IR 05000334/1980025
| ML19345F130 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 12/05/1980 |
| From: | Beckman D, Hegner J, Mccabe E, Simonetti G NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML19345F126 | List: |
| References | |
| 50-334-80-25, NUDOCS 8102060783 | |
| Download: ML19345F130 (41) | |
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U.S. NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFCRCEMENT Region I Ri;0 r
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~0 174/co 7:
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Cccket No.
50-334 License No.
CPR-66 Priority Categcry e
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Licensee:
Docuesne Licht Comcanv
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435 Sixth Avenue
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Pittsburch, Pennsv1vania 15219 Facility Name:
Beaver Vallev Power Station, Unit 1 Inspection at: Shippingport, Pennsylvania Inspection conducted: Auaust 16-September 12, 1980 Inspectors:
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l2 D.' A SecM,an Sr., Resident Inspector da te / signed
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MegnEr' Resi
- Irspector date' signed
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2/4 /20 M~
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- r Ingpector
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/6 Approved by:
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E. C. McCace,/ Chief, Reactor Projects
- tate signed Secticn No. 2, RO&NS Branch
Ins::ection Summary:
Inscections on Aucust 16-Sectember 12,1980 (Inscection Recort No. 50-334/80-25)
Areas Inspected:
Routine regular and backshift inspections by the resident inspectors (56.5 nours) and a regionally based inspector (32.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />) of: open items; plant operations and incidents; licensee etent reports; physical protection; licensee events; IE Bulletin and Circular followup; and load center locations.
Results:
No items of nonc0mpliance were identified.
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Region I Forn 12 (Rev. April 77)
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DCS Identification Nos. - IE Inscection 50-334/80-25 No.
Report Paragraph 50-334 - 80-02-7 2 LER 79-48
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50-334 - 79-09-10 2 LER 79-36 LHSI 50-334 - 80-0805 6 LER80-47)
50-334 - 80-0823 6 tER 80-59)
50-334 - 80-0827 6(LER80-60)
50-334 - 8C-0829 6 LER 80-61)
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50-334 - 80-0908 6 LER 80-62 -63)
50-334 - 80-0825 6 LER 80-64)
50-334 - 80-0912 6 (LER 80-65, -66, -67)
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DETAILS 1.
Persons Contacted
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R. Balcerek, Nuclear Engineering and Refueling Supervisor R. Surski, Senior Licensing Engineer J. Corey, Director of Nuclear Cperations M. Coppula, Results and Test Coordinator R. Hansen, Maintenance Supervisor J. Kosmal, Radeon Supervisor
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F. Lipchick, Senior Compliance Engineer L. Schad, Operations Supervisor
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J. Suber, Superintendent, Licensing and Compliance J. Werling, Station Superintendent H. Williams, Chief Engineer Other licensee personnel were contacted also.
2.
Licensee Action on Previous Inscection Findings The NRC Outstanding Items List was reviewed with responsible licensee personnel.
Items selected by the inspectors were subs 2quently reviewed through discussions with licensee personnel, documentation review, and field inspection to determine whether licensee actions specified in the OI's had been satisfactorily completed. Outstanding items are addressed below.
The overall status of previously identified inspection findings was reviewed, and planned and ccmpleted licensee actions were discussed for those items not re?)rted below.
a.
(Closed) Unresolved Item 80-01-09: Review Comoletion of Corrective
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Actions for Solic Racwaste Area Contamination Proclem. On January 12 anc 13, 1980, loose surface contamination was tracKec from the Solid Radwaste Disposal Area into previously uncontaminated areas on the 735' elevation of the Primary Auxiliary Building (PAB).
Personnel involved in operations within the Solid Radwaste Disposal Area also experienced contamination of their clothing and skin.
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corrective actions (see Inspection Report No. 50-334/80-01) included:
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A frisker station has been setup near the Solid Waste Area
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exit. A frisker had not been located there before due to
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transient high background radiation levels.
Safety meeting training material was to be provided and onshift
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training was to be conducted to review high radiation area entry rules, respiratory protection procedures, radiation work permit l
cer.trols and proper frisking techniques.
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A Radcon Department Memorandum had been forwarded to Radcon
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Foremen directing that all Radeon Technicians (RCTs) review requirements for familiarization of workers with radiation work permits prior to area entry, techniques for obtaining worst case samples for radiological evaluation, remote sampling techniques, and proper presentation of radiological data; and An Engineering Memorandum (dated January 29, 1980) had been sub-
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mitted to initiate a design change which would relocate the Solid Waste Ventilation Controls to the Solid Waste Control Room to expedite ventilation isolation should high airborne radioactivity levels be identified.
The inspector reviewed licensee records and held discussions with the Radeon Supervisor and other licensee personnel to determine the status of corrective actions. On A; gust 21, 1980, during a tour of the PAB, the inspectors verified that a frisker station had been placed near the Solid Waste Area Exit on the 735' election of the PAB.
Because of the potential high background problem, the frisker station was placed at a location approximately 30 ft. down a hall and not directly opposite the Solid Radwaste Area exit.
The licensee had marked the floor using striped tape to indicate the path personnel should follow to the frisker after exiting the Solid Radwaste Area.
The inspectors comented to a Radcon Foremen then in the area that the tape outlining the walkway was subject to wear during routine activities within the PAB not associated with the Solid Radwaste Area. The foreman stated that as part of the routine preparations for activity in the Solid Radwaste Area, Radcon personnel inspect the area and routinely replace old, worn, or dirty tape prior to comencing any activity.
Signs and barriers are also used to direct personnel to the frisker.
Independent d
inspector review confirmed that the BVPS Radcon Manual Chapter 3, Rad-con Procedure 4.1, Requirements for Setting Up a Frisker Station, Revi-sion 1, provides for these specific actions.
i On August 22, 1980 the inspector reviewed documentation of Safety Meetings conducted onshift by the shift Radcon Technicians (RCTs)
for operations personnel and DLC memorandum BVPS:JAK: 180 dated February 16, 1980 which directed shift RCTs to conduct shift train-ing which included the following information: proper frisking tech-niques; respiratory protection measures; radiation work permits, and High Radiation area rules. The following rosters were reviewed by the inspector:
Daily Training-Roster dated June 4, 1980, 02:00 hours; and
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Daily Training Roster dated June 4, 1980, 21:15 hours.
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The Radcon Supervisor said that a training session had taken place on the previous day but that documentation was not immediately avail-able.
The inspector later confirmed, through discussions with Opera-tions Depart:nent personnel, that the training had taken place.
In addition the Radcon Supervisor stated that a final training session would be performed to include those operations personnel that had not yet received the training.
On August 22, 1980, the inspector reviewed licensee memorandum BVPS:
JAK:167 dated January 25, 1980 which directed the Radcon Foremen to conduct training for all Radcon Technicians, specified topics to be covered and provided detailed training guidance which included:
Informing personnel signing onto RWPs to read the RWP prior
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to area entry; Procedures for obtaining liquid samples to assure that samples
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were representative of all material in the liquid; procedures for obtaining remote samples where possible; and
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normalizing radiological data and attention to units.
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The inspector reviewed the Daily Training Rosters attached to the refe-renced memo and verified that all Radtechs had received the specified training review.
Through discussions with the Radcon Supervisor and a station senior engineer on August 22, 1980, the inspector ascertained that licensee management had decided not to implement the design change for reloca-tion of the Solid Waste Ventilation Controls to the Solid Waste Con-
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trol Room.
Instead, the licensee is considering a different design change approach that would use portions of the fuel building ventila-tion system which were retired in place as a result of Design Change Packages 201/202 - Station Ventilation System Modifications.
The Radcon Supervisor stated that in the interim, whenever operations were to take place in the Solid Waste Area, easily accessible, portable ventilation systems would be provided to augment existing ventilation.
The inspector had no further questions on this item.
b.
(Closed) Unresolved Item 73-27-07: Review Licensee Evaluation and Actions for Foreion MaterMI Founo in EDG Fuel Oil Tanks.
Tne licensee conducted a revied of construction and preoperational test records to determine the source of foreign material found during December,1979 in the engine mounted EDG Juel Oil tanks but was unable to determine a specific cause or source.
The licansee also inspected wall-mounted fuel oil day tanks under Maintenance Work Requests Nos. 806351 and 806352 and concluded that no additional foreign material was present in any of the tanks.
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On August 18-19, 1980 the inspector reviewed the following documents:
Licensee Event Report No. 79-48/03L submitted February 7, 1980;
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DLC Incident Report No. IR-1-79-84 dated January 10, 1980;
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Cnsite Safety Comittee (OSC) Meeting Minutes No. SV-05C-87-79;
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and, NSQC General Inspection Reports dated December 13,18, and 21,
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1979; April 26, 1980; and July 14, 1980.
The inspector found that the licensee had adequately inspected all EDG fuel oil systems and could reasonably conclude that no additional foreign material was present.
The inspector noted that the referenced OSC minutes stated that the licensee intended to inspect or survey other safety-related systems during the current outage to ensure that no similar problems existed.
On August 20, 1980 the inspector selected several safety-related systems (on a sampling basis) and requested that the licensee provide documentation that those systems were also free of foreign material.
The following systems were selected by the inspector: Auxiliary Feedwater Pump Lube Oil sumps; Centrifugal Charging Pumps Lube Oil sumps; Diesel Driven Fire Pump Luce Oil sump; and Security System Diesel Generator Lube Oil sump.
