IR 05000244/1995021
| ML17264A391 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 02/29/1996 |
| From: | Doerflein L NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML17264A390 | List: |
| References | |
| 50-244-95-21, NUDOCS 9603060418 | |
| Download: ML17264A391 (23) | |
Text
U. S.
NUCLEAR REGULATORY COMMISSION
REGION I
Inspection Report 50-244/95-21 License:
DPR-18 Facility:
R.
E. Ginna Nuclear Power Plant Rochester Gas and Electric Corporation (RGSE)
Inspection:
December 3,
1995 through January 27, 1996 a
e Inspectors:
P.
D. Drysdale, Senior Resident Inspector, Ginna E.
C. Knutson, Resident Inspector, Ginna J.
D. Noggle, Senior Radiation Specialist, Division of Reactor Safety (DRS),
Region I G. Bagchi, Chief, Civil Engineering and Geosciences Branch, Office of Nuclear Reactor Regulation (NRR)
S.
K. Chaudhary, Senior Reactor Engineer, Materials, DRS, Region I J.
S.
Ha, Structural Engineer, NRR E.
H. Gray, Technical Assistant, DRS, Region I Approved by: ~~~~~ ~!n/
< >>
'l(o oer eon,
>e Reactor Projects Branch
Division of Reactor Projects Ins ection Summar:
Core, regional initiative, and reactive inspections performed by resident and specialist inspectors during plant activities are documented in the areas of plant operations, maintenance, engineering, and plant support.
Results:
See Executive Summary.
.96030604i8 960229 PDR ADOCK 05000244
EXECUTIVE SUMMARY R.
E. Ginna Nuclear Power Plant Ins ection Re ort No. 50-244 95-21
~0erati ons:
The plant operated at full power (approximately 97 percent)
throughout the inspection period.
The facility was operated safely and in accordance with licensee procedures.
Operating logs and records accurately identified equipment status or deficiencies.
No operational inadequacies or concerns were identified.
For the entire report period, the plant maintained a steady full power output without experiencing any significant abnormal conditions or transients.
The effort to have plant operating and test procedure changes completed prior to implementation of Improved Technical Specifications appears to be on schedule.
The number of procedures effected by this effort is significant, but it appears that all necessary changes have been identified and will be available when needed.
The plant operating organization and the Plant Operations Review Committee are effectively adopting the conservative operating philosophy policy and program initiated within the site groups last year.
Their conservative decisions on several occas ons demonstrated an operational judgement that reflected a clear priority for plant and personnel safety.
Maintenance:
The performance of maintenance activities was good.
This included troubleshooting of DB-50 air circuit breaker for the B-CCW pump; replacement of the diesel driven fire pump engine; and the ongoing maintenance related to the charging pumps, especially their vari-drive units.
The licensee has devoted significant effort to identifying and correcting performance and maintenance problems with the pumps.
The corrective actions for these problems have been effective.
Surveillance testing conducted during this period was well controlled.
Personnel were alert to the operating conditions of equipment being tested, and promptly evaluated abnormal conditions.
A review of the engineering backlog identified a relatively large number of engineering work tasks in progress.
This item was also identified by RGEE auditors and is currently being addressed by engineering management.
New goals have been established for maintaining open engineering items at manageable levels, and the licensee is working on several new initiatives in the area of engineering services.
Some engineering work processes are being revised and will be trended for effectiveness.
The inspector considered that
the engineering backlog has been effectively managed, in that it has not adversely affected plant operations and has not significantly affected the functioning of any licensee organizations.
The planning and preparations for the replacement of the steam generators are being well executed.
The planning appears thorough and slightly ahead of schedule.
The licensee is placing an appropriate emphasis on safety of the operation and plant systems affected by this project.
A review of the preparation of radiation safety activities in support of the planned l996 steam generator replacement project (SGRP)
noted that excellent research and use of lessons learned from steam generator replacement projects at other nuclear facilities had been incorporated into the licensee's radiation controls plan.
The inspector determined that very good planning and preparations were in progress to accomplish the SGRP safely and with relatively low personnel exposures, RGItE has incorporated a mixture of permanent licensee radiation controls staff and experienced contracted personnel into the steam generator replacement project.
The licensee has acquired additional radiation survey instrumentation, additional mockup training facilities, and has constructed a temporary access control facility for containment personnel access.
There were no significant radiation safety issues identified.
The inspector also reviewed the licensee's actions relative to a licensee-identified plant in-leakage of contaminated water into the containment perimeter sump.
Preliminarily, the licensee is investigating the possibility that the leakage source is the spent fuel pool and has plans to quantify the total leakage from the spent fuel pool.
The licensee has performed tritium and gamma analyses on various liquid samples collected from accessible plant drainage and environmental monitoring wells.
These results indicate that there is slight tritium contamination seeping into the containment perimeter water collection system.
The licensee has initiated an investigation to identify the source and extent of the leakage.
No regulatory limits have been exceeded and no violations were identified.
Safet Assessment ualit Verification:
The NSARB completed its regularly scheduled quarterly meeting on January 24-25 at the Ginna station.
The meeting included a full agenda of review items and presentations from plant management.
The board also devoted a significant review effort to the five major program activities including new steam generators, Improved Technical Specifications, reactor coolant system average temperature (T.,) reduction, 18-month fuel cycle, and feedwater regulator valve trim changes.
All issues before the NSARB were discussed, and the board was effective in identifying areas needing followup actio TABLE OF CONTENTS EXECUTIVE SUMMARY.
TABLE OF CONTENTS
.
1. 0 OPERATIONS l. 1 Operations Overview 1.2 Control of Operations 1.3 Preparations for Implementation of Improved Technical Specifications
.
.
1.4 Implementation of Conservative Operating Philosophy 2.0 MAINTENANCE.
2.1 Maintenance Activities.
.
