IR 05000244/1996005
| ML17264A574 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 07/26/1996 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML17264A572 | List: |
| References | |
| 50-244-96-05, 50-244-96-5, NUDOCS 9608060038 | |
| Download: ML17264A574 (88) | |
Text
U. S. NUCLEAR REGULATORY COMMISSION
REGION I
Docket No:
License No:
50-244 DPR-18 Report No:
Licensee:
Facility:
Location:
50-244/96-05 Rochester Gas and Electric Corporation (RGRE)
R. E. Ginna Nuclear Power Plant 1503 Lake Road Ontario, New York 14519 Dates:
May 5 - June 15, 1996 Inspectors:
P. D. Drysdale, Senior Resident Inspector, Ginna E. C. Knutson, Resident Inspector, Ginna J. E. Carrasco, Reactor Engineer, Division of Reactor Safety (DRS),
Region I
J. C. Pulsipher, Reactor Systems Engineer, Office of Nuclear Reactor Regulation, (NRR)
J. M. D'Antonio, Operations Engineer, DRS, Region I
R. J. Lewis, Nuclear Engineer, Office of Nuclear Material Safety and Safeguards, (NMSS)
A. R. Johnson, Project Manager, NRR Approved by:
"L. T. Doerflein, Chief, Reactor Projects Branch
Division of Reactor Projects 9608060038 960726 PDR ADOCK 05000244
EXECUTIVE SUMMARY R. E. Ginna Nuclear Power Plant lns ection Re ort No. 50-244/96-05 This integrated inspection included aspects of licensee operations, engineering, maintenance, and plant support.
The report covers a four week period of resident inspection.
In addition, it includes the results of announced inspections by a regional reactor engineer and operations engineer, a NRR reactor systems engineer and a NMSS nuclear engineer.
~Oeratione:
At the beginning of the inspection period, the plant was shut down with the reactor defueled and the reactor coolant system drained for steam generator replacement.
The refueling was completed on May 23, 1996. After restoration of the containment access penetrations made for SG replacement, the licensee conducted a containment integrated leak rate test (ILRT) and a containment structural integrity test (SIT).
Reactor plant heatup was conducted on June 3, 1996; however, the plant was returned to Mode 5 the following day due to failure of a reactor coolant'system (RCS) resistance temperature detector (RTD). 'Following completion of this repair, the plant entered Mode 3 on June 7, 1996.
Reactor startup was commenced on June 8, 1996.
The plant entered Mode 2 early the following day, and criticality was achieved on June 9, 1996.
The main generator was paralleled to the grid on June 10, 1996.
Later that morning, operators manually tripped the main turbine from about 20 percent power due to a loss of main condenser vacuum while shifting the steam generator (SG) blowdown lineup.
The main generator was paralleled back onto the grid on June 10, 1996.
On June 11, 1996, the licensee commenced a plant shutdown to perform main turbine overspeed trip testing and to repair a leak from a high pressure turbine governor control valve.
The main generator was taken off line and the reactor trip breakers were opened on
'une 12, 1996.
During the subsequent reactor startup, a rod control system abnormality aborted the startup.
Later that day, the plant achieved criticality. The main generator was paralleled to the grid the following day. At the close of the inspection period, the plant was operating at approximately 99 percent power.
The licensee demonstrated good control during the plant startup activities.
Operator communications were precise, although better coordination of the blowdown system realignment may have precluded a loss of main condenser vacuum that resulted in a manual turbine trip. Shift supervisors demonstrated strong command and control, and responded promptly and conservatively to the off-normal situations.
The licensee's decision to shut down for a control valve repair was prudent, although it appeared closer examination of the valve prior to original reassembly should have identified the existing
Executive Summary (continued)
defect.
Insertion of a manual trip in response to the rod control system abnormality demonstrated the licensee's conservative operating philosophy.
