IR 05000259/1989019
| ML18033A853 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 07/18/1989 |
| From: | Carpenter D, Little W, Patterson C NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML18033A852 | List: |
| References | |
| TASK-3.D.3.4, TASK-TM 50-259-89-19, 50-260-89-19, 50-296-89-19, NUDOCS 8908070235 | |
| Download: ML18033A853 (44) | |
Text
~Pg REMI
+
~
0 cs O
I
+r sp ~*~0 UNITED STATES NUCLEAR REGULATORY,COMMISSION R EG ION II 101 MARIETTASTREET, N.W.
ATLANTA,GEORGIA 30323 Report Nos.:
50-259/89-19, 50-260/89-19, and 50-296/89-19 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 OPR-68 Facility Name:
Browns Ferry Units 1, 2, and
Inspection at Browns Ferry Site near Decatur, Alabama Inspection Conducted:
April 15 - May 15, 1989 Inspectors:
C D.
R. Carpenter, NR Site Manager c
C.
A. Patterson, NR Restart Coordinator
/AS I Q')
Date Signed
'7l(S.-./P'T Date Signed Accompanied by:
E. Christnot, Resident Inspector M. Bearden, Resident Inspector K. Ivey, Resident Inspector A. Johnson, Project Engineer Approved by:
C
<- P
~
~
~
~
~
M.
S. Little, Sect>on Chief, Inspection Programs TVA Projects Division Date Signed SUMMARY Scope:
This routine resident inspection included the areas of operational safety verification, surveillance observation, reportable occurrences, maintenance observation, restart test program, licensee action on previous enforcement matters, and followup of open inspection items.
8908070
, 05000'259 35 8907i8 PDR AD CK
'DC
Results:
One unresolved item was identified:
259,, 260, 296/89-19-01:
Possible Failure to Control Mork Activities on System Important to Safety,, paragraph 3.
The Restart Testing Program was reviewed and evaluated during the inspection period.
Selected completed tests were reviewed to insure the test results were acceptable.
There is concern about the changes in organizational responsibility for restart testing that are planned by the licensee.
This concern will be pursued further in future inspections.
Several open item packages prepared by TVA were reviewed and found incomplete for closure; The NRC staff and the licensee have reached agreement on the review requirements for closure packages.
The licensee has been responsive in this issu REPORT DETAILS Persons Contacted Licensee Employees:
0. Kingsley, Jr., Senior Vice President, Nuclear Power C.
Fox, Jr., Vice President and Nuclear Technical Director
"J.
Bynum, Vice President, Nuclear Power Production
"0. Zeringue, Site Director
~G.
Campbell, Plant Manager R. Smith, Project Engineer
"J. Hutton, Operations Superintendent
"A. Sorrell, Maintenance Superintendent
"D. Mims, Technical Services Supervisor G. Turner, Site guality Assurance Manager
~P. Carier, Site Licensing Manager
"J.
Savage, Compliance Supervisor J.
Corey, Site Radiological Control Superintendent R. Tuttle, Site Security Manager T. Bradish, Plant Report Section L. Retzer, Fire Protection Supervisor D.
Hosmer, Restart Test Manager
"Attended exit interview Other licensee employees or contractors contacted included licensed reactor operators, auxiliary operators, craftsmen, technicians, and public safety officers; and quality assurance, design, and engineering personnel.
NRC Attendees at Exit D. Carpenter, Site Manager E. Christnot, Resident Inspector W. Bearden, Resident Inspector K. Ivey, Resident Inspector Acronyms used throughout this report are listed in the last paragraph.
Surveillance Observation (61726)
An NRC inspector observed the performance of SI procedure 2-SI-4.5.C.1 (2)
"EECW Pump Operation".
The inspection consisted of a review of the SI for technical adequacy and conformance to TS, observation of the conduct of the test, confirmation of proper removal from service and return to service of the system, and a review of the test data.
The inspector also verified that limiting conditions for operation were met, testing was accomplished by qualified personnel, and the SI was completed at the required frequenc During the performance of the test, the operator had difficulty lining up the system for the test due to the current plant lineup.
The SI was not written to allow easy performance in all modes of operation.
The desired lineup was achieved by using the system operating instruction.
The operator indicated that the SI was being revised to resolve these problems.
No additional discrepancies were identified during the performance of this test.
No violations or deviations were identified in the Surveillance Observation area.
3.
Maintenance Observation (62703)
Plant maintenance activities involving 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 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; Technical Specification adherence; 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.
The inspectors observed the below listed maintenance activities during this report period:
On April 6 and 7, 1989, four MRs were written to document missing bolts on four manhole covers located in the intake building, RHRSM rooms A and 0, as follows:
MR874629 4/6/89 Missing manhole cover bolts intake building, RHRSM room A MR874630 4/6/89 Missing manhole cover bolts intake building, RHRSM room A MR874649 4/7/89 Missing manhole cover bolts intake building, RHRSM room D
MR874650 4/7/89 Missing manhole cover bolts intake building, RHRSM room D
The NRC inspector reviewed the MRs and noted that all four were originated by the Mechanical Maintenance group and were signed by a supervisor.
On April 9,
1989, the system engineer noted the condition of the manhole covers and that each cover had an associated
e b.
deficiency tag referencing the above MRs.
This issue was reviewed by BFNP personnel and was initially documented as not reportable by the use of LRED No. 89-02-071.
The system engineer generated a CA(R No.
BFP890330, dated April ll, 1989, which identified that the manway hold down bolts in the A and D
RHRSW pump rooms were discovered missing and this loss of leak tightness could result in the flooding of the RHRSW pump rooms.
It was later determined that a mechanical maintenance walkdown group, tasked with identifying missing bolts on plant equipment had identified this item, written the MRs and hung the deficiency tags.
It was determined from review of the MR 908217 package that the manway covers in rooms B and C were removed and replaced on February 3, 1989, and the manway covers for rooms A and D
were removed on February 9, 1989, and there is no indication that the covers were replaced any time before April 6 or 7, 1989.
The inspector noted that MR 908217 indicated that the manway covers were considered as being a non-critical system structure or component (non-CSSC) yet when the covers for the A and D rooms were removed, this resulted in half of the RHRSW and EECW pumps being inoperable.
The inspector was informed by licensee representatives that the original specifications required that the missing bolts be made of brass.
A walkdown of all four RHRSW/EECW pump rooms was made and it was noted that both of the manway covers in rooms A and D were in place with bolts made of brass.
