IR 05000259/1988010

Insp Repts 50-259/88-10,50-260/88-10 & 50-296/88-10 on 880401-30.Violation Noted.Major Areas inspected:Q-list, Operational Safety,Maint Observation,Surveillance Testing Observation,Ros & Restart Test Program

ML20155J072
Person / Time
Site:	Browns Ferry
Issue date:	06/03/1988
From:	Ignatonis A, Paulk G NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML20155H976	List: IR 05000259/1988010 ML20155H973 ML20155J041
References
50-259-88-10, 50-260-88-10, 50-296-88-10, NUDOCS 8806200425
Download: ML20155J072 (24)

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. UP$1TED STATES

[r.n otop o NUCLEAR REGULATORY COMP.81SSION

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Report No /83-10, 50-260/S8-10, and 50-296/88-10 Licentee: Tennessee Valley Authority 6N 3SA Lookcut Place 1101 Market Street Chattanooga, TN 37402-2S01 Docket No , 50-260, and 50-296 License Nos. DPR-33, DPR-52, and OPR-68 Facility Name: Browns Ferry Nuclear Plant inspection at Browns Ferry Site near Oecatur, Alabama Inspection Conducted: April 1-30, 1933 Inspectors: [.aulk, Shor Resident f Inspettor _f_3_Jf f1 G. D te Signed Accompanied by: C. R. Brooks, Resident Inspector E. F. Christnot, Resident Inspector W. C. Bearden, Resident Inspector A. H. Johnson, Project Engineer i Approved by:

A. J.'Ig Stonis, dd 4 _f. .g 6 ection Chief

$ gg Di.e Gigned Inspection Progr ms, TVA Projects Division SUMMARY Scope: This routine inspection was in the areas of Q-list, operational safety, maintenance observation, surveillance testing observation, reportable occur-rences, restart test program, personal dosimetry, ,nd fuel reconstitution.

l Results: One violation was identified for fa'~ re to have an adequate l administrative procedure for controlling the eparation of licensing documents.

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t 8806200425 880603 PDR ADOCF 05000259 Q T:Cn

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REPORT DETAILS Licensee Employees Conta:ted

• J. G. Walker, Plant Manager *

J. D. Martin, Assistant to the Plant Manager

• R. M. McKeon, Operations Superintendent T. F. Ziegler, Superintendent - Maintenance D. F. Mins, Superintendent - Technical Services J. G. Turner, Manager - Site Quality Assurance M. J. May, Manager - Site Licensing

• J. A. Savage, Compliance Supervisor A. W. Sorrell, Site Radiological Control Superintendent R. M. Tuttle, Site Security Manager L. E. Retzer, Fire Protection Supervisor H. J. Kuhnert, Office of Nuclear Power Site Representative T. C. Valen7ano, Director - Restart Operations Center Other licensee employees contacted included licensed reactor operators, auxiliary operators, craftsmen, technicians, public safety officers, i quality assurance, design and engineering personne )

• Attended exit interview l l

Acronyms and abbreviations used throughout this report are listed in the last paragrap . Licensee Action on Previous Enforcement Matters (92702) l

(CLOSED) Violation (259/260/296/84-15-02), Control Air System Operating Instructions (0132) valve checklist in Operating Instruction 32/32A was deficient. Tne licensee supplemental response of October 24, 1936, was reviewed to assure corrective actions were addressed and implemente Although walkdowns were performed and "as-constructed" drawings were issued for system 32 (Control Air) Units 1, 2, and 3, it was determined that these corrective steps were insufficien As a result of the incorrect drawings, 01 32 was incorrectly update These discrepancies .

were due, in part, to the lack of a formal procedure being issued to provide for systenatic and controlled verifications of drawing correc-tion Since the initial walkdowns and procedure corrections on the control air system, the BFN Nuclear Performance Plan was initiated with Volume III,Section II.2.4., addressing Procedures Upgrade and Section III.2.2. showing plans for an improved Design Baseline Program. Both the Procedures Upgrade Program and Design Baseline Program were initiated and thus resulted in corrections to the control / station air and drywell control air systems drawings and procedures. Specifically, on March 12, 19S3, BFN Site Director's Standard Practice 9.1 was revised to implement i and control mechanical system walkdowns, as well as conform with the BFN 1

