IR 05000259/1985049
| ML18030A841 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry, 05002598  |
| Issue date: | 11/06/1985 |
| From: | Brooks C, Cantrell F, Patterson C, Paulk G NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML18030A840 | List: |
| References | |
| 50-259-85-49, 50-260-85-49, 50-296-85-49, NUDOCS 8511130198 | |
| Download: ML18030A841 (28) | |
Text
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UNITED STATES NUCLEAR REGULATORY COMMISSION
REGION II
101 MARIETTASTREET, N.W.
ATLANTA,GEORGIA 30323 Report Nos.
50-259/85-49, 50-260/85-49, and 50-296/85-49 Licensee:
Tennessee Valley Authority 500A Chestnut Street Tower II Chattanooga, Tennessee 37401 Docket Nos.
50-259, 50-260, and 50-296 License Nos.
DPR-33, DPR-52, and DPR-68 Facility Name:
Browns Ferry Nuclear Plant Inspection Conducted:
October 1-25, 1985 Inspectors:
G.
L.
P lk, Senior R
id t C. A.
Pa erson, Resi t
C.
R. Br ks, Reside Approved by:
F.
S; Cantrell, Secti hief Division of Reactor rojects
//ba D
e igned
//
Da e
igned
// /
Da e
igned'/8 F3 D te Signed SUMMARY Scope:
This routine inspection involved 240 resident inspector-hours in the areas of operational safety, maintenance observation, reportable occurrences, surveillance, regulatory performance, modifications, and Nuclear Safety Review Board.
Results:
No violations or deviations were identified.
85f ffQoi98 Sff07 PDR ADOCK 0 pDR
REPORT DETAILS Licensee Employees Contacted:
T. F. Ziegler, Acting Site Director R.L. Lewis, Plant Manager (Acting)
J.E.
Swindell, Superintendent
- Operations/Engineering TED. Cosby, Superintendent Maintenance (Acting)
J.H.
Rinne, Modifications Manager J.D. Carlson, Quality Engineering Supervisor D.C. Mims, Engineering Group Supervisor R.M. McKeon, Operations Group Supervisor C.G.
Wages, Mechanical Maintenance Supervisor J.C.
Crowell, Electrical Maintenance Supervisor (Acting)
R.E.
Burns, Instrument Maintenance Supervisor A:W. Sorrell, Health Physics Supervisor R.E. Jackson, Chief Public Safety T.L. Chinn, Senior Shift Manager J.R. Clark, Chemical Unit Supervisor B.C. Morris, Plant Compliance Supervisor A.L. Burnette, Assistant Operations Group Supervisor R.R.
Smallwood, Assistant Operations Group Supervisor S.R.
Maehr, Planning/Scheduling Super'visor G.R. Hall, Design Services Manager W.C. Thomison, Engineering Section Supervisor C.E. Burke, Radwaste Group Controller Other licensee employees contacted included licensed reactor operators, auxiliary operators, craftsmen, technicians, public safety officers, quality assurance, design and engineering personnel.
Exit Interview (30703)
The inspection scope and findings were summarized on October 24, 1985, with the Plant Manager and/or Assistant Plant Managers and other members of his staff.
The licensee acknowledged the findings and took no exceptions.
The licensee did not identify as proprietary any of the materials provided to or reviewed by the inspectors during this inspection.
T.
F. Ziegler assumed duties as Acting Site Director, relieving J.
A. Coffe Licensee Action on Previous Enforcement Matters (92702)
(Closed)
Violation (259/260/296/84-29-01)
The licensee response and corrective action to this violation were reviewed and the inspector has no further questions.
This item is closed.
(Closed)
Inspector Follow-Up (259/84-26-19)
All trip reports are routinely reviewed.
This item is closed.
(Closed)
Violation (259/260/296/84-38-03)
Procedures have been checked since the initial problems and no additional problems noted.
This item is closed.
(Closed)
Violation (259/260/296/84=38-01)
The inspector reviewed the corrected copies of the test analysis reports.
Purchase specifications are now reviewed before a delivery truck is allowed entry into the protected area.
This item is closed.
(Closed)
Unresolved Item (259/260/296/85-25-04)
A violation (85-28-.05)
was issued for the battery racks not being mounted as required by drawings.
This item is closed.
