ML20054C547
| ML20054C547 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 03/18/1982 |
| From: | Beckman D, Greenman E, Lazarus W, Reynolds S, Troskoski W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20054C544 | List: |
| References | |
| 50-334-82-01, 50-334-82-1, NUDOCS 8204210307 | |
| Download: ML20054C547 (54) | |
See also: IR 05000334/1982001
Text
THE INFORMATION ON THIS PAGE IS
DCS Nos. 50-334 820114
820102
DEEMED TO BE APPROPRIATE FOR PUBLIC
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820114
320106
DISCLOSURE PURSUANT TO 10 CFR 73.21.
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820128
1
U.S. NUCLEAR REGULATORY COMMISSION
or' v,E OF INSPECTION AND ENFORCEMENT
Region I
Report No.
50-334/82-01
Docket No.
50-334
License No.
Priority
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Category
C
Licensee:
Duquesne Light Company
435 Sixth Avenue
The report details contain
Safeguards Info
Pittsburgh, Pennsylvania
(Pages 49-51)
Facility Name: Beaver Valley Power Station, Unit 1
Inspection at: Shippingport, Pennsylvania
Inspection cond c d: J uary 4 - February 16, 1982
3,7 87e
Inspectors:
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D AnB kmal, Senior Res' dent Inspector
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W. M.sTroskoski, Resident) Inspector
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W.'Jr1.azarus Reactor Inspector
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S. D. R ynolds, Rejetor Inspector
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Approved by:
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jvE. G. Greenman, Chief, Reactor Projects
date signed
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Section No. 2A, Projects Branch No. 2
Inspection Summary:
Inspection on January 4-February 16,1982 (Report No. 50-334/82-01).
Areas Inspected: Routine inspections by the resident inspectors (191 hours0.00221 days <br />0.0531 hours <br />3.158069e-4 weeks <br />7.26755e-5 months <br />) and two
region-based inspectors (26 hours3.009259e-4 days <br />0.00722 hours <br />4.298942e-5 weeks <br />9.893e-6 months <br />) of licensee action on previous inspection findings,
followup on NRC Performance Appraisal Section findings, plant operations, housekeeping,
fire protection, radiological controls, :,urveillance testing, maintenance, physical secur-
ity, radwaste system operation, in-office review of licensee event reports, onsite event
followup, refueling preparations, IE Bulletin followup, TMI lessons learned followup,
EPP drill observations, potential design deficiency review, piping NDE review,, and ar
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devitalization.
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Results:
Violations:
None in seventeen areas.
Seven in two areas (Failure to
post fire watches for nonfunctional penetration, paragraph 3.e(3); Failure to
administer / document maintenance training, paragraph 2.b; ORC failed to review
violations required by TS, paragraph 2.b; Failure to establish and execute inspection
program for operating activities, paragraph 2.b; Failure to document bases for
10CFR50.59(b) safety evaluations, paragraph 2.b; QA Audit deficiencies, para
and, Inadequate ORC / Management Audit and ORC Training Audit, paragraph 2.b.) graph
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DETAILS
1.
Persons Contacted
D. Beron, Warehouse Supervisor, Stores Dept.
F. Bissert, Manager, Nuclear Support Services
J. Carey, Vice President, Nuclear Division
K. Grada, Superintendent of Licensing and Compliance
R. Hansen, Maintenance Supervisor
H. Harper, Security Assistant
T. Jones, Manager, Nuclear Operations
J. Kosmal, Radcon Supervisor
W. Lacey, Chief Engineer
V. Linnenbom, Radiochemist
J. Lukehart, Security Director
J. McGee, Director, Administrative Services
L. Schad, Operations Supervisor
J. Sieber, Manager, Nuclear Safety and Licensing
J. Vassello, Training Supervisor
H. Williams, Station Superintendent
J. Wenkous, Reactor Control Chemist
The inspectors also contacted other licensee employees and contractors
during this inspection.
2.
Licensee Action on Previously Identifiet Inspection Findings
a.
The NRC Outstanding Items (01) List was reviewed with responsible
licensee personnel.
Items selected by the inspectors were sub-
sequently reviewed through discussions with licensee personnel,
documentation review, and field inspection to determine whether
licensee actions specified in the OIs had been satisfactorily
completed. The overall status of previously identified inspection
findings was reviewed, and planned and completed licensee actions
were discussed for those items not reported below.
(Closed) Unresolved Item (79-12-02): Type C penetration ncn-
conscrvative test results.
In several Type C tests performed
using OST 1.47.4, Containment Isolation Valve Leakage Test, Type
C, the licensee recorded the leakage as zero, which is non-
conservative in that: 1) the rotameters used to measure leakage
are only accurate above a given minimum sensitivity, and 2)
the accuracy of leakage measurement using the " downstream
method" is not assured because no verification is done for leak
tightness of downstream boundries. The licensee comited to
revising the procedure to specify the use of the tests minimum
sensitivity when appropriate and to specify that additional veri-
fications be made when using the " downstream method."
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The inspector reviewed OST 1.47.4, Containment Isolation Valve
Leakage Test, Type C, Revision 39. This procedure includes a
note requiring that, if no indication of flow can be obtained
using the most sensitive rotameter then the minimum sensitivity
for the flow leakage shall be assigned as 0.01 SCFH, which
corresponds to the minimum sensitivity for the most accurate
rotameter used. Another note also specified that the make-up
air test method is to be used instead of the downstream method
for verifying test boundry leakage. The downstream method was
outlined and available for trouble-shooting only. The inspector
had no further questions at this time.
(Closed) UnresolvedItem(79-12-04): Type B containment leakage
test - nonconservative results. MSP 47.01, Type B Containment
Leakage Test - Electrical Penetrations, was detemined to be
inaccurate when computing leak rate by the Pressure Decay Method
in that temperature variations over the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> test period were
not measured nor compensated for in the calculations.
In addition,
pressure gauges used in the test were not calibrated. The inspector
reviewed Revision 2 to MSP 47.01 and determined that the use of
both calibrated test gauges and thermometers were specified for
conducting the test. The inspector also verified that the tempera-
ture compensation was included in leak rate calculations. The
inspector had no further questions on this item.
(Closed) Unresolved'cem(79-12-06): Airlock door bypass leakage.
During performance of the last containment integrated leak rate
test (CILRT), the airlock innet door equalizing solenoid valves
leaked by even though they had satisfactorily passed their local
leak rate test (LLRT).
In order to maintain containment integrity,
a manual isolation valve in the line 1-VS-153 must be kept closed.
The licensee comited to include this valve in the Type C LLRT
program and to incorporate checks of this valve position into a
periodic containment integrity verification.
The inspector reviewed surveillance test OST 1.47.1, Containment
Air Lock Test, Revision 22, perfomed on a 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> basis to verify
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that detectable seal leakage is within containment integrity limits.
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To limit any leakage from the inner door equalizing solenoids, the
OST requires valve 1-VS-153 to be closed.
The inspector also re-
viewed surveillance test OST 1.47.30, Personnel Air Lock Isolation
Valve (1VS-153)TypeCLeakTest, Revision 42,andverifiedvalve
testing per the licensee's Type C leak rate test program. The
inspector had no further questions on this item.
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(Closed) UnresolvedItem(79-12-07):
Electrical penetration
surveillance.
During a field inspection of electrical penetra-
tions, the inspector noted that all pressure indicating gauges
were isolated and not indicating the status of the penetration
canisters o r 0-ring seals. When several of the isolation valves
were opened, two were found to be at zero pressure and two others
at 15 psig. Although these canisters are nomally pressurized
at or above 45 psig, no routine surveillance was performed to
verify proper pressurization. The licensee had commited to
issuing a routine surveillance to monitor the pressure.
TheinspectorreviewedMaintenanceSurveillanceProcedure(MSP)
47.01, Type B Containment Leakage Test - Electrical Penetrations,
Revision 2.
The MSP provides detection of local leakage from the
containment electrical penetrations by periodically testing their
pressure integrity and comparing test data to the overall contain-
ment integrated leakage rate acceptance criteria specified in
TS 4.6.1.2, Containment Leakage - Surveillance Requirements.
The MSP is perfomed during each refueling outage to meet the
requirement of 10CFR50, Appendix 1, Primary Reactor Containment
Leakage Testing for Water-Cooled Power Reactors. The inspector
had no further questions on this item.
(Closed) Unresolved Item (79-22-01): Licensee to evaluate cali-
bration program for MCB PR-NI. During a review of the Main Control
ector noted that the MCB
Board (MCB) instrumentation,theinsp(NI)meterswerereadingup
power range (PR) nuclear instrument
to 6% full power greater than the NI rack indicators. This
difference is attributed to the MCB indicators being driven by an
isolation amplifier that was not routinely recalibrated during
daily NI calorimetric calibrations. Calorimetric calibrations
result in adjustment of the circuit's principal summing and level
amplifier which provides signals to the NI rack indicators and
all protection functions. The meters' isolation amplifier receives
its signal from the summing and level amplifier and will deviate
from the NI rack indication if it is not adjusted after a summing
amplifier gain adjustment.
The inspector reviewed MSP 2.04, Power Range Neutron Flux Channel
(N-NI-42) Quarterly Calibration, Revision 13: This revision
includes adjusting the isolation amplifier signal to the main
control board instrumentation whenever a deviation greater than
plus or minus 1.5% exists between the NI rack indicators and MCB
meters during calorimetric calibration. The inspector had no
further questions on this item.
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(Closed) Unresolved Item (80-16-02):
Review licensee action
for NRC identified gouges on cold leg safety injection line.
During a tour of the containment building on June 4,1980, the
inspector noted several gouges on a 6 inch cold leg safety
injection line adjacent to vent valve SI-330. This was brought
to the attention of the QC Supervisor for evaluation and corrective
action. A NSQC General Inspection Report, dated June 25, 1980
identified the line as SI-73-1502-Ql, a 6 inch schedule 160 pipe
and reported a dial depth micrometer (No. 644, calibrated January
1,1980) measurement of gouge depth as a maximum wall penetration
of 0.011 inches. The condition was accepted as-is pending a
final Station Engineering evaluation of minimum wall thickness
acceptance criteria.
Engineering Memorandum (EM) 30086, approved December 21, 1981
compared the as-found wall thickness to Power Piping Code B.31.1
requirements applicable to BVPS-1.
Nominal wall thickness for
6 inch, schedule 160 pipe is 0.719 inches, minus 12.5% for
conservatism tolerance equals a minimum expected wall thickness
of 0.629 inches for standard pipe.
Subtracting the 0.011 inch
maximum 9 mge depth yields a wall thickness of 0.618 inches.
This exceeded the B31.1 acceptance criteria of 0.484 inches.
The inspector noted that the request for evaluation of minimum
wall thickness acceptance criteria was not forwarded to engineering
via EM 30086 until December 15, 1981.
b.
Followup on NRC Perfomance Appraisal
On October 19-30 and November 12-20, the NRC Performance Appraisal
Section (PAS) conducted a Performance Appraisal Inspection (50-334/
81-29) at Beaver Valley, Unit 1.
The following items were identified
as possible violations of NRC requirements and were reviewed by
the resident inspectors to determine the need for additional
enforcement action.
(Closed) Unresolved Item (81-29-02): ORC failed to review certain
documents containing violations of license requirements per TS 6.5.2.7.e.
That TS requires the Offsite Review Coonittee (ORC) to
review violations of applicable statutes, codes, regulations, orders,
TS, license requirements, or of internal procedures or instructions
having nuclear safety significance.
The PAS found that the ORC
did not review Nonconformance and Action Reports, incident reports,
and QA Surveillance Reports, all of which contain examples of such
violations. The PAS further found that the CRC may have relied
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upon Onsite Safety Committee minutes review to meet this require-
ment but that the information contained in the OSC minutes was
inadequate to pemit a thorough review.
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The inspector discussed this matter with the DLC Manager,
Nuclear Safety and Licensing, (ORC Chaiman) and reviewed
the following:
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ORCMinutes, Meeting #80(June 27,1980) through #101
December 22,1981).
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Operations Quality Control Nonconfomance and Corrective
Action Reports (NCARs), #234 (March 4, 1980) through-
- 275(December 3,1981).
BVPS Licensee Event Reports (LERs) 80-01 through 81-100.
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BVPS Incident Reports (not issued as Licensee Event Reports)
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Nos. 81-10 (January 12,1981) through 81-52 (April 2,1981).
Quality Assurance Department Outage Surveillance Reports,
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1980 Outage. All reports reviewed (about 250).
