ML20039D101
| ML20039D101 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 09/24/1981 |
| From: | Brownlee V, Orlenjak R, Beverly Smith, Stetka T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20039D085 | List: |
| References | |
| TASK-1.A.2.1, TASK-2.B.4, TASK-2.F.1, TASK-TM 50-302-81-15, IEB-80-03, IEB-80-09, IEB-80-18, IEB-80-24, IEB-80-3, IEB-80-9, IEB-81-12, IEC-79-13, IEC-79-20, IEC-79-22, IEC-79-23, IEC-79-4, IEC-80-04, IEC-80-14, IEC-80-4, NUDOCS 8112310317 | |
| Download: ML20039D101 (22) | |
See also: IR 05000302/1981015
Text
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UNITED STATES
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NUCLEAR REGULATORY COMMISSION
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5
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REGION il
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101 MARIETTA ST
N.W.. SUITE 3100
[
ATLANTA, GEORGIA 30303
Report No. 50-302/81-15
Licensee:
Florida Power Corporation
3201 34th Street, South
St. Petersburg, FL 33733
Facility Name: Crystal River 3 fluclear Generating Plant
Docket No. 50-302
License No. DPR-72
Inspection at Crystal River site near Crystal River, Florida
C!2N[8-/
Inspectors: l >LNA.t
/
T. F. Shef.ka, Senior Residdnt Inspector
Date ' Signed
ZL/ff-/
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D . Srfith, R sident Inspdctor
Da'te Sitjned
avl Q L
w /w
R. V.
lenja , Reactor Insp#ctor
Date Sfsned
Approved by:
,/.4A%vf
2-V
/
V. L. Bfownlee, Acting Chief, Reactor Projects
Da~te Signed
Section 2B, Division of Resident and
Reactor Project Inspection
SU!!!t\\RY
Inspection on July 21 through August 25, 1981
Areas Inspected
Routine inspection by the resident inspectors of plant operations, security
radiological controls, procurement, Licensee Event Reports (LERs) and Non-
conforming Operations Reports (NCOR's), non-routine events, licensee action on IE
Bulletins and Circulars, reactor coolant system leak rate verification,
Inplementation of T111 Action Plan Items, and licensee action on previous
inspection items. Numerous facility tours were conducted and facility operations
observed.
Some of these tours and observations were conducted on back shifts.
The inspection involved 284.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> on site by two resident inspectors and one
regional inspector.
Results
Two violations were identified (Failure to perform plant modifications in
accordance with procedures as required by 10 CFR 50, Appendix B, Criterion V,
paragraph 5.B.(11)b; Failure to perform safety evaluations for plant modifi-
cations as required by 10 CFR. 50.59 paragraph 9.b.
8112310317 811223
PDRADOCK05000g
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DETAILS
1.
Persons Contacted
Licensee Employees
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P. Baynard, Manager, Nuclear Support Services
- G. Boldt, Technical Services Superintendent
- C. Brown, Nuclear Compliance Supervisor
J. Buchner, Security Officer
- J. Bufe, Compliance Auditor
M. Collins, Reactor Specialist
- J.
Cooper, QA/QC Compliance !!anager
B. Crane, Planning Engineer
- W. Herbert, Technical Specification Coordinator
V. Hernandez, Compliance Auditor
S. Johnson, itaintenance Staff Engineer
W. Kemper, Plant Training Manager
- K. Lancaster, Quality Assurance Auditor
T. Lutkehaus, Technical Assistant to the Nuclear Plant Manager
- P. McKee, Operations Superintendent
G. Perkins, Health Physics Supervisor
- D. Poole, Nuclear Plant Manager
R. Rogers, Nuclear Stores Supervisor
- G. Ruszala, Chemistry / Radiation Protection Manager
J. Smith, Senior Quality Programs Auditor
D. Smith, Technical Services Superintendent
D. Todd, Materials QC Inspector
- J. Lander, Maintenance Superintendent
G. Williams, QA/QC Supervisor
K. Wilson, Licensing Specialist
Other personnel contacted included office, operations, engineering,
maintenance, chem / rad, and corporate personnel.
- Present at the Exit Interview conducted on August 19.
- Present at the Exit Interview conducted on August 25.
- Present at both Interviews.
2.
Exit Interview
The inspectors met with licensee representatives (denoted in paragraph 1) on
numerous occasions during and at the conclusion of the inspection on
August 25, 1981.
During this meeting, the inspector surmiarized the scope
and findings of the inspection as they are detailed in this report.
During
this meeting, the violations, an unresolved item, and inspector followup
items were discussed.
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3.
Licensee Action on Previous Inspection Findings
(0 pen) Unresolved Item (302/81-02-06): The licensee has begun developing
new vz ve lineup proced"res that will include instrument valves and is
completing revisions to system flow diagrams. The licensee will complete
drawing revisions and implement the new valve lineup procedures prior to
returning the plant to operation from the refuel outage scheduled to begin
on September 27.
(0 pen) Inspector Followup Item (302/81-05-12): The licensee was notified by
the Clark Relay Company that a relay rebuild kit is now available that will
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replace the relay shafts with a new design. The licensee has purchased
these rebuild kits and will replace the relay shafts on all safety-related
Clark relays during the upcoming refuel outage scheduled to begin
September 27.
The replacement will be performed in accordance with flAR
81-5-20 and Work Request 24403.
(0 pen) Inspector Followup Item (302/80-42-12):
The licensee has added signs
on all equipment in the fossil plants (CR-1 & 2) that effect operation of
the nuclear plant (CR-3) to warn fossil plant personnel to notify CR-3
personnel prior to removing said equipment from service.