(Closed)
Open Item (259/84-48-02)
The procedure MMI-87 now specifies the type grease to be used for each operator.
This item is closed.
(Closed)
Unresolved Item (259/84-07-03)
The licensee has implemented Standard Practice BF-2.'. 1 to control computer software changes.
This item is closed.
(Closed) Violation (259/260/296/84-10-01)
The response, revised response, and two supplemental information letters concerning this violation were reviewed.
This item is closed.
(Closed) Violation (259/84-15-03)
Standard Practice BF-14.25 was reviewed and second-party verification requirements for clearance tags have been implemented.
This item is closed.
(Closed)
Violation (259/260/296/84-20-02)
No notice of violation was issued for this in the civil penalty.
This item is closed.
(Closed) Violation (259/260/84-26-02)
The licensee's corrective action was reviewed and the inspector has no further question.
This item is closed.
(Closed)
Violation (259/260/296/84-26-03)
Usage of Form 126 has been reviewed during routine tours of the control room and found acceptable.
This item is closed.
(Closed) Violation (259/260/84-26-05)
The Control Room Emergency. Ventila-tion damper linkage has been modified with a threaded rod and lock nuts.
This item is closed'
(Closed)
Violation (260/84-38-04)
The response to thi s violation was reviewed and the inspector has no further questions'his item is closed.
(Closed) Violation (260/84-35-05)
The corrective action for this violation has been reviewed in past inspections and this item is closed.
(Closed)
Violation (259/260/296/84-02-01/02/03/04)
The licensee response and corrective action to this violation and other 84-02 report violations concerning rod control were reviewed and the inspector has no further questions.
This item is closed.
(Closed)
Deviation (259/84-26-06)
The licensee has developed a
safety i.ssues list to track LER dates.
This item is closed.
(Closed)
Inspector Follow-Up (259/84-26-07)
The licensee has initiated a
general valve maintenance program and is working toward INPO accreditation.
This item is closed.
(Closed)
Inspector Follow-Up (259/84-26-08)
The inspector toured the pump room and the B fire pump pressure gage has been securely fastened to the panel.
This item is closed.
(Closed)
Inspector Follow-Up (259/84-26-10)
This resulted.
in violation 84-33-01 'his item is closed.
(Closed)
Violation (259/260/296/84-33-01)
The licensee now performs a
quarterly inspection of locked valves.
This item is closed.
(Closed)
Inspector Follow-Up (259/84-26-15)
This problem has not been observed in recent routine tours of the facility.
This item is closed.
(Closed)
Open Item (259/84-33-02)
This item will be tracked in the future under 85-15-11 which discusses recent LER problems.
This item is closed.
(Closed)
Open Item (259/83-33-04)
Mechanical Results Instruction
was reviewed and the inspector has no more questions.
This item is closed.
(Open)
Unresolved Item (259/260/296/85-45-09)
The wiring discrepancies in reactor protection system (RPS) trip panels (panels 9-15 and 9-17) are still indeterminate at this point.
The licensee has traced several errors to the performance o$- engineering change notice (ECN)
P0126 which was issued for installation of the analog transmitter trip system (ATTS).
Many work plans are in an unknown status of completion and the responsible group of personnel originally involved in the installation of the ECN has been disbanded.
An effort is underway to reassemble the work team to field verify the present configuration.
The damaged wire from the RPS MG set breaker to the hot bus of panel 9-17 has been replaced with a larger capacity cable.
The No.
AWG cable was apparently undersized and was
replaced by a No.
AWG wire.
This discrepancy is the subject of 10 CFR 21 reportability determination being performed by both the plant and manufac-turer (General Electric).
This unresolved item will remain open pending further investigation by the licensee.
4.
Unresolved Items" (92701)
New unresolved items are described in paragraph 5, 8, and 11.
5.
Operational Safety (71707, 71710)
The inspectors were kept informed on a daily basis of the overall plant status and any significant safety matters related to plant operations.
Daily discussions were held each marning with plant management and various members of the plant operating staff.
The inspectors made frequent visits to the control rooms such that each was visited at least daily when an inspector was on site.
Observations included instrument readings, setpoints and recordings; status of operating systems; status and alignments of emergency standby systems; onsite and offsite emergency power sources available for automatic operation; purpose of temporary tags on equipment controls and switches; annunciator alarm status; adherence to procedures; adherence to limiting conditions for operations; nuclear instruments operable; temporary alterations in effect; daily journals and logs; stack monitor recorder traces; and control room manning'his inspection activity also included numerous informal discus-sions witn operators and their supervisors.