However, both of the manway covers in rooms B and C did not appear to be fastened with brass bolts.
The licensee reported this event as LER BFRO-50-260/89013 dated May 9, 1989.
This item is considered an Unresolved Item (URI 259, 260, 296/89-19-01)
Failure to Adequately Control Work Activities Involving Systems Important to Safety.
Storage of Main Generator Hydrogen The inspector determined the storage location of the hydrogen in relation to air intakes and the location of safety-related equipment.
The storage location is outside the turbine building adjacent to the east wall and just north of the condensate storage tanks.
The storage structure is described in the FSAR Section 12. 2. 11, Hydrogen Trailer Port (Class II).
The concrete structure is composed of a ground supported slab, walls on three sides, and roof when needed.
Two trailers c'ontaining cylinders of hydrogen gas would be stored in the structure.
Each trailer has a capacity of 48,260 cubic feet with a
system design pressure of 125 psi at 95'F.
A concrete wall is between the trailers and the condensate storage tanks.
The FSAR states that the structure is oriented such that if the gas cylinders become missiles, they will not strike any Class I structures.
This information was passed on to NRC Headquarter.
Operational Safety Verification (71707)
The 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 inspectors made routine visits to the control rooms.
Inspection 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 also included numerous informal discussions with operators and supervisors.
General plant tours were conducted.
Portions of the turbine buildings, each reactor building, and general plant 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; and radiation protection controls.
Informal discussions were held with selected plant personnel in their functional areas during these tours.
The inspectors made routine plant tours for the purpose of observing general housekeeping.
Areas toured for this purpose included the control bay, diesel generators, turbine building, refueling floor and all three reactor buildings.
The inspectors did not identify any examples of significant accumulations of dirt, debris or unused/discarded material.
There appears to have been an increase in management attention in this area and this is evident by the overall improvement in cleanliness observed by the NRC inspectors in selected areas observed.
Although overall housekeeping has improved, there are still several noticeable areas for improvement such as the presence of layers of dust and construction debris in cable trays and in such areas as the cable spreading rooms.
The inspectors followed licensee activities associated with installation of the steam separator assembly into the Unit 2 reactor vessel.
On April 24, 1989, the licensee positioned the gates in the fuel handling cavity above the reactor vessel, lowered the water level in the cavity, transferred the steam separator from the equipment storage pool to the reactor vessel and reflooded the cavity to normal level again.
This activity was conducted in accordance with Special Operation Instruction 2-SOI-27, Reactor Cavity Level Control During Steam Separator Assembly Insertion.
The inspector reviewed 2-SOI-27, and observed actual licensee work including preparations for this activity.
At all times throughout this operation, operations personnel maintained proper communications between the refuel floor and the control room, and were in control of water level.
The activity was conducted in a controlled and professional
manner with an adequate amount of coordination of all planned activities.
Technical Specifications 3.5.B.9 and 3.5.A.4, associated with required ECCS systems, were verified by the licensee to be satisfied prior to commencing the vessel draining activities in accordance with prerequisite step 4.4 of 2-SOI-27.
No violations or deviations were identified in the Operational Safety Verification area.
5.
Action on Previous Inspection Findings a 0 (OPEN)
URI 87-09-02:
Inadequate Evaluation Of The Threat To The Plant From A Barge Shipment Of Explosives.
Significant shipments of explosive substances occur on the Tennessee River near the plant.
Inspection Report 87-09 addressed the concern that these shipments were not adequately evaluated in the FSAR under FSAR question 2.3.
In response to FSAR question 2.3, TVA stated
"Corps of Engineer records show no shipment of munitions or explosive chemicals pass through Mheeler Lock downstream or through Guntersvi lie Lock upstream of the Browns Ferry site. "
The plant design was examined to determine the maximum explosion that structures could withstand.
The FSAR calculation found that the reactor building super structure was limiting and could withstand a
50-ton TNT explosion at the center of the river near the plant.
The inspector questioned that the 50 ton TNT value was sufficiently conservative.
Known shipments of 2000 or more tons of gasoline each are documented in licensee memos from the 1970's and 1980's.
Applying the guidelines of Regulatory Guide 1.91, TNT equivalent values well in excess of the 50 ton level could be postulated.
The licensee response to this concern addressed the impact on plant structures of an explosion of a 3200 ton barge carrying gasoline in the channel of the Tennessee River near the plant.
The evaluation was done in accordance with RG 1.91,
"Evaluations of Explosions Postulated to Occur on Transportation Routes Near Nuclear Power Plants."
That response estimated the effect of such a shipment to be enveloped by the 50 ton TNT original FSAR evaluation.
The inspector reviewed the licensee's calculation, ND-f0000-88120 (BFNAP52-018),
and found that it failed to fully implement RG 1.91.
Assumptions used in the calculation are as follows:
Assum tions Evaluation Basis 3200 ton barge SATISFACTORY Reference 9 in the licensee calculation uses river traffic data
10K of the barge content vaporizes and remains in a suitable cloud configuration for a detonation, SATISFACTORY Reference 6 in the calculation use AMOCO Pipeline Rupture Data The TNT equivalent of vaporized gasoline is 10K mass equivalent of the gasoline UNSATISFACTORY The discussion section of Regulatory Guide 1.91 addresses explosions of vapors in a cloud and a 10K TNT equivalent example is mentioned.
However, the RG warns that the equivalence concept is not well documented and could range up to a 100'ass equivalence.
Therefore, the RG specifies the use of the conservative 240K value derived from
experimental confined space testing unless an analysis including consideration of actual cargo, site topography, and prevailing meteorological conditions justify a lower effective yield.
Furthermore, section C of the RG clearly states:
"In calculating TNT equivalents, assumptions of 100X TNT (mass)
equivalence for solid energetic materials and 240K TNT (mass)
equivalence for substances subject to vapor phase explosions are acceptable upper bounds when effective yield generated from test data do not exist.
Lower effective yields may be justified by analysis accounting for reaction kinetics, site topography, and prevailing meteorological conditions when the hazardous cargos can be identified."
On Hay 3, 1989, the inspector contacted NRR reviewers in the Structural
Geosciences Branch (responsible for reviewing evaluations for conformance to RG 1.91) to ensure no misunderstanding existed in the intent of the RG equivalence discussion.
The NRR reviewers concurred with the interpretation of the inspector.
The 240K value is used because it is based on experimental data.
The discussion on the unconfined space values was an illustration to show how little is known of such detonations and therefore provide a
justification for using the confined space test data as a basis for TNT equivalence estimates.
