IR 05000219/1985036
| ML20137K926 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 01/03/1986 |
| From: | Anderson C, Bissett P, Chung J, Gregg H, Murphy K NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20137K904 | List: |
| References | |
| 50-219-85-36, NUDOCS 8601240273 | |
| Download: ML20137K926 (33) | |
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U.S. NUCLEAR REGULATORY COMMISSION l
REGION I
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Report No.
50-219/85-36 l
Docket No.
50-219 i
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j License No. DPR-16 Priority Category C
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Licensee: GPU Nuclear Corporation Oyster Creek Nuclear Generating Station P. O. Box 388 l
Forked River, NJ 08731
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l Facility Name:
Oyster Creek Nuclear Generating Station
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Inspection At:
Forked River, N. J.
)i Inspection Conducted: November 7 - 22, 1985 I
Inspectors:
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i cam Leader -
eac r Engineer
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K. G" u p.
,i nicpAssistant
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. H /lii ett, Reactor Engineer
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H. Gregg, LeadV acto Engineer date
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Approved by:
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C. And on, hl W Plant 5ystems Section date
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Ins Inspection on November 7-8/18_-22, 1985 (Report No, i
50 pection Summary:
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219/8590 l
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Areas inspected:
Special announced inspection of eautpment and activities important to prevent or mitigate potential overpressurization event of low l
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pressure ECCS systems.
Specifically, the core spray system was inspected.
The
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inspection included 225 inspector hours on-site, 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> off-site and 26 hours3.009259e-4 days <br />0.00722 hours <br />4.298942e-5 weeks <br />9.893e-6 months <br /> l
l at the NRC regional office by four region-based inspectors.
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i Results: No violations or deviations were identified.
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DETAILS 1.0 Persons Contacted D. Barnes, Plant I&C Engineer
- K. Barnes, Licensing Engineer K. Bass, Mechanical Engineering Supervisor J. Barton, Deputy Director, Oyster Creek J. Boyle, Operating Supervisor R. Brown, Operations Control Manager P. Cervenka, Supervisor, Operating Experience Assessment F. Cimino, M&CF Planner P. Crosby, Plant Engineering Supervisor R. Fenti, Site Manager-Nuclear Assurance
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Fiedler, Vice President & Director R. Fittz, M&CF Production Director
- V. Foglia, Preventive Maintenance Manager
- J. Frew, M&CF Production Director T. Gaffney, Material Manager-Electrical /I&C R. Harkleroad, Plant Engineering
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- D. Holland, Licensing Manager T. Jenkins, M&CF Area Supervisor L. Leitman, Plant Engineering J. Litrakis, Spare Parts Engineer
- R. Markowski, Manager, QA Program Development & Audit J. Maloney, Manager, Plant Materials J. Molnar, Core Engineer D. Pino, Electrical Material Engineer B. Pitman, Special Functions Engineer
- W. Popow, Director, Maintenance & Construction
- T. Quintenz, Manager, Plant Engineering A. Rone, Manager, Operations Engineering C. Schilling, Mechanical Maintenance Manager
- G. Simmonetti Quality Manager
- W. Smith, Director, Plant Engineering
- *J. Sullivan, Director, Plant Operations R. Tilton, Engineering Assurance Manager-Technical Functions V. Willet, M&CF Planning Manager R. E. Weltman, Material Manager-Mechanical
- F, Weinzimmer, Manager, Special Projects USNRC R
- J. Durr, Branch Chief, Division of Reactor Safety Region I
- W. Bateman, Senior Resident Inspector
- J. Wechselberger, Resident Inspector
- A, Blough, Section Chief, Division of Reactor Projects, Region I
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A The inspectors also held discussions with other licensee employees during the inspection, including operations, technical support and administra-tive personnel.
- Denotes those present at the exit meeting on November 22, 1985.
2.0 Scope of Inspection 2.1 Objective L
The inspection objective was to assure that the low pressure portions of the systems and components interfacing with the high pressure reactor coolant system wouid not be subjected to over-
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pressurization inadvertently and that operator actions would mitigate the consequence should an overpressurization event occur.
The potential overpressurization of the low pressure side has been referred as " Event V" in the Probabilistic Risk Analysis (PRA) study and could lead to a '.0CA.
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2.2 Scope
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This inspecti4n was organized into two distinct phases:
1) an initial phase'wes conducted by the resident inspector, as documented
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in inspection report 50-219/85-15, and was largely fact finding in nature, designed to determine if obviout deficiencies existed and to i
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identify high-low pressure interfacing systems and components; and 2) a follow-on inspection extended the inspection scope in more detail based on the findings of the initial phase of the inspection
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and the lessons learned from previous failure experiences.
l This report documents the findings of the follow-on inspection, which focused on the availability of the interfacing systems and components, and preventive human actions / factors to prevent or mitigate the potential overpressurization event,
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l 2.3 Conduct of Intpection Potential contributors to equipment unt.vailability, the system's l
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"As Bui't".cnfiguration, potential hardware and procedural weak-nesses, and buman actions which could lead to or to cope with the poten+.fil ove'rpressurization event were evaluated.
The inspection i
also ircluded programmatic aspects of the administrative controls, inp).enentation of site QA/QC activities, and housekeeping in general.
L 2.3.1 2AS_ BUILT"Configurationof_IsolationInterface
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Walk-down, inspection tours and P&ID reviews were conducted to i
evaluate and verify the high-low pressure interfaces of the
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feedwater system, isolation condenser, shutdown cooling system,
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and core spray system.
Each system was evaluated for their "AS BUILT" and "AS FOUND" conditions against the plant P& ids and design criteria, and the potential for system overpressuriza-tion.
For the hardware associated with the core spray system, the following aspects relating to human factors, equipment accessibility, and testability were evaluated.
2.3.2 Equipment Availability The availability and operability of the selected equipment for the core spray system were evaluated based on the adequacy of the plant maintenance and surveillance programs and their implementation.
The inspection focused on the station activities and initiatives to prevent, detect, correct and re-cover from equipment failures, deficiencies and inoperable con-ditions.
Supporting records were reviewed to assure that the preventive measures, corrective maintenance and periodic surveillance were performed effectively in accordance with the prescribed procedures, and that recurring failures or generic problems were identified and resolved in a timely manner.
