IR 05000259/1989003
| ML18033A705 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 04/10/1989 |
| From: | Carpenter D, Little W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML18033A704 | List: |
| References | |
| 50-259-89-03, 50-259-89-3, 50-260-89-03, 50-260-89-3, 50-296-89-03, 50-296-89-3, NUDOCS 8904200182 | |
| Download: ML18033A705 (25) | |
Text
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Wp0 cu gp 0**pW UNITED STATES NUCLEAR REGULATORY COMMISSION
REGION II
101 MARIETTASTREET, N.W.
ATLANTA,GEORGIA 30323 Report Nos.:
50-259/89-03, 50-260/89-03, and 50-296/89-03 Licensee:
Tennessee Valley Authority 6N 38A Lookout Place 1101 Market Street Chattanooga, TN 37402-2801 Docket Nos.
50-259, 50-260, and 50-296 License Nos.
DPR-33, DPR-52, and DPR-68 Facility Name:
Browns Ferry Nuclear Plant Units 1, 2, and
Inspection at Browns Ferry Site near Decatur, Alabama Inspection Conducted:
January 1-31, 1989 Inspecto
. Car e
r, NRC Site Manager Date igned Accompanied by:
W. Bearden, Resident Inspector K. Ivey, Resident Inspector E. Christnot, Resident Inspector A. Johnson, Project Engineer A.
Lo g, Pro 'ect E gin er Approved by:
W. S. Little, ect'
Chief, Inspection Programs, TVA Projects Division Date igned SUMMARY Scope:
This routine resident inspection was conducted in the areas of operational safety verification, refueling activities, review of fuel load procedures, surveillance observations, maintenance observation, restart test program, cold weather preparation, health physics survey, followup of open inspection items, and licensee action on previous enforcement matters.
This inspection included continuous NRC onsite coverage of Unit 2 core reloading activities.
Results:
No violations or deviations were identified.
One inspector followup item was identified:
260/89-03-01:
Interference Between Electrical System and Source Range Monitor (SRM) (restart),
paragraph
8904200182 8904i0 PDR ADOCK 05000259
The inspector followup item must be resolved prior to restart of Unit 2.
During the observation of control room and core loading activities, the NRC inspectors noted a
general improvement in the level of professionalism and seriousness of licensed operators in the performance of their. duties.
In paragraph 3,
a review of the DBVP calculation supporting the incorrect opening time on the SBGT Train 8 inlet damper is documented.
The NRC inspectors were concerned that the licensee relied upon the low decay heat level of the fuel in this and other evaluations.
Although this calculation was adequate to support continuing with core reload, additional analysis 'is required to support Unit 2 restart.
In paragraph 3, observation of Unit 2 core reloading activities is documented.
Although
.various delays occurred due to hardware problems and significant weaknesses were identified associated with fuel loading procedures and
CFR 50.59 Safety Reviews, the licensee's corrective actions were considered proper and thorough and the core reload was resumed with no further weaknesses noted.
Throughout the core reload effort observed licensee activities were carried out in a professional manner.
In paragraph 10, a
review of an attempted falsification of a
RWP sign-in record is documented.
Although falsification would have constituted a violation of NRC requirements, it is noteworthy to mention that the attempt was promptly identified by the licensee and prevented through the professional behavior of licensee employees.
In paragraph 15, a concern is documented relating to the licensee's change and implementation of system configuration contro REPORT DETAILS Persons Contacted Licensee Employees 0. Kingsley, Jr.,
Senior Vice President, Nuclear Power C.
Fox, Vice President and Nuclear Technical Director J
~
Bynum, Vice President, Nuclear Power Production
- C. Mason, Acting Site Director G. Campbell, Plant Manager H. Bounds, Project Engineer
- J. Hutton, Operations Superintendent
"D. Mims, Technical Services Supervisor G. Turner, Site guality. Assurance Manager
"P. Carier, Site Licensing Manager J.