This topic was discussed with the licensee on May 3, 1989, and will remain unresolved until valid technical justification for the assumptions used in the calculation is provided.
(OPEN)
URI 89-11-02, Single Failure Criteria.
During this inspection period, an additional single failure problem was identified in that the EECM coolant for both H
analyzers for each unit, discharge into a common discharge header and a failure of
one check valve could result in subsequent failure of both analyzers for that particular unit.
The NRC inspector reviewed Browns Ferry Nuclear Plant Design Criteria, BFN-50-729, Revision 0, Single Failure Criteria for Fluid and Electrical Safety-Related Systems, dated June 12, 1987, (B45 870612 258).
Although this current TVA design criteria document appears to satisfy the intent of 10 CFR 50, Appendix A, General Design Criteria, the licensee document has been in effect for less than two years.
Based on discussions held with licensee personnel, the NRC inspector determined that no licensee formal documented design criteria existed prior to June 12, 1987.
This item will remain open pending further review to determine adequacy of the licensee's program for identification and correction of single failure deficiencies.
This item must be resolved prior to restart of each respective unit.
(CLOSED) Deviation 260/85-36-04:
Torus Flood Level Switches This deviation concerned lack of seismic qualification for water level switches used to detect a flooding condition in the low lying safety related areas of the plant.
The seismic requirements are documented in FSAR Section 10. 11.5. l.
The licensee's corrective actions specified in its response were as follows:
Perform a design change to provide seismic mounting requirements for the Reactor Building and RHRSW pump room level switches, Install the level switches to meet the new design requirements, Develop a periodic test instruction to veri.fy the proper functioning of the components, Place the periodic test on the PM schedule to ensure future testing.
The licensee response also stated that all switches had been tested.
The inspector reviewed Design Change Request BF-DCR-D3211Rl, dated April 30, 1986, and the associated ECN P5357, and seismic calculation CD-02077-871705R1, and associated document changes to ensure that the documented actions were performed and completed.
No concerns were identified.
Field observation by the NRC inspector of 6 level switches was performed on April 14, 1989, to verify that actual installation met design.
No deficiencies were identified.
EMI-90 "Reactor Building and RHRSW Pump Rooms Flood Level Switches Functional Test,"
was reviewed and found adequate for RHRSW and U-2 LS-77-25 A,B,C,D,E,F.
Completed test data for U-2 switches from September 27, 1988, were reviewed and found to be satisfactory.
The
0'
NRC inspector reviewed the licensee's PM program and verified that EMI-90 (Listed on the PM schedule as RO 1393, 1394, 1395, 1396, 1397, and 1398, for U-2 Level Switches)
was scheduled to be performed once per cycle with the next due date being cycle 5.
RHRSW Pump Room Level Switches O-LS-23-76A, B; -77 A, B, and -79 A, B, were also visually inspected for proper mounting.
No deficiencies were identified.
The PM schedule provides for future testing of the RHRSW Pump House switches also.
Deviation 85-36-04 is considered closed for Unit 2.
(CLOSED) Violation 50-259, 260, 296/86-32-01, SI 4. 7. E. 5, CREV Flow Rate Test Inadequacies This violation was concerned only with the data gathering instrumentation and techniques of SI 4.7. E. 5.
Subsequent to this violation a
number of issues regarding overall adequacy of the CREV system design, proper system alignments, and other items have been identified.
(See IE Report 50-259, 260, 296/87-14).
This item does not address those issues.
The inspector reviewed ANSI N510-1975 and the American Conference of Governmental Industrial Hygenists (ACGIH) Industrial Ventilation Documents to verify that acceptability of the licensee proposed test equipment was addressed for existing field conditions.
The NRC inspector found that the ACGIH document did address the use of a pitot-tube to measure air flow in ducting with total flow rates less than 1000 FPM.
The reason for the 1000 FPM limitation is that the pressure differentials sensed by a pitot tube are very small at low flows and are difficult to measure with great accuracy.
The licensee changed SI 4.7.E.5 in 1986 to specify the use of the more accurate instruments.
In the same change the SI was modified to, require 16 transverse velocity measurement data points.
The previous test specified only 12 points which did not meet the recommendations of Section 9 of the ACGIH.
The inspector found that sufficient test points, three penetrations in each train, were located in the duct work to support testing and confirmed that the geometry was such that they could not be located 7.5 duct diameters downstream of any airflow disturbances.
The NRC inspector found that the data gathering techniques and the t4DE used in SI 4. 7. E. 5 for measuring air flow rates at velocities of less than 1000 FPM meet the requirements of ANSI N510-1975 and therefore considers Violation 50/259, 260, 296/86-32-01 to be closed.
(OPEN) Violation 260/84-34-03, Core Spray Relief Valves This violation involved the failure to test the core spray system relief valves per ASME Code Subsection IWV-3510 requirements.
The relief capacity is used by the licensee as a basis for setting isolation valve leakage limits between high and low pressure piping
to assure the low pressure piping is not overpressurized.
In its response letter to the violation TVA stated that Surveillance Instruction, SI 3.2,
"Inservice Valve Testing required by ASME Section NI,"
was revised to include the core spray valves in question, 2-75-543A and 2-75-543B.
This violation was part of a civil penalty (Enforcement Action 84-108)
for the core spray overpressurization event for Unit 1.
This item was also previously addressed for closeout in NRC IE Report 259, 260, 296/86-40.
The item was not closed at that time because the relief valves on all
Units had not been tested although test procedures wer e in place.
The inspector reviewed SI 3.2.9, and verified that the current revision (REV 1) still contained test requirements for the CS relief valves.
Completed SI 3.2.9 data sheets documenting satisfactory completion of the ASME testing were reviewed.
The data package was dated July 2,
1987 for the Unit 2 valves.
Minor administrative concerns with the data were discussed with appropriate licensee personnel.
The licensee was unable to provide completed test data sheets for the performance of SI 3.2.9 for relief valves on units one and three (a Unit j. valve had been set point tested in 1984 on MR 307977)..
The inspector determined that cur rent SI tracking systems and control of system operability status provides sufficient assurance that required testing of U-1 and U-3 core spray relief valves wi 11 be performed prior to placing those components into service.
The inspector observed
.the Unit 2 and Unit 1 relief valves in the field to ensure proper reinstallation and noted that valve 2-75-543A had two quality tags affixed to it.
The tags were for CA(R 88-07-69 which documented licensee identified concerns regarding relief valve flow capacity calculations.
