IR 05000334/1979016
| ML19250C286 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 09/14/1979 |
| From: | Beckman D, Keimig R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML19250C268 | List: |
| References | |
| 50-334-79-16, NUDOCS 7911230150 | |
| Download: ML19250C286 (25) | |
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U.S. NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND E.' FORCEMENT Region I Report No. 50-334/79-16 Docket No. 50-334 License No.
DPR-66 Priority Category C
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Licensee:
Duquesne Light Company 435 6th Avenue Pittsburgh, Pennsylvania 15219 Facility Name:
Beaver Valley Power Station Unit 1 Inspection at:
Shippingport, Pennsylvania Inspecticn conducted: July 2-6 and 10-13, 1979 Inspectors:
b4& W[I
D. A. Beckman, Reactor Inspector date signed
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date signed d
e signed Approved by
A R'R.KeibiighChief,ReactorProjects
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dat'e sicfned Section No. 1, RO&NS Branch Insoection Summary:
Inspection on July 2-6, and 10-13, 1979 (Report No. 50-334/79-16)
Areas Inspected:
Routine, unannounced inspection, beginning on a backshift, of licensee action on previous inspection findings, followup on licensee event reports, review of licensee periodic reports, action in response to IE Bulletins and Circulars, review of plant operations, followup on loss of RHR flow events, review of administrative controls, and review of a 10 CFR 21 report. The inspection involved 78 hours9.027778e-4 days <br />0.0217 hours <br />1.289683e-4 weeks <br />2.9679e-5 months <br /> onsite by one NRC regional based inspector.
Results:
Of the eight areas inspected, no items of noncompliance were identified in seven areas. One item of noncompliance was identified in one area (Deficiency - failure to maintain Jumper and Lifted Lead Log, paragraph 3).
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DETAILS 1.
Persons Contacted Mr. J. Carey, Technical Assistant - Nuclear Mr. D. Crouch, Shift Supervisor
- Mr. W. Glidden, QA Engineer Mr. K. Grada, Shift Supervisor Mr. M. Haddad, Engineer Mr. S. Lacey, Shift Supervisor
- Mr. F. Lipchick, Station QA
- Mr. L. Schad, Operations Supervisor Mr. J. Turner, Shift Supervisor
- Mr. J. Werling, Superintendent
- Mr. H. Williams, Chief Engineer Mr. R. Woodling, Senior Engineer The inspectors also interviewed other licensee representatives during the inspection, including members of the operations, maintenance, guard, stores, and general office staff.
- denotes those present at the exit interview.
2.
Licensee Action on Previous Insoection Findings
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(0 pen) Unresolved Item (76-26-03):
Licensee to determine the number of allowable safety injection cycles for Unit 1 design.
The licensee has issued Engineering Memorandum (EM) No. 20165, dated June 7, 1979 which presents the DLC Engineering Department position on the need for performing the fatigue analyses which would be required to determine the number of allowable injection cycles.
The EM states that all analyses required for components in accordance with commitments to ASME Boiler and Pressure Vessel Code, Section 3 and those required for piping in accordance with commitments to Power Piping Code B31.1 have been completed in conformance with the applicable FSAR requirements.
The memorandum recommends that the subject fatigue analysis not be performed.
This matter will remain unresolved pending further review by NRC of the licensee's position.
(Closed) (Unresolved Item 78-26-07): Licensee to evaluate procedures for control of Safety Injection System (SIS) reset feature.
The licensee has reviewed the Emergency Operating Procedures of the BVPS Operating Manual, Chapter 1.53.4, which provide the specific operator actions required to manually restart /reinitiate the required Engineered Safety Features (ESF)
if a loss of offsite power or an accident occur after the SIS has been 13'6 209
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reset and before the ESF equipment is returned to a lineup for automatic initiation.
Revisions have been made to the Emergency Operating Procedures for loss of coolant accidents and loss of offsite power which instruct the operator to manually reinitiate SIS if the above conditions are experienced.
Inspector review of these procedures, ESF logic and elementary diagrams for the Safety Injection System, Main Steam Line Isolation, Feedwater Isolation, Containment Isolation Phases A and B, and Emergency Diesel Generator start and sequencing confirmed the following:
If an accident occurs after the reset of SIS, a reinitiation can be
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effected using either one of the two main control board Manual SIS Actuation pushbuttons.
Review of individual valve and pump logic diagrams indicates that all components will automatically resequence into a safety injection alignment on such a manual actuation signal.
Each component control switch is equipped with a spring-return-to-neu-tral feature which ensures that individual control switch positions cannot override reinitiation.
If a loss of offsite power occurs after reset of the SIS, undervoltage
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on the Emergency Busses will reinitiate a fast start of the diesel engines, even during an engine cooldown, and will resequence generator breaker closure and sequential loading.
The licensee has added precautionary notes to each applicable emergency operating procedure instructing the operator to manually reinitiate SIS, if required, and has provided procedural notes describing the response of emergency power systems to the loss of offsite power as discussed above.
These have been incorporated into Revision 16 to the BVPS Operating Manual Chapter 1.53.4, Emergency Ope. rating Procedures.
Tb2 inspector's review utilized controlled copies of the logic and elementary diagrams contained in the BVPS Operating Manual.
(0 pen) Unresolved Item (79-04-03):
Maintenance of proper environmental conditions in the Auxiliary Feedwater Pump Room.
Extremely high humidity conditions had been previously experienced in the room due to the Steam Driven Auxiliary Feedwater pump steam isolation valve seat leakage.
