IR 05000413/1993024

Insp Repts 50-413/93-24 & 50-414/93-24 on 930808-0904.No Violations or Deviations Noted.Major Areas Inspected:Review of Plant Operations,Surveillance Observations,Maint Observations & LERs

ML20058N214
Person / Time
Site:	Catawba
Issue date:	09/24/1993
From:	Freudenberger, Hopkins P, Lesser M, John Zeiler NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML20058N212	List: IR 05000413/1993024 ML20058N208
References
50-413-93-24, 50-414-93-24, NUDOCS 9310080279
Download: ML20058N214 (17)

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UMTED STATES

1. NUCLEAR REGULATORY COMMISSION ye #mnec%[

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)g ?> 101 MARIETTA STREET. N.W., SUITE 2WJO ATLANTA, GEORGIA 30323-0199 g * . ,

s Report Nos.: 50-413/93-24 and 50-414/93-24 Licensee: Duke Power Company 422 South Church Street Charlotte, Docket Nos.: 50-413 and 50-414 License Nos.: NPF-35 and NPF-52 Facility Name: Catawba Nuclear Station Units 1 and 2 Inspection Conducted: August 8, 1993 - September 4, 1993 Inspector: a / o Ig/.J['f3 y R.'J. Freudenberger, Senior Resident Inspector ta'te Signed Inspector:

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P.M . Hopkins, Resident Inspector NMMTEo _/a/.M'f5 Date Signed

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Inspector: u ffh/fri 3 71/. 7 [ 95 J. Neiler, Resident Inspector l ate Signed

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Approved by: CO(

Lesser, Chief f /!f T Date Signed M. S'.

Projects Section 3A Division of Reactor Projects SUMMARY Scope: This resident inspection was conducted in the areas of review of plant operations, surveillance observations, maintenance observations, the licensee's employee concerns program (TI 2500/28), and licensee event report Results: One Unresolved Item was identified involving potential discrepancies in the licensee's low Temperature Over Pressure

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Protection setpoints (paragraph 3.c).

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REPORT DETAILS Persons Contacted .

Licensee Employees S. Bradshaw, Shift Operations Manager

• K. Evans, Senior Mechanical Engineer

• J. Forbes, Engineering Manager ,

R. Futrell, Regulatory Compliance Manager

• L. Jones, Nuclear Generation Engineer

• W. McCollum, Station Manager i W. Miller, Operations Superintendent

• K. Nicholson, Compliance Specialist '

• D. Rehn, Catawba Site Vice-President Other licensee employees contacted included technicians, operators, [

mechanics, security force members, and office personne NRC Resident Inspectors i R. Freudenberger, Senior Resident Inspector

• P. Hopkins, Resident Inspector  ;

J. Zeiler, Resident Inspector

• Attended exit intervie !

Acronyms and abbreviations used throughout this report are listed in the ;

last paragrap ;

. Plant Status and Activities j Unit 1 Status Summary  ;

Unit 1 operated at or near full power for the entire report period f with no major problem Unit 2 Status Summary Unit 2 operated at or near full power for the entire report period with no major problem Inspections and Activities of Interest i During the inspection period one inspection was conducted by specialist inspectors from the NRC Region II Office as follows:

Date Inspectors Functional Area Report 8/23-27 M. Hunt Engineering 93-25 M. Miller )

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3. Plant Operations (71707) General Observations The inspector reviewed plant operations throughout the report period to verify conformance with regulatory requirements, TS and administrative controls. Control Room logs, the Technical Specification Action Item 1.og, and the R&R log were routinely reviewed. Shift turnovers were observed to verify that they were conducted in accordance with approved procedures. The number of licensed personnel on each shift inspected either met or surpassed the requirements of Technical Specifications. Furthermore, daily plant status meetings were routinely attended, and plant tours were performed on a routine basi During the plant tours, the inspector verified by observation and interviews that proper measures were taken, and procedures were '

