IR 05000334/1987015
| ML20236R733 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 11/16/1987 |
| From: | Lester Tripp NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20236R727 | List: |
| References | |
| 50-334-87-15, IEB-79-02, IEB-79-14, IEB-79-2, IEIN-87-032, IEIN-87-034, IEIN-87-32, IEIN-87-34, NUDOCS 8711230452 | |
| Download: ML20236R733 (13) | |
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U. S. NUCLEAR REGULATORY COMMISSION
REGION I
Report No.
50-334/87-15
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Docket No.
50-334 Licensee:
Duquesne Light Company One Oxford Center 301 Grant Street Pittsburgh, PA 15279
Facility Name: Beaver Valley Power Station, Unit 1
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Location:
Shippingport, Pennsylvania l
Dates:
September 24, 1987 - October 29, 1987 Inspectors:
F. I. Yo g, Senior Resident Inspector, BV-1
M. Piidale, Resident Inspector, BV-1 Approved By:
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(L. E. Trip), Chief, Reactor Projects Section 3A
'Date i
Inspection Summary:
Inspection No. 50-334/87-15 on September 24, 1987 -
October 29, 1987.
Areas Inspected:
Routine inspections by the resident inspectors (93 hours0.00108 days <br />0.0258 hours <br />1.537698e-4 weeks <br />3.53865e-5 months <br />) of licensee actions on previous inspection findings, plant operations, physical l
security, radiological controls, housekeeping and fire protection and surveil-lance testing.
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Results:
No violations were identified.
Three unresolved items were identified concerning inadvertent liquid waste discharge (Detail 4.2.2),
Technical Specifications regarding containment isolation valves (Detail 4.2.3),
and improper shipment of radioactive waste (Detail 6).
Three previously open NRC items were closed during this inspection.
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TABLE OF CONTENTS Page 1.
Persons Contacted........................
2.
S umma ry o f Fa c i l i ty Ac t i v i t i e s.................
3.
Followup on outstanding Items..................
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Plant Operations........................
4.1 General..........................
4.2 Operations............
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4.3 Plant Security / Physical Protection............
4.4 Radiation Control s........
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4.5 Plant Housekeeping and Fire Protection
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Surveillance Testing.
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6.
Improper Shipment of Radioactive Waste.
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7.
Review of Non-essential Diesel Generator Trips.........
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Followup on NRC Information Notices...
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Review of Periodic Reports.......
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I 10. Unresolved Items......
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11.
Exit Interview
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l DETAILS
1.
Persons Contacted During the report period, interviews and discussions were conducted with l
members of licensee management and staff as necessary to support inspec-l tion activities.
i 2.
Summary of Facility Activities
At the beginning of the inspection on September 24, 1987, the plant was operating at 100% power and had recently surpassed its standing continuous
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run record of 104 days.
For this entire inspection period, the plant I
remained at or near full power and extended its continuous run record to
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142 days.
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3.
Followup on Outstanding Items The NRC Outstanding Items (OI) List was reviewed with cognizant licensee personnel.
Items selected by the inspector were subsequently reviewed through discussions with licensee personnel, documentation reviews and field inspection to determine whether licensee actions specified in the OIs had been satisfactorily completed.
The overall status of previously identified inspection findings were reviewed, and planned / completed l
licensee actions were discussed for those items reported below:
l 3.1 (Closed) Inspector Followup Item (83-21-03):
Design packages for IE
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Bulletins 79-02 and 79-14 lack proof that design changes incorpora-ting IE Bulletin requirements are controlled documents.
For IE Bulletin 79-02 and 79-14, two design change packages (DCPs), Nos. 305
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and 418 respectively, were generated to address the issues contained
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in those bulletins.
The licensee completed the final engineering reviews for the DCPs, and approximately 700 documents were identified to be updated.
As of February 2, 1987, the licensee completed the update of the 700 documents and considered the required actions associated with Bulletins 79-02 and 79-14 to be complete.
The inspector reviewed DCPs 305 and 418 to verify that the required updating of records had been performed.
No deficiencies were noted.
The inspector also verified that a program was in place to ensure
that the licensee would perform the necessary updates and corrections l
to drawings for all DCPs. Nuclear Engineering and Construction Pro-cedure 6.7, Records Update Control, describes the system in place to ensure that proper reviews were conducted for DCPs. In addition, the inspector sampled two DCPs to ensure that the system is effectiv.
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While it is apparent that the licensee is actively pursuing the DCP records update program, significant work still remains on updating other DCPs, and continues to be a significant workload for the engineering staff.
Currently, the licensee has contractor engineers on site to assist in reducing the record update backlog.