This leakage resulted in steam being released through the pump turbine's exhaust duct.
Since the time of IE Inspection No. 334/79-04, the licensee has improved the ventilation lineup for the area, installed temporary sealing material in the pump turbine exhaust duct, and has performed repairs on the pump turbine steam valve.
The inspector has toured the area during the current shutdown and confirmed the licensee's action in regard to the above.
The inspector also reviewed a Nuclear Services Quality Control General Inpsection Repc/t, dated March 5, 1979, which details the work performed on the leaking valve.
This item will remain open pending inspection of the area by NRC:RI at power to confirm the adequacy of the actions described above.
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(0 pen) Unresolved Item (79-04-02):
Licensee to review long term entries in Jumper and Lifted Leads Log and process as design changes in accordance with 10 CFR 50.59 and the BVPS Quality Assurance Program.
The licensee has identified those jumpers which are to be consideced as permanent plant modifications and has provided a listing of those items for review and initiation of design change packages by the DLC Engineering Department.
T%a licensee stated that the subject listing will be reviewed by the Jnsite Safety Committee pursuant to 10 CFR 50.59 prior to plant startup.
The inspector stated that this review should include the purpose of the jumper or lifted lead and a technical evaluation of the actual circuit configuration to ensure that the installation accomplishes its intended function with no other detrimental effects to the safety of operation.
The licensee acknowledged the inspector's comments.
This matter will remain unresolved pending completion of the Onsite Safety Committee review and reinspection by NRC:RI.
(0 pen) Unresolved Item (79-04-05):
Licensee to review and augment the program for surveillance and control of penetrations (other than containment penetrations).
The licensee, in conjunction with implementation of the recently issued facility fire protection Technical Specifications, is contir.uing to review and implement controls for penetrations which act as fire barriers or pressure boundaries within the NSSS structures.
Revision 7 has been issued to Operating Manual Section 1.48.6.1.9 and provides additional controls for construction craft work via use of Area Clearance Permits.
These permits are issued directly by the Operations Supervisor for individual work items.
Review of the BVPS Maintenance Manual, Chapter 1, Section T.1, for control of penetrations and discussion with licensee Maintenance Department supervision indicate that additional measures have been identified as necessary to ensure control of penetrations during routine maintenance.
The licensee's progress in this regard appears satisfactory and will continue to be monitored by NRC:RI during future inspections.
(Closed) Unresolved Item (79-09-01):
Licensee to revise Clearance (tagging)
Procedures of Operating Manual Chapter 1.48 to specify that a licensed Senior Operator will review sys+em alignment changes resulting from clearances which affect safety re nted systems.
The licensee has imple-mented Revision 7 to Operating Manual Chapter 1.48, Section 1.48.6.b.3, which requires a Nuclear Shift Supervisor or Nuclear Shift Operating Foreman to specifically review and authorize the switching or valving required during tagout or restoration of safety related equipment.
Additionally, the licensee issued Special Operating Order No. 79-2 on April 25, 1979, which ensures that key operations personnel and watch-standers are specifically informed of system alignment changes made via clearance procedures.
The inspector observed implementation of the above provisions several times during this inspection and found it to be satis-factory.
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(Closed) Unresolved Item (79-09-08):
Licensee to incorporate outstanding Operating Manual Change Notices (OMCN's) into procedures via formal revisions.
The licensee has eliminated the backlog of OMCN's which had been outstanding for more than 90 days.
The inspector reviewed the official OMCN log and determined that all OMCN's except those processed within the past 90 days had been incorporated into the applicable pro-cedures or cancelled, as appropriate.
The OMCN log was reviewed for the period January 1978 through July 1, 1979.
(Closed) Unresolved Item (79-09-06) Licensee to clarify acceptance criteria values in Operating Surveillance Test 1.33.3, Fire Protection System Drain Test.
The licensee issued OMCN 79-49 on May 1, 1979, which provides clear guidance for acceptance of static and flow conditions for fire system line pressures.
The OMCN was subsequently incorporated into Revision 15 of Operating Manual Chapter 1.55. A on May 22, 1979.
(Closed) Unresolved Item (79-13-02):
Licensee to revise alarm response instruction for Steam Generator Feed Pump Suction Pressure Low Alarm.
The licensee issued OMCN 79-61 on July 12, 1979, revising the subject procedure to provide more specific guidance for the operators regarding rates of plant load reduction to avoid steam pump operation.
Additional guidance has also been included which requires the operator to manually trip the reactor if the load reduction results in excessive transient response such as excessively high steam flows.
Provisions are also included for manually securing the steam dumps should a steam dump mal-function be observed.
(0 pen) Unresolved Item (78-30-01) Errors in SI System Pipe Support analysis.
This matter is the subject of the NRC Show Cause Order dated March 13, 1979.
The inspector informed the licensee that a supplemental report for LER 78-53 should be submitted prior to restart to present the current status of licensee corrective actions.
This item will continue to be unresolved pending final disposition of the matter by NRC and the licensee.
3.
Review of Plant Operations a.
Shift Logs and Operating Records The following logs and records were reviewed for the periods indicated.
Gl-1, S1-1 to Sl-9, LI-l to L1-5, and L5-13 and -14 for the
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period of June 8 to July 3, 1979.
Temporary Operating Procedures 79-2 through 79-25 covering the
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period from January 15, 1979 to June 29, 1979.
