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{{Adams | |||
| number = ML20135G039 | |||
| issue date = 09/06/1985 | |||
| title = Insp Repts 50-327/85-26 & 50-328/85-26 on 850706-0805. Violation Noted:Failure to Follow Procedures for Whole Body Frisking & to Follow Procedures to Document & Correct Individual RPI Module Deficiency | |||
| author name = Jenison K, Watson L, Weise S | |||
| author affiliation = NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) | |||
| addressee name = | |||
| addressee affiliation = | |||
| docket = 05000327, 05000328 | |||
| license number = | |||
| contact person = | |||
| document report number = 50-327-85-26, 50-328-85-26, IEIN-84-31, NUDOCS 8509180072 | |||
| package number = ML20135G032 | |||
| document type = INSPECTION REPORT, NRC-GENERATED, INSPECTION REPORT, UTILITY, TEXT-INSPECTION & AUDIT & I&E CIRCULARS | |||
| page count = 16 | |||
}} | |||
See also: [[see also::IR 05000327/1985026]] | |||
=Text= | |||
{{#Wiki_filter:. - . . . . .-. . -. | |||
e - | |||
UNITED STATES | |||
4 #p>* MGoq'o NUCLEAR REGULATORY COMMISSION | |||
[ - | |||
'n REGION ll | |||
5 101 MARIETTA STREET N.W. | |||
* *" *j ATLANTA, GEORGI A 3o323 | |||
, | |||
\...../ | |||
i Report Nos.: 50-327/85-26 and 50-328/85-26 | |||
Licensee: Tennessee Valley Authority | |||
6N11 B Missionary Ridge Place | |||
Chattanooga, TN 37402 | |||
Docket Nos.: 50-327 and 50-328 | |||
' | |||
License Nos.: DPR-77 and DPR-79 | |||
Facility Name: Sequoyah.1 and 2 | |||
Inspection Conducted: July 6 - August 5, 1985 | |||
Inspectors: 6 ./) . - d w h J, 9/4/85 | |||
q K.M.Jenis66,SenfofResidentInspector Dat( Signed | |||
G 0. n v: c/C/85 t | |||
t | |||
. J. Watson,LResidep ~Insp c~ tor DatV Sfgned | |||
Accompanying Insp ct r: J. W. York, Senior Resident Inspector, Bellefonte | |||
Approve by: h7 A 7 | |||
S. P. $1se, Section Chief Date S'igned | |||
Division of Reactor Projects | |||
SUMMARY | |||
, | |||
Scope: This routine, announced inspection involved 256 resident inspector-hours | |||
' | |||
onsite in the areas of operational safety verification including operations | |||
performance, system lineups, radiation protection, security and housekeeping | |||
inspections; surveillance and maintenance observations; review of previous | |||
inspection findings; followup of events; review of licensee identified items and | |||
. review of licensee response to NRC IE Information Notice 84-31. | |||
Results: In the areas inspected, two violations were identified (Failure to | |||
follow procedures for whole body frisking after exit from a contaminated zone | |||
(paragraph 5); and Failure to follow procedures to document and correct an | |||
Individual Rod Position. Indication module deficiency (paragraph-5)). | |||
, | |||
8509180072 850906 | |||
i PDR ADOCK 05000327 | |||
G PDR | |||
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. | |||
REPORT DETAILS | |||
1. Persons Contacted | |||
Licensee Employees | |||
*H. L. Abercrombie, Site Director | |||
*P. R. Wallacc, Plant Manager i | |||
*L. M. Nobles, Operations and Engineering Superintendent | |||
*B. M. Patterson, Maintenance Supervisor | |||
M. R. Harding, Engineering Group Supervisor | |||
; J. M. Anthony, Operations Group Supervisor | |||
D. C. Craven, Quality Assurance Supervisor , | |||
D. E. Crawley, Health Physics Supervisor ' | |||
*J. L. Hamilton, Quality Engineering Supervisor | |||
*G. B. Kirk, Compliance Supervisor , | |||
*D. H. Tullis, Mechanical Maintenance Group Supervisor , | |||
*D. L. Love, Mechanical Maintenance Engineering Section Supervisor | |||
*J. T. Crittenden, Public Safety Supervisor | |||
f | |||
Other licensee employees contacted included technicians, operators, shift | |||
engineers, security force members, engineers, and maintenance personnel. | |||
* Attended exit interview | |||
2. Exit Interview | |||
1 | |||
The inspection scope and findings were summarized with the Plant Manager and | |||
members of his staff on August 7, 1985. Violations described in paragraphs | |||
5.a. and 5.d. were discussed. The licensee acknowledged the inspection | |||
findings. The licensee did not identify as proprietary any material | |||
reviewed by the inspectors during this inspection. During the reporting < | |||
period, frequent discussions were held with the Site Director, Plant Manager | |||
and his assistants concerning inspection findings. At no time during the | |||
inspection was written material provided to the licensee by the inspector. | |||
3. Licensee Action on Previous Enforcement Matters (92702) | |||
(Closed) Violation 328/83-29-01. The licensee's response of February 8, | |||
1984, was reviewed and the indicated corrective actinns were audited. The | |||
licensee took administrative disciplinary action in this case which involved | |||
an operator who failed to follow established system alignment procedures. | |||
System Operating Instruction S01-14.3, Condensate Demineralizer Waste | |||
Disposal, was amended to include independent verification. Plant personnel | |||
have also received additional training in. procedural compliance. The | |||
licensee's corrective actions are considered complete. | |||
t | |||
.-. .. . | |||
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. | |||
2 | |||
(Closed) Violation 328/83-29-03. The licensee's response of February 8, | |||
1984, was reviewed and the indicated corrective actions were audited. The | |||
licensee took administrative disciplinary action in this case which involved | |||
an operator who failed to follow established system operating procedures. | |||
Also, plant personnel have received additional training in procedural | |||
compliance. The licensee's corrective actions are considered complete. | |||
(Closed) Violation 327, 328/84-25-02. The licensee response of November 23, | |||
1984, was reviewed and the indicated corrective action was audited. The | |||
licensee submitted a revision to LER 84055 describing the events which led | |||
to the inadequate LER and providing additional information on the breaches | |||
of the ABSCE. LER 84055, Revision 1, was closed in Inspection Report 327, | |||
328/85-16. In addition, the inspector noted an improvement in the quality | |||
of LER submittals. The licensee's corrective actions are considered | |||
complete. | |||
J | |||
(Closed) Violation 327, 328/84-29-01. The licensee's response of | |||
January 18, 1984, was reviewed and the indicated corrective action was | |||
audited. The licensee revised Surveillance Instruction SI-256.2, Inspection | |||
of Molded Case and Lower Voltage Circuit Breakers, to provide alternate | |||
power to breaker 213 en the 125 volt Vital Battery Board III when the | |||
4 | |||
breaker is being tested and to include a signcff to ensure that the alter- | |||
nate power is provided to the fuse column. | |||
(Closed) Violation 327, 328/84-35-01. The licensee's response of | |||
January 21, 1985, was reviewed and the indicated corrective action was | |||
audited. The licensee reviewed System Operating Instruction SOI-63.1 and | |||
verified that the procedure had been revised to include the two missing | |||
level transmitter root valves for the cold leg accumulators. The licensee's | |||
corrective actions are considered complete. | |||
4. Unresolved Items | |||
Unresolved items are matters about which more information is required to | |||
determine whether they are acceptable or may involve violations or devia- | |||
tions. One unresolved item identified during this inspection is discussed | |||
in paragraph 8. | |||
5. 0perational Safety Verification (71707) | |||
a. Plant Tours | |||
The inspectors observed control room operations, reviewed applicable | |||
logs, conducted discussions with control room operators, observed shift | |||
turnovers, and confirmed operability of instrumentation. The | |||
inspectors verified the operability of selected emergency systems, | |||
reviewed tagout records, verified compliance with Technical Specifica- | |||
tion (TS) Limiting Conditions for Operations (LCO) and verified return | |||
to service of affected components. The inspector verified that | |||
maintenance work orders had been submitted as required and that | |||
followup activities and prioritization of work was accomplished by the | |||
- - - | |||
- | |||
. | |||
3 | |||
licensee. Tours of the diesel generator, auxiliary, turbine buildings | |||
were conducted to observe plant equipment conditions, including | |||
potential fire hazards, fluid leaks, and excessive vibrations and plant | |||
housekeeping / cleanliness conditions. | |||
During the performance of a routine Unit 2 Control Room tour the | |||
inspectors noticed a wedge of paper pressed between two Individual Rod | |||
Position Indication (IRPI) modules on the main control panel 2-M-4. | |||
This same wedge of paper was' still in the panel three days later on a | |||
succeeding inspection tour. When the assigned Reactor Operator was | |||
questioned about the foreign object, he stated that the paper was to | |||
keep the module from vibrating. Vibrations cause the rear contacts on | |||
the module to become loose, resulting in a loss of indication and an | |||
inoperable module. No Maintenance Request (MR) had been written to | |||
repair the module, and there was no Temporary Alteration Control Form | |||
(TACF) prescribing the use of the paper wedge. Administrative | |||
Instruction 12 requires that personnel formally identify deficiencies | |||
and that such deficiencies be promptly identified and corrected. | |||
Failure to follow procedures .foi ' entification and correction of the | |||
vibration deficiency affecting o,mrability of the subject IRPI is a | |||
violation (328/85-26-09). | |||
The inspectors walked down accessible portions of the following | |||
safety-related systems on Unit I and Unit 2 to verify operability and | |||
proper valve alignment: | |||
'' | |||
Safety Injection System (U$its 1 and 2) | |||
: Turbine Driven Auxiliary Feedwater System (Units 1 and 2) | |||
Upper Head Injection System (Units 1 and 2) | |||
Diesel Generators (Units 1 and 2) | |||
Emergency Gas Treatment System (Units 1 and 2) | |||
On July 17, 1985, during a walkdown of diesel . generator 1A1, the | |||
inspectors observed an air leak in pressure reducing valve | |||
0-FCV-82-172A. The licensee wrote MR A-537220 to correct the | |||
deficiency. It was determined that the body of the valve was cracked. | |||
A new valve has been ordered and the valve will be replaced during the | |||