2. 1. 1 Routine Observations 2. 1.2 A-and C-Charging Pumps Maintenance and Testing
.
2.2 Surveillance and Testing Activities 2.2. 1 Routine Observations 2.2.2 B-Component Cooling Water Pump Breaker Malfunction 3.0 ENGINEERING
.
3.)
Review of Engineering Backlog 3.2 (Update)
Steam Generator Replacement Project 3.2. 1 Project Status 3.2.2 Observations
~
~
~
~
1V
1
1
4
5
6
9
10
11 4.0 PLANT SUPPORT
,
4. I Steam Generator Replacement Radiation 4.2 (Open)
Inspector Follow Item - Investi Leakage (50-244/95-15-01)
~
~
~
~
~
~
~
~
Controls Prepara gation of Spent t1on Fuel Pool
5.0 SAFETY ASSESSMENT/EQUALITY VERIFICATION 5. 1 Nuclear Safety Audit and Review Board 5.2 Periodic Reports
.
5.3 Licensee Event Report 6. 0 ADMINIST RATIVE 6. 1 Senior NRC Management Site Visit
.
6.2 Exit Meetings (NSARB) Meet 1 llg
15
16
16
DETAILS 1.0 OPERATIONS (Inspection Procedure (IP) 71707)'.
Operations Overview The plant operated at full power (approximately 97 percent)
throughout the inspection period.
There were no significant operational events or significant challenges to plant equipment during the inspection period.
Operator and plant performance was assessed as good during the reporting period.
1.2 Control of Operations The inspectors observed plant operation and verified that the facility was operated safely and in accordance with licensee procedures and regulatory requirements.
This review included tours of the accessible areas of the facility, verification of engineered safeguards features (ESF)
system operability, verification of proper control room and shift staffing, verification that the plant was operated in conformance with technical specifications and that appropriate action statements for out-o'-service equipment were implemented, and verification that logs and records were accurate and identified equipment status or deficiencies.
No operational inadequacies or concerns were identified.
1.3 Preparations for Implementation of Improved Technical Specifications In Nay 1995, RG&E developed a comprehensive plan to assure overall organizational readiness to implement the Improved Technical Specifications (ITS) expected in February 1996 (i.e., prior to the receipt of new fuel for the April outage).
The plan included the development of several new programs required by tne ITS, and an implementation schedule for the key aspects of these programs.
One major program identified was the need to.upgrade plant procedures to reflect the new ITS requirements, especially for plant equipment operability and surveillance testing.
An ITS review sub-committee was formed and a lead engineer was assigned responsibility for-coordinating the procedure change effort among the various plant departments.
Representatives from each of the principle site departments participated on the sub-committee to develop a set of procedure change criteria, to identify new operability and surveillance test requirements in the ITS, and to develop a schedule for changing existing procedures or writing new ones needed to implement the ITS, The sub-committee also developed a preliminary procedure matrix to identify what existing procedures were expected to already satisfy the ITS requirements.
An operations supervisor lead the effort to change all necessary operating procedures, and the Hanager of Results and Test led the effort for surveillance tests.
The sub-committee lead engineer was responsible for changing all site administrative procedures and other higher level documents.
A total of 1380 new and changed station procedures will be required for ITS implementation, of which approximately 465 are operations and I
The NRC inspection manual procedure or temporary instruction that was used as inspection guidance is listed for each applicable report sectio test procedures.
This includes approximately two new operating procedures, and four new surveillance test procedures.
Approximately ten old operations/
test procedures will be deleted.
One significant difference that the ITS imposes on the Ginna station is standard operating modes (I 6), where these were previously not clearly defined in terms of specific plant conditions.
Consequently, the operating procedures needed to be changed to reflect these modes and the appropriate minimum conditions required for a mode change.
The ITS also contains an updated LCO and action statement logic which required substantial revisions to the licensee's administrative procedures A-52.4 "Control of Limiting Conditions For Operating Equipment,"
and A-52. 12, "Inoperability of Equipment Important To Safety."
Another significant difference in the ITS is that all equipment operability requirements now have an explicit surveillance test associated with them.
Surveillance tests that were previously performed on a monthly basis are now only required to be performed quarterly.
The licensee has evaluated the surveillance test intervals on a case-by-case basis and adjusted the interval to quarterly only where equipment performance clearly warrants this change.
The remaining surveillance tests will continue on a monthly basis until the licensee is satisfied that the interval can be extended.
Other surveillances that were previously performed annually (i.e.,
each refueling outage) will now be performed at least biennially.
The ITS also includes operability and surveillance test requirements for some equipment that is not included in the existing TS.
This has required the development of new surveillance test procedures for that equipment.
For example, the containment hydrogen (H,) recombiners are not in the existing TS; however, the ITS contains operability and functional verification requirements for the recombiners.
Some surveillance tests and oper ability requirements no longer reside in the ITS and have been relocated to a new "Technical Requirements Hanual,"
(TRN).
These tests and operability requirements were removed from the ITS because they are not necessary to support any accident analysis.
However, the licensee will continue to maintain all of the past requirements and procedures, as necessary, until they are altered or deleted through the
CFR 50.59 process.
The licensee also developed a procedure matrix that correlated all new ITS surveillance requirements (SRs) with specific procedure and paragraph numbers were a particular SR was implemented.
This index will be incorporated into the TRN for use by the operating shift and the results and test organization.
Many operating and test procedure changes are "Reference Only" changes.
These include a large number of procedures that did not require any technical change for the ITS, but that contain a reference to, for example, a specific paragraph in the ITS.