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Jumper and Bypass log entries for the period from January 1,
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1979-July 1, 1979.
Clearance log for the period June 1, 1979 to July 10, 1979.
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The logs and records were reviewed to verify that:
log sheet entries are filled out and initialed;
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log entries involving abnormal conditons are sufficiently
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detailed; log book reviews are being conducted by the plant staff;
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operating orders and temporary procedures do not conflict with
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the Technical Specifications; and, jumper log entries do not conflict with the Technical Speci-
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fications.
Acceptance criteria for the above review included inspector judgement, the requirements of applicable Technical Specifications, and the following procedures:
BVPS OM Chapter 48, Conduct of Operations;
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OM 1.48.3, Section H, Temporary Procedures;
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OM 1.48.5, Section D, Jumpers and Lifted Leads;
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OM 1.48.6, Clearance Procedures;
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OM 1.48.8, Records; and, OM 1.48.9, Rules of Practice.
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b.
Plant Tour Inspection tours of the following plant areas were conducted at various times during this inspection:
Auxiliary Building; Control Room; Switchgear Rooms; Safeguard and Auxiliary Feedwater Pump Areas; and Cable Tunnels.
The following determinations / observations were made.
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Control room and local monitoring instrumentation were reviewed
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to verify that instrumentation and systems required to support Mode 5 operation were in conformance with Technical Specifications LC0 requirements.
This review included the radiation monitoring instrumentation, nuclear instrumentation, RWST, BAST, emergency boration system lineups, diesel generator lineup, offsite power system lineup, and residual heat removal system lineup.
Radiation controls established by the licensee, including
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posting of radiation areas, the condition of step-off pads, and the disposal of protective clothing were observed.
Several radiation work permits used for entry into radiation areas were reviewed.
Plant housekeeping conditions including general cleanliness
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conditons and storage of materials to prevent fire hazards were observed.
Equipment caution and danger tags were inspected for proper
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post.ing and logging.
Systems and equipment in all areas toured were observed for the
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existence of fluid leaks and abnormal piping vibrations.
Mechanical snubbers and hangers installed on the Quench Spray,
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Service Water, Recirculation Spray, Safety Injection, and Auxiliary Feedwater system piping were observed for proper setting and conditions.
The inspector noted that baseplate anchor bolt inspections were in progress during this tour.
The control board was reviewed for annunciators that normally
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should not be lighted during the existing plant conditions.
The reasons for existing annunciators were described by a plant operator.
The licensee's policy and practices regarding plant tours were
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reviewed and no changes from previous practices were noted.
Control room manning was observed on several occasions during
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the inspection.
Shift turnovers were observed on several occasions to confirm that continuity of system status is main-tained.
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Acceptance criteria for the above items included inspector judgement, and the requirements of the Technical Specifications, 10 CFR 50.54(k), and the following procedures:
BVPS Unit 1, Systems Valve Lists, and Markups;
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OM 1.48.5, Safety Related Systems Valves and Equipment;
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MM Chapter 1, Section J, Housekeeping;
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MM Chapter 1, Section H, Cleaning and Maintenance Cleaning;
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and, SAD-25, Housekeeping and Cleanliness Procedure.
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c.
Findings (1) The inspector reviewed all entries in the control room jumper and lifted lead logbook for the systems listed below to determine that each entry was properly completed and authorized.
The following systems' entries were reviewed.
Reactor Control and Protection
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Reactor Excore Instrumentation
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Chemical and Volume Control System
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Residual Heat Removal System Safety Injection System
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Containment Vacuum System
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Containment Depressurization Systems
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Reactor Plant Component Cooling System
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Supplementary Leak Collection and Release System
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Main Steam System Main and Auxiliary Feedwater Systems
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River Water System
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4 KV Station Service System
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480 V Station Service System
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120 VAC Distribution System
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125 VDC Control System
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The BVPS Operating Manual, Chapter 1.48.5.0.2.b, Revision 5, requires, in part, that eacn entry include:
Elementary control schematic number where the jumper or
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lifted lead effect on the control circuit can be seen.
The reason jumper is placed or lead lifted with sufficient
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information to clearly identify location on the schematic.
The location of jumper or lead (breaker compartment / panel
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board).
The following log entries are examples of failure to comply with the above procedure:
Failure to provide elementary control schematic numbers:
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Tag Nos. 1792-1793, 1965 1967, 2190, 1834-1837, 1863-1865, 2065-2069, 2106-2118, and 2136-2137.
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Failure to identify reason for placement and information to clearly identify location schematic:
Tag Nos. 1834-1837, 2065-2069, 2080-2087, 2106-2118, 2158-2170.
Failure to specify location of jumper or lifted lead
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(breaker compartment or Panel Board):
Tag Nos. 1792-1793, 1965-1967, 2065-2069, 2080-2087, 1519-1520.
The above examples were extracted from in-service jumper and lifted lead entries and did not include numberous similar items which are no longer active in the official log.
The inspector noted that numerous other entries were only marginally acceptable based on the requirements of the licensee's procedure.
Failure to properly implement procedures for jumper and lifted lead control is contrary to Technical Specification 6.8.1.a, Regulatory Guide 1.33, November 1972; and, the BVPS Operating Manual, Section 1.48.D.2.b(3), (5), and (7), and constitutes an item of noncompliance at the deficiency level.
(79-16-01)
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(2) During the tours of the Auxiliary Building, the inspector observed a number of Caution Tags posted as part of the licensee's procedure for plant cooldown and entry into Mode 5 conditions.
