ML20058P158
| ML20058P158 | |
| Person / Time | |
|---|---|
| Site: | Maine Yankee |
| Issue date: | 07/29/1993 |
| From: | NRC |
| To: | |
| References | |
| NUDOCS 9312230104 | |
| Download: ML20058P158 (4) | |
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ATTACIIMENT A x
EMPLOYEE CONCERNS'PRObRAMS#
PLANT NAME: Vermont Yankee LICENSEE: Vermont Yankee Nuclear Power Corporation DOCKET #: 50-309 NOTE:
Please circle yes or no if applicable and add comments in the space provided.
A.
PROGRAM:
1.
Does the licensee have an employee concems program? (Yes er No/ Comments)
Yes 2.
Has NRC inspected the program? Yes, Report No. 92-12 (Section 8.5). SALP 91-99 (III.G.1), and Report No. 93-19 B.
SCOPE: (Circle all that apply) 1.
Is it for:
a.
Technical? (Yes, No/ Comments) Yes b.
Administrative? (Yes, No/ Comments) Yes c.
Personnel issues? (Yes, No/ Comments) Yes 2.
Does it cover safety as well as non-safety issues? (Yes g No/ Comments) Yes 3.
Is it designed for:
a.
Nuclear safety? (Yes, No/ Comments) Yes b.
Personal safety? (Yes, No/ Comments) Yes c.
Personnel issues - including union grievances? (Yes er No/ Comments)
Yes 4
Does the program apply to all licensee employees? (Yes Dr No/ Comments) Yes 5.
Contractors? (Yes Dr No/ Comments) Yes
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- T1ssue-Date-07/29/9A A-1 2500/028 Attachmenf x
2000$6 9312230104 930729 PDR ADOCK 05000309:
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Does the licensee require its contractors and their subs to have a similar program? -
(Yes a No/ Comments) No 7.
Does the licensee conduct an exit interview upon terminating employees asking if they have any safety concerns? (Yes a No/ Comments) No' i
C.
INDEPENDENCE:
1.
What is the title of the person in charge?. Director of Human Resources for-administration, the Internal Auditor for confidential issues..
2.
Who do they report to? President and Chief Executive Officer 3.
Are they independent of line management? Yes 4.
Does the ECP use third party consultants? No 4
5.
How is a concern about a manager or vice president followed up? By referral to the Internal Auditor.
D.
RESOURCES:
1.
What is the size of the staff devoted to this program? No one is devoted to the -
program. It is a collateral duty for all involved.
j 2.
What 'are ECP staff qualifications (technical training, interviewing. training.
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investigator training, other)? The Internal Auditor has 13 years of experience.
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E.
REFERRALS:
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1.
Who has followup on concerns (ECP staff, line management, other)? Plant.
Manager is usually notified of concerns. Ile is then responsible to coordinate resolution.
F.
CONFIDENTIALITY:
1.
Are the reports confidential? (Yes E No/ Comments) Yes, confidential concerns -
are generally maintained confidential.. All of her improvement suggestions are.
publicized.
2.
Who is the identity of the alleger made known to (senior management, ECP staff, i
line management, other)? (Circle, if other explain) For confidential concerns,'.
the Internal Auditor and the President and CEO. For other suggestions and -_
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. concerns, the person may be identified.'
Issue Date: 07/29/93 A-2 '
2500/028 ' Attachment t
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3.
Can employees be:
a.
Anonymous? (Yes, No/ Comments) Yes b.
Report by phone? (Yes, No/ Comments) Yes. The Internal Auditor takes phone calls.
The Human Resources Department receives suggestions and concerns via their VAX system.
G.
FEEDIIACK:
1.
Is feedback given to the alleger upon completion of the followup? (Yes a No -
If so, how?) Yes - if the person is identified. In one case, an anonymous concern results were publicized due to the nature of the issue.
2.
Does program reward good ideas? Yes, in that succestions that provide improvement or savings may result 'n a cash reward of from $HK).00 to
$2,000.00.
3.
Who, or at what level, makes the fmal decision of resolution? The Evaluation Committee for Suggestions and Concerns.
The Internal Auditor and President for Confidential Concerns.
4.
Are the resolutions of anonymous concerns disseminated?
Generally not; however, in one case the results of an anonymous allegation were included in department training due to the nature of the issue.
5.
Are resolutions of valid concems publicized (newsletter, bulletin board, all hands meeting, other)? The status of all non-anonymous, no confidential issues are posted on company bulletin boards, 11.
EFFECTIVENESS:
1.
How does the licensee measure the effectiveness of the program? Although there is no objective measure of program effectiveness, corporate officers and senior managers are involved with the resolution of issues.
2.
Are concerns:
a.
Trended? (Yes n No/ Comments) No, suggestions are listed as part of the suggestion program, b.
Used? (Yes or No/ Comments) Yes issue Date: 07/29/93 A-3 2500/028 Attachment
3.
In the last three years how many concerns were raised? =6 Of the concerns raised, how many were closed? All. What percentage were substantiated? =75%
Approximately eleven confidential concerns were received. All are closed and fifty percent substantiated.
4 How are followup techniques used to measure effectiveness (random survey, interviews, other)? There are no effectiveness measurements.
5.
How frequently are internal audits of the ECP conducted and by whom? There are no internal audits of the programs.
I.
ADhflNISTRATION/ TRAINING:
1.
Is ECP prescribed by a procedure? (Yes nr No/ Comments) Yes. Vermont Yankee Improvement Suggestion and Safety Program, VYP:225.
2.
How are employees, as well as contractors, made aware of this program (training, newsletter, bulletin board, other)? The program is addressed in the initial and annual generic employee training for employees and contractors. It is also publicized in bulletin board notices and within the company newsletter.
ADDITIONAL CONINfENTS:
(Including characteridics which make the program especially effective, if any.)