Specific features inspected included:
Surveillance programs, procedures, records, test
witnessing, and technical adequacy.
Local Leak Rate Testing (LLRT), accessibility, local
equipment operations, indicators, alarms, and snubbers.
Component cooling, electrical support, logics,
ventilation, and environmental qualification.
Calibration of indicators and transmitters, and witnessing
calibration activities.
QA/QC participation.
- 2.3.3 Plant Operations and Emergency Responses The readiness and effectiveness of plant operations were evaluated based on the ability of the plant staff to respond to and recover from a potential overpressurization event.
This was accomplished by the following specific plant activities:
Demonstration of equipment and system operability: remote,
automatic, manual and local operations.
Station operations:
operator ability to utilize the con-
trol room and local panel indications and parameters to determine the plant abnormal and emergency conditions, and to respond to a potential overpressurization event. Opera-tor interviews and system walk-throughs were conducted.
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Station operating procedures:
normal, abnormal, alarm
response, and emergency (symptom-oriented) procedures and their utilization. Procedural reviews, including technical content, identification of equipment and their status, and clarity of instructional steps were also conducted.
Operator understanding and human factors engineering:
equipment identification, detection of inoperable equipment, and communications.
2.3.4 Previous Core Spray System Events Previous events associated with the Oyster Creek core spray system were reviewed, including inadvertent actuation of the
core spray system or inoperable conditions which could have resulted in a ov'erpressurization event.
3.0 Summary of Findings j
The inspection findings demonstrated that the plant programs designed to assure hardware availability and their implementations were adequate and the plant staff exhibited an excellent knowledge of plant operations, equipment, and procedures.
These were indicative of a high degree of con-fidence to prevent or to recover from a potential overpressurization event of the core spray system.
The operator responses were indicative of the effectiveness of training j
on procedures and equipment, and the high degree of equipment availability i
and operational adequacy were indicative of the effective management and administrative controls within the scope of this inspection.
The inspection team identified four unresolved items and made one obser-
vation.
Prompt remedial actions on these findings could improve the availability of the equipment and effectiveness of the plant operations.
Human Factors Alarm Response Procedures
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A' safety concern was identified;on the manual corrective actions in response to a core spray system overpressurization alarm for both systems I and II. Alarm response procedures, B-6-e for system I and B-6-f for system II, require that pressure be relieved to the torus through motor-operated valve (MOV), V-20-26 or V-20-27, depending upon which system the overpressurization occurred.
However, opening
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of the valve, V-20-26 or V-20-27,' would require the operation of a key-lock which is located in the same vicinity as the hub drains for
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relief valves of the core spray system. Operation of the key-locks i
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for the valves would therefore require sending an operator to an area that could possibly be filled with steam (details in paragraph 4.2.1).
Equipment Surveillances Testable Check Valve Test
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Surveillance procedure 610.4.011 requires a periodic leak testing of core spray system testable check valves in accordance with the re-quirements in Technical Specification 4.3(G) and the intent of the NRC's Order of April 20, 1981, which was incorporated into the spec-ifications. However, the test frequency of the leakage test appears to be inconsistent with the intent of the Order which requires the test upon startup after valve is exercised or otherwise moved. Also, the purpose of the leak testing is not simply a leak test but to assure that the testable check is properly closed and seated. This was not clearly indicated in the procedure 610.4.011 (details in paragraph 6.3.2).
Preventive Measures
Previous failure experiences in the industry indicated that the root
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causes of such failures were often traced to inadequate procedural steps and subsequent failures by individuals performing the surveil-lance activities. This includes precautions necessary to protect the equipment and personnel, confirmation and verification of the actions taken during activities, and utilization of proper parameters and indications. Three examples of procedural weaknesses were identi-fled, all related to core spray system surveillance testing.
Procedure 610.3.006 could be improved by adding precautions and veri-fication of valve positions during and after the test and utilization of control room core spray system overpressure alarm.
Verification instructions of closure and adequate seating of core spray testable check valve was not clearly stated in the surveillance procedure, 610.4.008.
Procedure 610.4.011 could be improved for parallel M0V's for leakage status by utilizing test gauge pressure readings before and after manipulation of valve V-1 (details in paragraphs 6.2.1, 6.3.1 and 6.4 respectively).
Post-Maintenance Testing - Documentation
Upon completion of a maintenance activity, it is required that both an operational test of a component and a system functional test, if applicable, be performed. However, a system functional test is often delayed because of additional work still in progress on the same or related system (s). Also, each work activity is controlled by and documented through various mechanisms, including M&C Short Form, Sur-veillance Test Procedures, Job Orders, Special Procedures,
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Surveillance Review forms, Switching and Tagging Log Sheets.
Assembling and reviewing of completed maintenance activities were major tasks.
The inspector could not verify completed post-mainte-nance testing associated with completed maintenance activities for
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several core spray system valves because of the complexity of the
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documentation process.
The licensee volunteered to assemble completed maintenance packages covered under M&C Short Form 06500, 07800 and 07801 but was unable to provide all of the necessary
documentation prior to the exit meeting (details in paragraph 7.1.7).
Observation
Air-operated testable check valves are tested for valve leakage
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periodically as required by Technical Specification 4.3(G). However, test records indicated that test leakage was often recorded as zero leakage and the results were not trendable due to a lack of accuracy.
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According to FSAR Table 6.3-4 this 8 inch check valve will have a normal seat leakage, not to exceed 2 cc/hr per inch of seat diameter at a differential pressure greater than 62 psi (details in paragraphs 6.3.3 and 7.1.1,(2)).
Verification methods for the check valve position indications were not clear.
Clarification of the procedural steps would be desirable (paragraph 6.4).
4.0 C,re Spray System Operations
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4.1 System Description The Core Spray System (CSS) is one of three separate systems that is part of the Emergency Core Cooling System and thus provides for the removal of decay heat from the core following a postulated Loss-of-Coolant Accident. The CSS is made up of two completely indepen-dent Ioops with each loop consisting of two main pumps, two booster pumps, two sets of parallel isolation valves inside and outside the
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drywell, a spray sparger and associated piping, instrumentation and
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controls (See Figure 1).
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Each loop has a test recirculation line to the torus, provided with
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motor operated test valves for full flow testing without discharge into the reactor vessel.
Flow and pressure instrumentation are provided in the control room for each loop.