Although approximately 439 design change packages have been accepted'since initial construction of the plant, only 243 of the design change packages have been fully completed and documented.
The licensee expects to maintain the contracted personnel until the remaining 196 DCP record updates are completed.
This item is closed, however, the effectiveness of the licensee's program to complete the DCP record updates will be followed in sub-l
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sequent inspections.
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3.2 (Closed) Violation (84-31-02):
Licensee's Quality Assurance Program did not conduct a comprehensive review for 10CFR50 Criteria I, II, IV, X, XII and XVII associated with radiological waste transportation packaging.
Subsequent licensee reviews confirmed that prior QA audits in this area did not address all criteria regarding radwaste transportation packages.
The licensee revised the checklists used for auditing in the area of radiological waste transportation pack-aging. The inspector verified that audits beginning in the second
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quarter of 1985 in the solid waste / transportation area addressed the applicable criteria of 10 CFR 50. This item is closed.
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f 3.3 (Closed) Violation (84-31-04):
The licensee did not conduct a quality control audit of radiological waste packages that were
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shipped from the site to ensure they did not contain any free stand-
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ing liquids, waste was properly classified or.that the waste was
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structurally stable. The licensee's review of the QA program in' this area in response to the violation determined that weaknesses existed and that station procedures needed to be revised. The licensee sub-sequently revised Operating Manual, Chapter 18 (Solid Waste Disposal)
and Radiological Control Procedure RP 3.31 (Determination of Radio-active Waste Class).
In addition, the licensee also conducted training for QA/QC personnel to ensure that they were fully aware of the requirements and could properly audit the above mentioned areas.
The inspector reviewed the procedure revisions and verified that the necessary revisions were made. The inspector also reviewed the QA/QC training material and instructor lesson plans.
No deficiencies were identified.
Based on the above, this item ~is closed.
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4.
Plant Operations
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4.1 General l
Inspection tours of the following accessible plant areas were con-ducted during both day and night shifts with respect to Technical
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Specification (TS) compliance, housekeeping and cleanliness, fire
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protection, radiation control, physical security / plant protection and operational / maintenance administrative controls.
-- Control Room
-- Safeguard Areas l
-- Auxiliary Building
-- Service Building i
-- Switchgear Area
-- Diesel Generator Buildings
-- Access Control Points
-- Containment
-- Protected Area Fence Line -- Yard Area l
-- Turbine Building
-- Intake Structure 4.2 Operations
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During the course of the inspection, discussions were conducted with operators concerning knowledge of recent changes to procedures, i
facility configuration and plant conditions. During plant tours, logs
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and records were reviewed to determine if entries were properly made
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and that equipment status / deficiencies were identified and. communi-I cated. These records included operating logs, turnover sheets, tag-out and jumper logs, process computer printouts, unit off-normal and draft incident reports. The inspector verified adherence to approved procedures for ongoing activities observed.
Shift turnovers were witnessed and staffing requirements confirmed. In general, inspector comments or questions resulting from these reviews were resolved by licensee personnel.
Inspections conducted during backshifts and weekends verified that plant operators were alert and displayed no signs of fatigue or inattention to duty.
4.2.1 Engineered Safety Feature (ESF) Actuations On October 22, ESF actuations occurred when a plant oper-ator inadvertently caused an electrical spike on vital bus No. 2.
The operator was attempting to change paper in the main generator watt recorder.
After the power plug was removed from the paper drive, the plug assembly dropped onto a metal part of the recorder which shorted to ground and caused a large spike on Vital Bus No.
2.
Numerous alarms were annunciated and several plant components were actuated.
Included were two ESF systems, supplementary leak collection and release (SLCR) system and the control room emergency bottled air pressurization (CREBAP) system.
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exceeded on separate radiation monitors. (both' spiked _ high-i when the. event was initiated).
The SLCR system actuation caused the auxiliary building ventilation system-to divert'
its exhaust flow through. the' main filter banks and then-to the elevated release path on top of the containment build-
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ing.
The CREBAP system actuation initiated a discharge of
.l the pressurized bottled air into the control room.
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j fected plant systems were-subsequently ' restored. to normal j
without further incident, The NRC' was notified. of this
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event per 10 CFR 50.72 -reporting requirements.
Licensee corrective action for ~ this_ event included' placing -
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caution tags on recorders. of similar design _ that provide'
instructions to the ' operators' to pull the power plug (to.
l de-energize the. recorder) from inside the-main control board before changing recorder' paper.
The - effectiveness of th:e licensee corrective actions will be reviewed during
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subsequent inspections.