The inspector found that the ORC meeting minutes only documented
review of NRC inspection findings and DLC responses, DLC QA Audit
findings, prompt (24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />) Licensee Event Reports, several 30 day
Licensee Event Reports, and the running status of NRC and QA
audit corrective actions. No NCARs, Incident Reports, or Sur-
veillance Reports were listed. Neither were onsite contractor
nonconformance reports. The ORC Chairman advised the inspector
that, although individual ORC members receive and review such
documents as part of their non-comittee duties, no fomal system
for referral to the ORC was used. The chairman also stated that
such items were occasionally discussed in committee but apparently
were not documented in the minutes. The inspector further noted
that the QA Surveillance Reports were not distributed outside the
QA Department. The chaiman further advised that, although OSC
minutes are routinely reviewed by the ORC, they were not used as
the document of record for review of violations per TS 6.5.2.7.e.
Reviews- intended to satisfy that requirement were conducted by
use of the actual document (LER, NRC Inspection Report, etc.),
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generally presented by a committee member.
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Typical violations of the type subject to ORC review but not
documented in the comittee's minutes include:
High Radiation Area / Contamination Area procedure and TS
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violations documented in QA Surveillance No. 007, February
26, 1980.
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Safety Injection System modification Weld Data Sheets not
complied with and Weld Data Sheets completed prior to weld
completion; documented in QA Surveillance Nos. 024, March
17,1980; 068, April 2,1980; and, 070, April 3,1980.
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Chronically recurrent cases of weld rod control procedure
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violations documented in several dozen QA Surveillances.
Unplanned or unmonitored radioactive releases documented
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in Incident Reports 81-11, January 12,1981 and 81-14,
January 22, 1981.
Personnel errors and procedure violations documented in
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LERs 81-06, 81-21, 81-17, 81-18, 81-53, 81-66, and others.
Recurrent procedure violations: QC Hold Tags, installation
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of material without receipt inspection, cable routing errors,
safety related rigging procedures, and safety related snubber
fill procedures, documented in QC NCARs.
The inspector advised the ORC Chairman that TS 6.5.2.7.e does not
require review of all violations but that an appropriate sample
of all types of violations and/or documents identifying violations
must be reviewed in committee with the goal of evaluating both the
significance of the violation and the effectiveness of preventive
and corrective 6ctions.
Failure of the ORC to review documents such as described above
constitutes violation of TS 6.5.2.7.e.
(82-01-01).
(Closed) Unresolved Item (81-29-03): No inspection of operating
activities as required by the 0QA Program. The PAS found that,
while certain activities such as maintenance and modification,
were subject to inspection by the licensee and contractor quality
control organizations, operating activities were not subject to
any in process inspection. These uninspected activities included,
but were not limited to:
Routine plant evolutions such as plant startup, shutdown,
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and routine system operation.
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Reactor engineering activities.
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Operating Surveillance Testing.
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Equipment lubrication activities.
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Chemistry activities.
Equipment clearance (tagout) activities, other than independent
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system operability and equipment position verifications.
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10CFR50, Appendix B, Criterion X, requires that a program for
inspection of activities affecting quality be established and
executed to verify confirmance with the documented instructions,
procedures, and drawings for accomplishing the activity. The
BVPS FSAR, Section A.2, contains the same requirement. BVPS
FSAR, Section A.2.2.2, states that the OQA Program applies to
plant operations associated with safety related (Category I)
structures, systems, and components. The DLC Quality Assurance
Policy, issued by the President, DLC, as part of the BVPS 0QA
Manual, requires, in Section 10, that a program for the inspection
of activities affecting quality be established and provide for
inspections during operations.
Through discussions with licensee management, including the
Manager, Nuclear Operations, and review of the BVPS 00A Manual,
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the inspector found that no program for the inspection of
operating activities had been established or executed.
Although
0QA Manual Procedure OP-9, Technical Procedure for Control of
Operations and Maintenance, Revision 1, assigns responsibility
for periodic test and inspection (Section 9.5.4),ther procedure
addresses only tests and inspections for equipment and does not
address inspection of the operating activities above.
No imple-
menting procedures for such inspections were identified.
The licensee does, however, provide for redundant and independent
verification of certain operating and surveillance test activities
such as clearance tagouts, jumper and lifted lead placement / removal,
valve lineup verifications, switch lineup verifications, etc.
These verifications are accomplished by individuals performing the
operations activity and do not meet independence requirements of
Criterion X and the 00A Program.
Similarly, the example activities
above are variously subject to supervisory review of documentation
and results and/or supervisory observation of performance. Post
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perfonnance results review does not satisfy requirements for
activity inspection.
Supervisory observation may not provide the
required inspection independence and generally does not meet the
0QA program requirements for a formalized, documented inspection
activity.
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Failure to establish and execute an inspection program for opera-
ting activities such as provided above constitutes a violation of
10CFR50, Appendix B, Criterion X and the licensee's 0QA Program.
(82-01-02).
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(Closed)
UnresolvedItem(81-29-04): Changes made to the
facility via Temporary Operating Procedures (T0Ps) and Special
Operating Orders (S00s) without adequate reviews as required
by 10CFR50, Appendix B, or 10CFR50.59. The PAS identified four
examples of potential ncacompliance:
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(1) TOP 80-27, Filling the RWST from the Boron Recovery Tanks.
The TOP provided a temporary hose flowpath for transferring
water between the tanks. The flowpath used was different
than the system arrangement described in Sections 9.1 and
9.2 of the Final Safety Analysis Report. These sections
describe the expected operation of the Chemical and Volume
Control System (CVCS) and Boron Recovery System, including
the normal flow paths for processing Boron Recovery Tanks.
The Figures of FSAR Section 9 (Flow diagrams) provide the
normal flow paths, including makeup flowpaths to the Refueling
Water Storage Tank.
The inspector found that TOP 80-27 had been approved via Onsite
Safety Committee (OSC) poll on August 15, 1980, that the poll
had been reviewed in committee on August 25,1980(Minutes
No. BV-0SC-104-80), and that the minutes did include a safety
evaluation as required by 10CFR50.59. The minutes stated that
the poll had been reviewed and that "no changes to procedures
or equipment as described in the SAR result and that no un-
reviewed safety questions results." No basis for the above
determination was provided nor was the temporary hose flow
path addressed relative to the systems as described by the
FSAR.
10CFR50.59(a)(1) permits the licensee to make changes in the
facility and procedures as described in the safety analysis
report unless a change to technical specifications or an
unreviewed safety questioncare involved.
requires that the licensee maintain records of these changes,
including a written safety evaluation which provides the bases
for the determination that the change does not involve an
unreviewed safety question.
Failure to document the bases for
the acceptability of TOP 80-27 constitutes a violation of
10CFR50.59(b). An additional example of further violation is
discussed below.
(82-01-03).
(2) TOP 81-27, Operating the Temporary Liquid Waste Demineralizer
(LW-I-2).
LW-I-2 was first installed and operated by TOP 80-33
(September 4,1980) to augment the capacity of the permanently
installed Liquid Radwaste System. The permanently installed
system was unable to process sufficient volume, resulting in
undesirable backlogs of unprocessed waste. The temporary
system, consisting of temporary filters, demineralizers and
plastic piping, had been installed as an " experiment" but
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had remained in use, with some modification, through 1981.
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TOP 81-27 was issued on July 17, 1981 to reflect the use
of a different style demineralizer. OSC Meeting Minutes
BV-0SC-84-81 (July 17,1981) documented review of the
original safety evaluation (T0P 80-33, Minutes BV-0SC-80-108
(September 4, 1980))and the committee's determinaticn that
the original evaluation remained valid. The inspector reviewed
the 80-33 safety evaluation, finding that it addressed con-
sideration of each of the items recomended by IE Circular
80-18,10CFR50.59 Evaluations for Changes to Radioactive
Waste Treatment Systems, and included the bases for the
comittee determination that no unreviewed safety question
nor technical specification change were involved.
During discussions with the DLC Chief Engineer, Station
Engineering Supervisor, and cognizant Operating Foreman, the
inspector learned that the system had not been fomally
designed but rather, had been assembled in place using
commercially available components and piping. The BVPS Opera-
tions Quality Assurance Program (FSAR Appendix A.2); QA
Procedure OP-4, Design Change Control, Revision 6; and the
BVPS 0QA Manual, Appendix B, Category I Systems, Structures
and Components, Revision 3, provide the requirements for
design change and modification control for safety related
systems. The Liquid Radwaste System is not considered a
safety related system by the BVPS QA Program and is therefore
not subject to the Quality Assurance Program controls.
Although application of fomal engineering and design change
controls to this installation were desirable, no violation
of NRC requirements was identified.
The extended use of a temporary system is similarly undesirable;
the licensee has experienced periodic problems with the in-
stallation. On July 14, 1981, Station Modification Request
(SMR) 404 was issued to DLC Engineering to install an improved
but again, interim system pending availability of a pemanent
system design. As of January 29, 1981, SMR 404 remained under
licensee review.
(3) TOP 81-31, River Water System Operation While Dredging Near
Intake. The TOP provided instructions for cross-connecting
the main River Water (RW) System with the Auxiliary River
Water (ARW) System during dredging operations at the main
intake structure to provide backup ccoling capability if
dredging silt fouled the main system. The inspector reviewed
FSAR Section 9.16; FSAR Question Response 2.30; BVPS Operating
Manual, Section 1.30, River Water System, and QA Manual,
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Appendix B, Revision 2, confirming that, although the
ARW System meets some of the design requirements of an
Engineering Safety Feature System (seismic design, single
failureanalysis,emergencypowersupplies,etc.)andis
capable of supporting a plant shutdown and cooldown on loss
of the main RW system, the ARW System is not considered QA
Category I (safety related). Additionally, the FSAR does
not describe extended operation of the ARW and RW systems
in a cross-connect mode during continued reactor operation.
TheinspectorreviewedOSCMinutesBV-0SC-93-81(August 7,
1981), finding that the OSC had determined that no unreviewed
safety question existed but provided no basis for that
detemination. Although the minutes noted that one ARW
header would be in service during dredging, no evaluation of
this alignment was provided. Failure to document the basis
for the detemination that an unreviewed safety question does
not exist is contrary to 10CFR50.59(b) and constitutes an
additional example of a violation.
(82-01-03).
(4) Special Operating Order (S00) 81-9, Pressurizer Relief Tank
Alarm Setpoints. Pressurizer Power Operated Relief Valve
seat leakage had caused abnomally high pressures and tempera-
tures in the Pressurizer Relief Tank. The licensee adjusted
the PRT high pressure alam and high temperature alarm setpoints
upward to reduce the frequency of alarm actuation and to reduce
the volume of water used to spray (quench) the tank each time
the alams actuated.
The inspector reviewed draft OSC Minutes BV-OSC-77-81 (June
25,1981) finding that S00 81-9 had been reviewed and found
not to involve an unreviewed safety question. The minutes
documented review of the setpoints relative to FSAR Section
4.2.2.3 and included consideration of the higher than nomal
tank temperatures and pressures. The minutes acknowledged
that such operation could result in PRT discharge to the
containment via the rupture disc on a design basis pressurizer
safety / relief valve actuation but noted that condition to be
an analyzed event. The DLC Chief Engineer also advised the
inspector that, although not documented in the meeting minutes,
he had consulted the pressurizer safety valve vendor and
determined that the abnomal operating conditions had no
adverse effect on safety valve operation. The inspector
also noted that, on July 17, 1981, misoperation of the PRT
spray system had resulted in PRT rupture disc failure and
discharge to the containment (Reference:
NRC Inspection
Report 50-334/81-18).
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The inspector also reviewed the procedures and data used
to reset the alanns finding that the PRT temperature and
pressure channels are not included in the licensee's safety
related calibration program, that the adjustments were made
on June 25, 1981 in accordance with generic Calibration
Procedures (CP515 and CP035) for the types of instruments
involved, and that the S00 provided the interim setpoint
information to plant operators.
Although the documented safety evaluation was considered
only marginally acceptable (discussions of safety valve
operation not included), no violation of NRC requirements
was identified.
(Closed) UnresolvedItem(81-29-05): Work Instructions not
included on Maintenance Work Requests (MWRs) required by Mainte-
nance Manual. The PAS found that minimal or no work instructions
had been provided on MWRs and that, in some cases, the scope of
work had changed but was not reflected on the MWR:
MWR 810190, Diesel Generator Lube Oil System. No written
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instructions were provided to repair an oil leak. A flange
was tightened to stop the leak.
MWR 810381, Channel 1 Loop RCS Flow. No written instructions
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were provided to replace the flow transducer. The instru-
ment was calibrated in accordance with MSP 6.26, referenced
on the MWR.
!!WR 810421, Loop Protection Tavg. No written instructions
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were provided on the MWR to perform the maintenance in
accordance with Corrective Maintenance Procedure (CMP)
1-75-45 which was used to perfonn the work.