The piping up to the test valve is carbon steel, designed for 400 psig and 350 F.
From the test valve into the reactor, the piping is fabricated of stain-less steel designed for 1250 psig and 575 F.
The low pressure piping of each CSS loop is equipped with a relief valve for overpressuriza-tion protection (See Figures 1 and 2).
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The water supply for the CSS is normally from the torus with the condensate storage tank and the fire protection system providing a backup supply of cooling water.
The CSS can be started manually or by automatic trip signals gen-erated when a low-low reactor water level and/or a high drywell pressure condition is detected.
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FIGURE 1 DIAGRAM 0F OYSTER CREEK CORE SPRAY SYSTEM V-20-153 I
NZO2D V-20-24 v'.
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w V-20-41 V-20-18 V-20-51 I
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NZ029 V-20-21 V-20-54 V-20-151 V-20-22 V-20-9 Spec.
O V-20-23 LO.
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V-20-55 C
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V-20-26 Drywell e To TORUS gy FR0u TORUS Csmu i
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V-20-152 V-20-27 I
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v-20-17 LO.
V-20-40 V-20-50
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V-20-16 V-20-8 a
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NZ02A V-20-15 V-20-12 V-20-53 v-20-150 C
spec.
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V-20-25 V-20-52 A
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- l 4.2 Conduct of Core Spray System Operations The inspector performed " walk-throughs" and " hands-on" simulation on portions of the core spray system operations (normal, abnormal and emergency operations) with a SRO, experienced with the Navy and at Oyster Creek, and with a RO with eight years of experience both at Oyster Creek and fossil plants.
The inspector questioned the oper-ators about the operations of the core spray system, control room front panel indicators, equipment operations and valve lineup re-quirements.
The operators then described coincidence logic and set-points associated with the core spray system actuation and manual operations. The inspector noted that the reactor water level set-points were posted by the control rod status board on the control room front panel.
They were also asked to discuss control room in-dications and verification of the reactor trip, and alternate means of reactor trips, including manual both inside and outside of the control room. They walked through specific steps necessary to close normally open (breaker racked out) MOVs during the overpressurization event of the low pressure side of the core spray system. There is one normally open MOV, with its circuit breaker racked out, for each
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train upstream of normally closed parallel MOVs, and they are high-I low pressure interface valves (see Figures 1 and 2).
The inspector was informed by the operators and concurred that the racking-in and control room closure of the MOVs would take less than a few minutes.
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The inspector asked the operators about the diagnostic and restora-tion procedures of abnormal and emergency event operations and di-agnostic control room indications, including reactor power, pressure, temperature, and water level. The operators explained and walked through various event symptoms and indications by pointing out alarms, recorders, instruments, and parameters, both at front and back panels in the control room. They also demonstrated the use of the color-coded, symptom-oriented Emergency Operating Procedures and various entry conditions to either Emergency Operating Procedures (EOP) or Abnormal Operating Event Procedures (A0EP).
Operator responses to a core spray system overpressurization alarm were discussed, including passage of the steam through a relief valve (one relief valve per train at 350 psig relief setpoint).
Specific
operations of valves and response times were discussed.
4.2.1 Findings During the simulation of operator action in response to a core spray system overpressurization alarm (300 psig), a safety con-cern was identified in respect to the operators manual correc-tive action, as denoted in the alarm response procedure 2000-RAP-3024.01.
Both RAPS, B-6-e (System I) and B-6-f (System II)
require that pressure be relieved to the torus through V-20-26
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or V-20-27, depending upon which system the overpressurization
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occurred.
However, to open V-20-26 or V-20-27 would require the operation of a key lock, which is located in the same vicinity as the hub drains for the core spray overpressure relief valves (350 psig).
Operation of the key locks for these valves would therefore re-quire sending an operator to an area that could possibly be filled with steam if system pressure had reached 350 psig.
Subsequent action, if the pressure builds up again, requires that test isolation valves V-20-18 and/or V-20-12 be closed.
The inspector noted that it might be more appropriate to initially close V-20-18 and/or V-20-12 since the system piping between these valves and the reactor pressure vessel is rated above normal operating pressure.
The licensee recognized the
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possible safety concern and agreed to further evaluate the sit-uation to determine a more appropriate corrective action,
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i Further evaluation and any appropriate revisions are to be
completed by April, 1986. This is an unresolved item pending licensee action and resolution on this item and subsequent l
NRC:RI inspection (50-219/85-36-01).
5.0 Core Spray System Instrumentation 5.1 Surveillance / Calibration The Core Spray system design includes instrumentation to provide indication, alarm, and low pressure or differential pressure permissive interlocks to prevent the Core Spray parallel injection valves from opening until reaching a preset pressure for system over-pressurization protection.
The licensee performs applicable calibra-tion and functional tests to meet or exceed Technical Specification requirements.
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The inspectors reviewed various surveillance procedures associated
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with the Core Spray system to verify that instrumentation was calibrated and tested as required by Technical Specifications.
Procedures reviewed are listed in the Appendix.
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The inspector's review included verifying logic sequences, as applicable, and independent verification of switches and valves after system restoration. Also reviewed were the appropriate history cards associated with the core spray system valves, pumps, power supply breakers, pressure switches, etc.
Those history cards reviewed were found to be maintained and up-to-date, t
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The inspectors also observed, in part, the performance of Procedure No. 610.3.105, " Core Spray System I Instrument Channel Calibration and Test," and 619.3.013 " Reactor Low Level Test and Calibration,"
to verify the following:
An approved, up-to-date procedure was used.
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The procedure was adequately detailed te asure sati,sfactory
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performance.
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Operational personnel were notified prior to and upon completion of test.
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Properly specified parts and materials were identified for the test.
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Calibrated test equipment was used.
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Proper restoration of barriers and covers was accomplished.
l It was noted that these tests involved the initial testing of Static-
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0-Ring (SOR) pressure switches which recently replaced Yarway level indicating switches. The SOR pressure switches are non-indicating
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level switches and for this reason, the licensee was required by the i
NRC to develop a method which would demonstrate direct communication
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between the reactor pressure vessel and the SOR level switches. The SOR level switches of concern provide Rx Lo Level, Rx Lo-Lo Level and/or Drywell High Pressure signals.
6.0 Assessment of Core Spray System Isolation The prime objective was to evaluate the adequacy of protection against overpressurization events in light of actual overpressurization events that have occurred in BWR's as expressed in IE Information Notice No.