Savage, Compliance Supervisor A. Sorrell, Site Radiological Control Superintendent R. Tuttle, Site Security Manager L. Retzer, Fire Protection Supervisor HE Kuhnert, Office of Nuclear Power, Site Representative Other licensee employees or contractors contacted included licensed reactor operators, auxiliary operators, craftsmen, technicians, public safety officers, quality assurance, design, and engineering personnel.
NRC Representatives
- D. Carpenter, NRC Site Manager
"E. Christnot, Resident Inspector
"W. Bearden, Resident Inspector
"Attended exit interview Acronyms used throughout this report are listed in the last paragraph.
Operational Safety Verification (71707, 71710)
The NRC inspectors were kept informed of the overall plant status and any significant safety matters related to plant operations.
Daily discussions were held with plant management and various members of the plant operating staff.
The NRC inspectors made routine visits to the control rooms.
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 operability; temporary alterations in effect; daily journals and logs; stack monitor recorder traces; and control room manning.
This inspection activity included numerous informal discussions with operators and supervisors.
The NRC inspectors conducted general plant tours including portions of the turbine buildings, each reactor building, and other plant areas.
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; and radiation protection controls.
Informal discussions were held with selected plant personnel in their functional areas during these tours.
The NRC inspectors walked down the SGTS using the following plant drawings and procedures:
0-47E610-65-1 Unit 0 (common):
Mechanical Control Diagram Standby Gas Treatment System 1-47E610-64-1 Unit 1:
Mechanical Control Diagrams Primary Containment Systems 2-47E610-64-1 Unit 2:
Mechanical Control Diagrams Primary Containment Systems 3-47E610-64-1 Unit 3:
Mechanical Control Diagrams Primary Containment System
0-OI-65 Standby Gas Treatment System Operating Instructions 0-47E865-ll Unit 0 (common):
Flow Diagrams Heating and Ventilation, Standby Gas Treatment System 1 "47E865-1 Reactor Building Unit 1:
Flow Diagram, Heating and Ventilation Air Flow 2-47E2865 Reactor Building Unit 2:
Flow Diagram, Heating Ventilation Air Flow 3-47E865-12 Reactor Building Unit 3:
Flow Diagram Heating and Ventilation Air Flow The NRC inspectors noted that control diagram drawing 3-47E610-64-1 indicated that the end of the Unit 3 SGTS ductwork was open to the Unit 3 reactor building, when in fact this end piece is closed and welded.
The NRC inspectors discussed this item with the plant Technical Staff and it is not considered significant because flow diagram drawing 3-47E865-12 showed the end piece to be capped off.
The System Engineering group was contacted and a
Drawing Discrepancy (DD) was submitted to correct the control diagram drawing.
The BFN facilities control diagrams show only
those items that have controls, such as remote switches, automatic actuators, and transmitters with remote indicators and the flow diagrams show only items that are directly attached to the piping.
The NRC inspectors used both types of drawings to verify that the operating procedures corresponded to the actual plant arrangements.
No major equipment discrepancies were identified.
However, the inspectors noted that the back draft damper on SGTS Trains A & B, numbers65-505 and 506 respectively appeared to be completely covered with insulation and not identified by external labeling.
This item was also discussed with the plant Technical. Staff for resolution and a review of the drawings for the system indicated that the back flow dampers were to be completely covered with insulation.
The components in the Reactor Buildings, Standby Gas Treatment Rooms, and the Control Room were adequately labeled and identified to support operation of the system.
No violations or deviations were identified in the area of Operation Safety Verification.
Refueling Activities (60710, 93702)
The licensee completed closure of all outstanding SPOC exceptions for systems required for fuel loading on January 1,
1989, and commenced reloading the Unit 2 reactor core on January 3,
at 10: 17 a.m.
NRC inspectors provided continuous on-site coverage during core reloading activities.
Core reloading continued until January 5,
when the licensee suspended the activities following the NRC's questions about the lack of adequate core neutron monitoring, and the technical adequacy of the Technical Specifications (TS) and reload procedures.
When fuel reloading was terminated, three of the four Source Range Monitors were reading less than zero and the fourth was reading about 0.2 cps with 74 of 764 fuel assemblies loaded in the core.