As with any procedure change at the Ginna station, all changes for ITS implementation will be processed through the licensee's screening controls for 10 CFR 50.59 and the necessary PORC approvals prior to issuance to ensure that all required technical changes are mad The inspector concluded that the licensee's ongoing efforts have identified the necessary operating and test procedure changes and that these changes will be issv~~
on time to support implementation of the ITS in February 1996.
1.4 Implementation of Conservative Operating Philosophy In response to an operating event at another nuclear facility in April 1994, where non-conservative operator decisions and plant equipment problems resulted in a reactor trip, RG&E initiated a formal policy and program on conservative decision making to institutionalize management's expectations.
The policy was aimed primarily at all plant staff, but was especially focused toward the on-shift operating crews.
The licensee planned to instill a conservative decision making philosophy within the operating organization emphasizing a high priority on safe operation of the plant when responding to abnormal conditions, and that plant and personnel safety will always take precedence over schedule or commercial considerations.
The Plant Hanager assumed overall responsibility to develop this program and the Operations Hanager pursued implementation within the onshift operating organization.
The licensee formalized their conservative philosophy by establishing a high level policy statement in the Nuclear Policy Directive.
The policy was reiterated during regular shift supervisors meetings, and was also incorporated into the routine simulator and classroom training for operators.
During a shift supervisor's meeting in January 1995, the Operations Hanager reinforced the expectation that conservative decisions have a higher priority for plant and personnel safety over normally scheduled shift work.
In a subsequent shift supervisor meeting in September 1995, station management openly reinforced the same policy by reiterating, without question, that shift supervisor and operator actions were appropriate when they I) manually tripped the reactor in response to the loss a circulating water pump; 2) avoided testing the B-EDG during a severe storm that eventually caused the loss of offsite power circuit no.
751 and automatically started the B-EDG; and 3)
started the B-EDG to energize Safeguards Bus No.
17 after the failure of a fuse in its undervoltage protection cabinet.
As a result of the above plant events, operations management followed-up to proceduralize the conservative philosophy by initiating administrative changes to the rapid load reduction procedure (AP-TURB-5) and the EOP/AP procedure users guide (A-503.1) to direct that a manual reactor trip is to be initiated whenever critical plant parameters approach their trip setpoints.
In addition, administrative procedure A-52. 1, "Shift Organization and Responsibilities,"
charges all shift crew personnel with the authority and responsibility For safe and deliberate reactivity management, and conservative operational decisions, to include taking the main turbine out of service and initiating a manual reactor trip when faced with uncertain conditions.
RG&E's quality assurance (gA) auditors have recently observed apparent weaknesses in the amount of operator classroom training that was designed to reinforce the conservative operating philosophy.
These observations were communicated to the station management as a potential shortfall in their expectations in this area; however, it did not appear to affect operator's knowledge of the conservative philosoph A conservative approach to operating decisions was recently evidenced within the operations department and the plant operations review committee (PORC)
during an evaluation of plant equipment out of service.
The A-Charging Pump was out of service and inoperable for maintenance when a slight level of water contamination above the vendor limit was discovered in the lube oil of the 8-EDG.
The 8-EDG provides the source of emergency electrical power to the operating 8 and C-charging pumps.
The operations department considered removing the 8-EDG from service to inspect and troubleshoot the source of water, but first consulted with licensing for a Technical Specifications interpretation and an evaluation of single-failure susceptibility for the operating charging pumps.
The licensing department appropriately concluded that the Technical Specifications requirements would be satisfied, and that another emergency power source to the operating charging pumps could be made available if the 8-EDG and offsite power sources were not.
However, the PORC made a conservative operating decision to leave the 8-EDG in service and available.
It further concluded that the slight water contamination was not high enough to make the diesel inoperable, and that the water source could be adequately evaluated through increased sampling and non-intrusive testing.
The PORC also developed an action plan to establish an administrative limit on water contamination in the lube oil, and other contingencies for maintaining the 8-EDG operable.
The source of the water is still under investigation.
The inspector considered that the licensee's conservative operating philosophy was effectively demonstrated in the above instance.
The PORC appropriately evaluated the potential risks and consequences of removing the 8-EDG from service with the A-charging pump unavailable, and concluded that the 8-EDG should remain in service.
Although the licensee has identified some weaknesses in operator training in the policy on conservative operating decisions, the inspector concluded that this policy is being instilled within the plant organizations and functioned well in this case.
2.0 MAINTENANCE (IPs 62703, 61726)
2.1 Maintenance Activities 2.1.1 Routine Observations The inspector observed portions of plant maintenance activities to verify that the correct parts and tools were utilized, the applicable industry code and technical specification requirements were satisfied, adequate measures were in place to ensure personnel safety and prevent damage to plant structures, systems, and components, and to ensure that equipment operability was verified upon completion of post maintenance testing.
The following maintenance activities were observed:
~
C-charging pump vari-drive belt replacement, observed December 28, 1995 This item was not completed due to an inability to achieve sufficient tension on the drive belts during reassembly.
As a result, the vari-drive unit was overhauled.
Concurrently, the charging pump plunger check valves were replaced as part of the effort to troubleshoot excessive vibration of the suction piping; this problem is further
discussed in section 2. 1.2 of this report.
Maintenance on the C-charging pump was completed on January 23, 1996.
DB-50 air circuit breaker troubleshooting, observed January 9,
1996 Troubleshooting was conducted as a result of problems encountered when attempting to secure the 8-component cooling water pump during a
surveillance test on January 3,
1996.
This item is further discussed in section 2.2.2 of this report.
~
Diesel driven fire pump engine replacement, observed over the course of the inspection period.
The diesel was replaced due to progressive degradation of the head gaskets and the water pump, as well as unavailability of replacement parts.
The replacement commenced on January 15, 1996 and was completed on January 24, 1996.
The Ginna Station Updated Final Safety Analysis Report only requires that, if one of the two station fire pumps
{one diesel driven and one motor driven) is inoperable for more than seven days, a 30-day special report be submitted to the NRC.