The Nuclear Engineering and Refuel-ing Supervisor stated that Preventive Maintenance Procedures (PMPs)
for the Auxiliary Feedwater and Centrifugal Charging pumps provided verification that those sumps were free of foreign material; exami-nation of the Diesel Driven Fire Pump lube oil sump would take place
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during upccming maintenance which is scheduled for the pump; and, the Security System Deisel Generator lube oil sump cleanliness would be confirmed during installation currently in progress.
Based on this information, the inspector reviewed the following records:
PMP 1-7CH-P-1A-3M PM 06, Pump Bearing Lube Oil System Oil
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Change, performed Fecruary 23, 1977; PMP 1-7CH-P-18-3M PM 06, Pump Searing Lube 011 System Oil
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Change, performed August 1, 1978; PMP 1-7CH-P-1C-3M PM C6, Pump Bearing Oil System 011 Change,
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performed February 15, 1978; PMP 1-24FW-P-2-1M (Turb. Drive) PM 01, Clean and Lube Pump,
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performed June 20, 1979;
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PMP 1-24FW-P-3A-1M PM01, Clean and Lube Pump, performed
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September 24, 1977; PMP 1-24F4-P-3A-1E Plt 02, Inspect. Test, and Lube Pump flotor,
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performed January 21, 1980; PMP 1-24P4-P-38-1M PM01, Clean and Lube Pump, performed
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September 22, 1977; PMP 1-24FW-P-38-1E PM02, Inspect, Test, and Lube Pump Motor,
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performed March 10, 1979.
The inspector concluded that licensee actions were adequate to assure that other safety-related systems that might be susceptible to similar problems had been or will be examined prior to startup and that no additional foreign material had been found.
The inspector had no further qucstions on this matter.
c.
(Closed) Unresolved item 79-20-03:
Review of Licensee Actions Regard-ing Foreign Material Found in LHSI Pumos As a result of deficiencies observed during performance of a surveil-lance test on the 1A low Head Safety Injection on September,1979, a check valve (1-SI-29) in the pump's recirculation flowpath was removed and examined on two separate occasions and found to contain a total of three pieces of red plastic material later identified to be fragments of a WILCO No. HN-4-L fire hose nozzle.
(Inspection Report No. 79-20, Section 4 refers).
The fragments constituted only a small portion of the fire nozzle, and tne inspector asked that the following concerns be included and documented tri the licensee's evaluation:
the possible locations of fragments not found and the poten-
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tial consequences, including possible entry into the core, damage to ECCS pumps / components, and further potential for line blockage; the occasions on which the noz:le could have entered the
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system and the adequacy of the administrative controls which should have prevented such entry; the effects of the nozzle material on reactor plant system
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materials and fluid including chemical content, behavior of nozzle materials in various system operating environments, with potential for adverse effects noted; and, documentation of plans for the search for and removal of any
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remaining nozzle fragments from reactor plant systems, includ-ing the method (s) to be employe D e
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From August 18-22, 1980, the inspector met with the Nuclear Engineer-ing ano Refueling Supervisor and other merters of the licensee staff to discuss and review records of licensee actions taken in response to the concerns raised by the inspector.
The Nuclear Engineering and Refueling Supervisor stated that a visual inspection and flush of the 1A LHSI pump and system had been perfomed that no additional foreign materials nor damage to the pump had been identified.
In addition, an evaluation of the chemir.al properties of the fire naz-zie and their effect on reactor plant system materials and fluids had been performed and had concluded that no detrimental effects would result due to the presence of the remaining fragments.
The inspector reviewed the following records provided by the licensee in support of this position:
Schneider Inspection Report fM189-0-1553, Final Flush of
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LHSI Pump 1A; LHSI Pump 1A Pump Can Inspection, performed' January 5,1980;
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Fire Nozzle Specifications for Model HN-4-L Portable 1 1/2 inch
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Ad,justable Spray, Solid Stream, Shut-Off Nozzle; and, Fire Nozzle Corrosion Study (evaluation of chemical and material
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properties of the fire nozzle in an RCS environment).
The inspector also reviewed (on a sampling basis) additional licens-ee records which documented inspections of cleanliness conditions of other safety-related tanks and equipment to determine whether any similar conditions adverse to quality had been identified:
d Schneider memo dated June 26, 1980 for SR-TX-6A;
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Schneider memo dated July 3,1980 for BR-TX-6A;
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Schneider Inspection Report No. Z-1006-0-4624 for BR-TX-10;
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Schneider memo dated July 20, 1980 for BR-TX-2A and 23;
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NSQC General Inspection Report dated July 14,1980 for #2 EDG
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- 2 Heat Exchanger; NSQC General Inspect.Jn Report dated July 29,1980 for #2 EDG
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Lube Oil Cooler.
This review confirmed, on a sampling basis, that the licensee had sur-veyed or inspected a reasonable sample o1 safety-related equipment to ensure the absence of foreign material and had found no additional instances of foreign material which could impair equipment operability.
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During the inspector's review, the NSQC Supervisor provided the results of a recent NSQC surveillance inspection on the Refueling Water Storage Tank, QS-TX-1 (the suction source for the LHSI pumps)
where a significant breakcown in cleanliness / tool control had been identified by NSQC inspectors (ref: NSQC General Inspection Report
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dated April 12, 1980 and Nonconformance and Corrective Action Report No. 238, dated April 17,1980).
Inadequate administrative controls
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and numerous tools, logs, and debris were observed on both occasions in and around tha tank. In addition, similar deficiencies were observed by NSQC inspectors near the 18 Low Head Safety Injection pump. The inspector ascertained through review of NSQC General Inspection report dated July 25, 1980 that the deficiencies noted i
in the referenced reports were corrected by the licensee and that the following corrective actions were initiated to prevent recurrence:
Two full time QC Cleanliness Control inspectors were assigned
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within the licensee contractor's organi:ation and provided appropriate training; Licensee Construction Division (CDN) personnel were provided
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additional training to emphasize requirtients of CDN clean 11-ness and housekeeping procedures; and,
_ Appropriate CDN procedures regarding housekeeping / cleanliness
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were revised to provide more explicit guidance.
This confirmed on a sampling basis that the licensee's management control system was identifying review conditions adverse to quality (i.e., that might affect safety-related equipment operability) and
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assuring that appropriata and timely corrective actions were being
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initiated. Adequacy of the corrective actions will routinely be observed during future inspections. The inspector had no further
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questions at this time.
d.
(Closed) Inspector Follow Item 80-12-01: Construction Deficiencies in Safety Related Tanks.
I Additional NFC review of licensee activities regarding repair programs i
I for the deficient welds in the RWST and DWST, review of analysis which
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indicated that tank / piping failure would not occur under design basis
accident conditions with the deficient welds present, and continuing review of QA/QC records associated with fabrication of the tanks are discussed in Inspection Report No. 50-334/80-17.
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THIS PAGE, CONTAINING 10 CFR 2.790 INFORMATION, NOT FOR PUBLIC DISCLOSURE, IS INTENTIONALLY LEFT BLANK.
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f.
(Closed) Unresolved I+em 79-27-04: Acceotability of UT orocedures used for IEB 79-17 pioing inscections.
During prior inspections, discussions between the inspectors and licensee personnel and review of licensee contractor UT calibration proce? res had indicated that the methods used for part of the piping examinations could yield potentially invalid results.
Additional review by tne licensee determined that the questionable methods had been applied only to piping which had a nominal pipe thickness of less than 0.250 inches.
Paragraph 2(b) of IES 79-17
Revision 1, required that only dye penetrant and visual examination be used for such piping.
The inspectors confirmed that UT Calibra-tion Technique "C" had been used only on piping of this size, that all inspection results (including dyde penetrant and visual) were otherwise acceptable, and that ce licensee's final IE3 submittal adeocately addressed this matter. Ccepletion of IE3 79-17 implemen-tation reviews is further discussed in paragraph 3 of this report.
3.
IE Bulletin Followuo '
The inspector reviewed licensee actions taken in response to the following IE Bulletins to assess whether:
the written response was submitted within the required time period; the response included the infor.ation required including adequate corrective action comitments; and licensee management had forwarded copies of the response to responsible onsite management. The review included discussions with licensee personnel and cbservations and review of items discussed below.
a.
IE3 79-17:
pi::e Cracks in Stagnant Borated ' dater Systems.
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From November 1974 to February 1977 a number of cracking incidents occurred at c-her facilities in safety-related stainless steel piping systems which contain oxygenated, stagnant or essentially stagnant borated water. Licensees were required to conduct a review of safety-related stainless steel piping systems using visual inspection and nondestructive examination and take appropriate corrective actions.
The DLC reply (January 3,1980 letter, C. N. Dunn to S. H. Grier, Subject:
IE3 79-17. Revision 1) to the subject bulletin provided a sumary (Table I) the inspection program conducted by the licensee.
A later OLC letter (August 18, 1980, C. N. Dunn to 3. H. Grier, same subject) submitted a revised Table I for which data on ultrasenic examination (UT) of welds on pipe thickness less than 0.250 inches had been deleted. The acceptability of the UT technicue used on these welds had been previously questioned by NRC:RI.
The licensee stated that these welds were not required to be UT examined by Revision 1-to the bulletin. The inspectors reviewed this licensee position as dis-cussed in paragraoh 2 of this report.