The Vermont Yankee Employee Concerns Program is unique in that it has been combined with an improvement suggestion program. Cash awards may be granted for suggestions that provide improvements or monetary savings. The program accepts confidential or anonymous concerns from employees or contractors. Confidential concerns are handled differently from others in that they are forwarded directly to the Corporate Internal Auditor or Company President, NAAfE:
TITLE:
PHONE #:
Issue Date: 07/29/93 A-4 2500/028 Attachment
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KING OF PRUSSIA. PENNSYLVANIA 1 6 1415 Docket No. 50-271 SEP 2 9 IN Mr. Donald Reid Vice President, Operations Vermont Yankee Nuclear Power Corporation RD 5, Box 169 Ferry Road Brattleboro, Vermont 05301
Dear Mr. Reid:
SUBJECT:
VERMONT YANKEE INSPECTION 93-19 This refers to the safety inspection conducted by Messrs. H. Eichenholz and P. Harris on August 8 - September 11, 1993, at the Vermont Yankee Nuclear Power Station, Vernon.
Vermont. The preliminary results were discussed with Mr. R. Wanczyk at the conclusion of the inspection. The inspection focused on issues important to public health and safety, and consisted of performance observations of the conduct of operations and maintenance, and independent evaluation of safety system status and quality records.
Our observations this period were that the plant was operated safely, and pre-staging of equipment for Refueling Outage XVII was adequately controlled. We found that your staff reviewed and scheduled the outage in a manner consistent with industry guidance and the implementation of defense-in-depth philosophies.
No reply to this report is necessary and your cooperation with r is appreciated.
Sincerely, l.
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(' Reactor Projects ' Branch 3 mes C. Linville, Chief.
Division of Reactor Projects
Enclosure:
NRC Inspection Repon 50-271/93-19
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SEP 2 9 Sg3 Mr. Donald Reid 2
cc w/ enc 1:
R. Wanczyk, Plant Manager J. Thayer, Vice President, Yankee Atomic Electric Company L. Tremblay, Senior Licensing Engineer, Yankee Atomic Electric Company J. Gilroy, Director, Vermont Public Interest Research Group Inc.
D. Tefft, Administrator, Bureau of Radiological Health, State of New Hampshire Chief Safety Unit, Office of the Attorney General, Commonwealth of Massachusetts R. Gad, Esquire G. Bisbee, Esquire R. Sedano, Vermont Department of Public Service T. Rapone, Massachusetts Executive Office of Public Safety Public Document Room (PDR) local Public Document Room (LPDR)
Nuclear Safety information Center (NSIC)
K. Abraham, PAO (2 copies)
NRC Resident Inspector State of New Hampshire, SLO Designee State of Vermont, SLO Designee Commonwealth of Massachusetts, SLO Designee t
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o Mr. Donald Reid 3
SEP 2 91993 bec w/ encl:
Region I Docket Room (with concurrences)
E. Kelly, DRP i
J. Shediosky, DRP.
J.12nning, TSS, DRP (Section 5.8, Attachment 1).
R. Rosano, OWFN 7-H-5 (Section 5.8, Attachment 1)
Allegation Coordinator (Section 5.8, Attachment 1) bec w/enci (VIA E-MAIL):
V. McCree, OEDO D. Dorman, NRR W. Butler, NRR G
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U.S. NUCLEAR REGULATORY COMMISSION REGION 1 Report No.
93-19 Docket No.
50-271 Licensee No.
DPR-28 Licensee:
Vermont Yankee Nuclear Power Corporation RD 5, Box 169 Ferry Road Brattleboro, VT 05301 Facility:
Vermont Yankee Nuclear Power Station Vernon, Vermont Inspection Period:
August 8 - September 11,1993 Inspectors:
Harold Eichenholz, Senior Resident Inspector Paul W. Harris, Resident inspector John T. Shediosky, Pro' Engineer Approved by:
M Eugene M. Kelly, Chief /
- Date Reactor Projects Section 3A Scope:
Station activities inspected by the resident staff this period included Operations, Maintenance, Engineering and Plant Support. Initiatives selected for inspection included:
installation of the Vernon Tie station' blackout modification, establishment of formalized Outage Guidelines, and the evaluation of industry experience regarding quick start capability and rupture disc design for the high pressure coolant injection system. Backshift and ' deep" backshift including weekend activities amounting to 53 hours6.134259e-4 days <br />0.0147 hours <br />8.763227e-5 weeks <br />2.01665e-5 months <br /> were performed on August 9-11,13.
16, 24-26, 31 and September 2-3, 6, and 9,1993. Interviews and discussions were conducted with members of Vermont Yankee management and staff as necessary to support this inspection.
Findings:
An overall assessment of performance during this period is summarized in the-Executive Summary. Unresolved items were initiated regarding configuration control for the high pressure coolant injection system exhaust piping debris screens (Section 4.3), and the screening process used for written ' safety evaluations of temporary modifications (Section 4.4).
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o EXECUTIVE
SUMMARY
Vermont Yankee Inspection Report 93-19 Operations Operating personnel responded well to an equipment failure that affected torus cooling operations. Evaluation of corrective actions to preclude the recurrence ofinadvertent Group IV isolations of the residual heat removal system continues. The implementation of daily Plant Manager meetings improved the licensee's assessment of emergent issues. Refueling Outage XVII was planned with defense-in-depth philosophies.
Maintenance and Testing Accurate and prompt recording of as-found conditions associated with the repeat failure of a service water valve enabled better root cause determination. Good coordination between technicians and vendor representatives was observed during the rebuilding of main steam isolation valve actuators; however, incorrect torque values were initially applied during assembly of the valve manifold, but later corrected prior to installation. Vermont Yankee appropriately implemented industry information regarding time-to-full flow testing for high pressure safety injection systems.
Engineering The installation and initial testing of the new Vernon tie line were completed satisfactorily this period. The design configuration of high pressure coolant injection system turbine casing vent lines was questioned, and is being evaluatni. Temporary modification screening criteria used to determine the need for a written safety evaluation were inappropriately applied to the jumpering of a refueling interlock. Fire barrier discrepancies were promptly evaluatal and corrected.
Plant Support 1
A downward trend was noted in the number of personnel radiological contamination events.
NRC Temporary Instruction 2500/028, " Employee Concerns Program" was conducted.