The issue in the CA(R implies that the relief valves are inadequately sized to protect the core spray low pressure piping from over.-pressurization that could result from leakage past the normally closed valves that isolate the low pressure piping from reactor pressure.
The licensee is reviewing the design criteria considering actual isolation valve leakage rates compared to the postulated rates, as well as discussing the design criter ia with GE.
The resolution of the CARR is required by the licensee to be complete prior to the restart.of Unit 2.
The inspector found the testing of the relief valves to have been completed satisfactorily but considers the relief valve sizing issue so closely related to the original concerns regarding overpressure protection that this item will remain open pending review of the completed CARR.
(OPEN) Violation 50-259, 260, 296/87-14-02:
CREV Train B Inop.
This violation involved the CREV system being found to be inoperable because of the system air flows being less than specified in TS 3. The licensee responded to the violation in its letter dated July 20, 1987.
The low flow was attributed to an erroneous surveillance test method.
In that response a statement was made that an evaluation of the Control Room Emergency Ventilation System (CREVS)
would be performed.
It is from that evaluation and other information that significant issues remain open regarding the CREVS.
The discussion of this violation and related issues consists of the following topics:
(1)
Review of Corrective Action For Example 2 of the Violation Example 2 of the violation was admitted by the licensee in it'
response and was attributed to improper use of newly specified test equipment.
Corrective steps taken included re-performing the test on the systems previously tested, training of personnel and revision of instructions on the use of manometers.
The NRC inspector reviewed the response and found the discussion regarding example 2 to be adequate.
The test equipment involved was a micro-manometer, a device which integrates a micrometer and
"hook gage" manometer along with a low voltage circuit to precisely measure water column changes due to minute pressure differentials sensed by a pitot tube.
Previous instructions gave a
general description on the use of pitot tubes and manometers.
The NRC inspector reviewed Mechanical Results Instruction MRI-27, "Pitot Tube-Manometer Training,"
and noted that it had not been updated to reference the new test equipment, but instead a
new procedure MRI-27A, "Microtector Training" was developed.
The NRC inspector found that MRI-27A satisfactorily implemented vendor manual instructions on the test equipment as well as lessons learned from events surrounding this violation.
Interviews with testing personnel revealed an awareness of the use and sensitivity of this new equipment.
Subsequent use of this M & TE has not resulted in any further documented problems.
Circumstances surrounding the return to operability and subsequent release for test of the redundant systems was reviewed and found adequate.
Example
of this technical specification violation is considered closed.
(2)
Review of Corrective Action For Example 1 of the Violation Example one of the violation was admitted by the licensee in it's response and was attributed to inaccurate testing techniques.
Part of this issue was the accuracy of measuring techniques for low flows in the CREV system.
This was addressed in the follow-up on Violation 50-259, 260, 296/86-32-01 which is closed in paragraph 5. d. of this repor In addition, the licensee corrective action included:
Surveillance Instruction (SI) 4. 7. E was to be revised to specify new test equipment.
A special test would be performed to verify the adequacy of SI 4.7.E.5 to correctly set the damper positions required to support the post-accident ventilation valve alignment of the CREVS.
The Restart Test Program would perform a test to prove capability of the CREV system to pressurize the control room.
The inspector reviewed the corrective actions and had the following comments:
SI 4.7.E.5.A and B, "Control Room Emergency Ventilation System Flow Rate Tests" were reviewed and were found to adequately address new test equipment requirements and damper/fan alignments.
This surveillance ensures confirmation of the required design flow rate (2 10K) as specified in TS 3.7.E.2.C.
The inspector had no further comments on SI 4.7.E.G.A and B.
The inspector reviewed Special Test (ST)
8726 which was designed to gather flow-rate data for the CREYS units for various line-ups/fan configurations of the control bay ventilation system.
The inspector found that ST 8726 met the basic intended purpose of the effort.
The results showed that while the flow rate through the CREVS varied with fan alignments, it was able to maintain flow rates within TS limits for all but one configuration.
CREVS Train "A" was initially set at 548 CFM (after SI 4.7.E.5.A)
and test recommendations now require the flow rate in the SI to be set at as close to 500 CFM as possible, therefore, a net change of 48 CFM would be expected.
Applying this 48 CFM difference to the test configuration in alignment 1 of the special test, which resulted in a flow of 617 CFM, results in a flow rate of 569 CFM.
This flow rate is in excess of the TS limit of 450-550 CFM permitted by SI 4.7.E.S.A.
This was not addressed in the discussion of results in ST 8726.
The importance of flow rates through the charcoal filtration units of the CREV system has to do with the residence time of methyl-iodide and the associated heat loads that are generated.
FSAR section 10 states that charcoal tray configuration of the filter trains will
permit processing of flows considerably in excess of the 500 CFM system capacity.
This issue is open pending the licensee's evaluation of the effect of flowrates found in ST 8726 for fan alignments that permit CREVS flows of approximately 550 CFM.
The NRC inspector reviewed ST 8731 which was designed to verify that the design differential pressure between the control room and the outside atmosphere can be established and maintained by CREYS under postulated worst-case conditions.
The test demonstrated that CREVS could pressurize the control room to 0. 125 inches H 0 greater than outside air pressure.
All test acceptance criteria were met.
However, the licensee identified that unfiltered air was leaking into the CREVS suction downstream of the filters from the board room supply system.
This unfiltered bypass flow, since it does not pass thru the filtration system, could present a
hazard to operators if it is laden with radionuclides or toxic chemicals.
This issue must be resolved before this item can be closed and prior to startup.
(3)
Conclusions and Criteria for Closing Example
From the preceding discussion it was obvious that several issues remain open with regard to the CREVS which need to be addressed prior to system operability.
The NRC inspector noted that the CREVS system in its current state may not meet its design intent, i.e.
to provide an isolated control room envelope, the integrity of which permits only out-leakage of the filtered air provided by the CREV system.
The basis for this position is the inability of ST 8731 to detect or quantify bypass leakage around the CREV system charcoal filters.
The system could be said to meet the Technical.Specification 3. 7. E because that section assumes the control room integrity was previously established and therefore states only the requirements to assure that the pressurization system components are maintained operable.
This position does not meet the intent of the Technical Specification.
In the inspection of this issue, the inspector could not establish that any analysis existed that addressed the ability of the control room envelope to protect operations personnel from the hazards of a toxic chemical spill on transportation routes near the site.