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f configuration mana.3ement progra Subsequently, the "common" (Unit 0)

portion of systera 32 ns walked down in accordance with this procedure Information obtained during this walkdown was then incorporated into configurstion control drawings. These drawings were rev wwed and approved [

by the walkdown group and than -evaluated by the plant for procedural- '

changes before restart. 01-3cA is now correc i Additionally, complete walkdowns for all unit's control / station air and drywell control air was scheduled to support Unit 2 restart. This ensuras i all unit crossties and interconnections are properly documented with i-procedural controls in olace for all portions of the system af fecting Unit 2. 01 32 was then corrected and reverified as correct by plant

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personnel and contains all system valves in the valve lineup checklists for Unit 2. This violation is therefore close l (CLOSED) Inspector Followup Item (296/85-09-03), Calibrations for Reactor [

Vessel Thermocouple This item relates to a problem encountered in January 1985, regarding the Unit 3 bottom reactor vessel drain temperature indic& tion in the control roo *

Indication errors were caused by an improper valve alignment in the Reactor Water Cleanup System (RWCU). The inspector expressed concerns about the calibration or verification of proper thermocouple readings during plant operation .

Due to its nature, a thermocouple cannot be calibrate There is no !

adjustment which could be made during a calibratio A calibration is ;

performed once per cycle en the temperature recorders which use the thermocouples as inputs. The procedures used to calibrate the temperature l recorders are Standard Calibration Instructions 501.1 and 501.2, These calibrations make the recorder function properl Surveillance Instructions 4.6. A.1, 4.6. A.2, 4.6. A.5, and 4.6. A.7 require temperature readings from reactor vessel thermocouples using tempercture recorders TR-56-4, TR-68-37, and Tit-68- TR-68-2 and TR-68-37 indicate upscale if the thermocouple fails. TR-56-4 will be replaced on all units with a new recorder which will indicate thermocouple failure by an upscale reading. This will be accomplished by ECN P062 Installation is complete on Unit 2 and Units 1 and 3 will be complete prior to startu l l

With procedures that keep the recorders calibrated and built in thermo-couple fail,ure indication on each recorder the licensee considers that adequate assurance of thermocouple accuracy is provided and that the plant will be operated within reactor vessel thermal limits during plant heatups, cooldowns, and recirculation pump start This item is considered close (CLOSED) Violation (260/87-37-01), Failure to Prevent Inadvertent Operation by Tagging Component The inspector identified a failure by the licensee to maintain an operational restriction on the electrical operation of RHR pump suction Valve 2-74-2 The vabe had been tagged

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under a maintenance hold order for an extended period of time prior to the

even Although the limit switches required readjustment prior to i

returning the valve to service, maintenance personnel released the hold ;

j order to allow system testing with the understanding that the valve was to i be manually operated and not suitable for electrical operation until the i limit switch readjustment was performed. This operational restriction was t l subsequently lost in the many turnovers that occurred over the course of a

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month and electrical power was eventually restored to the valve, resulting i in the valve haing operated against its close seat without limit switch l

protection ut...I the associated breaker tripped on overloa The valve

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was later determined to have not been damaged.

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The inspector reviewed the iicensee's response to the violation dated January 15, 1983, and documentation to support completion of the proposed corrective action ~

The inspector determined that Site Director Standard Pra.:tice (SDSP)-14.9,

Equipment Clearance Procedure had been revised to include an allowable method for temporary lifting of a hold order on a particular piece of

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equipment. The temporary lift is controlled and documented by use of form 50SP-217, and may not exceed one eight hour shift.

Additionally, the inspector reviewed BFN Unit 2 Superintendent memorandum

dated November 30, 1987 (R41 871130 945) to all operations personnel, j This memo consisted of operations critique 87-059 and covered the events j leading up to the event and the conclusions, lessons learned and correc-

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tive actions resulting from the event.

The inspector feels that the currective actions as implemented should be

adequate to prevent recurrence. This item is closed.