first surveillance (i.e. , the next time the diesel will be _out of | |||
service) after the valve is received. The licensee stated that the DG | |||
was operable due to the redundant air start system for the engine. The | |||
DG was started successfully on August 7, 1985. Repair of this valve is | |||
identified as Inspector Followup Item (327, 328/85-26-01). | |||
b. Verification of Ice Condenser Door Operability | |||
The inspector's review of ice condenser door operability is documented | |||
in Inspector Report 327, 328/85-23. During this inspection period, the | |||
inspector discussed the methods for independent verification of removal | |||
of door blocks from the lower ice condenser doors with the licensee. | |||
The licensee stated that the lower ice condenser doors would be | |||
numbered and that a double visual verification of the removal of the | |||
i | |||
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. | |||
4 | |||
door blocks would be conducted and double signoffs recorded for each | |||
door. This is identified as Inspector Followup Item (327, 328/ | |||
85-26-02) until licensee actions are completed. | |||
c. Security | |||
During the course of the inspection, observations relative to protected | |||
and vital area security were made, including access controls, boundary | |||
integrity, search, escort, and badging. Two security concerns were | |||
identified during the inspection period which will be reviewed by NRC | |||
Region II security inspectors for potential violations. These concerns | |||
involved movement of materials into the protected area and the | |||
visibility of the protected area boundary. | |||
d. Radiation Protection | |||
The inspectors observed Health Physics (HP) practices and verified | |||
implementation of radiation protection control. On a regular basis, | |||
radiation work permits (RWPs) were reviewed and specific work | |||
activities were monitored to assure the activities were being conducted | |||
in accordance with applicable RWPs. Selected radiation protection | |||
instruments were verified operable and calibration frequencies were | |||
reviewed. | |||
On July 16, 1985, after observing a functional test on the B train | |||
waste gas compressor, the inspector exited the contaminated zone and | |||
went to the personnel frisking station on EL690. After frisking, the | |||
inspector noted that two individuals who had also exited the | |||
contaminated zone had not arrived at the frisking station. The | |||
inspector found the individuals outside the regulated area. The | |||
inspector interviewed the two individuals and determined that they had | |||
not conducted a whole body frisk when exiting the contaminated zone. | |||
The individuals had used a hand and foot monitor at the exit of the- | |||
regulated zone. The inspector discussed the incident with Health | |||
Physics. The two individuals were recalled, a whole body frisk | |||
performed, and surveys conducted of the areas the individuals had | |||
entered after leaving the contaminated zone. The hand and foot monitor | |||
and whole body and area surveys performed by the licensee after the | |||
incident indicated that the individuals were not contaminated. Radio- | |||
logical Control Instruction RCI-1, Radiological Hygience Control, which | |||
was established to implement the requirements of Technical Specifica- | |||
tion 6.11, states that individuals exiting a contaminated zone are | |||
required to perform a whole body frisk to prevent the spread of | |||
contamination to other areas. RCI-14, Radiation Work Permit (RWP) | |||
Program, states that it is the responsibility of each employee to | |||
adhere to the requirements of RWPs and RWP Timesheets. RWP 02-0-85663, | |||
which was issued to control access to the waste gas compressor room on | |||
July 16, 1985, required employees to perform a whole body frisk upon | |||
exit from the contaminated room. The failure to perform a whole body | |||
frisk is a violation (327, 328/85-26-03). | |||
- | |||
. | |||
5 | |||
On August 5, 1985, during a routine tour of the Auxiliary Building, the | |||
inspector observed several yellow poly bags on an Instrument Mainte- | |||
nance Section cart which were labeled only with the sections's name and | |||
telephone number. The material was unattended in a remote location and | |||
not marked with survey data or radiation warning tape. The inspector | |||
requested Health Physics (HP) to survey the bags. The inspector | |||
discussed the observation with the licensee. The licensee stated that | |||
the bags contained instruments which had been used inside a contami- | |||
nated zone during the morning and were to be used in another | |||
contaminated zone during the afternoon. The technicians using the | |||
instruments had left the regulated area for lunch. Although the bags | |||
were unattended and not marked, the licensee felt that the technicians | |||
were cognizant of the location and content of the bags and were | |||
controlling the material properly. The inspector discussed the | |||
observation with the NRC Region II office and determined that the | |||
procedure was act.aptable in cases where work is still in progress, the | |||
articles are contained in yellow poly bags with identification of the | |||
responsible section on the bag, and licensee employees have been | |||
trained to recognize that material in yellow poly bags are potentially | |||
contaminated. Since the procedure which controls the movement of | |||
radioactive material inside the regulated area, RCI-1, was not clear in | |||
how this situation should be handled, the licensee stated that the | |||
procedure would be revised to address these requirements. The licensee | |||
was cited previously in Inspection Report 327, 328/85-20 for not | |||
labeling a yellow poly bag containing contaminated material which had | |||
been left unattended in the Auxiliary Building on May 22, 1985. | |||
Corrective actions for this previous violation are still under review. | |||
6. Engineered Safety Features Walkdown (71710) | |||
The inspector verified operability of the centrifugal charging pump flowpath | |||
through the boron injection tank on Units 1 and 2 by performing a complete | |||
walkdown of the accessible portions of the systems. The following specifics | |||
were reviewed and/or observed as appropriate: | |||
a. that the licensee's system lineup procedures matched plant drawings and | |||
the as-built configuration; | |||
b. that equipment conditions were satisfactory and items that might | |||
degrade performance were identified and evaluated (e.g. , hangers and | |||
supports were operable, housekeeping, etc., was adequate); | |||
c. with assistance from licensee personnel, the interior of the breakers | |||
and electrical or instrumentation cabinets were inspected for debris, | |||
loose material, jumpers, evidence of rodents, etc.; | |||
d. that instrumentation was properly valved in and functioning and cali- | |||
bration dates were appropriate; | |||
e. that valves were in proper position, breaker alignment was correct, | |||
power was available, and valves were locked as required; and | |||
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6 | |||
f. local and remote instrumentation was compared and remote instrumenta- | |||
tion was functional. | |||
No violations or deviations were identified. | |||
7. Monthly Surveillance Observation (61726) | |||
The inspectors observed Technical Specification (TS) required surveillance | |||
testing and verified that testing was performed in accordance with adequate | |||
procedures; that test instrumentation was calibrated; that Limiting | |||
Conditions for Operation were met; that test results met acceptance criteria | |||
requirements and were reviewed by personnel other that the individual direc- | |||
ting the test; that deficiencies identified, as appropriate, and that any | |||
deficiencies identified during the testing were properly reviewed and | |||
resolved by management personnel; and that system restoration was adequate. | |||
For complete tests, the inspector verified that testing frequencies were met | |||
and tests were performed by qualified individuals. | |||
The inspector witnessed / reviewed portions of the following surveillance test | |||
activities: | |||
SI-7, Electrical Power System: Diesel Generators | |||
SI-64, Boric Acid Flow Paths - Valve Position Verification | |||
SI-281, Functional Tests of Radiation Effluent Monitors with Automatic | |||
Actuations (Quarterly) | |||
ST -'5, Channel Calibrations for Radiation Monitoring Systems | |||
No violatict.s or deviations were identified in this area. | |||
8. Monthly Maintenance Observations (62703) | |||
a. Station maintenance activities of safety-related systems and components | |||
were observed / reviewed to ascertain that they were conducted in | |||
accordance with approved procedures, regulatory guides, industry codes | |||
and standards, and in conformance with TS. | |||
The following items were considered during this review: LCOs were met | |||
while components or systems were removed from service; redundant | |||
components were operable; approvals were obtained prior to initiating | |||
the work; activities were accomplished using approved procedures and | |||
were inspected as applicable; procedures used were adequate to control | |||
the activity; troubleshooting activities were controlled and the repair | |||
record accurately reflected what took place; functional testing and/or | |||
calibrations were performed prior to returning components or systems to | |||
service; quality control records were maintained; activities were | |||
accomplished by qualified personnel; parts and materials used were | |||
properly certified; radiological controls were implemented; QC hold | |||
points were established where required and were observed; fire | |||
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. | |||
7 | |||
prevention controls were implemented; outside contractor force | |||
activities were controlled in accordance with the approved Quality | |||
Assurance (QA) program; and housekeeping was actively pursued. | |||
b. Corrective maintenance was reviewed on containment isolation valve | |||
2-67-580D, Essential Raw Cooling Water supply to upper containment | |||
cooler 20. The following documents were reviewed: | |||
Maintenance Requests: A-291918 A-024413 | |||
A-533609 A-112021 | |||
A-128336 A-151118 | |||
A-024411 A-024414 | |||
A-024415 | |||