In a related issue the resident office
determined from licensee supplied documents that significant shipments of toxic chemicals occur frequently in barge traffic near the 'site.
(One barge at 1200 tons per barge every 2 weeks for one type and 3 barges per 24 days for another.)
Toxic chemical shipments are addressed in Regulatory Guide 1.78 - 1974
.=..
"Assumptions For Evaluating The Habitability Of A Nuclear Plant Control Room During A Postulated Hazardous Chemical Release."
The licensee had previously evaluated the Chlorine issue in its response to NUREG 0737 III.D.3.4.
The NRC inspector reviewed the licensee response and NRC acceptance.
In subsequent reviews of the issue of hazardous chemicals the inspector and TVA determined that river traffic included hazardous chemicals other than chlorine.
This fact was addressed in NRC letter to TVA dated October 21, 1988.
That letter addressed the fact that toxic chemicals in addition to ch'lorine are shipped near the plant, and that these shipments were large and at a frequency of
barges per year for a single type of hazardous chemical listed in table C-1 of Reg.Guide 1.78.
Total barge shipment of hazardous chemicals exceeded 50 per year.
The NRC is awaiting a TVA.response.
This issue of the bypass flow leakage and the CREVS capability to protect the control room personnel from potential hazardous chemical spills will remain open.
g.
(OPEN) IFI 259,260,296/89-11-03 The inspectors continued to follow the licensee activities associated with identification and replacement of cables that are subject to premature deterioration of conductor insulation.
The licensee notified the resident inspectors of this condition on March 17, 1989, and it is documented under CA(Rs BFP 890290, BFP 890291, and BFP 890292.
Since the last reporting period, the licensee has completed the replacement of this type of cable in all Unit 2 Control Room NI panels.
Additionally, GE has completed their review of records, and has identified to TVA all known applications of this type 'cable.
In addition to usage as power cables in NI panels, APRM panels, main steam line radiation monitor panels, and various process recorders located in the control room, the cables are also used in similar applications in recorders and other instrumentation located in the radwaste control room and the remote shutdown panels.
The licensee is continuing with their evaluation and determination of corrective actions for this condition.
The-NRC inspectors will continue to monitor licensee activities in this area.
This item must be resolved prior to restart of each respective unit.
No violations or deviations were identified in the area of Licensee Action on Previous Enforcement Matter,
8.
Design Control (37700)
As a result of a recent EA audit (BFT 89901) significant problems were found with the implementation of the design control process.
This audit focused on looking at the final product.
Four recently closed ECNs (P3118, P7043, P3092, and P3098)
and one civil DCN (M1347) were reviewed.
The systems covered were the CS, RHR, and RMCU systems.
Four CAgRs were issued by EA and three by the site quality assurance organization.
The Site Director instructed that no further ECNs be issued until a quality product could be obtained.
Discussions were held on May ll, 1989, with the EA Audit team leader and other TVA managers regarding the audit findings.
The audit found that the program and procedures were adequate, but had not been implemented effectively for the items reviewed.
For example, one of the findings was that the transitional design change control program did not ensure that the design criteria, system requirements calculations and the modification criteria reflected the actual plant configuration.
One of the concerns with ECN P3118 was that a diesel generator loading calculation was issued but the updated load information was not issued.
The licensee's corrective action was still being developed at the close of the inspection period.
Part of the action plan will be to review over 300 ECNs for five systems using
people for several weeks.
This will provide a larger sample size and better indication of the magnitude of the problem.
System operability must be considered in determining the necessary corrective action.
An NRC design change inspection is planned for May 22-26, 1989.
The design change control process will receive further assessment during this inspection.
Followup of Open Inspection Items (92701)
(OPEN)
IFI-50-259, 260, 296/86-05-07:
Reactor Building Radiation Isolation Monitor This issue addressed an inspector concern regarding the length of time a radiation monitor instrument channel could be left inoperable during a
surveillance before the affected channel had to be tripped or declared inoperable.
There appeared to be contradictory notes in the technical specifications.
Initially, TVA proposed a technical specification change to rearrange and clarify notes of Table 4. 2.A and 3. 2.A to resolve this issue.
Following discussions between the NRC and the licensee it was agreed that the best course of action was to change the SI and make it more specific concerning the testing and operability of the radiation monitors rather than change the technicial specifications.
The NRC inspector reviewed the following surveillance instructions that were changed to clarify the difference between time allowed for correction
of deficiencies and time allowed for verification of operability.
The SI's reviewed were:
U-2 SI-4.2. A-10FT Reactor Building Ventilation Radiation Monitors RM-90"140, 141, 142, 143 Instrument Functional Test, Rev 0.
U-2 SI"4.2.A. 10 Reactor Building and Refuel Floor Ventilation Radiation Monitor Calibration and Functional Test, Rev 6.
Both of the SI's placed the trip system in the tripped condition at the beginning of the test.
This initiates the action required by the technical specifications for instrumentation which exceeded any of its action statements or surveillance grace periods.
This conservative approach eliminates any concern over time periods, however, it does introduce other technical specification problems.
Specifically, when the reactor building rad monitor trip system is tripped for either surveillance, a
Group VI isolation occurs, (along with other system realignments).
Among the Group VI valves which are shut, are both the Division
and Division 2 isolation valves for the drywell and torus hydrogen analyzers.
With both analyzers isolated, T. S.
3. 7. H. 3 limiting condition applies and a
shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> must occur.
No provisions for alternate sampling is specified in the T. S.
or T.S.
interpretations.
The plant operating requirements. section of OI-76 states one of the two Hydrogen analyzer systems will continuously monitor the drywell and torus.
Furthermore OI-76 prohibits the use of the isolation bypass switches for surveillances.
The SI does not address the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> limitation or the associated operational restriction.
The licensee needs to thoroughly evaluate the implications of performing the affected SI under all operating conditions.
The inspector considers the changes to the SIs as incomplete in that they do not clarify how long a
RB radiation monitor is allowed by Technical Specification to remain inoperable for surveillances.
During research for this issue the inspector discovered the following technical specification and surveillance instruction deficiencies that are related to'this issue:
T.S.
Table 3.2.A -- Unit 2 Notes (2), (5) and (13) are not attached to any item in the table Paragraph 6. 2 of SI 4. 2.A. -10FT, which describes acceptance criteria is not accurate.
SI 4.2.A Data Tables 7-140B, 141B, 142B, 143B should denote the value of the downscale trip as an acceptance criteria since the downscale trip is a technical specification functio ~
~
These items were discussed with licensee personnel on May 10, 1989.