General plant tours were conducted on at least a weekly basis.
Portions of the turbine building, each reactor building and outside areas were visited.
Observations included 'valve positions and system alignment; snubber and hanger conditions; containment isolation alignments; instrument readings; housekeeping; proper power supply and breaker; alignments; radiation area controls; tag controls on equipment; work activities in progress; radiation protection controls adequate; vital area controls; personnel search and escort; and vehicle search and escort.
Informal discussions were held with selected plant personnel in their functional areas during these tours.
Weekly verifications of system status which included major flow path valve alignment, instrument alignment, and switch position alignments were performed on the standby gas treatment and emergency diesel systems.
"An Unresolved Item is a matter about which more information is required to determine whether it is acceptable or may involve a violation or deviatio A complete walkdown of the accessible portions of the reactor protection system was conducted to verify system operability.
Typical of the items checked during the walkdown were:
lineup procedures match plant drawings and the as-built configuration, hangers and supports operable, housekeeping adequate, electrical panel interior conditions, calibration dates appro-priate, system instrumentation on-line, valve position alignment correct, valves locked as appropriate and system indicators functioning properly.
Technical Specification Interpretations In order to promote uniform and consistent interpretation of Technical Specifications, tPe licensee established in June 1985, a
program for developing and documenting Technical Specification interpretations.
The program as outlined in Standard Practice 12.23, Technical Specifi-cation Interpretations, involves Technical Specification interpretation request (TSIR) forms which may be initiated by anyone in the plant when a situation arises which might warrant clarification.
The TSIR is acted on by a Technical Specification Interpretation Committee (TSIC)
composed of engineering and operations personnel and Chaired by the compliance supervisor.
Once the interpretation is developed, it is sent to the Plant Operations Review Committee (PORC)
and the Plant Manager, for review and approval.
Approved interpretations are main-tained in controlled manuals which are distributed to all holders of Technical Specifications.
The co'mplexities, inconsistencies and errors in the Browns Ferry custom Technical Specifications have been a matter of concern for some time and several Regulatory Performance Improvement Program (RPIP) action items were developed to address these concerns.
The new interpretation program is being flooded with initial requests containing statements to the effect that the subject Technical Specifi-cation is worded poorly and has been the cause of much confusion in the past, and the wording of the Technical Specification invites errors in
.application.
Some of the significant interpretations presently being pursued are listed below:
( 1)
Technical Specification 1.L defines cold shutdown as "the reactor is in the shutdown mode and the reactor coolant temperature equal to or less than 212 degrees F."
Technical Specification 1.M.3 defines the shutdown mode as
"placing the mode switch to the shutdown position...."
There seems to be no provisions elsewhere in Technical Specification for considering that the plant can be maintained in the cold shutdown condition while the mode switch is cycled to the refuel or startup/hot standby positions for sur-veillancee and maintenance requirements.
This is further aggra-vated by the present plant conditions which require invoking Technical Specification 1.C. 1 which states that if LCOs and/or associated requirements cannot be satisfied because of circum-stances in excess of those addressed in the Technical Specifica-tions, the unit must be placed in cold shutdown within 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
The consequence of the restrictive language in the Technical Specifications is that the mode switch must remain in shutdown yet required survei llances necessitate taking the mode switch to other position (2)
Technical Specification 1.C.2 provides flexibility in the appli-cation of the definition of operability (Technical Specification, 1.E)
when components are inoperable solely because its onsite power source is inoperable.
It further allows operation to be governed by the time limits of the LCO for the diesel power source, not the individual requirements for each component that is determined to be inoperable solely because its diesel power source.
The Technical Specification contains a
statement that
",this is not applicable if the unit is already in cold shutdown or refueling". It is unclear if the term "this is not applicable" is referring to the entire specification or to the sentence immediately preceeding this statement which requires the unit be placed in cold shutdown if certain requirements are not met.
(3)
Technical Specification 3.5.C.7 requires that there shall be at least
RHRSW pumps, associated with the selected RHR pumps, aligned for RHR heat exchanger service for each reactor vessel containing irradiated fuel.