The inspector concluded that the above activities were well controlled.
Morker s were proficient in trade skills and utilized approved procedures.
The inspector observed excellent support by the radiological protection technician during the C-charging pump maintenance.
Engineering personnel were directly involved in the DB-50 air circuit breaker troubleshooting.
2. 1.2 A-and C-Char ging Pumps Naintenance and Testing The chemical and volume control
{CVCS) system at Ginna uses three high head positive displacement charging pumps for reactor coolant system inventory control and chemical addition.
These pumps are not part of the emergency core cooling system, but are safety significant in that at least one pump is required to maintain injection flow to the reactor coolant pump shaft seals.
During normal operations, two pumps are operated.
A minimum of one pump is required to be operable by the technical specifications
{TS).
During this operating cycle, the licensee experienced performance problems with the charging pumps on several occasions.
As discussed in inspection report 50-244/95-17, the A-charging pump vari-drive speed control unit, which had been overhauled during the 1995 annual refueling outage, again required overhaul in July 1995 due to maintenance-related problems.
In August 1995, the C-charging pump discharge relief valve lifted, as indicated by flow noise and vibration in the discharge line to the volume control tank.
Following replacement of the relief valve, operators noted that the C-charging pump suction line, and other associated piping, continued to vibrate more than normal.
The relief valve was twice again replaced due to seat leakage, and, on November 24, 1995, the C-charging pump was declared inoperable due to a
cracked weld in the charging pump plunger hood vent line.
Problems with the C-charging pump are further discussed in inspection report 50-244/95-2 During this inspection period, the C-charging pump plunger hood vent line was repaired and the pump was returned to service on December 12, 1995.
A troubleshooting procedure was conducted to determine the cause of the abnormal suction line vibrations.
Haintenance procedure EH-788, "Testing Charging System Suction Line Vibrations," measured parameters such as vibration, flow, pressure, temperature, and pump acoustics, with various combinations of charging pumps in operation.
This testing identified the C-charging pump as having a different acoustic signature than the other charging pumps.
On Oecember 26, 1995, operators noted that the C-charginig pump vari-drive was making excessive noise
~
As pump speed was decreased in an effort to reseat the drive belts, the pump plungers were observed to stop.
The pump was declared inoperable and work was initiated to replace the drive belts.
This maintenance, however, failed to correct the problem with low speed operation, and a complete overhaul of the vari-drive was performed.
Lessons learned from earlier work on the A-charging pump vari-drive were employed during this maintenance.
Specifically, a revised maintenance procedure was used, and the vari-drive rebuild was performed in the shop to allow for better fitup of parts.
Additionally, the licensee noted that, due to personnel turnover in the machinist shop, the level of experience with vari-drive units had declined; therefore, on-the-job training received additional emphasis.
In addition to the vari-drive overhaul, the pump plunger suction and discharge check valves were replaced in an attempt to eliminate the suction piping vibration.
Haintenance on the C-charging pump was completed on January 23, 1996.
The pump vari-drive operated satisfactorily and the lack of vibration in the suction piping indicated i,liat the check valve replacements had corrected the cause of the vibration.
Initially, the pump was to be allowed to run in for several days prior to acceptance testing; however, later that same day, the A-charging pump vari-drive began to make excessive noise.
The A-charging pump was declared inoperable, which placed the licensee in a 72-hour TS action statement.
As a result, the licensee accelerated its schedule for acceptance testing of the C-charging pump, and declared the pump operable on January 23, 1996.
Following disassembly and inspection, the licensee commenced an overhaul of the A-charging pump vari-drive; this maintenance was in progress at the close of the inspection period.
The inspector considered that the licensee's corrective actions for problems with the A-and C-charging pumps were effective.
Troubleshooting of the suction piping vibration was thorough and appears to have identified the cause of the problem.
Overhaul and assembly of the vari-drive unit in the shop facilitated proper fitup and balance of key components, and allowed for increased on-the-job training.
2.2 Surveillance and Testing Activities 2.2. 1 Routine Observations Inspectors observed portions of surveillance tests to verify proper calibration of test instrumentation, use of approved procedures, performance of work by qualified personnel, conformance to limiting conditions for
operation (LCOs),
and correct post-test system restoration.
The following surveillances were observed:
Periodic Test (PT)-2. IH, "Safety Injection System Honthly Test," observed December 28, 1995
~
Instrument Calibration Procedure (CPI)-AXIAL-N43, "Calibration of Nuclear Instrumentation System Power Range N43 Axial Offset," observed January 9,
1996 Output from the high voltage power supply (NQ-303)
was initially found to be out-of-tolerance low.
A technician adjusted the voltage as allowed by the procedure and continued with the calibration.
However, following final adjustment at the conclusion of the procedure, the technician noted that the output from NQ-303 continued to slowly increase.
A work request was initiated and the high voltage power supply was replaced (work order 19600171) prior to returning N43 to service.
~
PT-3Q,
"Containment Spray Pump Quarterly Test," observed January 10, 1996 While the A-containment spray (CS)
pump was operating, level in the auto-fill oil reservoir for the pump bearing slowly decreased from approximately 1/4-full to empty.
The technician immediately reported this to operations personnel and the pump was secured.
The oil reservoir was refilled and the pump restarted.
After initially continuing to decrease, level in the reservoir stabilized at approximately 2/3-full.
The surveillance was terminated upon completion of the A-CS pump test to determine whether additional corrective action was required.
The condition was determined to be due to a very slow oil leak from the pump bearing which led to normal (although infrequently observed and therefore not common knowledge) operation of the reservoir to make up oil to the bearing housing.