J (CLOSED) Violation (259/S0-47-03 and 296/80-41-03), Work Plan had no Final QA Review. The inspectors had identified a failure to document the performance of a quality assurance review for completion associated with

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the installation and testing of certain high density spent fuel storage racks into the Units 1 and 3 spent fuel pools. The modification work was j associated with work plan numbers 6371 and 7703.

$ ine inspector reviewed the response to the violation dated April 6,1981, in which the licensee committed to complete the required quality assurance reviews and to conduct necessary training for modifications personnel .

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The inspector reviewed completed work plan numbers 6371 and 7703 and verified that the documents contained the appropriate reviews by the res?onsible engineer section and QA section supervisor supervisor Additionally, the inspector determined that Browns Ferry Site Director Standard Practice (SDSP)-8.3, Plant Modifications and 50SP-17.2, Post j Modification Test Program were revised to clarify requirements for test

review and closecut. Discussions with supervisory personnel that were associated with the modifications group during the subject period revealed

, that the required training called for in the response was accomplished not through formal training classes with attendance records, but through

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i individual emphasis by managers. Subsequent to completion of the training  !

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the modification program has evolved through the RPIP effort and newer i

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formal training requirements now exis The inspector feels that the l corrective actions taken due to the original violation and the existence of newer upgraded Post Modification Training requirements form an adequate ,

basis for closure of this ite This item is close (CLOSED) Violation (259,260,296/85-53-02), Main Steam Line Radiation .

Source Check Inadequat The inspector identified that the licensee ,

failed to perform required routine instrument calibrations on the four -

channels of main steam high radiation instrumentation. The licensee had been performing routine surveillance 5 on the instrumentation channels which amounted to performing a source response check (no acceptance ,

criteria) rather than an instrument calibration as defined in technical specification The inspector also noted that the licensee failed to maintain records of actual applied voltage calculations and radiation monitor output indication The inspector reviewed the licensee's response to the violation dated January 6, 1986, and supplemental response dated March 20, 198 .

Additionally, the inspector reviewed new surveillance instruction  !

(SI)-4.1.B-10.1, Main Steam Line Radiation Monitor Source Calibration, l which contained sufficient instructions and detail to allow for per-formance of a channel calibrations using a known source field as required by T.S. Table 4.1.6 SI-4.1.B-10.1 includes comparison of test results to acceptance criteria, and data sheets for recording actual applied voltage calculations and radiation monitor output indications. The inspector ,

determined from conversations with licensee employees that the surveil- '

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lance instruction is PORC approved and has been walked down as part of the SI upgrade program but is not presently scheduled for actual performance.

The licensee's planning and scheduling section has the SI included on the

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list of outage SI, required prior to Unit z startup. The inspector feels that the above corrective actions are adequate to prevent recurrenc This item is close l Followup of Open Inspection Items (92701)

i (CLOSED) Unresolved Item (259/260/296/86-05-09), Basis for radiation monitor trip set point No justification could be located for the

, licensee's selection of 92 mr/hr as the trip setpoint for the Reactor Zone

, and Refuel Zone Radiation Monitors. Technical Specifications limit this setpoint to 100 mr/hr maximum; however, the potential +100*. inaccuracy was

not considered in establishment of the actual setpoin The licensee responded to this item in it's April 18, 1986 letter which included a justification for the instrument setpo 5t. Also in this letter, the

licensee stated that a more rigorous program for establishing instrument l setpoints was being initiated in accordance with the Instrument Society of j America Standard 67.0 Since then, another open inspection item l (259/260/296/S6-32-04) has been established related to the program for i

Rector Protection System (RPS) setpoints. The programmatic concerns

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raised by the radiation monitor setpoint issue will be tracked and closed ,

through the RPS setpoint open item. Therefore URI 86-05-09 is close l (CLOSED) Unresolved Item (259/260/296/87-02-07),. Problems identified during surveillance testing of the Standby Gas Treatment System (SGTS).