Technical Specification 3.5.B.9 requires that when the reactor vessel pressure is atmospheric and irradiated fuel is in the reactor vessel, at least one RHR loop with two pumps or two RHR loops with one pump per loop shall be operable.
It is not clear whether RHRSW pumps assigned to one unit can also be given credit for meeting 3.5.C.7 for another unit.
(4)
Technical Specification 3.7.H. 1 requires that whenever the reactor is not in cold shutdown, two gas analyzer systems shall be operable for monitoring the drywell and torus.
This restrictive language requires that two analyzers be maintained operable and surveillance requirements satisfied during refueling operations.
The licensee has indicated that Technical Specification Amendments will be submitted to clarify these areas on a permanent basis once the need is identified.
b.
Hissed Surveillance Requirements On October 8, 1985, the licensee reported that due to invalid interpre-tations of Technical Specifications and various errors in surveillance instructions',
several surveillance requirements had not been satisfied for Unite
and 3.
Subsequent investigation into the generic aspects of the event also discovered that similar surveillances were missed during fuel off-loading of Unit 2 in September 1984.
The root cause of the missed survei llances was an over reliance on SI-1, Surveillance Program, by personnel involved in routine scheduling and approval of missed survei 1 lances.
SI-1, Appendix C, contains a list of survei l 1-ance requirements which were allowed to be skipped during refueling outages.
This list contained errors and was generally not in accor-dance with Technical Specifications.
The following table summarizes the missed survei llances:
(>
Surveillance Instruction No.
Descri tion Unit
Dates Missed 4.2.8-3 4.2.B-4 Reactor Low Mater Level Channel Channel Functional Test Orywell High Pressure Channel Functional Test 7/13/85 7/27/85 4.2.B-5 Drywell High Pressure Calibration and Functional Test 7/27/85 4.2.8-7 Reactor Low Pressure Channel Functional Test 7/28/85 4.2.B-8 4.2.B-69 Reactor Low Pressure Channel Functional Test Reactor High Pressure Channel Functional Test 7/14/85 7/07/85 4.3.F.1.b 4.7.H Scram Discharge Volume Vent and Drain Valve Operability Gas Analyzer Channel Calibration Unit 2 7/31/85 8/30/85 9/27/85 7/15/85 8/15/85 9/11/85 4.2.B-3 4.2.B-8 4.7.H 4.2.B-2 Reactor Low Water Level Sensor Calibration and Functional Test Reactor Low Pressure Sensor Calibration and Functional Test Gas Analyzes Channel Calibration Unit 3 Reactor Low Mater Level Functional Test 9/17/84 9/18/84 10/03/84 9/23/85 4.2.B-4 4.2.B-5 Orywell High Pressure Channel Functional Test Drywell High Pressure Channel Functional Test 9/16/85 9/16/85
(cont'd)
4.2.B-6 Drywell High Pressure Calibration and Functional Test 9/16/85 4.2.B-7 Reactor Low Pressure Channel Functional Test 9/17/85 4.2.B-8 Reactor Low Pressure Channel Functional Test 10/06/85 4/2/B-69 Reactor High Pressure Channel Functional Test.
9/24/85 The licensee is still evaluating the causes and corrective action to prevent recurrence.
This item will be unresolved for evaluation and follow-up of the licensee's corrective action.
(259/260/296/85-49-01).
C.
Storage of Fuel Related Components During a routine daily tour on October 18, 1985, the inspector noted that a crate of fuel channels was being stored outside of the reactor building near the equipment access hatch.
The side of the crate was damaged, allowing the top of the crate to fall and rest on the channels contained within.
A tarp was p'artially covering the crate; however','his had been ineffective in protecting the channels from rain and the weather.
Several miscellaneous
.pieces of hardware were resting on top of the caved-in crate.
The crate was marked with Order 083P67-196476-1.
A check of Power Stores records indicated that the last parts issued against this contract were drawn on December 20, 1984 by mechanical maintenance.
Contract documentation contained the General Electric storage requirements.
The channels should have been maintained in an indoor, unheated storage space.
Power Stores personnel stated that these storage requirements were not passed on to the users since users are only authorized to draw equipment for use and are required to return unused portions back to power stores.
The plant superintendent, maintenance, quality assurance supervisor, power stores supervisory personnel and mechanical maintenance personnel were made aware of this condition on October 18, 1985.