followed, to ensure that physical protection of the facility met current requirements. Items inspected included the adequacy of the security organization; the establishment and maintenance of gates, doors, and isolation zones in the proper condition; and the use of access control badgin In addition, the areas toured were observed for fire prevention and protection activities and radiological control practices. The inspector also reviewed PIPS to determine if the licensee was appropriately documenting problems and implementing corrective action Offsite Communications Degradation On September 3, at 12:00 p.m., the inspector noticed that the NRC resident office (commercial) telephone lines were not operating properly. Based on concerns over the status of the ENS telephone system, the inspector contacted the control room and requested that the licensee check the operability of this system. The licensee subsequently discovered that the NRC's FTS system and commercial long distance telephone lines were out of service. The licensee determined that the emergency communications systems (Selective Signaling System) used to notify and activate the state / county emergency response network was operating properl Following contact with the commercial phone carrier, the licensee learned that a fiber-optic cable had been cut during road construction work, causing the loss of telephone servic l In accordance with 10 CTA 50.72.b.l.v, the licensee notified the NRC operations center of the loss of FTS and commercial communications via a microwave link through the load dispatche In addition, the licensee established limited backup communication via a cellular phone. All telephone communications were restored i at 3:25 p.m., following repair of the telephone cable. The i licensee's actions to implement compensatory methods of l communication and re-establish normal communication were l appropriat l l

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c. Potential Nonconservatism in Low Temperature Over Pressure Setpoints During this inspection period, the inspectors reviewed the licensee's evaluation of the potential nonconservatism in the LTOP protection setpoint calculation for the Westinghouse-designed cold '

over pressure mitigation system. A Westinghouse letter dated March 15, 1993, alerted licensees of the potential problem and '

described two factors that may not have been properly considered when Westinghouse originally developed the LTOP setpoint curve These factors included: (1) the failure to take into consideration the pressure difference between where the pressure transmitters for LTOP are located (in the hot leg piping) and the actual pressure in the reactor vessel, and (2) the failure to consider the pressure difference in the reactor vessel when the reactor coolant pumps are operating. If these factors were not properly considered, the dynamic pressure in the reactor vessel could be higher than that which is sensed by the LTOP pressure sensor This could result in nonconservative PORV setpoints, which could allow reactor vessel pressures to exceed the design limit for LTOP operatio The licensee reviewed their LTOP setpoint calculations and determined that the above factors were not considered in the setpoint development. The licensee determined that the pressure drop between the reactor vessel downcomer region and the reactor coolant hot leg (where the LTOP pressure sensors are located) was 44.5 psi when all four reactor coolant pumps are operating. Their review of the original LTOP calculation and the reactor vessel pressure / temperature curves from their TS revealed that there is adequate margin in the current PORV setpoint of 400 psig, provided that no more than one reactor coolant pump is in operation and a cooldown rate of 20*F/hr is not exceeded. At the end of the report period, the licensee was developing operating conditions necessary to limit the number of reactor coolant pumps in operation and the allowable cooldown rates for LTOP operatio The operating procedures will be revised to incorporate these limitations. The inspectors will review the adequacy of these limits following their implementatio During review of the licensee's original and revised LTOP PORV setpoint calculations contained in CNC-1223.03-00-0005, the inspectors identified two problems discussed below:

1) The licensee's original PORV setpoint evaluation was performed verifying the acceptability of a 400 psig PORV setpoint. Likewise, during LTOP conditions, the licensee administratively sets the PORVs at 400 psig. However, TS 3.4.9.3 allows the PORVs to be set at a maximum of 450 psi Based on review of the LTOP calculations, it appeared to the inspector that at a PORV lift setpoint of 450 psig, the Reactor Coolant System pressure and temperature limits

