ML20023D976
| ML20023D976 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 05/17/1983 |
| From: | Lester Tripp, Troskoski W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20023D959 | List: |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-1.C.6, TASK-TM 50-334-83-08, 50-334-83-8, IEB-82-01, IEB-82-02, IEB-82-03, IEB-82-04, IEB-82-1, IEB-82-2, IEB-82-3, IEB-82-4, IEB-83-02, IEB-83-04, IEB-83-2, IEB-83-4, NUDOCS 8306060214 | |
| Download: ML20023D976 (22) | |
See also: IR 05000334/1983008
Text
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U. S. NUCLEAR REGULATORY CO MISSION
REGION 1
Report No.
50-334/83-08
Docket No.
50-334
License No.
Priority
Category
C
--
Licensee:
Duquesne Light Company
435 Sixth Avenue
Pittsburgh, Pennsylvania
,
Facility Name:
Beaver Valley Power Station, Unit 1
Inspection at:
Shippingport, Pennsylvania
Inspection Conducted: April 5 - May 2,1983
Inspector:
5
. 4@/)
7
3
. M. Tro&Koski, Resident Inspector
date signed
/
Approved by:
/-
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73
L. E. Tripp, Chief, Reactor Projects
date signed
Section No. 2A, Reactor Projects
Branch 2
Inspection Summary:
Inspection an April 5 - May 2,1983 (Inspection No.
50-334/83-08).
Areas Inspected:
Routine inspections by the resident inspector (100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />) of:
licensee action on previous inspection findings, plant operations, housekeeping,
fire protection, radiological controls, physical security, maintenance
activities, surveillance testing, engineered safety features verification,
refueling preparation, radioactive waste transportation activities, safety
and quality classification of equipment, changes, tests and experiments
program, and IE Bulletins.
Results: One violation (failure to demonstrate ECCS valve operability
within specified surveillance interval - detail 5).
8306060214 830523
PDR ADOCK 05000334
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DCS NUMBERS
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821229
821014
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810519
821028
800822
821203
830322
830304
830328
830311
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820331
830329
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820507
830502
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820602
830420
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DETAILS
1.
Persons Contacted
F. Bissert, Manager, Nuclear Support Services
J. Carey, Vice President, Nuclear Division
M. Coppula, Superintendent of Technical Services
K. Grada, Superintendent of Licensing and Compliance
R. Hansen, Maintenance Supervisor
J. Indovina, I&C Supervisor
T. Jones, Manager, Nuclear Operations
J. Kosmal, Radiological Operations Coordinator
W. Lacey, Station Superintendent
V. Linnenbom, Radiochemist
J. Lukehart, Security Director
L. Schad, Operations Supervisor
E. Schnell, Radcon Supervisor
J. Sieber, Manager, Nuclear Safety and Licensing
R. Swiderski, Superintendent of Nuclear Construction
N. Tonet, Manager, Nuclear Engineering
T. Zyra, Plant Performance and Testing Supervisor
The inspector also contacted other licensee employees and contractors
during this inspection.
2.
The NRC Outstanding Items (0I) List was reviewed with cognizant licensee
personnel.
Items selected by the inspector were subsequently reviewed
through discussions with licensee personnel, documentation review and
field inspection to detennine whether licensee actions specified in
the OIs had been satisfactorily completed. The overall status of
previously identified inspection findings were reviewed, and planned
and completed licensee actions were discussed for those items reported
below.
(Closed) Violation (82-25-03): Failure to review and approve tank curve
notebook. By response dated December 29, 1982, DLC committed to review
the accuracy of the reference curves of tanks used for liquid waste
discharges and surveillance testing, and to submit the entire curve book
to the Onsite Safety Conmittee for review and subsequent approval by the
Station Superintendent. The inspector reviewed the control room tank
curve book and verified that it was approved and controlled per OM
Chapter 1.48.9, Procedure M, Revisions or Reissues to Operating Manual
Flow Diagrams. This item is closed.
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(Closed)UnresolvedItem(81-04-01):
Review DLC evaluation of PAB
ventilation system perfomance. This item was to provide further
review of a November 25, 1980 event (discussed in NRC Inspection Report
No. 50-334/80-30) where the maximum pemissible concentration of Xenon
and various daughter products in a restricted area was detemined to
be about 28%, due to primary leakage from the reactor coolant system
filter. Completion of corrective actions to insure design performance
of the PAB ventilation system were documented in that inspection report.
In addition to those items, the licensee rebalanced PAB ventilation
system and attempted to correlate any higher than normal background
radiation / airborne radioactivity levels to specific plant evolutions.
Preliminary infomation was assembled by the Radcon staff and forwarded
to Operations for identification of process evolutions occuring during
the same time period. No meaningful correlation could be obtained.
During the course of 1983, the inspector observed several instances
where Xenon and various decay products were detected in the PAB.
In
each case, the maximum concentrations were significantly below MPC
limits. Because those particular isotopes decay to background levels
in approximately 30 minutes and no radiological safety concern is present,
this item is closed.