Core neutron monitoring was being conducted using only SRMs.
With the spiral loading sequence started at the center of the core, the fuel array was neutroni-
.
cally decoupled from the SRM detectors because of the distance from the fuel to the SRMs.
This raised the concern that the core was not monitored for reactivity changes.
The TSs permitted loading in this manner, but did not specify a minimum core neutron count rate that had to be maintained during the core loading.
There had apparently been an inadequate safety analysis to support the associated TS amendment.
Review of this issue is documented in NRC Inspection Report 259, 260, 296/89-04.
To provide monitoring of core neutrons, the licensee installed movable Fuel Loading "Dunking" Chambers.
Additionally, the procedures associated with fuel reloading activities were revised, requiring the FLCs or SRMs to indicate a
minimum of 3 cps.
They were reviewed by PORC, and operator training was conducted on the procedural changes.
The NRC resident inspector staff reviewed the revised procedures, attended the PORC meetings, evaluated the training effectiveness, and witnessed the installation and calibration of the FLCs prior to the licensee's resumption of fuel. movemen The licensee resumed fuel reloading on January 16, 1989, with core monitoring initially provided by two FLCs until the SRMs could provide adequate response to core neutrons.
Fuel loading proceeded until completion on January 30, 1989.
During the core reloading, the licensee experienced several delays.
With the exception of the delay due to the failure to provide adequate core neutron monitoring, the problems were mostly, minor hardware related problems with the refueling bridge crane and fuel grapple.
During the continuous onsite coverage of core reloading activities, the NRC.inspectors witnessed fuel handling and movement, refueling support activities such as surveillance testing, maintenance, and refuel floor radiation controls as well as routine shift activities.
The NRC inspectors observed ongoing activities in the control room, on the refueling floor, and made routine plant tours to acquaint themselves w'ith plant conditions'he NRC inspectors observed that licensee personnel performed periodic testing and verification of the operability of refueling related equipment and systems as required by the TS and licensee procedures.
Various tests such as SRM response checks and refueling platform interlock verifications were directly observed and/or reviewed to confirm that all required checks were performed within the time constraints stipulated.
Applicable portions of the prerequisites contained in 2-GOI-100-3,
"Refueling Operations",
were performed prior to resumption of refueling activities after greater than an eight hour delay as required.
Various licensee operations and management personnel were interviewed concerning their responsibilities, understanding of procedural requirements and status of ongoing refueling activities with acceptable results.
Continuous communications were maintained between the control room and the refueling floor any time fuel was being moved and control room personnel were located so as to directly observe SRM indication.
The adequacy of site management involvement in the ongoing activities was evident in that a
licensee duty manager was onsite, normally in the control room area, throughout the period.
Each of these managers was a
SRO licensed individual who was in addition to the normal shift complement.
Additionally, the Shift Operations Supervisor (SOS)
work station had been recently relocated into the control room directly adjacent to the Unit 2 controls area.
The NRC inspectors noted that both the duty manager and the SOS along with other operations personnel appeared to be fully"cognizant of refueling procedures, the status of core refueling activities and any required ongoing testing or maintenance.
Management attention was further evident in that the NRC inspector noted recent improvement in the noise levels and traffic in the control room.
During the refueling activities, SRM Channel C was declared inoperable due to s'ignificant spiking.
The licensee monitored the spiking and determined that an unknown phenomenon is causing the spiking whenever the Primary Suppression Chamber (PSC)
head tank pump number 2A starts and stops.
The
PSC head tank maintains the core spray and RHR low pressure injection piping filled to prevent water hammers.
The licensee tagged out the 2A pump and the spiking stopped.
The NRC inspector reviewed drawing 2-47E814-1,
"Flow Diagram Core Spray System,"
and could not determine any relationship between the PSC head tank system and the neutron monitoring system.
This item is identified as Inspector Followup Item (IFI)
260/89-03-01, Interference Between Plant Electrical Systems and SRMs.
This item applies to Unit 2 only and must be addressed prior to restart.
The licensee initiated a
MR to troubleshoot and determine the cause of this unknown phenomenon.