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imposed by TS could be exceeded if the maximum pressure overshoot beyond the PORV setpoint were to occur. This '

pressure overshoot could result from time delays in signal processing and valve opening, instrument uncertainties, and single failure. Operation with a PORV setpoint at the maximum pressure allowed by TS should ensure that these limits are not exceeded. The inspector requested that the licensee perform an evaluation to determine if a PORV setpoint of 450 psig could ensure that the pressure and temperature limits allowed by TS would not be exceede Until this evaluation can be completed by the licensee and reviewed by the inspector, this issue will be followed as part 1 of URI 413, 414/93-24-01: Review of Potential Discrepancies in LTOP PORV Setpoint ) Based on review of TS 3.4.9.1 for Reactor Coolant System pressure and temperature limits, the inspector noted that during the maximum cooldown rate of 100*F/Hr (per TS), the reactor coolant pressure limits during LTOP operation potential . ' could be exceeded. As is shown in TS Figure 3.4-3, at this cooldown rate there appears to be no LTOP protection at reactor coolant system temperatures between approximately 95'F and 115'F with a PORV setpoint of 400 psig. This vulnerability is exacerbated by the potential for PORV setpoint overshoot and setpoint nonconservatism discussed earlier in this section. The inspector requested that the licensee evaluate this potential problem. Until this evaluation can be completed by the licensee and reviewed by the inspector, this issue will be followed as part 2 of URI 413, 414/93-24-01: Review of Potential Discrepancies in LTOP PORY Setpoint . Maintenance (62703)

During the reporting period the inspector reviewed maintenance activities to verify compliance with the appropriate procedures and T Methods used in this inspection included direct observation, interviews with personnel, and review of records. The activities associated with the following W0s were considered noteworthy: WO 93060298: Replace Valve INI-482 with New Valve In June, a small orifice and and fine throttling valve (1NI-482)

were installed downstream of vent valve INI-210 in the NI discharge piping to provide a means of bleeding off pressure trapped between the Unit 1 Cold Leg discharge check valves and the NI pump discharge valves. On August 16 a problem developed in this bleed path that indicated an obstruction in the orific The orifice was cut out, cleaned, and welded back in place. When i adequate flow still could not be obtained through the orifice and valve, the licensee replaced both with a new type valve that was more suitable for the fine throttling neede .

WO 93060298 was initiated to implement minor modification CE 4211 for this replacement. On August 18 the licensee implemented this modification. The inspector witnessed portions of this maintenance, noting that the licensee made no attempt to preserve the "as found" condition of the removed pipe section to identify the root cause of the obstruction in the orifice. The inspector discussed this with the NI System Engineering staff; they indicated that they had intended to witness the removal of the orifice but were not contacted by the person who performed the work. The inspector considered the coordination of this aspect of the work to have been wea The inspector reviewed the WO documentation associated with this repair work and concluded that it was properly performed. The modification package was prepared in accordance with station requirements. The 10 CFR 50.59 safety evaluation for the modification was reviewed; it adequately addressed the safety concerns associated with the valve replacemen b. Steam Generator PORY Stroke Time Problems In 1989 Catawba, San Onofre, and Palo Verde each had failures of '

Steam Generator PORVs to open on demand. Control Components Incorporated (CCI), the manufacturer of the valve, issued a vendor information letter to all affected plants to acknowledge the problem and to offer suggestions for solving the failure ,

mechanis The valve is a pilot operated valve. The pilot valve is opened by the actuator in the first 1/2-inch stroke allowing steam that is in the valve bonnet to vent through the valve plug and equalize ,

pressure across the plug. This pressure balance allows the  ;

actuator to lift the plug and thereby complete the valve strok The pressure in the bonnet is required to maintain tight seating of the valve. If the pressure is not vented through the plug by the pilot valve, the bonnet will remain pressurized and the valve will not stroke open. The actuator thrust cannot overcome the force of a pressurized valve plu j CCI determined that the piston ring, located around the valve plug <

to limit leakage of steam into the bonnet during valve actuation, j was allowing excess steam flow into the bonnet. The bonnet remained pressurized because more steam leaked into the bonnet than could be relieved by the pilot valve. CCI redesigned the piston ring to further limit the steam flow in 1989 and increase the flow capacity of the pilot valve. These two modifications were installed throughout the industry to prevent a recurrence of the failure to open scenario described in this section. Catawba Nuclear Station started live-steam testing on a periodic basis to prove that the PORVs would open. This testing has demonstrated that the valves open reliably; however, stroke time problems continu l I