Training activities conducted in preparation for startup with the replacement steam generators and attendant procedure changes was well conducted both in the simulator and classroom.
The nature of operating parameter changes, the reasons for them, and their effects were appropriately presented.
The one operating crew observed had no problem operating the simulator with these changes.
An appropriate review of plant procedures was conducted to determine needed changes due to the steam generator replacement and other modifications.
Although some of these changes were being made three days before the Mode in which they would be applicable, the changes were accomplished and the operating crews were informed.
Also, the inspector determined that the facility operations department had performed an effective review of procedures to identify changes required by the steam generator replacement and was in the process of making those changes.
A lack of refueling integrity existed for more than three hours during the movement of irradiated fuel in containment.
Fuel movement was immediately-stopped.
The immediate corrective actions to restore an open containment penetration were adequate to restore refueling integrity. Although the actual safety consequences of the incident are minimal, failure to properly implement the applicable procedures was a violation of 10 CFR 50, Appendix B, Criterion V. (VIO 50-244/96-05-01)
Operators made the auto start feature of the auxiliary feedwater (AFW) pumps inoperable with the reactor in Mode 2 at approximately 2% power.
The on-shift operations personnel had not been adequately briefed on the purpose of a procedure change.
Operator unfamiliarity with the Improved Technical Specifications (ITS) contributed to this event.
The shift supervisor acted promptly and decisively when he determined that a problem existed with the AFW system alignment.
It is not clear whether the motor drive auxiliary feedwater (MDAFW) pump automatic start feature is, or is not, assumed to be operable for any accident analysis or any analyzed condition while in Mode 2.
RGSE considers that the ITS requirement to maintain this function operable in Mode 2 is not necessary, represents a
potential hinderance to safe plant operation, and should be removed from their TS.
This'tem remains unresolved (URI 50-244/96-05-02)
No safety-significant deficiencies were noted during a detailed walkdown of safety related systems.
The material condition of the systems was good and the systems were properly'ligned.
All system component identification and labelling was consistent with controlled drawings.
The Nuclear'Safety Audit Review Board (NSARB) QA/QC Subcommittee discussions were thorough and indepth.
Presentations by the individual members and guests were very informative for senior management and NSARB members.
The Subcommittee Chairman and the members were able to assess the effectiveness of both QA/QC activities and activities of the plant line organizations over the previous quarte Executive Summary (continued)
Maintenance:
Observed maintenance activities were adequately controlled, personnel properly adhered to maintenance procedures, and the maintenance was accomplished in accordance with specified requirements.
Surveillance activities were performed in accordance with procedures and problems were dealt with conservatively.
.Omission of the B-RHR pump secondary shaft seal during maintenance received proper evaluation and disposition by engineering.
Operability of the B-RHR pump was not affected by the omission of this seal.
Management actions to reenforce lessons learned from this event to plant maintenance personnel were appropriate.
~En ineerin The containment dome was restored near the end of the refueling outage and the licensee ensured the structural integrity of the containment and the leak tightness of the containment liner.
Each sequential step in the structural restoration of containment and the leak tightness of the liner plate was implemented in an excellent manner by competent personnel.
RGSE and Bechtel engineering provided good oversight and resolution of issues related to the SGRP project activities.
RGSE maintained good control of welding, rebar splicing, and nondestructive examinations of the liner, and ensured the adequacy of the liner's restoration.
During the placement of new concrete, RGKE 5 its contractor displayed excellent corporate, project, and engineering management involvement.
The containment dome concrete was restored to its specified compressive strength, and successfully passed the integrated leak rate test (ILRT)
and the structural integrity test (SIT).
RGRE and contract engineers demonstrated experience, technical know-how, and safety responsibility prior to and during the ILRT and the SIT.
The licensee's search for, and response to, actual or suspected containment leaks during the ILRT was commendable.
The test director's performance was excellent and the contractor's supervisor demonstrated considerable knowledge and expertise.
The test was successful and all acceptance criteria were met.