Maintenance Instructions: 6.20, Configuration Control During | |||
Maintenance Activities | |||
11.4, Maintenance of CSSC Valves | |||
6.21, Repairs and <: placements of | |||
ASME Section XI Components | |||
6.15, General Procedure, Tightening | |||
bolts | |||
Temporary Alteration Control Forms: 2-85-2012-67, 1-85-5007-67 | |||
Surveillance Instruction: SI-158.1, Containment Isolation | |||
Valve Leak Rate Test | |||
Isolation check valves 580A through 0_ are scheduled for replacement | |||
during the next outage period. These valves normally allow ERCW flow | |||
through the upper compartment air cooler and become crud traps as a | |||
result of the low flow rates and the raw condition of the ERCW. Review | |||
of MR-A151118 indicated that a waxy substance had been found during | |||
maintenance on the disc of valve 2-67-580D. After interviews with | |||
those technicians that performed the disc replacement maintenance on | |||
valve 2-67-5800, it appeared that no foreigr. material was placed into | |||
the valve. Several tests were conducted in order to get the valve to | |||
seat correctly and to pass the leak test. This brings the repro- | |||
ducibility of the test into question. In addition, post modification | |||
test data was not taken and treated as a Quality Record for those tests | |||
that failed leak testing. The licensee is committed to ANSI standard | |||
18.7-1976, which requires that' post maintenance test failures be | |||
treated as quality records. Such records provide maintenance history | |||
data for trending and planning purposes. Tests that are intended only | |||
as troubleshooting diagnostic devices and on which operability of the | |||
valve would not be based appear not to require retention. Because of | |||
the iterative nature of lapping valve internals, the status of the | |||
failed tests is unclear as a quality record. The valves were turned | |||
over to a separate group (Engineering Test) for test evaluation four | |||
times. During interviews with the technicians that performed the | |||
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8 | |||
maintenance on 2-67-580D, the inspector also learned that its internal | |||
flapper arm was out of round, but not replaced due to lack of parts. | |||
This valve is scheduled to be retested in October 1985 and the results | |||
will be evaluated by the inspector. This is an Unresolved Item (327, | |||
328/85-26-04) until the inspector further evaluates licensee mainte- | |||
nance controls and retention of test records associated with this and | |||
other maintenance. | |||
c. On July 24, 1985, the inspector observed a post maintenance functional | |||
test on the B train waste gas compressor. The following documents were | |||
reviewed: | |||
Maintenance Request (MR) A526712 | |||
Hold Order 993 | |||
Maintenance Instruction MI-8.20 | |||
The compressor was identified on the maintenance request as CSSC | |||
. equipment. The licensee stated that the equipment was QA controlled | |||
but was not safety-related. Maintenance Instruction MI-8.20 provided | |||
detailed steps on performance of the maintenance; however, the | |||
procedure only stated that a functional test was to be performed after | |||
the maintenance. The procedure did not provide any criteria for the | |||
test. This concern was discussed with the licensee. The licensee | |||
stated that functional test criteria would be included in tests for | |||
CSSC equipment that is not safety-related. The licensee is also | |||
examining assignment of responsibility for functional tests to assure | |||
appropriate test control. This is identified as Inspector Followup | |||
Item (327, 328/85-26-05). During the system lineup, the inspector | |||
observed an Auxiliary Unit Operator (AU0) misalign several root valves. | |||
The AU0 later corrected the misaligned valves prior to completion of | |||
the test. The licensee stated that the AVO has received additional | |||
training in the requirements for system alignment. The licensee is | |||
evaluating the need for control of root valves on System Operating | |||
Instructions. The licensee presently controls the position of root | |||
valves by Hold Orders and with Instrument Maintenance procedures. This | |||
is identified as Inspector Followup Item (327, 328/85-26-06). | |||
9. Licensee Event Report (LER) Followup (92700) | |||
The following LERs were reviewed and closed. The inspector verified that: | |||
reporting requirements had been met; causes had been identified; corrective- | |||
actions appeared appropriate; generic applicability had been considered; the | |||
LER forms were completed; the licensee had reviewed the event; no unreviewed | |||
safety questions were involved; and violations of regulations or Technical | |||
Specification conditions had been identified. | |||
a. LERs Unit 1 | |||
' | |||
327/83012 Feedwater Flow Channel Declared Inoperable | |||
, 327/83019 Rod Position Indicator Declared Inoperable | |||
.. __ _ _ _ _ . _ _ __ | |||
f | |||
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1 | |||
l | |||
9 | |||
327/83021 Failure of Heat Tracing on BIT | |||
' | |||
327/83025 Inoperable Flow Rate Monitor for Shield Building | |||
Exhaust | |||
327/83026 Failure of ERCW Valve | |||
> | |||
; 327/83056 Inoperable Upper Containment Personnel Inner Airlock | |||
; Door | |||
. | |||
t | |||
327/83065 Inoperable Train of EGTS | |||
327/83070 Diesel Generator Declared Inoperable | |||
i | |||
327/83080 Inoperable Rod Position Indicator | |||
327/83110 Failure of EGTS Cooldown Valves | |||
~ | |||
' | |||
327/83137 Diesel Generator 28-B Inoperable | |||
: | |||
. 327/83180 Thermal Overload Devices Failed to Trip Check | |||
327/84013 Reactor Trip on Low-Low Steam Generator Level | |||
! 327/84018 Fire Protection Deluge Valve Isolated | |||
327/85008 Failure to Meet Fire Watch Requirements | |||
i 327/85010 Containment Ventilation Isolation | |||
327/85011 Failure to Meet Fire Watch Requirements | |||
, | |||
327/85012 Failure to Meet Fire Watch Requirements | |||
t | |||
i 327/85013 Failure to Meet Fire Watch Requirements | |||
: | |||
' - | |||
327/85015 Failure to Meet Fire Watch Requirements | |||
327/85017 Inadvertent Auxiliary Building Isolations | |||
, | |||
327/85026 Inadvertent Feedwater Isolation | |||
b. LERs Unit 2 | |||
4 | |||
328/83004 RWST Low Boron Concentration | |||
; and Rev. 1 | |||
328/83051 BIT Valve Inoperable | |||
i 328/83059 Rod Position Indication Inoperable | |||
i | |||
i | |||
!- | |||
-. , . ,_ . - - . _ __ _ _ _ - --- - _ _ _ _ _ _ - _ _ - . _ ~ . _ - . . . . _ . _._, | |||
- - . . . . . .. . | |||
. | |||
- | |||
10 | |||
328/83068 RWST Low Water Level | |||
328/83081' Upper Containment Airlock Door Inoperable | |||
J | |||
328/83082 RHR Remote Shutdown Channel Inoperable | |||
4 | |||
328/83093 Inoperable Flow Rate Monitor | |||
328/83121 Failure of Lower Containment Airlock to Meet Leakage | |||
Criteria | |||
328/83124 Inoperable Con' denser Vacuum Exhaust Flow Rate Monitor | |||
328/83131 Inoperable Condenser Vacuum Exhaust Flow Rate Monitor | |||
328/83153 Containment Airlock Door Not Fully Closed | |||
328/83161 Rod Bottom Light Bistable Inoperable | |||
l 328/83163 Inoperable Rod Position Indicator | |||
328/84003 Inadvertent Containment Building Ventilation Isolations | |||
328/85005 Debris Inside Containment | |||
c. Followup on LER 328/85001 - Unit 2 Reactor Trip on January 14, 1985 | |||
, | |||
Surveillance Instruction (SI-80), Power Range Neutron Flux Channel | |||
Calibration and Functional Test, was conducted on Unit 2 Nuclear | |||
Instrument Power Range Channel N-41. SI-80 referenced Instrument | |||
. | |||
' | |||
Maintenance Instruction (IMI-92-PRM-CAL), NIS Power Range, as the | |||
required calibration procedure. Step 5.2.1.6 of IMI-92-PRM-CAL | |||
required that the instrument power fuses be removed and independently | |||
verified. When the instrument power fuses are removed from a Power | |||
Range (PR) Nuclear Instrument (NI) drawer, the high voltage supply to | |||
that drawer is removed and a negative rate flux trip bistable is | |||
activated. The negative rate flux trip bistable does not automatically | |||
reset when the instrument power fuses are replaced and must be manually | |||
' | |||
reset. If two separate NI PR negative rate flux trip bistables are | |||
activated at the same time, a reactor trip results. | |||
A Senior Instrument Maintenance (IM) technician removed the instrument | |||
power fuses from PR NI N-42 in error, replaced the fuses, but did not | |||
reset the reset the channel. When the IM technician subsequently | |||
deenergized PR NI drawer NI-41, by removing the instrument power fuses, | |||
, | |||
the corresponding negative rate was activated. This satisfied a | |||
] two-out-of-four reactor protection system logic, resulting in a Unit 2 | |||
i reactor trip from 30's power. The removal of the instrument power fuses | |||
, was determined to be a personnel error by the licensee. The inspectors | |||
; verified this through interviews with personnel involved and a review | |||
i of the licensee's IM training and qualification system. However, | |||
. __ _. __ | |||
_ __ __ _ _ _, | |||
_ | |||
. - | |||
11 | |||
sev-:11 actions taken by the technician after removal of the fuses are | |||
examples of a failure to follow procedure. These actions are explained | |||
: below. | |||
Following the removal of the PR NI drawer N-42 instrument power fuses, | |||
the IM technician failed to follow procedures in two instances. First, | |||
he attempted to place PR NI N-42 back in service by replacing its | |||
instrument power fuses. The appropriate section of IMI-92-PRM-CAL | |||
(Section 7) was not implemented to ensure that N-42 was correctly | |||
returned to service. Second, after the IM technician had returned PR | |||
NI drawer N-42 to what he assumed to be an operable condition, he | |||
commenced surveillance activities on PR NI N-41 without reporting to | |||
the RO or SRO that he had not complied with an assigned procedure and | |||
had affected safety-related equipment not authorized for su,rveillance | |||
activities. | |||
Administrative Instruction AI-12, Adverse Conditions and Corrective | |||
Actions, states that plant personnel shall report any suspected | |||
abnormal plant condition adverse to quality in the performance of their | |||
regular work duties. A failure to follow procedure is defined in | |||
section 4.0 of AI-12 as a condition adverse to quality. The inspector | |||