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On August 21, 1980 the inspector cet with the NSQC Supervisor to discuss and review NSQC inspection reports which documented the required examinations and provided the bases for the information summarized in Table I.
The inspector selected several NSQC inspection reports on a sampling basis and determined that the inspections had been performed in accordance with the bulletin requirements.
NSQC inspection reports for the following lines were examined:
6"-RS-30-1538-Q2 6"-SI-44-153W-Q2 6"-RS-29-1538-Q2 (modified per DCP 162)
12"-RH-14-152-02 6"-RH-14-152-Q2 10"-RH-24-1502-Q1 6"RH-14-152-Q2
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In addition, the inspector examined drawings of the Residual Heat Removal, Recirculation Spray and Safety Injection systems and selected lines which would normally contain stagnant, borated water. The inspectors then confirmed that those lines had been included in the licensee's inspection program by ascertaining whether the selected lines appeared in Table I.
The acceptability of the licensee's nondestructive examination tech-niques were previously reviewed during IE Inspection No. 50-334/79-27.
The inspector found that the licensees actions were acceptable.
b.
IEB 79-21: Temoerature Effects on Level Measurements.
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DLC supplemental response (July 24, 1980 letter, C. N. Dunn to B. H.
Grier, Subject:
IES 79-21) stated that the following actions were taken in response to the bulletin:
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To minimize reference leg heatup due to increased containment temperature caused by a LOCA or secondary break inside contain-i ment, insulation was applied to the reference leg water columns in accordance with instructions provided by the Westinghouse Electric Corporation.
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The steam generator low-low trip setpoint has been set at 12 i
percent of level span as reconnended by Westinghouse for the l
insulated condition of the legs.
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3.
The review and revision of emergency procedures has been completed to reflect the new reference guidelines developed by Westinghouse and the Utility Owners Group and the opera. tors have been instructed on the potential for and magnitude of erroneous level signals caused by adverse environments."
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License actions in response to the bulletin are being coordinated by Design Change Package (DCP) 2/7, Insulated Steam Generator Reference Legs.
Inspector review of licensee actions specified in Item I was conducted during previous inspectior (Reference:
Inspection Report
No. 50-334/80-20). Those actions were found to be in accordance with
bulletin requirements.
The inspector had no further questions regard-
ing Item 1.
On August 28 and September 4,1980 the inspector discussed Items 2 and
3 with the station engineer responsible for DCP 277. On September 8,
1980, the inspector asked the engineer to explain how the revised
setpoint had been determined.
The engineer stated th'at DLC action was based on guidance provided
by the vendor (Reference: June 27, 1980 Westinghouse letter DLW-80-77,
F. Noon to J. Werling, Subject: Steam Generator Reference leg Heatup
Options). The vendor recommended the installation of insulation. Based
on that solution, the vendor letter stated that the only instrument
errors that had to be incorporated in the low-low Steam Generator Reactor
level trip setpoint were those due to normal channel accuracy and
transmitter deviation.
According to the letter, the specified insu-
lation requirements ensured that the water level error due to reference
leg heatup would not exceed 2 percent of span.
The licensee concluded
that a revision of the low-low trip setpoint from its former 10 percent
to 12 percent of level span was appropriate and in-accordance with the
vendor guidance. The inspector reviewed the referenced document and con-
cluded that the licensee's actions were acceptable.
The engineer stated that recalibration of the setpoints was controlled
by means of Maintenance Surveillance Precedures and stated that the
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following procedures were applicable:
MSP:24.-17 L-4741A Steam Generator Level Loop 1 Protection
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Channel I Calibration, Revision 5;
MSP 24.18 L-4751A Steam Generator Level Loop 1 Protection
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Channel I Calibration, Revision 5;
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' MSP 24.19 L-4761A Steam Generator Level Loop 1 Protection
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Channel I Calibration, Revision 5;
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MSP 24.20 L-4841B Steam Generator Level Loop 2 Protection
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Channel II Calibration, Revision 5;
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MSP 24.21 L-485 IB Steam Generator Level Loop 2 Protection
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Channel II Calibration, Revision 5;
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MSP 24.22 L-486 LB Steam Generator Level Loop 2 Protection
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Channel II Calibration, Revision 5;
MSP 24.23 L 4941C Steam Generator Level Loop 3 Protection
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Channel III Calibration, Revision 5;
MSP 24.24 L-4951C Steam Generator Level Loop 3 Protection
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Channel III, Calibration, Revision 5;
MSP 24.25 L-4961C < team Generator Level Loop 3 Protection
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Channel III Calibration, Revision 5;
MSP 24.01 L-474 Steam Generator 1A Level Protection Channc' I
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Test, Revision 4;
MSP 24.02 L-475 Steam Generator 1A Level Protection Channel I
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Test, Revision 4;
MSP 24.03 L-476 Steam Generator IA Level Protection Channel I
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Test, Revision 4;
MSP 24.04 L-484 Steam Generator 1B Level Protection Channel II
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Test, Revision 4;
MSP 24.05 L-485 Steam Generator 1B Level Protection Channel II
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Test, Revision 4;
MSP 24-06 L-486 Steam Generator IB Level Protection Channel II
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Test, Revision 4;
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MSP 24.07 L-494 Steam Generator 1C Level Protection Channel III
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Test, Revision 4;
MSP 24.08 L-495 Steam Generator 1C Level Protection Channel III
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Test, Revision 4; and
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MSP 24.09 L-496 Steam Generator 1C Level Protection Channel III
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Test, Revision 4.'
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On September 10, 1980 the inspector examined the above procedures and
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found that the new setpoint had not yet been incorporated. When ques-
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tiened by the inspettor as to why the procedures had not been revised,
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the station engineer stated that revision of the precedures had not
yet taken place penoing receipt of other unrelated information to be
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incorporated in the procedures.
The inspector questioned the Proce-
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dures Engineer assigned to revise the above procedures and confirmed
that the individual was aware of the changes required for the steam
generator low-low level setpoints. Revisions to and implementation
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of the revised procedures will be confirmed during future inspections
(79-8U-21).
IEB 79-21 Item 4 asked licensees to review and revise
Emergency Operating Precedures (E0Ps) as necessary to address the
results of actions required by tre bulletin.
On September 4,1980
the inspector asked the Compliance Engineer to furnish evidence that
such a review had been performed and necessary corrective actions had
been initiated.
The Compliance Engineer referred the inspector to
the Training Specialist assigned responsibility for review / revision
of E0Ps.
The inspector confirmed through discussions with the reviewer that
a review and draft revision of E0Ps was conducted, as stated in the
licensee's supplemental bulletin response, and using the new reference
guidelines developed by Westinghouse and the Utility Owners Group.
Additional licensee review was continuing. However, the licensee was
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unable to show that an E0P review pursuant to the bulletin require-
ments had been performed.
The licensee had assumed that a review
of the E0Ps using the new reference guidelines met the bulletin
requirements since the same vendor had been the originator-of both
the guidelines and the solution to the issue raised by the bulletin.
When questioned by the inspector neither the Compliance Engineer nor
the Training Specialist was able to confirm that the Luthors of the
reference guidelines had incorporated or considered the effects of
increased containment temperatures on reference leg water columns.
The licensee has elected to perform an additional review in accordance
with the bulletin requirements.
NRC:RI review of licensee actions in
this regard will be conducted during future inspections (80-25-04).
Item 4 of the bulletin also asked licensees to assure that operators
were instructed on the potential for and magnitude of erroneous level
signals.
In its supplemental response dated July 24, 1980 the licens-
ee stated that the required instruction had been completed.
On August 28, 1980 the inspector met with the Training Supervisor
to review documentation that the required training had been performed
in accordance with bulletin requirements.
The Training Supervisor
stated that the information required by the bulletin was included in
Lesson Plan LP-DCP-2, Design Changes to Beaver Valley Power Station,
dated June 16, 1980.
The inspector reviewed the portion cf the lesson
plan entitled Steam Generator Narrow Range Reference leg and confirmed
that the required infonnation had been included.
The subject lesson
plan was presented several times over the period July 14-September 2,
1980. The inspector reviewed attendance rosters for the days the
referenced lesson plan was presented to determine whether all licensed
operators had received the training.
Inspector review of Daily Train-
ing Rosters entitled Plant Design Change Training, LP-0CP-1, 2, 3, 4
.
or Plant Design Change Training, LP-DCP-1-14 determined that all licensed
l
operators had received the specified training.
Except as noted, the
inspector had no further questions regarding licensee implementation
of IES 79-21.
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c.
IEB 80-19:
Failures of Mercury Wetted Matrix Relays in Reactor Protec-
tive Systems of Operatina Nuclea. Power Plants Desicned by Comuustion
Engineering
The bulletin asked licensees to review their facilities to detennine
whether mercury wetted relays were used in the Reactor Protection System
(RPS).
The licensee's response to the bulletin (DLC letter dated
August 18, 1980; C. N. Dunn to B. H. Grier; Subject:
stated that, based on a review of the RPS, no mercury wettd relays
were installed in the RPS.
On August 21, 1980 the inspector reviewed documentation which provided
the bases for the licensee's letter and discussed the scope and metho-
dology of the review with the responsible station engineers.
The
inspector also reviewed DLC memorandum No. BVPS:JES:31, dated August 8,
1980, which documented the licensee's review. That review consisted
of examination of the applicable systems' technical manual description
-
and parts lists for the Solid State Protection System, the Primary
Process Control System, and the Nuclear Instrumentation System.
