, they

were evaluating the pipe stresses resulting from the

operation of the pump in the recirculation mode and had

temporarily readjusted the overspeed trip setpoint for the

pump to 105% of rated speed until completion of the pipe

stress analysis.

Inspector Followup Item (302/86-09-02): Review the

licensee's activities to place an oil plug in the EFP-2

bearing and determine the proper overspeed trip setpoint as a

result of the piping analysis.

(b) During the repair on the DHP-1B rotating assembly, the

licensee discovered that the pump shaft had broken in the

impeller region of the shaft. The licensee is evaluating i.hc

failure mechanism for this break and plans to have this

evaluation complete by April 4,1986. During the alignment

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of the replacement pump, the licensee identified several pipe

hangers in the "B" decay heat pit that were damaged. These

pipe hangers support the Decay Heat Removal system inlet

piping to DHP-18. They have been repaired by the licensee

and DHP-1B has been satisfactorily- aligned and declared

operable. The licensee is evaluating the possible effect

these pipe hangers had on the failure of the pump shaft.

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Inspector Followup Item (302/86-09-03): Review the

licensee's determination of the shaft breakage on DHP-18.

(c) On February 26, while observing the performance of preventive

maintenance on the "A" Emergency Diesel Generator (EDG) in

accordance with PM-123, Periodic Electrical Checks of

Emergency Diesel Generators, the inspector noted a

discrepancy between the insulation resistance (megger)

readings taken for the rotor and the stator in that three

resistance readings were taken on the stator and two

resistance readings were taken on the rotor. Since the EDGs

are three phase machines, the inspector questioned why three

resistance readings were not taken on the rotor. The

maintenance personnel responded that they were following the

procedure (PM-123) and that step 7.5.2 required the rotor to

be meggered as described in step 7.5.1.1. Step 7.5.1.1

indicated that the meggering should be done from the brush

spring retainers (of which there are only two on the

machine).

Subsequent discussions with supervisory personnel indicated

that the insulation resistance readings were taken

incorrectly, i.e., the measurement should have been taken

from the rotor and not from the brush spring retainers. If

the readings had been taken from the rotor, only a single

resistance reading would have been recorded on the data

sheet.

A review of the data from PM-123 that was performed on the

"B" EDG appeared to indicate that this data had been taken

cor-ectly since only a single resistance reading was

recorded. However, further investigation indicated that this

PM-123, which was performed approximately a month earlier by

the same individual, also may have been performed incorrectly

since the plant's general meggering procedure (PM-105) allows

multiple leads to be tied together for the insulation

resistance check. Threfore, the two brush spring retainers

could have been tied together to provide the single

resistance reading and the rotor still would not have been

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Failure to have an adequate procedure to perform insulation

resistance checks is contrary to the requirements of TS 6.8.1.a and Regulatory Guide 1.33 and is considered to be a

violation.

Violation (302/86-09-04): Failure to have an adequate

procedure for conducting preventive maintenance on the

Emergency Diesel Generators.

(10) Radioactive Waste Controls - Selected liquid releases

and solid waste compacting were observed to verify that approved

procedures were utilized, that appropriate release approvals were

obtained, and that required surveys were taken.

On February 12, 1986, while reviewing liquid release permit number

L-1986-95 for the reiease of Evaporator Condensate Storage Tank

(WDT-10A) contents, the inspector noticed special instructions on

the permit that required valve RW-33 to be closed. This valve is

one of two cross connects which tie together the discharges of the

two Decay Heat Seawater trains. The Decay Heat Seawater system is

the ultimate heat sink for the Decay Heat Removal system. The

inspector reviewed the operating procedure, OP-407A, used to

perform this release and did not find instructions in this

procedure to position valve RW-33. Following verification that

RW-33 was closed, the inspector discussed this matter with

licensee representatives to determine how the positioning of the

valve was being controlled. As a result of this discussion,

licensee representatives made an immediate temporary change to

OP-407A to account for repositioning of this valve and to ensure

that the valve was returned to the open position following.the

completion of the release.