As of'ctober 23, 1985, there were still no controls to prevent the inadvertent use of these nonconforming fuel channels even though the channels had been uncrated and moved to the refuel floor.
The Nuclear guality Assurance Manual (NIZAM) Part III, Section 7. 1 requires that if defects, variations, deficiencies, or deviations appear in an item to make it either unacceptable or of questionable quality, the item may not be used or installed until proper review and dispositions are obtained.
Items which are unaccept-able for use shall be identified, documented, and segregated.
The item shall be identified by use of a nonconforming item (NCI) tag (form TVA 7830).
This item will be left unresolved pending results of the licensee inspections and disposition of the nonconformance (259/260/296/85-49-02).
On September 23, 1985, the licensee suspended fuel handling operations on Unit
due to a
concern over the stroke time requirements for Reactor Building Ventilation dampers which are actuated upon secondary containment isolation signals.
Section 5.3.4.2 of the FSAR (Standby Gas Treatment Instrumentation and Control) discusses the time required to switch from the normal containment ventilation system to the Standby Gas Treatment System upon detection of high radiation.
The radiation monitor response time is stated to be
second and the reactor zone isolation dampers are stated to close in
seconds.
The
second radiation monitor response time plus the 2 second response time of the isolation dampers is less than the transport time from the surface of the fuel pool to the isolation damper and thereby assures that the release from a
fuel handling accident will be contained within the secondary containment.
The licensee identified that reactor zone isolation dampers were isolating in approximately 5 seconds as opposed to the required 2 seconds.
These response times are not technical specification requirements and have not been verified in past surveil-lance tests.
Although the licensee did not declare the isolation dampers inoperable or close the dampers, fuel handling operations were suspended and an evaluation was initiated to determine the basis for the FSAR statements.
Secondary containment and the Standby Gas Treatment System (SGTS) were considered fully operable.
The inspectors expressed their concern over this non-conservative approach.
On October 22, 1985, the inspector-Observed movement of 'blade guides into the Unit
spent fuel pool.
Since Technical Specification 3.7.C.4 requires that if refuel zone secondary containment is not maintained, handling of spent fuel and all operations over the spent fuel pool shall be prohibited.
The inspectors once again questioned the status of the secondary containment isolation-dampers.
Repairs to the dampers had not been completed and the engineering evaluation into the 2 second isolation damper response time was not completed.
Licensee representa-tives continued to consider the isolation dampers operable although fuel handling operations were still prohibited until the dampers were repaired or evaluated.
This item will be tracked as an unresolved item (259/260/296/85-49-03)
pending completion of the evaluation regarding the 2 second time requirement.
The additional concern over inadequate technical specifications in the area of secondary containment damper isolation times will also be tracked under this item.
Although standard technical specifications contain an LCO and surveillance requirements on secondary containment isolation times, the Browns Ferry Technical-Specification has no such requirements.
6.
Maintenance Observation (62703, 62700)
.Plant maintenance activities of selected safety-related systems and components were observed/reviewed to ascertain that they were conducted in accordance with requirements.
The following items were considered during this review:
the limiting conditions for operations were met; activities were accomplished using approved procedures; functional testing and/or calibrations were performed prior to returning components or system to
1~)
servi ce;.
qual ity control records were maintained; acti vities were accom-pli shed by qualified per sonnel; parts and material s used were properly certified; proper tagout clearance procedures were adhered to; Technical Specification adherence; and radiological controls were implemented as required.
Maintenance requests were reviewed to determine status of outstanding jobs and to assure that priority was assigned to safety-related equipment maintenance which might affect plant safety.
The inspectors observed the below listed maintenance activities during this report period:
a.
Diesel Generator Battery Rack Repairs b.
Residual Heat Removal Service Cater Pump Pit Cleaning c.
Standby Gas Treatment d.
Electrical Cable Tray Seismic gualification Repairs e.
Fuel Channel Cleaning While following the progress of repairs on the diesel generator battery racks (refer to paragraph 6 of Inspection Report 85-45),
the inspector learned that a personnel error occurred during reinstallation of the battery cells on the Units 1/2 C diesel generator battery.
Three cells were installed with 'reversed polarity during the maintenance performed during the evening shift.
Although the work package contained a cognizant engineer's sign-off step to verify proper polarity of the cells, this step was omitted.