This item contained multiple issues of which the majority were addressed and closed in Inspection Report 259,260,296-88-05. With regard to the operability question on the relative humidity (RH) heaters, another ;

surveillance test verifies proper performance of the low flow heater :

cutoff switch and a Design Change Notice (H 6140A) has been initiated to replace the relative humidity indicating controller. With regard to a ,

suspicious temporary change to the flow calculations, the Plant Operations ,

Review Committee (PORC) reviewed this issue as documented in meeting ;

minutes 6349 and approved the change. To address this issue in total, the '

inspector witnessed the performance of SI 4.7.C, Secondary Containment Capability, performed on April 13, 1988, and found thet the problems i evident during the past performance had been acceptably addressed. This item is close ,

t (CLOSED) Unresolved Item (259/84-53-01) and (CLOSED) Inspector Followup Item (260/296/85-15-08) and (OPEN) Inspector Followup Item (259/85-15-08), :

Limitorque Valve actuator inspection progra The history on this

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inspection program can be traced through, IE Circular 79-04, Inspection Reports 84-52, 85-53, 85-15, and 85-3 The concern surrounds proper installation and orientation of various Limitorque actuator components such as pinion gears, set screws, retaining rings and split rings, which if not properly installed have been shown through experience to create premature failures. The licensee initiated a 100*4 inspection program with l independent verification of all emergency core cooling system (ECCS) valve !

actuators which are susceptible to this problem. On Unit 2, nine of the valves inspected needed some type of corrective action. All Unit 2 and Unit 3 valves have been inspected; however, some have not been released since motor changeout and pinion gear installation will be necessitated by the environmental qualification (EQ) program. The final check on pinion gear orientation will not occur until this maintenance is complete. Since this activity is controlled by detailed written instruction which depict proper gear orientation, these open items will be closed out for 'Jnit 2 and Unit 3. The inspection program for Unit I has not yet started, therefore, the Unit 1 Inspector Followup Item 85-15-08 will remain open to track completion of the program for Unit (OpEN) Inspector Followup Item (260/86-40-05), Several Walkdown !

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discrepancies and houseketping problems in the Main Steam Valve Vaul The licensee presented a closure package for this item which included MRs generated by the plant to resolve the problems. The inspector toured the i area on April 15, 1988, and found conditions still unacceptable. It was l obvious from the material laying about that post work housekeeping ,

inspections and routine periodic housekeeping inspections were 1 ineffectiv The following materials were noted; discarded rings of old l valve packing material, loose nuts and bolts not under in-use material control, burned out light bulbs, nails, stripped wire insulation, strips

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of banding material from removed pipe insulation, cotton rags, linen tags, '

poly bags, wads of used radiological control tape, ink pens, wire brushes, goggles, carpenters level, an empty bottle of "snoop" leak detector. A layer of dust and particulate material existed on the floor. and 'all horizontal surfaces. In addition to the_ housekeeping, additional _ concerns were identified as follows: A section cf 4-inch pipe in the northwest corner of the roon wa, being supported by a hand-operated chain hoist hanging from a tailing lug in lieu of a hanger which was removed. This condition appeared to have been in place for years with no tags or other controis to identify the work necessary to repair the hange The main steam tunnel blowout panels were ruste The panel's l protective coating had flaked off exposing about 90% of the bare :

meta The metal exhibited so much corrosion that the wastage may ,

need to be evaluated in light of the panel's secondary containment '

integrity function. CAQR 880293 was written to address blowout panel concern The previously identified concern over missing paint of the main steam tunnel temperature element junction boxes has yet to be fixe The previously identified concern with graffiti has yet to be cleaned, An abandoned wire rope sling, which was rusted to the point of being unusable, was found hanging from a structural member near TS-1-17 This item will remain open pending cleanup by the licensee and reinspec-tion by the NRC prior to Unit 2 Restar l (CLOSED) Inspector Followup Item (259/260/296/87-09-03), Residual Heat '