In addition a
revision to Administrative Instruction (AI)-1300, Crystal River
Units 1 and 2 Interface with Crystal River Unit 3, is being written to list
this equipment. This revision will be issued by September 30, 1981. This
item remains open pending revision of AI-1300.
(Closed) Violation (302/81-05-02):
The licensee revised procedure RP-101,
Radiation Protection 11anual, as Revision 14 on July 29, 1981, to include an
enforcement policy for Radiation Work Permit violators. The licensee's
actions on this item are complete.
(Closed) Unresolved Item (302/81-02-15): The licensee has revised flodifi-
cation Approval Record (itAR) 80-11-74 on June 10 to provide for partial
replacement of solid state triac relays in the ES system.
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(Closed) Inspector Followup Item (302/81-07-11): The licensee revised
procedures CP-115, In-Plant Clearance and Switching Orders, as Revision 35
on liay 28, 1981 to include a listing of safety-related equipment.
(Closed) Inspector Followup Item (102/81-11-09):
The licensee has revised
procedures QC-200, Training and Qualification of Nuclear Quality
Assurance / Quality Control Inspection Personnel, and QC-201, Training of
Nuclear Compliance Personnel to specify that training will be completed
within two years.
(Closed) Unresolved Item (302/81-05-06): Work instruction cards utilized by
the Chem / Rad Technicians as a guide for performing their duties were revised
to alert the technician that if a Standing Radiation Work Permit (SRWP) is
going to expire prior to the next update, the technician should notify a
Chem / Rad Supervisor so that a determination of renewing or cancelling the
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SRWP can be made.
The inspectors' review of the outstanding SRWP's
indicates that the revised system is working.
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(Closed) Unresolved Item (302/81-11-06): The licensee performed a second
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level supervisory review of the completed data from procedure SP-417 on
June 30, 1981 and determined that the results were satisfactory. Trainir
sessions were conducted during the period of August 3-7, 1981 and on
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August 13, 1981, for all plant staff personnel except the operations staff
to re-enforce the supervisory revits requirements of AI-400. The operations
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staff personnel will receive their training during requalification training.
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The licensee's action on this item is considered to be canplete.
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(Closed) Inspector Followup Item (302/81-07-03): The licensee has revised
preventative maintenance procedure PM-119, Inspection and Cleaning of 480
Volt Motor Control Centers, to include a requirement to check the cabinet
filters and to clean or replace them as necessary.
(Closed) Inspector Followup Item (302/81-01-02):
Reviews of the Clearance
Order Book indicates thu'. the required clearance audits are being performed.
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(Closed) Inspector Followup Itan (302/81-02-19): The licensee's requalif-
ication training syllabus now includes the interrelationship between the
waste disposal system and the reactar coolant system.
(Closed) Violation (302/81-07-12): The inspector reviewed the licensee's
response to this violation and verified training completion and procedural
revisions to SP-187 and HP-106.
The licensee's actions on this item are
considered complete.
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4.
Unresolved Items
Unresolved items are matters which more infonnation is required to determine
whether they are acceptable or may result in violations. A new unresolved
item identified during this inspection is discussed in paragraph 5.B(12).
5.
Review of Plant Operations
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The plant continued with Mode 1 power operations until July 31 at which time
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a reactor trip occurred due to personnel error. The plant returned to
Mode 1 power operations on July 31 and continued in this mode for the
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duration of the inspection period.
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a.
Shift Logs and Facility Records
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The inspectors rev,ewed the records listed below and discussed various
entries with operations personnel to verify compliance with T.S. and
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the licensee's administrative procedures.
-Shift Supervisor's Log;
-Operator's Logs:
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- Equi pmen t-ou t-o f-Service-Log ;
-Shif t Relief Checklist;
-Control Center Status Board;
-Short Term Instructions;
-Auxiliary Building Operators' Log; and
-0perating Daily Surveillance Log.
In addition to these record reviews, the inspector independently
verified selected clearance order tagouts.
b.
Facility Tours and Observations
Througout the inspection period, facility tours were conducted to
observe operations and maintenance activities in progress. Some
operations and maintenance activity observations were conducted during
back shifts. Also during this inspection period, numerous licensee
neetings were attended by the inspectors to observe planning and
management activities.
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The facility tours and observations encompassed the following areas:
-Security Perimeter Fence;
-Turbine Building;
-Meteorological Tower;
-Control Room;
-Emergency Diesel Generator Rooms;
-Auxiliary Building;
-Intermediate Building;
-Battery Rooms; and,
-Electrical Switchgear Rooms.
During these tours, the following observations were made:
(1) Monitoring instrumentaticn - The following instrumentation was
observed to verify that indicated parameters were in accordance
with the Technical Specifications for the current operational
mode:
-Equipment operating status;
-Area, atmospheric and liquid radition monitors;
-Electrical system lineut;
-Reactor operating parameters; and
-Auxiliary equipment operating parameters.
(2)
Safety Systems Walkdowns
The inspectors conducted walkdowns of the following safety systems
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to verify lineups were in accordance with license requirements for
system operability:
-Emergency Diesel Generator (EDG) Fuel Oil Transfer System;
-EDG Compressed Starting Air and Engine Exhaust System;
-fiuclear Services and Decay Heat Seawater System; and
-4160V and 480 Engineered Safeguard Busses
(3) Shift Staffing
The inspectors verified by numerous checks that operating shift
staffing was in accordance with Technical Specification require-
ments.
In addition, the inspectors observed shift turnovers on
different occasions to verify the continuity of plant status,
operational problems,and other pertinent plant information was
being accomplished.
(4) Plant Housekeeping Conditions
Storage of material and components and cleanliness conditions of
various areas throughout the facility were observed to determine
whether safety and/or fire hazards exist. The general house-
keeping conditions are acceptable.