The oil leak had been previously documented by a work request and is scheduled to be repaired during planned maintenance on the A-CS pump in February 1996, PT-12.2B,
"Emergency Diesel Generator B," observed January 11, 1996
~
PT-16Q-B, "Auxiliary Feedwater Pump B Quarterly," observed January 16, 1996
PT-12.2A,
"Emergency Diesel Generator A," observed January 19, 1996
~
PT-16Q-T, "Auxiliary Feedwater Turbine Pump - Quarterly," observed January 22, 1996 The inspector determined through observing the above surveillance tests that operations and test personnel adhered to procedures, that test results and equipment operating parameters met applicable acceptance criteria, and that redundant equipment was available during testing for emergency operation.
Operations personnel and test technicians demonstrated conservatism in
pursuing the power supply drift problem that was identified during the N43 nuclear instrument calibration and the low level in the auto-fill oil reservoir for the A-CS pump that occurred during the CS pump surveillance.
2.2.2 8-Component Cooling Water Pump Breaker Malfunction On January 3,
1996, the licensee performed PT-2.8g,
"Component Cooling Water (CCW)
Pump quarterly Test."
When testing of the B-CCW pump was completed, the technician requested that operations personnel switch running pumps to the A-CCW pump, to continue the surveillance.
From the main control board, the operator started the A-CCW pump and attempted to secure the B-CCW pump.
The B-CCW pump breaker indicated open when the switch'as taken to the stop position, but immediately went back to the shut indication when the operator released the switch.
The operator attempted to stop the 8-CCW pump two additional times, with the same results.
However, approximately 10 seconds after releasing the switch after the third attempt, the breaker shut indication extinguished, and the breaker open and position disagreement indicators lit (indicating that the actual breaker position was not as required by the position of the control switch).
An operator locally verified that the B-CCW pump breaker was open, and the B-CCW pump was declared inoperable.
This placed the licensee in a 24-hour action statement per TS 3.3.3.2.a
~
The B-CCW pump breaker was racked out for troubleshooting.
No mechanical problems were evident and when the breaker was cycled shut and open, it operated cleanly.
Electrical troubleshooting of the breaker and the control circuit up to the NCB switch did not identify any problems.
The breaker was reinstalled and operated satisfactorily during a trial run of the B-CCW pump.
The licensee theorized that the cause of the malfunction had been corrosion on
.
the switchboard-to-breaker connection for the breaker auxiliary contacts.
This resulted in reduced voltage being supplied to the breaker trip coil, which, in turn, did not have enough power to unlatch the tripper bar.
The repeated attempts to open the breaker, however, had moved the tripper bar into a "hair-trigger" condition, such that normal vibrations in the equipment finally caused the breaker to trip.
This scenario could not be repeated, because corrosion on the switchboard-to-breaker connections would h~.'e been wiped off (due to the friction fit) when the breaker was removed for troubleshooting.
Based on the satisfactory trial run, the B-CCW pump was declared operable on January 3,
1996.
The licensee continued efforts to positively identify the mode of failure.
In-shop testing, using a spare breaker and a simplified control circuit, failed to reproduce the problem.
However, this testing did demonstrate that the original "hair-trigger" scenario was highly unlikely.
A historical review identified that nearly the same problem had occurred
years earlier.
Since the same breaker was still in service, the licensee decided to replace the B-CCW pump breaker to allow in-depth troubleshooting of the suspect breaker.
Prior to the breaker replacement, a test was performed on January 23, 1996, to attempt to reproduce the problem with the B-CCW pump, This test included CCW system hydraulic monitoring, motor electrical monitoring, and examination of a
suspect pressure switch/relay (PIC-617) in the pump automatic start control circuit.
The problem did not recur during the test, and all test data indicated normal system/component operation.
Following the test, the breaker and PIC-617 relay were replaced, and the B-CCW pump was returned to service.
At the close of the inspection period, the licensee was preparing to disassemble these components for further inspection.
The inspector considered that the licensee's initial response to the problem with the B-CCW pump breaker was adequate.
Performing additional in-shop troubleshooting was appropriate, given that the cause of the problem had not been definitively established.
The decision to replace the breaker when additional testing was unsuccessful in identifying the cause of the problem was conservative and displayed the licensee's strong safety perspective.
3.0 ENGINEERING (IPs 71707, 37551)
3.1 Review of Engineering Backlog The inspector reviewed the status of requests for engineering support to determine the magnitude of the engineering workload and assess the department's effectiveness at dispositioning these requests.
In conducting this review, the inspector examined two of the principal processes for requesting engineering support at Ginna.
These two processes are Technical Staff Requests (TSRs)
and Engineering Work Requests (EWRs).
The TSR process is governed by administrative procedure A-300, "Preparation and Disposition of a Technical Staff Request,"
and is generally used to request routine engineering assistance.
The EWR process is governed by A-55, "Requests for Engineering Services,"
and engineering procedure gE-342,
"Requests for Nuclear Engineering Services,"
and is generally used when more extensive engineering assistance is required.
EWRs, in turn, are processed either as "short form" or "long form" requests.
A short form EWR is used for non-modification requests that will require short term support.
A long form EWR is used for larger projects and modifications.
The inspector determined that, as of December 22, 1995, the licensee had 512 open TSRs.
Of these, 266 were back-logged (that is, not yet being addressed)
and 134 were in process; the remainder were either complete through engineering, in the process of being closed out, or were in an indeterminate status.
Also as oF December 22, 1995, the licensee had 191 open long form EWRs, of which seven were back logged and 105 were in process, and 112 open short form EWRs, of which 14 were back logged and 83 were in process.
The inspector observed that the licensee has recently audited and continuously monitors the engineering work backlog.
Goals have been established to reduced the number of open TSRs and EWRs.
The licensee noted in a recent self-assessment that, as of the end of 1995, little progress had been made on reducing the backlog of TSRs and EWRs.