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Because the piston ring seal was tighter after the modification, l greater friction force was exerted on the plug. Valve stroke-time failures occurred more frequently in 1990 and increased dramatically for 3 valves in particular in 199 Subsequent valve and actuator disassembly revealed no evidence of binding in the valve or actuator. In most cases no root cause was identifie Part of the problem is that Catawba specified a 20-second closure time when the valves were procured in the 1970's, whereas the TS !

requires a five-second closure time. CCI has been involved with the problem and maintains that the valve is not designed to stroke ;

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in five seconds and that the force output of the actuator is marginal for a five-second closure. The licensee technical basis for a five-second closing time, presently required by TS 4.6.3.2, Table 3.6-20, was not clear. Consequently, Catawba is developing the technical basis to change the TS to lengthen the stroke time requiremen In December 1992, diagnostic testing of the valves was performed by the licensee. Results indicated that the piston ring was the >

cause of the increased friction. The ring is a two-piece wedge-style design that is activated by pressure and flow. The rush of steam through the piston ring area during valve actuation creates a differential pressure across the two rings, allowing them to '

wedge together and seal tighter. Initial conclusions from the testing revealed that additional actuator thrust would evercome the piston ring drag. The actuator thrust was doubled in early 1993 by the addition of a third spring in the actuator. CCI '

concurred with and designed this modificatio The inspectors have reviewed documentation and observed extensive work being performed on the SG PORVs. Three valves, ISVI, 2SV1 and 2SV13, recently had increased stroke times. Their stroke times have ranged from 3 to 12 seconds. Fisher Services was

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contracted to perform diagnostics on 2SV13 in August. The testing verified that the piston ring was the cause of the friction in the '

valve. Valve 2SV13 is disassembled at this time and the licensee is taking measurements of the valve internals. The intent is to determine why ISVI, 2SV1 and P6V13 do not stroke consistentl The hypothesis is that there is a dimensional problem or dimensional stack-up that differs from the other six PORV The following is a chronology of problems that have occurred on steam generator PORVs for both unit Type of Valve: Double acting piston actuator 6x10 inch pilot operated Pressurized seat 1970's: Initial PORY valves were procured with operating stroke times specified as 20 second . - .

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'E Late 1980's: Valve operator EQ upgraded and Nitrogen I control added by licensee commitment to !

have valve assurance to operate during ;

Steam Generator Tube Ruptur ;

Prior to 1989: Valve IWV stroke times normal for 5 ~ 'I seconds. Multiple seat leak problems were ,

identified and solve !

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Early 1989: Failure of PORVs at Palo Verde drew more ;

attention to PORVs. Consequently, CCI' ;

issued a Vendor Information Lette Catawba and San Onofre also experience

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PORV valve problem ;

Mid 1989-Mid 1990: Catawba initiated a 2-year rebuild j modification program of the PORY valves t

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under CCI recommendations' that included (1) pilot holes enlargement; .(2) piston ,

ring design change, and (3) other

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maintenance enhancements and seat angle changes that were expected to prevent. seat leakage and decrease stroking time Seat leakage under control. Random. stroke

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Post 1989/90:

failures with root cause evaluations were .

varying. Results of internal inspections !

appeared to be normal. The license ;

started rebuilding actuators at this tim Mid/ Late 1992: Because of repetitive stroke time problems-for ISV1, 2SVI, and 2SV13, the licensee applied in-house diagnostics. TSMs were i installed with the CCI field engineer j presen :

Unit 2 1993: A third spring was installed to double the actuator thrust. Valves ISV1, 2SVI, and j 2SV13 were inspected, and measurements were sent to CCI for further evaluatio i The valve piston rings have been sent to !

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CCI for measurement and/or honing as necessar The licensee's action plan for short-term operability of the SG ]

PORVs includes (1) disassemble and inspect the valve with the -

longest stroke time (2SV13); (2) compare 2SV13 to a good valve -

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and (3) replace 2SV13 and 2SV1 internals if no problem (other than dimension stack-up) can be foun .