This IFI will remain open pending further licensee action and must be resolved prior to Unit 2 restart.
No violations or deviations were identified during the Followup of Open Inspection Items.
9.
Restart Test Program (99300B)
During this inspection the test procedures and test results for 2-BFN-RTP-031,
"Control Bay HVAC,"
and 2-BFN-RTP-099,
"Reactor Protection,"
were reviewed.
ao RTP Test of System 031, Control Bay Heating Ventilation and Air Conditioning (HVAC).
The control bay HVAC consists of numerous ventilation motors, air conditioning units, cooling water, chilled water, and heating systems that cover all three units with certain equipment in the system being important to safety.
This also included new air handling units for the shutdown board rooms "C" and "D."
As part of this review the NRC inspector utilized the following drawings:
0-47E610-90" 2 2-47E610-90" 2 0-45E769-7 thru ll 1 and 2-730E927-18 0-47E865"4 2"47E2865-2 0 and 3-47E961-6 47E866-3, 5 and
Units 0, 1, 8 3, Mechanical Control Diagrams Radiation Monitoring System Unit
8 0,
Mechanical Control Diagram Radiation Monitoring System Unit 0, Miring Diagram 480V Common Auxiliary Power Schematic Diagram Primary Containment Isolation System Flow Diagram Ventilation and Air Conditioning Air Flow Flow Diagram Cooling Air Flow and Air Conditioning System Mechanical Heating Ventilation and Air Conditioning Controls Flow Diagram Heating and Air Conditioning Hot and Chilled Mater 3-45E779-20 1,
2 and 3-47E844-2 Miring Diagram 480V Shutdown Auxiliary Power Schematic Diagram Flow Diagram Raw Cooling Mater
45N620-8 Wiring Diagrams Annunciator System" 'Key Diagram The following were utilized in conducting this review:
The Baseline Test Requirement Document BFN-BTRD-014, Rev. 4; System Test Specification 2-BFN-STS-031,
"Control Bay Air Conditioning";
RTP Test Procedure 2-BFN-RTP-031A, "Control Bay Heating, Ventilating and Air Conditioning System";
and RTP Test Procedure 2-BFN-RTP-031B, "Control Bay HVAC."
The following are the specific reviews of the four items:
Baseline Test Requirement Document The BTRD listed nine modes of system 31.
Four modes selected for review were:
Mode Definition 031-01-M-S 031".02" M-S Isolate normal supply paths and provide fresh filtered air to Control Bay.
Pr ovide ventil ation to Reactor Building Board Rooms and Control Bay Mechanical Equipment Rooms.
031"05-M"S
, Provide control rooms.
recir cul ation air handl ing to rooms and auxiliary instrument 031-06-M-S Provide ventilation to battery rooms.
The BTRD listed a total of 14 required tests.
These selected for review were:
Mode Test 031-Ol-M-S 031-01-M"S Demonstrate upon receipt of any of the isolation signals or loss of control air, that Unit 1 and 2 dual Control Room and Unit 3 Control Room isolation dampers close.
Demonstrate the capability of each emergency pressurization Unit "A" or "B" to deliver adequate air flow, treat the air with HEPA and charcoal filtration and establish a
positive 0. 125 inch of water gauge pressure in the Unit 1 and 2 dual Control Room and the Unit 3 Control Room after an isolation signal is received and/or power supply load shed occur Ol-M"S and 031"02-M-S Demonstrate how treated outdoor air is supplied to the control room air pressurization units A and B;
how air is supplied to, and exhausted from the Unit 1 and 2 and Unit 3 Elevation 593'echanical Equipment Rooms; supplied to the Unit 1 and
and Unit 3 Elevation 593'ir handling units; and supplied to and exhausted from the Unit 1 and 3 Shutdown Board Rooms.
031-05-M-S Demonstrate the capability of the air handling units to circulate the required conditioned air to Unit 1 and 2 dual Control Room, Unit 3 Control Room at Elevation 617',
the Unit 1, 2 and 3 Auxiliary Instrument and associated rooms at Elevation 593'.
031-06-M-S Demonstrate the capability of Unit 1, 2 and 3 Battery and Board Room exhaust systems to remove sufficient air and potential hydrogen accumulation from the Elevation 593'attery Rooms.
The inspector noted that Table 6. 1 of the BTRD provides a direct r'eference for various system modes to a particular attachment.
These attachments are test scoping documents.
RTP System Test Specification 2-BFN-STS-031, Control Bay Air Conditioning The NRC inspector noted that two revisions of this system test specification (STS) were made as of this reporting period.
The revisions involved a
change in the Unit 1, 2 and 3 control room pressure from
.25 inches to
. 125 inches water column, the addition of a required test from BFN-BTRD-010, Raw Cooling Mater
~ System and a
change to BFN-BTRD-014.
Section 5 of the STS outlines test requirements and the NRC inspector noted that all the test requirements listed in the BTRD's are also listed in the STS under subsection 5.4.
In addition to the BTRD requirements this section also indicated additional testing required by ECN's, vendor recommendation, employee concerns, etc.
and the NRC inspector noted that a special test involving the CREVs would be performed to satisfy an NRC commitment tracked by the licensee as number NC00870213002.
This special test referenced SI 4. 7. E and a
demonstration that proves capability of CREVs to pressurize the control rooms.
The NRC inspector noted that based on the amount of testing activities and the type of system being tested, the licensee divided the test procedure into two parts RTP-31A and 31 RTP Procedure 2-BFN-RTP-031A, Contro1.. Bay Heating Ventilating and Air Conditioning The RTP procedure test results were reviewed by the inspector to determine the acceptability of the method of testing, control of testing activities, compliance with test specifications, and documentation of test exceptions.
This determination consisted of a review of four of the modes and applicable testing as indicated in the BTRD and the STS for system 031.
Due to the fact that the actual testing was performed as two tests, overlapping between the tests existed.
The overall purpose of RTP-31A was to test and verify the operation of chillers, temperature controls, chilled water pumps, coolers and associated cooling water systems such as Raw Cooling Mater, System 23, i.e.
testing involved with the water side of the system.
The inspector reviewed the four selected modes and the specific tests of section 5.0 of the RTP procedures which verified app 1 icab 1 e porti ons of the modes.
Modes 031-01-M-S and 031-02-M-S were tested in sections 5.1 and 5.3.