(5) Radiation Areas
Radiation Control Areas (RCA's) were observed to verify proper
identification and implementation. These observations included
review of step-off pad conditions, disposal of contaminated
clothing, and area posting.
Area postings were verified for
accuracy through the use of the inspector's own radiation
monitoring instrument.
fio problems were identified in this area.
(6)
Fluid Leaks
Various plant systems were observed to detect the presence of
leaks.
fio problems were identified in this area.
(7) Piping Vibration
flo excessive piping vibrations were noted.
(8) Pipe Hangers / Seismic Restraints
Several pipe hangers and seismic restraints (snubbers) on
safety-related systems were observed.
fio problems were identified
in this area with the exception of the issue discussed in Section
6.B(2) of this report.
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(9) Security Controls
Security controls were observed to verify that security barriers
are intact, guard forces are on duty and access to the protected
area is controlled in accordance with the facility security plan,
fio problems were identified in this area.
(10) Operating Procedures - Operating Procedures were observed to
verify that:
-approved procedures were being used;
-qualified personnel were perfoming the operations; and
-Technical Specification requirements were being followed.
The following Operating Procedures were observed:
-0P-407, Liquid Waste Disposal System, and associated liquid
release permit for the release of "A" Evaporator Condensate
Storage Tank water.
-0P-210, Reactor Startup, and OP-203, Plant Startup, fcr plant
restart following the reactor trip on July 31, 1981.
(11) Tour of Auxiliary and Intermediate Buildings
a.
During a tour of the Auxiliary Building the inspector sted
that the handwheel of pneumatically operated itakeup Valve
(HUV)-253, outside reactor building isolation from reactor
coolant pump controlled bleed-off, was removed from the valve
stem and secured to the piping below the valve.
It appeared
that a pipe hanger located near the valve stem would
interfere with the installation of the valve handwheel. The
inspector discussed this issue with the licensee and was told
that the necessary actions would be taken to reinstall the
!!UV-253 handwheel .
Inspector Followup Item (302/81-15-01): Review the
licensee's actions to reinstall MUV-253 handwheel.
b.
During a tour of the Intermediate Building on August 13, the
inspector noted that the position of the bearing oil level
sight glasses on the turbine driven emergency feedwater pump
(EFP)-2 appeared to be different than that described in
Modification Approval Records (MAR) 78-1-1 and 78-1-1A.
The
inspector obtained copies of the MARS and verified that the
installed sight glass positions on EFP-2 were not as
indicated in the f1ARs. This finding was presented to licensee
representatives and the licensee conducted an investigation.
The original modification was perfomed in accordance with
itAR 78-1-1 and the work request (W/R) for performanc of the
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f1AR was signed off as complete on September 28, 1978.
Apparently during some subsequent routine maintenance, the
sight glasses were returned to their original configuration.
(The licensee does not have documentation to reflect this
change.) During the period of February 12-16, 1979, NRC
Inspection 50-302/79-04 was conducted and the repositioned
sight glasses were identified by the inspectors following a
review of MAR 78-1-1.
At this time the licensee was issued
an item of noncompliance. The licensee responsed to this
item and stated that the sight glasses were returned to the
configuration required in MAR 78-1-1.
This re-modification
was accomplished by issuing a new 14AR 78-1-1A and the W/R for
this modification indicates this work was completed on
!! arch 22, 1979.
MAR 78-1-1A was then sent to the licensee's engineering
department as required by the licensee's procedures.
The
engineering department approved the itAR on January 25, 1980,
but did not close out the package.
The engineering depart-
ment also added a requirement in the MAR to scribe low oil
level lines on the turbine casing to allow sighting from each
sight glass and to remove all high/ low level marks from the
sight glasses proper.
This additional modification was not
perfomed.
Sometime between the completion of MAR 78-1-1A and August 13,
apparently tne EFP-2 sight glass configuration was again
changed as identified by the inspector. The licensee is
unable to produce documentation to reflect this change.
Failure to accomplish plant modifications in accordance with
procedures, instructions, and drawings is contrary to the
requirements of 10 CFR 50 Appendix B, Criterion V governing
activities affecting quality and the requirements of
procedure CP-114, Procedure for Preparation of Permanent and
Temporary Modifications, with regard to verifying that a MAR
is properly completed, is considered to be a violation.
Violation (302/81-15-02):
Failure to perfom plant modi-
fications as required by 10 CFR 50 Appendix B Criterion V,
FSAR section 1.7.6.7.1, and CP-114.
The additional violations against 10 CFR 50 Appendix B
Criterion V identified in paragraphs 8(a) and 9.a following
are considered to be further examples of the licensee's
failure to adhere to procedures for the control of plant
modifications.
Similar violations of 10 CFR 50, Appendix B, Criterion V with
respect to plant modifications were identified during the
inspection period of February 12-16,1979 (NRC Report
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50-302/79-04) and therefore this violation is considered to
be recurrent and uncorrected.
(12) Surveillane Testing - Surveillance testing was observed to
verify that:
-approved procedures were being used;
-qualified personnel were conducting the tests;
-testing was adequate to verify equipment operability;
-calibrated equipment, as required, were utilized; and
-Technical Specification requirements were being
followed.
The following tests were observed:
-SP-110, Reactor Protective System Functional Testing
(Channel "A" and "B");
-SP-140, IN-Core Neutron Detector System Calibration (Data
review only);
-SP-317, Reactor Coolant System Water Inventory B:ilance;
-SP-334, Spent Fuel Pool Pump Operability;
-SP-335, Radiation Monitoring Instrumentation Functional Test
(Section 6.3); and,
-SP-340, ECCS Pump Operability (Return to operability of "A"
Decay Heat ump and "A" Building Spray Pump following
maintenance .