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4.2.2 Inadvertent Liquid Waste Discharge j
Upon. initiation of the CREBAP. and. SLCR systems: on October 22, 1987 (Detail 4.2.1),. plant' operators immedi-ately terminated a liquid waste system _: discharge (from liquid waste tank 7A) that was in progress at the time.
After the ESF : systems were restored to normal, ' the dis-charge was restarted. After approximately.nine_ minutes, a
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reactor operator noticed that the liquid - waste tank.7B outlet valve (M0V-LW-112B)_ was opun instead of the 7A out-
-let valve (MOV-LW-112A).
The 'incorrecti discharge was
.immediately stopped and.the Radiological Control' Department (RadCon) was notified. It was subsequently determined that
261 gallons of water was discharged fro'm tank 78.
RadCon
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completed an Abnormal Release Record which identified that no discharge / dose action levels or Technical ' Specification limits were exceeded.
On June 30, 1987, operator error Jcaused an. inadvertent l
gaseous waste discharge of the. wrong waste gas decay _. tank.'
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To. prevent recurrence 'of similar. events, the Nuclear Sta-tion Operating Supervisor (NSOS)' issued a Special Operating-
Order (S00).which specified that an operations supervisor be notified prior to' the initiation of any liquid or gas-eous waste discharge so that. details of the discharge ~could.
be discussed and confirmed.
At _the time of reinitiating
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i-the liquid waste discharge on October 22, the control room '
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l ities and the S00 was. not followed.
Additionally,4 the
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licensee stated that a large~ number.of discharges had pre -
viously. been performed, including a recent discharge 'of-liquid waste tank. 78. LThis may have been the reason that
7B was inadvertently opened.
The licensee issued a new S00 on October 28,.1987,.that was.
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more detailed' regarding pre-discharge requirements' and l
applicability.
Proper use of the 500 was also. emphasized
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to. all Operations ' Control Room personnel.
The licensee plans to revise station procedures to impose strongers administrative controls associated with both. liquid and gaseous waste discharges.. The inspector. also questioned
the development of post-initiation checkoffs ~(changes' in-R I
tank level, pressure, flow for the associated equipment)
to confirm the proper discharge.
The licensee will evalu-ate the benefits of such checklists as a part of their long'
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term corrective actions.
Various similar' events caused by
operator inattention to detail have occurred at BV-1, and j
licensee corrective' action and its effectiveness : will be l
closely monitored by. the inspectors.
Pending implementa-i tion of licensee corrective action' to prevent recurrence o
for. this type of ' events, - this item will be tracked' as i
Unresolved Item 50-334/87-15-01.
J 4.2.3 Containment Isolation Valve Technical Specifications
On October 2, 1987, the licensee identified that several Hydrogen Analyzer system valves listed in Technical.Speci-
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fication (TS) Table 3.6-l',
Containment Penetrations, are
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periodically opened during routine analyzer calibrations.
TS 3.6.3.1 (Containment Isolation Valves, CIVs) requires
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that the CIVs specified in-Table 3.6-1 be' operable in Modes 1 through 4.
The normal system alignment requires the-affected valves to be closed.
.TS. Table 3.6-1 allows test-
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ing for periodic opening of specific CIVs under administra -
tive controls.
The hydrogen analyzer valves do not have that provision noted.
The-licensee identified that TS ' 3.6.3.1-and Table 3.6-1 do not allow periodic opening of Hydrogen Analyzer System valves for testing.
However, the licensee is required to perf orm Hydrogen Analyzer calibrations and the valves' must'
be opened during calibration evolutions. Additionally, the same TS and associated table were unclear with ' respect to
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identification of CIVs in the table.
That is, plant y
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operations personnel.were not sure whether all valves listed in the table were CIVs. In the above case, when the valves were opened, they remained operable and conse-quently, TS 3.6.3.1 was not violated.
The licensee con-cluded that there may be additional valves which require the special testing provision in the table. A review was initiated to determine if there are other valves that require the provision. Any identified valves and any addi-tional TS 3.6.3.1 and Table 3.6-1 clarification changes, including CIV identification, will be the subject of a future TS change request.
Since the BV-2 TSs were modeled after the BV-1 TSs, a similar review is planned for BV-2.
Pending completion of the liceasee's review of the contain-ment isolation valve technical specifications for both BV-1 and BV-2, this is Unresolved Item 50-334/87-15-02.