Section 5.1 required the performance of SI-4.2.G-2, which was performed in May, 1988 in preparation for the Loss of Offsite Power/Loss of Coolant Accident Testing.
The NRC inspector noted that the SI verified that the. "A" and "B" CREVs would start and the combined Unit 1 and 2 control room and Unit 3 control room outside ventilation would isolate; the control rooms would go to a
recirculation air operation; and that filtered air would be supplied to the Control Rooms.
The inspector noted that this SI did not document the pressurization of the control rooms.
Section 5.3 verified the operability and interlock functions of various control bay and shutdown board room chillers.
This verification consisted of.documenting the amount of RCM flow through the chillers, the interlocks between the chilled water pumps and the chillers, that temperature control'alves respond and that check valves are functional.
Mode 031-05-M-S required the most extensive testing in that five sections of the RTP procedure were performed to satisfy the mode:
Section 5.4; verified the operability of the Unit 3 control bay chillers from their respective control panels and demonstrated the i'nter locks between the chiller and chilled water pumps; Section 5.5 verified the chilled water flow rates delivered from the chilled water pump and flow rates to each air handling unit (AHU) for the applicable chilled water system; Section 5.6 verified the performance of the chilled water pumps; Section 5. 7 verified that the automatic pump switch over functions operate for the chilled water circulating pumps and the chillers; and Section 5.8 verified the performance of the shutdown board rooms emergency cooling syste A total of nine Test Exceptions were identified during the performance of this test and TEs
and
were still outstanding during the review.
TE 08 involves the performance of procedure EMI-60, "Inspection and Preventive Maintenance of Control Bay Chillers,"
and was not completed.
TE 07 involves the pump performance of Unit 1 and 2 control bay chilled water pumps in that they did not meet the acceptance criteria.
RTP Procedure 2-BFN-RTP-031B, Control Bay HVAC The RTP procedure and test results were reviewed by the inspector to determine the acceptability of the method of testing, control of testing activities, compliance with test specifications and documentation of test exceptions.
This determination consisted of a review of four of the modes and applicable testing as indicated in the BTRD and the STS for system 031.
Due to the fact that the actual testing was performed as two tests, overlapping between the tests existed.
The over all purposed of RTP-31B was to test and verify the operation of dampers, fans, air filters, air flow rates and temperature controls, i.e. testing involved with the air side of the system.
The inspector reviewed the four selected modes and the specific tests of Section 5.0 of the RTP procedure which verified applicable portions of the modes.
Modes 031-01-M-S and 031-02-M-S were tested in eight parts of Section 5.0.
Section 5. 1 verified the control pressurization and leakage by requiring that special test 8731,
"Control Bay Pressurization and Leak Rate Test,"
be performed.
Section 5. 2 verified control room isolation by requiring that SI-4.2.G-2,
"Control Room Isolation and Pressurization,"
be performed.
Section 5. 3 verified the control room emergency ventilation functional performance by requiring that a total of eleven SI-4.7.E series of SI be performed.
Section 5.4 verified that the control bay isolation would occur on loss of control air.
Section 5.7 verified the operation of supply and exhaust fans and Section 5.8 verified the functional operation of the new air conditioning units installed for shutdown broad rooms C and D which are located above room C
by the performance of Post Modification Test (PMT)-161,
"Shutdown Board Rooms HVAC System," for Engineering Change Notice P0956.
Further review of Section 5. 1 indicated that the Special Test ST8731,
"Control Bay Pressurization and Leak Rate Test,"
was successfully performed on August 25, 1988.
However, additional information supplied by the licensee indicated that the pressurization portion of the test was due to excessive leakage in the air ducting which is presently routed through the habitability zone.
This excessive leakage is being addressed as a separate issue.
See paragraph S.f of this repor b.
RTP Test of System"'099, Reactor Protection As part of this review, the NRC inspector utilized the following drawings:
45W641"4 45W671-25 thru 46 791E245-1 thru 3 791E247-1B thru
2-730E927-7 and
RPS Schematic Analog Trip Units RPS Trip System A Cab RPS Trip System B Cab Primary Containment Isolation 2-730E321-3, 5 and
Reactor Manual Control System 2-730E915 Shl thru Sh20 Reactor Protective System Elem.
Baseline Test Requirement Document, BFN-BTRD-045, Rev.
2; System Test Specification, 2-BFN-STS-099, Rev.
1; and RTP Test Procedure 2-BFN-RTP-099, were reviewed as follows:
Baseline Test Requirement Document The BTRD liste'd seven modes of system 99.
Four selected for review were:
Mode 099-01"I-S 099-02-I-S 099-03"I"S 099-05"I"S Definition Provide Automatic Scram and SDV Vent and Drain Valve Isolation signals to the Control Rod Drive System (CRD)
Provide Manual Scram signal to the CRD Provide
"RUN" mode signal to PCIS for low Steam Line Pressure Isolation permissive Provide trip signal to Reactor Recirculation Pump motor breakers The following tests for the four selected modes were reviewed:
099"01"I-S 099-02-I-S Demonstrate the capability of the RPS to provide a
scram signal to the CRD system upon receipt of trip signal inputs Demonstrate the capability of the RPS to provide a
manual scram signal to the CRD system.
The manual signal is initiated
,
I t /
~
$
~
'
~
either from the control room or from outside the Control Room 099"03"I"S 099-05-I-S Demonstrate that the Reactor Mode Switch will provide an open contact to the Primary Containment Isolation System in the RUN posi tlon Demonstrate that the RPS provides Recirculation Pump Trip (RPT) signals to the RTP switch gear logic The inspector noted that each test referenced a test scoping document which was an attachment to the BTRD.
RTP System Test Specification 2-BFN-STS-099, Reactor Protection System Section 5 of the STS outlined the test requirements and all the test requirements listed for the four selected modes in the BTRD were also listed in the STS under Section 5. 4.
In addition to the BTRD test requirements, this section listed under additional test requirements relay/contactor voltage pick-up and drop out tests.
The NRC inspector noted that mode 099-Ol-I-S required the greatest amount of testing and interface with other systems.
RTP Procedure 2-BFN-RTP-099, Reactor Protection System The RTP procedure and test results were reviewed by the NRC inspector to determine the acceptability of the method of testing, control of testing activities, compliance with test specifications and documentation of test exceptions.
This determination consisted of a review of four of the modes and applicable testing as indicated in the BTRD and the STS for System 099.
The inspector reviewed the four selected modes and the specific tests of Section 5. 0 of the RTP procedure which verified applicable portions of the modes.