(Closed) Unresolved Item (81-29-01): Bases for 10 CFR 50.59 procedure
reviews not documented by OSC. The inspector reviewed the Onsite
Safety Committee meeting minutes of 1983 and verified that written
evaluations of new procedures, or changes to existing procedures, were
provided per the provisions of 10 CFR 50.59 that detail the bases for
detemining that no unreviewed safety question is involved. Administrative
chaages to procedures that did not affect the intent were so identified
on a generic basis. Those written evaluations now allow the Offsite
Review Committee to better perform their subsequent TS 6.5.2.7 required
reviews. This item is closed.
(Closed) Unresolved Item (80-24-01):
Inadequate review of design changes
by OSC for technical specification impact. Various design change packages
and associated safety analysis were reviewed and accepted by the OSC
without properly identifying technical specification changes needing prior
Commission approval. Subsequent inspection of this area indicated that
the design concepts continued to evolve after the preliminary safety
evaluation was perfomed; some to the point where they impacted on the
technical specifications. To correct this program deficiency, the
licensee revised Station Engineering Procedure 2.3, Design Change
Coordination, to require a new safety evaluation on any design concept
that is substantially changed after the initial review. The new safety
evaluation is then forwarded to the OSC for their review and approval.
The inspector reviewed a sampling of DCPs performed in 1982 to verify
proper program implementation. Additionally, the licensee has performed
a systematic review of all pre-refueling Category 1 design changes per
the revised SEP 2.3 criteria (as detailed in Unresolved Item 80-09-14
and discussed in this report section). This item is now closed.
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(Closed)UnresolvedItem(80-09-14): DLC to access quality of as-built
infomation. To access the quality and accuracy of station drawings,
procedures, and engineering infomation, the licensee perfomed a
systematic review of all pre-refueling Category 1 design changes. This
,
review was conducted per Station Engineering Procedure 2.3, Design
!
Change Coordination, which identifies and assigns responsibility for
!
various turnover activities to applicable station groups. The inspector
reviewed selected DCPs referenced in the Engineering Matrix Program
and confimed that the reviews (including TS, SAR, and 10 CFR 50.59)
were conducted between March and July,1981. This item is closed.
(Closed) Unresolved Item (80-24-02): Adequate content of bases used
for unreviewed safety questions. A previous inspector concern questioned
the adequacy of the documented basis provided for the 10 CFR 50.59 review
of several design change packages. The licensee infomed the inspector
that in the future, guidance provided by IE Circular 80-18, Safety Evaluations
!
for Changes to Radioactive Waste Treatment Systems, would be employed
during the preparation of all. bases used to detemine whether or not the
design change involved an unreviewed safety question. Both SEP 2.3, for
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the Station Engineering Group and NEMP 2.3, for the Nuclear Engineering
Division, were revised to provide this guidance for evaluating each
proposed station modification. This item is closed.
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(Closed)UnresolvedItem(82-11-01): Technical Specification Table 3.6 - 1
to be updated to reflect modifications to the containment system boundries.
!
During containment local leak rate testing conducted during May,1982, the
inspector noted that several plant modifications necessitated a change to
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TS Table 3.6-1. Containment Penetrations. TS Amendment No. 65 approved
March 22, 1983, waives the need for Type C testing of certain valves that
do not represent potential containment atmosphere leakage paths, and
therefore, are not subject to 10 CFR 50, Appendix J, Section C require-
ments. This item is closed.
'
3.
Plant Operations
!
a.
General
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Inspection tours of the plant areas listed below were conducted
during both day and night shifts with respect to Technical Spec-
ification (TS) compliance, housekeeping and cleanliness, fire
protection, radiation control, physical security and plant pro-
.
tection, operational and maintenance administrative controls.
Control Room
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-- Primary Auxiliary Building
Turbine Building
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Service Building
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-- Main Intake Structure
Main Steam Valve Room
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Purge Duct Room
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East / West Cable Vaults
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-- Emergency Diesel Generator Rooms
Containment Building
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Penetration Areas
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Safeguards Areas
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Various Switchgear Rooms / Cable Spreading Room
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Protected Areas
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Acceptance criteria for the above areas include the following:
BVPS FSAR Appendix A, Technical Specifications (TS)
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-- BVPS Operating Manual (0M), Chapter 48, Conduct of Operations
OM 1.48.5, Section D, Jumpers and Lifted Leads
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OM 1.48.6, Clearance Procedures
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OM l.48.8, Records
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OM 1.48.9, Rules of Practice
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-- OM Chapter 55A, Periodic Checks - Operating Surveillance Tests
BVPS Maintenance Manual (MM), Chapter 1, Conduct of Maintenance
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10 CFR 50.54 (k), Control Room Manning Requirements
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-- BVPS Site / Station Administrative Procedures (SAP)
BVPS Physical Security Plan (PSP)
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Inspector Judgement
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b.
Operations
The inspector toured the Control Room regularly to verify compliance
with NRC requirements and facility technical specifications (TS).
Direct observations of instrumentation, recorder traces and control
panels were made for items important to safety.