4.3 Plant Security / Physical Protection Implementation of the Physical Security Plan was observed in various plant areas with regard to the following:
Protected Area and Vital Area barriers were well maintained and
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not compromised;
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Isolation zones were clear; Personnel and vehicles entering and packages being delivered to
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the Protected Area were properly searched and access control was in accordance with approved licensee procedures;
I Persons granted access to the site were badged to indicate l
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whether they have unescorted access or escorted authorization,
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l Security access controls to Vital Areas were being maintained
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and that persons in Vital Areas were properly authorized.
l Security posts were adequately staffed and equipped, security
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personnel were alert and knowledgeable regarding position
requirements, and written procedures were available; and l
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Adequate illumination was maintained.
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No concerns were identified.
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-4.4 Radiological ~ Controls Posting and control. of radiation and high radiation areas were inspected.
Radiation Work' Permit compliance 'and use of personnel-i monitoring ' devices were ' checked.
Conditions of _ step-off pads,. dis-posal _ of prote'ctive clothing,. cleanliness. of work areas, radiation control job coverage, area monitor. operability and calibration (portable and ' permanent)' and ' personnel frisking were observed on a '
i sampling basis.
No concerns were identified.
4.5 plant Housekeeping and Fire Protection Plant housekeeping conditions including general cleanliness ~ condi-tions and control and storage of. -flammable material and other poten-tial safety hazards' were observed in various areas during : plant
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tours. Maintenance of fire. barriers, fire barrier penetrations',- and
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verification-of posted fire. watches in these' areas were also.
observed.
No significant deficiencies were identified.
5.
Surveillance Testing The inspector witnessed / reviewed L selected surveillance tests.to determine-whether properly approved procedures were in use,: details were adequate, test instrumentation was properly calibrated and used, technical specifi-cations were satisfied testing was performed by qualified personnel and test results satisfied acceptance criteria or were properly dispositioned.
The following surveillance testing activities.were reviewed:
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OST 1.5A.1 Delta Flux Alarm Program Operability Check OST 1.7.1 Boric Acid Transfer Pump Operational Test
OST 1.11.6 LHSI Loop A - ECCS - Flow Path and Valve Position Checks
OST 1.24.2/3 Motor Driven Auxiliary Feedwater Pump Monthly Tests
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No deficienices were identified.
6.
- mproper Shipment of Radioactive Waste On September 8,1987, the. licensee shipped a compacted. radioactive. waste '
i (radwaste) load to an outside contractor for further' ' proce e si ng.
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shipment was to be compacted further as a measure to reduce radwaste volume prior to burial.
The load was shipped as radioactive. low' specific-activity (LSA).
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On October 13, 1987, the ifcensee was notified by the contractor that the shipment contained four more 55 gallon drums of radwaste than were identi-
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fied on the shipping papers.
The shipping manifest identified 80 drums
(by serial number) as the total. number shipped.
However, the contractor reported'that they processed 84 drums.
The licensee performed a check of their computerized inventory and found that the four extra drums should
have been in storage for a future radwaste shipment.
The licensee immediately conducted a meeting with 'both Operations and j
Radiological Control (RadCon) personnel to evaluate this event regarding j
cause, potential radiological consequences and immediate corrective ac-tions necessary to prevent recurrence.
The four drums were determined to-have' contained low level contaminated polyethylene. The maximum radiation level was 0.08mR/ hour and that' drum contained 89 microcuries of activity.
j The total activity of all four drums was 179 microcuries.
Additionally, the shipment was surveyed per plant procedures prior to leaving the site and all radiation levels were below regulatory requirements.. Therefore,
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the licensee determined that the radiological consequences of this event were minimal. The cause of the shipping error could not be determined at
.j the meeting. Since there were no outstanding radwaste shipments, and none planned in the near future, corrective actions for this event were to be determined at a critique meeting scheduled for the next day.
On October 14, a critique meeting was held with all available personnel.
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involved in the shipment.
The licensee concluded the following from the j
critique meeting:
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The trailer was supervised at all times during loading activities, j
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The trailer was loaded on September 3 ' and shored on September 4.
No additional drums were loaded prior to shipping the load.
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Computerized drum radwaste inventory sheets were used during loading activities to enable check-offs of drum serial numbers.
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A hand written tracking sheet was discarded after the trailer was loaded and the computerized inventory drum storage data was updated.
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Three signatures verified package inspections.
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Package inspection verifications were not compared to the total num-ber of packages.
Following extensive questioning and discussions, it could not be deter-mined how the error occurred.
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The licensee initiated two procedure changes to prevent recurrence of this event:
(1) Radcon Procedure RP 3.6, Radioactive Shipment Record will be revised to require that all hand written material generated for radio-
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active shipments be retained as part of the shipping record package, and.