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The licensee's long-term course of action includes: (1) requesting a TS change to increase stroke times, and (2) considering a design change of the valves to eliminate piston ring bindin The licensee is still in the process of evaluating the PORV stroking problems in conjunction with CCI's technical engineer The inspector reviewed historical data, PORY modifications, post-modification testing, and maintenance. In addition, inspectors conducted an in-depth review of engineering input by the licensee, '

the valve vendor and interface utilities, and observed testing, inspection and maintenance. The inspector concluded that the licensee has applied sufficient resources to the SG PORV stroke time proble .

5. Review of Licensee's Employee Concerns Program (TI 2500/28)

The objective of this inspection was to obtain information on the characteristics of the licensee's employee concerns program used to provide employees a means to express any safety concerns without fear of retributio The licensee's process for handling formal and informal employee concerns is described in DPC Management Procedure entitled " Employee Recourse." The recourse process can be used by employees to deal with personal , business , and safety-related concerns. In this procedure the licensee encourages the re:olution of employee concerns via informal discussions with their management or with personnel in the Human Resources department who are specially trained to handle such issue For employees electing to use formal recourse, a two-step process is involved whereby the concern initially is communicated in writing to any level of management and/or the Human Resource Department personne Human Resource personnel then conduct an investigation of the employee's concern. After Human Resources completes the investigation the results are reviewed by management, and a written response is provided to the ,

employe If the employee is dissatisfied with the response, the next '

step in the process involves further review of the concern by the Executive Vice President of Power Generation and the Human Resource staf The inspectors determined from their review that the licensee has implemented a program whereby employees can informally and formally discuss concerns with personnel other than line management. The enclosed attachment to this report provides characteristics of the licensee's progra . Licensee Event Reports (92700)

l The LERs listed below were reviewed to determine if the information I provided met NRC requirements. The inspector considered the adequacy of I description, verification of compliance with Technical Specifications and regulatory requirements, corrective action taken, existence of

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potential generic problems, fulfillment of reporting requirements, and the relative safety-significance of each even (Closed) LER 413/91-19: Reactor Trip on Turbine Trip Due to Loss of Both Main Feedwater Pumps on High discharge Pressur On September 11, 1991, a Unit I reactor trip occurred as a result of a turbine trip. The turbine tripped because both main feedwater pumps tripped on high discharge pressure. The high discharge pressure occurred when the feedwater Control System sensed that there was a loss of two out of three of the Nuclear Power Channel inputs and called for the rapid closure of the main feedwater control valves. The cause of the loss of Nuclear Power Channel input was attributed to IAE personnel failure to follow procedures for calibrating the Nuclear Power Channel N-41, which was in progress at the time of the trip. The IAE technicians manipulated the wrong switch during this testing and possibly caused the loss of Nuclear Power Channel input to the feedwater control system. NRC violation 413/91-21-02 was issued for this event for personnel failure to follow proper independent verification techniques. The inspectors previously reviewed the licensee's corrective action for this aspect of the event in Inspection Report No. 413,414/93-03. The licensee also enhanced the Nuclear Power Channel testing procedures to allow testing of the system without degrading the median selected input to the Feedwater Control System. The inspector verified that the revisions to these procedures were complete (Closed) LER 413/91-23: Technical Specification 3.0.3 Entered as a Result of Both Trains of the Control Room Ventilation System Being Inoperable Due to Equipment Failur On November 17, 1991, at 10:40 a.m., with Unit 1 in Mode 1 at 100%

power, and Unit 2 in No Mode (Defueled), Operations personnel ,

discovered that the VC damper 1-CR-D-10, suction isolation damper for A Train Control Room Air Handling Unit, in the Control Room Ventilation and Chilled Water System had failed closed. With VC Damper 1-CR-D-10 closed, A Train VC/YC was inoperable. Train B of the Control Room Ventilation and Chilled Water system had been declared inoperable on November 11, 1991, at 1:45 a.m., because of maintenance on B Train of the Nuclear Service Water syste Consequently, with both trains of VC/YC inoperable, Unit I entered TS 3.0.3, at 10:40 VC Damper 1-CR-D-10 was manually positioned and blocked cpen to restore operability of Train A VC/YC at 10:45 a.m., and TS 3. Unit I was exite The Control Room Area Ventilation and Chilled Water Systems combine to form one system that is designed to maintain a suitable environment in the Control Room, Cable Room, Battery Rooms, Switchgear Rooms, Motor Control Center Rooms, and the Electrical