4.
Fuel Load Procedure Review (60705)
Prior to the resumption of fuel loading, the NRC inspectors reviewed the proposed procedure changes associated with the source range monitoring issue (see paragraph 3).
A general comment was made concerning the procedures in that the licensee uses the term
"operable" when
"responding",
or
"monitoring",
are more applicable.
The following specific comments were noted.
'a ~
2-GOI-100-3:
Refueling Operations (1)
Section
made no mention of waiting to ungrapple a
fuel assembly until the responding SRM has settled out.
This i s considered to be good practice.
(2)
Section 3.21 and Section 4.21 both deal with placing the SRM in the noncoincidence scram mode (all shorting links removed),
and they did not agree with each other.
(3)
Section 3.25, which outlines the assignment of personnel to the refuel floor, was not in keeping with the TS.
(4)
Section 3.36 directed personnel when to halt refueling.
NRC inspectors determined that this section should use the words responding and/or monitoring instead of operable.
Subsection 3.36. 10 was difficult to understand and should be simplified.
(5)
Section 3.38 directed personnel where to place a fuel assembly in the process of being moved if core alterations are suspended.
This section should stipulate that the assembly is not to be placed into the core.
(6)
Section 3.51 discussed jumpering or bypassing interlocks.
It shou'ld indicate what specific plant procedure to use in doing this type of activity.
(7)
Section 3.56, due to the system interaction between the SRM and the fire/medical emergency alarm, directed that fuel handling operations be suspended while the alarm is sounding.
However,
it does not direct who may authorize the resumption of fuel load and under what criteria.
(8)
Section 4.24 indicated which SRMs or FLCs are to be operable for fuel or control rod movement; however, this section did not indicate what the minimum count rate should be.
(9)
Section 4.28 discussed control rod indication, but did not indicate which of the three methods of indication was acceptable.
( 10) Section 5.4 discussed operations during refueling or movement within the Spent Fuel Storage Pool (SFSP),
but was not clear on defining
"unexpected subcritical multiplication."
Subsection 5.4.7 discussed recording the SRM or FLC reading, but did not mention SRM or FLC response time.
I b.
TI-14:
Special Nuclear Mater ial Control The NRC inspectors reviewed this procedure and noted that Steps
thru 100 were changed in order to utilize the FLCs.
c.
S-II-2-XX-92-095:
Fuel Loading Chamber Instructions (1)
All of Section 7.2, unlike Sections 7. 1 and 7.3, had no initial blocks where individuals performing the activity of assembling the FLCs could sign off each step performed.
(2)
Section 7.3 discussed the SRM to FLC conversion.
Subsection 7.3. 13, discussed the recording of steady state indication and referenced attachment 3.
However, a review of the attachment did not indicate where the recording was to be documented on the attachment.
d.
2-SI-4. 10.B:
Demonstration of SRM System Operability During Core Alternations
.The NRC inspectors reviewed this procedure and did not have any comments.
e.
2-SI-4.2.C-4(A)
and (B):
Instrumentation That Initiates Rod Block/Scrams SRM Calibration and Functional Test The NRC inspectors reviewed these two procedures and noted that in both procedures the number of operable SRMs as well as their location was not consistent with the fuel loading procedur f.
2-TI-147:
Fuel Loading After a Complete Core Reload (1)
Section 4.2.8 and 7.6, which discussed the per formance of subscriticality checks and the SRM readings, needed to be clarified in that they were difficult to understand.
(2)
Appendix 3 dealt with subcriticality checks.
Step 1 directed the nuclear engineer to choose a control rod to do a subcriti-cality check; however, it did not give clear instructions on how to select a control rod.
One other general comment made by the NRC inspectors involved information given to the inspectors during briefings by TVA Staff and Senior Managers to the NRC Staff and Senior Managers.
The specific information had to do
'ith clarifying TSs by adding further administrative controls and included the following:
the number and locations of operable SRMs (should be responding or monitoring) in the quadrant where core alterations are being made (TS 3. 10.8. 1);
core alterations suspended if RHR and Core Spray systems are inoperable (TS 3.5.A and 3.5.8);
reactor building isolation functions are to be operable when secondary containment integrity is required (TS 3.2.A);
and the number of Emergency Equipment Cooling Water pumps necessary while refueling (TS 3.5.C. 1.).