During observation of SP-110, Reactor Protective System (RPS)
Functional Testing, the inspector noted that the method used
for functionally testing RPS analog channel bistables did not
include verification of bistable trip point by vcitage
measurement at the bistable input. The present method uses
an indicator for verification of bistable trip setpoint
(example: RCS low pressure bistable trip point is verified by
the RCS pressure indicator in the RPS cabinet).
In addition,
the acceptance tolerances for the bistable trip setpoints in
the procedure allow values to be accepted that could
potentially be beyond that allowed in Technical Specifi-
cations for (TS) "!!aximum allowable trip setpoints."
The inspector also noted that this same concern (bistable
setpont verification) exists similarly in Procedur SP-130
Engineering Safeguards fionthly Functional Test. This issue
was discussed with the licensee and the inspector's concerns
were acknowledged.
The inspector stated that adherence to
Technical Specification requirements for functionally testing
analog channel bistable trip functions thould include
verification of bistable setpoints by voltage measurement
with suitable acceptance tolerances to ensure bistable trip
setpoints do not exceed Technical Specification maximum
allowable trip setpoints. The licensee will revise
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functional testing procedures, where required, to ensure
analog channel bistable setpoints are verified.by voltage
measurements at the histable input and that suitable
acceptance criteria are provided.
Unresolved Item (302/81-15-03):
Revise functional test
procedures to ensure analog channel bistable trip setpoints
do not exceed TS required limits.
(13) Maintenance Activities - The inspector observed maintenance
activities to verify that:
-Correct equipment clearances were in effect;
-Work Requests (W/R's), Radiation Work Permits (RWP's),
and Fire Prevention Work Pemit, as required, were
issued and being followed;
-Quality Control personnel were available to inspection
' activities as required; and,.
-Technical Specification requirements were being
followed.
The following maintenance activities were observed:
Preventive Maintenance Procedure (Pli)-133, Equipment Lubri-
cation (Decay Heat Pump (DHV)-1A and Building Spray Pump
(BS)-1A oil replacement).
Maintenance Procedure (MP)-131, Disassembly and. reassembly of
DHP-1A (Replacment of DHP-1A pump due to bearing failure).
Troubleshooting activities on Auxiliary Building Air handling
fans to correct damper problem.
Maintenance Procedure (MP)-149, Check Valve Cap Removal and
Reinstallation (Repair of Nuclear Services Closed Cycle : Check
Valve (SWV-101).
Troubleshooting meteorological tower 175' and 33' temperature
indication.
No problems were identified in this area.
(14) Prerefueling Acitivities
The inspectors witnessed the unloading and transfer to the
new fuel storage area of several burnable poison rods (BPR's)
in accordance with refueling procedure (FP)-303, New Control
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Component Receipt, Inspection, Storage and Container
Reclosing.
In addition, the inspectors witnessed Quality
Control inspection of several BPR's.
No problems were identified in this area.
6.
Review of Licensee Event Reports and Nonconfonning Operations Reports
(NCORs)
a.
The inspector reviewed Licensee Event Report (LER's) to verify that:
-The reports accurately described the events;
-The safety significance is as reported;
-The report satisfies requirements with respect to information provided
and timing of submittal;
-Corrective action is appropriate; and
-Action has been taken.
LER's 80-43, 80-47, 81-33, 81-37, 81-38, 81-39, 81-41, 81-42, 81-43,
81-44, 81-45, 81-47, 81-48, 81-49, and 81-51 were reviewed. As a
result of this review the following items were identified:
1.
LER 81-36 stated as a part of the corrective actions that an
evaluation would be performed to determine if failed fuel was
present. A test was run during the period of July 29, 1981
through July 31, 1981 of approximately 43.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> duration with
the reactor at steady state full power and the purification system
isolated.
During this time period, samples were taken at
approximately four hour intervals and were analyzed for dose
equivalent iodine 131. An excessive increase in Iodine 131 would
be indicative of fuel degradation.
The results of this testing were reviewed by the licensee's
corporate engineering group and the reactor vendor (Babcock and
Wilcox) and it was determined that no abnormal criteria were
observed. The inspector discussed this determination with
licensee representatives and reviewed some data results. The
inspectors have no further questions on this item at this time.
2.
LER 81-38 reported an increase in the Strontium 90 (Sr-90) levels
in green leafy vegetable samples analyzed at station C48. This
increase exceeded ten times the control station value. This
increase in Sr-90 could not be explained and the licensee has
conducted a resample to confirm the initial findings.
The results
of this resample were examined on August 25, 1981 and since both
the control station and critical station exhibited a Sr-90
increase of similar magnitude, the licensee attributes the higher
Sr-90 level to nuclear weapons testing conducted by the Peoples'
Republic of China in December,1980.
As of August 25, the
licensee has determined that the increase in Sr-90 at the critical
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station was less than 10 times the control station (actual
magnitude about 3.5 times) and therefore the LER was not required.
The licensee is considering withdrawing this LER. The inspectors
have no further questions on this item at this time.
3.
LER 91-51 was issued as a followup to a prompt report of August 14
which reported an inadequacy of the design of the Cable Spreading
room floor. The licensess's Architech-Engineer (AE), Gilbert
Associates Incorpoated (GAI), determined, during an engineering
nyestigation for a fire protection modification, that concrete
embedments in the wall supporting the cable spreading room floor
were not adequate to sustain dead load, live load,and seismic
requirements.
These embedments were detemined to be adequate as
long as the live loading was restricted.
The licensee took immediate action to restrict live loads by
securing the room with lock and key.
Key control is via the Shift
Supervisor who has explicit instructions on the amount of
permissible live loading. The licensee is developing a modif-
ication (tiAR 81-8-15) that will be installed to utilize the walls
beneath the ficor for added support.