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TABLE OF CONTENTS EXECUTIVE
SUMMARY
ii TABLE OF CONTENTS.......................................iii 1.0
SUMMARY
OF FACILITY ACTIVITIES I
2.0 OPER ATIONS (71707).................................... 1 2.1 Operational Safety Verification............................. I 2.2 Operator Response to an Equipment Failure................... 2 2.3 Group IV Primary Containment Isolation..................... 3 2.4 Daily Plant Manager Meetings 3
2.5 Outage Safety Review 4
3.0 MAINTENANCE AND TESTING (62703, 61726)................... 5 3.1 Maintenance
...................................... 5 3.2 Failure of Service Water Valve Anti-Rotation Key............... 6 3.3 Main Steam Isolation Air Actuator Rebuild.................... 6 3.4 Thermal Performance Monitoring 7
3.5 Surveillance (61726) 7 3.6 Operational Assessment Feedback - HPCI Quick Starts............ 8 4.0 ENGINEERING (71707, 37828) 8 4.1 Review of Written Reports.
8 4.2 Installation of the Vernon Tie Line Modification 9
4.3 High Pressure Coolant Injection System Debris Screens...........
10 -
4.4 Refueling Interlock Temporary Modification..................
I1 4.5 Fire Barrier Grouting Criteria...........................
12 5.0 PLANT SUPPORT (71707, 61726, TI 2500/028)..
13 5.1 Radiological Controls................................
13 5.2 S ecu ri ty.......................................
14 5.3 Employee Concerns Program (TI 2500/028)..................
14 6.0 EXIT M EETIN G.......................................
14
Attachment:
Employee Concerns Program Sun ey Note: Procedures from NRC Inspection Manual Chapter 2515. " Operating Reactor Inspection Program" which were used as inspection guidance are parenthetically listed for each applicable report section.
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DETAILS 1.0
SUMMARY
OF FACILITY ACTIVITIES Vermont Yankee Nuclear Power Station was operated at power in a safe manner during this inspection period. Outage preparation continued and on August 27, VY commenced reducing power to enter maintenance / refueling outage (RFO) XVII completing a 16.5 month operating period. Offgas activity just prior to shutdown was approximately 19,000 uCi/sec indicating a minor fuel element defect. This value was significantly less than the 56,800 uCi/sec value observed at the end of the last operating cycle, and indicates the decay of transuranic isotopes plated on reactor pressure vessel internals. Vermont Yankee plans to sip all reload fuel and all discharge assemblies, except peripheral bundles, in an attempt to locate and characterize the leaking fuel pin.
During the first phase of refuel operations, two separate refuel incidents occurred. The first event occurred on September 3 at 12:23 p.m. when a third cycle spent fuel assembly (LYN-831), in transportation from the core to the spent fuel pool, inadvertently uncoupled from the fuel grapple. The assembly fell approximately 10 feet back into its original core position. The second event on September 9 at 4:10 a.m. occurred when the refuel bridge operator inadvertently lowered fuel assembly LYV-667 onto an internal component of the reactor vessel.
During both events, no measurable increases in radiation levels on the refuel floor of the reactor building were identified.
An NRC Augmented Inspection Team was sent to the site to independently assess both events, and findings will be documented in NRC Inspection Report 93-81.
The outage was originally scheduled for 37 days. Major planned activities include refueling of one-third of the core, replacement of feedwater heaters 4A/4B and cross-around piping, and inspection of reactor vessel head cladding, low pressure turbine, and drywell and suppression chamber paint. Corrective and preventive maintenance will be conducted on emergency core cooling systems (ECCS) and the emergency diesel generators. Modifications to reactor vessel level instrumentation to improve level indication reliability, and to the residual heat removal service water (RHRSW) system to improve flow and control characteristics will be performed.
A load test of the new seismically qualified Vernon tie line (Section 4.2) and an integrated ECCS test are also scheduled.
2.0 OPERATIONS (71707) 2.1 Operational Safety Verification Daily, the inspectors verified adequate staffing, adherence to procedures and Technical Specification (TS) limiting conditions for operation (LCO), operability of protective systems, status of control room annunciators, and availability of emergency core cooling systems. Plant tours confirmed that control panel indications accurately represented safety system line-ups.
Eafety tagouts properly isolated equipment fer maintenance.
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The inspector toured the reactor building to observe the status of outage preparations and its i
affect on the operability of the ECCS.
Field observations confirmed that installation of
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scaffolding and temporary equipment did not render the systems inoperable and did not impede operator access to equipment. Emergency lighting was verified to be properly positioned and not obstructed. Housekeeping near motor control centers feeding the sub-components of the ECCS systems was commensurate with the work in progress; no fire or safety concerns were identified.
On August 24, during torus cooling, the inspector verified that corner room temperatures were not affected by the temporary work platforms installed in the vicinity of the room coolers. The platforms were installed approximately six feet off the floor and covered approximately 20 percent of the floor area. During residual heat removal (RHR) and service water (SW) system operation, differential pressures, cooling water flows, and seal pressures were verified to be within acceptable operating bands. In addition, the Shift Engineer (SE) was cognizant of the status of the fire detection system in these areas. One deficiency was noted, in that, the operating chain for a 1 1/2 ton chain fall was draped over the motor operator for RHRSW-89A.
The chain fall was pre-staged for the modification of the RHRSW system (Section 3.2 and NRC '
Inspection Report 93-13) and did not affect system operability. An Auxiliary Operator (AO) repositioned the chain fall and inspected the other sub-system to assure that a similar situation did not exist. Similar occurrences ofimproperly positioned chain falls have not been observed.
2.2 Operator Response to an Equipment Failure On August 24, the inspector observed control room operator (CRO) response to a stuck open RHRSW valve being used during torus cooling.
The valve (RHRSW-89A) failed at -
approximately 30 percent open and did not respond to operator control. The valve is used to.
throttle service water flow from the RHR heat exchanger. The operators promptly declared the system inoperable, entered the applicable TS action statements, and notified the Maintenance Department. This failure was identical to a previous event involving the same valve (NRC Inspection Report 93-13). The maintenance performed to restore the valve to operation is described in Section 3.2.