The licensee needs to update the test procedure and the UFSAR in accordance with their recent license amendment regarding Appendix J.
The containment structural integrity test (SIT) was performed in an excellent manner and in conformance with regulatory requirements, the updated final safety analysis report (UFSAR), and the approved SIT procedure.
Control rod drag testing conducted by RGSE adequately demonstrated that the rods currently used in the reactor do not experience the high drag conditions reported in Bulletin 96-01.
The rod testing in the spent fuel pool was well controlled and the drag data was
Executive Summary (continued)
captured with accurate instrumentation.
The licensee's test methodology and data review were generally adequate, and the results were reported to Westinghouse and the NRC in a timely fashion.
Disposition of the service water pump fastener thread engagement analysis was adequate.
A difficultywith addressing the potential existence of a nonconforming condition could not be fully resolved because applicable design documents did not have any specifications, or because original design requirements were not available to RGSE Mechanical Engineering.
Lack of QA reviews could allow a nonconforming condition to go unidentified and unreviewed by the licensee.
Very good and effective engineering support of operations was noted by eliminating nearly all operator workarounds prior to completing the refueling outage.
The remaining workarounds do not impose a significant burden on operators.
RG&E's efforts for preparation of the original steam generators (OSGs) are consistent with the revisions to the NRC and DOT regulations.
The licensee was cognizant of the important regulatory revisions which will affect OSG shipments, and has identified the relevant technical issues.
'he licensee fully and effectively restored the proper vital area access controls for the containment building prior to conducting refueling operations.
All visible portions of plant systems inside containment were properly evaluated for integrity and the vital area was properly restored.
The licensee took prompt and adequate corrective actions after discovering an inattentive security officer. A check for unauthorized containment entries was immediately undertaken.
Adequate administrative controls were in place to avoid further incident J
SUMMARYOF SPENT FUEL POOL (SFP) RERACKING/STATION MODIFICATIONS/SYSTEMATICEVALUATIONPROGRAM (SEP)
The original Ginna Station SFP cooling system consists of a single permanently installed cooling loop containing one stainless steel horizontal centrifugal circulating pump, one shell-and-tube heat exchanger and associated piping, valves and instrumentation.
A small purification loop, and a loop-surface skimmer loop, were also provided.
The spent fuel assembly storage arrangement was modified in 1977.
The original spent fuel racks were replaced with higher density flux trap type racks.
The increased storage capacity from 210 to 595 fuel assemblies.
In 1979, a skid-mounted back-up system, designed and procured in the previous year, was installed.
It was comparable in capacity to the original SFP cooling system.
The designs of the pump, heat exchanger, were similar to the original. The skid-mounted back-up system used heavy-duty hoses to interconnect the components and to interface with existing plant systems.
Improvements were made to the portion of the Auxiliary Building ventilation system serving the SFP in 1981.
This modification installed a set of fire dampers in the ventilation system ductwork.
P Also in 1981, to accommodate future SFP storage needs, and to address NRC staff concerns regarding the Integrated Plant Safety Assessment Systematic Evaluation Program (SEP) Topic IX-1, "Fuel Storage" a new, larger, fully-qualified SFP cooling loop was proposed.
The proposed system, which was designed to ASME Code class and Seismic Category I criteria, was installed in 1988.
It consisted of another permanently installed cooling loop containing one stainless steel horizontal centrifugal circulating pump, one shell-and-tube type heat exchanger and associated piping, valves and instrumentation.
The new system was installed in parallel with the original SFP cooling system. This new SFP cooling system (loop 2), together with the combination of the original SFP cooling system (loop 1), and the skid -mounted SFP cooling system (loop 3), continues to provide the Ginna plant with two trains of 100% capacity SFP cooling.
In.-1 985, six of the flux trap type spent fuel racks were replaced with new. higher density fixed poison type spent fuel racks.