These tags, specified by the cooldown procedure, are being administered as " permanent" caution tags but did not, in all cases, consistently meet the information requirements of BVPS Operating Manual, Section 1.48.6, for maintenance of " permanent" tags.
The information required to be included on these tags was readily available from the control room caution tag logbook as an alternate source.
The licensee acknowledged the inspector's comments and stated that the subject tags would be reviewed and corrections made to the inconsistencies noted.
This item will be reviewed during future inspections (79-16-02).
Except as noted above, the inspector had no further questions in this area.
4.
Loss of Residual Heat Removal (RHR) System Flow On two occasions during this inspection, RHR system flow to the core was temporarily lost due to malfunctions in the facility's emergency power systems.
The circumstances surrounding these events are further described below, a.
On July 3, 1979, while switching the No. 3 120 VAC Vital Bus from its inverter power supply to the alternate 480/120 VAC transformer source, a voltage transient induced by the switching actuated a false trip signal on two of four Containment High-High Pressure ESF logic trains.
These trip signals resulted in actuation of Phase B Containment Isolation.
The voltage transient appeared to be caused when the operators placed the additional load (No. 3 Vital Bus) on the same 480/120 VAC alternate source that was already carrying the No. 1 Vital Bus.
The false trip signals were generated when voltage momentarily surged to the two Containment High-High Pressure compar-ators which are powered from busses No. 1 and 3.
Field testing of these units in 1976 confirmed that such transients will consistently trip these comparators.
If the containment is at its nor:li, negative pressure, the trip will be momentary and and the comparators will automatically reset.
If the containment is at atmospheric pressure, as in the current Mode 5 operations, the trip will lock in as it did in this instance.
In either case above, ESF actuation of Phase B Containment Isolation equipment will occur.
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Upon receipt of two of four Containment High-High pressure signals, the following should normally occur:
Containment Phase B Isolation of Reactor Plant Component Cooling
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Water and Steam Generator Blowdown Lines Tripping of 4160 VAC " Stub" Busses which supply non-emergency
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loads including RHR and Component Cooling Water Pumps Initiation of Quench Spray and both Recirculation Spray Systems
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for containment spray and depressurization.
The actuation on July 3 did not result in Quench Spray and Recircu-lation Spray Systems initiation because these systems have been blocked for maintenance during the current outage.
Containment
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isolation valves appeared to function as required.
Additionally, RHR flow was interrupted for approximately 21 minutes resulting in an increase in coolant temperature from 107 - 1140 F.
The operators reset the tripped condition and restored cooling system alignments with no evidence of additional problems.
The inspectors review of this occurrence, including data review and discussions with plant personnel, indicated that the potential for recurrence during power oper+ tion does exist, possibly resulting in the above ESF actuations.
The licensee has taken administrative action to preclude recurrence as a result of this event.
BVPS Operating Manual Change Notice No. 79-67, issued on July ll, 1979, instructs operators to place one Containnent High-liigh Pressure channel in " bypass" prior to switching to alternate power.
Thus, temporarily precluding false coincident trip, signals during the transfer.
The channel is to be removed from " bypass" upon completion of the power supply transfer.
The vital power transfer panels are also posted with warning signs which should alert the operator to the additional actions required.
The inspector reviewed these actions and found them to be consistent with the facility's Technical Specifications and apparently suitable for interim operation.
Continuing acceptability of these actions and the reliability of emergency power systems is discussed further below.
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b.
On July 10, 1979, offsite power supplies were lost to 4160 VAC Nonvital Busses Nos. lA and 18, resulting in a loss of power to 4160 VAC Emergency Bus lAE.
This loss of power appears to have been due to a transformer differential current breaker trip caused by the starting current of a coolino tower (circulating water) pump while aligned to backfeed the Unit's fiain Transformer for testing.
The Emergency Bus No. lAE was deenergized and the No. 1 Emergency Diesel Generator (EDG) started, but did not sequentially load to reenergize Bus lAE as required.
The failure to sequentially load was due to the No. 1, 120 VAC Vital Bus inverter, which directly powers the EDG load sequencer circuits, being out of service.
Alternate 120 VAC power i:. not provided for the load sequencer.
Although the redundant Emergency Bus IDF did not deenergize, the No.
2 EDG which serves Bus IDF was out of service for maintenance at the time.
As a result of the loss of power to Emergency Bus lAE and its associated loads, the operating RHR pump was deenergized, stopping RHR flow to the core for approximately five minutes and resulting in a negligible increase in reactor coolant temperature.
The redundant RHR subsystem was operable and available but was not required to operate to restore cooling.
The affected RHR subsystem was returned to service with no evidence of additional problems.
The inspector reviewed the event with regard to the facility's safety analysis and Technical Specifications and determined that no violation of regulatory requirements had occurred.
Inoperability of the EDG load sequencing circuits due to inverter inoperability is, however, a concern for operations other than cold shutdown.
In such cases, inoperability of the associated vital power inverter will also render the EDG inoperable with respect to Technical Specifications.
The licensee has t' ken administrative action to preclude inadvertent a
disabling of the EDG load sequencing circuits at power due to inverter inoperability.
BVPS Operating Manual Change Notice 79-67, issued July 11, 1979, instructs operators that the load sequencers are unavailable under such circumstances and that additional operator action is required to assure immediate availability of EDG cooling water.
(The sequencer affects the River Water System cooling supply to the operating engine.) The vital power transfer panels are posted with warning signs to alert operators to the additional actions required by inverter inoperability.