The inspector concluded that the CRO's actions were appropriate. Accurate communications between the CROs and the AO regarding field conditions such as valve position, motor temperature, circuit breaker position helped in making the initial failure assessment. The supervisory CRO independently verified valve position and motor operator position modulator' to gain additional information. The Shift Supervisor (SS) directed the SE to initiate a Potential Reportable Occurrence report and to independently review TS requirements. In addition, the SS assured that the failure conditions in the field were not altered prior to inspection by maintenance personnel. This demonstrated appropriate regard for the preservation of as-found conditions and the importance of such conditions on root cause determinations. The control-r room staff reviewed and followed plant procedure OP 2120, Rev. 30, " Residual Heat Removal System" during the restoration from torus cooling.
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2.3 Group IV Primary Containment Isolation A primary containment Group IV isolation occurred on August 28 while operators were initiating _
the shutdown cooling' mode of the "B" residual heat removal (RHR) system. A pressure transient that resulted from starting the "B" RHR pump caused the Group IV isolation of the shutdown cooling suction isolation valves (RHR 17 and _18). No damage to the equipment -
occurred. This event was similar to four other occurrences in 1991 (March 14 April 23 June 15, and September 8). The occurrences were reported in LER 91-006, wherein the licensee stated that the pressure transients were caused by the depressurization of a section of the piping between the inboard low pressure coolant injection (LPCI) check valve (RHR-46B) and the outboard injection valve (RHR-27B) after the injection loop was flushed. These earlier events were inspected and results documented in NRC Inspection Reports 91-07,91-19, and 92-01.
Vermont Yankee revised operating procedures to pressurize the LPCI injection piping before initiating shutdown cooling flow. Instrumentation was installed to record the dynamic pressure of different parts of the system. These points included the shutdown cooling suction line pressure outside the containment isolation valves, the "B" RHR pump suction pressure, the "B" RHR sub-system pressure down stream of the RHR heat exchanger, the LPCI injection line pressure between valves RHR-25B and RHR-46B, and the "B" recirculation loop pressure downstream of the "B" recirculation pump discharge valve.
The inspector reviewed the procedures for operation and surveillance of the RHR system in its various modes. Technical Specification 4.5.I requirements for maintenance of filled LPCI sub-system discharge piping are incorporated in the normal operating and surveillance procedures.
At the conclusion of this inspection period, VY was analyzing the pressure data and had not reached a definite solution to this problem. Vermont Yankee intends to continue evaluating the dynamic pressure characteristics of the RHR system.
2.4 Daily Plant Manager Meetings In an effort to facilitate effective inter-departmental communications and to allow plant management better direction regarding emergent issues, the Plant Manager implemented the Daily Plant _ Manager Meeting. Attended by all department heads and the Quality Support Group, the daily meetings focus on plant status, priority maintenance items, engineering design modifications, and Potential Reportable Occurrence reports. Discussions also include regulatory information and industry issues.
The inspectors attended several meetings, reviewed the meeting minutes, and concluded that this initiative represents an opportunity to review plant and industry issues. Effective instances included modification of the Vernon tie line, corrective maintenance for an emergency diesel generator fuel oil leak, and leakage from the reac'or building closed cooling water system.
Department managers accepted responsibility, committed to corrective. actions, and provided diverse perspectives for issues outside their expertise.
4 2.5 Outage Safety Review RFO XVII marks the first use of the VY Outage Guideline as the primary document governing 3
the safe conduct of refueling by establishing formal processes to plan, review, and implement an outage schedule. The guideline was approved by the Plant Manager, reviewed by the PORC, and will be assessed prior to future refueling outages incorporating lessons learned, and industry experience. He Quality Support Group plans to independently verify that the guideline conforms to NUMARC and industry guidance regarding the conduct of outages.
Backcround The primary goal of the Outap Guideline (OG) is to plan and conduct a refueling outage in the safest plant configuration p>ssible, consistent with the work in progress.
Duties and responsibilities have been assigned for significant outage tasks, such as contractor processing, surveillances, fire protection and housekeeping, leak rate testing, and water inventory.
The OG defines the " preferred," " normal," and " minimum" plant configurations acceptable during a refueling outage. For the purposes of planning, " minimum" describes the systems necessary to meet TS requirements. For the cold shutdown condition, this means that no ECCS systems are required (by TS) to be operable. In this case, VY applies defense-in-depth philosophies to assure that compensatory actions enhance this operating condition. This may include enhanced operator surveillance of equipment and systems required for cavity cooling and makeup, assignment of additional rc;ponsibilities, and increased process control. When diverse.
alternate, or redundant systems are available above the minimum requirements. VY enters the
" normal" plant configuration. The third and final plant configuration reflects VY policy to plan outages to a " preferred" system configuration for key safety functions (defined below) whenever possible. In this mode, VY intends to embrace the highest defense-in. depth philosophy, assuring that all available systems are available. and/or operable prior to and during the outage activities.
Communication of Manacement Expectations A daily Planning Report documents the status of outage activities and safety system f
configurations, and is distributed to deph ment managers. This report itemizes safety system status with respect to the safety functions performed. The " key" safety functions identified and used by VY are based on NUMARC and industry guidance, and consist of decay heat removal methods, inventory control, reactivity control, primary and secondary containment, and availability of electrical power. In addition, the report states _whether the current plant /systemL lineups correspond to the " minimum," " normal,".or " preferred" configurations. Information regarding the installation of temporary modifications, planned system availability, time-to-boil i
considerations, and maintenance is also provided. The Planning Report also lists an anticipated sequence of events expected to take place within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, based oa the current status of maintenance, critical path activities, and plant configuration. A detailed listing of open work orders is also provided.
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5 Outace Safety Review Committec This four member committee consists ofindividuals with diverse backgrounds, plant knowledge, and familiarity with guidance for the conduct of safe outages. Chanered to identify safety issues associated with the outage schedule and to make recommendations to the Plant Manager, the committee utilized probabilistic risk assessment (PRA) insights, NRC shutdown guidance, and experiences gained from previous outages. The committee reviewed 'he different outage phases against the TS, support systems required for safety system operability, availability of alternate systems, contingency plans, and training. Refueling Outage XVII will consist of seven phases, cach bounded by the shutdown condition of the reactor and the status of primary and secondary containment.