Included in the
reviews are the rod position indicators, nuclear instrumentation
systems, radiation monitors, containment pressure and temperature
parameters, onsite/offsite emergency power sources, availability
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of reactor protection systems and proper alignment of engineered
safety feature systems. Where an abnormal condition existed (such
as out-of-service equipment), adherence to appropriate TS action
statements were independently verified. Also, various operation
logs and records, including completed surveillance tests, equip-
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ment clearance permits in progress, status board maintenance and
temporary operating procedures were reviewed en a sampling bases
for compliance with technical specifications and those administrative
controls listed in paragraph 3a.
During the course of the inspection, discussions were conducted
with operators concerning reasons for selected annunciators and
knowledge of recent changes to procedures, facility configuration
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and plant conditions. The inspector verified adherence to approved
procedures for ongoing activities observed. Shift turnovers were
witnessed and staffing requirements confirmed. Except as noted
below, inspector comments or questions resulting from these daily
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reviews were acceptably resolved by licensee personnel.
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.
(1) During perfomance of OST 1.1.11, Safeguards Protection
System Train A Test, on April 11, 1983, safety injection
transfer relay K 641, failed a contact continuity check.
Investigation indicated that M0V-SI860A (low head SI pump
suction valve from containment) themal overload had tripped.
The themal overload relay was checked and Preventive
Maintenance Procedure 1-75-MOV-lE was performed on the
valve breaker, but the problem could not be duplicated.
The OST was reperformed satisfactorily and the system
declared operable.
A review of the equipment history
list did not identify any similar failures. The inspector
had no further concerns.
(2) From about April 10th to 17th, 1983, the containment sump
pump-out rate (as indicated on FTO-DA-102) increased from
about 1 gallon per minute to approximately 2 gallons per
minute. The licensee conducted a containment entry on
April 16, 1983, to identify the source of increased in-
leakage. The suspected leak path was identified as coming
from valve packing on a manual isolation valve (80-14) in
the steam generator blowdown line at the B cubicle 738'
elevation. This leakage was in the fom of steam from the
secondary side and did not present a radiological hazard
as indicated by no measured increase in containment parti-
culate or gaseous activity. The licensee informed the
inspector that another containment entry was planned in
an attempt to Furmanite the valve packing. The inspector
discussed licensee plans for maintaining exposures during
this job to levels that are as low as reasonably achievable
(ALARA). Radiation Work Permit 11008 and associated documentation
were reviewed to verify that proper containment air sampling
was made to calculate the maximum pemissible concentration
(MPC) hours, that exposure limits were properly reviewed and
authorized and that contractor personnel were properly rad-
worker qualified. Pocket dt.,imetry records indicated that
the total man-Rem exposure for the job was approximately
1.4 R.
When comparing this data to the initial pre-job
survey of the area which indicated a 5 to 7 R field, the
licensee's pre-planning efforts were effective in minimizing
total exposure. No individual
exceeded the DLC administrative
guidelines. Additionally, the results of the maintenance
efforts were successful in reducing the secondary leakage
rate back down to about 60 gallons per hour.
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(3) On April 21, 1983, it was detennined that 26 of 50 axial
flux thimbles were blocked. Technical Specification 3.3.3.2, Moveable Incore Detectors, requires that at least
50% of the detector paths are passable, whenever the system
is used to:
(1) recalibrate the axial flux offset system;
(2) monit'.,r the quadrant power tilt ratio; or, (3) to measure
the Nuclear Enthalpy and Heat Flux hot channel factors.
Through discussicns with the reactor engineer, the inspector
detennined that the next required hot channel factor measure-
ments and recalibration of the axial flux offset detector
system would be due on May 16, 1983. Monitoring of the
quadrant power tilt ratio by this system is not necessary
as all four power range nuclear instruments are operable.
This item will receive continued attention.
(4) At about 4:45 a.m. on April 29, 1983, boronometer relief
valve RV-CH-103 (located down stream on the nonregenerative
heat exchanger in the 722' elevation of the PAB) lifted,
releasing a small amount of RCS water to the PAB sump that
off-gased an estimated 3.8 curies of Noble gas (Xe-133).
No iodines were detected and maximum particulates (Rb-88,
Cs-138) were 8 E-8 microcuries per cc.
PAB ventilation
monitors increased about 250 counts per minute above back-
ground before tailing off to normal levels at 8:00 a.m.
There was no spread of contamination and gas samples
indicated less than minimum detectible activity (about 3.6
E-8 microcuries per cc).
The inspector reviewed the various
gas samples taken during the event. Calculations indicated
that the gaseous release was less than 1.5% of the
instantaneoas technical specification values and less than
1.7% for skin dose.
Regional NRC health physics specialists
had no further radiological concerns about this event.
During this event, the licensee was performing two compatable
evaluations in parallel; an RCS inventory calculation per
OST 1.6.2, and an RCS dilution to maintain T-average. This
involved filling the volume control tank (VCT) to its upper
level with primary makeup water and isolating the degasifier
flow path. Total head on the boronometer (recently unisolated
for corrective maintenance) was estimated to be no more than
65 psi when RV-CH-103 lifted
(normal lift point should have
been 200 psi), setting off PAB radiation monitors. The boron-
ometer was reisolated to prevent RV-CH-103 from failing again.