MWR 810535, Number 2 Emergency Diesel Generator. There were
--
no written instructions on the MWR to replace the turbo-
charger in accordance with CMP 1-75-E08, which was used to
,
replace the turbo charger.
'
MWR 810566, Number 1 Emergency Diesel Generator. There
--
were no written instructions provided to repair the exhaust
bonnet for the electrical control panel.
MWR 811783, Containment Vacuum Pump. No written instructions
--
were provided to troubleshoot or repair bench board lights.
MWR 810949, Low Head Safety Injection Pump Recirculation
--
Pressure.
No written instructions were provided to replace
the pressure gauge.
y
.
13
The inspector reviewed:
BVPS QA Procedure OP-10, Maintenance and Modification
--
Planning, Revision 3;
BVPS QA Procedure OP-9, Technical Procedure Control for
--
Operations and Maintenance, Revision 3; and,
--
BVPS Maintenance Manual (MM), Chapter 1, Conduct of Maintenance,
Section A, General Rules for Implementation, Revision 14.
OP-10 provides requirements for work orders (Maintenance Work
Requests) and supplemental procedures or instructions for main-
tenance activities affecting quality. The OP requires that the
MWRs and/or supplemental procedures prescribe what is to be
accomplished and provide for a record of the quality achievements
andverifications(Section10.3.5). Any changes to the work order
or procedures must also be controlled, including review and approval
in the same manner as the original (Section 10.4.7).
MM Chapter 1, Section A, provides equivalent requirements.
Section
4.a.7 pemits instructions to be written in the " Additional Work
Instructions" section of the MWR for simple tasks requiring only
a few steps, otherwise Section 4.a.9 requires use of a written
procedure. Section 7.c requires that work shall be accomplished
as described in the instructions or procedures. Section 7.d requires
that any changes to instructions or procedures shall follow the
same approval process as the original procedures (except for on-
the-spot procedure revisions).
The inspector's review of the specific MWR examples identified no
actual problems resulting from the absence of work instructions
or procedure references. The work specified by these MWRs involved
troubleshooting / repairing.
In the case of sample repairs, the
repair was accomplished during the act of trouoleshooting.
Detailed
work instructions were inappropriate because the task was considered
to be within the skills nomally possessed by qualified maintenance
personnel.
In those cases where the troubleshooting identified a
more significant repair task, appropriate maintenance / calibration
procedures were subsequently referenced on the MWR and used in
the repair.
Discussions with the DLC Maintenance Supervisor
established that infomal oral comunications are normally used
to accomodate the inspector's concerns. However, the inspector's
review identified the following additional concerns:
(1) The Quality Control Department initially reviews MWRs and
detemines whether QC inspection coverage is required or de-
s t rable. A change in MWR scope does not appear subject to
additional QC review, particularly if a change in scope is
not documented on the MWR or is not subject to other specific
maintenance procedures.
_
.
.
14
(2) Similarly a change in scope may affect the post-maintenance
testing to be performed by the Operations Department. Failure
to document such changes do not pemit adequate Shift Super-
visor review upon closeout of the MWR.
(3) A change in work scope may also require a change in clearance
(tagout) boundaries. The licensee's procedures for MWR
control do not address such changes.
(4) The specific " additional work instructions" or reference
to supplemental procedures would nomally establish an authorized
scope of work and limit the worker's ability to extend the
scope of work without change control.
(5) None of the MWR examples included " additional work instructions"
or reference to supplement procedures or instructions (except
MWR 810381 which referenced a precedure for calibration of
thenewtransmitter).
The licensee acknowledged the inspector's concerns. The station
Maintenance Supervisor advised the inspector that procedure revisions
were in preparation at the close of this inspection and would be
issued by March 1, 1982. This matter will remain unresolved pend-
ing NRC:RI review of the procedures when implemented.
(82-01-07).
(Closed) UnresolvedItem(81-29-06): Vital Battery cells jumpered
without perfomance of 10CFR50.59 safety evaluation. On October
17, 1981, the number of cells in the No.1 bP.tery was reduced
from 60 (the number specified in the FSAR)to 58 in accordance
with MWR 817707 and Corrective Maintenance Procedure (CMP) 1-39DC-
Bat 1, 2, 3, 4-2E, Revision 1.
The cells had failed to meet the
voltage requirement of TS 4.8.2.3.2.b.1.
The remaining 58 cells
appeared to meet the minimum voltage requirements of TS but a written
safety evaluation had not been perfomed to establish that the
remaining battery capacity met FSAR and TS requirements.
During December,1981 - January,1982 the licensee was unable to
provide documentation of a written safety evaluation predating
the above change. The inspector reviewed OSC Minutes BV-0SC-68-79
(October 4, 1979) and -18-80 (January 30,1980) which documented
OSC review of Revision 0 and 1, respectively, of the jumper
installation CMP. The inspector found that the OSC had determined
no unreviewed safety questions existed but had failed to provide
the bases for that determination.No additional information was
available.
..
.
.
15
Although failure to provide a written safety evaluation per
10CFR50.59(b) is a violation of NRC requirements, a Notice of
Violation will not be issued. NRC Inspection 50-334/81-28
identified a similar violation for other jumper installations.
The licensee's response letter (dated February 8,1982) to
NRC:RI for that violation specifically addresses the corrective
and preventive actions regarding battery cell jumpers.
After the PAS finding, the licensee prepared a Safety Evaluation
Report, dated December 17, 1982, properly documenting the
determination that no unreviewed safety question exists for
jumpered battery cells. The safety evaluation addressed FSAR
and TS 4.8.2.3 capacity requirements and recommended a limit of
2 cells per battery be jumpered at any time. The safety evaluation
was reviewed by the OSC at Meeting BV-0SC-137-81 on December 31,
1981.
The minutes of this meeting were in preparation at the
end of this inspection.
The OSC also recommended that the Corrective Maintenance Pro-
cedures be revised to include the limit of two cells per battery
jumpered. The inspectors confirmed that the revised procedures
were approved on January 12, 1982 and were pending final typing
and distribution at the close of this inspection.
During this and prior inspections, the inspectors had discussed
TS 4.8.2.3.2.b.1 with licensee management, including the Chief
Engineer and Superintendent of Technical Services. The TS
requires that the voltage of each connected cell be greater than
2.02 volts under float charge and that the voltage has not decreased
more than 0.05 from the value observed during the original acceptance
test. The inspectors noted that IEEE Standard 484-1975, Installa-
tion of Large Lead Storage Batteries, considers an individual cell
acceptable if its voltage is within 0.4 volts of the average of
all cells. The licensee has experienced difficulty in maintaining
both original and replacement cells within 0.05 volts of their
original acceptance test values. The inspectors advised the
licensee that consideration of a TS change consistent with the
guidance of the IEEE Standard appeared appropriate.
(Closed) Unresolved Item (81-29-07): Audit program plan not
developed to assure coverage of applicable quality assurance pro-
gram.
10CFR50, Appendix B, Criteria XVIII requires the licensee
to establish a comprehensive system of planned and periodic audits
to verify compliance with all aspects of the quality assurance
program. ANSI N45.2.12, paragraph 3.3, Audit Planning, endorsed
by FSAR, Appendix A, Attachment to Table A.2, further requires
that the audit system be planned, documented, and conducted to
assure coverage of the applicable quality assurance program.
The audit system is to be periodically reviewed and revised as
necessary to assure that coverage and schedule reflect current
activities.
_ _ _ _ . .
_ _ _
.
.
16
QA Procedure OP-16, Revision 2, Section 16.4, Audit Planning,
provides guidance for scheduling audits that are planned, con-
ducted and reported in accordance with written procedures.
These procedures are to be revised, approved for use, and updated
as necessary. QA Procedure OP-16, does not provide for system
plans to assure coverage of all aspects of the quality assurance
program. QA Instruction 18.1.1, Revision 6, implements the QA
audit schedule requirements. The Senior QA Engineer is responsible
for assuring that audit responsibilities and obligations are met.
These responsibilities include detennining the areas to be audited
and the maintenance of audit schedule. The QA Manager approves
the audit schedules. The audit schedule is prepared annually
and reviewed quarterly by the QA Manager and lists all areas
requiring a QA audit. The audit schedule by itself does not
provide assurance that all aspects of the applicable QA program
have been covered, nor can it identify those areas that still
need coverage. The inspector held discussions with licensee QA
representatives to determine whether an alternate means existed
for assuring that all aspects of the program were covered by the
end of the year. None were identified.
Failure to develop an audit program system plan that can assure
coverage of all aspects of the QA program is a violation of
10CFR50, Appendix B, Criteria XVIII.
(82-01-04).
(Closed) Unresolved Item (81-29-08):
1980 0QA Program Manage-
ment Audit did not address all activities subject to 10CFR50,
Appencix B.
TS 6.5.2.8 requires that an audit of 0QA Program implementation be
performed under the cognizance of the Offsite Review Connittee (ORC)
at least once per 24 months. The audit must encompass the per-
formance of all activities required by the 0QA program to meet
the criteria of 10CFR50, Appendix B.
Criterion II, and BVPS OQA Procedure OP-1, Revision 4, also
require a management review (biennially per OP-1) to assess the
status and adequacy of the 0QA Program.
The PAS found that the 1980 Management Audit, performed by a
contractor, was intended to satisfy both the 0QA Program and TS
requirements. That audit addressed only five of sixteen major
0QA procedures used to implement the eighteen criteria of 10CFR50,
Appendix B.
The PAS also found that, although the audit report
was submitted to the licensee in December,1980, the report was not
distributed to the Offsite Review Connittee until September 29,
1981.
.
.
17
The inspector reviewed the 1980 audit, the licensee's management
auditplan(Memo,C.N.Dunn,datedAugust 18,1978),the1977
and 1978 management audits, and applicable sections of the 0QA
Procedures (ops). The inspector found that the licensee's audit
plan provided for the audit of the implementation of four or five
0QA Procedures during each biennial audit.
(Note: The OP-1 audit
requirements were changed from annual to biennial in 1979). The
inspector confimed that each audit report documented confomance
with the audit plan.
Implementation of the audit plan would result
in complete coverage of all 0QA procedures over a five year period.
Perfomance of biennial cudits for only portions of the 0QA proce-
dures does not constitute an adequate management review pursuant
to 10CFR50, Appendix B, Criterion II nor an adequate ORC audit
pursuant to TS 6.5.2.8.d and is considered a violation.
(82-01-05).
An additional example of similar violation is discussed in Item
81-29-09 below.
The inspector found that the late distribution of the 1980 Audit
Report to the ORC was apparently the result of administrative
oversight. Cognizant managers of the audited areas had apparently
received distribution of the report upon its receipt from the
contractor and at least once again in mid-1981. The report had
not, however, been fomally distributed to and reviewed by the
ORC. The inspector did not review the adequacy or timeliness of
corrective actions or ORC audit followup.
(Closed) Unresolved Item (81-29-09): Operation training audit
inadequate in scope and depth. The QA Department under the cog-
nizance of the ORC, conducted audit BV-1-81-4 on March 24 - April
4, 1981, to comply with the requirements of TS paragraph 6.5.2.8.b,
for the annual audit of the performance, training and qualification
of the entire facility staff. The inspector reviewed this audit
and discussed its content and scope with the lead auditor. The
audit was conducted to cover: BVPS training; training for Con-
struction Department - Nuclear, Schneider, Incorporated, Sargent
Electric Company and Dick Corporation personnel under their
Program requirements, as related to BVPS Unit 1 modification
program; and, to meet the requirements of TS 6.5.2.8.b.
The
audit consisted mainly of training record reviews for the above
groups, requalification training of licensed operators and
radiation technician training. The audit did not address the
staff areas of: mechanical and electrical maintenance, instru-
ment and control, testing and plant perfomance, reactor control
chemistry, station engineering, or the technical advisory group.
Though the licensee was undergoing a reorganization during the
period of audit, the organization of the above plant groups
remained essentially unchanged.
_
,
.
18
The scope of the audit was insufficient in that it did not address
the qualifications of any staff group, other than licensed operators
and radiation technicians, nor did it address the performance of
any group to assure that personnel performing a specified job
function were trained and qualified to do those jobs.
Failure to audit the performance of any facility staff group,
and failure to audit the training and qualifications for some
facility staff groups (mentioned above) constitutes another
example of violation of TS 6.5.2.8.
(82-01-05).
(Closed) UnresolvedItem(81-29-10): QA audits did not include
observations of perfomance of operating and maintenance activities
for 1980 and 1981. The inspector reviewed the following QA audits
of maintenance and operations activities perfomed in 1980 and
1981:
BV-1-80-33, Operations
--
BV-1-80-15, Maintenance
--
--
BV-1-89-39, Maintenance
BV-1-81-10, Maintenance
--
BV-1-81-28, Operations
--
BV-1-81-30, Maintenance
--
Through this review and discussions with licensee QA personnel, the
inspector detemined that the QA audits of operations and mainte-
nance activities did not include documented observations of any
activities in either audit checklists or results.