(2) Radcon Procedure RP 3.29, Inspection of Radioactive Material Prior to Shipment, will be revised to include a double verification signoff as to the total number of packages in a shipment.. This verification will be in addition to the package inspection verification.
The licensee plans to complete the above procedure changes prior to the next radwaste shipment.
This event is a potential violation of NRC requirements.
Pending NRC Region I specialist review and followup of this i
event, and completion of the proposed corrective actions, this is Unre-solved Item 50-334/87-15-03.
7.
Review of Non-essential Diesel Generator Trips A potentially generic issue was identified at another site wherein emerg-ency diesel generator (EDG) engine non-essential protective trips were found not to be bypassed on loss of power (dead bus) conditions. A'non-essential diesel engine trip, which is bypassed on a loss of coolant accident (LOCA) signal but not on a loss of offsite power (LOOP) signal, was actuated during EDG testing and resulted in a trip of the EDG due to a spuriously generated signal.
There are two EDGs at BV-1, each of which automatically start upon' receipt of a safety injection (SI) signal or an emergency bus undervoltage signal.
The EDG output breakers automatically close on a bus undervoltage signal
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when the EDG is up to rated speed and voltage.
The non-essential EDG l
trips are automatically disabled on a loss of power to the emergency I
buses, leaving only the engine overspeed, generator differential and over-current trips enabled.
The EDGs are only loaded when a LOOP occurs and the non essential trips are disabled under the same conditions.
The test of automatic bypass function of the non-essential EDG trips is performed on a refueling interval and verifies that on a loss of power to the emerg-ency buses, all EDG engine trips are automatically disabled except for overspeed, generator phase differential current and overcurrent.
BV-1 uses a different design than that identified at the other site where-in non-essential EDG trips are not bypassed under certain conditions or actuations. The EDGs automatically load during accident conditions only when a LOOP occurs, -and the non-essential trips are also bypassed at that time. Based on the above, a similar safety concern was not identified at BV-1.
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Followup On Information Notices-(ins)
8.1 IN 87-32, Deficiencies in the Testing of Nuclear-Grade Activated
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Charcoal
,IN 87-32 was prov'ided to license'es _ to call attention to' deficiencies
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found in the testing of nuclear grade activated-charcoal. used for.
accident mitigation in nuclear facilities.
Periodic testing of the
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BV-1 nuclear grade activated ' charcoal is performed, by an outside-vendor contracted through Duquesne' Light Company.
Subsequent to the.
l initial identification of test result inconsistencies regarding;
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charcoal testing, the NRC issued revised and expanded results of i
inter-laboratory comparison studies. : The report _ documented testing
results' and concluded that the vendor contracted by the; licensee
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exhibited excellent monitoring and control of testing. activities.
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The purchase order for that vendor was recently amended to include testing of BV-2 charcoal.
No' concerns were identified.'
8.2 IN 87-34, Single Failures in Aux'iliary Feedwater Systems
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IN 87-34 was provided to' alert licensees of potential. single _ failures-of auxiliary feedwater (AFW) pump start 'and. protective - pump trip circuitry that could-cause partial or. complete-loss of capability to.
supply AFW. The IN cited examples of single failures in portions of
AFW pump start circuitry that are common to both ' motor-driven ~ AFW pumps and which could prevent both pumps from starting automatically.
i The licensee reviewed these concerns previously as similar-informa-tion was provided from the NSSS vendor" (Westinghouse).
The results of the licensee's review showed that no relays common to both motor driven AFW pump start circuits were identified.
The inspector inde-
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pendently confirmed the licensee conclusions.
Based ' on the above, the concerns identified in IN 87-34 are not applicable:.to BV-1.
9.
Review of Periodic Reports Upon receipt, periodic reports submitted pursuant to Technical Specifi-cation 6.9 (Reporting Requirements) were reviewed-The review assessed
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whether the reported information was valid, included the NRC required' data and whether results and supporting information were consistent with design predictions and performance specifications.
The inspector. also ascer-tained whether any reported information should be classified, as an abnor-mal occurrence.
The following periodic report was reviewed:
a BV1/BV2 Monthly Operating Report for P.lant. Operations from September 1 --
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30, 1987.
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10. Unresolved Items Unresolved items are matters for which more information is required in order to determine' whether they are' acceptable items or violations. Unre-solved items identified during this inspection are discussed in paragraphs 4.2.2, 4.2.3 and 6.
11. Exit Interview Meetings were held with senior facility management periodically during the course of this inspection to discuss the inspection scope and findings. A summary of inspection findings was further discussed with the licensee at the conclusion of the report period on November 4, 1987.'
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