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Inspector review of the licensee's action determined it was acceptable for continued, interim operation.
Continued acceptability of these actions and the operability of emergency power systems is discussed further below.
c.
As discussed above, and in paragraph 6 of this report, the facility's-emergency power systems have displayed failures which have resulted in operational problems in various modes of plant operation.
The inspector's review of the equipment's maintenance history appears to confirm the need for additional licensee attention to equipment reliability and its affect on the safety of operation.
The inspector addressed the following concerns to the licensee:
Through licensee event reports and maintenance records, the
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licensee has documented a continuing history of EDG breaker closure failures during surveillance testing.
Although corrective maintenance has been performed following such failures, the maintenance does not appear to have corrected the cause(s) of failure as evidenced by their recurrence.
Additional equipment failures / malfunctions in EDG auxiliary systems have not displayed the recurrent nature of the above breaker problems but their numbers appear to indicate the potential for auxiliary system malfunctions detracting from the reliability of the EDG's.
Both EDG's appear to be effected by breaker closure failures, although No. 2 EDG has not had a documented failure in over one year.
Licensee records indicate a similar recurrent trend for 120 VAC
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inverter malfunctions.
Corrective maintenance performed on these units over the past two years does not appear to have resulted in a major improvement in system reliability.
Principal problems have degradation or loss of output voltage and unstable output frequencies.
Discussion with plant personnel also indicated a sporadic
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problem history with miscellaneous emergency power system equipment and switchgear which, while not producing an identi-fiable trend, should be the basis for additional review by the licensee.
Such failures / malfunctions include apparently isolated cases of switchgear malfunctions and the presence of low resistance to ground conditions on emergency buses and associated transformers.
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The inspector informed the licensee that the following aspects of this matter would remain unresolved pending further action by the licensee and review by NRC:RI:
(1) Because actions taken to date have apparently been ineffective in correcting the cause of EDG breaker closure failures when the units are operated in the exercise mode, the operability of both EDG's is considered unresolved pending additional action by the licensee to establish the operability necessary to support plant startup.
The licensee stated that additional troubleshooting will be conducted in order to identify and correct this problem and that the results of this work will be reported to NRC:RI as a supplementary report to LER 79-09.
(79-16-03)
(2) The overall reliability and performance of all emergency power systems will be the subject of additional NRC:RI review.
The licensee stated that the DLC Engineering Department would be directed to perform an inhouse review of system reliability and performance in conjunction with related activities onsite.
The inspector stated that such a review should be completed on a timely basis with at least preliminary conclusions and recommendations available within six to eight months.
The licensee acknowledged this statement and informed the inspector that this would be incorporated into their planning.
This matter will remain unresolved pending continuing NRC:RI review of the licensee's actions in this regard (79-16-04).
5.
In Office Review of Licensee Event Reports (LER's)
The inspector reviewed LER's received in the NRC:RI office to verify that details of the events were clearly reported including the accuracy of the description of cause and adequacy of corrective action.
The inspector determined whether further information was required from the licensee, whether generic implications were indicated, and whether the event warranted onsite followup.
The following LER's were reviewed:
LER 79-02/3L, One of Two Containment Vacuum Pumps Inoperable
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LER 79-03/3L, Two of Three Charing Pumps Inoperable
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LER 79-04/99X, Inadvertent ECCS Actuation
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LER 79-06/3L, 120 VAC Vital Bus Inoperable
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LER 79-07/1T, Steam Driven Auxiliary Feedwater Pump Inoperable and Steam Generator Steam Supply Check Valve Inoperable
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LER 79-08/1T, Potential for A Single Dropped Rod to Lead to Calculated DNB Ratios Lower than Reported in the Safety Analysis
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LER 79-09/3L, Failure of Diesel Generator Output Breaker to Close
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LER 79-10/1T, Potential for Failure of the Inside Recirculation Spray Pump Motors LER 79-13/3L, Control Room Emergency Bnttled Air Pressurization
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System Inoperable
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LER 79-15/lT, Potential Steam Generator Level Instrument Error Due to High Energy Piping Break Inside Containment LER 79-16/3L, Incorrect Routing of Motor Operated Valve Power Cable
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- denotes those LER's selected for onsite followup.
6.
On Site Licensee Event Followup For those LER's selected for on site followup (denoted in paragraph 5 above), the inspector verified that reporting requirements of the Technical Specifications and Procedures SAD 14 and SAD 23 had been met, that appro-priate corrective action had been taken or planned, that the event was reviewed by the licensee as required by Technical Specifications and Procedure SAD 21, and that continued operation of the facility was conducted in accordance with Technical Specifications and did not constitute an unreviewed safety question as defined in 10 CFR 50. 59 (a) (2).
The licensee's internally issued incident reports associated with each of these events were each reviewed by the inspector.
The inspector's findings regarding these events were acceptable except as discussed below:
a.
LER 79-06 The inspector had no questions regarding the specific event but reviewed the generic implications of this and other similar vital bus inverter / power supply failures documented by LER's and the station's maintenance history.
This review indicated a consistent history of inverter malfunctiions including 17 instances of inopera-IP6 222
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bility requiring maintenance or repair since 1977.
In at least three of the instances identified by the inspector, the inverter malfunction resulted in a plant trip or inadvertent ECCS actuation due to electrical transients or loss of voltage on the 120 VAC Vital Busses.
This matter is further discussed in paragraph 4 of this report.
b.