84-74 and related documents. To assess the concerns, procedures and associated test records were reviewed to determine the adequacy of the written instructions, test frequencies, and test conduct.
6.1 Interface Configuration Oyster Creek has two independent core spray trains. The components that comprise the interface between high design pressure (1250 psig)
and the low design pressure core spray piping (400 psig) are shown in Figure 2.
Two such configurations are present, one for core spray System I and one for System II.
The locked open maintenance valve and parallel air operated check valves are located in the drywell while the balance of the components are in the reactor building. The check valves have position indicators in the control room as well as a control to test the open function of the valves via air actuation. The parallel, normally closed MOV's, as well as the
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normally open MOV, have position indicators and hand controls in the control room.
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FIGURE 2
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DIAGRAM OF OYSTER CREEK CORE SPRAY ISOLATION VALVE CONFIGURATION (orie of two identical configurations shown)
Pressure Alarm
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Parallel
>300 PSIG In In Control Room Control Room Normally Closed MOVs Locked Open Parallel p
p Maintenance Spray Air Operated Valve Sparger Check valves Ax a^
Normally Open MOV With Locked Out Breaker
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Local Change in Pressure Pressure / Temp. Spec.
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Cauge 1250 PSIG @ 575 F to:
00 SG @ M0 F Calibrated k '
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55 GAL Drum l F7
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F 7 Temporary
Test Configuration s For Check Valve Leak Testing
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A single hand control actuates both parallel MOV's. The control room has OPEN and AUTO positions.
In the AUTO position the valves will receive an open signal upon core spray actuation having the following logic:
One out of four low-low reactor water level indicators
O_R One out of four high containment pressure indications.
- The control room operator can normally send an open signal to the valves by transferring the hand operator to OPEN. Whether manual or
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automatic open signals are received, the valves will remain shut until enabled by a pressure interlock that prevents opening of the valves when reactor pressure is greater than 300 psig. The parallel valves have two reactor pressure sensors and the enabling condition is that at least one of the two sensors activate as the pressure falls below 300 psig. The normally open MOV has its power removed by
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opening and locking its power breaker. This valve receives the same open signal and pressure enabling signal as the parallel MOV's upon
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i actuation of core spray. The control room has CS system pressure indication (booster pump discharge pressure gauge) as well as a sys-tem overpressurization alarm with a set point of 300 psig. Also
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shown on Figure 2 is a temporary test configuration composed of
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small diameter piping for check valve leakage measurements.
6.2 Core Spray Isolation Valve Actuation Test and Calibration i
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The purpose of this procedure (610.3.006, Revision 14) is to test the actuation of the parallel MOV's by simulation of low reactor pressure at the respective pressure sensors.
6.2.1 Test Procedure Review
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Starting with System I, the normally open MOV is closed and power removed (refer to Figure 2). A core spray actuation signal is simulated and each of the two pressure sensors are tested in turn by simulating low reactor pressure. The as-found trip pressures are recorded as are the parallel valve opening times. The parallel valves are then closed (closure time is also recorded). The test is then repeated for the second pressure sensor, after which, the valves of System I are returned to normal.
The entire procedure is repeated for System II.
Several improvements were suggested to the detailed procedures involving verification of valve positions and assurance of vaive seating, as follows:
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(1) Revisions 13 and 14 of the procedure added verification check-offs to assure independent checks by a second individual
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at critical steps in the evolution.
However, manipulation of the normally open MOV did not require this independent verification.
As the closure of this valve during the test is a vital pre-
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caution against system overpressurization, the inspector judged that independent verification was important.
In addition, the assurance of the reopening of the valve at the end of the test is as important since it assures functioning of core spray and
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then should also be independently verified.
(2) The Precautions section of the procedure does not mention the possibility of overpressurization if errors in following the procedure are made.
Such a precaution would sensitize the operators to the concern and further assure their careful atten-tion to the procedure.
(3) Overpressurization experience indicates that MOV closure lights in the control room periodically fail to indicate complete valve seating. The inspector reviewed the procedure with this concern in mind and determined that two additional steps in the procedure would provide independent assurance of MOV seating, as well as a gross indication of valve leak tight-ness.
First, a requirement to record the time between the open-ing signal to the normally closed parallel valves and the over-pressure alarm indication in the control room should be added.
It was determined by operating staff interviews that the alarm does sound at this point in the procedure if the reactor is at pressure.
Sounding of the alarm is caused by the small amount of leakage past the parallel air operated check valves.
The sounding of the alarm is indicative of a properly closed and seated discharge MOV (the normally open MOV in Figure 2).
If this valve was open or leaking badly, the alarm would not sound at this point.
The second step is the confirmation that the overpressurization alarm clears after the parallel MOV's are closed and the discharge valve opened. This will assure that
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the operator becomes immediately aware of closure or leakage problems involving the parallel MOV's.
The above enhancements in Procedure 610.3.006 were discussed with utility personnel, and they were in agreement with the valve of upgrading the procedure as discussed in paragraph 6.5.1.
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6.2.2 Records Review Table 6.1 summarizes the test dates for Procedure 610.3.006.
The detailed check off sheets and data sheets were reviewed.
No inconsistencies in testing frequency or documentation were noted. However, as shown in the table the majority of these tests have been performed while the reactor was at pressure; thus the need for precautions against overpressurizatio.
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AB _ E 6
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COR E S3 RAY SO E ON VA_VE AC"UA'~ 0 s ~ES As ) CA_ B RA" O s (Oyster Creek Procedure 610.3.006)
Test Procedure Reactor Press.
Date Revision
>350 PSIG 1/15/85
yes 2/16/85
no 3/19/85
yes 4/15/85
yes 5/13/85
yes 6/06/85
yes 7/12/85
yes 8/14/84
yes 9/85 nir
10/85 nir
11/14/85
no
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nir= not as yet in records file
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.
6.3 Core Spray System Check Valve Leakage and In-Service Test The purpose of this procedure (610.4.011, Revision 4) is to measure the leakage of the parallel air operated check valves. The test is
to satisfy the April 20, 1981, " Order for Modification of License Concerning Primary Coolant System Pressure Isolation Valves" as well as IST requirements.