is not satisfied that licensee personnel clearly understand their - | |||
i responsibility to report to the shift operating staff any unintended | |||
* | |||
affects on safety-related equipment during planned activities. This | |||
issue will be reviewed as Inspector Followup Item (327, 328/85-26-07). | |||
The LER is closed. | |||
10. Event Followup (93702, 62703, 61726) | |||
a. Unit 1 Reactor Trip Due to Loss of Main Feedwater Pump Turbine (MFPT) | |||
Oil Pump | |||
On July 19, 1985, Unit I tripped from 100 percent power on low-low | |||
steam generator level. The licensee was utilizing normal procedures to | |||
search for a ground on a turbine unit board. In this process, 480V | |||
Unit Board 1B was switched from the normal power supply to an alternate | |||
supply. The operator closed the alternate supply breaker to the board, | |||
but the breaker did not close the breaker completely. He then | |||
deenergized the normal supply. He immediately noted that the MFPT oil | |||
pump light on the board went out and jerked up on the alternate supply | |||
breaker completely closing it; however, the oil pump tripped. | |||
A backup oil pump started automatically on the trip signal; however, | |||
the momentary drop in oil pressure caused partial completion of. logic | |||
associated with the MFPT oil pressure switch. This logic includes | |||
closing of the train associated MFPT stop valves, a main turbine | |||
runback, and start of the motor driven (MDAFWP) and turbine driven | |||
(TDAFWP) auxiliary feedwater pumps among others. The logic was | |||
completed for the closing of the stop valves resulting in coasting down | |||
of the MFPT; however, the turbine runback logic was not completed and | |||
steam generator levels started to decrease. The MDAFWPs started, and | |||
-. . .-. . - - . | |||
. _ _ _ .. . _ _ . . | |||
. | |||
12 - | |||
i | |||
the TDAFWP did not. The operators immediately put the control rods in | |||
auto and attempted to manually runback the main turbine. This was | |||
unsuccessful, and the reactor tripped on low-low stcam generator level. | |||
All reactor protection system functions reacted normally on the reactor | |||
trip including start of the TDAFWP. | |||
The logic associated with the MFPT pressure switch is not a reactor | |||
protection or engineered safety feature signal. The purpose of the | |||
switch is to provide anticipatory features, such as turbine runback, | |||
for events which could lead to low-low steam generator reactor trips. - | |||
The inspector reviewed the logic with a Senior Reactor Operator and | |||
discussed the failure of the logic circuit to completely energize with | |||
the licensee. This problem was attributed to the fast operation of the | |||
pressure switch resulting in only a momentary contact in the logic | |||
circuit. The mechanical contacts eif.her did not have time to makeup or | |||
the seal-in circuit was not energized. | |||
! | |||
.The licensee attempted to duplicate the initiating conditions through | |||
. | |||
testing to ascertain if any problems existed within the logic | |||
1 | |||
circuitry; however, the condition could not be duplicated. This | |||
testing and performance of Surveillance Instruction SI-110.1, TDAFWP | |||
and Valve Automatic Actuation, resulted in completion of the full | |||
logic. The licensee adjusted the pressure switch which starts the | |||
backup oil pump to a higher setting to preclude a significant drop in | |||
oil pressure on trip of one of the oil pumps. This should prevent a | |||
similar event. | |||
No violations or deviations were identified. | |||
b. Unit 2 Notification of Unusual Event Due to Excessive Reactor Coolant | |||
System (RCS) Leakage | |||
On July 29, 1985, operators noted a decrease in volume control tank | |||
level. Subsequently, a high radiation alarm on the Auxiliary Building | |||
Vent particulate monitor caused an Auxiliary Building isolation. The | |||
operators followed the requirements of Abnormal Operating Instruction | |||
AOI-31, Abnormal Release of Radioactive Materials, and AOI-6, Small | |||
Reactor Coolant System Leak. The Auxiliary Building was evacuated and | |||
i posted as an airborne radiation area. The operators placed excess | |||
letdown in service and secured the normal charging and letdown flow- | |||
paths. | |||
. A few minutes later, a fire watch monitoring temperatures in various | |||
: | |||
pipe chases reported a high temperature in the Unit 2 EL690 pipe chase. | |||
The Unit 2 Assistant Shift Engineer (ASE) responded to the area and | |||
encountered two electricians exiting the area. The electricians | |||
reported that there was a steam leak in the pipe chase. (The | |||
electricians were subsequently decontaminated as discussed below.) The | |||
ASE investigated the report and determined that weld upstream of sample | |||
valve 62-674 was leaking. The weld was located at the point where the | |||
- .- . . - . - , ._. , . - _ - - - - . | |||
. | |||
. - - - - . - -. - . . , | |||
. _ - | |||
. * | |||
13 | |||
one inch sample line to the hot sample room joined the three inch | |||
normal CVCS letdown line. | |||
The leak rate was determined to be about 15 gpm. The licensee | |||
* | |||
estimated that approximately 600 gallons of RCS water was released | |||
during the event. In addition, due to backleakage across a check valve | |||
between the Volume Control Tank and the break, a small amount of | |||
hydrogen was released from the VCT to the pipe chase. This leak was | |||
promptly isolated. The licensee repaired the line and returned the | |||
normal charging and letdown flowpaths to service on August 1, 1985. | |||
, The licensee removed the segment of the sample line affected and sent | |||
the weld to a metallurgical lab for analysis. It was determined that | |||
the break which was in the heat affected zone above the weld was due to | |||
; fatigue from high cycle vibration. Two initiation points approximately | |||
' | |||
180 degrees apart were observed. The licensee stated that these points | |||
were due to high cycle, low stress fatigue. The break propagated about | |||
300 degrees. The licensee has evaluated the configuration and has | |||
installed a hanger on the line. The installation will be evaluated to | |||
assure that the vibration is corrected. The weld on the Unit I sample | |||
line was visually examined, and no defects were identified. An | |||
appropriate hanger configuration will be installed on Unit I when the | |||
Unit 2 analysis is complete. Completion of the harger work will be | |||
: followed by the inspector and is identified as Inspector Followup Item | |||
(327,328/85-26-10). | |||
Eleven personnel suffered skin contamination as a result of tne primary | |||
leak described above. The most severe case was 11,000 dpm on one | |||
person's face. All personnel were decontaminated using normal methods | |||
(showers) and had no detectable contaminatica after they were | |||
decontaminated. | |||
On July 30, 1985, the licensee posted the control building and an area | |||
adjacent to the entrance to the Auxiliary Building on EL690 as airborne | |||
areas. The air samples from these areas had been analyzed in the plant | |||
chemistry lab. The readings were later determined to be erroneous due | |||
to the high background in the lab by sending samples to the Power | |||
Operations Training Center laboratory for analysis. These samples | |||
indicated no airborne radiation was present. | |||
While placing the Auxiliary Building general supply fans back in | |||
service, three Auxiliary Building isolations occurred due to spiking on | |||
0-RM-90-1018. The licensee stated that the setpoint on this monitor | |||
was reset to a higher value which was still within the range required | |||
to meet Technical Specification 3.11.2.1. Verification of the setpoint | |||
change by the inspector is identified as Inspector Followup Item 327, | |||
328/85-26-08. The Auxiliary Building Vent System was returned to | |||
service with the Auxiliary Building Gas Treatment System operating to | |||
reduce airborne levels in the Auxiliary Building. | |||
No violations or deviations were identified. | |||
i | |||
.-. - _ - - _ - - , - - | |||
_ | |||
l | |||
. * | |||
14 | |||
11. Review of Licensee Actions on NRC Information Notice 84-31 (92703) | |||
The inspectors examined the licensee's handling of. NRC Information Notice | |||
- | |||
84-31, Increased Stroking Time of Bettis Actuators Because of Swollen | |||
Ethylene - Proplyene Rubber Seals and Seal Set, dated April 30, 1984. The | |||
; | |||
information notice described the problem as follows: | |||
The G. H. Bettis Company is a supplier of actuators used principally in | |||
heating, ventilating and air-conditioning (HVAC) safety-related systems. | |||
The G. H. Bettis Company notified the. NRC via a Part 21 report that their | |||
NCB series, N52X, N72X, N73 series, and the NT310-SR4 and 5 and NT312-SR5 | |||
; actuators had potential stroking times of greater than the required 15 | |||
seconds because the EP elastomers in contact with the Mobil 28 grease | |||
l lubricant could swell. (The 15-second stroking time was used by Bettis as | |||
typical of customer requirements.) The actuator seals swell when in contact | |||
with the Mobile 28 grease currently used in the manufacture of "N" series | |||
actuators. Where it is necessary to replace swollen seals, Bettis | |||
, | |||
' | |||
recommended replacing them with new seals and using Dow-Corning Molykote 44 | |||
grease, a silicon based lubricant which Bettis states has been shown to | |||
7 | |||
cause no seal degradation and adequate lubrication. | |||
. | |||
The G. H. Bettis Company also identified another problem that could , | |||
adversely affect stroking time. Their report states that ". . .the magnitude ; | |||
of stroking time degradation is related to the elapsed time between actuator | |||
i | |||
cycles. The longer the actuator remains stationary the more " set" the seals | |||
, take. The set characteristic causes the seal to form an intimate contact | |||
1 | |||
with the sealing surfaces, further increasing the time required to | |||
initialize stroke. Once the actuator begins to stroke, the seals begin to | |||
* | |||
recover their original shape, thus freeing the unit up. Stroking the | |||
actuator three or more complete cycles using pressurized gas will cause the | |||
seals to recover sufficiently to reduce stroking time to a minimum. No seal | |||