ANSI N18.7-1972, Administrative Controls for Nuclear Power Plants,
Section 4.4 requires audits to include observations of operations
and maintenance activities in addition to reviews of procedures
and records and interviews. ANSI N18.7-1972 is endorsed by the
BVPS FSAR, Appendix A.2, 00A Program, Section A.2.2, via Endorse-
ment of NRC Regulatory Guide 1.33-1972.
Quality Assurance Procedure OP-16, Audits, Revision 2, does not
include requirements for such observations during audits. Failure
to establish and implement procedure requirements for audit observa-
tions of such activities is a violation of 10CFR50, Appendix B,
Criterion XVIII; the BVPS FSAR, Appendix A.2 and ANSI N18.7-1972
as endorsed by Regulatory Guide 1.33.
(82-01-0 4).
.
.
19
(Closed) UnresolvedItem(81-29-11):
Personnel conducting audit
of Plant Operations did not have training or experience in nuclear
plant operations. The inspector reviewed the personnel folders
of the auditors who performed Quality Assurance (QA) audit
BV-1-81-28, Operations, during September 21 - October 7, 1981.
Prior to this audit, the team leader had performed 11 other audits
in 1981, none of which involved operations. This individual met
the audit team leader qualification requirements specified in QA
Instruction (QAI)
2.1.3, Training and Qualification of Auditors,
Revision.4, issued March 16, 1979. He also participated in a
continuing auditor training program as specified in QAI 2.1.2,
Training of QA Personnel, Revision 5, issued February 23, 1981.
However, this individual has had no previous experience in nuclear
plant operations nor other specialized training. The other audit
team member had received approximately 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> of operations
related training between 1976 and 1981; however, he too had
no previous experience in plant operations. Neither QAI 2.1.2
nor QAI 2.1.3 require specialized training or experience to
establish that the auditors' qualifications are commensurate with
the special nature of the activities to be audited.
ANSI N45.2.12-1974 as endorsed by the BVPS FSAR, Attachment to
Appendix A-2, requires that QA personnel or technical specialists
be selected for an auditing assignment based on experience or train-
ing which establish that their qualifications are comensurate with
the complexity or
special nature of the activities being audited.
Although both individuals participating in audit BV-1-81-28 were
_
qualified auditors, neither had plant operation experience nor
specialized training in this area. Failure to provide require-
ments for selection of auditors or use of technical specialists
having qualifications commensurate with the complexity or special
'
nature of activities to be audited is a violation of ANSI Standard
N45.2.12. _(82-01-04).
(0 pen) Unresolved Item (81-29-12): Written training, reh &ing,
and replacement training program for all unlicensed personnai not
provided and fully implemented.
(0 pen) UnresolvedItem(81-29-13):
Key personnel not provided BVPS plant specific training as required
The PAS found that established training programs did not appear
to meet all requirements of the licer.see's OQA program and Technical
Specifications and that existing programs specified in the BVPS
Training Manual had not been fully implemented. Additionally,
key personnel reassigned to CVPS from other DLC facilities had not
been provided plant specific training.
.
.
.
20
The review of these two items was in progress at the end of this
inspection. The inspector had reviewed the applicable regulatory
requirements and was reviewing a sample of facility staff member
training records for compliance with the licensee's existing
programs and license requirements.
Items 81-29-12 and -13 will
remain unresolved pending completion of the inspector's review.
During the review, the inspector found that BVPS Maintenance
Department required reading assignments had not been p(MM) contains
roperly
performed. Chapter 1 of the BVPS Maintenance Manual
administrative procedures for the control of maintenance activities.
Chapter 1. Section A.5.b, Revision 14, requires electricicans,
mechanics, and meter and control repairmen (instrument technicans)
to read pertinent sections of Chapter 1.
The table below shows
inspection results for a sample of 5 individuals whose records
were reviewed:
MM Section
N
Q
H
J
W
Y
Z
0
Legend
Electr. A
C
C
I
C
X
C = Complete
Electr. B
C
C
I
C
X
I = Training
incomplete or
Mech. A
X
X
C
C
C
I
C
C
not documented
X = Not required
Mech. B
X
X
I
I
C
I
C
I
for job classi
fication
Tech. A
C
X
C
C
C
I
C
C
- = Training in-
complete but
Tech. B
I
X
I
I
I
I
I
C
identified by
DLC internal
memo dated
4/16/81
Section N - Relay Testing
Section Q - Motor Repair
Section H - Cleaning & Maintaining Cleanliness
Section J - Housekeeping
Section 0 - Calibration Program
Section W - Maintenance Procedure Control
Section Y - Control & Maintenance of Respiratory Equipment
Section Z - General Work Practices
Each individual is rated as a "first class" craft worker normally
assigned to safety related maintenance activities. Of the 40 re-
quired reading assignments sampled, only 20 had been recorded as
complete.
.
21
10CFR50, Appendix B, Criterion II, requires that the OAQ program
provide for indoctrination and training of personnel performing
activities affecting quality as necessary to assure that suitable
proficiency is achieved and maintained. The BVPS FSAR, Appendix
A.2.2.2, states that indoctrination and training measures assure
that all responsible individuals are aware of quality policies,
procedures and manuals and have an adequate understanding of these
requirements. QA Procedure OP-14, Indoctrination and Training,
Section 14.4.1, Revision 3, requires that station personnel shall
be trained, as appropriate, to achieve special skills required in
the performance of equipment protection, process, and test pro-
cedures and that retraining will be provided as necessary to main-
tain adequate proficiency.
The BVPS MM, Chapter 1, Section A.5.b, General Rules for Implemen-
tation, Revision 14, requires individuals to receive indoctrina-
tion on specific sections of the manual (as represented in the
above table) and to document completion of the training. Failure
to complete and document this training constitutes a violation.
(82-01-06).
(Closed) UnresolvedItem(81-29-14):
Training records not stored
in accordance with QA Program requirements and ANSI N45.2.9. The
PAS found that training records were improperly stored at the Nuclear
Division Training Center (the Johnson Street School) and that de-
partmental training records maintained by managers and supervisors
were not subject to QA record requirements.
The inspector reviewed records at the Johnson Street School and in
various station departments with respect to Technical Specification 6.10.2; QA Procedure OP-15, QA Records, Revision 1; and ANSI
N45.2.9-1974Property "ANSI code" (as page type) with input value "ANSI</br></br>N45.2.9-1974" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process., Requirements for the Collection, Storage, and Main-
tenance of Quality Assurance Records for Nuclear Power Plants.
Discussions were held through the inspection with Nuclear Division
supervisors including the Manager, Nuclear Support Services, and
the Nuclear Division Director, Adininistrative Services.
The inspector determined that the licensee had established an on-
going QA records review and upgrade program, already addressing plant
drawings, technical manuals, maintenance work requests, and various
procedure records; establishment of satelite record centers; and,
division-wide categorization and indexing of QA records.
Each
activity resulted from prior licensee identification of records
program deficiencies or record discrepancies. Corrective actions
for each problem were either in progress or in planning.
,
- , - -
'
22
The inspector found that single copy training records at the
Johnson Street School and in BVPS-1 station departments were
not stored or maintained in accordance with the above require-
ments but that the licensee had initiated corrective actions
similar to the others above. Other areas requiring similar
licensee attention identified by the inspector were: station
incident reports, operating procedure and valve lineup records,
radwaste and transportation records, and certain maintenance
records. By the end of this inspection the licensee had begun
addressing each of these items as part of the overall effort.
The Nuclear Division Director, Administrative Services advised
the inspector that:
1) a fire loading / rating survey had been
completed for the Johnson Street School records room; 2) engineer-
ing action was in progress to upgrade that facility to meet ANSI
N45.2.9Property "ANSI code" (as page type) with input value "ANSI</br></br>N45.2.9" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process. requirements; 3) the use of computerized duplicate
training records was being evaluated; and 4) an expected date of
completion of actions for the Johnson Street facility would be
provided to the inspector by about March 15, 1982. The Nuclear
Division Director, Administrative Services further advised that
a general review of in-plant and divisional records would be
performed to ensure that all single copy records were properly
stored; the expected date of completion would also be provided
by about March 15, 1982. The acceptability of the licensee's
actions in this matter will remain unresolved (82-01-08).
Unresolved Item 81-29-14 will be closed.
(Closed) Unresolved Item (81-29-15): Appropriate acceptance
criteria not available for activities affecting quality.
PAS
review of Welding Manual Procedure No.1.4, Welding Electrode
Control, and NSQC 8.1, Storeroom Quality Control, found that
certain activities affecting quality were not governed by appro-
priate acceptance criteria.
(1) QA Procedure OP-12, Control of Measuring and Test Equipment,
Revision 3, requires the calibration of instruments and
measuring devices used in activities affecting quality,
regardless of the owner or user. Welding Manual Procedure
No.1.4, Welding Electrode Control, Revision 1, requires
that ovens used to store covered electrodes be maintained
0
at 250-300 F. During the PE> inspection the inspector observed
I
that thermometers used to determine the temperature of the
weld electrode oven in the Maintenance Department had not
been calibrated.
The licensee has since calibrated the thermometer and has
scheduled it for periodic recalibration in accordance with
their administrative controls. The inspector had no further
questions on this item.
.
.
.
23
(2) General Stores (GS) Procedure 204.0, Storage Functions,
Revision 2, requires that temperature and humidity be
controlled within specified limits for the Level A storage
room. The inspector noted that the temperature and humidity
recording instruments had not been calibrated. A sampling
review of purchase requisitions by the inspector failed to
identify any piece of equipment that required Level A storage.
Although the licensee had stored electronic instruments and
parts in this storage area, the inspector verified that this
equipment did not require this level of storage. The licensee
has since initiated a calibration program for the storeroom
instrumentation. As Level A storage was apparently not required
up to this time and the licensee has instituted a routine
calibration program for those instruments, the inspector had
no further questions on this matter.
(3) The temperature and humidity limits discussed in GS 204.0
are not specified by the procedure. Through discussions
with Stores personnel, the inspector determined that these
limits had not been established at the time of the PAS. inspection.
The licensee has since issued a memorandum that specifies
appropriate limits, subject to change as required to meet
special material storage requirements. As there has apparently
never been equipment stored in the Level A storage room that
required these special temperature and humidity controls,
the inspector had no further questions at this time.
(Closed) Unresolved Item (81-29-16): The PAS identified the
following items related to storage and inspection of materials:
(1) Bags of cement were stored within 10-15 feet of reactor
plant equipment such as pumps, motors, valves, and spare
parts in the warehouse. Several cement bags had been punctured.
A plastic tarp used for covering one skid of cement was not
in place.
The tarp was immediately lowered over the exposed
cement bags to contain any dust. The inspector had no
further questions on this item.
(2) Openings in four Limitorque valve operators did not have
covers or seals to prevent foreign material from entering
the valve operators.The PAS inspector further noted that
attached vendor instructions directed that desiccant be
added if the operators were not immediately installed.
Discussions with storeroom personnel confirmed that instruc-
tions from Engineering specifying storage and maintenance
requirements were not available and that stores personnel
__
.
.
24
had not been instructed to look for vendor storage require-
ments.
During this inspection, the inspector confirmed that the
open valve parts had been covered and that requirements for
installation inspections and cleaning would assure proper
cleanliness levels if and when the valves are used in reactor
plant systems. When the licensee disassembled the valve
operator to add disiccant, three packs were found inside the
operator, apparently added by the vendor before shipping.
The licensee's overall program receipt and storage of equip-
ment and material will be inspected in the near future as
part of the prescribed inspection program. The inspector
had no further questions on this item.
(0 pen) Unresolved Item (81-29-01): Bases for 10CFR50.59(b) deter-
minations not documented by Onsite Safety Committee (OSC) for
procedure changes. The PAS found that the committee did not
make written evaluations of procedures or changes to procedures
to document the bases for their detennination that no unreviewed
safety question existed.
The inspector reviewed a sample of 50 procedure changes documented
in OSC Meeting Minutes BV-0SC-71-81 (June 10,1981) through
BV-0SC-131-81 (December 17,1981), including changes to operating
procedures, surveillance procedures, calibration procedures, and
system descriptions. The inspector also reviewed Section 12.5,
Procedures, of the BVPS FSAR and selected other sections discussing
test and inspection procedures. The inspector found that none of
the fifty procedure changes sampled appeared to constitute a change
to the procedure as described by the FSAR. This item remains
open pending additional NRC:RI management review.