6.3.1 Test Procedure Review The test is performed during reactor pressurization prior to exceed-ing 600 psig and with a pressure differential across the valve disk of greater than 150 psid. A temporary test connection is made as shown in Figure 2.
A pressure gauge, valve, and a calibrated 55 gallon drum are installed. Upon opening the test connection, water is drained into the drum.
Based on a measured increase in drum level within a one minute period, the gpm leakage is calculated and re-corded.
-
,
As a result of reviewing the written procedure and discussion with
,
the utility staff, the following changes in the procedure were agreed
upon:
(1) The _" Purpose" section of the test procedure should convey that the underlying purpose of leakage measurements as described in the 1981 Order, is "to verify that each valve is seated properly and functioning as a pressure isolation device".
(2) As there is no requirement to leak test the parallel, normally closed MOV's, a step should be added that provides assurance that the M0V's are indeed seated and are not grossly leaking. After the leakage measurement of the check valve is made and acceptable leakage
,
is calculated, the valve outboard of the local pressure gauge should i
be closed and the gauge read once a constant pressure is attained.
!
Any significant difference between reactor pressure and the local pressure gauge reading would be indicative of MOV leakage.
The above changes will appear in the next procedure revision.
The utility has committed to issuing the revision by April 1, 1986.
6.3.2 Test Frequency
'
The records of leakage tests were reviewed.
Table 6-2 sum-
'
marizes the leakage tests reviewed.
In addition, a review of
,
events that moved the check valves in comparison with the t
dates of leak testing was made. Table 6-3 summarizes this data.
This tables shows that the check valve discs had been moved twice within the recent past without subsequent leakage testing.
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CORE SPRAY SYSTEM CHECK VALVE LEAKAGE AND IN-SERVICE TEST (Oyster Creek Procedure 610.4.011)
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Test Proced ure Reactor Gauge Level Leck Comrnents i
,
i Date Revision Press u re Pressure Change Rate j
(psig)
(psig)
(inches)
(gprn)
l 5/28/81
-
-
-
-
Maint. on l
502 O
O O
Sys. 11 i
)
4/05/82
502 O
O O
502 O
O O
.
j 8/23/84
540
2 3.4
500
1 1.7 i
i
8/23/84
530
0.3 0.5 Retest of
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-
-
-
-
Sys.I f
,
j 6/18/85
518
0.016 0.03
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l 525
O O
11/17/85
-
-
-
-
Sys, ll only;
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+75 O
O O
to assure
- l NZO2D seated j
Note: System I dato on top I
q Systern ll data on bottom j
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.
TABLE 6-3 RELATIONSHIP BETWEEN CHECK VALVE MOVEMENT AND LEAK TESTING (Oyster Creek Nuclear Plant)
Reactor Check Valve Check Valve Cornments Shutdown Movement LeakogeTest Periods **
Dates Dates 2/12/83-11/1/84 3/17/84-air operator repfoced 4/05/84-valve maint.
8/23/84-during hydro 10/29/84-inadvertent
CS injection 11/04/84-11/09/84 11/10/84-11/22/84 11/30/84-12/03/84 12/03/84-12/04/84 2/02/85-2/14/85 2/10/85-volve
operability 2/15/85-2/17/85 2/17/85-2/23/85 2/24/85-2/25/85 2/25/85-3/04/85 6/12/85-6/18/85 6/18/85 7/08/85-7/09/85 7/22/85-8/03/85 8/09/85-8/10/85 10/19/85-11/16/85 11/17/85-Sys.ll only
- Valve movement without subsequent leak testing
- Bosed on Monthly Reports File f 20.0008.0010.0020.03
.
.
Though the current Technical Specification requirements (4.3-G)
do not state the need for leak testing after valve movement, it was judged to be a prudent measure. Air operated check valves have failed to seat because of debris being lodged on their seats after opening due to flow conditions.
Valves are also known to stick open upon actuation of the air piston from a variety of causes.
In addition, Section 2.2.1 of the Franklin Research Center report, TER-C5257-252 (Rev. 1), attached to.the 1981 Order recommends testing after disc movement.
This test frequency concern was discussed with the utility staff. During a meeting between the utilities engineering and licensing staff, the licensee concluded that the concern would be best handled by issuing a request for a Technical Specifica-tion change that adds to Section 4.3(G), the requirement to per-form leak testing whenever the check valves are moved, whether manually actuated or due to flow conditions.
6.3.3 Test Accuracy for Trending The following is made as an observation concerning the accuracy of leak testing of the testable check valves.
Based on the test record review and discussions with engineering staff members, the inspector raised questions concerning test accuracy.
Specifically, the leak measurements (refer to Table 6-2) appear suspect with five measure-ments of zero leakage; since, for metal to metal seated valves of this type, zero leakage is not expected. Also, during the testing of the parallel MOVs, with the reactor at pressure, the overpressure
- protection alarm routinely actuates.
This indicates that the check valves do leak.
Thus the leak rates of the testable check valves of the leak data would appear suspect.
The licersee agreed to look into the leak measurement method of these valves.
6.4 Core Spray Testable Check Valve Operability Test The purpose of this procedure (610.4.008, Revision 2) is to verify operability (the opening function) of the testable check valves and operability of their remote position indicators.
A review of this procedure revealed several statements of possible misinformation that deal with the concern of overpressurizatio.
._..
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.
-
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,
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.
First, the procedure does not clarify that the valve disc closure indication, using the magnetic reed switch, is not accurate enough to
assure valve seating and effective pressure isolation.
The assurance of seating can only be effectively checked by valve leakage measure-ments.
The second concern involves Sections 6.2.7 and 6.3.7, which states
" verify each valve shuts by observing pivot arm rotation". This
- .
statement could be misleading since the pivot arm rotation to the closed position does not necessarily assure disc closure as the disc must drop by gravity to the closed position and, in fact, can be stuck open for a variety of causes.
Direct visual observation of the valves, thus cannot assure closure.
Notation on the test record of 610.4.011, conducted November 17, 1985 indicates that direct obser-vation could have convinced certain individuals that check valve NZO2D was closed (the control room position indicator of this valve
!
was not working). However, as noted on Table 6-2, a leak test was
-
also performed.
However the above could mislead the operator and a procedural change and training sessions with the engineering staff might be desirable.
This observation was discussed with a licensee representative, and the licensee stated that the inspector's obser-vation would be incorporated in the procedure, if applicable.