degradation has been traced to periodic actuator stroking, quite the | |||
' | |||
opposite has been experienced. Frequent stroking tends to extend seal life , | |||
resulting in longer actuator cycle life." | |||
The inspectors reviewed a memorandum from H. A. Abercrombie, Director of | |||
. | |||
Nuclear Services, to Sequoyah and other sites, dated August 22,. 1984 | |||
! indicating that all safety-related valves and dampers of the model series | |||
! described in the notice should have the seals and lubricant replaced at the | |||
j first available outage. The memorandum requested that a search be made of | |||
site documentation in order to identify the valves and dampers involved. .t | |||
! Twenty seven (27) valves and dampers were identified at Sequoyah and three | |||
l Nonconformance Reports (NCRs) were . written. The NCRs with pertinent | |||
, information are as follows: | |||
a. NCR EEB 84-12 dated December 1984 involved valves 1-FCV-77-420 and | |||
2-FCV-77-421. Both of these valves were still in the warehouse and | |||
have been tagged as not to be used. | |||
l b. NCR MEB 84-07 dated September 1984 involved 17 valves used in three | |||
j separate areas. | |||
- - _ - _ . _ ~ _ - - . - | |||
. , . _ - . | |||
. - | |||
; | |||
15 | |||
1. Three valves are used to actuate the vacuum breaker system for | |||
containment and come under ASME Section XI rules for testing. | |||
These valves have a maximum allowed stroking time of 25 seconds | |||
and must be tested quarterly. The latest test (June 26, 1985) | |||
indicated stroking times for the three valves ranged from 6.4 to | |||
6.8 seconds. The three valve numbers are 2-FCV-30-46, | |||
2-FCV-30-41, and 2-FCV-30-40, | |||
2. Ten valves are used to actuate the Emergency Gas Treatment System. | |||
Each train is tested every other month but each of the valves is | |||
not tested individually to determine stroke time. The fastest the | |||
system has to actuate is 38 seconds and no problems have been | |||
roted during the testing. | |||
3. Four valves are Control Room HVAC isolation valves. This system | |||
is required to be tested every 18 months and was last tested | |||
August 17, 1984 with no problems. The four valve numbers are | |||
0-FCV-31A-105A, 0-FCV-31A-105B, 0-FCV-31A-106A and 0-FCV-31A-106B. | |||
The failure evaluation / engineering report (FE/ER) classified the | |||
deficient condition to be a Category 1 (acceptable for all modes of | |||
operation and design condition) based upon a sampling of the valves for | |||
stroking time, | |||
c. NCR MEB 84-08, Rev. 1, dated October 1984 involved 8 valves installed | |||
in the fifth vital battery room for protection from tornado depressuri- | |||
zation. The eight valve numbers- are FCO-51-485, FCO-31-486, | |||
FC0-31-488, FCO-31-489, FCO-31-493, FCO-31-494 and FC0-31-501. The | |||
FE/ER classified the deficient condition to be a Category I based on | |||
the fact that the valves were not installed in the system, and the | |||
; seals and lubricant could be changed before the valves would experience | |||
operational use. This was not done and this violates 10 CFR 50 | |||
LAppendix B Criterion XVI in that effective corrective action dictated | |||
by the FE/ER was not taken. However, no violation will be issued since | |||
programmatic corrective actions are currently in progress at TVA due to | |||
, | |||
an Order Modifying Licenses issued June 14, 1985 (EA 85-49). | |||
; Correction of this specific deficiency is an -Inspector Followup Item | |||
(327,328/85-26-09). | |||
! | |||
, | |||
* | |||
v +, - - ,- , -w ,,-w | |||
}} |
Latest revision as of 23:54, 27 October 2020
ML20135G039 | |
Person / Time | |
---|---|
Site: | Sequoyah |
Issue date: | 09/06/1985 |
From: | Jenison K, Linda Watson, Weise S NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
To: | |
Shared Package | |
ML20135G032 | List: |
References | |
50-327-85-26, 50-328-85-26, IEIN-84-31, NUDOCS 8509180072 | |
Download: ML20135G039 (16) | |
See also: IR 05000327/1985026
Text
. - . . . . .-. . -.
e -
UNITED STATES
4 #p>* MGoq'o NUCLEAR REGULATORY COMMISSION
[ -
'n REGION ll
5 101 MARIETTA STREET N.W.
- *" *j ATLANTA, GEORGI A 3o323
,
\...../
i Report Nos.: 50-327/85-26 and 50-328/85-26
Licensee: Tennessee Valley Authority
6N11 B Missionary Ridge Place
Chattanooga, TN 37402
Docket Nos.: 50-327 and 50-328
'
License Nos.: DPR-77 and DPR-79
Facility Name: Sequoyah.1 and 2
Inspection Conducted: July 6 - August 5, 1985
Inspectors: 6 ./) . - d w h J, 9/4/85
q K.M.Jenis66,SenfofResidentInspector Dat( Signed
G 0. n v: c/C/85 t
t
. J. Watson,LResidep ~Insp c~ tor DatV Sfgned
Accompanying Insp ct r: J. W. York, Senior Resident Inspector, Bellefonte
Approve by: h7 A 7
S. P. $1se, Section Chief Date S'igned
Division of Reactor Projects
SUMMARY
,
Scope: This routine, announced inspection involved 256 resident inspector-hours
'
onsite in the areas of operational safety verification including operations
performance, system lineups, radiation protection, security and housekeeping
inspections; surveillance and maintenance observations; review of previous
inspection findings; followup of events; review of licensee identified items and
. review of licensee response to NRC IE Information Notice 84-31.
Results: In the areas inspected, two violations were identified (Failure to
follow procedures for whole body frisking after exit from a contaminated zone
(paragraph 5); and Failure to follow procedures to document and correct an
Individual Rod Position. Indication module deficiency (paragraph-5)).
,
8509180072 850906
i PDR ADOCK 05000327
G PDR
-
. - - . . . - . - - - - - -
. , - . , - . . . - - , - . - , - , , - . - , , . .
'
.
REPORT DETAILS
1. Persons Contacted
Licensee Employees
- H. L. Abercrombie, Site Director
- P. R. Wallacc, Plant Manager i
- L. M. Nobles, Operations and Engineering Superintendent
- B. M. Patterson, Maintenance Supervisor
M. R. Harding, Engineering Group Supervisor
- J. M. Anthony, Operations Group Supervisor
D. C. Craven, Quality Assurance Supervisor ,
D. E. Crawley, Health Physics Supervisor '
- J. L. Hamilton, Quality Engineering Supervisor
- G. B. Kirk, Compliance Supervisor ,
- D. H. Tullis, Mechanical Maintenance Group Supervisor ,
- D. L. Love, Mechanical Maintenance Engineering Section Supervisor
- J. T. Crittenden, Public Safety Supervisor
f
Other licensee employees contacted included technicians, operators, shift
engineers, security force members, engineers, and maintenance personnel.
- Attended exit interview
2. Exit Interview
1
The inspection scope and findings were summarized with the Plant Manager and
members of his staff on August 7, 1985. Violations described in paragraphs
5.a. and 5.d. were discussed. The licensee acknowledged the inspection
findings. The licensee did not identify as proprietary any material
reviewed by the inspectors during this inspection. During the reporting <
period, frequent discussions were held with the Site Director, Plant Manager
and his assistants concerning inspection findings. At no time during the
inspection was written material provided to the licensee by the inspector.
3. Licensee Action on Previous Enforcement Matters (92702)
(Closed) Violation 328/83-29-01. The licensee's response of February 8,
1984, was reviewed and the indicated corrective actinns were audited. The
licensee took administrative disciplinary action in this case which involved
an operator who failed to follow established system alignment procedures.
System Operating Instruction S01-14.3, Condensate Demineralizer Waste
Disposal, was amended to include independent verification. Plant personnel
have also received additional training in. procedural compliance. The
licensee's corrective actions are considered complete.
t
.-. .. .
-
.
2
(Closed) Violation 328/83-29-03. The licensee's response of February 8,
1984, was reviewed and the indicated corrective actions were audited. The
licensee took administrative disciplinary action in this case which involved
an operator who failed to follow established system operating procedures.
Also, plant personnel have received additional training in procedural
compliance. The licensee's corrective actions are considered complete.
(Closed) Violation 327, 328/84-25-02. The licensee response of November 23,
1984, was reviewed and the indicated corrective action was audited. The
licensee submitted a revision to LER 84055 describing the events which led
to the inadequate LER and providing additional information on the breaches
of the ABSCE. LER 84055, Revision 1, was closed in Inspection Report 327,
328/85-16. In addition, the inspector noted an improvement in the quality
of LER submittals. The licensee's corrective actions are considered
complete.
J
(Closed) Violation 327, 328/84-29-01. The licensee's response of
January 18, 1984, was reviewed and the indicated corrective action was
audited. The licensee revised Surveillance Instruction SI-256.2, Inspection
of Molded Case and Lower Voltage Circuit Breakers, to provide alternate
power to breaker 213 en the 125 volt Vital Battery Board III when the
4
breaker is being tested and to include a signcff to ensure that the alter-
nate power is provided to the fuse column.
(Closed) Violation 327, 328/84-35-01. The licensee's response of
January 21, 1985, was reviewed and the indicated corrective action was
audited. The licensee reviewed System Operating Instruction SOI-63.1 and
verified that the procedure had been revised to include the two missing
level transmitter root valves for the cold leg accumulators. The licensee's
corrective actions are considered complete.
4. Unresolved Items
Unresolved items are matters about which more information is required to
determine whether they are acceptable or may involve violations or devia-
tions. One unresolved item identified during this inspection is discussed
in paragraph 8.
5. 0perational Safety Verification (71707)
a. Plant Tours
The inspectors observed control room operations, reviewed applicable
logs, conducted discussions with control room operators, observed shift
turnovers, and confirmed operability of instrumentation. The
inspectors verified the operability of selected emergency systems,
reviewed tagout records, verified compliance with Technical Specifica-
tion (TS) Limiting Conditions for Operations (LCO) and verified return
to service of affected components. The inspector verified that
maintenance work orders had been submitted as required and that
followup activities and prioritization of work was accomplished by the
- - -
-
.