Mode 099-01-I-S was tested in 21 of the 30 subsections of Section 5 of the procedures.
These sections included Section 5. 1 which verified mode switch in shutdown functions; Section 5.3 verified the high reactor vessel pressure automatic scram; Section 5.6 thru 5.8 verified the high water level in the scram discharge volume tank automatic scram and other automatic scrams; Section 5.9 verified the RPS turbine first stage pressure permissive interlock; and Section 5. 10 thru 5. 14 verified additional automatic scram functions.
Mode 099-02-I-S was tested in Section 5. 1, 5.2 and 5.25; Section 5.2 verified the manual scram from the control room; and Section 5.25 verified scram of RPS from outside the control room.
Mode 099-03-I-S was tested in Section 5.27 which verified the proper
i
operation of the reactor mode switch contacts.
Mode 099-05-I-S was tested in Section 5. 18 which was to verify the recirculation pump trip circuit; however, TE-07 was written indicating that this verification would be performed in procedure 2-BFN-RTP-068, Reactor Recirculation System and credit would be taken at that time.
C.
The inspector has found that the BFN RTP contained the essential elements needed to achieve its intended goals and objectives as outlined in the NPP Volume III.
However, the inspector is concerned that test exceptions may not be adequately documented in all cases or tracked as open hardware issues.
This item will be reviewed in future NRC inspections.
The NRC inspector also reviewed the status of the Restart Test Program.
The program is estimated at 90K complete with a scheduled completion date of August 25, 1989.
Thirty-four of the forty-three systems being tested have been completed.
Six tests are in progress and three tests have not begun.
Fifty-two test exceptions are open and need to be resolved.
The licensee is planning to transfer several test exceptions to the power ascension program.
Each of these items will be reviewed by the NRC inspector to determine the acceptability of transferring the exceptions.
Presently there are six engineers and one supervisor remaining in the restart test group.
The group will be eliminated once the remaining testing is complete.
Power Ascension Testing The NRC inspector reviewed the licensee's plans for power ascension testing.
The program is comparable to near term operating plants and to facilities with recent long outages.
The test program includes plant shutdown from outside the control room and loss of offsite
, power tests.
Thirteen power ascension test procedures are being developed along with the review of sixteen refueling. test instructions.
The overall test instruction procedure is Plant Manager's Instruction PMI-26. 1, Refueling Test Program.
The power ascension test program is being performed under the cognizance of the technical support superintendent and reactor engineering group supervisor.
This is a departure from the original intention of maintaining the restart test group to perform the power ascension testing.
The controls used to perform the restart test program appeared to be adequate for the program, however the controls for the power ascension test program have not been proven.
10.
Site Management and Organization (36301, 36800, 40700)
Ouring the reporting period, a discussion was held with BFN managers concerning the restart test program.
The NRC inspectors were informed that the testing program was 90K complete and that arrangements were being made to close the program and shift the remaining work to the plant operational programs.
This planned activity will involve cancelling Site
Directors Standard Practices (SDSP)
12. 1, Restart Test Program and 12.2, Development of System Test Specifications and using Plant Managers Instruction (PMI) 17. 1, Conduct of Testing, to complete the program.
The program has had the high visibility that it deserves and this change will make it a less visible program i.e. going from the site director s level to the plant manager's level.
Furthermore, the RTP is a volume III program, has been reviewed by the NRC, has been continuously monitored, and the NRC has issued a
SER documenting the adequacy of the program.
The operational readiness of which PMI 17. 1 is a part, has not been reviewed by the NRC, has not been monitored continuously and no SER documents the adequacy of the operational readiness program.
The current structure of the Restart Test Program is such that it can remain in place as a program until BFN Unit 2 commences commercial operation and can also remain intact if Unit 1 or 3 are to be brought back on line.
The NRC inspector has discussed this item with BFN TVA senior managers as well as NRC, ADSP managers and coordinators.
The restaffing of selected middle management is continuing.
The Maintenance Manager has been named from within the BFN staff.
The new management at BFN that has been phased in over the last six months is starting to function as a team.
This is evidenced by how BFN deals with issues.
The NRC staff has observed a more cooperative, and conservative and realistic attitude from BFN staff.
Ex& Interview (30703)
The inspection scope and findings were summarized on May 15, 1989 with those persons indicated in paragraph 1 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 259, 260, 296/89"19-01 Acronyms Descri tion Unresolved Item.
Possible Failure to Control Work Activities Involving Systems Important to Safety (Paragraph 3).
ACGIH ADSP AHU ANSI ASME ATU BFNP BTRD American Conference of Governmental Industrial Hygrenists Associate Directorate f'r Special Projects Air Handling Unit American National Standards Institute American Society of Mechanical Engineers Analog Trip Units Browns Ferry Nuclear Power Plant Baseline Test Requirements Document
Cl
CAQR CRD CREVS CS CSSC DCN DBVP EA ECCS ECN EECW EMI FSAR GE HEPA HVAC IE IFI IR LRED MRI MR M8(TE NI NPP.
NRC NRR OI PCIS PM PMI PMT QA RCW RG RHR RHRSW RPS RPT RTP RWCU SDSP SDV SER SI STS TE TS TVA URI ystem mponents gram ning ion Maintenance Request Measuring and Test Equipment Nuclear Instrumentation Nuclear Performance Plan Nuclear Regulatory Commission Nuclear Reactor Regulation Operating Instruction Primary Containment Isolation System Preventive Maintenance Plant Manager Instruction Post Maintenance/Modification Test Quality Assurance Raw Cooling Water Regulatory Guide Residual Heat Removal Residual Heat Removal Service Water Reactor Protection System Recirculation Pump Trip Restart Test Program Reactor Water Cleanup Site Director Standard Practice Scram Discharge Volume Safety Evaluation Report Surveillance Instruction System Test Specification Test Exception Technical Specifications Tennessee Valley Authority Unresolved Item Condition Adverse to Quality Report Control Rod Drive System Control Room Emergency Ventilation S
Core Spray Critical Structures, Systems, and Co Design Change Notice Design Baseline and Verification Pro Engineering Assurance Emergency Core Cooling Systems Engineering Change Notice Emergency Equipment Cooling Water Electrical Maintenance Instruction Final Safety Analysis Report General Electric High Efficiency Particulate Activity Heating, Ventilation, 8 Air'onditio Inspection and Enforcement Inspector Followup Item Inspection Report Licensee Reportable Event Determinat Mechanical Results Instruction
~~
0