Procedure OP-404, Decay Heat Removal System, .ection 11.0,

specifies the Engineered Safeguards (ES) normal standby mode valve

lineup for the Decay Heat Seawater system. This lineup requires

valve RW-33 to be open. Technical Specification (TS) 6.8.3

allows temporary changes to be made to operating procedures

provided the original intent of the procedura is not altered and

the change is approved by two members of the plant management

staff, at least one of whom holds a Senior Reactor Operator (SRO)

license. The licensee has a method for implementing a temporary

change that is called an Immediate Temporary Change (ITC); this

change method is delineated in Administrative Instruction AI-401.

The positioning of valve RW-33 by the direction of release permit

instructions represents a change made to procedure OP-404 without

the approval of two members of the plant management staff, at

least one of whom holds an SR0 license; this is considered to be a

violation.

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Violation (302/86-09-05): Failure to have two members of the

plant management staff approve a change to procedure OP-404 as

required by TS 6.8.3.

(11) Pipe Hangers and Seismic Restraints _- Several pipe hangers and

seismic restraints (snubbers) on safety-related systems were

checked to ensure that fluid _ levels were adequate and no leakage

was evident, that restraint settings were appropriate, and that

anchoring' points were not binding.

No violations or deviations were identified.

6. Review of Licensee Event Reports and Nonconforming Operations Reports

a. Licensee Event Reports (LERs) were reviewed for potential generic

impact, to detect trends, and to determine whether corrective actions

appeared appropriate. Events, which were reported immediately, were

reviewed as they occurred to determine if the TSs were satisfied.

LERs 83-50, 85-04, and 86-02 were reviewed in accordance with current

NRC policy and are closed.

LER 83-50 was held open pending resolution of testing of the Low

Pressure Injection (LPI) system and testing of components actuated by

the ES system. Amendment #79. to the TS and a revision of the

licensee's commitments regarding low temperature overpressure

protection resolved this issue.

LER 85-04 was held open pending verification of the extent of hanger

discrepancies and until a plant operability determination could be

made. Revision 4 to this LER provided this verification and

determination. The licensee's consulting engineer's report was

reviewed by the inspector to verify these conditions.

b. The inspector reviewed Nonconforming Operations Reports (NCORs) to

verify the following: compliance with the TS, corrective actions as

identified in the reports or during subsequent reviews have been

accomplished or are being pursued for completion, generic items are

identified and reported as required by 10 CFR Part 21, and items are

reported as required by the TS.

All NCORs were reviewed in accordance with the current NRC Enforcement

Pol. icy.

NCORs.86-29, 86-30, 86-36 and 86-40 reported several remote shutdown,

post-accident monitoring, and ES transmitters that were found out of

calibration during performance of surveillance tests. These tests are

generally conducted every refueling outage (every 18 months) and were

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performed during the licensee's last refueling ~ shutdown which ended in

July 1985. The licensee is reperforming these tests now to extend the

operating cycle of the plant once the present maintenance outage is

completed. The occurrence of these out of tolerance instruments in the

relatively short time interval since the last calibration

(approximately 7-8 months) creates the possibility that these

instruments may drift excessively and may not remain accurate for the

duration of the 18-month calibration interval. The licensee is

evaluating the accuracy of these instruments to determine appropriate

corrective action.

Inspector Followup Item (302/86-09-06): Review the licensee's

evaluation of the accuracy of . instruments identified in NCORs 86-29,

86-30, 86-36 and 86-40.

7. Design Changes and Modifications

The installation of 'new or modified systems was reviewed to verify that the

' changes were reviewed and approved in a:cordance with 10 CFR 50.59, that the

changes were performed in accordance with technically adequate and approved

procedures, that subsequent test results met acceptance criteria and

deviations were resolved in an acceptable manner, and that appropriate

drawings and facility procedures were revised as necessary. This review

included selected observations of modifications and/or testing in progress.

The following Modification Approval Record (MAR) was reviewed and/or

associated testing observed:

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MAR 84-01-16-01, Replacement of makeup system vents and drains with

butt welded valves.

No violations or deviations'were identified.

8. Review of IE Information Notices (IEN) and Bulletins (IEB)

The inspectors reviewed the following IEN and IEB and verified that these

items were reviewed by the licensee and appropriate actions were taken:

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IEB 85-01, Steam Binding of Auxiliary Feedwater Pumps; and

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IEN 85-71, Containment Integrated Leak Rate Tests.