-
- _
1
.
.
25
3.
Plant Operations
a.
General
The facility was shutdown for all of the inspection.
Inspections
and plant tours were conducted during day and night shifts with
respect to outage activities and maintenance of safe shutdown
conditions. Acceptance criteria for those inspections included:
--
BVPS FSAR Appendix A, Technical Specifications
BVPS Operations Manual, Chapter 48, Conduct of Operations
--
--
OM 1.48.5 Section D Jumpers and Lifted Leads
OM 1.48.6, Clearance Procedure 3
--
OM 1.48.8, Records
--
OM 1.48.9, Rules of Practice
--
BVPS Operations Manual, Chapter 55A, Periodic Checks -
--
Operating Surveillance Tests
BVPS Maintenance Manual, Chapter 1, Conduct of Maintenance,
--
Section J, Housekeeping
BVPS Radcon Manual, various sections
--
10CFR50.54(k), Control Room Manning Requirements
--
--
Inspector Judgement
--
BVPS Physical Security Plan
Findings resulting from these inspections are discussed in
paragraph 31 below.
b.
Areas Inspected
Primary Auxiliary Building, including High Radiation Areas
--
and Loose Surface Contamination Areas
--
Service Building
Main Steam Valve Room
--
Purge Duct Room
--
--
East / West Cable Vaults
--
Emergency Diesel Generator Rooms
Containment Building, including High Radiation Areas
--
Penetration Areas
--
Safeguards Areas
--
,
Various Switchgear Rooms, Cable Spreading Room
l
--
Protected Area
l
--
l
The inspectors also toured the Control Room regularly to review
logs and records and conduct discussions with operators concerning
l
reasons for selected lighted annunciators and knowledge of recent
'
changes to procedures, facility configuration and plant conditions.
,
!
,
_
_
_
.
.
26
c.
During daily Control Room tours the inspectors made the following
observations:
(1)
Instrument and recorder traces for systems required during
shutdown were observed for abnomalities. Systems included:
--
Residual Heat Removal (RHR) System
Chemical and Volume Control System (CVCS)
--
Fuel Pool Cooling and Purification System
--
Supplementary Leak Collection and Release (SLCRS) System
--
Liquid (LW) and Gaseous (GW) Radioactive Waste Systems
--
--
Fire Protection Systems
NuclearInstrument(NI) System
--
--
ProcessandAreaRadiationMonitors(RMs)
Offsite and Onsite Electrical Power Systems
--
(2) Proper Control Room and shift manning were confimed. Control
of personnel access was confimed to be in accordance with the
(3) The inspectors verified operator adherence to approved operating
procedures for partial RCS draindown per BVPS OM Section 1.6.4.N.
Draining RCS to Centerline of Hot Leg Loops for Maintenance,
Revision 3.
These inspections activities are further discussed
in paragraph 5 of this report.
(4) The following licensee logs and documents were reviewed daily
on a rotating basis during the inspection to obtain infomation
on plant conditions, determine compliance with regulatory re-
quirements and assess the effectiveness of the communications
provided by the documents:
Nuclear Shift Supervisors Logs
--
Nuclear Control Operator Logs
--
Equipment Clearance Logs
--
--
Caution Tag Log
Special Operating Orders
--
Waste Handling Systems 7 Day Running Logs
--
--
Chemistry Log Sheets
Nuclear Shift Operating Foreman Logs
--
Radcon Foreman Logs
--
--
EquipmentOutofService(00S) Logs
--
Temporary Operating Procedures & Log
Temporary Logs Sheets (for special surveillance or
--
operations)
--
Nuclear (auxiliary) Operator Logs
- Note:
Each of these logs was reviewed for the entire inspection
period. All other logs were reviewed at least weekly.
_ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ . . _ _ _ . .
.
.
27
(5) The inspectors observed Control Room instrumentation,
controls, and indicators to verify that ongoing operations
and maintenance of shutdown conditions were in conformance
with Technical Specification (TS) Limiting Conditions for
Operations (LCOs). Portions of the below TS LCOs confirmable
from the Control Room were observed on the dates shown:
TS
Title
Date
3.1.1. 2
Reactivity Control System,
January 6, 1982
0
Tevg less than 200 F
January 14,1982
February 2,1982
3.1.2.3
Reactivity Control System,
January 7,1982
Charging Pump - Shutdown
February 2,1982
February 4,1982
3.4.1.3
January 8,1982
Shutdown
January 11,1982
January 13,1982
January 19,1982
February 4,1982
3.8.2.2
Electrical Power System -
January 11,1982
A.C. Distribution - Shutdown
January 14,1982
January 19,1982
3.8.1.2
Electrical Power System -
January 13,1982
Shutdown
January 28,1982
January 29,1982
February 2,1982
3.1. 2.1
Reactivity Control System,
January 14,1982
Boration System
January 22,1982
February 2,1982
3.1.2.7
Reactivity Control System,
January 18,1982
Borated Water Sources - Shutdown
January 22,1982
February 2, 1982
3.1.2.5
Reactivity Control System, Boric
January 22,1982
Acid Transfer Pump
February 2,1982
. _ - _ _ - _ _ _ _ _ . _ _ _ _ _ _
.
.
28
(6) The inspectors reviewed completed surveillance tests to
verify that:
the tests were completed as scheduled; test
results were reviewed by responsible supervisors, and that
corrective actions were initiated for test identified
deficiencies:
OST 1.43.1 - TS Required Area and Process Monitors
--
Channel Functional Test, Revision 5, performed
January 6, 1982, following corrective action on
RM-RW-100 per MWR 816045.
OST 1.43.2 - Area and Process Monitors Channel
--
Functional Check, Revision 16, completed January 6,
1982 to return RM-VS-105 to service.
--
OST 1.39.lD - Weekly Battery Check - Battery No. 4,
Revision 7, performed January 21, 1982.
OST 1.20.1, Spent Fuel Pool Level Verification,
--
Revision 0, perfomed January 21, 1982.
OST 1.49.2, Shutdown Margin Calculations, Revision 11,
--
perfomed January 6,13, and 22,1982.
OST 1.20.2, FC-P-1A Fuel Pool Pump Operability Test,
--
Revision 7, initiated January ll,1982 and put on hold
for review after the pump failed to meet the differential
pressure acceptance criteria.
MSP 21.06, P-495 1C Steamline Pressure Protection (Loop
--
3) Channel III Test, Revision 3, perfomed December
24, 1981.
MSP 13.09, L-100A RWST Level Loop Channel III Test,
--
Revision 2, perfomed December 24, 1981.
--
MSP 10.02, RHR Automatic Isolation and Pressure Interlock
Test, Channel II, Revision 3, performed January 1, 1982.
MSP 2.10, Nuclear Instrument Source Range N-32
--
Calibration, Revision 4, performed December 29, 1981.
OST 1.7.3, Boric Acid Transfer Pump Operational Test,
--
Revision 21, performed January 22, 1982.
--
OST 1.39.lE, Weekly Station Batter Check, Battery No. 5,
Revision 8, performed January 22, 1982.
-.
-
.
_
_ _-
.
0
29
--
OST 1.7.8, Boric Acid Storage Tanks and RWST Level
and Temperature Verifications, Revision 6, perfomed
January 22, 1982.
OST 1.32.1, Chemical Waste Sump, pH Monitor Operability
--
Check, Revision 5, and OM Change Notice 82-12, performed
January 14 and 22,1982.
OST 1.16.2, Supplementary Leak Collection and Release
--
Exhaust Fan and Remote Damper Component Test (Train B),
Revision 8, performed January 6, 1982.
OST 1.48.1, Mode 5 and 6 ESF Train Operability, Revision
--
5, performed January 6 and 13,1982 for Train "B."
OST 1.36.7, Offsite to Onsite Power Distribution System
--
Breaker Alignment Verification, Revision 18, perfomed
January 9, 1982 for Mode 5 conditions.
OST 1.36.9, AC Power Source Breaker Alignment Verification
-
--
During Shutdown, Revision 1, performed January 9, 1982.
OST 1.11.10, Boron Injection Flow Path Power Operated
--
Valve Exercise, Revision 30, perfomed January 12, 1982.
OST 1.7.3, Boric Acid Transfer Pump Operational Test,
--
Revision 21, perfomed January 29, 1982, OMCN 82-03,
January 5,1982.
OST 1.1.10, Cold Shutdown Valve Exercise, Revision 29,
--
performed January 28, 1982.
OST 1.7.8, Boric Acid Storage Tanks and RWST Level and
--
Temperature Verification, Revision 17, performed January
28, 1982.
OST 1.33.3, Fire Protection System Drain Test, Revision
--
33, performed January 23, 1982.
OST 1.39.1A, Weekly Battery Check - Battery No. 1,
--
Revision 7, performed February 8,1982.
--
OST 1.39.1B, Weekly Battery Check - Battery No. 2,
Revision 7, perfomed February 8,1982.
_
.
30
OST 1.49.2, Shutdown Margin Calculation, Revision ll,
--
perfonned February 9 and 10,1982.
OST 1.16.5, Fuel Building Ventilation System Verifica-
--
tion - Fuel Storage, Revision 7, performed February 10,
1982.
OST 1.36.2 - Diesel Generator No. 2 Monthly Test,
--
Revision 25, perfomed February 4,1982.
d.
The following activities were inspected during tours of the plant
areas listed in paragraph 3.b:
(1) Safety related tagouts (below) were verified to be properly
posted with equipment properly positioned and redundant equip-
mentoperable(ifrequired):
--
"A" Emergency Diesel Generator, Equipment Clearance Tag
No. 473726 and 466270, placed January 6, 1982.
Pressurizer Heaters, Equipment Clearance Tag No. 466277,
--
observed January 18, 1982.
"C" Component Cooling Water Pump, Equipment Clearance
--
Tag No. 466140 and 466214, observed January 8, 1982.
M0V-RC-557 A&B, Equipment Clearance No. 466315 and
--
444316, placed February 11, 1982.
(2) The inspectors independently verified plant conditions and
equipment status required for confomance with the following
TS LCOs during inspection tours outside the Control Room:
--
TS 3.4.1.3 - Reactor Coolant Systems - Shutdown
TS 3.8.1.2 - Electrical Power Systems - Shutdown
--
--
TS 3.8.2.2 - Electrical Power Systems - AC Distribution
TS 3.4.2 - Reactor Coolant Systems - Safety Valves
--
TS 3.7.15 - Fire Barrier Penetrations
--
TS 3.7.14.4 - Fire Hose Stations
--
(3) General plant / equipment conditions including operability
and verification of standby equipment, pipe hanger / seismic
restraint settings and oil levels, and instrumentation and
recorders functional.
(4) The inspectors verified that Maintenance Work Requests (MWRs)
had been initiated for equipment in need of maintenance and
that proper priorities had been assigned to the repairs.
Examples include:
.
.
31
MWR 816045, completed January 6,1982 to repair
--
RM-RW-100
MWR 820046, completed January 6, 1982 to repair
--
MOV-RH-758
MWR 820045, completed January 6,1982 to repair
--
FCV-CH-ll4
MWR 817917, observed maintenance activities on
--
January 29, 1982 to rebuild PCV-MS-106A
(5) Toured areas were observed for fire hazards, availability
and operability of fire fighting equipment and emergency
equipment, and general condition of fire alarms and actuating
controls. The inspectors verified that observed ignition
sources were being controlled in accordance with BVPS OM
Section 1.56.
(6) The following ongoing activities outside the Control Room
were observed to confinn that they were conducted in
accordance with applicable administrative controls:
Rebuild Steam Dump Valve TCV-MS-106A6 per:
(i)MWR817918
--
for mechanical work; and, (ii) MWR 817917 for actuator
work. The inspector verified proper equipment clearance
per Clearance No. 479124.
Completion of OST 1.39.lC, Weekly Battery Check - Battery
--
3, Revision 7, performed January 21, 1982.
(7) Plant housekeeping conditions and cleanliness were observed
to confirm that:
--
Critical clean areas are controlled.
Excess materials and materials are returned to storage
--
areas.
Combustible materials and debris are promptly removed
--
from the facility.
(8) The inspectors observed implementation of the Physical Security
Plan, including:
Proper manning of the security organization.
--
Security personnel were capable of performing their
--
assigned functions.
--
Protected Area barriers were not degraded.
Isolations Zones were clear.
--
Persons and packages were properly checked prior to
--
Protected Area entry.
.
.
32
Vehicles are properly authorized, searched and
--
escorted or controlled within the Protected Area.
'
Persons within the Protected and Vital areas display
i
--
photo identification badges and are properly escorted
if required.
Communications checks were conducted and proper com-
--
munications devices were available.