In the procedure changes reviewed, the NRC inspectors could not identify where these specific clarifications were spelled out, However, further review by the NRC inspectors indicated that the operations night orders were used to clarify these items.
The above review comments were discussed with BFN Senior Management, and the NRC comments were adequately resolved.
Surveillance Observation (61726)
The NRC inspectors observed and/or reviewed the surveillance instructions described below.
The inspection consisted of a review of the procedures for technical adequacy, conformance to TS, 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 conditions for operation met, testing accomplished by qualified personnel, and that the SI was completed within the required frequency.
During fuel loading, the NRC inspectors observed the periodic performance of 2-SI-4. 10.8,
"Demonstration of Source Range Monitor System Operability During Core Alterations."
This SI was performed every eight hours during fuel movement.
2-SI-4.2.C-4 (A) and 2-SI-4.2.C-4 (8) were performed to replace SRMs A and 8 with fuel load chambers.
2-SI-4. 108 required that the incore SRMs be driven out until a
25% minimum change was detected in the count rate or the FLCs be withdrawn from the core by the fuel handling equipment for the same 25% change.
In all cases observed the change was always greater than 25% before the incore SRMs were driven back up into the core or the FLCs were lowere No violations or deviations were identified in the area of Surveillance Observation.
Maintenance Observation (62703)
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 systems to service; quality control records were maintained; activities were accomplished by qualified personnel; parts and materials used were properly certified; proper tagout clearance procedures were adhered to; TSs were adhered to; and radiological controls were implemented as required.
Maintenance requests were reviewed to determine the status of outstanding jobs and to assure that priority was assigned to safety-related equipment maintenance which might affect plant safety.
MR869786 was written by the licensee to provide as needed mechanical maintenance support on portions of the refueling platform that are not safety related and was kept open to cover.repairs unti 1 core reload was completed.
The NRC inspectors observed ongoing work including replacement of the main hoist hose reel, hose, and air hose connector, and grapple solenoid valve.
Additionally, the NRC inspectors observed adjustment of the grapple mechanical stop mechanism per GE recommendations.
MR912716 and MR912714 were written to allow replacement of the Unit 2 refueling platform hose take up reel with the take up reel from the Unit
refueling platform.
The NRC inspector reviewed the completed work packages and noted no discrepancies.
For the above maintenance requests associated with the refueling platform, the NRC inspector noted that the applicable portions of MMI-34, Refueling Platform Inspection and Repair, were performed as post-maintenance testing prior to returning the refueling platform to service.
An NRC inspector observed ongoing work associated with MR912107 which verified that limit switch settings for FCO-65-25 (B train
correctly reflected the full open and full closed position.
During the work licensee personnel removed the damper cover and directly observed damper operation.
The damper had appeared to operate slower during the performance of SI-4.7.B.3C, SBGT Operability, than the
seconds specified in FSAR Section 5.3.
The.licensee determined that the damper limit switches were properly set and that the actual cycle time was closer to
seconds.
This issue is covered in more detail in Paragraph 2.
During the performance of the Design Baseline Verification Program the
second stroke time requirement was eliminated as a func-tional criteria.
This change must be reflected in an update to the FSAR at the completion of the DPVP.
The licensee performed a safety evaluation
which determined that the SBGT System operability was not affected by the increased stroke time.
The PORC reviewed the SE and concurred.
The NRC inspector noted that continuous health physics coverage of the work was provided due to the potential for airborne radioactivity.
No violations or deviations were identified in the area of Maintenance Observation.
7.
System Return To Service (71711)
The licensee completed the system return to service under the System Pre-operational Checklist (SPOC)
process.
The NRC inspectors had the following concerns, which were adequately resolved by the licensee:
a.
The initial walkdowns and system statusing process seemed somewhat confusing in that outstanding issues such as thermal overloads were not adequately screened for system impact.
b.