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Penetration Rooms at elevation 594. The Control Room Area Ventilation and Chilled Water System is shared between both Unit There are two 100% redundant trains of VC/YC equipment. Each is capable of being powered by Unit 1 or Unit 2 Essential Auxiliary l Power, but under normal conditions both trains are aligned to j Unit 1. Two Diesel Gi.nerators are provided per Unit to energize :

the Essential Auxiliary Power buses during emergency condition The inspector examined the records and found that a review of the Operating Experience Program data base by the licensee for the previous 24 months prior to this event revealed that VC/YC system inoperability as a result of equipment failure (including damper hydromotor failure) is a recurring problem. Hydromotor failure in the past also caused Unit I and Unit 2 to enter TS 3.0.3 due to both trains of VC/YC system being inoperabl In response to the November 17, 1991, event, the licensee's imediate actions were appropriate and resulted in the Unit exiting T.S. 3. Subsequently, the licensee modified the VC/YC system controls to improve the reliability of the system and, subsequently, simplify maintenance. The inspectors have reviewed the licensee's responses and NSM's, and observed the ongoing modification work and the post-modification, maintenance and operability testin The inspector concluded that the short-term and long-term responses appropriately addressed the issues involve ,

7. Exit Interview The inspection scope and findings were summarized on September 7, 1993, with those persons indicated in paragraph 1. The inspector described the areas inspected and discussed in detail the inspection findings listed below. No dissenting comments were received from the license The licensee did not identify as proprietary any of the materials provided to or reviewed by the inspector during this inspectio Item Number Description and Reference  !

URI 413, 414/93-24-01 Review of Potential Discrepancies in LTOP PORV Setpoints (paragraph 3.c).

8. Acronyms and Abbreviations CE -

Catawba Exempt (modification)

CCI -

Control Component Incorporated t

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DPC -

Duke Power Company ENS -

Emergency Notification System EQ -

Environmental Qualification FTS -

Federal Telecommunications System IAE -

Instrument and Electrical IWV -

Inservice Valve Test

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LER -

Licensee Event Report LTOP - Low Temperature Overpressure Protection NI -

Safety Injection NRC -

Nuclear Regulatory Commission  ;

NSM -

Nuclear Station Modification PIP -

Problem Investigation Process (report)

PORY - Power Operated Relief Valve ,

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R&R -

Removal and Restoration SG -

Steam Generator TI -

Temporary Instruction TS - Technical Specifications TSM -

Temporary Station Modification URI -

Unresolved Item VC -

Control Room Ventilation ^

VC/YC - Control Room Ventilation and Chill Water System WO -

Work Request j

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Attachment EMPLOYEE CONCERNS PROGRAM PLANT NAME: CATAWBA LICENSEE: Duke Power Company DOCKET #: 50-413. 414 PROGRAM: Does the licensee have an employee concerns program?

Yes

, Has NRC inspected the program?

No SCOPE: (Circle all that apply) Is it for: Technical?

Yes Administrative?

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Yes Personnel Issues? g Yes Does it cover safety as well as non-safety issues?

Yes Is it designed for: Nuclear safety?

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Yes, but more closely designed for handling personnel issues Personnel safety? ,

f Yes, but more closely designed for handling personnel issues Personnel issues - including union grievances?

Yes i

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2 , Does the program apply to all licensee employees?

Yes Contractors? i No, however the licensee indicated that contractor employees may raise safety concerns to the station's Safety Assurance Manage . Does the licensee require its contractors and their subs to have a similar program?