The error was detected during battery charging prior to the post-maintenance capacity test.
The three cells involved were removed and replaced.
One of the cells was found to have a damaged case and was scrapped.
Manufacturers'ecommendations were followed to restore the remaining two cells to a usable status.
A Discrepancy Report was written on the personnel error and disci-plinary action is being contemplated.
Failure to adhere to to procedures is a recurring deficiency which has been difficult to correct although several Regulatory Performance Improvement Program (RPIP) action items address this area.
On October 17,
.1985, licensee representatives discussed their intention to remove residual heat removal service water (RHRSW)
pumps Dl and D2 from service for pN cleaning.
Concurrence with their Technical Specification interpretation on this issue was requested.
Technical Specification 3.5.A.4 requires that when the reactor vessel pressure is atmospheric and irradiated fuel is in the reactor vessel, at least one core spray loop with one operable pump and associated diesel generator shall be operable except with the reactor vessel head removed as specified in 3.5.A.5.
Since the asso-ciated diesel generator was inoperable for satisfying Technical Specifica-tion requirements due to missed survei llances, Technical Specification 3.5.A.4 was not satisfied.
Technical Specification 3.5.A.S states that when irradiated fuel is in the reactor vessel and the reactor vessel head is removed, core spray
.is not required provided one RHRSW pump and associated valves supplying the standby coolant supply are operable and additional requirements are met.
The RHRSW pumps which supply the standby coolant supply to Unit
are the
and
pumps.
Inference from Technical Specification 3.5 AC.4 would indicate that for the standby coolant supply pump to be considered operable, its associated diesel generator must also be operable.
The diesel generators associated with both the Dl and
RHRSW pumps were also inoperable for the same reasons as the diesel for the core spray pump.
Although the diesels were inoperable for satisfying Technical Specification requirements, they were available and would presumably operate if called upon to do so.
The licensee planned to remove the 01 and
RHRSW pumps from service since neither Technical Specification could be rigorously satisfied and the lack of compliance with either 3.5.A.4 or 3.5.A.5 would be essentially equivalent.
The inspectors indicated that it would not be prudent to intentionally remove the redundancy provided by the standby coolant supply without profer justification.
Before guidance and clarification could be obtained from the Regional Office and NRR, the resident learned that the Dl and
RHRSW pumps had been taken out of service for suction pit cleaning and that a one-hour ENS phone report had been made by operations personnel once this had been discovered.
The licensee took immediate action to restore the
RHRSW pump to service but maintenance related problems caused delays until on October 21, 1985, the pump was again declared operable.
This lack of coordination between main-tenance, operations and management has been a continuing problem and is being addressed by the licensee in his response to the
CFR 50.54 (F)
request for information.
Surveillance Testing Observation (61726)
The inspectors observed and/or reviewed the below listed surveillance procedures.
The inspection consisted of a
review of the procedures for technical adequacy, conformance to technical specifications, verification of test instrument calibration, observation on the conduct of the test, removal from service and return to service of the system, a review of test data, limiting condition for operation met, testing accomplished by quali-fied personnel, and that the surveillance was completed at the required frequency.
a.
S I-4. 1. B-16 b.
OI-99 c.
EMI-B Reactor Protection System MG Set Reactor Protection System Reactor Protection System MG Set Inspection, Maintenance, and Repair d.
BF"10.9 No violations or Handling of Test Deficiencies.
deviations were found in the above are Reportable Occurrences (90712, 92700)
The below listed licensee events reports.(LERs) were reviewed to determine if the information provided met NRC requirements.
The determination included:
adequacy of event description, verification of compliance with technical specifications and regulatory requirements, corrective action taken, existence of potential generic problems, reporting requirements satisfied, and the relative safety significance of each event.
Additional in-plant reviews and discussion with plant personnel, as appropriate, were conducted for those reports indicated by an asterisk.
The following licensee event reports are closed:
LER No.
Date Event 259/85-24
~259/85-39 6-12-85 Relay Failure Causes PCIS Signal 8"15-85 MSIV Leakage Excessive
"259/85-41 8-14-85 Diesel Generators Inoperable Licensee event report (LER) 259/85-48 was reviewed and the information contained in paragraph two appeared misleading.