The use of deferrals and SPOC exceptions did not at first appear to be well understood by the licensee personnel engaged in the SPOCing process.
However, as the SPOC program progressed, the system engineers, the SROs and the supervisors became more aware of the use of deferrals and exceptions.
The NRC inspectors observed that during the walkdowns of a system it was not clear to the SROs where one system ended and another started, i.e.,
the DG 125 volt DC control panels were at one time part of System 57,
"Auxiliary Electrical,"
and also part of System 82,
"Diesel Generators."
This item was discussed with BFN management and the NRC inspectors were informed that the interface would be a shared responsibility, required by both systems.
The licensee stated that the lessons learned in performing the SPOCs for fuel load systems would be applied to the SPOCs to be done for the systems yet to be returned to service for criticality.
8.
Restart Test Program (RTP) (99030B)
The NRC inspector reviewed the test summary of the BFNP - Diesel Generator Evaluation Report, dated January 20, 1989.
This report discussed the actions taken as follows; Provide the diesel generator load summary report Evaluate the effects of reduced voltage starts on the residual heat removal pump motors to demonstrate adequacy of operation for plant life Provide summary and resolution of restart record problem
Provide analysis and closeout of problem with the diesel power available time sequence exceeding sequence used in
CFR 50, Appendix K analysis.
The NRC inspector noted that in the Diesel Generator Voltage Analysis,Section IV, Subsection 4,
480V Board Feeders and MG Set Feeders, no specific mention is made of the four Unit
Low Pressure Core Injection ( LPCI)
Motor Generators (MG).
The NRC inspector determined during conversations with various licensee operations personnel that there has been a history of frequent supply breaker tripping during attempted starts of the LPCI MGs.
These MG sets are equipped with flywheels which apparently result in large electrical transients when started from a
stationary condition.
The MG sets were included in the Bechtel Load study and the NRC inspector believes that the effects of the voltage dips and the interim loss of power on the MG sets should have been addressed in the licensee's load summary report in more detail.
Although no other concerns associated with the licensee's load summary report were noted during the review by the NRC inspector, the NRC inspector believes that this issue should receive additional clarification.
The NRC inspector observed many of the DG restart tests as they were being performed.
Additionally, the NRC inspector reviewed initial as well as completed data and attended numerous meetings spanning a
one year period when aspects of the eight BFN DGs were discussed.
Other than the LPCI MG sets, the NRC inspector found nothing in this report that would indicate any unknown or undiscovered DG issues.
No violations or deviations were identified in the Restart Test Program area.
Cold Weather Preparations (71714)
The NRC inspector reviewed the licensee's program to protect plant systems and equipment important to safety from cold weather conditions.
At the RHRSW/EECW intake structure the inspector noted that all heat trace wiring appeared to be intact and'hat new insulation on all piping systems was in place.
The inspector had been observing work activities in this area over the past months.
The heat tracing material was new and appeared to be acceptable and was adequately spaced around the large piping and laid on small piping.
An unresolved item was documented in a previous inspection report concerning the RHRSW/EECW structure (URI 260/87-46-04)
and this item will remain open until CARR BF 870018, written to document the freeze protection deficiency, can be reviewed for adequate closeout.
Additional observations documented in NRC Inspection Report 88-16 were made concerning heat tracing which included procedure upgrade, listing of the system as important-to-safety and the acceptability of a calculation.
The inspector reviewed procedures EMI-46, Revision 3,
Freeze Protection Program, and GOI 200-1, Revision 02.
The NRC inspector noted that EMI-46 is extensive and covers over 37 areas of the BFN facility.
Procedure
GOI-200-1 is equipped with a checklist for insuring that the system is lined up adequately.
MRs 897220, 902073 and 902072 were written to correct out-of-calibration switches and alarm problems in the freeze protection system.
The inspector reviewed the MRs and noted that each problem was adequately documented and corrective action was in place.
The inspector also walked down 480 V auxiliary DG boards A and B and noted that all circuit breake>
s for the heat trace system were closed.