This further expanded the storage capacity from 595 to 1016 fuel assemblies.
In 1986, the fuel rods from 16 assemblies were consolidated into eight canisters (two additional canisters were filled with non-fuel-bearing hardware resulting from disassembly)
~
GINNA NUCLEAR POWER PLANT
B. SUMMARY OF CURRENT LICENSING BASIS (CLB)
REQUIREMENTS/LIMITS/COMMITMENTS RE:
SPENT FUEL POOL (SFP)'DECAY HEAT REMOVAL/REFUELING OFFLOAD PRACTICES
~ TS 3.9.11:
Water level at least 23 feet above top of stored irradiated fuel, TS 3.9.12:
Spent fuel pool boron concentration.
TS 3.9.13:
Storage Region II burnup/enrichment restrictions.
2.
The "normal" basis heat load for an'inventory of fuel assemblies from annual refueling discharges through 1998 and offload of one-third core in 1999 under refueling conditions is 7.9 x 106 BTU/hr (FSAR 9.1.3).
The corresponding temperature limitfor this scenario is 120 'F based on operator comfort.
Operating time, decay time, and other parameters important to heat load calculation are'not identified in the FSAR or other CLB documents.
3.
The "safety" basis heat load for an inventory of fuel assemblies from, annual refueling discharges through.1998 and offload of a full core in 1999 under refueling conditions is 16 x 106 BTU/hr (FSAR 9.1.3)
~ The corresponding temperature limitfor this scenario is 150 F based on providing alternate cooling before temperature exceeds 180 F
following a loss of the primary cooling loop (loop 1).
Full core offloads are described as occurring on 10 year intervals to support inservice inspection and "on other occasions when it is deemed necessary" (FSAR 9.1.3.1).
Again, operating time, decay time, and other parameters important to heat load calculation are not identified in the FSAR or other CLB documents, 4. The loop 2 (primary loop) spent fuel heat exchanger B is sized to remove the safety basis and normal basis heat loads using service water at 80'F with the service water temperature rise constrained to 20'F and 15 F respectively.
5. Adequate redundancy is provided to address a single failure.
Loop 2 (the primary loop)
can handle the safety basis heat load alone with a maximum. temperature of 150 F.
Loops 1 and 3 (the skid mounted loop) together can handle the safety basis heat load with a maximum temperature of 150'F.
Each of Loops 1 or 3 can alone handle the normal basis heat loads alone with a maximum temperature of 120 F. Normally, either Loop 1 or Loop 2 is operated alone to maintain the desired pool temperature.
Loop 3 is connected and operated only when high heat loads are expected.
6.
Pool structural analysis based on bulk pool temperature of 180 GINNA NUCLEAR POWER PLANT B.1 SUMMARY OF CURRENT LICENSING BASIS (CLB) DISCREPANCIES/RESOLUTIONS CLB COMPLIANCE REVIEW Discrepancy:
1. The NRC staff's assessment of spent fuel pool storage in 1995 determined that the licensee lacked administrative controls regarding verification that actual decay heat load is below the appropriate level for the type of offload.
Determine what corrective actions the licensee has taken prior to refueling.
Resolution:
Full-core offloads are not a regular practice.
The licensee has accomplished full core offloads during refueling outages in the spring of 1971, 1979, 1981, 1989, 1990, 1993, and 1994.
The licensee will perform a full-core offload during the spring 1996 refueling outage to accommodate steam generator replacement.
The licensee has offloaded a full-core 7 times in 26 years.
Not including the 1996 refueling outage the licensee will have offloaded a full core twice in five years.
All full-core offloads at Ginna were in support of major inspection and maintenance activities.
Therefore the "normal" basis heat load will continue to be 7.9 MBTU/hr and the "safety" basis heat load will continue to be 16 MBTU/hr.