3
licensee. Tours of the diesel generator, auxiliary, turbine buildings
were conducted to observe plant equipment conditions, including
potential fire hazards, fluid leaks, and excessive vibrations and plant
housekeeping / cleanliness conditions.
During the performance of a routine Unit 2 Control Room tour the
inspectors noticed a wedge of paper pressed between two Individual Rod
Position Indication (IRPI) modules on the main control panel 2-M-4.
This same wedge of paper was' still in the panel three days later on a
succeeding inspection tour. When the assigned Reactor Operator was
questioned about the foreign object, he stated that the paper was to
keep the module from vibrating. Vibrations cause the rear contacts on
the module to become loose, resulting in a loss of indication and an
inoperable module. No Maintenance Request (MR) had been written to
repair the module, and there was no Temporary Alteration Control Form
(TACF) prescribing the use of the paper wedge. Administrative
Instruction 12 requires that personnel formally identify deficiencies
and that such deficiencies be promptly identified and corrected.
Failure to follow procedures .foi ' entification and correction of the
vibration deficiency affecting o,mrability of the subject IRPI is a
violation (328/85-26-09).
The inspectors walked down accessible portions of the following
safety-related systems on Unit I and Unit 2 to verify operability and
proper valve alignment:
Safety Injection System (U$its 1 and 2)
- Turbine Driven Auxiliary Feedwater System (Units 1 and 2)
Upper Head Injection System (Units 1 and 2)
Diesel Generators (Units 1 and 2)
Emergency Gas Treatment System (Units 1 and 2)
On July 17, 1985, during a walkdown of diesel . generator 1A1, the
inspectors observed an air leak in pressure reducing valve
0-FCV-82-172A. The licensee wrote MR A-537220 to correct the
deficiency. It was determined that the body of the valve was cracked.
A new valve has been ordered and the valve will be replaced during the
first surveillance (i.e. , the next time the diesel will be _out of
service) after the valve is received. The licensee stated that the DG
was operable due to the redundant air start system for the engine. The
DG was started successfully on August 7, 1985. Repair of this valve is
identified as Inspector Followup Item (327, 328/85-26-01).
b. Verification of Ice Condenser Door Operability
The inspector's review of ice condenser door operability is documented
in Inspector Report 327, 328/85-23. During this inspection period, the
inspector discussed the methods for independent verification of removal
of door blocks from the lower ice condenser doors with the licensee.
The licensee stated that the lower ice condenser doors would be
numbered and that a double visual verification of the removal of the
i
-
.
4
door blocks would be conducted and double signoffs recorded for each
door. This is identified as Inspector Followup Item (327, 328/
85-26-02) until licensee actions are completed.
c. Security
During the course of the inspection, observations relative to protected
and vital area security were made, including access controls, boundary
integrity, search, escort, and badging. Two security concerns were
identified during the inspection period which will be reviewed by NRC
Region II security inspectors for potential violations. These concerns
involved movement of materials into the protected area and the
visibility of the protected area boundary.
d. Radiation Protection
The inspectors observed Health Physics (HP) practices and verified
implementation of radiation protection control. On a regular basis,
radiation work permits (RWPs) were reviewed and specific work
activities were monitored to assure the activities were being conducted
in accordance with applicable RWPs. Selected radiation protection
instruments were verified operable and calibration frequencies were
reviewed.
On July 16, 1985, after observing a functional test on the B train
waste gas compressor, the inspector exited the contaminated zone and
went to the personnel frisking station on EL690. After frisking, the
inspector noted that two individuals who had also exited the
contaminated zone had not arrived at the frisking station. The
inspector found the individuals outside the regulated area. The
inspector interviewed the two individuals and determined that they had
not conducted a whole body frisk when exiting the contaminated zone.
The individuals had used a hand and foot monitor at the exit of the-
regulated zone. The inspector discussed the incident with Health
Physics. The two individuals were recalled, a whole body frisk
performed, and surveys conducted of the areas the individuals had
entered after leaving the contaminated zone. The hand and foot monitor
and whole body and area surveys performed by the licensee after the
incident indicated that the individuals were not contaminated. Radio-
logical Control Instruction RCI-1, Radiological Hygience Control, which
was established to implement the requirements of Technical Specifica-
tion 6.11, states that individuals exiting a contaminated zone are
required to perform a whole body frisk to prevent the spread of
contamination to other areas. RCI-14, Radiation Work Permit (RWP)
Program, states that it is the responsibility of each employee to
adhere to the requirements of RWPs and RWP Timesheets. RWP 02-0-85663,
which was issued to control access to the waste gas compressor room on
July 16, 1985, required employees to perform a whole body frisk upon
exit from the contaminated room. The failure to perform a whole body
frisk is a violation (327, 328/85-26-03).
-
.
5
On August 5, 1985, during a routine tour of the Auxiliary Building, the
inspector observed several yellow poly bags on an Instrument Mainte-
nance Section cart which were labeled only with the sections's name and
telephone number. The material was unattended in a remote location and
not marked with survey data or radiation warning tape. The inspector
requested Health Physics (HP) to survey the bags. The inspector
discussed the observation with the licensee. The licensee stated that
the bags contained instruments which had been used inside a contami-
nated zone during the morning and were to be used in another
contaminated zone during the afternoon. The technicians using the
instruments had left the regulated area for lunch. Although the bags
were unattended and not marked, the licensee felt that the technicians
were cognizant of the location and content of the bags and were
controlling the material properly. The inspector discussed the
observation with the NRC Region II office and determined that the
procedure was act.aptable in cases where work is still in progress, the
articles are contained in yellow poly bags with identification of the
responsible section on the bag, and licensee employees have been
trained to recognize that material in yellow poly bags are potentially
contaminated. Since the procedure which controls the movement of
radioactive material inside the regulated area, RCI-1, was not clear in
how this situation should be handled, the licensee stated that the
procedure would be revised to address these requirements. The licensee
was cited previously in Inspection Report 327, 328/85-20 for not
labeling a yellow poly bag containing contaminated material which had
been left unattended in the Auxiliary Building on May 22, 1985.
Corrective actions for this previous violation are still under review.
6. Engineered Safety Features Walkdown (71710)
The inspector verified operability of the centrifugal charging pump flowpath
through the boron injection tank on Units 1 and 2 by performing a complete
walkdown of the accessible portions of the systems. The following specifics
were reviewed and/or observed as appropriate:
a. that the licensee's system lineup procedures matched plant drawings and
the as-built configuration;
b. that equipment conditions were satisfactory and items that might
degrade performance were identified and evaluated (e.g. , hangers and
supports were operable, housekeeping, etc., was adequate);
c. with assistance from licensee personnel, the interior of the breakers
and electrical or instrumentation cabinets were inspected for debris,
loose material, jumpers, evidence of rodents, etc.;
d. that instrumentation was properly valved in and functioning and cali-
bration dates were appropriate;
e. that valves were in proper position, breaker alignment was correct,
power was available, and valves were locked as required; and
- - . -__ ,
-
.
6
f. local and remote instrumentation was compared and remote instrumenta-
tion was functional.
No violations or deviations were identified.
7. Monthly Surveillance Observation (61726)
The inspectors observed Technical Specification (TS) required surveillance
testing and verified that testing was performed in accordance with adequate
procedures; that test instrumentation was calibrated; that Limiting
Conditions for Operation were met; that test results met acceptance criteria
requirements and were reviewed by personnel other that the individual direc-
ting the test; that deficiencies identified, as appropriate, and that any
deficiencies identified during the testing were properly reviewed and
resolved by management personnel; and that system restoration was adequate.
For complete tests, the inspector verified that testing frequencies were met
and tests were performed by qualified individuals.
The inspector witnessed / reviewed portions of the following surveillance test
activities:
SI-7, Electrical Power System: Diesel Generators
SI-64, Boric Acid Flow Paths - Valve Position Verification
SI-281, Functional Tests of Radiation Effluent Monitors with Automatic
Actuations (Quarterly)
ST -'5, Channel Calibrations for Radiation Monitoring Systems
No violatict.s or deviations were identified in this area.
8. Monthly Maintenance Observations (62703)
a. Station maintenance activities of safety-related systems and components
were observed / reviewed to ascertain that they were conducted in
accordance with approved procedures, regulatory guides, industry codes
and standards, and in conformance with TS.
The following items were considered during this review: LCOs were met
while components or systems were removed from service; redundant
components were operable; approvals were obtained prior to initiating
the work; activities were accomplished using approved procedures and
were inspected as applicable; procedures used were adequate to control
the activity; troubleshooting activities were controlled and the repair
record accurately reflected what took place; functional testing and/or
calibrations were performed prior to returning components or systems to
service; quality control records were maintained; activities were
accomplished by qualified personnel; parts and materials used were
properly certified; radiological controls were implemented; QC hold
points were established where required and were observed; fire
-
.