Until this modification is
completed, live loading in the cable spreading room will be
restricted.
The inspectors examined and are satisfied with the licensee's
actions.
Inspector Followup Item (302/81-15-04):
Review the licensee's
progress in modifying the cable spreading roon floor to increase
floor loading capacity.
b.
The inspector reviewed NCOR's to verify the following:
-Compliance with the Technical Specifications;
-Corrective actions as identified in the reports or during subsequent
reviews have been accomplished or are being pursued for completion;
-Generic items are identified and reported as required by 10 CFR Part 21; and,
-Items tre reported as required by the Technical Specifications.
The following NCOR's were reviewed:
81-43
81-259
81-272
81-300
81-117
81-262
81-276
81-301
81-148
81-263
81-278
81-302
81-150
81-264
81-279
81-303
81-153
81-265
81-282
81-304
81-250
81-266
81-284
81-305
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81-251
81-267
81-287
81-306
81-252
81-268
81-280
81-252
81-269
81-291
81-288
81-289
81-254
81-270
81-292
81-290
81-257
81-299
81-288
81-295
81-297
Based on this review, the following items were identified:
(1.) NCOR 81-0265 reported the failure of decay heat system valve
DHV-41 to open in 120 seconds during performance of Surveillance
Procedure SP-370, Quarterly Cycling of Valve.
Review of this
event indicates that SP-370 was developed to satisfy the quarterly
valve cycling requirement of ASME Section 11 Inservice Inspection
(ISI) and the isolation requirement of Technical Specification (TS) 3.6.3.1.
The Licensee has detennined that DHV-4: does not
receive an engineered safeguards (ES) closure signal, the valve
is normally closed in operational Modes 1, 2 and 3, and is only
used during normal decay heat removal operations. Therefore the
isolation time specified in TS 3.6.3.1 is not applicable and the
TS appears to be incorrect. The licensee has also determined that
the following valves are also apparently incorrectly specified in
-BSV-3 and BSV-4, only get an open signal at 4 psi Reactor
Building Pressure;
-DHV-42 and 43, does not receive an ES closure signal and are
maintained closed during Itodes 1, 2, and 3;
-DHV-4, does not receive an ES closure signal and is maintained
closed during Modes 1, 2, and 3, and only has an interlock to open
at 284 psi for decay heat operation; and,
-DHV-5 and 6, receives an ES signal to open for low pressure
injection when RCS pressure is less than 500 psig.
The licensea is presently stroking these valves either open or
closed to verify TS stroking times and to meet AMSE ISI require-
ments.
The licensee also determined that auxiliary spray isolation valve
DHV-91 was not being tested for closure time by SP-370 or SP-435,
Valve Testing During Cold Shutdown. The licensee verified that
this valve was tested for closure time during post-maintenance
testing. The inspector reviewed the data from this valve testing
,
and verified the stroke time was in accordance with TS require-
men ts .
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13
The licensee will submit a TS change request to revise section
3.6.3.1 to be consistent with actual plant operation.
The
licensee will also revise procedures SP-370 and SP-435 to provide
for monitoring the closure time for DHV-91 and to insure closure
times on the previously listed valves are monitored until a TS
change is approved.
Inspector Followup Item (302/81-15-05):
Review licensee action to
resolve inaccuracies in TS 3.6.3.1 and procedures SP-370 and
(2.) NCOR 81-279 reported a potential generic problem with aluminum
adapter bushings installed in hydraulic snubbers manufactured by
Power Piping Co..
Since July 1979, eleven aluminum adapter
bushings have been found cracked during surveillance testing of
safety-related snubbers. Subsequent testing of the affected
snubbers showed that six of the snubbers passed and five failed
the operability criteria. All of the cracked aluminum adapter
,
bushings were replaced with stainless steel bushings.
In
addition, the licensee has been replacing the aluminum adapter
bushings with stainless steel bushings on any snubbus removed for
other reasons (testing, bent piston rods, etc). At this time
approximately 8% of the safety-related snubbers have stainless
steel bushings. The inspectors discussed the extent of the
cracking problem and what actions were being taken to correct the
problem with licensee representatives. The licensee stated they
were unable at this time to classify the cracked aluminum bushings
as a generic problem based on the failure rate being less than 1%
for the total number of visual examinaticns performed on the
In addition, an ongoing failure analysis study
of the cracked bushings indicated the failures do not appear to be
thermal or fatigue induced or a casting problem. The licensee is
presently conducting a special surveillance inspection of all
accessible snubbers outside of the reactor building that will
include vibration measurements as a possible aid in the failure
analysis study.
If this inspection does not indicate exceptional
cracking problems then the licensee will continue with their
original plans for replacing approximately 25% of the snubbers in
the reactor building during the refueling outage, scheduled to
begin September 27, 1981, with snubbers having stainless steel
bu shings.
If exceptional cracking is evident the replacement
amounts will be re-evaluated. The inspectors are closely
monitoring this issue and concur with the licensee's action at
this time.
Inspector Followup Item (302/81-15-06):
Review licensee's action
regarding cracked aluminum adapter bushing on hydraulic snubbers.
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14
(3.) NCOR 81-299 reported that core quadrant power tilt exceeded the
Technical Specification (TS) steady-state value for the Incore
Instrumentation following a power reduction.
The out of core
instrumentation did not indicate evidence of an incore tilt.
The
inspector's investigation of this event revealed that the licensee
had just completed the incore instrumentation check portion of
surveillance procedure SP-140, In-core Neutron Detector System
Calibration, but that the procedure did not direct the instrument
technicians to substitute a zero value into the plant computer for
any failed detectors. As a result the computer utilized the
failed detector values in the incore tilt calculation, thus giving
an erroneous incore tilt. When zeros were substituted for the
failed detectors, the incore tilt returned to a value consistent
with the out of core detectors. The inspector verified that the
number of failed detectors did not exceed TS required number for
operability. The licensee will revise SP-140 to direct instrument
technicians to substitute a zero in the plant computer for each
failed detector.