No Does the licensee conduct an exit interview upon terminating !

employees asking if they have any safety concerns?

Yes, for Duke Power employees C. INDEPENDENCE: What is the title of the person in charge?

l An employee who chooses to use the " formal Recourse Process" must i outline the concern in writing to his/her second level supervisor. '

This supervisor provides a written response to the employee. If ;

the employee is dissatisfied with this response, he/she may l request a further review of the concern by writing the Executive j Vice President over the employee's department. This Executive Vice President prepares a written response to the employe An employee may also forward written concerns to the Human Resources personnel (Human Resources Consulting) or discuss the options in handling any concerns with this grou . Who do they report to?

Ultimately, the Human Resource Consulting group reports to the Catawba Site Vice Presiden . Are they independent of line management?

No, the Human Resources Consulting Manager reports to the Human Resources Manager who in turn reports to the Catawba Site Vice Presiden . Does the ECP use third party consultants? .

No I

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3 How is a concern about a manager or vice president followed up?

Based on discussions with the Catawba Human Resource Consulting department, if a concern about a manager is brought to their attention they would investigate and followup with higher management personnel as needed

, RESOURCES: What is the size of staff devoted to this program?

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The Catawba Human Resource Consulting department has 10 employees, however, they are not devoted solely to the recourse function What are ECP staff qualifications (technical training, 3 interviewing training, investigator training, other)?

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The Catawba Human Resource Consulting job qualifications include 1) typically 5 - 8 years experience in human resource expertise, 2) demonstrated project management skills, presentation skills, analytical and problem-solving skills, interpersonal human relations verbal / written communication skills, and organization and time management skills. Duke Power Company offers human resource development training such as 1) Interviewing Skills, 2)

Exit Interviewing, 3) How to Conduct an Investigation, 4) Problem Solving and Decision Making, and 5) Effective Documentation and Writing Skill REFERRALS: Who has followup on concerns (ECP staff, line management, other)?

Both line management and the Human Resource Consulting group ,

depending on the circumstances involved in the concern

, CONFIDENTIALITY: Are the reports confidential?

Yes

, Is the identity of the alleger made known to (senior management, ECP staff, line management, other)?

Designated Human Resource personnel only depending on the nature of the concern Can employees be: l

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4 Anonymous?

Yes Report by phone?

Yes

. FEEDBACK: Is feedback given to the alleger upon completion of the followup?

Yes, if the allegation is identified in writing Does program reward good ideas? ,

No Who, or at what level, makes the final decision of resolution?

This is determined by the level of recourse initiated, ultimately, the Executive Vice President of Power Generation could get involved and make the final resolution decisio l Are the resolutions of anonymous concerns disseminated?

No Are resolutions of valid concerns publicized (newsletter, bulletin board, all hands meeting, other)?

Maybe, depending upon the nature of the concer EFFECTIVENESS: How does the licentee measure the effectiveness of the program?

i There does not appear to be a good mechanism for measuring the I effectiveness of the program, however, the licensee conducts employee feedback and employee opinion surveys that might offer an employee an opportunity to voice dissenting comments i Are concerns: Trended?

Yes l

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5 Used?

Yes In the last three years how many concerns were raised?

Closed? What percentage were substantiated? Catawba Personnel Concerns: 14, all were closed and all were substantiate Catawba Technical Concerns: 0, however, there was one employee exit interview in 1991 that involved a concern of a technical nature and was appropriately resolved with the employee How are followup techniques used to measure effectiveness ,

(random survey, interviews, other)?

None How frequently are internal audits of the ECP conducted and by '

whom?

None have been conducted ADMINISTRATION / TRAINING: Is ECP prescribed by a procedure?

Yes ,

f How are employees, as well as contractors, made aware of this program (training, newsletter, bulletin board, other)?

The licensee reported that employees are made aware of the employee recourse process through 1) orientation for new employees, 2) general employee training, 3) employee benefits material, 4) company procedure manual, 5) outage handbook for employees and vendors, and 6) through periodic postings on the bulletin boards and site team notes.