The status of the system for the freeze protection heat trace appeared adequate to prevent damage to systems important to safety due to cold weather.
The RHR/EECW rooms A and B were covered over with canvas and plastic and equipped with large portable heaters in order to protect workers in these rooms.
The NRC inspector could not visually verify the thermostat settings for the strip heaters located inside instrument cabinets or 'for the room heaters such as those located inside the OG rooms, however they appeared to be working satisfactorily.
These observations were discussed with various BFN plant senior managers, who confirmed that they were following the scheduled TVA cold weather program.
The inspectors had no further questions.
No violations or deviations were identified.
Health Physics Surveys The licensee identified a condition on January 25, 1989 in which a member of the Radiological Control staff unsuccessfully attempted to falsify a radiological survey as being performed.
The i'ndividual transferred social security numbers, entry times, arid exit times from an original sign-in sheet for RWP 89-8003 and while doing so placed his name on the duplicate in such a
manner that it would appear as if he had performed a survey on January 24, 1989, at 1:00 a.m.
The licensee determined that the name of the RadCon personnel was not on the original sign-in sheet, a survey of the area was not performed at that time, and the last survey of the area expired on January 23, 1989, at 12:00 a.m.
The licensee further determined that the next actual survey was performed on January 25, 1989, at 1:00 a.m.
The licensee instructions allow for a 25 percent extension on radiological surveys and with a periodicity of 7 days for this survey, the extension would expi re at 6:00 p.m.,
on January 25, 1989.
Thus the surveillance frequency was not exceeded.
A licensee review of the individual's work history and job assignments did not indicate any clear example of previous falsifications or unsatis-factory job performance.
Although falsification would have constituted a
violation of NRC requirements, it is noteworthily to mention that is was prevented through the professional conduct of licensee employees.
TVA took disciplinary action against the individua.
Technical Support Center (TSC) Staff Training A
NRC inspector attended the TSC Staff Training course given on January 12, 1989.
This training is given to personnel who may be required to respond to the TSC in the event of a plant emergency.
The lesson included various aspects of emergency response and TSC operation such as:
(1) the purpose of the Radiological Emergency Plan and its planning basis, emergency classifications, and procedures; (2)
TSC and emergency re.ponse staffing, procedures, and responsibilities; and (3) onsite and offsite emergency facilities, their organization, and locations.
No deficiencies were identified.
No violations or deviations were identified during the TSC Staff Training session.
12.
Licensee Action on Inspector Followup Items (92701)
(OPEN)
Inspector Followup Item 259, 260, 296/87-09-04:
Resolution of RHRSW Pump Vibration Caused By Resonance.
This issue was adequately addressed for fuel load in NRC Inspection Report 259, 260, 296/88-24, but requires followup action.
Therefore, this item is being reopened until review is complete of the licensee permanent corrective actions for the RHRSW pump vibration problem.
This item is required to be resolved prior to Unit 2 restart.
13.
No violations or deviations were identified in the area of Licensee Action on Inspector Followup Items.
Licensee Action on Previous Enforcement Matters (92702)
(OPEN)
Unresolved Item 260/87-46-04:
Followup On Generic Implications Of Heat Trace CAQR Disposition.
This item was originally identified due to numerous deficiencies in the installation of the heat trace in the RHRSW/EECW systems located in the intake pumping station.
This item was further expanded in NRC Inspection Report 250, 260, 296/88-16 to include procedure upgrades, Q-List status of heat trace and acceptability of calculation MEB-BWR-M2-751-1.
The NRC inspector-reviewed the status of the issues and determined that they are not fully resolved.
This item will remain open and must be resolved as part of the next scheduled review of cold weather preparations.
This is not a restart item.-
No violations or deviations were identified in the area of Licensee Action on Previous Enforcement Matters.
14.
Site Management and Organization (36800)
The licensee successfully completed the reloading of 764 fuel assemblies into the Unit 2 Re'actor Vessel during this reporting period.
After an extended shutdown of approximately four years this must be recognized as a
significant milestone.
However this was not accomplished without
problems.