However, several initiatives in the area of engineering services are under development.
A single entry point program to replace TSRs and EWRs is scheduled to be in place by mid-1996.
Interface procedure IP-DES-1,
"Engineering Support Requests,"
is intended to provide for more effective allocation of engineering resources through request
screening and prioritization.
The licensee is also developing new performance indicators and an Engineering Work Tracking System (EWTS) to enhance trending of engineering services.
The inspector considered that, although relatively large, the engineering workload has been effectively managed.
Within the inspector's experience, the engineering backlog has not adversely affected plant operations and has not significantly affected the functioning of any licensee organizations.
The inspector considered that the licensee's actions to reduce the size of the backlog through improved allocation of engineering resources was a good initiative.
3.2 (Update)
Steam Generator Replacement Project 3.2. 1 Project Status RGKE's project to replace both steam generators (SGs) during the next refueling outage (April 1996) is currently on schedule and onsite preparation activities are ongoing.
Overall, the project continues to be well managed and coordinated between RGKE and Bechtel.
The major cohstruction activities now completed or currently in progress are as follows:
~
The test excavation of concrete and steel rebar on the full-scale containment dome mock-up is complete.
The mock-up liner plate has been cut, additional plate steel reinforcements are installed, and a custom made lifting strongback has been attached for removal of the plate.
~
Installation of the equipment support and work platform ("CROWN" ) on the containment dome is essentially complete.
Grouting of the structural column baseplates is the last task to complete the CROWN assembly.
~
The new containment access facility (CAF) is essentially complete and the building is occupied by radiological controls and various contractor personnel.
The bridge walkway from the upper level of the CAF to the intermediate building is still in progress.
Actual penetration of the intermediate building wall for containment access is not planned before mid Harch 1996.
e All of the components of the Lampson TransiLift crane are onsite and assembly of the crane is approximately SOX complete.
The main structural members are assembled and are ready for erection.
The support cables for the main boom are being strung, and the forward crawler is ready for connection to the rear crawler.
Final assembly of the main lifting cables and load blocks will proceed after all structural members are assembled.
~
Fabrication of both new steam generators (SGs) is complete.
Both SGs were transported to the Port of Hamilton, Ontario, Canada.
One SG was loaded onto a barge and transported to Rochester, NY after weather conditions permitted.
The licensee has started to assemble a temporary work facility to receive and prepare the new SGs for installatio.2.2 Observations On Decem"..r 7 and 8, 1995, staff from NRR - Division nf engineering and Region I - Division of Reactor Safety performed a joint audit and inspection of the proposed construction of two large openings in the containment dome to permit crane access for steam generator replacement.
One of the objectives of the inspection was to assess the effectiveness of a "Brokk" hydraulic jackhammer being tested and qualified on a mockup of the concrete dome.
The mockup is a test piece made of reinforced concrete of the same thickness as the dome (2.5 feet) with a steel liner on the under side.
The mockup has a
single curvature, as opposed to the double curvature of the actual containment dome.
The Ginna containment structure consists of a flat basemat, a cylindrical upper part prestressed only in the vertical direction and a reinforced concrete dome - all lined with a leak-tight steel plate liner on the inside surface.
Based on prior licensee meetings and discussions concerning SG replacement, the inspectors expressed a concern about the potential for a jackhammer impact to propagate cracks throughout concrete into areas away from the construction openings.
During the inspection of the dome mockup the inspectors noted that, except for the area to be cut by the jackhammer, the entire dome mockup was closely and continuously supported by steel frame shorings.
Thus, the load transmission path for the jackhammer blows would not be simulated in the dome mockup, because the containment dome was not supported in a similar manner.
Although the test was effective in establishing the cutting and removal process of concrete, cutting and fit up of the liner plate, and training for crafts people in general, it was not fully effective in demonstrating that the impact forces would not propagate cracks.
A close scrutiny of the mockup concrete block did show some cracks in the vicinity of the jackhammer impact points, suggesting a need to include crack mapping and close observation of crack propagation, if any, throughout the use of the jackhammer on the dome.
The licensee agreed to consider this point.
The inspectors noted that construction activities with respect to the steam generator replacement project were ahead of schedule, and the planning and engineering of the project were satisfactory.
4.0 PLANT SUPPORT (IPs 71750, 84750)
4.1 Steam Generator Replacement Radiation Controls Preparation As discussed in section 3.2 of this report, the licensee will undertake SG replacement during the 1996 refueling outage.
Due to the radiological hazards associated with replacement of these major reactor coolant system components, a readiness review of the radiation controls program was performed.
The licensee has spent greater than two years in planning the radiological support for the steam generator replacement project (SGRP).
Various exposure reduction and radiation protection (RP) contract cost incentives have been established for the SG replacement contractors.
During the previous 1995
refueling outage, 12.9 person-rem was attributed to the SGRP.
During the 1996 outage, an additional ll5 person-rem is estimated for,a total project expenditure of 128 person-rem as currently projected.
A final SGRP exposure goal is expected to be somewhat lower than this estimate, but was not available at the time of this inspection.
The inspector reviewed the licensee's radiation control staffing plans for the SGRP.
Approximately 49 RP technicians and 22 decontamination technicians are currently planned for the project.
The inspector determined that the licensee had developed a very good RP organizational team with a good mixture of experienced contractors and permanent RP personnel.
Personnel resources appear to be adequate and properly organized to meet the needs of the steam generator replacement project.
The licensee has provided for significant additional radiation survey instrumentation to support the SGRP.
In addition, approximately 14 closed circuit television cameras and additional radio communications will be utilized for the project.
A separate personnel access facility has been built to facilitate all containment entries for the outage through a temporary two-story building that provides change rooms, protective clothing, dosimetry issue, RP briefing areas and closed circuit television-monitoring stations.