The recommendations made by the committee provided defense-in-depth for a number of outage i
activities. For example, plant management accepted committee recommendations regarding:
(1) rescheduling of "A" RHR system maintenance until after reactor cavity flooding, due to time-to-boil and decay heat removal considerations; (2) throttling of RHRSW valves to minimize system vibration (Section 3.2); (3) prioritization and establishment of an individual responsible for the coordination of switchyard-maintenance to assure h availability and reliability of indepeadent power supplies; and, (4) the establishment and availability of service water fiaw to the "B" emergency diesel generator during system modifications.
Conclusions RFO XVII has been planned with defense-in-depth as a primary consideration. The Outage Guideline formally established responsibilities and management expectations.
Use of the Planning Report by control room personnel and department managers demonstrated the utility of this document. The philosophy of " preferred" plant configurations was reflected in the final draft of the outage schedule. The Outage Safety Review Committee met its charter and VY implemented committee recommendations that improved plant safety by increasing the -
- availability of alternate systems and control of outage activities.
3.0 MAINTENANCE AND TESTING (62703,61726) 3.1 Maintenance The inspectors observed selected maintenance on safety-related eqHpmed to determine whether these activities were effectively conducted in accordance with VY Technical Specifications (TS),
and administrative controls (Procedures AP-0021 and AP-4000). Interviews were conducted with the cognizant engineers and maintenance personnel and vendor equipment manuals were reviewed. Inspections specifichily evaluated work in accordance with approved procedures, safe tagout practices and appropriate industry codes and standards.
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3.2 Failure of Service Water Valve Anti-Rotation Key The repair of residual heat removal service water system valve RHRSW-89A was well y
controlled. Maintenance Department personnel initiated troubleshooting in accordance with the emergency work order, documented as-found conditions, and briefed the cognizant engineer.
Field conditions were reviewed by department management and a maintenance plan was developed.
The Maintenance Department identified that this failure was identical to a previous event t!iat occurred on June 15, in which the motor pinion gear key, located in the motor operator portion of the valve, carae loose. This key splines the drive motor pinion gear to the motor shaft to prevent rotation, and was again found in the motor grease and excessively worn. The motor pinion set screw (which pins the gear to the motor shaft to prevent axial motion) was also worn, but intact. Similar to the previous corrective maintenance, both components were replaced and installed in the identical manner as before, by staking the key in three positions and the use of lock wire on the set screw. To provide further assurance that RHRSW-89A would not fail within the four days prior to the start of RFO XVII, the drive motor and gear assembly were also replaced. The applicable motor operator valve (MOV) maintenance procedure was being revised to include periodic inspection of the motor gear assembly and retaining. devices.
Training was conducted for the maintenance personnel responsible for motor operator inspections. Vermont Yankee attributed the root cause to cyclic stress induced on the key by the motor cycling and system vibration. The licensee also expects that the RHRSW modification scheduled during RFO XVil should preclude recurrence.
3.3 Main Steam Isolation Air Actuator Rebuild During this inspection period, VY rebuilt main steam isolation valve (MSIV) actuators for installation during RFO XVil.
The work was conducted as preventive maintenance in accordance with OP 5303, Rev.15, "MSIV Preventive Maintenance / Calibration Test," and is performed in five year intervals. The actuators, made by the Ralph A. Hiller Company (Hiller),
control the extension and retraction of the MSIVs. Nitrogen is introduced above and below the valve actuating piston through a series of solenoid operated pilot valves and 2-way and 4-way.
valves. The scope of the rebuild activity included replacement of o-rings, gaskets. 2-way/4-way -
valves, and solenoid valves as required by the VY environmental qualification program.
Quality controls were implemented and maintained for replacement parts and lubricants.
Appropriate cleanliness was maintained while valve internals were open during rebuild.
Documentation in the field was complete including correspondence between Hiller.and VY regarding rebuild techniques, torque values, and minor machining required to correct a misalignment problem between the 2-way/4-way valves and valve manifold. One-for-one-Evaluation No.93-079 documented VY's engineering assessment that the machining and.
subsequent bolt change did not represent a change to form, fit, or function. The misalignment L
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t has been evaluated by VY and Hiller as a fabrication issue caused by tight tolerances associated with the forging and casting of the valves and valve manifold; actuator operation was not affected.
Good communications between the I&C technician, engineer,. and Hiller were observed.
However, the inspector observed that the Hiller actuators were being assembled with the old torque values in OP 5303, instead of the new values specified by Hiller and noted in the body of the work package. The MSIV actuators were subsequently rebuilt with the correct torque values. Vermont Yankee acknowledged the inspector's observation and discussed other work control issues with the inspector, such as a change of work scope to address the deviation from OP 5303, and the one-for-one evaluation. No further concems were noted.
3.4 Thermal Performance Monitoring Vermont Yankee implemented a task team to evaluate the thermal performance monitoring program at VY and to compare it to other industry programs. The multi-disciplined task team identified a number of deficiencies with the current program for balance of plant equipment and provided recommendations intended to improve VY's monitoring and assessment capabilities.
In the areas identified as deficient (namely: assessment capabilities, inter-departmental communications, and prioritir.ation of maintenance and design changes that are related to improved thermal performance), recommendations for improvement included: (1) enhanced surveillance, calibration, and maintenance processes for the 500 or so components identified as being important to improved thermal performance; (2) assignment of program responsibility to a single superintendent to lend focus and credibility to a concerted plant-wide effort: and, (3) the use of available thermal monitoring techniques, such as thermography, chemistry, surveillance, and the trending capabilities of the emergency response information system (ERFIS) and the maintenance planning and control system (MPAC). The final disposition of the task team's recommendations is currently in review by corporate management.
3.5 Survei!!ance (61726)
The inspector reviewed procedures, witnessed testing in-progress, and evaluated completed surveillance record packages.