The
inspector asked the station engineers if their review had considered
modifications to the RPS, e.g., Westinghouse Reliably and Availability
Program (WRAPS) modifications.
The station engineers stated that the
Westinghouse WRAPS manual had been examined as part of the review and
that the onsite vendor representative had been contacted for confirma-
tion of their review. As a result of that discussion and review, the
licensee determined that no mercury wetted switches were involved in
WRAPS modifications.
The inspector concluded that the licensee's actions described above
in response to the bulleting requirements were acceptable.
d.
IEB 80-20:
Failures of Westinghouse W-2 Sprirq Return to Neutral
Control Switches.
The subject bulletin described malfunctions of Westinghouse Type W-2
'
control switches and asked licensees to determine whether such switches
'
l
were used in safety-related applications at their facilities.
The DLC
l
response to the bulletin (DLC letter dated August 18, 1980, C. N. Dunn
l
to B. H. Grier, subject:
IEB 80-20) stated that a review had been con-
l
ducted and had concluded that no Westinghouse Type W-2 switches were
'
used in safety-related applications at Beaver Valley Unit 1.
l
On August 21, 1980 the inspector reviewed DLC memorandum BVPS:JES:30,
l
dated August 8,1980, which provided the basis for the licensee's
'
response.
In addition, the scope and methodology of the review was
discussed with the station engineer responsible for the review.
The
memorandum described a document review of:
Elementary Diagrams
l
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O
~-,
w
e
w
,
w
,.
.
.
.
(RE-21 Series); Instruction Scok 1.12-139A, Main Control Board Switches
Supplied by Westingnouse; Purcnase Order BV-467, Miscellaneous Control
and Relay Panels; and the CLC Cesign Change Package Indes. Whenever
the reviewer believed that a DCP could have involved control switch
replacement, a detailed review of the CCP was conducted.
The inspector noted that no field verification was conducted by the
licensee as part of the review process and questioned the responsible
individual as to why no field checks had been performed. The station
engineer stated that, considering the time constraints imposed by the
bulletin and that no switches of interest had been identified through
dcument review, a field check would not have been of any additional
value.
On August 22, 1980 the inspector incependently reviewed the following
elementary drawings:
S&W-RE-21-JK-1, Quench Spray, Sheet 1;
--
S&W-RE-21-JL-1, Quench Spray, Sheet 2;
--
S&W-RE-21-JU-1, Residual Heat Removal System, Sheet 1;
--
S&W-RE-21-JW-1, Recirculation Spray, Sheet 1; and,
--
S&W-RE-21-KK-1-A-3, Safety Injection, Sheet 2.
--
The inspector found no W Type W-2 control switches on the selected
drawings.
In addition, the inspector observed several control switches
on the Main Control Scard and determined tnat they were not W Type W-2.
'
The inspector also verified through examination of appropriate elemen-
tary drawings that the selected switches were not of the type described
in the IE3.
The inspector's observations are listed below:
Switch Observed
S_ witch,Ty3
Drawing Examined
TV-1MS-111A (1AMS/ Drain
Isolation)
W Type OT2
S&W-RE-21-HY-1A4
MOV-1Fd-156A (IASGF4/Contain-
ment Isolation)
W Type 072
S&W-RE-21-HF-1
TV-ISI-8848 (BIT Recirc to
Boron Injection Surge Tank
Isolation)
W Type OT2
S&W-RE-21-KL-1
IWR-P-1C (IC Reactor Plant
River Water Pump)
W Type W
S&W-RE-21-KX-1-
ISI-P-3A (3A Baron Injection
Recir Pumo)
W Type W
S&W-RE-21-Ci-1-
.
f
.
..
Based on review of actions taken by the licensee in response to the
bulletin and independent sampling review by the inspector, the inspec-
tar found that there was reasonable assurance that the switches of
concern were not installed at Unit 1.
4.
IE Circular Followuo
The inspector reviewed licensee actions taken on the following IE Circular
in order to determine that it was received by licensee management, that a
review for applicability was performed, and that corrective actions, if
appropriate, have been taken or planned for implementation.
IEC 80-18:
10 CFR 50.59 Safety Evaluation for Chances to Radioactive Waste
Treatment Systems.
NRC inspection experience at other facilities, licensee failures to perform
adequate safety evaluations for radwaste systems changes made pursuant to
During discuss 10ns among the inspectors, the Plant Superin-
tendent, and the station senior taff at the exit meeting for this inspec-
tion, the Superintendent stated 'that the IEC had been reviewed by the
Onsite Safety Comnittee and was considered to be of great value for
structuring the committee's reviews of all changes, tests, and experiments
pursuant to 10 CFR 50.59. A permanent committee secretary has recently
been appointed and has been instructed to include the IEC guidance in all
future connittee review activities.
The new committee secretary has been
charged to assure that all committee reviews and safety evaluations pursuant
to 10 CFR 50.59 comply with the IEC guidance and are properly reflected in
committee records and minutes.
The inspectors will routinely review the licensee's implementation of the
s
IEC during future inspections of radwaste and design change activities.
5.
Review of Plant Ooerations
a.
General
Selected plant areas were inspected on the dates noted during the day
and night shifts to assess housekeeping, cleanliness, fire protection,
j
radiation control, physical security and plant protection, operational
l
and maintenance administrative controls, and Technical Specification
compliance.
Acceptance criteria for the above areas included the following:
BVPS FSAR Appendix A, Technical Specifications;
--
i
OM 1.48.5 Section D, Jumpers and Lifted Leads;
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CM 1.48.6 Clearance Procedures;
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OM 1.48.8 Records;
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OM 1.48.9 Rules of Practice;
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BVPS Operations Manual, Chapter 55A, Periodic Checks - Operating
--
Surveillance Tests;
BVPS Operations Manual, Chapter 54, Station Logs;
--
BVPS Maintenance Manual, Chapter 1 Conduct of Maintenance,
--
Section J, Housekeeping;
BVPS Radcon Manual, various sections;
--
10 CFR 50.54(k) Control Room manning requirement; and
--
BVPS Physical Security Plan.
--
Station Administrative Procedures
--
Nuclear Operations Directives
--
Inspector Judgement
--
.
The Station Administrative Directives (SADs) we e superseded on
August 29, 1980 by the Station Administrative Pitcedures and Nuclear
Operations Directives.
As stated in the transmittal letter (No. BVPS:
GWM:267) distributing the new procedures and direc*.ives: "The Station
Administrative Procedures are intended to gather together all mandatory
administrative requirenents for activities involving Tategory I safety-
.
related items for which individuals under the direct supervision of
'
the BVPS Unit i Station Superintendent are responsible... Nuclear Opera-
tions Directives are provided to assure that activities are clearly.
understood and properly interfaced by individuals with nuclear respon-
sibilities within the Power Station Department including those who are
not on the BVPS Unit 1 staff."
.
These procedures and directives will be reviewed for compliance with
the various applicable NRC requirements during future routine inspec-
tions.
The newly issued procedures and directives were issued in
fulfillment of DLC commitments made during IE Enforcement Conference
No. 50-334/79-21.
b.
Areas Insoected
Primary Auxiliary Building, except High Radiation Areas and Loose
--
Surface Contamination Areas (August 19, 20, and 21)
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,
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Service Building (August 19, 20, and 21)
--
Main S*aam Valve Room (August 20)
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Purge Duct Room (August 20)
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East /Wnt Cable Vaults (August 21)
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Containment Building, including High Radiation Areas (September 11)
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Safeguards Area (August 20)
--
Various Switchgear Rooms; Cable Spreading Room (August 20)
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Protected Area (August 19, including accessible rooftops;
--
September 11)
Fuel Handling Building (August 20)
--
The inspectors also toured the Control Room on a daily basis to review
logs and records and conduct discussions with operators concerning
reasons for selected lighted annunciators and knowledge of recent
changes to procedures, facility configurations and plant conditions.
On August 20, 1980 the inspectors conduc:ed discussions with Control
Room operators to ascertain operator awa"eness of plant status.
Rea-
sons for selected lighted annunciators which were of significance
undering existing plant conditions were discussed.
t
Reasons Lighted
'
,
Radiation Monitor High/ Radiation
I&C Technician performing sur-
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i
Monitor Power Supply Failure
veillance test on radiation
monitors.
Containment Sump Level High
-
Condensate from containment
cooling coils gathering in
j
sump.
Pump in av.to and running.
,
Main Transforcer F.re
Fire suppression system deluge
valve isolated for maintenance.
Fuel Pool Level Alarm
Water leaking by isolation
valves during perfornance of
TOP 80-27 raised level (refe-
rence: paragraph 5.c., this
report).
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No items of noncompliance were identified.
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c.
Observations
(1) Conformance with Technical Soecifications
Control Room monitoring instrumentation and equipment conditions
were observed to verify that instrumentation and systems required
to support Mode 5 operations were in conformance with Tecnnical
Specification (TS) Limiting Conditions for Operations (LCOs). The
following equipment / instrument conditions were observed on the
specified dates with respect to the LCOs indicated:
Boric Acid Flowpath (August 19)
--
Boric Acid-Storage Tank Level and Temperature
TS 3.1.2. 7
--
(August 20)
Source Range Nuclear Instruments (August 20)
--
.