Appropriate NRC notifications were made via the ENS system.
Licensee actions were satisfactory.
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c.
Plant Security / Physical Protection
Implementation of the Physical Security Plan was observed in the
areas listed in paragraph 3a above with regard to the following:
Protected area barriers were not degraded;
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Isolation zones were clear;
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Persons and packages were checked prior to allowing entry
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into the Protected Area;
Vehicles were properly searched and vehicle access to the
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Protected Areas was in accordance with approved procedures;
Security access controls to Vital Areas were being maintained
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and that persons in Vital Areas were properly authorized;
Security posts were adequately manned, equipped and security
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personnel were alert and knowledgeable regarding position
requirements, and that written procedures were available; and,
-- Adequate lighting maintained.
The inspector identified no deficiencies.
d.
Radiation Controls
Radiation controls, including posting of radiation areas, the
conditions of step-off pads, disposal of protective clothing,
completion of Radiation Work Permits, compliance with Radiation
Work Pennits, personnel monitoring devices being worn, clean-
liness of work areas, radiation control job coverage, area monitor
operability (portable and pennanent), area monitor calibration,
and personnel frisking procedures were observed on a sampling basis.
No violations were identified.
e.
Plant Housekeeping and Fire Protection
Plant % sekeeping conditions including general cleanliness
conditions and control of material to prevent fire hazards were
observed in areas listed in paragraph 3a. Maintenance of fire
barriers, fire barrier penetrations, and verification of posted
fire watches in these areas was also observed. No inadequacies
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were observed.
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f,
Chemistry Sampling Program
The inspector reviewed DLC's chemistry log sheets for technical
specification required sampling of the reactor coolant system
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chemistry (dissolved oxygen, chloride, fluoride), gross activity,
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and dose equivalent Iodine - 131 for the period of April 1 - 25,
,
1983. Additionally, boric acid sampling logs of the BIT, RWST,
SI accumulators, and boric acid storage tanks were also audited
for~that time period. From this review, it was determined that
the licensee's sampling program was being properly implemented
per applicable technical specification requirements and that the
sampling frequency and individual parameters were within limits.
No discrepancies were noted.
g.
Liquid Waste Discharges
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References:
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1.
Updated FSAR, Section ll, Radioactive Waste and Radiation
Protection.
2.
Regulatory Guide 1.21, Measuring, Evaluating and Reporting
Radioactivity in Solid Wastes and Releases of Radioactive
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Material in Liquid and Gaseous Effluents from Light Water
Cooled Nuclear Power Plants.
3.
BVPS Technical Specification Amendment No. 66.
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4.
10 CFR 50 Appendix I, Numerical Guides for Design Objectives
in Limiting Conditions for Operations to meet the Criterion
"As Low As is Reasonably Achievable" for Radioactive Material
in Light Water Cooled Nuclear Power Effluents.
5.
Operating Manual, Chapter 1.17, Liquid Waste Disposal System.
Amendment No. 66 to the BVPS Technical Specifications updates the
radiological effluent technical specifications that are outlined
,
in the ETS. This amendment is effective as of March 28,1983, and
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is scheduled to be fully phased in by January 1, 1984.
It imple-
ments the requirements of Appendix I to 10 CFR 50 and establishes
new limiting conditions for operation for the quarterly and annual
average release rate and revises environmental monitoring programs
to assure conformance with Commission regulations. The inspector
<
conducted discussions with cognizant licensee representatives on
phasing in the various aspects of the new RETS program, in particular,
i
with respect to the Offsite Dose Calculation Manual and the radioactive
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waste discharge program. Responsibilities and authorities have
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already been assigned in this regard to meet the January 1,1984,
,
deadline. This item will receive further review as the program
progresses.
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The duty on the liquid waste system was increased due to the higher
than nomal containment sump pirip-out rate (130-140 gallons per
hour) in April. The inspector reviewed liquid radioactive waste
discharge authorization (RWDA) Numbers 2220 to 2253 to verify
chemistry limits on chromates, minimum dilution flows, proper
recirculation time prior to sampling, rad monitor alarm setpoint
calculations, isotopic sampling and recording, and proper reviews
and approvals from the Operations, Radcon and Chemistry Departments
were obtained per the above administrative guidelines. The inspector
also reviewed the requirements of Regulatory Guide 1.21 to verify
that they were properly translated into the Operating Manual
Chapter 17 Procedures. Licensee actions were found to be consistent
with those requirements.
4.
Engineered Safety Features (ESF) Verification
A.
The operability of the low head safety injection, chemical addition,
containment depressurization and cooling systems were verified by
performing a walkdown of accessible portions that included the
following as appropriate:
1.
System lineup procedures match plant drawings and the as-
built configuration.
2.
Equipment conditions were observed for items which might
degrade perfomance. Hangers and supports are operable.
3.
The interior of breakers, electrical and instrumentation
cabinets were inspected for debris, loose material, jumpers,
etc.
4.
Instrumentation was properly valved in and functioning; and
had current calibration dates.
5.
Valves were verified to be in the proper position with power
available. Valve locking mechanisms were checked, where
required.