The inspector observed that all tests were performed by qualified and knowledgeable personnel, and in accordance with VY Technical Specifications (TS), and administrative controls (Procedure AP-4000), using approved procedures.
The surveillances which follow were reviewed and were found satisfactory with respect to meeting the safety objectives of the surveillance program:
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OP 4126, Rev. 29, " Diesel Generator Surveillance" OP 4116, Rev.14, " Secondary Containment Surveillance" AP 0125, Rev. 5, " Plant Equipment Control" OP 4120, Rev. 26, "High Pressure Coolant Injection System Surveillance" OP 4121, Rev. 29, " Reactor Core Isolation Cooling System Surveillance" OP 4613, Rev. 20, " Sampling and Testing of Diesel Fuel Oil"
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3.6 Operational Assessment Feedback - HPCI Quick Starts On August 24, VY verified that the high pressure coolant injection (HPCI) and reactor core.
isolation cooling (RCIC) systems reached rated flow wi hin establish?J time requirements. The t
tests were recommended by General Electric (GE) based on their review of HPCI and RCIC testing performed by boiling water reactor utilities. Service Information Letter (SIL) 336, 1
" Surveillance Testing Recommendations for HPCI and RCIC Systems," dated July 11,1980, and SIL 336, Revision 1, dated December 8,1989, were issued recommending quick start (QS) capability and proper valve sequencing testing. Both revisions essentially contained the saine q
information; however, the testing was not implemented into plant procedures until after licensee review of Revision 1. The FSAR describes the capabilities of the RCIC and HPCI systems to reach rated flow, but time-to-full-flow testing is not required by TS.
Vermont Yankee evaluated SIL 336, Revision 1, and took exception to certain of GE's recommendations. One notable exception was VY's evaluation of the QS transient. General Electric stated that it is difficult to monitor the transient visually, and recommended the installation of transient recorders to monitor system performance parameters on a real-time basis (such as turbine acceleration, speed peak, speed oscillations, speed controller outputs, and stop valve position). Vermont Yankee visually observes system operation, qualitatively assesses
- arbine operation, records TS-and IST-required values, and calculates time to full flow. The -
additional trending and predictive activities recommended by the vendor are not quantitatively performed.
Based on the inspector's review of surveillance procedures, TS and SIL requirements.-and test results, the surveillances were conducted in a manner that verified QS capability of both systems. Appropriate acceptance criteria, precautions and limitations, and test requirements were established.
The testing provided further assurance that the systems meet design requiremerits. System flow rates, pressures, flow stability evaluations, and time-to-rated-flow confirmed that the systems were operable.
Vermont Yankee plans to perform this GE recommended testinh once per operating cycle.
4.0 ENGINEERING (71707,37828) 4.1 Review of Written Reports
- l The inspectors reviewed Licensee Event Reports (LERs) submitted to the NRC to verify accuracy, description of cause and adequacy of corrective actions. The inspectors considered f
the need for further information, possible generic implications, and whether the events warranted l
further onsite followup. The LERs were also reviewed with respect to the requirements of 10 CFR.50.73 and the guidance provided in NUREG 1G22.
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9 LER 93-08:
"A" Residual Heat Removal Service Water (RHRSW) Pump Was Not Declared inoperable as Required by the Inse, vice Testing (IST) Program On July 1 during an historical review of the "A" RHRSW pump (P 8-1 A) surveillance test data, VY determined that the pump differential pressure of 151 psid obtained during a May 1991 test should have resulted in a determination that the pump was in the Required Action Range (RAR).
This designation occur:,, in part, when the pressure is less than 152 psid. The cause of this event was the failure of the system engineer (SE) to correctly log the pump performance on the procedure data sheet (VYAPF 0164.02). A contributing cause was the lack of an independent review of test data. Had the pump been declared in the RAR, VY would have been required to declare the pump inoperable and perform alternate testing per TS 4.5.C.2. For reasons other than being in the RAR, alternate testing was accomplished. Subsequent test data for May 1991 and August 1991 indicated that the pump was appropriately performing outside of the IST RAR.
LER 93-08 accurately described the event details and documented appropriate corrective actions including the revision of procedure AP 0164 to include an independent review of the SE's determination. Weaknesses in the IST program area were identified in the NRC's Operational Safety Team inspection (OSTI) Report 93-80, Section 3.3.2. Some of the program changes enumerated above were the result of OSTI-identified concerns, and are being followed as part of Unresolved item 93-80-01.
Periodic and Soecial Reports Vermont Yankee submitted the following periodic and special reports which were reviewed for accuracy and found to be acceptable:
Failed Fuel Action Plan Report for July 28 - August 10 Monthly Statistical Report for July,1993 4.2 Installation of the Vernon Tie Line Modification During the week of August 15, VY made the final electrical connection, completed post-
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maintenance testing, and declared the new Vernon tie line operable. The power line was i
modified to carry full emergency station blackout load (2300 kW,2700 kVA) while protected l
from weather-related damage..Unlike the original tie line, the new power supply line runs underground (for all but a short section at the Vernon switchyard ring bus), is of higher load carrying capability, and is seismically qualified. Termination, isolation, and instrumentation of.
i the 4160 Vac power cable remains unchanged at VY safety buses 3 and 4.
Modifications -
implemented at the power source, the Vernon hydro-station, include a new electrical connection
.I point to the hydro-station ring bus with six independent electrical power sources. A new 13.2-4.16 kV transformer on VY property was also installed.
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Vermont Yankee implemented the moditication to take credit for the Vernon tie line as an alternate ac source as described by 10 CFR 50.63, "less of All Alternating Current Power,"
the station blackout rule. This rule requires, in part, that the ac source be reliable and available in the event of a loss of offsite power concurrent with a turbine trip and loss of the onsite emergency diesel generator capability. The load carrying capacity and capability shall be sufficient to ensure that the reactor is cooled and appropriate containment integnty is maintained in the event of a station blackout. Demonstration of the' time to startup and alignment of the
.i system as an alternate ac source (approximately 10 minutes) is planned for RFO XVII.