Compensatory measures were employed when required by
--
security equipment failure or impairment and were
effective.
Security access controls to vital areas are properly
--
implemented.
Response force composed of specified amed individuals
--
available and responded in a timely manner on January
19, 1982.
Explosives Detector Alam checked prior to returning to
--
service on January 13, 1982.
--
Vital Area controls for outage activities are also
discussed in paragraph 14 of this report.
(9) Shift turnovers of the following work groups / departments
were periodically observed to ensure continuity of infoma-
tion between shifts:
Group
Date
Operations
January 6, 7, 11, 14, 18, 21,
and 26, 1982
Health Physics
January 8,13, 22 and 28,1982
Security
February 11, 1982
Test Group
January 21, 1982
(10) The following radiological control activities were observed
on day and night shifts:
(a) Portions of the following licensee perfomed surveys
were observed for confomance with BVPS Radcon Manual
requirements:
Survey of Blender Cubicle High Radiation Area
--
in the Primary Auxiliary Building (discussed
further in paragraph 3.i(3).
Survey conducted during Type C local leak rate
--
test on penetration 20 inside containment and the
Primary Auxiliary Building on January 21, 1982.
--
Survey of Steam Generator Drain Tank Building during
decontamination efforts (discussed in paragraph 6.b).
Survey of Solid Waste Areas conducted as part of
--
TOP 82-07 and subsequent decontamination (discussed
furtherinparagraph6.d).
_
_
-
__
_
._
_.
.
.
33
(b) Radiation protection instruments below were inspected
during plant tours to verify operability and adherence
to calibration frequency requirements.
(c) The following Radiation Work Permits and Radiation
Access Control Permits were reviewed for completeness.
The permit (s) denoted by asterisk were reviewed in the
'
field to verify that the permit requirements were being
followed:
RWP B009351, Repair Line on LW-TK-7B, January 19,
--
1982.
RWP B009359, Decontaminate Outside Area, January
--
19, 1982.
RWP B009133, Dewatering Demineralizers, January
--
8, 1982.
PACP 81-11-H, Inspection and Surveillance, January
--
14, 1982.
RWP B009122, Clean / Repair / Test Motor Operators,
--
January 1,1982 for SI-850A.
--
RWP B009299, Install Hydrogen Recombiner System
at735'S/G(Maint), January 14, 1982.
RWP B009501 Decon LW-P-1A8B Cubicle, February 4,
--
1982.
RWP B009501, Decon Solid Waste Area, February 8,
--
1982.
(d) The inspector audited the licensee's Jumper and Lifted
Lead controls on January 15 and 19,1982 in accordance
with BVPS OM Section 1.48.5 to ensure no conflicts with
Technical Specifications, that the licensee is actively
pursuing correction of conditions requiring the jumpers
or lifted leads, and that the installation and removal
were proper. The following sample was audited:
--
2021, Area Ventillation, to allow VS-C-1B1 to run
with V S-E-4B out-of-service, placed March 4,1978.
2122 and 2123, Gaseous Waste System, to disable
--
Alam A-2-1, placed February 3,1979.
2766, Miscellaneous Safety Related System, to disable
--
alam A6-51, placed October 12, 1981.
(e) The inspector witnessed selected portions of the below
radioactive releases to verify conformance with approved
procedures, that required release approvals have been
obtained, that required sampling was accomplished, and
that effluent release instrumentation was operable:
RWDA 01736, Steam Generator Drain Tank 7A, discharged
--
to Unit 2 Blowdown, observed February 1, 1982.
.
.
34
(f) The following records of liquid and gaseous radioactive
releases were also reviewed to assure that approvals
were obtained, sampling was accomplished,fand release
limits satisfied:
RWDA (Liquid) 1724, Steam Generator Drain Tank 7B,
--
December 20, 1981.
RWDA 1724, Steam Generator Drain Tank 7A, January
--
1, 1982.
RWDA 1727, Laundry, January 1,1982.
--
RWDA (Gaseous) 0560, Gas Decay Tank 1A, January
--
3, 1982.
RWDA 0559, Gas Decay Tank 1C, January 2, 1982.
--
RWDA 0553, Containment Purge, December 30, 1981,
--
(Reviewed by OSC meeting 136-81).
(g) The inspector observed solid radioactive waste disposal
activities to verify implementation of administrative
controls.
Spent resin transferred to lined waste cask for
--
offsite disposal on February 7 - 10, 1982.
(h) The inspector observed portions of on-going maintenance
activities to verify that:
These activities did not violate TS Limiting
--
Conditions of Operation.
Required administrative approvals and tagouts
--
were obtained prior to initiating work.
Approved procedures were being used or the
--
activity was within the " skills of the track."
c
QC reviewed the proposed work to define any desired
--
hold points.
Activities observed included:
Rebuilding Steam Dump Valve TCV-MS-106A6 per
--
(i) MWR 817918 for mechanical work, and
(ii) MWR 817917 for work on the valve actuator.
Instrument Calibration of Rod Area Monitor RM-215A
--
per MSP 43.08, Containment Particulate Calibration,
Revision 3, performed February 11, 1982.
_ _ _ _
_
_ _
__
-
-
---
35
e.
Findings
(1) Fire Barrier Penetrations Inoperable
Technical Specification 3.7.1.5 requires all penetration fire
barriers protecting safety related areas to be functional at
all times. The Action Statement of TS 3.7.15 requires that,
with penetration fire barrier nonfunctional, a continuous
fire watch must be established on at least one side of the
affected penetration within one hour.
During this inspection,
three examples of violation of TS 3.7.1.5 were identified.
(a)
TOP 82-6 (pection of temporary cable installations per
During ins
paragraph 6c, this report) the inspectors
found that the temporary 480 volt pcwer supply passed
through the wall between the AE and DF Vital Switchgear
Rooms via a standard conduit penetration. The fire
barrier packing for the penetration had apparently been
removed for the cable installation but not replaced. No
fire watch was present. The inspectors found the condition
at about 10:50 a.m. on January 28, imediately notified
the Control Room, and observed repacking of the pene-
tration with acceptable fire stop material at about
11:10 a.m.
Inspector discussions with the cognizant
Electrical Maintenance Foreman and review of operator
logs established that the cable had been run through the
penetration at or before 9:00 p.m. on January 27,
apparently without proper replacement of the fire stop
material or fire watch.
(b) During a February 5,1982 tour of the PAB Safeguards
Area, the inspectors found an 8" conduit penetration
- WCV-735'-111 between the West Cable Vault and the
Auxiliary Feedpump Room open and not stuffed with fire
retardant material. No fire watch was posted as required
by TS 3.7.15.
The inspectors imediately notified the
Nuclear Shift Supervisor and witnessed repacking of the
penetration with fire retardant batting.
(c) Additionally, the inspectors noted that similar finding
was documented in the Nuclear Shift Supervisor's log
for 8:25 p.m., February 5.
A Nuclear Shift Operating
Foreman on plant tour found welding cables passing through
a fire barrier penetration between the AE and DF Vital
Switchgear Rooms without fire retardant stuffing or a
The licensee was unable to establish the date and/or time
the penetrations of examples 2) and 3) were unpacked to make
them inoperable or whether fire watches had been posted at
any time during the inoperability.
Failure to comply with TS 3.7.15 constitutes a violation.
(82-01-09).
..
_
-
_-
. _ _
.
.
.
36
(2) Security Picture Badge Procedures
During a February 5, 1982 tour of the #1 and #2 Emergency
Diesel Generator Rooms, the inspectors observed two
construction workers not wearing their security picture
badges as required by the BVPS Physical Security Plan. The
individuals, inside the #1 EDG Room Vital Area, had left
the badges en their coats when the coats were removed. When
addressed by the inspectors, both individuals put the badges
on as required.
This matter was discussed with the DLC Security Supervisor
and Security Assistant on February 5-6.
Based on no recent
-
similar observations, the inspectors considered the above
apparent violation of security plan requirements to be an
isolated case and advised the licensee representatives that
any future observations of improperly displayed security
badges would be considered as violations. The licensee was
ensuring that construction craft workers were made aware of
the need to properly wear badges via contractor safety meetings
and communications. The inspectors noted that about 1000
construction contractor personnel are onsite to support the
current outage and that prior observations showed no widespread
abuse of badge requirements.
(3) High Radiation Barrier Unlocked
At 2:45 p.m., January 28, 1982, during a plant tour, the
inspectors found the radiation barrier door to the CVCS
blender cubicle unlocked and ajar about 8 inches. The
door is a fence-type gate properly posted as a High Radiation
Area and Contamination Area /RWP required for entry. The
inspectors immediately notified the cognizant Radcon Foremtn
who responded to the scene, insured no personnel were within
the cubicle, and relocked the gate.
The inspectors and licensee reviewed survey data, Radiation
Access Control Pemit logs, Radiation Work Permit Logs, Instru-
ment Issue Logs, and Radiation Area key logs, confiming that
only two parties had entered the area since it was last
verified locked at about 11:30 a.m., January 28. Neither of
the two parties (an operator and a housekeeping crew) recalled
finding or leaving the gate unlocked. The inspectors con-
fimed that both parties were properly authorized entry and
both were equipped with required radiation survey meters.
Survey data showed that general area radiation levels within
the cubicle were less than about 60 mrem /hr. except at and
below grating level. At grating level, localized fields were
about 100 mrem /hr; below grating level, maximum fields were
600 mrem /hr. or less.
'
.
_
-
_
.
. _ _
.
.
37
TS 6.12.1.a and BVPS Radcon Manual, Radcon Procedure 9.2,
Radiation Area Control
Issue 1, Revision 2, require entry
to high radiation areas of more than 100 mrem /hr. but less
than 1000 mrem /hr. be controlled via RWP and barricades.
Locked barricades are only required if the radiation intensity
is greater than 1000 mrem /hr. The licensee locks the lower
intensity cubicles to provide additional positive entry control
beyond that required by TS. Based on the above and the
isolated nature of this finding, no violation of NRC require-
ments was identified.
4.
In Office Review of Licensee Event Reports (LERs)
The inspector reviewed LERs 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 actions.
The inspector determined whether further information was required from
the licensee, whether generic implications were indicated, and whether
the event warranted onsite followup. The following LERs were reviewed:
- --
LER 81-103/03L
Over Current Trip of IDF Emergency Bus due
to ground from 1C Component Cooling Water
Pump Motor.
LER 81-104/03L
Loss of Core Cooling Monitor due to power
--
supply failure.
LER 81-105/03L
Hydrogen Recombiner Pre-heater discharge air
--
temperature failed surveillance test.
LER 81-106/03L
Drift of 7 Main Steam Safety Valve Setpoints.
--
LER 82-01/03L
Supplementary Leak Collection and Release
--
System Train A failed surveillance test.
No unacceptable conditions were identified.
LER 81-103/03L was submitted by the licensee after the 1C Component
Cooling Water Pump Motor Bearings overheated and caught fire. A
resulting electrical ground caused an over current trip of the IDF
emergency bus.
Prior onsite inspector followup of this event is dis-
cussed in NRC Inspection Report No. 50-334/81-31.
LER 81-106/03L was submitted by the licensee after 7 of 15 Main Steam
Safety Valves experienced setpoint drift. Prior onsite inspector
followup of this event is discussed in NRC Inspection Report No.
50-334/81-31.
Denotes those reports selected for onsite followup as discussed belcw.
.
.
38
5.
Immediate Action Letter Followup - Operation With RCS Partially Drained
References:
--
(1)
Imediate Action Letter (IAL) 81-14, March 9,1981.
(2) Duquesne Light Company letter to the Nuclear Regulatory
Commission, January 6, 1982.
(3) Nuclear Regulatory Commission, Region I, letter to Duquesne
Light Company, January 8, 1982.
In March,1981, while the RCS was partially drained to mid-loop level,
inaccurate level indication and Residual Heat Removal Pump air binding
resulted in a partial loss of core flow. Licensee commitments to ensure
adequate Residual Heat Removal (RHR) system capability while in partially
drained conditions were documented in IAL 81-14.
Proposed changes to
those commitments based on a licensee evaluation of the event were
forwarded to the NRC in reference (2) for review and concurrence.
Clarifications of those comitments are contained in reference (3).
The licensee proposal and clarifications included redundant remote
level indications in the Control Room, requirements for instrument
checkout and periodic verification, data logging and parameter limits,
and actions for abnonnal conditions. On January 8 and 11 the inspector
reviewed implementation of the licensee commitments after the RCS had
been partially drained.
Operating Manual Change Notice (OMCN) issued
to support those commitments and reviewed by the inspector were:
OMCN 82-04 to BVPS OM Section 1.6.4 issued, January 7,1982.
OMCN 82-09 to BVPS OM Section 1.10.4, issued January 8, 1982.