6.
Technical Specification required surveillance testing was
current.
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Findings:
(1) Technical Specification 4.6.2.3 requires that each valve (manual,
power operated, or automatic) in the chemical addition flow path
that is not locked, sealed or otherwise secured in position, be
checked for correct alignment on a 31 day frequency. A review
of OSTs related to OM Chapter 13, Containment Depressurization
System, and the Pad Lock log book indicated that manual valve
QS-186 (the manual isolation valve to the chemical addition
tank, QS-TK-2) was not listed as being checked on a 31 day
frequency. Through discussions with the Operations Supervisor,
the licensee was able to demonstrate correct alignment by virtue
of running OST 1.13.10A(B), Chemical Addition System Valve
Position and Pump Operability Check - Train A(B), on a monthly
frequency (staggered every other week between ESF trains).
NUREG-0737, Clarification of TMI Action Plan Requirements Section
I.C.6, Guidance on Procedures for Verifying Correct Performance
of Operating Activities, provides guidance for the return to service
of equipment important to safety that includes proper system align-
ment verification by second qualified operator or perfomance of
a functional test which can prove that all equipment, valves and
switches involved are correctly ali
Because line flow is
fully demonstrated by OST 1.13.10A(gned.B),theinspectordetermined
that the licensee is in compliance with the surveillance require-
ment, and has no further questions on this item.
(2) During perfomance of a low head safety injection system walkdown
on April 22, 1983, the inspector entered the Safeguards Valvc Pit
and noted that M0V-SI862 A and B, (low head SI pump suction valves
from the RWST) had boric acid buildup on the valve bolts. The
inspector brought this to the licensee's attention and reviewed
station efforts to identify, evaluate and clean up boric acid
buildup on safety related valves and equipment.
For the primary reactor coolant system, residual heat removal
system and portions of the safety injection system located inside
containment, IE Bulletin 82-02 specifies required actions to be
undertaken by the licensee during the upcoming refueling outage
scheduled for June,1983 (see detail 8 of this inspection report).
To assure that high energy ECCS lines and associated boundries
outside of containment do not degrade, a quarterly walkdown is
performed per OST 1.48.2, High Energy Line and ECCS Inspection
(last completed March 26, 1983).
Identified deficier.:ies are
corrected per the maintenance work request system. Additionally,
QC is scheduled to perform an ASME Code Section XI inspection
during the planned June outage. The scope and content of these
inspections are currently being reviewed by the inspector.
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5.
Surveillance Activities
To ascertain that surveillance of safety-related systems or components
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is being conducted in accordance with license requirements, the
inspector observed portions of selected tests to verify that:
a.
The surveillance test procedure confoms to technical
specification requirements.
b.
Required administrative approvals and tagouts are obtained
before initiating the teste
c.
Testing is being accomplished by qualified personnel in
accordance with an approved test procedure.
d.
Required test instrumentation is calibrated.
e.
LCOs are met.
f.
The test data are accurate and comnlete. Selected test
result data was independently reviewed to verify accuracy.
g.
Independently verify the system was properly returned to
service.
h.
Test results meet technical specification requirements and
test discrepancies are rectified.
i. The surveillance test was completed at the required frequency.
OST 1.13.2,1B Quench Spray Pump Flow Test, perfomed
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April 6,1983.
OST 1.11.6, ECCS Flow Path and Valve Position Check
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(LHSI Loop A), perfomed April 21, 1983.
OST 1.13.10B, Chemical Addition System Valve Position and
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Pump Operability Check - Train B, performed April 26, 1983.
MSP 2.09, Nuclear Instrument Source Range Calibration,
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perfomed May 2,1933.
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Findings:
(1) The cover sheet of OST 1.11.6 references TSs 4.5.2.b.3, 4, 5
and 4.5.3 for demonstrating operability of the ECCS. TS 4.5.2.b.5 requires the licensee to verify that each ECCS sub-
system is aligned to receive electrical power from separate
operable emergency buses; yet the purpose of OST 1.11.6 is only
to exercise specified valves and verify their correct position.
Electrical bus alignments are checked per OM Chapter 36
surveillance tests. DLC is currently reviewing procedures that
perfom TS surveillance requirements to verify proper review and
approval. This is being tracked as unresolved item 83-07-04 and
should result in the development of a matrix that cross-references
each surveillance requirement with an appropriate procedure.
Verification that the OST cover sheets accurately reflect
actual TS surveillance requirements tested is an unresolved item
(83-08-01).
(2) The inspector had previously reviewed Section 6, Engineered Safety
Features, of the Updated FSAR, OM Chapter 11 and related OSTs, and
P&ID No. 8600-RM-167A as part of the low head safety injection
system walkdown. During perfomance of OST 1.11.6, it was noted
that MOV-SI 862 A and B, the low head SI pump suction valves
from the RWST, were not being cycled on a 31 day frequency as
required by TS 4.5.2.b.3, but on a quarterly basis as part of
the inservice inspection program per OST 1.47.3A, Three Month
Containment Isolation and ASME Section XI Tests. These valves
would be required to automatically close during changeover to
the recirculation phase of a safety injection. Failure to cycle
each testable power operated valve in the ECCS flow path on a
31 day frequency is a Violation (83-08-02) of TS 4.5.2.b.3.