The inspector conducted a field inspection of the final cable splice connecting the hydro-station 4
to the VY feeder cable and reviewed the engineering design change request (EDCR)90-412, "Vernon Tie Line Improvement." Plant procedure OP 0023, Rev. 7, " Installation and Testing i
of Cable and Conduit" and ANSI Standards N45.2.2-1972 and N45.2.4-1972 were used as inspection guides. Quality Control (QC) inspections were conducted by the Mercury Company QC staff to assure that the cable splice was in accordance with vendor requirements, and that the installation and test procedures were followed. Parts and materials were free of defects and appropriate cleanliness was maintained.
Oversight by cognizant engineers contributed. to effective communications between VY and the Vernon hydro-station, assuring the electrical safety of maintenance personnel. The post-maintenance testing (PMT) required by EDCR 90-412 and OP 0023 were adequately performed and documented. Prior to declaring the tie line operable, the Operations Department staffindependently verified the acceptability of the PMT results. The inspector considered the modification well-controlled and, for the areas reviewed, l
the modification was installed in accordance with design package EDCR 90-412.
t 4.3 High Pressure Coolant Injection System Debris Screens On August 11, the inspector performed a system walkdown of the turbine casing vent piping for the high pressure coolant injection (HPCI) and reactor core isolation cooling (RCIC) systems, and determined that debris screens installed on the downstream ends of the vent pipes were not described in plant procedures,- drawings, or vendor technical manuals. This condition was brought to the attention of the shift supervisor (SS), inspected by operations personnel, and the control room staff determined that system operability and personnel safety were not compromised. The inspector performed the system walkdown in response to a June 9 event at Commonwealth Edison's Quad Cities Station, Unit 1, in which equipment damage and personnel injuries occurred when the HPCI rupture disks burst during quarterly surveillance testing,_
venting reactor steam directly into HPCI room.
The design at VY differs from Quad Cities in that the rupture disk vent pipe directs turbine exhaust to the air space of the torus room in the reactor building to minimize the potential for personnel injury. The rupture disks (two in series and integral within the vent pipe) are t
designed to protect the turbine casing and associated exhaust piping from an overpressure event if r9 normal exhaust path to the suppression chamber becomes blocked. No concerns for the 1
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4 11 rupture discs themselves were identified, but the debris screens installed at the end of the vent piping were questioned. The status of a preventive maintenance inspection or replacement of the rupture discs was indeterminant at the time of this inspection.
VY was unable to identify the design basis of the debris screens or when they were installed.
On August 31, the Mechanical Engineering & Construction Manager informed the inspector that --
additional evaluation by VY was necessary to determine the adequacy of the installed configuration, for the debris screens at the end of the rupture dise exhaust piping. The initial a
assessment lacked appropriate depth in the area of corrective actions, and a corrective action report (CAR) was initiated. VY was also in the process of completing evaluations for the debris cover design, vent capacity and back pressure effects, rupture disk aging, additional personnel safety concerns, and the controls implemented to install the covers. An assessment is expected to be competed prior to reactor startup from RFO XVII. Pending review of the CAR and the licensee's evaluation of NRC Information Notice 93-67 issued on August 16, 1993, this item remains unresolved (URI 93-19-01).
4.4 Refueling Interlock Temporary Modification With the Reactor Mode Switch (RMS) in the " Refuel" position and following a change out of control rod 26-39, a loss of function of the rod-in signal for this rod occurred on September 1, 1993. This condition prevented an "all rods in" signal from being generated. This resulted in actuation of refueling interlocks that prevented movement of refueling equipment or withdrawal of control rods.
These refueling interlocks are required to be operated in this mode in accordance with TS Section 3.12.A. The bases for this TS is to minimi7e the possibility of loading fuel into a cell containing no control rod. All control rods are required to be fully inserted when fuel is being toaded into the reactor core, to prevent inadvertent criticality.
According to the Final Safety Analysis Report (FSAR), Section 7.6.3, the rod-in condition for each control rod is established by the closure of a magnetically-operated reed switch in the rod l
indicator probe. The rod-in switch must be closed for each rod position before the "all rods in" signal is generated. Initial troubleshooting by the I&C Department determined that the switch position signal was not being transmitted to the rod position indication circuitry. To facilitate the continuation of refueling activity, VY developed Temporary Modification (TM) No. 93-49, which installed a single wire jumper in the rod position indicating system cabinet to provide a simulated full in indication for rod 26-39. The precaution section of the TM specified: (1) rod 26-39 is fully inserted; (2) the rod is electrically disarmed and tagged out; (3) the hydraulic control unit for the rod is isolated and tagged out; and, (4) the tag out cannot be cleared until this TM is restored.
As required by the TM Control Sheet, VY APF 0020.03 and plant administrative procedure AP 0200, Rev.15,." Control of Temporary Modifications," a safety review was conducted that utilized screening criteria to determine if a written safety evaluation would be required. For TM 93-49, a determination was made by VY that no safety evaluation was required. The review was intended to determine if the proposed TM has "the potential to impact the safe operation of the
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12 facility." Although some of the screening questions were answered in the affirmative - the TM (1) operates outside the system design, (2) modifies required system interfaces that support safety system functions, and (3) affects system / equipment protective features-in each case VY referred i
to the precautions enumerated above as the assurance for the rod to remain full in. At the September 1 meeting of the Plant Operations Review Committee (PORC), the TM was approved, as written, with no written safety evaluation.
The inspector reviewed TM 93-49 and determined that VY had made a change in the facility as described in their FSAR, and as such, was required in accordance with 10 CFR 50.59(b)(1)'to develop a written safety evaluation which provides the bases for the determination that the change does not involve an unreviewed safety question as determined in 10 CFR 50.59(a)(2).
Tha inspector was informed by the licensee that the precautions stated in the TM are an integral J.o.i of the modification, which results in "no potential to impact safe operation of the facility."
Therefore, in VY's view, they did not change the facility as described in the FSAR.