The temporary facility appeared to be more than adequate for the outage needs.
The mockup facilities were reviewed and consisted of a full-scale steam generator channel head and a full-scale reactor coolant system (RCS) pipe loop.
In addition, a containment dome concrete/rebar mockup had been built to allow demolition and reconstruction techniques to be tested.
Mockup training qualifications require both a work supervisor and an ALARA sup'ervisor approval.
The mockup training activities that are planned include:
RCS pipe cutting, welding and radiography; removal and installation of the steam generators; RCS pipe end decontamination, debris dam and shielding installation; dome concrete and liner replacement; and lead shielding installation.
The inspector noted that the mockup training schedule for the general workforce was one-two weeks earlier than the availability of contract RP technicians.
The licensee indicated that the schedule would be reevaluated to optimize mockup training with RP technicians.
The overall project shielding, ventilation, and decontamination planning had been developed and was reviewed by the inspector.
The shielding plans incorporated the steam generator and RCS sources, as well as Ginna-specific interfering sources such as residual heat removal piping and pressurizer spray piping.
Transit areas, as well as work areas, were planned for with a total projected shielding effort of 63,000 pounds of lead.
The inspector determined that appropriate efforts were planned to reduce work area dose rates for the SGRP.
Ventilation planning specified one 2500 cubic feet per minute (CFH)
high efficiency particulate air (HEPA) unit per loop area; one 2000 CFH HEPA unit for the RCS pipe decon evolutions; and six 1000 CFN HEPA units for various project requirements.
Approximately eight HEPA vacuum cleaners are currently planned to support the major SGRP work locations.
A tool decontamination trailer will be leased to accommodate the additional tool and equipment decontamination requirements of the project.
The inspector determined that adequate planning and resources were devoted to the shielding,
ventilation, and decontamination aspects of the steam generator replacement project.
Implementation of the above-mentioned radiological. engineering controls will be reviewed during the project.
At the time of this inspection, the licensee was finalizing the work plan and inspection records (WPIRs) for the SGRP.
The WPIRs provided the work instructions to the craft personnel to complete each job requirement.
Several RP/ALARA holdpoints or instructions to contact RP were included in these work instruction documents.
In general, radiological controls were not specified in the WPIRs.
Details of job coverage and radiological control of work had not been documented at the time of this inspection.
Radiation work permits will be written by the licensee at a later date and will be reviewed by the inspector during the outage.
SGRP radioactive waste (radwaste)
has been estimated to be less than 600 ft'.
The two removed steam generators will be stored on site in a concrete bunker.
A burial option for the steam generators is still being actively explored by the licensee.
Excellent radwaste minimization initiatives have been incorporated into the project.
In summary, the inspector determined that very good planning and preparations were in progress to accomplish the SGRP safely and with relatively low personnel exposures.
No significant radiological controls issues were identified.
4.2 (Open) Inspector Follow Item - Investigation of Spent Fuel Pool Leakage (50-244/95-15-01)
As noted in NRC inspection reports 50-244/95-15, 50-244/95-17, and 50-244/95-20, the licensee had identified contaminated water leakage into the residual heat removal (RHR)
pump room.
The licensee speculated that this was due to leakage from the spent fuel pool (SFP),
based on sample analysis.
Approximately one cup of water per day was leaking into the room.
Onsite environmental well sampling has not confirmed any contamination from the SFP.
In addition to the gamma measurements, the NRC suggested that the licensee evaluate plant leakage water samples for tritium content.
During this inspection, this additional sampling data was reviewed.
Direct observation of the sampling locations, and the licensee's research of the local site area hydrology and geology was also performed.
Samples taken in mid-November 1995 indicated detectable tritium contamination in 1) the perimeter water collection system at the base of the containment building, 2)
on the north wall of the intermediate building subbasement (IBSB), and 3) in the RHR pump room on the southwest side of the plant (adjacent to the spent fuel pool).
The tritium concentration found in the RHR pump room is similar to the concentration found in the spent fuel pool (1,8E-l uCi/ml).
The pump room is located adjacent to the spent fuel pool.
The water
~samples collected from the containment perimeter water collection system indicated low levels of tritium contamination, i.e.,
Normal tritium background is approximately 1E-7 uCi/ml.
The ground water in-leakage is collected in two sumps and pumped into a retention tank for sampling prior to discharge.
A third area of water in-leakage was from
the north wall of the IBSB.
The seepage through the north wall indicated a
tritium concentration of 1.2E-4 uCi/ml.
During this inspection, the inspector estimated by visual observation that the total water in-'eakage to the IBSB was approximately five gallons per minute.
The inspector concluded that the relative concentrations and proximity to the spent fuel pool suggest that other potential sources of contamination should be investigated since the fuel pool is located adjacent to the south wall of the IBSB.
The licensee agreed to evaluate this.
The licensee has three on-site environmental monitoring wells located northeast, southeast, and southwest of the plant.
Several samples from each well have been taken and, as of the date of this inspection, no detectable contamination has been found based on both gamma and tritium analyses.
The inspector reviewed the licensee's preliminary action plan to address the leakage into the RHR pump room, and to identify the source(s)
of tritium.
Several actions are currently underway that include:
~
Continued cleaning of the RHR pump room walls and divert all'in-leakage to a leakage collection system.
~
Installation of a sensitive water level indicator in the spent fuel pool to quantify water.losses.
~
guantification of evaporative losses from the spent fuel pool, with consideration of total pool water makeup to establish the net water loss due to pool leakage.
~
Continued water sampling of the IBSB and environmental monitoring wells on at least a monthly basis.
~
Further testing of the turbine building drainage trench to determine if this is the source of in-leakage to the north wall of the IBSB.