Removal Service Water (RHRSW) system maintenance instructions were inadequate to address vendor manual and operational requirements. The :

inspector noted that RHRSW pump maintenance over the past several years was ex'cessive and could be potentially attributed to improper maintenance i activitie A review of the governing maintenance instructions by the i licensee confirmed that many deficiencies did exist. The licensee I upgraded the applicable maintenance instruction to represent vendor recommendation MCI-0-023-PMP002, (RHRSW/EECW Pump; Disassembly, Inspection, Rework and Reassembly Instruction) was updated on July 14, 1937, to incorporate the NRC concerns. Also, retaining nuts were added to l

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the pump baseplates to prevent loosening due to vibration. This item is considered close (CLOSED) Inspector Followup Item (259,260,296/84-41-01), Criteria for l HPCI Walkdown Inspections. During an inspection of the High Pressure l Coolant Injection (HPCI) system, the inspector identified that no criteria was specified in plant procedures for performance of walkdowns on HPCI system piping supports following an injection. As the result of this the

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licensee committed to provide detailed instructions with checklist for walkdown inspections of HPCI following an injectio .

The inspector reviewed the licensees response to the item and found the corrective actions adequate to address the inspector's concerns as stated i in the original inspection repor l

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Browns Ferry Standard Practice 12.S, Unit Trip and Reactor Transient Analysis, has been revised to include a requirement for a HPCI Pedestal and pump discharge piping inspection by qualified personnel in accordance with SI-4.6.G, HPCI pump inspection. The inspection is to be performed by l Mechanical Maintenance of Inservice Inspection personnel and documented by '

signing the scram / event report within 3 days following an inspectio I i

SI-4.6.G specifies the piping supports required for inspection and j requires that the inspection be performed in accordance with DPM NS0E3

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procedure N-VT-1, Inservice Visual Inspection which contains adequate inspection criteria. This item is close . Q-List Concerns (35003)

The inspector identified deficiencies with regards to the Q-list imple-mentation program for Unit 2 (Unresolved Item 250/260/296/88-05-07) during the last monthly inspection. Additional meetings with licensee program managers were held during this inspection period. The licensee made the following commitments to bolster and improve the current Q-list imple-mentation program. If properly implemented the inspector considers the additional actions adequate. The item will be lef t unresolved; however, to assure proper program completion and until a more thorough QA inspec-tion in this area can be performe The corporate organization responsible for Q-list implementation must assure the proper definition of equipment that is safetv-related, important-to-safety, or limited-Q These terms appear not tv De well-defined or consistent with NRC ,

definition )

i To fulfill the commitments in the NPP Volume III, the BFN Phase I Unit 2 Q-List is a listing of nuclear safety related components, systems, and structures. Some system components are only required for the mitigation of abnormal operating transients and special events. These components are not included in the Q-List because of the present Q-List definitio However, the following steps are being taken to alleviate the inspector concerns regarding those component A review of the Q-List Design Review File, the BFN Safe Shutdown Analysis, and the associated System Requirements Calculations shall be performed to determine the operating modes (and components) not included in the Q-List because they were required to function in the mitigation of abnormal operating transients and special events, For those systems which have operating modes (and components) for the mitigation of abnormal operating transients and special events that

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are determined not to be included on the Q-List because they are not safety-related, the system designations shall be compared to the BFN CSSC to determine that all systems originally specified on the CSSC are considered in this evaluatio t A comparative review and evaluation of components within the operating modes of steps 1 and 2 will be performed to reduce the total set due to any components that appear common to safety-related operating modes, The set of ccmponents developed through step 3 will be added to the i Q-List on a systematic revision basis with definition of limited QA

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program requirements, A review of the general boundaries of the CSSC and the included operating modes of the SSA shall be performed to determine whether the Q-List for each system is enveloped by the CSS If not, CAQRs will be generated as appropriat Once all systems have been considered, as indicated in steps 1 through 5 above, Q-List procedures will be revised to indicate the Unit 2 Q-List will stand alone independent of the Unit I and 3 CSSC list.