OMCN 82-08 to BVPS OM Section 1.10.4, issued January 8, 1982.
OMCN 82-10 to BVPS OM Section 1.10.4, issued January 8,1982.
OMCN 82-05 to BVPS OM Section 1.10.4, issued January 8, 1982.
OMCN 82-06 to BVPS OM Section 1.10.4, issued January 8, 1982.
OMCN 82-07 to BVPS OM Section 1.54.3, issued January 8,1982.
The inspector also observed control board system alignments and
indications, held discussions with operations personnel, and
reviewed data to confirm licensee compliance with their commitments.
These items will receive continued review by the inspector during
routine inspection activities.
6.
Onsite Event Followup
a.
Automatic Actuation of Safety Injection System
A safety injection signal inadvertently actuated at 5:07
.m.,
January 4, 1982. MaintenanceSurveillanceProcedure(MSP 1.04,
Solid State Protection System, Train "A," Revision 16, was being
used as a guide to return the Solid State Protection System to
l
-
.
.
39
normal after replacing a fuse (replaced MB0-15 fuse with Buss-
man ABC-15 as specified by EM 41393) per MWR 800775. The safety
injection signal resulted from resetting the system out cf the
sequence specified in MSP 1.04.
Duringtheincident,thereactorwasinColdShutdown(Mode 5)
with two Reactor Coolant Pumps (RCP) and two Residual Heat Removal
(RHR) loops in operation. No actual water injection into the
Reactor Coolant System occurred due to Mode 5 (Cold Shutdown)
system lineups (pumps in pull-to-lock). However, the signal did
isolate the charging pump makeup flow path and the RCPs' No.1
Seal leakoff path (Containment Isolation Phase A). The two operating
RCPs were tripped to prevent possible damage, leaving the two
Residual Heat Removal loops in operation, meeting Technical Speci-
fication requirements (one loop in operation, and one lcop operable
in Moh 5). After the Safety Injection signal was confirmed to be
spurious, normal system alignments were reestablished.
The inspector discussed corrective actions taken by the licensee
with the Instrument and Control Supervisor. Actions taken included
revising MSP 1.04 for clarity and discussion of the event and the
importance of adhering to procedures during a safety meeting with
all Meter and Control Repainnen. The inspector had no further
questions on this item.
b.
Liquid Radwaste Piping Rupture
On January 19, 1982, about 3:30 a.m., water was observed leaking
down an exterior wall of the Steam Generator Drain Tank Building.
A recirculation line between Steam Generator Drain Tanks (LW-TK-
7A & -78) had ruptured, spilling about 300-500 gallons of water
on the roof, walls, and ground around the building. The line was
immediately isolated and the spillage stopped. Survey results
indicated maximum contamination in the surrounding earth of about
12,000 cpm as measured by an RM14/HP210 frisker. Liquid activity
from takn and spilled water samples was about 6E-5 uCi/cc. Most of
the spilled water had frozen on the walls and earth due to extremely
cold weather. The snow and ice surrounding the tank was drumed
and disposed of as liquid radwaste.
Post decontamination surveys
showed no contamination above the licensee's adninistrative limits
(450 uuCi/100 .m+2).
The line apparently ruptured due to freezing. The inspector observed
the failed line and portions of repair activities per MWR 820164.
The cause of freezing appears to be deficient temporary electric
.
heat tracing and damaged piping insulation, both of which were
also repaired.
.
40
J
The inspector became aware of the event about 8:00 a.m.
January
19, during a routine log review.
10CFR50.72, Notification of
Significant Events, requires prompt telephone notification to
NRC of any accidental radioactive release within one hour of
occurrence. The inspector reviewed this event relative to
10CFR50.72 and the licensee's reporting procedures of BVPS OM
Section 1.48.9.D.
At the time of discovery, the area affected
by the spill was apparently well defined in the snow around the
tank.
Discussions with licensee witnesses confimed that no
spillage to stonn drains was observed (abo iater confirmed by
sample) . The onduty Shift Supervisor considered the event to be
a " spill" rather than a " release" based on the absence of a path-
way to the environment. This appears appropriate based on the
BVPS FSAR which defines the restricted area (in the context of
10CFR20) as the company owned property at the Beaver Valley and
Shippingport sites.
The inspector questioned whether the absence of a release had
actually been confirmed within the one hour reporting time of
10CFR50.72. While not actually confimed by sample, the inspector
concluded that the licensee's assessment was reasonable. The
inspector, however, advised the Station Superintendent that a
notification to NRC per 10CFR50.72 would have been prudent based
en the circumstances. The Station Superintendent acknowledged
the inspector's concern and, on February 2,1982, issued Memo
NDlSSl:482 to Operations Department Supervisors emphasizing the need
for both internal (DLC) and external (NRC and other) notifications
for potentially reportable or sensitive matters.
The inspectors had no further questions on this matter; no
unacceptable conditions were identified.
c.
4160 VAC Cable Failure and Partial Loss of Offsite Power
While in Cold Shutdown, a 4160 VAC feeder cable failed at 2:15 p.m.,
January 27, 1982, resulting in arcing and smoldering insulation in
a switchgear room cable tray, and caused loss of 4160 VAC supply
to two of four (the A" and "B") main electrical busses and the
"A" Train (lAE) Emergency Bus. The "A" Train Emergency Diesel
Generator (EDG) is out-of-service for modification. The smoldering
insulation was immediately extinguished with portable equipment.
The "B" Train busses and EDG remained operable throughout the
event. The Reactor Coolant System (RCS) was partially drained.
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41
"B" Train safety systems were aligned for " priority train service"
and remained operable. The inspector confimed that Technical
Specifications for boration flowpath, onsite and offsite electrical
systems, residual heat removal, and reactivity control systems
were satisfied, including Action Statements as applicable. Power
was lost to the standby Residual Heat Removal (RHR) loop, RH_R
flow control valves, temporary Reactor Coolant System Level Instru-
ments, and miscellaneous instrumentation and controls.
Core flow
and level were maintained using local control and monitoring. The
cable fire, in the Service Building switchgear room, did nr
avolve
radiation or contamination.
Power was partially restored (to 480 VAC busses) via a temporary
cable about 9:15 p.m., January 27, restoring most of the instru-
ment and control functions. The "A" 4160 VAC Bus was reenergized
by a second temporary cable connection about 4:00 a.m., January
28.
Inspector review of the temporary cable connections is further
discussed below.
On January 28, the inspectors reviewed the sequence of events,
inspected the damaged cable trays, and reviewed the licensee's
response to the event. At the time of the incident, the "A"
and
"B" 4160 VAC busses were being powered via backfeed from the Main
Transformer through the 1C Unit Station Service (USS) Transfomers.
The fault, apparently a phase-to-phase short, occurred in the feeder
cables between the 1C USS Transformer and the "A" 4160 VAC Bus.
Inspector review of Sequence of Events computer printouts and
discussions with onduty operators established that the electrical
fault protection equipment operated as designed. The operators
properly responded to the event in accordance with the Abnormal
Procedures of the BVPS Operating Manual, Sections 1.36-1.39.
Power was lost to RHR flow control valves and temporary RCS level
to manually adjust RHR flow if needed (y dispatched to containment
instruments. Operators were immediatel
valves failed as-is) and
to monitor the local RCS level standpipe. Both temporary RCS
level instruments were powered from the "A/AE" busses via Instru-
ment Bus 3 and were deenergized. The instruments use Channel 3
Steam Generator level transmitters circuitry, temporarily repiped
to indicate RCS loop level. The Channel 3 instruments, powered
from the same Instrument Bus, were selected because they are the
only channels equipped with Control Rod recorders. The
inspectors confirmed that both RCS level and flow were properly
controlled and monitored through the event.
On the evening of January 27, 1982 the licensee issued and im-
plemented two Temporary Operating Procedures (TOPS) to provide
temporary power supplies to the "A" Train electrical busses.
TOP 82-5, Temporary Supply of Power to MCCl-E9, provided for
temporary cross connection of the 480 VAC emergency busses.
TOP 82-6, 4160 VAC Temporary Station Service Supply to AE
Emergency Bus, provided a similar cross connection for the 1A
4160 VAC bus. On January 28, the inspectors walked down each of
the installations to verify their confomance with the requirements
of the TOPS and good installation practice. The inspectors found
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42
that the temporary cables were of the specified size, properly
supported, and routed to avoid jeopardizing safety-related cables
in adjacent conduit and cable trays.
The inspector noted, however, that the temporary 480 volt power
supply passed through the wall between the 1 AE and 1 DF Vital
Switchgear Rooms via a standard conduit penetration. The fire
barrier packing for the penetration had apparently been removed
for the cable installation but not replaced. The matter is further
discussed in paragraph 3e(1) of this report.
On January 28, 1982, the inspectors discussed the licensee's safety
evaluation for the two TOPS with the DLC Chief Engineer, confirming
that conson mode failures, cable separation, and circuit protec-
tion were properly addressed by the procedures and the Onsite
Safety Committee's (OSC) review. On February 4, 1982, the
inspectors reviewed the draft safety evaluation included in the
draft OSC Meeting Minutes BV-0SC-5-82. The inspectors had no
further questions on this matter.
At the close of this inspection, temporary cables continued to
supply the "A" 4160 VAC normal and emergency busses. The 480
VAC temporary cross connect remained installed but deenergized.
Preparations for repair of the failed 4160 VAC cable were in
progress.
On January 27, 28, and 30, the inspectors observed the failed
cables and damaged cable trays.
Two cables appeared to be the source
of the failure (phase to phase and/or phase to ground). The cables
failed about midway between wooden supports and did not appear
to be in contact with each other at the point of failure.
Failure damage included evidence of heavy arcing with burnt cable
insulation and melted cable shield and conductors on both cables.
Evidence of arcing from the cables to the tray was also observed.
The cables are located in a horizontal covered tray about 20 feet
above floor level in the Service Building switchgear room. To the
extent permitted by the damage, the inspectors observed no
evidence of pre-failure damage or tampering. On February 2, 1982,
the licensee issued Engineering Memoranda (EM) 44398 and 20928
for repair of the damaged cables and 1C transformer respectively.
EM 44398 also provided for analysis of the failed cable by the
cable vendoc. At the close of this inspection the failure causes
for the cable and the potential for similar failures of other
cable had not been established, pending vendor analysis. This
matter will remain unresolved pending review of the licensee's
l
actions.
(82-01-10).
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d.
Spent Resin Spill
Approximately 1/2 gallon of spent resin was spilled in the Solid
Waste Area on February 5,1982. No personnel internal or external
contamination or offsite release occurred.
Temporary Operating Procedure (T0P) 82-07 Resin Removal and Trans-
fer of CH-I-1A, -1B, and -38, was in progress at the time of the
spill. The procedure involved 'he transfer of spent resin in slurry
form from the ion exchangers to a shipping line within a cask atop
a flatbed truck for offsite disposal.
Through discussions with licensee personnel, the inspectors deter-
mined that the spill occurred after the initial transfer was
essentially complete.
Involved personnel interviewed by the
inspectors stated that when the transfer pump reached shutoff
head due to backpressure from the full line, a residual amount
of resin remained in the transfer hose coupled to the liner. The
hose was disconnected and the residual resin drained into a poly
bag. The bag slipped out of the operator's hand, spilling resin
onto the outside of the cask, the bed, and tires. The inspector
reviewed TOP 82-07 and associated RWP 9516 package to evaluate the
licensee's preparations and preventive measures for coping with
possible spills. The inspector found the licensee preparations
generally acceptable including:
a confinement structure around
the SWA door and trailer; radiation barriers and postings; sealing
of storm sewers; butiding of floor dikes; placement of spill kits;
temporary ventilation and radiation monitors; and radtech coverage.
These efforts were successful in containing the actual spill until
decontamination was complete.
The inspector noted that TOP 81-07 did not provide guidance for
handling abnonnal conditions encountered during the transfer such
as unexpectedly high radiation levels or line plugging. Additional
improvements in the transfer rig such as isolation capability at
the cask end of the hose were also identified as desirable. The
DLC Radcon Supervisor acknowledged the inspectors'consnents and
stated that the procedure would be revised to include stop-work
provisions for abnormal conditions and transfer rig improvements
prior to its next use. These provisions would be included in the
licensee's programmatic procedures to ensure that they are con-
sidered for any similar evolution. This matter will remain un-
resolved ending NRC:RI review of the licensee's actions.
'
(82-01-11.
The inspector reviewed , pre- and post-spill survey and sample data.
Continuous air monitor samples varied from E-9 to E-10 uCi/m1,
,
and showed no increase of airborne activity above initial area
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44
background during the spill. Contact readings on the resin-filled
bag showed 15 R/hr.