This paragraph implies the reactor protective system cable with the charred insulation was identified while performing Surveillance Instruction (SI) 4.2.C-1A on September 29; 1985'he problem was first identified by the resident inspector on September 24, and discussed wit) the licensee (Reference IE Report 85-49).
Since a separate LER was being prepared by the licensee concerning the cable problem, the whole paragraph was unrelated.
Additionally, the Energy Identification System component function identifier and system name of each component or system referred to in the LER was missing from the LER.
This is one of the items to be contained in the LER as discussed in 10 CFR 50.73.
A review of the completed SI data cover sheet found no mention of the arcing or cable problem.
The block on the coversheet to indicate
"delays or problems" was checked
"No".
Data cover sheets of completed SIs which do not accurately reflect delays or problems continue to be a recur ring problem at the plant.
This area will remain an unresolved item pending a
complete review of plant procedures in this area - BF 10.3, Corrective Action Program and BF 10.9, Handling of Test Deficiencies.
(259/260/296/85-49-04).
Due to several-licensee event reports (LER) being issued that concern the isolation of containment during reactor protective system (RPS)
power supply shifts, a review of this area was conducted.
The RPS system is normally supplied by A and B RPS motor generator (MG) with an alternate transformer supply capable of supplying either the A or B
RPS bus.
The transfer from normal to alternate power supply must be performed manually with a temporary loss of power to one RPS bus.
Loss of an RPS bus causes a half scram and loss of power to part of the Primary Containment Isolation System (PCIS)
logic.
Several recent LERs discuss the loss or transfer of the RPS MG set
and corresponding containment isolation.
The LERs are 260/85-08, 296/85-18, and 259/85-22.
For LER 259/85-22 on June 10, 1985, a plant Surveillance Instruction, SI 4. 1.B. 16 RPS MG Set, was revised to include a precaution that a half scram would occur along with various isolations.
The following actions will occur:
1)
2)
3)
4,)
5)
6)
7)
Half-scram Control room emergency ventilation initiation Standby gas treatment initiation PCIS group II (shutdown cooling) isolation PCIS group III (reactor water cleanup)
isolation PCIS group YI (primary containment vent and purge including reactor building ventilation) isolation PCIS group VIII (Tip valves) isolation The LERs stated that all safety systems performed as designed, and no adverse effects were noted.
Of particular concern to the inspector for a unit in cold shutdown was the isolation of shutdown cooling due to a planned power supply shift.
A similar event occurred on October 11, 1985, when the RPS power supply was shifted from the MG set to the transformer in order to perform maintenance on the MG set.
The shift was accomplished using Electrical Maintenance'nstruction EMI-13.
This instruction did not contain the precaution
'concerning containment isolations.
The inspector thought all the transfers should have been accomplished using
'.he plant Operating Instruction OI-99 for the RPS system, however, a review of SI 4. 1.B. 16 and EMI-13 found no reference to OI-99.
Likewise, the precaution about the various containment isolations and initiation of safety equipment was not addressed in the Operating Instruction.
The inspector concerns, that the containment isolations occur during RPS power supply transfers, that the transfers can be made using three different procedures with different precaution statements, and that the isolations result in a loss of shutdown cooling and initiation of safety systems during planned activities, were discussed with plant management on October 16, 17, and 21, 1985, respectively.
This area will remain unresolved pending resolution of the inspector concerns (259/260/296/85-49-05).
During comparison of the LERs addressed above the inspector found some inconsistencies.
LER 260/85-08 addressed only a
PCIS group III (reactor water cleanup isolation)
upon loss of RPS power.
LER 296/85-18 listed all of the containment isolations and equipment initiations for a loss of RPS power.
LER 259/85-22 stated only that a containment isolation occurred and did not list any of the details.
In
CFR 50.73 (b)(2)(i) it is stated that the content of the LER shall contain a clear, specific, narrative description of what occurred so that knowledgeable readers conversant with the design of commercial nuclear power plants, but not familiar with the details of a particular plant, can understand the complete event.
The mentioned LERs do not meet the intent of this requirement as the inspector who is familiar with the details of the particular plant can only understand the complete event after considerable researc Regulatory Performance Improvement Program (RPIP)
The responsible section chief reviewed the status of RPIP and actions taken by TVA to implement specific items as required by NRC Confirmatory Order EA 84-34 dated July 13, 1984, during a site visit of October 15-17, 1985.