6.
Maintenance Activities
The inspector observed portions of selected maintenance activities on
safety-related systems and components to verify that those activities
were being conducted in accordance with approved procedures, technical
specifications and appropriate industrial codes and standards. The
inspector conducted record reviews and direct observations to determine
that:
- Those activities did not violate a limiting condition
for operation.
- Redundant components were operable.
- Required administrative approvals and tagouts had been
obtained prior to initiating work.
- Approved procedures were used or the activity was within
the " skills of the trade."
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- The work was performed by qualified personnel.
- The procedures used were adequate to control the activity.
- Replacement parts and materials were properly certified.
- Radiological controls were properly implemented when necessary.
- Ignition / fire prevention controls were appropriate for the
activity.
- QC hold points were established where required and observed.
- Equipment was properly tested before being returned to service.
- An independent verification was conducted to verify that the
equipment was properly returned to service.
(1) Continued repa'r/rebuildir
f river u ter pump WR-P-1A
per MWR 807135.
7.
Refueling Preparation
>
The inspector reviewed DLC quak ty assurance audits, BV-1-82-41, 83-03,
and 83-17, that evaluated the Wstinghouse Nuclear Fuel' Division's
compliance with those requiremeats contained in the DLC QA Nuclear
Fuel Program and Westinghouse WCAP 7800, Nuclear Fuel Quality Assurance
Program. These onsite audits covered fuel pellet manufacturing,
vendor qualifications, design controls, testing, inspection, account-
ability and shipping, and fulfill DLC's obligation to conduct onsite
inspection of the nuclear fuel vendor.
8.
IE Bulletins
Licensee actions on IE Bulletins issued in 1982 and 1983 were inspected
to verify that they were reviewed by licensee management for applicability,
and appropriate corrective actions were taken or scheduled for those
items pertaining to BVPS, as specified in Nuclear Division Directive
No.13, Administration of NRC Bulletins, Circulars and Information Notices.
1.
IEB 82-01:
Rev. 1, Alteration of Radiographs of Welds in Piping
Subassemblies. This bulletin was received by the licensing and
compliance group and forwarded to the Quality Assurance and Quality
Control Departments for general information and review by cognizant
personnel in documented training sessions. This bulletin is closed.
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2.
IEB 82-03: Stress Corrosion Cracking in Thick-Wall, Large-
Diameter Stainless Steel, Recirculation System Piping .
at BWR Plants. This bulletin pertaining to operating boiling
water reactors, was forwarded to DLC for information only.
Cognizant licensee personnel received and reviewed this bulletin.
3.
IEB 82-04:
Deficiencies in Primary Containment Electrical
Penetration Assemblies. Review by the Nuclear Engineering Division,
documented in EM 30214, January 10, 1983, and field observations by
the inspector, detennined that no Bunko Ramo primary containment
electrical penetration assemblies are used at BVPS (Viking assemblies
are employed). This IE Bulletin was added to the engineering equip-
ment qualification file by the licensee for future reference on
possible procurement and installation of the subject assemblies.
This bulletin is closed.
4.
IEB 83-02:
Stress Corrosion Cracking in Large-Diameter Stainless
Steel Recirculation System Piping at BWR Plants. This is an
infonnation type bulletin not directly applicable to the Beaver
Valley Power Station, that was reviewed for information purposes
only.
5.
IEB 83-04: Failure of Undervoltage Trip Function of Reactor Trip
Breakers. This bulletin is not applicable to Beaver Valley Power
Station Unit #1 as Westinghouse Model DB-50 type breakers are usedin
the reactor protection system. Licensee actions on this generic
problem are governed by IEB 83-01, which was inspected in NRC
Inspection Report No. 50-334/83-04,
6.
IEB 82-02:
Degradation of Threaded Fasteners in the Reactor
Coolant Pressure Boundry of PWR Plants. This bulletin was issued
to notify all PWR reactor facilities about severe degracation of
threaded fasteners (bolts and studs) encountered in the reactor
coolant pressure boundry and to require appropriate action.
Included
were steam generator and pressurizer manway covers, valve bonnets,
pump flange connections installed on lines having a nominal diameter
of 6" or greater, and control rod drive flange and pressurizer
heater connections that do not have sealed welds to provide leak
tight intregrity. The inspector reviewed the licensee's response
to this bulletin dated August 2, :982, and could not determine
whether the control rod drive flang3 and pressurizer heater
connections were seal-welded (0mega seal
weld design) and therefore,
excluded from this bulletin action, or whether DLC is connitted to
Reg Guide 1.65, Materials and Inspection for Reactor Vessel Closure
Studs, which would negate further action on the vessel head studs.
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Thisitemisunresolved(83-08-03) pending such detennination.