7
prevention controls were implemented; outside contractor force
activities were controlled in accordance with the approved Quality
Assurance (QA) program; and housekeeping was actively pursued.
b. Corrective maintenance was reviewed on containment isolation valve
2-67-580D, Essential Raw Cooling Water supply to upper containment
cooler 20. The following documents were reviewed:
Maintenance Requests: A-291918 A-024413
A-533609 A-112021
A-128336 A-151118
A-024411 A-024414
A-024415
Maintenance Instructions: 6.20, Configuration Control During
Maintenance Activities
11.4, Maintenance of CSSC Valves
6.21, Repairs and <: placements of
ASME Section XI Components
6.15, General Procedure, Tightening
bolts
Temporary Alteration Control Forms: 2-85-2012-67, 1-85-5007-67
Surveillance Instruction: SI-158.1, Containment Isolation
Valve Leak Rate Test
Isolation check valves 580A through 0_ are scheduled for replacement
during the next outage period. These valves normally allow ERCW flow
through the upper compartment air cooler and become crud traps as a
result of the low flow rates and the raw condition of the ERCW. Review
of MR-A151118 indicated that a waxy substance had been found during
maintenance on the disc of valve 2-67-580D. After interviews with
those technicians that performed the disc replacement maintenance on
valve 2-67-5800, it appeared that no foreigr. material was placed into
the valve. Several tests were conducted in order to get the valve to
seat correctly and to pass the leak test. This brings the repro-
ducibility of the test into question. In addition, post modification
test data was not taken and treated as a Quality Record for those tests
that failed leak testing. The licensee is committed to ANSI standard
18.7-1976, which requires that' post maintenance test failures be
treated as quality records. Such records provide maintenance history
data for trending and planning purposes. Tests that are intended only
as troubleshooting diagnostic devices and on which operability of the
valve would not be based appear not to require retention. Because of
the iterative nature of lapping valve internals, the status of the
failed tests is unclear as a quality record. The valves were turned
over to a separate group (Engineering Test) for test evaluation four
times. During interviews with the technicians that performed the
. - - . - , - - - - - -
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.
,
8
maintenance on 2-67-580D, the inspector also learned that its internal
flapper arm was out of round, but not replaced due to lack of parts.
This valve is scheduled to be retested in October 1985 and the results
will be evaluated by the inspector. This is an Unresolved Item (327,
328/85-26-04) until the inspector further evaluates licensee mainte-
nance controls and retention of test records associated with this and
other maintenance.
c. On July 24, 1985, the inspector observed a post maintenance functional
test on the B train waste gas compressor. The following documents were
reviewed:
Maintenance Request (MR) A526712
Hold Order 993
Maintenance Instruction MI-8.20
The compressor was identified on the maintenance request as CSSC
. equipment. The licensee stated that the equipment was QA controlled
but was not safety-related. Maintenance Instruction MI-8.20 provided
detailed steps on performance of the maintenance; however, the
procedure only stated that a functional test was to be performed after
the maintenance. The procedure did not provide any criteria for the
test. This concern was discussed with the licensee. The licensee
stated that functional test criteria would be included in tests for
CSSC equipment that is not safety-related. The licensee is also
examining assignment of responsibility for functional tests to assure
appropriate test control. This is identified as Inspector Followup
Item (327, 328/85-26-05). During the system lineup, the inspector
observed an Auxiliary Unit Operator (AU0) misalign several root valves.
The AU0 later corrected the misaligned valves prior to completion of
the test. The licensee stated that the AVO has received additional
training in the requirements for system alignment. The licensee is
evaluating the need for control of root valves on System Operating
Instructions. The licensee presently controls the position of root
valves by Hold Orders and with Instrument Maintenance procedures. This
is identified as Inspector Followup Item (327, 328/85-26-06).
9. Licensee Event Report (LER) Followup (92700)
The following LERs were reviewed and closed. The inspector verified that:
reporting requirements had been met; causes had been identified; corrective-
actions appeared appropriate; generic applicability had been considered; the
LER forms were completed; the licensee had reviewed the event; no unreviewed
safety questions were involved; and violations of regulations or Technical
Specification conditions had been identified.
a. LERs Unit 1
'
327/83012 Feedwater Flow Channel Declared Inoperable
, 327/83019 Rod Position Indicator Declared Inoperable
.. __ _ _ _ _ . _ _ __
f
-
.
1
l
9
327/83021 Failure of Heat Tracing on BIT
'
327/83025 Inoperable Flow Rate Monitor for Shield Building
Exhaust
327/83026 Failure of ERCW Valve
>
- 327/83056 Inoperable Upper Containment Personnel Inner Airlock
- Door
.
t
327/83065 Inoperable Train of EGTS
327/83070 Diesel Generator Declared Inoperable
i
327/83080 Inoperable Rod Position Indicator
327/83110 Failure of EGTS Cooldown Valves
~
'
327/83137 Diesel Generator 28-B Inoperable
. 327/83180 Thermal Overload Devices Failed to Trip Check
327/84013 Reactor Trip on Low-Low Steam Generator Level
! 327/84018 Fire Protection Deluge Valve Isolated
327/85008 Failure to Meet Fire Watch Requirements
i 327/85010 Containment Ventilation Isolation
327/85011 Failure to Meet Fire Watch Requirements
,
327/85012 Failure to Meet Fire Watch Requirements
t
i 327/85013 Failure to Meet Fire Watch Requirements
' -
327/85015 Failure to Meet Fire Watch Requirements
327/85017 Inadvertent Auxiliary Building Isolations
,
327/85026 Inadvertent Feedwater Isolation
b. LERs Unit 2
4
328/83004 RWST Low Boron Concentration
- and Rev. 1
328/83051 BIT Valve Inoperable
i 328/83059 Rod Position Indication Inoperable
i
i
!-
-. , . ,_ . - - . _ __ _ _ _ - --- - _ _ _ _ _ _ - _ _ - . _ ~ . _ - . . . . _ . _._,
- - . . . . . .. .
.
-
10
328/83068 RWST Low Water Level
328/83081' Upper Containment Airlock Door Inoperable
J
328/83082 RHR Remote Shutdown Channel Inoperable
4
328/83093 Inoperable Flow Rate Monitor
328/83121 Failure of Lower Containment Airlock to Meet Leakage
Criteria
328/83124 Inoperable Con' denser Vacuum Exhaust Flow Rate Monitor
328/83131 Inoperable Condenser Vacuum Exhaust Flow Rate Monitor
328/83153 Containment Airlock Door Not Fully Closed
328/83161 Rod Bottom Light Bistable Inoperable
l 328/83163 Inoperable Rod Position Indicator
328/84003 Inadvertent Containment Building Ventilation Isolations
328/85005 Debris Inside Containment
c. Followup on LER 328/85001 - Unit 2 Reactor Trip on January 14, 1985
,
Surveillance Instruction (SI-80), Power Range Neutron Flux Channel
Calibration and Functional Test, was conducted on Unit 2 Nuclear
Instrument Power Range Channel N-41. SI-80 referenced Instrument
.
'
Maintenance Instruction (IMI-92-PRM-CAL), NIS Power Range, as the
required calibration procedure. Step 5.2.1.6 of IMI-92-PRM-CAL
required that the instrument power fuses be removed and independently
verified. When the instrument power fuses are removed from a Power
Range (PR) Nuclear Instrument (NI) drawer, the high voltage supply to
that drawer is removed and a negative rate flux trip bistable is
activated. The negative rate flux trip bistable does not automatically
reset when the instrument power fuses are replaced and must be manually
'
reset. If two separate NI PR negative rate flux trip bistables are
activated at the same time, a reactor trip results.
A Senior Instrument Maintenance (IM) technician removed the instrument
power fuses from PR NI N-42 in error, replaced the fuses, but did not
reset the reset the channel. When the IM technician subsequently
deenergized PR NI drawer NI-41, by removing the instrument power fuses,
,
the corresponding negative rate was activated. This satisfied a
] two-out-of-four reactor protection system logic, resulting in a Unit 2
i reactor trip from 30's power. The removal of the instrument power fuses
, was determined to be a personnel error by the licensee. The inspectors
- verified this through interviews with personnel involved and a review
i of the licensee's IM training and qualification system. However,
. __ _. __
_ __ __ _ _ _,
_
. -
11
sev-:11 actions taken by the technician after removal of the fuses are
examples of a failure to follow procedure. These actions are explained
- below.
Following the removal of the PR NI drawer N-42 instrument power fuses,
the IM technician failed to follow procedures in two instances. First,
he attempted to place PR NI N-42 back in service by replacing its
instrument power fuses. The appropriate section of IMI-92-PRM-CAL
(Section 7) was not implemented to ensure that N-42 was correctly
returned to service. Second, after the IM technician had returned PR
NI drawer N-42 to what he assumed to be an operable condition, he
commenced surveillance activities on PR NI N-41 without reporting to
the RO or SRO that he had not complied with an assigned procedure and
had affected safety-related equipment not authorized for su,rveillance
activities.
Administrative Instruction AI-12, Adverse Conditions and Corrective
Actions, states that plant personnel shall report any suspected
abnormal plant condition adverse to quality in the performance of their
regular work duties. A failure to follow procedure is defined in
section 4.0 of AI-12 as a condition adverse to quality. The inspector
is not satisfied that licensee personnel clearly understand their -
i responsibility to report to the shift operating staff any unintended
affects on safety-related equipment during planned activities. This
issue will be reviewed as Inspector Followup Item (327, 328/85-26-07).
The LER is closed.
10. Event Followup (93702, 62703, 61726)
a. Unit 1 Reactor Trip Due to Loss of Main Feedwater Pump Turbine (MFPT)
Oil Pump
On July 19, 1985, Unit I tripped from 100 percent power on low-low
steam generator level. The licensee was utilizing normal procedures to
search for a ground on a turbine unit board. In this process, 480V
Unit Board 1B was switched from the normal power supply to an alternate
supply. The operator closed the alternate supply breaker to the board,
but the breaker did not close the breaker completely. He then
deenergized the normal supply. He immediately noted that the MFPT oil
pump light on the board went out and jerked up on the alternate supply
breaker completely closing it; however, the oil pump tripped.
A backup oil pump started automatically on the trip signal; however,
the momentary drop in oil pressure caused partial completion of. logic
associated with the MFPT oil pressure switch. This logic includes
closing of the train associated MFPT stop valves, a main turbine
runback, and start of the motor driven (MDAFWP) and turbine driven
(TDAFWP) auxiliary feedwater pumps among others. The logic was
completed for the closing of the stop valves resulting in coasting down
of the MFPT; however, the turbine runback logic was not completed and
steam generator levels started to decrease. The MDAFWPs started, and
-. . .-. . - - .