6.5 Summary of Findings The following will be carried as unresolved items:
l 6.5.1 Inconsistency on Check Valve Leak Test Frequency (Unresolved Item No. 50-219/85-36-03) - As discussed in
,
Section 6.3.2 a utility generated Technical Specification change will be submitted to the NRC adding the requirement to perform leak testing whenever the check valves are moved. This submittal will be received before April 1, 1986.
.'
6.5.2 Test Procedures to be Revised (Unresolved Item No. 50-219/
85-36-04) - Test procedures 610.3.006, 610.4.008, 610.4.011 i
will be upgraded as discussed in Sections 6.2.1, 6.3.1 and 6.4, and issued before April 1, 1986.
,
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7.0 Core Spray System Valves, Piping and Pumps
!
The inspectors evaluation of components in the Core Spray (CS) system was for the most part, directed to the valves in the system. There was also some evaluation of the piping and pumps in this system.
,
,
During the initial phase of the inspection, the CS System Piping and In-
strument Drawings (P&ID) were reviewed and a walk-down of accessible portions of the system performed.
In subsequent phases of the inspection,
,
a study and evaluation was made of valve operational information, manu-
facturer's data, design characteristics, maintenance, operability sur-
veillances, In-Service Tests (IST), component specifications, LLRT re-
!
quirements, and the licensee's performance of a valve operability test.
Discussions were held with operational, engineering and craft personnel
-
throughout the inspection.
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7.1 Valves The inspection focused on all valves in the through path from the CS pumps to the reactor vessel (see Figures 1 and 2).
The following subparagraphs describe the details of the valve inspection areas.
7.1.1 Valves - Informational Data Based on the inspectors review of the licensee's drawings, the plant walk-down and appraisal of the components to be in-cluded in the inspection a system outline sketch (Figure 2) was
,
prepared. The valve data, including Manufacturer, valve size, type, and pressure class, as determined from observations of the valve and from the licensee's record information was reviewed and is described in the following subparagraphs (1) through (6).
,
The inspector determined the isolation valves were manufactured prior to ASME Section III and were made to ASA B31.1.
Based on observations of the valves and review of the manufacturers drawings and information, the inspector concluded the valve design characteristics, pressure class, and sturdy construction would provide for acceptable operation over long time periods.
The valve data and several valve items noteworthy of comment
,
are:
(1) The Atwood & Morrill testable swing check valves V-20-150 (152) and V-20-151(153) are 8"-680# St. Stl. and have 5" side mounted air cylinders. They were purchased with two means of valve position indication:
1) a magnetic switch arrangement which is directly connected to the disc arm and provides disc position indication and 2) an external limit
switch arrangement that is connected to the actuator arm to indicate air piston actuator position.
The licensee util-izes only the direct disc position indication and the in-spector concluded this provides the more positive indica-tion of valve position.
(2) The Atwood & Morrill testable swing check valves were purchased and supplied to a rigid 2 cc/hr/in seat leakage requirement.
The inspector concluded that the licensee's broader seat leakage procedural requirement for an installed valve is appropriate since it does not represent a brand new valve
,
undergoing factory test. This was further verified by the
'
!
Office of NRR imposed TS seat leakage requirements of 1 gpm max (and 5 gpm max with additional proviso's).
!
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1 (3) Valves V-20-12(18), V-20-15(21), V-20-40(41), are Anchor
,
'
Company, 8"-600# St. Stl., one piece disc, gate valves with Limitorque Operators. The valves have a clamp ring body to bonnet connection and have pressure seal bonnets.
(4) Valves V-20-17(23) are Anchor Company, 8"-600#, St. Stl.
Gate Valves with pressure seal bonnet and with 4:1 bevel gear handwheel.
(5) The relief valves, V-20-25(24) are J. E. Lonergan, Model
W303, 2 x H x 3 flanged steel, 300# inlet, 150# outlet, enclosed spring, 2 ring valves set at 350 psig. These valves are the only flanged valves in the system (all others have butt weld ends).
(6) The check valves in the lower pressure piping from the CS
pumps to the low pressure side of V-20-12(18) are steel 300# general variety of swing check and were furnished by the Crane Company and the Anchor Company.
7.1.2 Operability Surveillance and IST Test Requirements
).
The inspector reviewed the master surveilitnce schedule and the TS requirements and verified that the motor operated isolation i
valves and the testable check isolation valv's have surveillance i
requirements specified.
From the IST Component Procedure list, i
the inspector determined all the valves reviewed within the
scope of this inspection also have IST requirements.
.
The inspector reviewed several valve test procedure data sheets of tests recently performed and verified that the tests were performed within the time frame required by the TS and IST.
This effort involved test procedures 610.4.003, 1610.4.008 and j
610.4.011.
The inspector noted that the IST program was clearly defined and was being implemented, and that these programs are the initi-
,
ators of problem identification and corrective repair action.
!
7.1.3 Witnessing the Core Spray Valve Operability and IST The inspector witnessed the Core Spray Valve Operability test of CS system I performed in accordance to licensee's procedure 610.4.003 at 6:00 PM on November 21, 1985.
This test verifies operability of the CS motor operated valves in accordance to TS requirements. The test
,'
also satisfies the operability requirement of the IST program for these CS motor and air operated valves.
!
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. _ - - -, - __- _
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- - -.- -_- - --.
.
.
The inspector observed the Control Room Operator perform the test and verified that each of the procedural steps were followed.
During the test, the inspector witnessed valve opening and closing time, stem position indications and the directions given regarding breakers to be closed or locked open.
Based on the step by step verification of the test and the recording of data, the inspector's observations verified that the test was performed correctly.
7.1.4 Preventive ' Corrective Maintenance The maintenance history for valves and related components was reviewed and discussed with licensee representatives. This review included a verification that proper administrative controls were in place to effectively control the conduct of any planned and/or corrective maintenance. Administrative controls reviewed are included in the Appendix.
Core spray system maintenance activities during the last two years were reviewed by the inspectors.
Documents reviewed included maintenance history cards, MC&F computerized history list, Short Forms, M&C Work Request and Job Order forms, main-tenance and surveillance data sheets, switching and tagging log sheets, surveillance review forms, and control room log sheets. The review of these activities was performed to verify the following.
--
GSS approval was obtained prior to initiation of work activities.