The inspector verified through review of the reactor coolant
system, residual heat removal system, and safety injection system
prints and diagrams that the remainder of the licensee's planned
inspection scope, as contained in Engineering Memorandum 30165,
meets the intent of the bulletin. The inspector also determined
that those threaded fasteners and closure connections identified
in the EM are scheduled to be opened for component inspection or
maintenance during the next refueling outage scheduled for mid-
June, 1983. This bulletin remains open pending completion of
those actions.
9.
Changes, Tests and Experiments
10 CFR 50.59, Changes, Tests and Experiments, allows the licensee
to make changes in the facility and procedures as described in the
Safety Analysis Report (SAR) and to conduct tests or experiments not
described in the SAR without prior Commission approval unless it involves
a change to the Technical Specifications or an unreviewed safety
question. The DLC program that controls such changes, tests and
experiments was inspected to insure that an adequate 10 CFR 50.59
review is conducted in accordance with established administrative
controls. These controls were also reviewed to verify that:
(1)
a formal method had been established to handle requests or proposals
for conducting plant tests and experiments involving safety related
equipment or modes of operation different from those described in
the SAR; (2) that provisions have been made to assure that all tests
and experiments are conducted in accordance with approved written
procedures; (3) that responsibilities have been assi
and approving tests and experimental procedures; (4)gned for reviewing
that responsibilities
have been assigned to assure that controls identified above will be
implemented; (5) that a formal system has been established to assure
that all proposed tests and experiments will be reviewed to detennine
whether they are described in the FSAR; (6) that responsibilities have
,
been assigned to assure that a written safety evaluation pursuant to
10 CFR 50.59 will be developed for each test or experiment not described
!
in the SAR to assure that it does not involve unreviewed safety questions
or a change in the Technical Specification, and, (7) that responsibilities
have been assigned to assure that all tests and experiments are formally
reported to the NRC in a timely manner.
Documents Reviewed:
1.
Appendix A, Quality Assurance, Updated FSAR.
2.
Technical Specification S.5.1, Onsite Safety Committee.
3.
Technical Specification 6.5.2, Offsite Review Committee.
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4.
BVPS-1 Station Administrative Procedures (SAP).
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Nuclear Division Directives.
6.
Station Engineering Procedures (SEP).
7.
Nuclear Engineering Management Procedures (NEMP).
8.
Nuclear Engineering Division (NED) Internal Instructions.
9.
A sampling of active and completed Design Change Packages.
10. A sampling of approved and completed Operational, Maintenance,
and Plant Test and Performance Procedares.
Findings:
The SAPS establish the administrative controls program that defines
the method and responsibilities for conducting plant activities and
functions, in accordance with the BVPS Technical Specification and
various ANSI standards and Reg Guides. This provides the basic
procedural requirements that apply to all work activities on all
equipment, components and systems at BVPS. Additionally, Nuclear
Division Directives provide supplemental general administrative
guidance to and control over various operations and support activities.
The SEPs provide the general guidelines to assure that maintenance
and operating procedures are prepared, approved and controlled.
Through sampling review of various maintenance and operation
procedures (Maintenance Surveillance Procedures, Calibration Pro-
cedures, Preventative Maintenance and Corrective Maintenance Pro-
cedures, Operating Surveillance Tests, ar.d Temporary Operating
Procedures), and Onsite Safety Comittee Meeting Minutes for 1983,
the inspector verified that all new procedures (and procedure changes)
were reviewed to determine their effect on nuclear safety.
Facility changes are controlled by the SEPs and reviewed to determine
whether they are safety related or constitute a design change.
Controls exist to forward all safety related modifications that
are not design changes to the OSC with written safety evaluations
for their concurrence. For those modifications that constitute
design change, a 10 CFR 50.59 Safety Evaluation is perfonried on
the design concept, which is then reviewed by the OSC and forwarded
to the Station Superintendent for final approval. The Safety
Evaluation can be performed by either the Station Engineering Group
per applicable SEP, or by the Nuclear Engineering Division per
applicable NEMP. Additionally, if the design concept is significantly
revised to the extent that it impacts on the initial Safety Evaluation,
a nc-w station modification request is issued and the review process
is initiated again.
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The OSC performs a review of all tests and experiments that
affect nuclear safety by virtue of the 10 CFR 50.59 review of
all station operation (this includes plant testing and per-
fomance) and maintenance procedures.
Facility changes, tests and experiments conducted in 1982 were
reported to the NRC by DLC letter of March 29, 1983. The material
contained in this document was reviewed by the inspector on a
sampling basis to verify program implementation. The licensee
is currently conducting an adequate review of all facility changes,
procedure changes and tests or experiments, that meets the intent
of 10 CFR 50.59. No discrepancies were identified.
10.
Radioactive Waste Transportation Activities
10 CFR 71.62(c) requires the licensee to maintain, during the life of
a package to which they pertain, quality assurance records of the
monitoring, inspection and auditing of work performance during modi-
fication, maintenance and repair of packages. The inspector reviewed
NSOC General Inspection Reports on Chem Nuclear System Casks, CN-6-75
(AL-33-90) and Cask No.14-195-H-7,12,14 for 1981 thru 1982. The
only two recorded instances of maintenance or modification activities
to those packages were:
(1) a linear indication on the eye to plate
weld which was repaired per CNS instruction on October 20, 1981
(CNS-6-75), and (2) a primary lid bolt that was drilled and tapped
per MWR 827153 on June 29, 1982. The inspector was satisfied that
the licensee is maintaining required quality assurance records of
repair and modification activities made to packages as required per
10 CFR 71.62(c). This item was discussed with a Regional health
physics specialist with regards to inspection findings detailed in
NRC Inspection Report No. 50-334/82-19.