Inspector Followup Item (302/81-15-07): Verify revision to SP-140
to require substitution of zeros into the plant computer for
failed incore detectors.
(4.) NCOR 81-300 reported that a procedure was not available to reduce
the Reactor Protection System (RPS) nuclear overpower trip
setpoints that are based on reactor coolant system flow and axial
power imbalance. The licensee is revising calibration procedure
SP-113 Power Range Nuclear Instrumentation calibration, to provide
for these setpoint reductions.
Inspector Followup Item (302/81-15-08):
Review the revision to
SP-113 that will include a setpoint reduction procedure for
nuclear overpower trip.
7.
Review of IE Bulletins and Circulars
The following IE Bulletins (IEB) and Circulars (IEC) were reviewed to verify
adequacy of the licensee's actions:
a.
Loss of Charcoal from Standard Type II, 2 Inch Tray
Absorber Cells
b.
Hydramotor Actuator Deficiencies
c.
Decay Heat Removal System Operability
d.
Maintenance of Adequate Minimum Flow Thru Centrifugal
Charging Pumps Following Secondary Side High Energy Line
Rupture
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15
e.
Prevention of Damage Due to Water Leakage Inside
Containment (October 17, 1980 Indian Point 2 Event)
f.
IEC 79-13
Replacement of Diesel Fire Pump Starting Contactors
g.
IEC 79-20
Failure of GTE Sylvania Relay, Type PM Bulletin 7305,
Catalog 3U12-11-AC with a 120V AC Coil
The GTE Relays identified in this Circular are similar
to the Clark P!1 Relays utilized at this facility (Clark
relays are now manufactured by Sylvania).
The licensee
has identified a " burring" problem on the shaft of
these relays that was reported to the NRC in Licensee
Event Report (LER) 81-14, that interferes with relay
operation. The licensee's actions in this regard are
being tracked in accordance with Inspector Followup
Item (302/81-05-12) (See paragraph 2 of this report).
h.
IEC 79-22
Stroke Times for Power Operated Relief Valves
i.
IEC 79-23
Motor Starters and Contactors Failed to Operate
j.
IEC 80-04
Securing of Threaded Lockfsg Devices on Safety-Related
Equipment
The inspectors review of the licensee's action ca this
circular indicates that the licensee intends to add
caution statements into various Preventive thintenance
Procedures (PM's) and flaintenance Procedures (MP's).
These additinns will be made, as applicable, when the
procedures come up for periodic review.
Since this
periodic review cycle occurs over a two year period, it
is conceivable that these procedure changes would not be
completed until 1983.
The inspector discussed this issue with licensee
representatives and stated that the two year cycle
appeared inadequate to assure that such events do not
l
occur at this facility especially since the facility is
about to enter an extensive refuel outage on
September 27.
Licensee representatives acknowledged the
inspectors' remarks and are reviewing the status of this
review.
This circular remains open.
k.
IEC 80-14
Radioactive Contamination of Plant Demineralized Water
Systen and Resultant Internal Contamination of Personnel
With the exception of IEC 80-04, which remains, open, no additional
inadequacies were identified and the licensee's actions on these Bulletins
and Circulars are considered to be complete.
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8.
Implementation of Three Mile Island (TMI) Action Plan Items
,
The NRC is requiring licensees to implement certain items tqa L were-
determined to improve plant safety as a result of the lessen learned from
the TMI accident that occurred on thrch 28, 1479.
These Task AcEion Plib '
(TAP) items were delineated in NUREG 0585, 0660, and 0137.1 in,variods
~"
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,
correspondence between Nuclear Reactor Regulation (HRR) and tne licensees,
'
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and in various Inspection and Enforcement (IE) 3uiletins and Circulars.
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The following TAP's were reviewed:
<-
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a.
II.F.1. (1A) Accident Monitoring
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NRC Inspection Report 50-302/80-25, paragraphs 12c, d, and 3 discusses
'
a status of review of NUREG 0578 pa, agrap),2.1.8. items.
At the time
'
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of the report (June,1980), the inspectors noted in paragraph 12c that
,
the licensee was in process of. installins 3 main steam line (MSL)
~
radiation monitoring system and that the Mystem would be canpleted by
4
the licensee's cannitment date of January 1,1981.
In addition, in a
Florida Power Corporation (FPC) letter to Mr. Rober W. Reid of Nuclear
.
)
Reactor Regulation (NRR) dated April 20, 1980, the licensae stated that
1
MSL monitors would be installed to monitor each of the two steam
genera tors .
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The inspector reviewed Modification Approval Records (HAL)~30-5-78,A,
s
B, and C that were developed to install (MAR 80-5-78 A.and B) and cest
~
(MAR 80-5-78c) the MSL monitors.
In addition, the inspeciars have _,
,,
examined the installed monitors to verify installation und operability.
'
_
Two MSL Monitors have bain installed and they monitor t00' of the four
MSLs. The installation is arranged juch that monitorirs or each steam
s
generator is provided.
The licensee will install two atsef MSL
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monitors on the remaining MSLs during the upcaning refuel catage
'
scheduled to begin September 27.
,
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During the inspectors review of MAR' 80-5-78c that provided a functional
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test of the MSL monitors, it was noted that a temporary change was made
to the test procedure and that the data sheet was incorrectly
- -
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comple ted.