Radiation Monitor Operability (August 20)
'
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RM 207 (Fuel Storage Pool Area:
--
RM 103A/S Fuel Storage Building Gross Activity)
--
Boric Acid Transfer Pumps Operability (August 19) TS 3.1.2.5
--
AC/DC Electrical system Availability (August 19
--
3.8.2.2; and
3.8.2.4.
d
No items of noncompliance were identified.
(2) Radiation Controls
Radiation controls, including posting of radiation areas, the
condition of step-off pads, disposal of protective clothing,
completion of radiatior, work permits, compliance with. radiation
work permits, personnel monitoring devices being worn, cleanli-
ness of work areas, radiation control. job coverage, area monitor
operability (portable and permanent), area monitor calibration,
and personnel frisking procedures were observed on a sampling
'
basis in the following areas on the dates indicated:
primary Auxiliary Building (PAB) (August 19, 20, and 21)
--
,
Containment Airlock (September 11)
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Containment (September 11)
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.
,
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.
.
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,
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.
The following Radiation Work Permits (RUP) were reviewed for
completeness:
RWP 6752 PAB/ Safeguards Butiding; Low Rad, All Elevations,
--
(August 19); and
RWP 6994, Remove / Replace M0V-SI-867C (August 20).
--
Conditions were found acceptable except as noted in subparagraph
5.d. below.
(3)
Plant Housekeeping
Plant housekeeping conditions including general cleanliness con-
ditions and control of material to prevent fire hazards were
observed in areas listed in paragraph 5.b.
Maintenance of fire
barriers, fire barrier penetrations, and verification of posted
,
fire watches in these areas was also observed.
.
No items of noncompliance were identified.
(4) Control Room Manning, Shift Turnover, and Log Review
Control Room manning was observed periodically during the inspec-
tion on daily vi. its during the normal work week, on backshifts,
and on weekends,.nd was confirmed to meet or exceed the require-
ments of TS and t 'e BVPS Operations Manual.
Shift turncvers at
various watch pos cions were routinely observed by the inspectors.
The inspectors raviewed shift logs on a daily basis, including
narrative and numerical logs maintained by the Shift Supervisor,
d
Nuclear Control Operator, Plant Operator, and Nuclear Operator.
Except as noted in paragraph 5.d. below, inspector comments or
questions resulting from these daily reviews were acceptably
resolved by licensee personnel.
(5)
Surveillance Testing
The inspectors reviewed completed surveillance tests to verify
that:
surveillance tests were being performed within acceptable
,
l
frequencies; test results were being reviewed according to
approved procedures; and, appropriate corrective actions were
initiated.
The following records of completed Operational Sur-
veillance Tests (OSTs) were reviewed:
OST.2.48.1, Mode 5 and 6 ESF Train Operability, Revision 1,
--
completed August 13, 1980;
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,
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OST 1.39.1, Weekly Station Battery Check, Revision 4, com-
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pleted August 4 and 18, 1980;
OST 1.7.8, Boric Acid Storage Tanks and RWST Level and
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Temperature Verification, Revision 10, completed August 7,
1980;
OST 1.11.3, Baron Injection Flow Path Valve Position
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Verification, Revision 20, completed July 21, 1980;
OST 1.11.10, Boron Injection Flow Path Power Operated Valve
--
Exercise, Revision 21, completed August 7 and 12, and
September 2,1980;
OST 1.16.1, Supplementary Leak Collection and Release Exhaust
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Fan and Remote Dac:per Component Test (Train A), Revision 3,
completed July 23, 1980;
OST 1.20.2, FC-P-1A Fuel Pool Pump Operability Test, Revi-
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sion 3, completed June 24 and July 2, 1980;
OST 1.7.3, Boric Acid Transfer Pump Operability Test, Revi-
--
sion 12, completed August 4,11, and 18,1980;
OST 1.7.6, Centrifugal Charging Pump Test (ICH-P-IC), Revi-
--
sion 14, completed August 6,1980;
OST 1.36.1, Diesel Generator No.1 Monthly Test, Revision 16,
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completed August 5 and September 2,1980;
.a
OST 1.36.7, Offsite to Onsite Power Distribution System
--
Breaker Alignment Verification, Revision 16, completed
August 2 and 16,1980;
OST 1.36.9, AC Power Source Breaker Alignment Verification
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During Shutdown, Revision 16, completed August 16, 1980;
OST 1.49.2, Shutdown Margin Calculation (Plant Shutdown)
--
(Upgraded for Cycle 2), Revision 8, completed September 3,
1980;
OST 1.15.2, Reactor Plant Component Cooling Water Pump
--
(ICCR-P-18) Monthly Test, Revision 9, completed September 2,
i
1980; and
OST 1.33.17, Portable Fire Pump Operational Test, Revision
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21, completed September 2,1980.-
I
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No items of nonccmpliance were identified.
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(6) Plant Security / Physical Protection
Implementation of the Physical Security Plan was observed during
inspection of the areas listed in paragraph 5.b for the following:
Protected and Vital Area barriers were not degraded;
--
Isolation Zones were clear;
--
.
Persons and packages were checked prior to entry into'the
--
Protected Area;
Vehicles were properly searched and vehicle access to the
--
Protected Area was in accordance with approved procedures;
Security access controls to Vital Areas were being main-
--
tained and the persons in Vital Areas were properly autho-
rized;.and,
The Protected Area was properly lighted as required by the
--
Physical Security Plan.
A temporary opening between the Unit 1 and Unit 2 Control Rooms
was constructed during the inspection period in order to deliver
and install security-related equipment for the Central Alarm
Station.
This opening provided an additional access point to
a Vital Area. On August 22, 1980 the inspector reviewed Security
Special Order #15, dated August 21, 1980 to determine whether
the compensatory measures initiated by the licensee during the
temporary degradation of the Vital Area boundary were acceptable.
<
In addition, the inspector verified that the personnel posted
in implementation of the Special Order were familiar with their
l
i
responsibilities and were equipped in accordance with the appli-
'
cable security requirements.
No items of noncompliance were identified.
(7) Temocrary Operating Procedures
The inspectors reviewed the following Temponary Operating Pro-
l
cedure (TOP) to establish that it had been properly reviewed,
l
approved, and issued; that the plant operations it directed were
!
in accordance with the requirements of the facility TS and QA
programs; and, that it was properly _ implemented and its perfor-
mance documented.
TOP 80-27, Fill of RWST from Baron Recovery Tanks. 'This
--
.
TOP provided instructions for transferring reactor coolant
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stored in the Boron Recovery Tanks (BR-TX-4A/B) to the
Refueling Waster Storage Tank (RWST) via Boric Acid Trans-
>
fer Pump (BR-P-28).
The procedure was reviewed and approved
by the OSC on August 15, 1980.
On August 20,1980 the
inspector observed the Nuclear Shift Operating Foreman
verifying valve lineups in accordance with the TOP.
Also
on that date, the i-' rectors walked down portions of the
flowpath establishea in accordance with the TOP.
Inspector
findings regarding that aspect of the review as discussed in
paragraph 5.d.
On August 21, 1980, the inspector noted during
'
review of the Shift Supervisor's logs that backleakage to the
Fuel Pool had taken place during the transfer because isola-
tion valves shut per the procedure were found to be leaking
by.
The inspector also noted that the Shift Supervisor had
terminated the evolution upon that discovery until an Opera-
tions Manual Change Notice No.80-114 had been submitted and
approved that provided for additional shut valves to prevent
further backleakage into the Fuel Pool.
No items of noncompliance were identified.
(8) Operational and Maintenance Administrative Controls
Equipment control procedures for controlling glant equipment
and activities were examined to verify that tags were properly
filled out, posted, and removed as required by approved 'proce-
dures, and that corrective actions were initiated where necessary.
The inspectors reviewed logs and records for completeness.
On August 20, 1980 the inspectors sampled Out of Service (00S)
d
stickers posted on equipment / annunciators in the Control Room for
verification as discussed above:
,
l
00S Sticker Number
Date Posted
EcuiDment Posted
Corrective Action
,
'
80085A
June 25, 1980
Annunciator A6-91
MWR 800867 submitted
80037A
March 14, 1980
Temperature Indica-MWR 800336 submitted
tor TI-RC-469
80079A
June 11, 1980
Flow Recorder Con-MWR 800790 submitted
troller FRC-LW-104
80034
March 5, 1980
Annunciator A2-8
MWR 800288 submitted
.
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.
.
.
.
The inspectors also sampled the following Caution Tags posted on
Control Room equipment:
Caution Tag Posted on
Oate Posted
Reason
Feedwater pump FW-P-3A
July 12, 1980
Breaker deenergized
Control Switch
Feedwater Pump FW-P-3B
July 14, 1980
Breaker deenergized
Control Switch
The inspectors observed that a Caution Tag for Reactor Contain-
ment Sump Pump, DA-P-48, dated August 8, 1980 was attached to a
bookcase next to the Caution tag logs. When questioned by the
inspectors, the Shift Super"isor stated that the tag had previ-
ously been posted on the appropriate Main Control Board control
swtich.
Earlier that morning the Shift Supervisor direct opera-
tors to start DA-P-48, and the tag was removed.
The supervisor
stated that the operator had been in the process of logging the
.
tag into the Caution tag log when interrupted.
The operator
'
stuck the tag to the bookcase and later failed to return to com-
plete the log entry.