The inspector reviewed Quality Assurance Audit Report No. VY 93-07 which documented an April 1993 audit of the design process. During this audit, an evaluation of the TM processes was conducted. The audit noted that changes to the facility which tre screened as "not having an impact on operational safety" may be made without a written safety evaluation, and without prior review by the PORC, even though these changes may affect the facility description in the FSAR. The audit also noted that TS 6.2.A.6.d requires that the PORC review proposed changes to systems which require a change in operating and maintenance procedures identified in the TSs. The audit identified an observation finding that VY's use ofits screening criteria may not be consistent with 10 CFR 50.59. Subsequent VY corrective actions in response to this audit resulted in all TMs being reviewed by PORC. A review of the use of the screening criteria is being conducted by a VY task team, and Yankee Nuclear Services Division managers. This review is scheduled for completion by November 19, 1993.
VY stopped refueling operations on September 1 and performed a 50.59 safety evaluation from TM 93-49. The inspector reviewed this safety evaluation and identified no concerns. Vermont Yankee management initiated an action to require all TMs (such as the jumpering of this refueling interlock) to have a safety evaluation conducted until the use of the screening criteria is resolved. This example associated with TM 93-49 was by itself oflittle safety significance, and both short and longer term corrective actions have been identified by VY to address this issue. Vermont Yankee's use of screening criteria in the TM is therefore unresolved pending further NRC review (URI 93-19-02).
4.5 Fire Barrier Grouting Criteria On August 18 during a training session for fire protection inspectors, VY identified fire j
penetration seal discrepancies on two block walls of the station battery room. The barriers are required by Appendix A of NRC Branch Technical Position 9.5-1 and have a 3-hour rating. The discrepancies were small annular gaps around conduit and strut attachments and minor spalling of grouted penetrations. The barriers were immediately inspected by the VY Fire Protection
a 13 Coordinator, declared inoperable, and an hourly firewatch was stationed. On August 19 fire protection engineers from Yankee Atomic Electric Company assisted VY in an evaluation that concluded that the barriers were operable, although requiring repair. The licensee inspected the other block wall fire barriers and identified similar, yet less pronounced, discrepancies.
The inspector physically inspected the discrepancies, reviewed Procedure OP 4019, Rev. 5,
" Surveillance of Vital Fire Barriers," and concluded that the licensee operability determinations were appropriate and the timeliness of maintenance was commensurate with the safety significance of the deficiencies. During this review, the inspector observed that OP 4109 detaliis specific design requirements; however, some acceptance criteria used during fire barrier surveillances werr, not explicit. For example, the acceptance criteria for grout involves no measurement or visual standard of acceptable grout characteristics. The surveillance criteria for caulk are similar. This was discussed with the cognizant department manager who stated that VY had already determined that engineering judgement may have been inconsistently applied during the performance of " enhanced" fire barrier surveillances (NRC Inspection Report 93-05).
A survey of other block wall penetrations identified no similar deficiencies.
A Potential Reportable Occurrence Report and corrective action report (CAR) were initiated to evaluate the barrier discrepancies, to document the engineering assessment, and to resolve long-term corrective actions. These actions include OP 4109 acceptance criteria, inspector training, and documentation of engineering judgements made in the ficid. The training session was being performed as part of the corrective actions for Nonconformance Report (NCR) No. 92-24,
" Insulated Lines Penetrating Fire Barriers," which, in part, is expected to resolve the fire barrier and surveillance concerns identified in NRC Inspection Report 93-05. The licensee expects to resolve NCR 92 24 by July 1994. The inspector concluded that VY promptly corrected the block wall discrepancies and initiated appropriate actions to further enhance their fire barrier surveillance program.
5.0 PLANT SUPPORT (71707, 61726, TI 2500/028) 5.1 Radiological Controls Inspectors routinely reviewed radiological controls and practices during plant tours. The inspectors observed that posting of contaminated, high airborne radiation, radiation and high radiation areas were in accordance with administrative controls (AP-0500 series procedures) and -
plant instructions. High radiation doors were properly maintained, and equipment and personnel were properly surveyed prior to exit from the radiation control area (RCA). Plant workers were observed to be cogniant of posting requirements and maintained good housekeeping..
The number of personnel contamination events (PCEs) during the first half of 1993 indicated an overall downward trend.
Hot particle contamination events were also down this period.
Vermont Yankee attributes this to improved decontamination and housekeeping practices, as well as better radiological awareness by the work force. Generally, field observations by the
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14 inspectors confirmed VY's assessment that workers are aware of radiological conditions.
Radiological practices have been good, and contaminated areas have been reduced (also refer to NRC Inspection Report 93-01).
5.2 Security The inspector verified that security conditions met regulatory requirements and the VY Physical Security Plan. Physical security was inspected during regular and backshift hours to verify that controls were in accordance with the Security Plan and approved procedures.
On August 24, the Technical Services Superintendent and members of the Security Department.
met with NRC Security Section staff members at the Region I office. The purpose of the meeting was to brief the NRC staff on results of enhancements to improve the security program at the plant and to discuss proposed revisions to the VY Physical Security Plan. Revision' 6 to ;
the VY Training and QualiDeation Plan, submitted by letter cated July 6, was also discussed.
The meeting was closed to the public due to the discussion of safeguards information.
5.3 Employee Concerns Program (TI 2500/028) t The inspector reviewed the Vermont Yankee Employee Concerns Program as part of a-nation-wide survey. The intent was to determine the characteristics of the programs at reactor facilities that provide a path for the expression and resolution of safety concerns. There are no regulatory requirements for these programs, hcwever, the Energy Reorganization Act, Section 211, and 10 CFR 50.7 prohibits discrimination against employees who raise safety concerns.
Vermont Yankee incorporated its safety concerns program within an improvement suggestion program that is designed to encourage identification of improvements to safety, performance, and ef6ciency. The program applies to all company employees and contractors. The program was previously reviewed and addressed in NRC Inspection Report 92-12. Although the program -
is of the same basic form as described in that report, the inspector found that the licensee has made an effort to publicize the program during general employee training and within company publications. The results of the survey conducted are documented in an attachment to this report.
6.0 EXIT MEETING Meetings were held periodically -with VY management during this inspection to discuss inspection Dndings. A summary of preliminary Endings was also discussed at the conclusion of the inspection in an exit meeting held on September 17. No proprietary information was identified as being included in this report.
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