II,
~
The licensee has enlisted the services of a local research scientist from the University of Rochester to further investigate the plant hydrology/
geology, and to investigate other possible sources of tritium contamination.
An off-site laboratory analysis will help facilitate this investigation.
The inspector determined that the licensee's present actions are appropriate, and that continued investigation into the source of contamination and mitigation of leakage are warranted.
No regulatory release limits or violations of regulatory requirements were identified.
Further review of the licensee's actions by an environmental and effluent specialist inspector is planned and will be documented in Inspection Report 50-244/96-0.0 SAFETY ASSESSNENT/EQUALITY VERIFICATION (IP 71707)
5. 1 Nuclear Safety Audit and Review Board (NSARB) Neeting On January 24-25, the NSARB held a scheduled quarterly meeting at the Ginna station to review recent issues related to plant safety, and to oversee ongoing efforts at the station to preparation for several major program changes planned before or during the upcoming refueling outage.
The board's normal agenda included 1)
a presentation of the recent good operating performance of the Ginna plant; 2) written reports of one LER concerning inservice testing not performed on a
PORV block valve, and one letter informing the NRC of the inappropriate removal of an Appendix R barrier near the refueling water storage tank; and 3) several presentations on the status of outage preparations, the SG replacement project, a proposed technical specification amendment addressing containment integrity during fuel movement, a notable increase in charging pump unavailability, results of the training accreditation self-evaluation report, the status of the plant's Conservative Operating Philosophy program, and results of recent gA/gC, PORC, and modification sub-committee reviews of plant activities.
The inspector found that board member discussions in all of the above areas were of sufficient depth to evaluate weaknesses and request additional actions within the line organizations to make changes or improvements to meet the board's expectations.
A significant amount of the NSARB meeting was also devoted to a detailed review of preparations in progress to assure the operational readiness of the plant organizations to start up from the next refueling outage.
The licensee is in the process of making several major programmatic changes that must be effectively managed and integrated to ensure a successful startup and operating run.
These new programs include I) the new Improved Technical Specifications to be implemented in February; 2) the new SGs to be installed during the next outage; 3) the reactor coolant system average temperature (T, ) reduction program resulting from the new SGs; 4) the new 18-month fuel cycle; and 5)
a change in the trim for the main feedwater regulating valves.
The board examined all of these programs and their associated activities in some depth in order to evaluate their progress and to ensure that they are being managed effectively,.
The board also reviewed several audits in progress on these programs and made recommendations to assure that audit results can be incorporated as necessary into these areas in a timely manner prior to startup.
The board agreed that the next meeting should be held before the quarterly session, and should be devoted solely to a review of startup readiness.
The board devoted a considerable effort to define new NSARB action items related to the management of these new programs.
However, due to their complexity and the very heavy agenda load in the prescribed meeting time, an agenda item to separately review all new or updated action items formally before the entire board was not accomplished.
The NSARB provides a
significant oversight function that assures full integration and operational readiness for the several major new program changes currently in development.
A review of all action items before the entire NSARB assures agreement and
full understanding among the members, and is their standard practice.
The inspector discussed this concern with several of the board members and the NSARB vice chairman given the complexity and significance of these programs, and the need for effective NSARB oversight.
All agreed that this item should have been accomplished.
As a result, all new and revised action items will be formally submitted to each of the board members for review, comment, and
~
incorporation into the meeting minutes prior to the next meeting.
The NSARB did perform a self-critique on this meeting prior adjourning.
The critique identified a desire to incorporate more industry-wide operating experience into their evaluations.
Overall, the inspector concluded that the discussion of safety issues among the NSARB members were objective and in-depth.
All issues were thoroughly evaluated and areas requiring followup actions were identified.
5.2 Periodic Reports Periodic reports submitted by the licensee pursuant to Technical Specification 6.9. 1 were reviewed.
The inspectors verified that the reports contained information required by the NRC, that test results and/or supporting information were consistent with design predictions and performance specifications, and that reported information was accurate.
The following reports were reviewed:
~
Honthly Operating Reports for November and December 1995 No unacceptable conditions were identified.
5.3 Licensee Event Report Licensee Event Report (LER)95-009,
"Surveillance Not Performed, Due to Improper Application of Technical Specification Requirements, Resulted in Violation of Technical Specification," regarding inservice testing of the PORV block valves, was submitted to the NRC on December 14, 1995, The inspector determined that the details were clearly reported and that the cause was properly identified.
While the corrective actions were appropriate,
~he inspector considered that additional corrective action may be required as a
result of the ongoing review of this issue.
LER 95-009 remains open pending resolution of open item 50-244/95-20-01.
6. 0 ADHINISTRATIVE 6. 1 Senior NRC Hanagement Site Visit During this inspection period, two senior NRC Region I managers visited Ginna Station.
On January 22-23, 1996, Hr. William F. Kane, Deputy Regional Administrator, and Hr. James T. Wiggins, Director, Division of Reactor Safety, toured the site and met with senior licensee management.
Topics of discussion included:
the steam generator replacement project; the improved technical specification project, including plans for implementation; the upcoming
month fuel cycle; the T,, reduction program; and preparations for the upcoming refueling outag.2 Exit meetings At periodic intervals and at the conclusion of the inspection, meetings were held with senior station management to discuss the scope and findings of inspections.
The exit meeting for the steam generator replacement radiation controls preparation inspection and investigation of spent fuel pool leakage (sections 4.1 and 4.2 of this report, conducted December 4-8, 1995)
was held by Hr. James Noggle on December 8,
1995.
The exit meeting for the steam generator replacement project engineering inspection-(section 3.2 of this report, conducted December 7-8, 1995)
was held by Hr. Goutam Bagchi on December 8,
1995.
The exit meeting for the current resident inspection report 50-244/95-21 was held on February 1,
1996.