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5. Operational Safety (71707,71710)

The inspectors were kept informed of the overall plant status and any e significant safety matters related to plant operations. Daily discussions were held with plant management and various members of the plant operating "

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The inspectors made routine visits to the control rooms when an inspector j was on site. Observations included instrument readings, setpoints and l

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j recordings; status of operating systems; status and alignments of

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emergency standby systems; onsite and of fsite emergency power sources available for automatic operation; purpose of temporary tags on equipment i

controls and switches; annunciator alarm status; adherence to procedures; adherence to limiting conditions for operations; nuclear instruments operable; temporary alterations in effect; daily journals and logs; stack l monitor recorder traces; and control room manning. This inspection activity also included numerous informal discussions with operators and i their supervisors.

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General plart tours were conducted on at least a weekly basis. Portions i of the turbine build , each reactor building and outside areas were i visited. Observation, 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 j activities in progress; and radiation protection control Informal discussions were held with selected plant personnel in their functional areas during these tours, i

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6. 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 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 require Maintenance requests were reviewed to determine status of outstanding jobs and to assure that priority was assigned to safety-related equipment maintenance which might affect plant safety. The inspectors observed the below listed maintenance activities during this report period:

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RPS MG Set preventive maintenance Secondary Containment blowout panel . periodic inspection

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RHRSW Dresser coupling replacements The inspector observed Electrical Preventive Maintenance Instruction (FPI) '

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EPI-0-099-M6-ZOO 2, 6 Month Maintenance for RPS MG Set, performed on the 2A RPS MG Set on April 4, 1988. This is a preventive maintenance item which is a general clean, inspect and lubricate activit The cleaning was limited to a wipe-down with lint-free cloths and a brushing of electrical components within the control pane No vacuum cleaner was used even though the PM card listed a vacuum in the special equipment required section. After the maintenance was completed, a large quantitysof dust and debris remained under the MG set mounting skid and throughout the room. Since excessive dust accumulation on the motor cooling vent screens and within the motor insulation has been determined to be the cause of motor failures by overheating, a more thorough cleaning of the m a would i be expected. While adding grease to the motor bearings, an excr nive j amount was require One 12-ounce tube was used for 3 bearings. The xisted with cognizant the lack ofengineer was contacted instruction and agreed detail contained thatEP in the a problen The9'roblem p was i caused by failure to break away the hardened grease at the grease drain j plug so that the new added grease could purge the old grease out through ,

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the drain hole. With the hole plugged, grease was forced out- the sides of ;'

the bearing cavity and thus, the quantity of grease added wat'nisleadin The cognizant engineer stated his intention to revise the EN to provide more details on both cleaning and lubricating instruction /

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As a result of both the inspectors concern and a licensee reportability evaluation for the year of 1987, the overall reliability of the RPS MG Sets was called into question. The failure trend was apparently on the ,

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rise with both motor insulation- failures and bearing failures being predominan The licensee's maintenance organi:ation performed a reliability study on April 4,1988. Some of the failures were attributed to improper maintenance which necessitated rework and therefore were not counted against the reliability statistic Bearing failures were.~

attributed to erd-of-life. A predictive monitoring program is being in:ti tutect to detect and repair impending bearing failure The preventive maintenance program was revised to eliminate motor failures from overneating due to dust accumulation. The study concluded that overall, the RPS MG Set reliability has been good and activities either in progress de planned are i expected to either maintain or improve on the

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curryt reliability statistic The finspeci.or discussed the results of '.he recently completed blowout pansi inspection with the cognizant engineer. This inspection is required once every 5 years per MMI-14. Problems werd detected with the Main Steam i tunnel blowout panels which serve a dua) function. These panels are designed te "blow-out" in the event of a steam line rupture in order to prevent exceeding design aifferential pressures for the Reactor Building walls. In normal r eration, the panels serve to maintain secondary containment integrit The inspection found large gaps between the blowout panels and the supporting members. These gaps had been filled in with silicon RTV sealant within the past few years apparently to reduce the secondary containment inleakage. The amount of RTV used poses the potential to defeat the blowout function of the panels. No sealant should be present, only a gasket should exist between the panels and support member Repair activities are labor intensive and must be coordinated during a time when secondary containment is not require Resolution ' 9f