The inspector questioned licensee representa6ives about
,
these discrepancies and also questioned, based upon review of this test
procedure, whether the MSL monitors were operational.
s
Licensee representatives investigated these descrepancies and
determined that an apparent unauthorized change was made t'o :he ucta
j
sheet and concurred with the inspector's findtog that the test'was
performed incorrectly. The licensee provideJ additional documentation
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to the inspector to demonstrate that-the instrumentation had beek
properly tested and therefore was considered to be operational. Th'e
inspector reviewed this additional. data and concurred with the
licensee's conclusions.
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Ba cd on"theke reviews, it appears the licensee has met their
5.nnitment with respect to the installation and operability of the itSL
menitors. The' licensee has not, however, complied with the require-
"
ments' of their own procedure CP-114, Procedure for Preparation of
Pemonedt and Temporary liodifications, with regard to making !!AR
_
- chsges in the field and to verify the fiAR is properly accomplished.
The .br ukdown in the MAR system is considered to be a violation. This
violation >is another example of the tiAR system violation that is
discussed in paragraph 5.b(11)6 of this report.
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b.
I. A.2.1(4) Irme(iath Upgrading of Reactor Operator and Senior Reactor
Operator Training and Qualifications
,,_
Reactot Operato'- (RO) and Senior Reactor Operator (SRO) training
^
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- rograir.s and qualification requirements have been upgraded to meet the
aew req;irements of TAP 1. A.2.1, Enclosures 1 and 2 as presently
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implemen ted. To verify this upgrading the inspector discussed program
and qu6lification upgrading with licensee representatives and reviewed
~ the following documentation:
,
,
-Resumes and qualification records of three licensed reactor operators
. that are in training for upgrading to senior reactor operator
,
,
Sicenses;
,
-Attendade sheets for the requalification program to verify attendance
by-training instructors; and,
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-Syllabuses for the SR0 training program and licensed operator
requalification program to verify that the program included training
in heat transfer, fluid flow, and thermodynamics.
,
Based upon thir review, the licensee's program appears to be consistent
with their short range commitments to the flRC. The inspectors have
,
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no further questions on this item at this time.
c.
II.B.(2A Training for Mitigating Core Damage
"
g
The, licensee 4eveloped a training program for mitigating core damage on
j.
January 1,1931 and comnenced the training program in April,1981.
Tnese dates are consistent with the commitments made to fiuclear Reactor
.
.
Regulation (NRR) on December 15, 1980. As of August 12, 1981, the
'
licensee has completed training 23 of the 37 licensed operators and all
_
of the hot licensed training class. The licensee has not yet commenced
training of managers and technicians in the Instrumentation and Control
_
and Chem / Rad departments. All this training is expected to be
completed by October 1,1981.
~
The inspector reviewed the syllabus in use by the licensee to
(
_ accomplish this training.
The syllabus was developed for the licensee
"
by the reactor Nen%r (Babcock and Wilcox) and includes all the
requirements ouilned in TAP 1.A.2.1, Enclosure 3.
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Based upon this review, the licensee's progress is consistent with the
n-
commitments to the NRC. Completion of this training- remains to be
. verified and is considered to be an inspector followup item.
. Inspector Followup Item (302/81-15-09): Verify completion of training
for citigating core damage to licensed' operators and nonlicensed
- nanagers and technicians.
9.
Plant Modification Review
The insheter reviewed two potential enforcement findings that were
l.
identified during the Management Appraisal inspection conducted in February,
'
1981 thatjconcerned modification 'and design changes.
_
a.
Modtf'ication Approval Record (MAR) 80-02-72B required pipe support
EFH-64 to be modified in accordance with GAI drawing 32064. This
x
' drawing' required the installation of a 12"' square baseplate
perpendicular to the pipe run and a 12" x 19" baseplate in line with-
the pipe.
Drawing 32064 required bolt to bolt spacing of 8" for
baseplate "D".
It also required installation of grout under the
!~
basepla tes.
Installation. records for pipe support EFH-64 showed that the baseplates
3"
were reversed. The records also showed that minimum bolt spacing for
~
baseplate "D" was 7 3/8 inches. The baseplates were grouted without a
"
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written procedure for grouting.- There was no record of an engineering
evaluation and' acceptance of these conditions.
'
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Failure to accomplish plant modifications in accordance to procedures,
instructions, and drawings is' contrary to the requirements of 10 CFR
'
.50, Appendix B, Criterion V and the requirements-of procedure CP-114,
Procedure for Preparation of Permanent and Temporary Modifications, and
is considered to be a violation.
This violation is another example of the MAR system violation that is
discussed =in paragraph 5.b.11.b of this report.
b.
' MARS 79-5-62 and 79-5-62A contained a modification safety evaluation
"
that was prepared by the Nuclear Operations Engineering Department
"
(N0E) Jnd approved by the Nuclear Technical Specification Coordinator.
%
The modification involved removal and plugging of one or seven
Emergency Feedwater (EF) system pipe runs between each EF system and
O
each Steam Generator. The modification safety evaluation indicated
that Babcock and Wilcox (B&W) had advised N0E that " sufficient EF can
i
be" injected through 6 nozzles (rather than old 7)."
No written
confirmation from B&W was available. No documented N0E verification of
+
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this was available.
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19
In addition, MAR 79-5-62 shows that the EF system piping to both steam
generators was modified on' July 15, 1980, but a safety evaluation of
the effect of this system configuration change on the seismic analysis
of the system had not been done.
10 CFR 50.59 requires that the holder of a license authorizing
operation of a production or utilization facility may make changes in
the facility as described in the safety analysis report without prior
comission approval unless the proposed change involves an unreviewed
safety question. FSAR Figure 4-5 shows seven auxiliary feedwater lines
to each steam generator.
FSAR paragraph 10.2.1.2 describes the
emergency feedwater head requirements.