Resin on the truck was reading approx-
imately 3-4 R/hr. at 6 inches, with a general area reading of
40-100 mR/H. Transfer and spill. response activities up to cask-
top decontamination and cask lid" installation resulted in personnel
exposures within the As Low As Re~asonably Achievable (ALARA) guide-
lines for RWP 9' 6.
Decontamination efforts perfomed under RWP
9501 (General Decontamination) were also accomplished with minimum
personnel dose.
When operations to seal the cask resumed, an additional written
ALARA review was not performed. Continuous health physics coverage
was, however, provided. The Radcon Foreman who surveyed the lid
area of the cask where some of the resin had spilled, was equipped
with two high range extremity dosimeters (one for each hand), a
low range dosimeter for whole body exposure control, and a TLD
badge. After assessing the situation, the foreman decided to per-
sonally complete the RWP 9516 work by manually cleaning the lid
sealing area of resin. The individual's low range pocket dosimeter
was periodically read during the work to control whole body exposure
and indicated a total exposure of 425 mrem. The individual's
thermoluminescent dosimeter (TLD), processed after the job showed
a total 1130 mrem whole body exposure for the calendar quarter,
900 mrem attributable to this activity. A calibration check of the
pocket dosimeter showed the instrument to be defective. The total
exposure for this individual is within the limits prescribed by
10CFR20, but did exceed the licensee's administrative guidelines
of 1000 mrem / quarter.
The high range dosimeters showed a total extremity dose of 4320
mrem gamma for the right hand and 3480 mrem gamma for the left
hand. These values were also within the 10CFR20 limits of 18 3/4
rcm per quarter.
The inspector discussed this event with Radcon Supervisor and
expressed his concet.' that work had been allowed to continue after
radiological conditions had substantially changed without parform-
ing an additional ALARA review. Though the Radiological Control
Manual Chapter 3 Radcon Procedure 8.1, Radiological Work Permit,
Revision 1, states that Radcon should initiate an ALARA review of
,
I
a dose of greater than 200 mR for an individual or 1000 mR for a
l
work party is expected, none is required. The inspector further
noted that Radcon Procedure 8.1 contains provisions for terminating
an RWP if warranted by changes in radiological status, as detemined
by Radcon personnel. The licensee acknowledged the inspector's
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comments, stating that future jobs would be carefully monitored to
prevent recurrence.
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The inspector observed portions a the decontamination of the
Solid Waste Area and cask truck, ,; ifying that radiological con-
l
trols were properly implemented.
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7.
Refueling Preparations - New Fuel Receipt
The inspector reviewed procedure NSQC 10.2, " Fuel Assembly and Shipping
Container Receipt Inspection," Revision 1, July 8,1980, to verify that
a technically adequate, approved procedure was used for the receipt,
inspection and storage of new fuel. No inadequacies were identified.
The inspector reviewed the Fuel Assembly Receiving and Inspection Reports
(NSQC 10.2, Attachment 6.1) for the 52 fuel assemblies received in
shipments DLCF-1 through DLCF-5, to verify that the receipt, inspection
and storage of the new fuel was accomplished in accordance with pro-
cedure NSQC 10.2.
No unacceptable conditions were identified by the
receipt inspectors. The inspector had no further questions in this area.
8.
IE Bulletin Followup
Licensee responses to IE Bulletins were inspected for timely submittal,
adequate corrective action, and dissemination to onsite management as
discussed below.
IE Bulletin 80-05:
Vacuum Condition Resulting in Damage to Chemical
and Volume Control System (CVCS) Holdup Tanks. The inspector reviewed
this bulletin and the licensee's response (DLC letter of June 16, 1980)
which adequately addressed the concerns to the bulletin for the Pressur-
izer Relief Tank, Primary Drains Transfer Tanks 1
and 2, Coolant
Recovery Tanks 4A and 4B, and the Volume Control Tank. The inspector
noted, however, that there were several other tanks which had not been
evaluated to detennine if they could be subjected to potentially dam-
aging vacuum conditions. The licensee agreed to extend the evaluation
to the tanks in question. The licensee's action in this regard will
be reviewed in a subsequent inspection.
IE Bulletin 80-18: Maintenance of Minimum Flow Through Centrifugal
Charging Pumps Following Secondary Side High Energy Line Rupture.
The inspector reviewed the licensee's response to this bulletin (DLC
letter of September 24,1980) and Emergency Operating Procedure E-0,
verifying that the bulletin had been received and evaluated, the
required modification made to prevent automatic closure of the coolant
charging pump (CCP) miniflow isolation valves on safety injection actua-
tion, and that emergency procedures had been modified to provide operator
guidance on opening and closing the CCP miniflow isolation valves. The
inspector had no further questions in this area.
IE Bulletin 80-23: Failure of Valves Manufactured by VALCOR Engineering
Corporation. The inspector reviewed the licensee's response to this
bulletin (DLC letter of December 15, 1980) which stated that no Valcon
solenoid valves were in use at Beaver Valley, Unit 1.
The inspector
reviewed the licensee's records documenting their review, consisting of:
.
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46
a memo itsting all safety-related solenoid valves and the manufacturers,
a memo of a phone contact with a Stone and Webster representative,
(Stone and Webster had been listed in the bulletin as receiving some of
the valves from Valcor), a memo of a phone contact with the W. J. Wooley,
Co. (airlock supplier), and a memo referencing a check of spare parts.
This documentation supports the licensee's conclusion that none of these
valves are used at the site. The inspector had no further questions in
this area.
9.
Review of TMI Action Plan Requirements
The inspector reviewed licensee implementation of TMI Action Plan Items
identified in NUREG 0737, Clarification of TMI Action Plan Requirements,
published November 1980, with respect to the licensee letters identified
below, the guidance of NUREG 0737, and other applicable documents as
referenced by NUREG 0737.
Item II.E.4.1, Dedicated Hydrogen Penetrations. The licensee's response
to NUREG 0378 guidance (DLC letter of January 26,1980), describes the
installation of the Hydrogen Recombiners at the site. A review of this
response, the licensee's FSAR Sections 6.5 and 14.3.4.4, and the BVPS
Operating Manual, Section 1.46, confimed that the licensee's installa-
tion meets the requirements of 10CFR50.44,10CFR50 General Design Criteria 54 and 56, and the NRR position as clarified in NUREG 0737, Item II.E.4.1.
Inspector walkdown of the penetration piping confimed the installation
to be as described by the licensee's submittals. The inspector had no
further questions in this area.
Item II.E.4.2.5a & b - Containment Isolation Dependability (Actuation
Setpoint). The licensee's submittal to NRC dated December 31, 1980
provided the bases for no reduction or modification to the containment
pressure setpoint for initiation of containment isolation. The submittal
discusses the facility's subatmospheric containment design and the con-
siderations for maintaining the existing setpoint. The inspector reviewed
BVPS Operating Manual Section 1.16 and 1.47, finding the infomation
contained in the letter to be consistent with existing system descrip-
tions and operating procedures. This item was reviewed by NRC:NRR and
found acceptable (NRC letter, Varga to Carey, dated December 11,1981).
Item II.E.4.2.7: Containment Purge and Vent Valves Close on High
Radiation. This requirement has been detemined not to be applicable
to Beaver Valley, Unit 1 because of the subatmospheric containment
design and was deleted by a letter from the NRC to the licensee dated
April 29,1981. Subatmospheric containment designs do not pemit open-
ing purge and exhaust valves except in Cold Shutdown or Refueling con-
ditions.
Containment purge and exhaust valves are equipped with high
radiation isolation signals for these conditions.
Containment vacuum
pump lines are separately equipped with isolation signals.
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10.
Emergency Preparedness Exercise - Personnel Accountability Drill.
On February 12, 1982 the inspectors witnessed portions of the licensee's
accountability drill, performed to meet the requirements of the BVPS
Emergency Preparedness Plan, Section 6.7.5 and NUREG 0654, Section J.
The drill was conducted separately from the full scale NRC/ FEMA drill
of February 17, 1982 to minimize the impact of the current outage on
both exercises. About 1000 additional non-DLC employees are currently
onsite for the refueling / modification outage and were not included in
the accountability drill.
Inspector drill observations were made from the Primary Assembly Area
(Men's Locker Room), the Control Room, the Central Alarm Station, and
the Nuclear Division administration building, with respect to:
--
EPP/ Implementing Procedure (IP) 7.2, Administration of EPP Drills
and Exercises, Issue 6, Revision 3;
EPP/IP 3.1.3, Administration Building Evacuation, Issue 6, Revision
--
4; and
EPP/IP 3.2, Personnel Accountability, Issue 6, Revision 4.
--
The drill was to include limited participation by BVPS Unit 2 con-
struction personnel (emergency / accountability coordinators).
Partici-
I
pation was less than expected due to a misunderstanding of drill
'
announcements and poor PA system performance. Unit 2 activities were
observed by the BVPS Unit 2 resident inspector.
The inspectors attended the licensee's post drill critique, confiming
thaj licensee observer findings were consistent with the inspectors'.
The licensee met the accountability goal of 30 minutes specified by
NUREG 0654. The inspectors' findings will be included in NRC:RI
Inspection Report 50-334/82-04 for the full scale emergency response
exercise.
11.
Containment Hydrogen Recombiner Design Deficiencies
On December 15, 1981, Rockwell International, manufacturer of the
BVPS-1 recombiners, notified NRC:HQ and the licensee of deficiencies
in recombiner heater lead wire insulation. The vendor's evaluation
concluded that nomal testing combined with actual post-LOCA operation
would result in failures after ten years in nomal service. The
BVPS-1 recombiners were delivered in 1974
The vendor will provide
the licensee instructions for correction of the deficiencies.
The inspector confirmed that the licensee had received the vendor's
notification and had initiated Station Modification Request (SMR) 543,
dated February 3,1982, to accomplish the vendor's recomendations.
The SMR notes that action must be completed prior to 1984. The
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48
inspector also confinned that the modification requirements had been
entered into the BVPS Commitment Action Control System, responsibility
assigned for completion, and a required completion date of December
30, 1983 established.
12.
Safety Injection Actuation Design
Because of a recent design error found at the Maine Yankee Nuclear
Pcwer Station which would have prevented actuation of the Safety
Injection System for certain single active components failures, the
inspector reviewed the design of the Westinghouse safeguards actuation
circuitry for BVPS-1 with the Westinghouse Site Representative. This
review verified that a similar design error does not exist in this
circuitry. The inspector had no further questions in this area.
13.
Review of NDE Techniques for Plant Modification Piping
On February 2-3, 1982, a region based inspector reviewed licensee and
licensee contractor actions taken in response to Schneider Power
Corporation Nonconformance Report (NCR) 288, regarding use of DLC Radio-
graphic Procedure RT2 for radiography of USAS/ ANSI B31.10 (1967) pipe
welds. Procedure RT2 has been used on examination of field pipe welds
used for safety realted modifications in progress at BVPS-1.
NCR 288, dated December 23, 1981, indicated that the determination of
sensitivity in B31.1 (which references ASME SCI) is different than that
required in RT2. The penetrameter requirements are more severe in B31.1.
The licensee conducted a series of radiographic tests utilizing the same
techniques used for the affected pipe welds on pipes of similar wall
thickness with the B31.1 and RT2 (SCIII) penetrameters. The inspector
reviewed the radiographs and Radiographic Interpretation Reports dated
January 27, 1982 for 6" x 0.432", 10" x 0.365", 3" x 0.360", 6" x 0.134",
and 6" x 0.240" tests. The radiographic techniques used are capable
of meeting the sensitivity levels required by both B31.1 and RT2. The
quality of the weld joints is therefore acceptable to B31.1. A review
of the assignment of allowable stresses in ASME SCIII and B31.1 indicates
that a higher allowable stress for SA106C is permitted in SCIII Table
J-1.1 (19.4 KSI 0 650 F) than is permitted for B31.1 (17.5 KSI O 650 F)
indicating further conservatism for B31.1 fabrication when compared with
SCIII.
(The higher allowable stress is permitted with the lower radio-
l
graphic sensitivity).
The inspector reviewed DLC Memo DLCQA-3459 dated January 29, 1982 on
,
the evaluation of the sample radiographs finding the licensee's evalua-
l
tion and actions acceptable.
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15.
Review of Periodic Reports
The inspector reviewed the BVPS Monthly Operating Reports for October-
December,1981 to verify that information required to be reported by
NRC requirements had been included. No inadequacies were identified.
16.
Unresolved Items
Unresolved items are matters about which more information is required
to determine whether they are acceptable, items of noncompliance or
deviations. Two unresolved items were identified and are discussed
in paragraphs 6.c and 6.d of this report.
17.
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 conclusion of the report period.
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