The entry was completti prior to the com-
i
pletion of the inspector's tour.
No items of noncompliance were identified.
,
d.
Findings - Plant Operations
.
(1) Worker Contaminat _a Incident. September 9,1980
Summary
.
On September 9, 1980 a licensee contractor employee working in a
!
Controlled Area in the Reactor Containment Building found, after
l
exiting containrent, that he had been externally contaminated on
the skin of his face and neck.
The individual was innediately
,
I
decontaminated to background level.
l
Details
The inspector was infonned at 9:30 a.m., September 10, 1980 by
the acting Radeon Supervisor that a licensee contractor construc-
tion worker had exited containment 9:20 p.m. on September 9,1980
with external contamination on portions of his face and neck.
The
individual alerted Radcon personnel of the contamination after
performing a frisk at a secondary frisker station located inside
the Primary Auxiliary Building '3AB) on the 735 ft. elevation.
.
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)
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adjacent to the Radcon Field Office.
Initial surveys by Radia-
tion Control Technicians (RCTs) using an R0-2A survey instrument
indicated dose rates of 6 mrem / hour (gamma) and Ja mrad / hour (br ta)
e
from loose surface contamination located on the left side of the
individual's face and neck.
The worker was decontaminated imre-
diately to background levels by s
'-4'a and scrubbing with soap
and water.
The worker was a member of a seven man crew involved in removing
scaffolding from the L4 Steam Generator cubicle on the 718 ft.
elevation inside contairrent.
The scaffolding had been erected
in support of hanger modifications and painting.
Initial nasal
smears were obtained from all 7 individuals.
Results of three
smears indicated slightly greater than 100 cpm.
The acting Radcon Supervisor stated on September 10, 1980 that the
following actions would be taken:
whole body counts would be performed on the entire crew :
--
because of the positive nasal smears;
the contamination levels of remaining scaffolding would
--
be verified throoc!. review of recent survey results to
verify accept % 'e levels; otherdse, work crews removing
scaffoldina woi.id be required to wear respirators as a pre-
cautionary anasure;
personnel in these work areas would be monitored by means
--
of breathing zone air samplers;
d
scaffolding handled by the work crew involved in the con-
--
tamination event would be resurveyed;
the exit route taken by the contaminated worker would be
--
surveyed;
a comprehensive survey of the LA Steam Generator cubicle
--
would be undertaken;
an isotopic analysis would be performed on tne nasal smears;
--
existing documentation on the work activity would be
--
reviewed for discrepancies;
the dose to the worker would be calculated and recorded in
--
his permanent record; and,
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28-
calibration of all survey instruments involved in the event
--
would be verified.
.
On September 11, 1980 the inspector met with the Radcon Foremen
to discuss and review the results of the actions taken by the
licensee as specified above.
The results were as follows:
Whole body counts for all individuals indicated less than
--
minimum detectable concentrations;
resurvey of scaffolding handled by the work crew determined
--
maximum contamination levels present to be significantly
below those found on the worker;
isotopic analysis of nasal smears did not unable to identify
--
any specific isotope because no activity was detected in the
smears (the licensee was unable to reconcile this analysis
with the prior positive results);
survey of the exit path taken by the worker revealed no
--
contamination;
a comprehensive survey on the Steam Generator cubicle was
--
underway.
(Later discussions with Radcon personnel confirmed
that the survey had been completed and that no levels of con-
tination were identified that would have accounted for the
contamination on the worker);
.
the licensee was in the process of calculating the worker's
--
skin dose
review of documentation and interviews conducted by the
--
licensee with the work crew revealed the following discre-
pancies:
the worker stated that he had frisked upon exiting
--
containment and again at the secondary frisker station
on the 735 ft. elevation of the PAB.
The licensee was
concerned about the adequacy of his initial frisk. As
a result, the licensee prohibited the individual from
returning to Controlled Areas pending conclusion of the
licensee's investigation into this matter.
In addition
it was noted that tne contaminated individual had failed
to sign off the Radiation Work Permit.
It was later
determined that this occurred as a result of the RCT's
actions to survey and decontaminate the individual, tnus
interrupting the normal sign-off sequence.
.
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examination of all survey instruments involved in the event
--
determined that all instruments were calibrated and operable.
Subsequent discussions among the inspector, the Plant Superin-
tendent, and the Radcon Supervisor involved the shared concern
that the incident may have resulted from a breakdown of one or
more Radcon administrative controls. Although preliminary eva-
luation of administrative controls for this incident was incon-
clusive, the Raccon Supervisor stated that he intended to evaluate
the information surrounding the event with the intent of iden-
tfying potential deficiencies in the existing program because
of questions regarding the adequacy of the worker's initial frisk.
In response to a NRC:RI request, the Radcon Supervisor agreed
to provide NRC:RI with the results of the dose assessment on
the contaminated worker.
Results of the NRC:RI review of the
OLC skin dose assessment and review of actions taken by the
licensee as a result of the Radcon Supervisor's evaluation of
the incident will be documented in a future inspection report
(80-25-01).
(2) Use of Temporary Hoses for Radioactive Liquid Transfers
During a backshift tour of the 722 ft. elevation of the PAB on
August 20, 1980 the inspectors walked down portions of a tempo-
rary hose installation associated with TOP 80-27, Fill of RWST
from Baron Recovery Tanks.
The procedure required the use of
the hose for the transfer of radioactive liquid from the Coolant
RecoveryTanks (BR-TK-4A/B) ~ via the Boric Acid Transfer Pump
(BR-P-28) located on the 722 ft. elevation of the PAB to the
fill header for the RWST located on the 735 ft elevation of Fuel
Handling Building.
The inspectors made the following observations:
The hose employed was an air hose laid along the floor
--
,
I
of the PAB in normal traffic. areas.
The hose ran up a
stairwell from the 722 ft. elevation of the PAB through
two doors on the 735 ft. elevation leading to the Fdel
Handling Building.
The hose was not tied or otherwise
supported to prevent damage by its hanging weight or acci-
dental damage from area traffic.
Slight chafing was apparent on the hose under the 722 ft.
--
elevation door which was periodically used by personnel
entering and exiting the 722 ft. elevation.
Both doors were ajar because of the hose routed through them.
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Liquid was observed on the floor under the quick-disconnect
--
fittings used on the hose sections.
The inspectors noted
the liquid on the 722 ft. elevation PA8 floor and in the
s tai rwell. The fittings were not bagged nor fitted with
a radiological containment or drip pocket.
The hose was not posted to indicate that it contained radio-
--
active materials.
The inspectors also observed a security watchpe.rson stationed
by the open fire doors to the Fuel Handling Building through
which the hose passed.
The inspectors confirmed that the watch-
person was aware of fire watch actions required of him in the
event of fire.
The inspectors' observations were immediately discussed with
the Shift Supervisor and the shift Radcon Foreman.
The Radcon
Foreman stated that the liquid on the floor had been sampled
and was found to be condensate dripping from adjacent piping.
The Shift Supervisor informed the inspectors that the transfer
would require at least 5 days based on the capacity of the tempo-
rary hose.
He also stated that the maximum system pressure would
be about 40 psig; the air hose was rated for 150 psig.
Discussions with the Radcon Foreman later that day provided the
following additional information:
the 722 ft. elevation stairwell door had been secured to
--
a
preclude any additional chafing; the upper door would remain
open with the watchperson posted;
the hose had been posted and hose joints bagged.
--
Verification on August 21, 1980 by the inspectors of the post-
'
ing determined that only the stairwell portion of the hose had
been posted.
Discussion with the Radcon Foreman who was present
at the time determined that he had misguided the RCT whom he had
l
directed to post the hose when he had directed him to post the
hose "in the stairwell".
That the remaining section of the hose
was posted was later observed by the inspectors during a routine
PAB tour.
Inspector review of the Radcon Mandal determined that no guiuance
was provided for: conditions under which hoses would be acce.t-
able for radioactive liquid transfers; acceptable types of hoses;
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required hose certifications; Radcon surveillance coverage for
such transfers; and routing / posting considerations.
These
inspector concerns regarding the lack of any such guidance were
initially discussed with the Acting Radcon Supervisor on August 20,
1980.
On August 28, 1980 the inspector examined OSC meeting minutes
which documented review of TOP 80-27 and noted that the OSC's
review had been performed by means of a Poll of Members of the
OSC, No.1399, and approved by the OSC Chairman on August 15,
1980.
The inspectors noted that no member of the Radcon staff
.
had been among those polled.
The inspector asked the Radcon
Supervisor whether failure to include Radcon participation in
the review of a temporary liquid waste transfer procedure still
provided for adequate consideration of possible Radcon concerns.
The Radcon Supervisor stated that other OSC members polled had
had adequate prior Radcon experience.
At the exit interview conducted September 15, 1980 the inspectors
reiterated their concern regarding lack of formal guidance and
requirements for use of temporary hoses for liquid transfers and
the adequacy of the OSC's review in light of the potential for
unreviewed safety questions.
The Plant Superintendent comitted
to assuring that appropriate guidance and requirements for tempo-
rary liquid transfer systems would be developed and incorporated
into the appropriate department manuals by December 31, 1980 with
reference to that guidance placed in other appropriate department
manuals.
In the interim, the Plant Superintendent comitted to
assuring that all procedures for use of temporary liquid transfer
systems would be reviewed by a qualified member of the Radeon
.*
department prior to OSC approval. This matter is unresolved pend-
ing completion of licensee actions (80-25-02).