FSAR paragraph 10.2.1.2 also
indicates that emergency feedwater piping is designated to seismic
category 1.
Failure to perform safety evaluations of facility changes is contrary
to the requirements of 10 CFR ~50.59 and is considered to be a
violation.
Violation (302/81-15-10): Failure to comply with the requirements of
10 CFR 50.59 safety evaluations for plant modifications.
A similar violation of 10 CFR 50.59 with respect to, safety evaluation
of plant modifications was identified during the inspection period of
February 12-16,1979 (NRC Report 50-302/79-04) and therefore, this
violation is considered to be recurrent and' uncorrected.
10.
Reactor Coolant System Leakrate Verification
The NRC has recently developed a computer program which is used with the
Hewlett Packard (HP) 41C system for verification of Reactor' Coolant System
(RCS) leakrates calculated by licensees. The program has been developed to
perfonn leakrate calculations .for the three general types of pressurized
water reactors.
Before the program can be used at a specific facility the
inspector must prepare a set of magnetic data cards containing the " plant
specific" data for the facility. Using these cards, along with the general-
program, the inspector now has a " custom" program 'for the facility, in this
case Crystal River Unit 3.
The inspector can now verify RCS leakrates
accurately in a matter of minutes.
On August 12, 1981, the licensee performed a RCS leakrate calculation in
accordance with Surveillance Procedure, (SP)-317, RCS Water inventory
balance. The procedure was comenced at 0900 and was completed at 1300 (4
hrs.). The inspector observed the plant instrumentation and independently
verified the final results using the NRC calculator, and compared these
results with the licensee's final data computed in accordance with SP-317.
The licensee's final results were withir plus or minus .2 gpm of the NRC
calcu'lator results.
.
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20
Test Results
NRCCalculator
Gross (gpa)
.706
.70
Identified (gpm)
.527
.53
Unidentified (gpm)
.178
.18
Based on .these results, it appears that the licensee's leakrate calculation
procedure, SP-317 is providing adequate results. The inspectors have no
further questions on this item at this time.
11.
Procurement
The inspector reviewed the licensee's procedures for procurement, storage,
and handling of Quality Materials.
Areas examined included:
(1) Qualified
personnel receiving safety-related items, (2) storage and packaging
requirements, (3) preventive maintenance, and (4) identification (trace-
ability).
The inspector witnessed a stores receipt examination of Quality Materials
performed by the licensee. The licensee conducted the examination in
accordance with their Quality Operating ilanual Proceudres. The licensee's
" hold compound" was also toured by the inspector and found to be satis-
factory.
Parts and materials were appropriately identified with release,
hold and discrepancy tags.
Records for two safety-related items (Quality 11aterial) were chosed at
random by the inspector to verify traceability.
Records inspected included:
(1) Purchase Order, (2) Receipt Record,
(3) Issue Record, and (4)
Certification Record. The locations of the items in storage were also
verified.
During this review the inspector noted a problem with a Production Quality
itaterial Issue document (QCI 21867) for material purchased under Purchase
Order No F10515Q.
On flay 8,1981, the QCI was signed for issuance, however it was not signed
for receipt (i.e., the material picked up by a mechanic for use in the
field) until June 1,1981. Therefore, the time difference between the issue
signature and receipt signature was almost three weeks.
During this three
week period, on May 12, a Quality Programs Department (QPD) audit was
conducted.
The audit selected P0 tio. F10515Q for review and detemined that
there was a discrepancy in the certification records supplied by the
material's vendor. The item was immediately removed from stores and the
certification proMem was corrected. The item was then received by the
Field on June 1.
The issued QCI 21867 did not reflect the removal and
replacement of the material from stores for the certification problem.
. _ _ _ - _ _ _ _ _ _ _ .
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21
The record keeping involved with this event gave the impression that
uncertified material was released to the field.
Though this did not happen,
the possibility of this occurring is enhanced by allowing an exceptional
time delay between signing a QCI for issuance and signing the QCI for
receipt. The inspector reviewed approximately fifty additional QCI's and
noted that the maximum time period between issuance and receiving of
material was approximately one day.
The licensee will revise Quality Control llanual procedure QOP 7.0, Issue and
Return of Quality Material, to specify that QCI's will not be signed for
issuance until they are ready for receipt.
Inspector Followup Item (302/81-15-11):
Revu the licensee's actions to
revise Q0P 7.0 to specify a time limit between issuance and receipt of a
QCI.
12.
Nonroutine Event
At 0730 on July 31, the reactor tripped from full power due to high reactor
coolant system (RCS) pressure.
The high RCS pressure was caused by an
increase in RCS temperature which resulted when the "B" liain Feedwater (MFW)
Pump tripped. Upon loss of B !!FM pump, the Intergrated Control System (ICS)
began a plant runback, however the runback was not fast enough to prevent
the reactor trip.
Preliminary investigation of this trip indicated that the B fiFW pump tripped
when it lost its pump control speed signal.
The signal was lost when an
instrument technician inadvertently opened a circuit breaker supplying speed
signal power. The technician had intended to open an adjacent circuit
breaker, however due to the close proximity of the two breakers, the speed
signal breaker was opened instead.
A normal plant shutdown to Mode 3 (Hot Standby) occurred. by 1040 the
reactor was again made critical and by 1321 the plant returned to 11 ode 1
(Power Operation).
The resident inspector arrived in the control room soon after the plant trip
and observed the plant shutdown.
The inspector reviewed this event and
noted that the circuit breakers involved were clearly marked.
Thus it
appears that the event was strictly the result of operator error. The
inspectors have no further question on this event at this time an will
review the licensee's post trip report.
Inspector Followup Item (302/81-15-12):
Review the lessons' learned post
trip report for the July 31 reactor trip.
. . . _ .