--
Approved procedures were used.
'
Procedures and appropriate data sheets were properly
--
completed.
--
Verification of redundant system operability when required.
--
Post maintenance testing was performed as applicable.
QC involvement was commensurate with the maintenance
--
performed.
Appropriate reviews were completed.
--
Selected Short Forms and associated documents were reviewed by the inspectors to verify that the above attributes had been accomplished.
Examples of Short Forms, Job Orders and related
-- -
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i documentation involving completed CS system maintenance activ-ities are listed below. The majority of maintenance involved valve packing replacement and/or adjustment, limitorque re-greasing, and circuit breaker replacement.
SF 03619 SF 07800
!-
-
-
SF 04953
- SF 07801
-
'
- SF 11888
- SF 06500 SF 00968
- SF 18522
-
SF 18522
- SF 21392
-
JO 4381E
- JO 51657
-
,
i During the inspectors review of the above completed maintenance activities, it was difficult to verify the execution of appli-cable post maintenance testing requirements.
For limitorque
'
valve operator maintenance performed under Short Form 07800,
07801, and 06500 the inspector was unable to verify performance
of a system functional test following the completed maintenance activity.
The licensee was unable to provide, prior to the exit, the necessary documentation which verified that system
!
functional tests had been completed.
The inspector also reviewed the maintenance history records for the testable isolation check valves V-20-150/152/151/153, isola-tion gate valves V-20-15/40/21/41, and the relief valves i
V-20-25/24. The records show relatively few problems.
From the repair records of the above isolation valves there was only one
,
repair which required disassembly of the valve bonnet / cover from the body.
t Most of the repairs for the testable check valves relate to the
,
!
actuator and most of the repairs to the gate valves are packing leaks or motor operator problems.
!
7.1.5 Licensee's use of NPRDS The inspector reviewed the licensee's activities related to
input of valve problems and their use of the Nuclear Plant
]
Reliability Data System (NPRDS). The inspector reviewed several of the valve problems described on the history records and verified that an NPRDS input was made.
The maintenance person-
,
nel also stated they frequently review the NPRDS input from
'
other facilities.
7.1.6 LLRT CS Isolation Valve Requirements The inspector reviewed the licensee's LLRT program relating to primary testable check valves inside the Drywell and the next motor operated valve outside the drywell.
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t The inspector verified that the safety evaluation identifies the CS systems as a system required to function throughout a post
'
accident period and that these valves are not required to be LLRT tested.
7.1.7 Findings i
t
'
The inspector's overview of the CS valves is:
they are accept-ably designed and of sturdy construction and have given good i
performance since start of operation; the extensive operability and IST program requirements provide continued means of monitor-ing operability and identifying problems and initiating cor-rective action; and the maintenance on these components has been
,
acceptable. One area of concern related to post maintenance
j testing was identified by the inspector. This matter is dis-cussed in more detail in the following paragraph.
Following the completion of any maintenance activity, the li-
censee performs both an operational test of the component and a system functional test, if applicable. Many times, an opera-i l
tional test will be performed, however a system functional test
.
I will be delayed because of additional work still in progress on
.
the same or related system (s). Work activity is controlled and
'
documented through various control mechanisms, including M&C
.
Short Forms, Surveillance Test Procedures, Job Orders, Special
!
Procedures, Surveillance Review Forms, Switching and Tagging Log Sheets, etc.
During the inspector's review of various completed
maintenance activities, it was difficult to verify the comple-i tion of any post maintenance testing associated with completed maintenance activities for several core spray system valves.
'
Specific examples of this problem included those maintenance activities covered under M&C Short Form 06500, 07800 and 07801.
The licensee was requested to perform whatever actions were necessary in order to provide added assurance that appropriate
!
post maintenance testing had been completed on the above and any other applicable maintenance activities.The licensee agreed to
complete the necessary actions by February,1986. This is an
unresolved item pending licensee's action and resolution on this t
j item and subsequent NRC:RI intpection (50-219/85-36-05).
'
I 7.2 Piping and Pumps
1 7.2.1 Piping
i The inspector reviewed the CS piping and instrumer.t drawing (GE-885781 P22) and the Burns and Roe piping specification i
i (S-2299-60) to determine where the piping specification design pressure temperature changes occur and to evaluate the condi-i
!
tions.
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The inspector determined from the piping specification that the design conditions from the torus to the suction side of the CS pumps is 300 psig/350F and the piping is carbon steel, from the CS pump to the inlet of valve V-20-12/18 is 300 psig/350F and is carbon steel, and from this valve to the reactor vessel is 1250 psig/575F and is stainless steel.
It was also noted that both the suction and discharge carbon steel piping was specified to the same pipe schedule, 7.2.2 Pumps The inspector observed several of the core spray pumps, reviewed the manufacturers installation and operation instructions, the pump list data and the pump curves for the CS pumps and the CS booster pumps.
The inspector determined that the pumps were:
CS pumps NZ-01-A/B/C/D - Ingersoll Rand, 18 x 21AL, 405'
head, 3700 gpm CS booster pumps NZ-03-A/B/C/D - Ingersoll Rand, 10 x 17A,
250' head, 3750 gpm.
The inspector's observations concerning the piping and pumps were:
They are acceptable for the design conditions; have had numerous tests within the operability and test program re-quirements; and have performed acceptably.
8.0 "AS BUILT" Configuration of Isolation Interface The feedwater system, isolation condenser, and shutdown cooling system are all connected to the reactor vessel primary coolant system and would be vulnerable to the full reactor coolant system pressure under adverse con-ditions or events.
Particular concerns were those check and isolation valves interfacing between high and low pressure piping systems.
In order to verify such vulnerability of the valve configurations and piping systems at the high-low pressure interfaces, system P& ids and other necessary documents were reviewed as listed in the Appendix, and " walk-down" system inspection was conducted for the interface components and piping, including tanks, heat exchanges, valves and pumps.
Based on the above reviews and inspections, it was concluded that those vulnerable systems, components, and piping associated with the feedwater system, isolation condenser and shutdown cooling system were all construct-ed at a design pressure of 1250 psig, and would not be subjected to the potential Event "V" and subsequent LOC O
,
9.0 Facility Tours The inspector observed Control Room operations for shift turnover, log sheets, and facility operations in accordance with the administrative control procedures and Technical Specifications.