LER 79-08 The licensee was informed by their NSS vendor of a potential nonconservatism in the safety analysis methodology for single dropped control rod which could result in exceeding allowable DN8R limits.
Based on the vendor's recommendation, the licensee has submitted proposed change to Technical Specifications which will reset the Power Range Neutron Flux High Negative Rate Trip setpoint from 5%
thermal power with a time constant of 2 seconds to 3% thermal power with a time constant ~of I second.
In anticipation of plant startup, the licensee has revised Maintenance Surveillance Procedures 2.03 through 2.06, Power Range Neutron Flux Channel N41 through N44 Quarterly Calibration with these new setpoints and is recalibrating the respective channels accordingly.
The licensee intends to proceed with the additional operating and surveillance procedure revisions to support operation with the new setpoints pending receipt of the approved Technical Specification amendment.
The licensee's safety evaluation of this matter concluded that the changes may be implemented on an interim basis under the provisions of 10 CFR 50.59 pending issuance of the Technical Specification amendments.
This appears to be consistent with informal discussions held among the licensee, the inspector, and NRC:NRR in regard to this matter.
The licensee stated that an updated LER will be submitted prior to plant startup summarizing the actions taken.
This matter will remain unresolved pending receipt and review of the supplementary report by NRC:RI (79-16-05).
c.
LER 79-09 The licensee has experienced a continuing history of Diesel Generator Breaker closure failures as discussed in IE Inspection Reports Nos. 334/78-24 and 334/79-13.
This event was reported to NRC:RI subsequent to those inspections and is similar to the failure previously discussed.
Inspector review of LER's and maintenance records identified at least thirteen individual occurrences of breaker closure failures on both diesel generators which each apparently involved operation with the breaker in the exercise (manual) mode.
Review of maintenance records and discussion with Operations, Main-tenance, and Engineering department personnel involved in the followup of this failure indicated that no clear solution has been identified and that trouble shooting has not been consistently pursued.
This matter is discussed further in paragraph 4 of this report.
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d.
LER 79-10 The licensee had been informed by the motor vendor that, as a result of pump testing at anotMr nuclear facility, the motor thrust bearings for the Inside Recirculation Spray Pump motors may seize or bind after several hours of running time.
The motors have been removed, modified by the vendor and returned to the station for reinstallation and testing.
The vendor apparently modified the motor bearing clearances to eliminate a problem of bearing overheating.
Reinstallation and testing of the modified motors is pending completion of design change documentation and satisfaction of BVPS Quality Assurance Program requirements.
The licensee had not determined, at the time of this inspection, the extent to which the motors must be tested prior to return to service.
The inspector informed that licensee that a supplementary report to this LER should be submitted prior to plant startup detailing the cort -tive actions taken and testing accomplished to return the pumps te cervice.
This matter will remain unresolved pendinq receipt and rev ew of the supplementary report by NRC:RI.
(79-16-06; e.
LER 79-14 The licensee is M the process of completing repairs to the defective welds identified oy this LER.
Additionally, the licensee has identified defects in adjacent feedwater piping during the inspections required by IE Bulletin 79-13, Cracking in Feedwater System Piping, and is currently planning repairs.
A supplemental report to this LER should be submitted prior to plant startup detailing the results of inspections not inc16ded in original LER and corrective actions taken to support the startup.
The licensee stated that a report will be submitted.
This matter will remain unresolved pending receipt and review of the subject report by NRC:RI (79-16-07).
f.
LER 79-15 This report identified a potentially nonconservative setpoint analysis for Steam Generator Low Low Level Trip instruments which would result from temperature induced error during a high energy piping break inside containment.
The licensee's NSSS vendor recommendations include raising the actual setpoint of the instruments by approximately 10% to account for temperature induced errors.
This action would place the trio setpoint veg near the normal operating level band, making plant startup and operation, using manual feedwater control, at low powers extremely difficult.
Under such conditions inadvertant reactor trips due to steam generator level oscillations will become very likely.
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In order to mitigate the effect of the temperature induced errors, the licensee is considering other methods for correction, rather than raising the trip setpoints.
These include insulation of the Steam Generator level instrument reference legs to reduce reference leg heatup and the associated signal error and/or reanalysis of the containment environmental conditions versus time for the postulated high energy pipe breaks.
The evaluation of these options was underway but incomplete at the close of this inspection.
The licensee informed the inspector that, if necessary to support a timely plant startup, they would conform to the NSS vendor's recom-mendation until complete evaluation of the options above could be accomplished.
The licensee is considering instituting administrative controls to accomplish the setpoint adjustments until such time that a comolete evaluation and a proposed Technical Specification amendment can be submitted to NRC.
This action is consistent with informal discussions among the licensee, the inspector, and NRC:NRR.
The implementation of corrective action for this matter is unresolved pending further review by NRC:RI.
The licensee stated that a supple-mental report to this LER will be submitted to NRC:RI to permit review prior to plant startup (79-16-07).
g.
LER 79-16 The LER was issued while this inspection was in progress.
The licensee was continuing investigation of the matter and performing corrective actions.
The inspector informed the licensee that a supplemental report should be submitted prior to plant startup detailing the corrective actions taken and the results of the investi-gation of other cable routings.
This matter is unresolved pending receipt and review of that report by NRC:RI.
7.
IE Bulletin and Circular Followup The inspector reviewed licensee actions taken in response to the following IE Circulars in order to determine that the Circular was received by licensee management, that a review for applicability to the facility was performed, and, that for those applicable to the facility, appropriate corrective actions have been taken or have been scheduled for implementation.