Full-Core Offloads:
Year Ma'or Ins ection Maintenance Activit 1971 1979 1981 1989 1990 1993 1995 1996 Fuel cladding inspection 10-year inservice inspection (ISI)
Reactor coolant pump bowl inspection 10-year ISI Fuel inspection Service water system (SWS) valve refurbishment SWS valve refurbishment Steam Generator replacement The licensee has no formal procedures which require an annual SFP heat load analysis to be performed.
Beyond the 100-hr minimum in-vessel decay time required by the Ginna Technical Specifications, there are no formal procedural requirements which restrict the timing of fuel movement from the vessel to the SFP.
The SFP cooling system heat load capacities are given in Table C.1 for the full-core offload of spring 1996.
Table C.1 demonstrates adequate redundancy with all three cooling loop availability during a full-core offloa Discrepancy:
2.
Because the last 5 refuelings have involved full core offloads and a full core offload is planned this year to support steam generator replacement, determine whether the licensee plans to alter the description of "normal" basis heat load.
Resolution:
It is the staff's understanding that the licensee does not plan to alter the description of
"normal" basis heat load.
The licensee has accomplished full-core offloads during refueling outages in the spring of 1971, 1979, 1981, 1989, 1990, 1993, and 1994.
The licensee will perform a full-core offload during the spring 1996 refueling outage to accommodate steam generator replacement.
The licensee has offloaded a full-core 7 times in 26 years.
Not including the 1996 refueling outage the licensee will have offloaded a full-core twice in five years."
The licensee has established a "normal" basis heat load of 7.9 MBTU/hr, and a "safety" basis heat load of 16 MBTU/hr for the SFP.
The licensee has based the heat load on projected spent fuel assembly inventory from normal refueling operations through 1998 combined with one-third core discharge at the end of 1999 (date at which the SFP will be filled). By letter to the NRC, June 9, 1981, the licensee responded to the staff's request for information regarding modification of the SFP cooling system and the licensee's projected heat loads for full-core discharges.
These projections indicate that the "safety" basis heat load limit would be satisfied by progressively extending the irradiated fuel decay time in the reactor vessel prior to initiation of fuel movement from 8 days in 1981 to 14 days in 2009 (end of life). The progressively increasing decay time reduces the decay heat from the full-core discharge to accommodate the increasing decay heat load from earlier refueling outage discharges.
During a SFP storage assessment, performed by the NRC staff from June 19-23, 1995, the team, reviewed a spring 1994 fuel cycle-specific SFP decay heat analysis that demonstrated the decay heat load would be below the "safety" basis value of 16 MBTU/hr after a 10-day in-vessel delay.
All full-core offloads at Ginna were in support of major inspection and maintenance activities (see item 1 above).
Therefore the "normal" basis heat load will continue to be 7.9 MBTU/hr and the "safety". basis heat load will continue to be 16 MBTU/hr.
The SFP cooling system heat load capacities are given in Table C.1 for the full-core offload of spring 1996.
Table C.1 demonstrates adequate redundancy with all three cooling loop availability during a full-core offloa <s~
GINNA NUCLEAR POWER PLANT C. SUMMARYOF COMPLIANCE WITH CURRENT LICENSING BASIS (CLB)
REQUIREMENTS A sustained loss of spent fuel pool (SFP) cooling or a significant loss of SFP coolant inventory is a remote event at Ginna due to certain design features and procedural controls, listed below:
(a) The availability of multiple loops for the SFP cooling.
(b) The extended period available to recover from a loss of SFP cooling to the onset of bulk boiling conditions in the SFP when the rea'ctor is at full power.
(c) The procedural controls that govern the SFP cooling and support systems when a full core offload to the SFP is accomplished.
(d) The anti-siphon protection provides for flow-paths capable of draining SFP coolant levels below the top of stored fuel.
(2)
Administrative requirements control the water level at least 23 feet above the top of the stored irradiated fuel by Section 5.2.4 of Operating Practices Procedure No. 0-15.1, "Administrative Requirements for Reactor Head Lift, Core Component Movement and Periodic Status Checks," Revision 4 (February 24, 1996).