. _ _ _ .. . _ _ . .
.
12 -
i
the TDAFWP did not. The operators immediately put the control rods in
auto and attempted to manually runback the main turbine. This was
unsuccessful, and the reactor tripped on low-low stcam generator level.
All reactor protection system functions reacted normally on the reactor
trip including start of the TDAFWP.
The logic associated with the MFPT pressure switch is not a reactor
protection or engineered safety feature signal. The purpose of the
switch is to provide anticipatory features, such as turbine runback,
for events which could lead to low-low steam generator reactor trips. -
The inspector reviewed the logic with a Senior Reactor Operator and
discussed the failure of the logic circuit to completely energize with
the licensee. This problem was attributed to the fast operation of the
pressure switch resulting in only a momentary contact in the logic
circuit. The mechanical contacts eif.her did not have time to makeup or
the seal-in circuit was not energized.
!
.The licensee attempted to duplicate the initiating conditions through
.
testing to ascertain if any problems existed within the logic
1
circuitry; however, the condition could not be duplicated. This
testing and performance of Surveillance Instruction SI-110.1, TDAFWP
and Valve Automatic Actuation, resulted in completion of the full
logic. The licensee adjusted the pressure switch which starts the
backup oil pump to a higher setting to preclude a significant drop in
oil pressure on trip of one of the oil pumps. This should prevent a
similar event.
No violations or deviations were identified.
b. Unit 2 Notification of Unusual Event Due to Excessive Reactor Coolant
System (RCS) Leakage
On July 29, 1985, operators noted a decrease in volume control tank
level. Subsequently, a high radiation alarm on the Auxiliary Building
Vent particulate monitor caused an Auxiliary Building isolation. The
operators followed the requirements of Abnormal Operating Instruction
AOI-31, Abnormal Release of Radioactive Materials, and AOI-6, Small
Reactor Coolant System Leak. The Auxiliary Building was evacuated and
i posted as an airborne radiation area. The operators placed excess
letdown in service and secured the normal charging and letdown flow-
paths.
. A few minutes later, a fire watch monitoring temperatures in various
pipe chases reported a high temperature in the Unit 2 EL690 pipe chase.
The Unit 2 Assistant Shift Engineer (ASE) responded to the area and
encountered two electricians exiting the area. The electricians
reported that there was a steam leak in the pipe chase. (The
electricians were subsequently decontaminated as discussed below.) The
ASE investigated the report and determined that weld upstream of sample
valve 62-674 was leaking. The weld was located at the point where the
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.
. - - - - . - -. - . . ,
. _ -
. *
13
one inch sample line to the hot sample room joined the three inch
normal CVCS letdown line.
The leak rate was determined to be about 15 gpm. The licensee
estimated that approximately 600 gallons of RCS water was released
during the event. In addition, due to backleakage across a check valve
between the Volume Control Tank and the break, a small amount of
hydrogen was released from the VCT to the pipe chase. This leak was
promptly isolated. The licensee repaired the line and returned the
normal charging and letdown flowpaths to service on August 1, 1985.
, The licensee removed the segment of the sample line affected and sent
the weld to a metallurgical lab for analysis. It was determined that
the break which was in the heat affected zone above the weld was due to
- fatigue from high cycle vibration. Two initiation points approximately
'
180 degrees apart were observed. The licensee stated that these points
were due to high cycle, low stress fatigue. The break propagated about
300 degrees. The licensee has evaluated the configuration and has
installed a hanger on the line. The installation will be evaluated to
assure that the vibration is corrected. The weld on the Unit I sample
line was visually examined, and no defects were identified. An
appropriate hanger configuration will be installed on Unit I when the
Unit 2 analysis is complete. Completion of the harger work will be
- followed by the inspector and is identified as Inspector Followup Item
(327,328/85-26-10).
Eleven personnel suffered skin contamination as a result of tne primary
leak described above. The most severe case was 11,000 dpm on one
person's face. All personnel were decontaminated using normal methods
(showers) and had no detectable contaminatica after they were
decontaminated.
On July 30, 1985, the licensee posted the control building and an area
adjacent to the entrance to the Auxiliary Building on EL690 as airborne
areas. The air samples from these areas had been analyzed in the plant
chemistry lab. The readings were later determined to be erroneous due
to the high background in the lab by sending samples to the Power
Operations Training Center laboratory for analysis. These samples
indicated no airborne radiation was present.
While placing the Auxiliary Building general supply fans back in
service, three Auxiliary Building isolations occurred due to spiking on
0-RM-90-1018. The licensee stated that the setpoint on this monitor
was reset to a higher value which was still within the range required
to meet Technical Specification 3.11.2.1. Verification of the setpoint
change by the inspector is identified as Inspector Followup Item 327,
328/85-26-08. The Auxiliary Building Vent System was returned to
service with the Auxiliary Building Gas Treatment System operating to
reduce airborne levels in the Auxiliary Building.
No violations or deviations were identified.
i
.-. - _ - - _ - - , - -
_
l
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14
11. Review of Licensee Actions on NRC Information Notice 84-31 (92703)
The inspectors examined the licensee's handling of. NRC Information Notice
-
84-31, Increased Stroking Time of Bettis Actuators Because of Swollen
Ethylene - Proplyene Rubber Seals and Seal Set, dated April 30, 1984. The
information notice described the problem as follows:
The G. H. Bettis Company is a supplier of actuators used principally in
heating, ventilating and air-conditioning (HVAC) safety-related systems.
The G. H. Bettis Company notified the. NRC via a Part 21 report that their
NCB series, N52X, N72X, N73 series, and the NT310-SR4 and 5 and NT312-SR5
- actuators had potential stroking times of greater than the required 15
seconds because the EP elastomers in contact with the Mobil 28 grease
l lubricant could swell. (The 15-second stroking time was used by Bettis as
typical of customer requirements.) The actuator seals swell when in contact
with the Mobile 28 grease currently used in the manufacture of "N" series
actuators. Where it is necessary to replace swollen seals, Bettis
,
'
recommended replacing them with new seals and using Dow-Corning Molykote 44
grease, a silicon based lubricant which Bettis states has been shown to
7
cause no seal degradation and adequate lubrication.
.
The G. H. Bettis Company also identified another problem that could ,
adversely affect stroking time. Their report states that ". . .the magnitude ;
of stroking time degradation is related to the elapsed time between actuator
i
cycles. The longer the actuator remains stationary the more " set" the seals
, take. The set characteristic causes the seal to form an intimate contact
1
with the sealing surfaces, further increasing the time required to
initialize stroke. Once the actuator begins to stroke, the seals begin to
recover their original shape, thus freeing the unit up. Stroking the
actuator three or more complete cycles using pressurized gas will cause the
seals to recover sufficiently to reduce stroking time to a minimum. No seal
degradation has been traced to periodic actuator stroking, quite the
'
opposite has been experienced. Frequent stroking tends to extend seal life ,
resulting in longer actuator cycle life."
The inspectors reviewed a memorandum from H. A. Abercrombie, Director of
.
Nuclear Services, to Sequoyah and other sites, dated August 22,. 1984
! indicating that all safety-related valves and dampers of the model series
! described in the notice should have the seals and lubricant replaced at the
j first available outage. The memorandum requested that a search be made of
site documentation in order to identify the valves and dampers involved. .t
! Twenty seven (27) valves and dampers were identified at Sequoyah and three
l Nonconformance Reports (NCRs) were . written. The NCRs with pertinent
, information are as follows:
a. NCR EEB 84-12 dated December 1984 involved valves 1-FCV-77-420 and
2-FCV-77-421. Both of these valves were still in the warehouse and
have been tagged as not to be used.
l b. NCR MEB 84-07 dated September 1984 involved 17 valves used in three
j separate areas.
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. , . _ - .
. -
15
1. Three valves are used to actuate the vacuum breaker system for
containment and come under ASME Section XI rules for testing.
These valves have a maximum allowed stroking time of 25 seconds
and must be tested quarterly. The latest test (June 26, 1985)
indicated stroking times for the three valves ranged from 6.4 to
6.8 seconds. The three valve numbers are 2-FCV-30-46,
2-FCV-30-41, and 2-FCV-30-40,
2. Ten valves are used to actuate the Emergency Gas Treatment System.
Each train is tested every other month but each of the valves is
not tested individually to determine stroke time. The fastest the
system has to actuate is 38 seconds and no problems have been
roted during the testing.
3. Four valves are Control Room HVAC isolation valves. This system
is required to be tested every 18 months and was last tested
August 17, 1984 with no problems. The four valve numbers are
0-FCV-31A-105A, 0-FCV-31A-105B, 0-FCV-31A-106A and 0-FCV-31A-106B.
The failure evaluation / engineering report (FE/ER) classified the
deficient condition to be a Category 1 (acceptable for all modes of
operation and design condition) based upon a sampling of the valves for
stroking time,
c. NCR MEB 84-08, Rev. 1, dated October 1984 involved 8 valves installed
in the fifth vital battery room for protection from tornado depressuri-
zation. The eight valve numbers- are FCO-51-485, FCO-31-486,
FC0-31-488, FCO-31-489, FCO-31-493, FCO-31-494 and FC0-31-501. The
FE/ER classified the deficient condition to be a Category I based on
the fact that the valves were not installed in the system, and the
- seals and lubricant could be changed before the valves would experience
operational use. This was not done and this violates 10 CFR 50
LAppendix B Criterion XVI in that effective corrective action dictated
by the FE/ER was not taken. However, no violation will be issued since
programmatic corrective actions are currently in progress at TVA due to
,
an Order Modifying Licenses issued June 14, 1985 (EA 85-49).
- Correction of this specific deficiency is an -Inspector Followup Item
(327,328/85-26-09).
!
,
v +, - - ,- , -w ,,-w