The following Circulars were reviewed:
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IEC 78-08, Environmental Qualification of Safety Related Electrical Equipment at Nuclear Power Plants
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IEC 78-16, Limitorque Valve Actuators
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IEC 78-14, Deterioration of Buna N Components in ASCO Solenoids
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IEC 79-01, Administration of Unauthorized Byproduct Material to
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Humans IEC 79-02, Failure of 120 VAC Vital AC Power Supplies
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This review included discussions with licensee personnel, examination of plant systems, and review of the following facility records:
The minutes of OSC Meeting Nos. 40-78, 3-79, and 13-79
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Station Memorandum No. BV: LGS-13, Selected Maintenance Work Requests and Equipment History Records
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No items of noncompliance were identified.
Except as noted below, the inspector had no further comments.
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IEC 78-08 - The licensee has cuapleted the applicable reviews noted by the IEC and appears to have incorporated the information into their response to IE Bulletin 79-01 as discussed in paragraph b below.
Further action in regard to this IEC will be reviewed as part of the licensee's activities associated with IE Bulletin 79-01.
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IEC 78-16 - This circular identified potential failures in certain types of valve motor operators following manual operation.
The licensee's review concluded that, while having experienced no similar failure history, valve operators of the type discussed in the circular should be electrically stroked following manual operation to confirm their operability.
Accordingly, the Operations Supervisor has issued a letter (BV: LGS:13, dated April 28,1979) to the plant operators directing that this be done after each manual valve operation.
The inspector questioned the long term effectiveness of such direction.
The licensee stated that the matter had been under consideration and a method for ensuring long term, cyclic review of Operations Department directive letters will be established to ensure the continuing awareness of the operators for items requiring long term attention.
The licensee further stated that such directives will also be reviewed for incorporation into permanent plant procedures where/when appropriate.
This item will remain unresolved pending further review of the licensee's actions by NRC:RI (79-16-08).
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IEC 79-02 - The inspector reviewed the licensee's evaluation of the
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IEC material which concluded that problems similar to those discussed in the IEC were not applicable to Beaver Valley.
While those specific problems do appear to be inapplicable, the information discussed in paragraphs 4 and 6 of this report addresses the general matter of failure of 120 VAC power supplies.
No items of noncompliance were identified.
b.
IE Bulletin 79-01, Environmental Qualification of Class IE Equipment The inspector reviewed the licensee actions taken in response to the above bulletin in order to determine that the written response was submitted within the required time period, that the response included the information required, including adequate corrective action commitments, and that licensee management forwarded copies of the response to responsible onsite management.
The inspector informed the licensee that the NRC technical review of the subject bulletin response was in progress and was being conducted by NRC personnel.
Additional comments resulting from that review may be addressed to the licensee in the future.
The inspector's review of the licensee's response noted that certain control signal cables within containment were identified as not having sufficient quality documentation to establish appropriate environmental qualification of the cable.
The licensee stated that the bulletin response was somewhat misleading in that the absence of quality documentation had not yet been confirmed.
The licensee plans to update this submittal as soon as additional research is completed.
The inspector noted the licensee's statements and emphasized that, should the absense of qualification data be confirmed, the matter is reportable as a prompt report to NRC in accordance with the facility's Technical Specifications.
The licensee's response also included identification of terminal boxes in containment which could sustain damage if subjected to post-LOCA pressure transients in containment.
The licensee stated that the subject boxes had been vented by drilling relief holes to preclude such damage.
The damage anticipated for such transients has been analyzed by the licensee to include only " denting" of the boxes with no possibility of collapse.
The inspector noted that venting of the boxes to the containment athmosphere could result in exposure of the wiring terminations and terminal boards within the boxes to the post-LOCA containment environment.
This will be the subject of additional review by NRC:RI.
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The acceptability of the licensee's response to IE Bulletin 79-01, the corrective actions taken by the licensee and compliance with Technical Specification reporting requirements for this matter will remain unresolved pending continuing review of the licensee's actions by NRC:RI (79-16-09).
c.
IE Bulletin 79-06A, Revision 1, Review of Operational Errors and System Misalignments Identified During the Three Mile Island Incident During this inspection, the licensee issued a supplemental response to the subject bulletin (C. N. Dunn letter of July 12, 1979) which addressed comments provided by the NRC task force whii:h is conducting the technical review of licensee responses to the bulietin.
Although formal technical review of this submittal had not been. completed by the task force, the inspector utilized the submittal to confirm that the commitments made therein were being implemented.
The inspector confirmed that the procedure revisions discussed in the licensee's response for bulletin items 2, 3, 7A, 78, 7C, 7D, 8, 11, and 12 had been incorporated into the BVPS Operating Manual (0M) procedures and were consistent with the commitments made in the July 12, 1979 letter.
The following revisions were reviewed:
BVPS OM Chapter 1 6, Reactor Coolant System, Section 4, Revision
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BVPS OM Chapter 1.11, Safety Injection System, Section 4,
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Revision 11 BVPS GM Chapter 1.48, Conduct of Operations, Section 1, Revision
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5; Section 5, Revision 5; Section 6, Revision 8; Section 7, Revision 7; Section 9, Revision 8; and, Section 10, Revision 12.
No items of noncompliance were identified.
The inspector informed the licensee that the review discussed above was not intended to supplant the review to be performed by the NRC task force and did not consititue NRC concurrence with the procedure changes reviewed.