(3)
Administrative requirements control the SFP boron concentration as outlined in Section 5.1 of Operating Practices Procedure No. 0-15.1, Rev 4. The SFP boron concentration is also controlled by the Core Operating Limits Report (COLR) located in the plant Technical Requirements Manual.
(4)
Administrative requirements control the SFP storage Region II burnup/enrichment restrictions outlined in Sections 3.9, 3.10, and 3.11 of Refueling Procedure No. RF-8.4,"Fuel and Core Component Movement in the Spent Fuel Pit," Revision 44 (February 22, 1996).
(5)
Table C.1 below summarizes SFP cooling system heat load and capacities for the current 1996 refueling outage during a full-cor'e offload.
(6)
Possible SFP leakage reported at Ginna Station:
'I The licensee has identified contaminated water leakage into the residual heat removal (RHR) pump room potentially due to leakage from the SFP- (NRC Region I
Inspection Report No.s 50-244/95-15, 50-244/95-17, 50-244/95-20 and Draft 50-244/96-01).
The RHR pump room is located on the southwest side of the plant adjacent to the SFP.
The licensee, based on sample analysis, estimates possibly one cup of water per day.
The licensee is pursuing onsite environmental well sampling to confirm any contamination from the SFP.
The licensee's action plan to address the potential leakage into the RHR pump room include:
(a) Continued cleaning of the RHR pump room walls and divert all in-leakage to a leakage collection system.
(b) Installation of a sensitive water level indicator in the SFP to quantify water losses, (c) Quantification of evaporative losses from the SFP with consideration of total pool makeup to establish the net water loss due to SFP leakage.
(d) Continued water sampling of the intermediate building subbasement and environmental monitoring wells on at least a monthly basis.
(e) The licensee has employed a consultant to further investigate the plant hydrology/geology (includes other groundwater leakage in other areas).
GINNA NUCLEAR POWER PLANT SUMMARYOF SFP COOLING HEAT LOAD AND CAPACITIES 1996 REFUELING OUTAGE Date 4/9 4/1 0 4/11 4/23 5/1 5/4 - 5/6 Heat Load (Mbtu/hr)
16.3 16.0 1 5.46 11.96 10.64 (10.64 Lake Temp Assume d
(deg F)
50
50
60
"B" Sys Capacity (Mbtu/hr)
21.0 21.0 21.0 Not available 1 9.0 1 9.0
"A" Sys Capacity (Mbtu/hr)
13.5 13.5 Not available 13.5 1 2.0 12.0 Modified Skid Capacity (Mbtu/hr)
17.1 17.1 17.1 17.1 1 5.0 1 5.0 Available Backup Capacity (Mbtu/hr)
30.6 17.1 17.1 17.1 27.0 12/15 Comments Earliest offload Lampson rotated MCC C Outage Bus 16 Outage Lake temp.
assump.
change to 60 deg MOV static testing 5/1 5 (1064 '0
>8
>16 Lake Temp Change to 80 deg
GINNA NUCLEAR POWER PLANT
, CLB COIVIPLIANCEREVIEW D. SINGLE FAILURE CRITERION EVALUATION RE:
NUREG 0800, "STANDARDREVIEW PLAN FOR THE REVIEW OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER PLANTS (SRP)."
'
The Ginna Nuclear Power Plant CLB does not require conformance to SRP 9.1.3.
However, the Ginna Station SFP cooling and cleanup systems compare with most of the SRP areas.
The latest upgrades to the SFP cooling system (Loop 2, Pump B),
although not active failure proof, is designed to seismic Category I and 'Appendix B requirements.
The original non-seismic SFP cooling system (Loop 1, Pump A) remains in place as a fully functional backup to Loop 2, Pump B.
The licensee has no plans to backfit SFP cooling system requirements into the CLB.