(3) Posting Temocrary Controlled Areas
During a backshift tour of the Protected Area on September 11,
1980 the inspector observed that posting for a temporary radiol-
ogically Controlled Area located between the Refueling Water
Storage Tank (QS-TK-1) and the Safeguards Building had been
covered or removed and was no longer observable by individuals
entering the area. The inspector informed the Shift Radcon Fore-
man who took immediate corrective actions to repost the area. A
precautionary survey of the Controlled Area was performed by
Radcon personnel; no levels of radiation or contamination back-
ground count rates were observed.
.
.
,
,
On September 12, 1980 during another Protected Area tour, the
inspector observed that the same posting had been removed and
that licensee contractor personnel were conducting work in the
vicinity of the RWST.
The inspector imediately informed the
Radcon Supervisor and Foreman.
The Radeon Foreman accompanied
that inspector to the area and ascertained that the work had
involved lifting concrete blocks in the vicinity of the RWST.
The posting was replaced at that time using chain rather than
rope.
In order to determine whether surveys had been performed to
-
support the existing posting, or whether additional posting was
required, on September 12, 1980 the inspector reviewed survey
data (on a sampling basis) for the temporary Controlled Area
discussed above.
The following survey data was reviewed by
the inspector:
QS-TK-1 (RWST) Walkway, Trailer, and Entrance to Safe-
--
guards; August 14, ".980;
Survey Canvas Outside Safeguards Trailer; September 3,1981;
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QS-TK-1 Cage area and Pit; September 6,1980;
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Cleanup of Manway Outside QS-TK-1; September 12,1980; and
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Prework/ Hanger Inspection /QS-TK-1 Walkway; September 12,
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1980.
The inspector found that the licensee's existing posting was
,
acceptable.
!
During the Exit Interview on September 15, 1980 the Radcon
l
Supervisor stated that a permanent barrier to better control
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access was to be erected between the RWST and the Safeguards
I
building wall. The Plant Superintendent comitted to posting
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temporary Controlled Areas during restart activities with
chains, rather than rope, to minimize deterioration / destruction.
The inspectors stated that observation of licensee posting would
continue to be observed during routine inspections.
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(4) Missing Calibration Sticker on personnel Survey Instrument
l
(FRISKER)
On August 21, 1980, while exiting the Controlled Area through
a primary frisker station located at an exit' portal in the Ser-
vice Building Locker Room, the inspectors observed that one of
the survey instruments (Friskers) provided for personnel moni-
toring had no calibration sticker.
The remaining instruments
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were examined by the inspectors and verified to have calibration
stickers indicating acceptable calibrations.
Since personnel are required to perform whole body frisks prior
to exiting the Controlled Area, the inspectors immediately
informed the Radeon Foreman of their observation and questioned
whether frisks had been performed using 'an uncalibrated instru-
ment. The Radcon Foreman directed that the frisker (RM-4 #1062)
be immediately removed from the frisker station and that a source
check be performed to determine its accuracy.
The instrument was
found to be in calibration. After the licensee reviewed calibra-
tion data to verify the status of the frisker, a new calibration
sticker was applied and the frisker returned to service. A
licensee technician later performed spot checks to verify that
the sticker remained in place.
Inspector review on August 22, 1980 of instrurent calibration
records confirmed that the instrument had been included in a
scheduled calibration program and had been most recently cali-
brated on August 15, 1980. A new calibration sticker had been
applied at that time.
In addition, the inspector reviewed a DLC
informal memorandum, dated August 21. 1980, in which the Radcon
Foreman informed the Radcon Supervisor that the adhesive quality
of the new metal calibration stickers currently in use was mar-
ginal, and that, in the future, transparent contact paper would
be applied over this type of calibration sticker to prevent the
sticker from being removed or falling off.
The inspectors had no further questions in this matter.
6.
In Office Review of Licensee Event Reports (LER's)
d
The inspector reviewed LER's submitted to the NRC:RI office to verify
that the details of the event were clearly reported, including the accuracy
of the description of cause and adequacy of corrective action.
The inspec-
tar 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 Number Event Date
Report Date
Subject
- 80-047/03L August 5,1980
September 4,1980
No. Failure to Close on
Emergency Bus
80-059/01P August 23, 1980
August 25, 1980
Diesel Fire Pump Starter Motor
Failure
- 80-060/03L August 27, 1980
August 28, 1980
Potential Charging Pump Damage
Due to Low Flow
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LER Number Event Date
Recort Date
Subiect
- 80-C61/03L August 29, 1980
September 26, 1930 Loss of BA Flowpath Operability
80-062/01P September 8, 1980
Septemcer 8, 1980
Seismic Overstress on BIT
Outlet Line
80-063/01P September 8, 1980
September 8,1980
Seismic Overstress on IC
Charging Pump Discharge Line
80-064/03L August 25, 1980
September 10, 1980 Missed Surveillance on RCS
Chemistry
80-065/01P Septemcer 12, 1980 September 15, 1980 Seismic Overstress on Reactor
Plant River Water Lines
80-066/01P September 12, 1980 September 15, 1980 Seismic Overstress on IA Steam
Generator Outlet Line
80-067/01P September 12, 1980 September 15, 1980 Seismic Overstress on Component
Cooling Line to RHR
~
- Reports selected for onsite followup.
7.
Cnsite Licensee Event Folicwuo
For those LER's selected for onsite followup (denoted by asterisks in
paragraph 6), the inspectors verified that the reporting requirements of
the Technical Specifications, and SAD 14 and 23 or N00 No.10 and SAP
d
Chapter 3B and 4 (as applicable) had been met; that appropriate correc-
tive action had been taken or planned; that the event was reviewed by
the licensee as required by Technical Specifications, and SAD 21 or SAP
Chapter 10 (as applicable); and, that the 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 following findings relate.to the LERs reviewed onsite.
a.
LER 80-47:
Emergency Diesel Generator (EDG) No.1 Failure to Close
on Emergency Bus
At 11:15 a.m. on August 5,1980, the ins;3ctor was notified by the
Shift Supervisor that No.1 EDG output breaker had failed to close
on repeated attemots made during a monthly surveillance test and
had been declared inoperable at 9:50 a.m.
The No. 2 EDG was simul-
taneously inoperable because of fuel transfer pump and lube oil
cooler problems.
The Shift Supervisor stated that the Action State-
'
ments of TS 3.8.1.2 and 3.8.2.2 were being implemented, requiring
suspension of core alterations or positive reactivity changes and
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establishment of containment integrity within eight hours.
Control
room observations and discussions with control room personnel con-
firmed that the required actions were being implemented.
The inspec-
tor additionally confirmed that two offsite power sources and a
baration flowpath were available, as required by TS.
At 2 p.m. the Shift supervisor informed the inspector that the cause
of failure had been isolateo to a contact block in the EDG manual
start relay circuits which are used only in the test or " exercise"
mode of operation. Therefore, the faulty circuit did not affect the
capability of automatic starting and. loading upon receipt of a valid
safety injection or loss of power signal.
The Shift Supervisor also
stated that all operators onshift had been reinstructed regarding
actions to be taken in the event that the EDG failed to start on
receipt of the a valid automatic start and load signal. The inspec-
tar later confirmed, through discussions with the OCL Senior Technical
Engineer, that a relay contact was found to be dirty and was replaced.
The diesel was satisfactorily tested and returned to service, by
1:22 p.m. on the same day,
b.
LER 80-60:
potential Charging Pumo Damage Oue to Low Flow
A potential for damage to one or more centrifugal charging pumps follow-
ing a secondary side high energy line break was reported to NRC on May 8,
1980 by Westinghouse Electric Corporation pursuant to 10 ~CFR 21, Report-
ing of Defects and Noncompliance. On July 24, 1980 NRC issued IE Bul-
letin 80-18, Maintenance of Adequate Minimum Flow Thru Centrifugal Charg-
ing Pumps Following Secondary High Energy Line Rupture.
Licensees were
asked to perform plant-specific calculations specified in the enclosed
Westinghouse Part 21 report to detennine whether adequate minimum flow
d
was assured under all conditions.
If adequate flow was not assured,
the IES asked licensees to made inodifications (such as those suggested
in the Westinghouse 1,0 CFR 21 report) to equipment and/or procedures
to insure availability of adequate minimum flow under all conditions,
and to justify any such actions.
Tne licensee had also received a Westinghouse letter dated July 2,
1980 that paralleled the 10 CFR 21 report, discussed the potential for
charging pump damage, and provided evaluation and corrective action
gufdance.
The recomended analyses were initiated by the licensee to-
determine whether the concern was applicable to BVPS-l'.
On August 28,
1980 the referenced LER was submitted identifying this potential con-
dition as applicable to the licensee's facility.
Since information
required in the LER was discussed in the IEB or required in the licens-
ee's IES response, the licensee stated that the LER was primarily sub-
mitted for records purposes to satisfy Technical Specification report-
ing requirements.
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In the LER, the licensee stated that a calculative, plant-specific
evaluation had been performed using the Westinghouse guidelines and
'ne pcssibility for the event existed at Beaver Valley.
Specifically,
.
following a secondary side high energy line rupture and subsequent SI
initiation, the charging pump recirculation (miniflow) line isolation
valves automatically close.