11. Safety and Quality Classifications
The inspector reviewed licensee activities for ensuring that the as-
built quality of structures, systems and components important to safety
is maintained through various plant modifications and maintenance
activities involving procurement of replacement components. This
includes the establishment of various classifications, the criteria
by which the classification as safety related or non-safety related is
determined, and how these classifications are used for obtaining
replacement components during maintenance and obtaining new items for
station modification.
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Documents Reviewed:
Nuclear Engineering Management Procedures (NEMP)
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2.2 Classification of Structures, Systems and Components
2.3 Requests for Station Modifications
2.6 Specifications
2.7 Vendor Infonnation, Processing and Control
2.8 Handling of Design Change Packages
2.16 Safety Evaluation Procedure
3.1
Purchase Requisitions
Station Engineering Procedures (SEP)
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2.1 QA Category, QC Level and Documentation Review
2.3 Design Change Coordination
4.1 Quality Classification of Materials, Equipment and Services
- Nuclear Engineering Department Internal Instructions
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- QA Procedures
0QA Appendix B
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BVPS Maintenance Manual
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Master Equipment List
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Q*5 Program
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- ANSI Standard N45.2.13, Quality Assurance Requirements for Control
of Procurement of Items and Services for Nuclear Power Plants.
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ANSI N45.2, Quality Assurance Program Requirements for Nuclear Power
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Plants, 1971.
ANSI N18.7, Administrative Controls for Nuclear Power Plants,1972.
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Findings:
The licensee currently employs a methodology whereby each plant system
is designated either safety related or non-safety related, as it is
specified in 00A Appendix B.
For maintenance purposes, system "Q
breaks" are not yet used, and all components within a system are
treated as the system is designated. The licensee does plan to
eventually make use of Q breaks after review and acceptance of the
Q*5 program. This unapproved program, initially accepted by the
Station Superintendent in 1979, breaks down each piece of equipment
into its component parts and classifies them according to safety
function.
From 1979 - 1981, Q*5 was not updated to reflect station
design changes. The Nuclear Engineering Division is currently up-
dating the program and performing a quality verification per applicable
NEDIs prior to certification of Q*5 as a quality document.
Final
acceptance should be by the end of 1983.
Currently, a majority of the design work that involves determining the
safety and quality classifications of equipment used in design changes
and modification is being perfonned by the Station Engineering Group
in accordance with the SEPs. As the Nuclear Engineering Division
continues to staff up, those responsibilities will be transfered to
them and implemented through NEMPs that parallel the requirements and
controls already specified in the SEPs.
All maintenance work is controlled through a Maintenance Work Request
(MWR) system, whereby Operations issues the MWR and routes it to QC
for review of work scope, quality category, replacement part specification
and assignment of hold points. The MWR is then forwarded to a main-
tenance engineer for a third level of review before work is initiated,
insuring a high degree of confidence that the safety category has been
accurately determined.
The inspector reviewed the MWR maintenance history of the reactor
control and protection system, the safety injection system, the chemical
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and volume control system, and the river water system for 1982 thru
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April ,1983.
From selected MWRs involving component replacement, the
inspector tracked replacement parts thru material requisitions to the
!
original purchase order and verified that procurement quality documentation
and vendor certificates of compliance were in conformance with procurement
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specifications established according to its safety function.
No
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deficiencies were noted.
Through discussions with the Superintendent of Licensing and Compliance,
the inspector also determined that potential generic equipment problems
identified through IE Circulars and Information Notices were reviewed
for applicability and fcrwarded to the Procurement Group for any necessary
action.
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12. Inoffice Review of Licensee Event Reports (LERs) and Onsite Followup
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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 action.
The inspector detemined.whether further infomation was required from
i
the licensee, whether generic implications were indicated, and whether
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the event warranted onsite' follow-up. The following;LER was reviewed:
LER 83-10/99X
Diesel Driven' Fire. Pump Inoperable
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This LER satisfys the special reporting requirement of TS 3.714.1.'
The inspector reviewed corrective maintenance activities (see
Inspection Report No. 50-334/83-07) and temporary compensatory measures
(use of a portable fire pump) used during April,1983. Licer.see
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actions were acceptable.
.
13. Unresolved Items
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Unr.esolved items are matters about which more information is. required
i
to detemine whether they are acceptable, items of noncomplience or
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deviations. Two unresolved items were identified and are discussed
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in cections 5 and 8 of this report. Followup on several previo'us
t
unresolved items are discussed in section 2.
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14. Exit Interview
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Meetings were held with senior facility management periodically ,\\
during the course of this inspection to discuss the inspection
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scope and findings. A sunnary of inspection findings'were also 1
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provided to the licensee at the conclusion of the report period.;
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