As a result of this review, IEN 85-71 is considered to be closed. With

regards to IEB 85-01, the licensee's action is considered to be complete.

However, upon reviewing the venting procedure for EFP-2 contained in

procedure OP-605, Feedwater System, the inspector noticed that the

= established vent path was inadequate to properly vent the pump. A vent

valve (EFV-28) located downstream of the pump isolation valve (EFV-8) was

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specified as the vent flow path instead of a vent path which would actually

vent the pump and not the downstream piping. The inspector discussed this

matter with operations personnel who concurred with the inspector's

observation. The licensee plans to change procedure OP-605 to reflect the

proper vent path through valve EFV-29. This item will continue to be

tracked by existing Inspector Followup Item 302/85-21-05.

9. Review of Offsite Review Committee Activities

The inspector attended meetings and reviewed the activities of the

licensee's offsite review committee, the Nuclear General Review Committee

(NGRC). This review included a determination that TS requirements were

being met with regard to the following:

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committee quorum;

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committee composition with respect to disciplines and expertise;

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qualification of committee members; and

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review activities of the committee.

No violations or deviations were identified.

10. Cold Weather Preparations

The inspector reviewed the maintenance history for space heaters and heat

tracing on systems which are susceptible to freezing. The inspector

verified that cold weather protection measures were reestablished after

maintenance was performed on these systems.

Also, as part of this inspection, discussions were held with licensee

personnel to verify that areas of the plant normally heated during plant

operation were adequately protected during periods of prolonged plant

shutdown.

Generally, the climate at the Crystal River facility is such that freezing

temperatures do not occur for extended periods of time and do not generally

create freezing problems.

No violations or deviations were identified.

11. Visit to the Public Document Room (PDR)

The inspector visited the community's PDR on February 13, 1986. The

inspector examined the type of information available, the condition of this

information, and the filing system used for access to this information. The

inspector selectively reviewed both the microfiche and hardcopy files of

various documents. The inspector found the PDR to be orderly and

l accessible.

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.12. Reactor Coolant Pump (RCP) Repairs

Since the shutdown of the plant on January 1,1986, due to the shaft breaking

on the "A" RCP (Ref: NRC Report 50-302/85-44, paragraph 8.b. ), and the

confirmation of cracks in the "B" RCP shaft (Ref: NRC Report 50-302/86-07,

paragraph 7.b. ), additional ultrasonic testing (UT) has revealed crack

indications in both the "C" and "D" RCP shafts. As a result of these

findings, the licensee decided, on March 5, to replace the shafts in these

pumps as well.

Preliminary investigation into the shaft breakage and cracking has revealed

the following:

a. The "A" RCP shaft breakage appears to have been caused by a combination

of residual shaft stresses (due to forging during manufacture) and

thermal shocking (due to seal injection flows). Since the replacement

shaft will be composed of the same material as the original shaft (A286

stainless steel), the licensee is considering either reducing the seal

injection flows or eliminating it entirely thus reducing or removing

the thermal stress aspect of the failure.

b. The "B" RCP shaft cracking appears to have been caused by improper

welding of the journal bearing to the pump shaft (which was an

unweldable material). The cracking occurred .in the weld area. The

reason for the improper welding is under investigation by the licensee.

The licensee is planning to replace the "B" shaft with one made of a

new material called "Nitronics 50" which is a type of enhanced A316

stainless steel,

c. The "C" and "D" RCP shaft cracking has not yet been evaluated since

these pumps have not been disassembled. The only crack indication

presently available is that determined by UT. Thtse shafts are also

constructed of A286 stainless steel.

In addition to shafts breaking and cracking, the four cap bolts that are

used to attach the pump impeli9r to the shaft were found to be broken on the

"A" RCP 'and cracked on the "B' RCP. Preliminary investigation indicates

that these failures were due to intergranular stress corrosion cracking

(IGSCC). The IGSCC occurred because the bolts were made of A286, Grade 660

stainless steel which is highly susceptible to this type of failure. The

licensee is replacing all of the cap bolts with bolts composed of an Inconel

material.

The inspector will continue to follow the licensee's activities in this

area.