Several inspection tours and system "walkthrough" observations were conducted based on the pre-scoped inspection plan, which included equipment tagging, lock-out, housekeeping in general, local panels and indications, posting, radio-logical controls, and plant operations.
The areas toured included:
Air operated core spray system check valves inside Drywell
Reactor Building in general
Feedwater pumps and condensate pumps
CR0 accumulators
Isolation Condenser return valves, piping, and tanks
Core Spray Systems I & II MOVs and relief valves
Core Spray System main and booster pumps
Switchgear room and MCC associated with Core Spray Systems
Shutdown Cooling system and RBCCW heat exchangers
Reactor water level instrumentation
9.1 Findings Housekeeping conditions, generally were excellent and equipment identifications were clearly labeled and posted.
Radiological control zones were clearly identified with proper postings and locked-out high radiation areas were posted in accordance with the station radiological control procedures.
During a routine walk-doqn" inspection tour, the following valve packing leaks were identified and the licensee was subsequently notified:
CRD accumulator valves including 26-23 and 22-43 CRDs
Core Spray System II valves: V-20-18, V-20-21, V-20-41
Shutdown cooling system MOV V-17-206
However, the leakages were relatively small and the inspector was informed that once the systems heated up, the valve packings would tighten up due to expansion. All leakages were collected into waste cans by funneling through plastic wraps, which is indica-tive of good housekeeping.
The licensee representative informed the inspector that the valve leaks, identified by the inspectors, would be reinspected and proper actions would be taken.
The inspector did not have any safety concern regarding the leak '
\\
,
,
I At 3:45 pm on November 21, 1985, the inspector observed a portion of worh 'n progress for Rosemont pressure transmitter for Drywell
-
pressure (post-TMI action it.em). A Patel seal and new 0-ring were installed, reclacing the olt ones.
The inspector also noted that a QC inspector was monitoring t5g, activities and a plant engineer was supervising the maintenance work.
10.0 Review of Events Theirh,icctorreviewedtwoeventsassociatedwiththecorespraysystem, both cttributable to" personnel errors. An operational event, 81-09, in September, 1981 occurred during a performance of the core spray system instrument surveillance.
The operator did'not adhere to the procedural steps and mistakenly opened the core spray system isolation valves, leading to the icadvertent injection of chromated torus water into the reactor vessel.
The licensee subsequently evaluated the cause of the
,
event and lessoas learned, and took appropriate corrective actions. The personnel errors due to poor communication and failure to adhere to the written procedure were the ccntributing factors.
On Monday, October 29, 1984, during a, calibration of reactor water level instrumentation for the core spray sy; tem II, core spray system I was inadvertently initiated and injected torus water into the reactor vessel
<
.
for approximately 20 seconds (LER No.84-025).
The event was again attributable to personnel errors, by performing the calibration section of the procedure out of' sequence.
Post-event analyses indicated that in both incidences, the safety sig-nificance was minical and no long term effects or equipment damage re-salted.
The licensee performed a detailed assessment of the two events and took appropriate long and shortsterm corrective actions.
Based on the above, the inspectcr concluded that the potential for overpressurization
'of the low pressure side of the core' spray system due to similar events is minimal.
,
11.0 Unresc bed Items Un esolved items are matters about which mor,e information is required to determine.if it is a violation, a deviation or acceptable. Unresolved items are discussed in paragraphs 4.2.1, 6'3.3, 6.4, 6.5.1, and 6.5.2.
'
12.0 Exit Meeting The inspectors met with the licensee representatives denoted in paragraph 1 o.n Nove aber 22, 1985, and summarized the purpose, scope and findings of the inspection.
The attendees are listed in paragraph 1 of the report
'
details.
No writter niaceriai was provided to the licensee by the inspectors.
'
>
_ _ _ _ _ _ _ _ _
,
t
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APPENDIX DOCUMENTS REVIEWED A.
DRAWINGS Reactor Shutdown Cooling System, GE 148F711
RCS Instrumentation, GE 148F712
Core Spray System, GE 885D781, NU 5060E6003
Cleanup System, GE 148F444
Feedwater System, BR2003
Isolation Condenser, GE 148F262
Atwood & Morrill DWG No. 20605-H, Core Spray Valve
IDC DWG 112C2808, Reactor Level Instrumentation
B.
Administrative Control Procedures and Other Documents GPUN Response Letter to NRC on Generic Letter 83-28, May 8, 1985
Procedure 105, Conduct of Maintenance, Revision 20
Procedure 106, Conouct of Operations, Revision 34
Plant Manual, Plant Material Department:
Procedure 118, Revision 8,
Preventive Maintenance Administrative Procedure; Plant Material Electrical /I&C Trending Report, 1985; Preventive Maintenance Check
~
Sheets; Station Information Management System (SIMS).
Oyster Creek FSAR and Technical Specifications
LER No.84-025; Operational Event Report 81-09
Procedure No. 732.2.004, 480V MCC Preventive Maintenance
Procedure No. 108, Equipment Control
Procedure No. 112.1, Calibration of Technical Specification
Installed Instrumentation Procedure No. 116, Surveillance Test Program
.
L
,
C.
Implementation Procedures Procedure No. 305 Shutdown Cooling System Operation
Procedure No. 610.3.105 Core Spray System I Instrument
Channel Calibration and Test Core Procedure No. 610.3.205 Spray System II Instrument
Channel Calibration and Test.
Procedure No. 619.3.013 Reactor Low Level Test and
Calibration Procedure No. 610.4.002 Core Spray Pump Operability Test
Procedurc No. 700.2.011 Preventive Maintenance of Motors
Procedure No. 610.3.001 Core Spray Pump Failure Pressure
Switch Surveillance and Calibration.
Proceaure Nr.. 610.3.004 Core Spray Header dp Test and
Calibration Procedure No. 610.3.001 Core Spray Pump Failure Pressure
Switch Surveillance and Calibration Procedure No. 610.3.006 Core Spray Isolation Valve Actua-
tion Test and Calibration.
Procedure No. 610.3.105 Core Spray System 1 Instrument
Channel Calibration and Test Procedure No. 610.3.205 Core Spray System 2 Instrument
Channel Calibration and Test Procedure No. 610.4.002 Core Spray Pump Operability Test
Procedure No. 610.4.003 Core Spray Motor Operated Valve
Operability Test i