TVA had assigned a senior manager as RPIP Coordinator at the site.
The senior manager was reassigned as of October 18, 1985, and no new RPIP Coordinator will be assigned.
His responsibilities include verifying that each task has been implemented as described, has met objectives, and that the necessary programs are in place to insure that objectives will continue to be met.
Most of the short term items have been indicated as complete, but have not been signed off as completed by the RPIP Coordinator.
Based on the above review, the following items are closed:
Short Term Item Number Descri tion 4.27 (84-SC-51)
Implement QA Organization Long Term Item Number Descri tion 4.2 (84-SC-69)
Determine Ma'gnitude Of Modification Work That Can
'e Performed Consistent With Maintaining Positive Control 5.1 (84-SC-70)
5.2 (84-SC-71)
5.3 (84-SC-72)
Review Procedures In Areas Of Weakness And Make Recommendations OQA Develop Audit Plan Emphasizing Areas Of Weakness OQA And QE Develop Coordinated Audit Plan 6.1
, (84-SC-73)
Establish Committee To Resolve Procurement Conflict Between QA, Government Regulations And Need To Assure Adequate Supply 8.1 (84-SC-76)
8.2 (84-SC"77)
9.4 (84-SC-81)
9.5 (84-SC"82)
Establish Additional Controls For Design Change Request Establish Long Term Modification Schedule Improve Participation In NPRDS Improve Maintenance Of Equipment History By Utilizing MR Tracking Data Base
(cont'd)
9.10 (84-SC-87)
9.11 (84-SC-88)
9.8 (84-SC-85)
7.3 (84"SC"56)
Expand Computerized Systems To Include Scheduling And Coordination Of All Maintenance Activities Complete Development And Implementation Of SI Packaging Program Outside Contractor Evaluate Modification Program Provide "Live Time" On-Shift Experience Review Training 10. Modifications (37701).
The inspector was updated on the backlog work team activities.
The scope of their activities were laid out in RPIP objective number 2.0, to ensure that the backlog of modifications paperwork is brought up-to-date and closed out.'he RPIP was initiated in February 1984.
Although a licensee report dated July 27, 1983, indicated drawing problems with closed out work plans, this item was not specifically addressed in the RPIP.
A survey of 64 work plans found that an estimated 50 percent contained drawing discrepancies.
It is the licensee's intention to perform a complete review of a11 closed out work plans associated with complete'8 engineering change notices as a
long-term RPIP item.
Since the report on July 27, 1983, no further correction or review of completed work plans had taken place and no date for completion of their review has been established.
This will remain open for further tracking.
(259/260/296/85-49-06).
ll.
Nuclear Safety Review Board (NSRB) (40701)
The NSRB functions to provide independent review and audit of licensee safety related activities.
The inspector reviewed activities related to the conduct and operation of NSRB functions as required by Technical Specification 6.2.
A review of the NSRB meeting minutes was conducted for 1985.
Four meetings (No. 209-212)
were conducted as of the date of this inspection.
The remainder of this inspection wi Wl be concluded at the Corporate Office at a later date; however, the items listed below were noted and should be addressed presently:
a.
The NSRB members shall be appointed in writing by the Manager of Power and shall be appointed for two-year tenures.
A review of appointment letters indicated that the current NSRB (as of October 21, 1985)
consists of five regular members and two alternate members.
Two members'embership expired October 13, 198 b.
Technical Specification 6.2 requires NSRB to review all reportable events.
The review indicated that
CFR 50.72 reports are not reviewed.
A review of meeting No.
212 (July 1985)
indicates that reviews of safety significant issues are not concluded in a timely manner.
At meeting No.
212 there were
NRC inspection reports reviewed dating back to 1983 and 161 Plant Operating Review Committee Meeting Minutes reviewed dating back to November 1984.
Failure to complete reviews timely could lead to inadequate reviews, failure to recognize safety deficiencies as they occur, and/or failure to ascertain that nuclear safety aspects are adequately considered.
d.
Review of technical specification changes not completed in a timely manner although amendment request issued to NRC in February 1985 and noticed in Federal Register in March 1985 (i.e.,
RHR system interunit crosstie ability deletion review not concluded as noted in July 1985, NSRB Meeting No. 212).
This item is unresolved pending further review (259/85-49-07).