The loading of fuel without adequate core neutron monitoring as described in NRC Special Reactive Inspection Report, 259, 260, 296/89-04, is indicative of a lack of adequate over sight and management attention to licensed activities.
Although licensee management was initially reluctant to acknowledge the full significance of this issue, the corrective actions were considered to be thorough.
After reloading activities were resumed on January 16, various delays occurred due to hardware problems.
In each case licensee management actions were fully adequate and conservative.
NRC inspectors observed the Plant Manager and licensee management con-ducting frequent plant tours in the Control Room, Turbine and Reactor Buildings.
15.
System Status Control As a
result of NRC findings, documented in Inspection Report 259,260,296/88-36 dealing with system configuration control, the licensee committed in a letter to the NRC dated December 29, 1988 to the following actions prior to fuel load of Unit 2.
Issue a
new PORC approved system status control procedure, plant managers instruction (PMI) 12. 15 "System Status Control."
Perform PMI-12 on five selected systems.
All deviations found during the initial system alignment using PMI-12. 15 will be documented and receive the same technical review as a procedure change.
All documentation required by PMI-12. 15 will be gA records.
This will allow for a clear and auditable trai 1 of system status control.
Fuel load was recommenced on January 3,
1989 at which time the licensee stated the above commitments had been completed.
NRC inspector follow up during the month of January identified deficiencies in the status file, including numerous administrative errors in the completed alignment checklists performed per PMI-12. 15 on the five selected systems.
The observed deficiencies did not reflect any mispositioned components.
The licensee did not appear adequately responsive to NRC questions and comments on this issue.
This was in part attributed to the problems encountered as a result the unmonitored fuel load event, event resolution and resumption of fuel loading.
Near the end of the inspection period, the licensee assigned a designated manager to this area.
As the designated manager began investigating the issues, the problems were recognized and acknowledged, and the licensee accelerated their efforts in this area.
By the end of the inspection period, the licensee appeared to have a grasp of the problem, and correc-tive actions were on track, The initial lack of licensee sensitivity and
attention to detail will be considered in the resolution and closeout of violations in Inspection Report 88-36 dealing with configuration control.
16.
Exit Interview (30703)
The inspection scope and findings were summarized on January 31, 1989, with those persons indicated in paragraph
above.
The inspectors described the areas inspected and discussed in detail the inspection findings listed below.
The licensee did not identify as proprietary any of the material provided to or reviewed by the inspectors during this inspection.
Dissenting comments were not received from the licensee.
Item Number Descri tion and Reference 89-03-01 Inspector Followup Item:
Interference
'Between Electrical Systems and SRMs (paragraph 3)
17.
Acronyms and Initialisms BFNP CAQR CPS CS CSSC DD DG DNE DBVP EECW FLC FSAR GE HVAC IFI JTG LIV LOP/LOCA LPCI MG MR NOV NRC OI PORC PSC RHR RHRSW RTP RWP SBGT Browns Ferry Nuclear Power Plant Condition Adverse to Quality Report Counts Per Second Core Spray Critical Structures, Systems, and Components Drawing Discrepancy Diesel Generator Department of Nuclear Engineering Design Baseline and Verification Program Emergency Equipment Cooling Water Fuel Loading Chamber Final Safety Analysis Report General Electric Heating, Ventilation,
& Air Conditioning Inspector Followup Item Joint Test Group Licensee Identified Violation Loss of Power/Loss of Coolant Accident Low Pressure Core Injection Motor Generator Maintenance Request Notice of Violation Nuclear Regulatory Commission Operating Instruction Plant Operations Review Committee Primary Suppression Chamber Residual Heat Removal Residual Heat Removal Service Water Restart Test Program Radiation Work Plan Standby Gas Treatment System
SFSP SI SLC SOS SPOC SRM SRO TE TS TSC TVA URI VCR Spent Fuel Storage Pool Surveillance Instruction Standby Liquid Control Shift Operations Supervisor System Pre-Operation Checklist Source Range Monitor Senior Reactor Operator Test Exceptions Technical Specifications Technical Support Center Tennessee Valley Authority Unresolved Item Video Cassette Recorder