Except with regard to administrative controls, as noted in paragraph 8 of this report, the inspector had no further comments on this matter.
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8.
Review of Administrative Controls For Proper Return to Service of ESF Following Maintenance The inspector reviewed the BVPS Maintenance Manual (MM), Chapter 1, Conduct of Maintenance, in order to determine that sufficient administra-tive controls have been established to ensure that, at the completion of maintenance, Engineered Safety Features will have been returned to an operable condition as discussed in IE Bulletin 79-06A.
This review found that Corrective Maintenance Procedures, Maintenance Surveillance Procedures, and Preventive Maintenance Procedures are each required to include the following specific features to ensure system operability:
Identification of system conditions or testing requirements for
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redundant equipment which are required before or during specific equipment outages.
Identification of inspections, surveillance test procedures, or
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operating procedures which must be performed to ensure that the equipment is operable and that it may be returned to its normal operating status.
Equipment alignment is controlled and returned to normal in accordance
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with the BVPS Operating Manual procedures for Clearance Control (Chapter 1.48).
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The inspector reviewed twelve randomly selected maintenance procedures of various types associated with each of the following systems and confirmed that the above provisions are being consistently implemented:
Chemical and Volume Control System
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Safety Injection System
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Containment Depressurization System
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Auxiliary Feedwater System
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Supplementary Leak Collection and Release System
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River Water System
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Emergency Electrical Power Systems.
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The inspector noted that the Corrective Maiatenance Procedures, Pre-ventive Maintenance Procedures, and Maintenance Surveillance Procedures reviewed did not provide specific valve / breaker / switch alignment or return to service instructions but, instead, required that the equipment be removed from and returned to service in accordance with Operating Manual Chapter 1.48 requirements for Clearances (safety tagouts). Inde-pendent verification provisions are similarly not specified.
The inspector reviewea the specific aspects of the BVPS OM, Section 1.48.6, Revision 7, for clearance procedures, in order to determine whether this procedure provided sufficient assurance that return to service valve / switch / breaker alignments would be specified and imple-mented in a manner that would ensure operability of affected equipment.
This procedure requires a Switching Order to be prepared for each operation or alignment change associated with placement or removal of safety tags.
This document is drafted by an operator and contains the specific alignment changes to be performed.
Each Switching Order is reviewed and signed by a Nuclear Shift Operating Foreman or Nuclear Shift Supervisor prior to its implementation and, in that form provides authorization for an operator to complete the manipulations listed on the Order.
The Switching Orders are not a detailed procedure but contain the necessary operating instructions for system alignment changes.
This evolution is required for both removal from and return to service.
When the valving or switching specified by the Switching Order is accomplished, the operator who has completed the evolution signs a control room copy of the Order signifying satisfactory completion of the required actions.
Switching Orders are directly traceable to the Clearance Permits which describe the reason for the work and which are the vehicle for control and coordination of the Clearance Tagging system.
The inspector reviewed the implementation of these instructions for approximately twenty-four active and inactive Clearance Permits and determined that the methods described above are being implemented consis-tently.
As previously noted, the administrative controls do not provide for predetermined alignment instructions as part of approved maintenance procedures nor do they provide for independent (multi party) verification of equipment status to ensure operability.
These matters are currently under review by NRC:NRR as part of the NRC Task Force review of licensee responses to IE Bulletin 79-06A.
Further inspection and review of this matter is pending finalization of review comments by the Task Force.
The Task Force was informally apprised of this matter subsequent to this inspection.
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9.
Review of Periodic Reports The inspector reviewed the following reports to verify:
the information required to be reported by NRC requirements had been
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included.
supporting information discussed in the reports is consistent with
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design predictions and performance specificatons.
planned corrective action is adequate for resolution of identified
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problems.
if any information should be classified as an abnormal occurrence.
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The reports reviewed were:
Monthly Operating Reports for January through June 1979
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Report of Facility Changes, Tests, and Experiments for 1978.
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Review of Monthly Operating Reports and NoC:RI inspection records by the inspector indicates that all periodic reporting requirements for operating statistics and shutdown experience required by Technical Specification 6.9.1.6 have been met for 1978.
No items of noncompliance were identified.
10.
10 CFR 21 Report - Hydrogen Recombiner Power Connectors During this inspection, the licensee was informed by the vendor that the portable hydrogen recombiner units provided at the facility for post accident containment hydrogen removal were equipped with power cable connectors of a type similar to that which experienced a failure at Three Mile Island Unit 2 during recovery from the incident of March 28, 1979.
The report detailed the results of the vendor's failure evaluation and provided a field instruction for inspection and repair (if required) of the BVPS connectors.
The report further noted that vendor personnel would be scheduled to visit BVPS to perform the above inspection and make any repairs resulting from the inspection.
At the close of this inspection, the licensee's plans included completion of this work prior to plant startup.
This matter will be followed during future inspections in order to determine the extent of repairs required and to confirm operability of the recombiner units prior to plant startup (79-16-10).
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11.
Unresolved Items Unresolved items are matters about which more information is required to determine whether they are acceptable, items of noncompliance or deviations.
Unresolved items addressed during this inspection are discussed in para-graphs 2, 4, 6, and 7 of this report.
12.
Exit Interview A management meeting was held with licensee personnel (denoted in paragraph 1) at the conclusion of the inspection on July 13, 1979.
The purpose, scope, and findings of the inspection were discussed as they appear in the details of this report.
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