If the pressurizer power-oriented relief
valves (PORVs) were not operable due to loss of offsite power or to an
adverse environment in containment; or if the PORV was in manual code;
or if the PORV block valve was in a closed position due to PORV leakage,
then the Reactor Coolant System pressure would increase because of
injection flow and core decay heat generation until it reached the
setpoint of the pressurizer code safety valves.
At that high back-
pressure, the charging pumps might & t be able to supply sufficient
flow to prevent damaging the pumps.
The stated corrective action involved removing the SI valve closure
signal from the charging pump miniflow isolation valves and to proce-
durally control applicable valve manipulations to ensure that the charg-
ing pumps remain adequately protected.
-
On September 10, 1980 the inspector conducted discussions with licensee
personnel and reviewed records in order to ascertain whether:
corrective actions stated in the LER were appropriate to correct
--
the cause of the event;
responsibility had been assigned for assuring completion of cor-
--
rective actions;
the LER accurately descrl bed the actual event;
--
.4
the reported cause is accurate for the LER and reflected the
--
proper cause code;
the report satisfied the TS reporting requirements with respect
--
to information provided and timing of submittal;
the event was reviewed and evaluated as required by approved
--
procedures and administrative controls; and
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personnel within the licensee organization were notified of the
--
event as required by TS, license conditions, or approved procedures.
The inspector reviewed the calculations entitled " Minimum Centrifugal
Charging Pump Flow During Two Pump Parallel Safety Injection Operation"
and confirmed that the condition discussed above was applicable to the
,
!
licensee's facility.
.
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The inspector ascertained through review of Onsite Safety Committee
meeting Minutes No. BV-0SC-98-80, dated August 12, 1980, that the
licensee was implementing the corrective actions specified in IEB 80-18, Westinghouse 10 CFR 21 reported dated May 8,1980, and the
Westinghouse letter dated July 2,1980.
The inspector verified that
appropriate department heads were aware of the problem and had been
directed to initiate the corrective actions specified in the IEB and
Westinghouse guidance, and that responsibility was assigned for con-
sideration of long tenn corrective actions through development of a
Station Modification Request.
Based on comparison of the subject LER with IEB 80-18 and the Westing-
house guidance, the inspector ascertained that the report accurately
described the event.
In addition, the reported cause was determined
to be accurate and the proper cause code reflected.
The report was
submitted August 28, 1980 in accordance with TS 6.9.1.9.c.
Verification of the completion and adequacy of the corrective actions
-
-
taken or to be taken by the licensee will be examined in #uture inspec-
tions during NRC review of DLC response to IEB 80-18. At the close of
this inspection period, the licensee's response had not been submitted.
c.
LER 80-61: Loss of Boric Acid Flowcath Goerability
The licensee did not maintainan operable boric acid flowpath required
by TS 3.1.2.1.a, Boration System Flowpaths-Shutdown, for approximaiely
16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> on August 28 and 29,1980 because of operator error.
On August 26 the licensee was refilling the Refueling Water Storage
Tank with borated water; the tank had been emptied earlier in the
current outage for modifications.
In order to assure that a TS
operable borated water source was available, the licensee isolated
the 1A Soric Acid Storage Tank to maintain its TS required level and
concentration.
The operator then aligned the 2A Boric Acid pump to
the IB Boric Acid Storage Tank and continued batching borated water
to the RWST via the 2A pump and the IB tank. On August 28 during the
performance of a surveillance test on the IB tank and pump, the opera-
tor terminated the cross-connect between the 1A pump and the 18 tank
when reestablishing Normal System Alignment per the test procedure return
to normal steps but failed to realign the isolated "1A" tank and pump
to an operable conclition.
(The IB train was not considered operable
because its associated emergency diesel generator was out of service.)
This misalignment was identified on August 29 when an operator attempted
to fill the Volume Control Tank using the 1A tank and pump but was unable
to es.ablish flow.
The shut manual outlet isolation valve to the 1A
was discovered, innediately reopened, and the surveillance test on the
"A" train flowpath performed to verify operability.
.
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.
.
The inspector became aware of these event during review of Shift
Supervisor logs on August 29, 1980. As a result of discussions with
the Station Operations Supervisor, Shift Supervisors, and Nuclear
Control Operators, and review of Shift Supervisor and Nuclear Control
Operator logs, the inspector assembled the following sequence of
events:
August 26, 1980
Morning
Operations personnel conducted boric
acid blending operations to the RWST
with Boric Acid Transfer Pump (CH-P-2A)
aligned to Boric Acid Storage Tank (CH-
TX-1A) according to procedures in BVPS
CM Chapter /, Chemical & Volume Control
System.
The Shift Supervisor and the
Station Operations Supervisor discussed
the possibility of losing the TS required
boric acid conceitration or water level
in the L; tar'. as a result of the blend-
ing operation.
The decision was made to
isolate the 1A tank in order to meet TS
requirements and continue batching
operations using the IB tank.
The Shift
Supervisor verbally directed the opera-
tors to realign the system.
12:58 p.m.
Realignment from the 1A to IB Boric Acid
Storage Tank took place in accordance
with verbal direction from shift supervi-
sor.
The L4 tank manual outlet isolation
valve (CH-71) to the 1A tank was shut.
The status board in the Control Rocm was
d
appropriately marked and a general entry
made in the operator's logs regarding the
new lineup.
Existing system operating
procedures did not cddress this specific
alignment.
The log entry and status
board markup were the only administrative
controls employed to track system status.
.
An operable beric acid flowpath via the
l
crossconnected tank and pump still existed.
Inspector review of OST 1.7.8, BA Storage
Tank and RWST Level and Temperature
Verification, Revision 10, completed
August 2S,1980 confirmed that the IB
tank was operable when credit was taken
far the cross-connected lineup as the
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operable flowpath.
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August 27, 1980
Refilling RWST continued.
'
August 28, 1980
1:42 p.m.
Operator performed surveillance test,
OST 1.7.2, Boric Acid Transfer Pump
(ICH-P-28) Operational Test, Revision 12,
and aligned the IB tank and pump to
perform the test.
Upon completion of
the OST, the lineup was left in the OST
directed lineup which was a recirculation
ficwpath between the 18 tank and pump.,
At that point:
no operable flowpath existed on the
--
A train since the 1A Boric Acid Stor-
age Tank manual isolation valve (CH-71)
was closed;
no operable flowpath existed on the B
--
train since its associated emergenc'y
diesel generator was out of service;
and
no cross-connected operable flowpath
--
using the 1A pump and the IB tank
existed since the operator.had termi-
nated the cross-connect during perfor-
mance of the OST.
August 29, 1980
5:35 a.m.
While attempting to add water to the Volume
control Tank using the A pump, the operator,
reported that he could not establish flow
from the 1A tank. Another operator was
sent out to check the lineup and found the
manual isolation valve (CH-71) shut. He
insnediately opened the valve and the sur -
veillance test on the A tank and pump was
performed to verify operability.
Inspector review of the LER submitted by the licensee'on September 26
l
after the close of the inspection period detennined that the infonnation~
supplied by the licensee was consistent with that obtained by the
inspector during the event.
The inspector had no further questions
regarding the LER.
As a result of this event and previous discussions with the inspectors
i
regarding procedure adherence, the Station Operations Supervisor issued
Special Operating Order No. 80-08, Procedure Compliance, dated September 8,
'
1980 which specified safety-related activities or procedures requiring-
strict procedure compliance.
Requirements for changes to procedures
.
.
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,
.
.
were discussed and clarification of Normal System Arrangement specified
under the Initial Conditions sections of operating surveillance proce-
dures was provided.
In addition, the S00 required that deviations frem
NSA or Initial Conditions be logged in the Operators log of surveillance
procedures and that the equipment be caution tagged.
Implementation of
the Special Operating Order will be reviewed by the inspectors during
future inspections (80-25-03).
.
8.
Mcation of Load Centers
.
D'uring an RCP seal leak event at Arkansas Nuclear, Unit 1, on May 10,1980,
it became necessary to eriter the containment to unlock and close breakers
which provided power to the isolaticn valves of the core flood tanks.
This
entry entailed exposure of personnel to adverse environmental and radiologi-
cal conditions in containment resulting from a primary system leak. The
resident inspectors reviewed BVPS-1 designs and installations to assure that
all relevant load centers and/or circuit breakers were located in normally
accessible areas and in conformance with an NRR Technical position which
states:
"Any valve which is required by TS to be locked in a particular
position during operation and requires entry into the containment
to actually unlock the valve-locking capability (such as'a locked
open circuit breaker) should have its locking capability located
outside the containment where access can be provided."
The Resident Inspector conducted discussions with the Operations Supervisor
and several Shift Supervisors regarding locations of load centers and operat-
ing, abnonnal, and emergency procedures required for shutdown.
In addition,
'
the inspector reviewed the following sections of the BVPS Operations Manual:
,
Chapter 36, 4 KV Station Service System
--
Chapter 37, 480 V Station Service System
--
Chapter 38,120 VAC Distribution and Lighting; and
--
Chapter 39,125 VDC Control System.
--
Independent verification of selected load center locations was perfonned
by the inspector during routine containment and auxiliary building tours.
The inspector determined that relevant load centers and circuit breakers
at Beaver Valley Unit I were located in conformance with the referenced
t
NRR Technical Position. No load centers were observed to be located inside
containment.
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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 paragraph 5
of this report.
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 also provided to the licensee at the con-
clusion of the report period.
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