, this deficiency will be tracked as an Inspector Followup Item (260/83-10-01) to ensure completion prior to fuel loa CAQR 880293 was written to address these concerns, i Surveillance Testing Ooservation (61726)

e The inspectors observed -atfd/or reviewed the below listed surveillance procedures. The inspecjion consisted of a review of the procedures fog technical adequacyr consarmance to technical specifications, verification i of test instrument calibration, obsnrvation on the conduct of the t(st, !

removal from service and returr to service of the system, a revir9 of test j data, limiting condition for opera; ion met, testing accomplished by qualified personnel, and that the surveillance was completed at the required frequenc The inspector-witnessed the performance of SI 4. Secondary Containaant Integrity, performed on April 13, 1988. This was the first performance of the upgraded procedure and was being performed as a validation run as wel'

as to support the secondary contaiment operability requirement for fuel reconstitution activitie One minor change was implemented as an immediate temporary change (ITC) during the conduct of the S ;

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) .5 \,\ la 23'of de frocedure, test switches t li{n.qiculated to

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3 s During-provide Reactor-Section }Xone t.V R*ql zone isolations. Following the

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procedure requires thn' 'pbyf on dampers be -chec@l1 clised[this, 3 If the dampers are not closed, thd procedure authorizes fpe perfonr1 Lto close the damoers and cont.fye w(th .the test. This dp w.e M riiN fently s

writtee te circumvent prior problerQwhich occur if som,e of thb tssociated ventiL) tion fans are off. In tif (case, some or phe sp hprQy,,not go closed on receipt M a test signa' This problem op s nopcgr \th the actual i solatic.<hogic SI 4.7.C acked any-guidance to the :c(:rt ibrmer so-that it inf Jrned judgement could %9ade as to whether a proJiem existed and nedbd to be thaired if a daipe s was found not to close automatically ,

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upon receipt clarifying of tod [in the next revision.The system engineer agreec' to add a statement

Another problem was noted with system cor / Oration control . The syste%

status fi?e and cWtrol room' panels both in.licated by caution order any 'MR ' '

sticker the t% power supply and trar fitte [for flow indicators FI-65-@ '

had been ' sect back to the manuf acturer,/for bepai This indicator was used durind 4% SI. .The system engi;,ir stated that a modification th6t was partially complete had installed 'a ,new transmitter and the flo indicator was funct W ,'a It was unclear .\ow the oper ators would be made aware of this and noW the operators would use this indication during operation of the SGTS System. 01-65 contains a sectio iiwhich requ'.les operators to periodically monitor the normal operating parameters during operation. FI-65-50 is[yst w forused t,y the operator to veri.fy sy stem flow rate is maintained at less tarv'12,000 cfm. The syste ,

engineer agreed to add an entry 11 'ihe tyrten status file whiih wouid clarify the status of the flow indicdto , 3

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The below listed licensee e"ents reports (LERs) wer e reviev to deternine if the information. provided met SRC requirement The detemination included: adequacy of event description, verf ficatic.) cf compiiince with technical specifications and regulatory requirem?nts, corrective acting taken, existence of potential generic problen:, repo. ting reauirements , ,

satisfied, and the relative safety significncA of each even The ,

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following licensee event rep >rts are closed: 4 s

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t LER N Date Ev ent .  ;

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84-13 Rev. 3 7/9/85 Limitorque motor pinion gear failure  :

87-02 3/13/87 Drywe11 Platform Steel i i U \

259/82-13 2/2/82 , Setpoint Drift of ', i j

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,s ( LER No; Date Event Vi (cont'd)

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259/84-30 9/23/84 Possible Block Wall Failure During a Tornado Due to Design Miscalculations of Loading 259/84-37 11/14/84 Reactor Protection System Wiring Error 259/85-04 2/23/85 Degradation of LPCI MG Set Generator Coil Clamps and Rectifier Rings 259/85-23 7/17/86 Switchyard Problems Lead to Reactor Scrams

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259/87-28 10/8/87 Unplanned Containment Isolations Due to Inadequate Procedures 6- 259/88-02 1/6/88 Engineered Safety

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Feature Actuation Due to ,

Personnel Error During Switch i Calibration 259/88-05 1/22/88 Diesel Generator and l Emergency Equipment Cooling I

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Water (EECW) Pump Actuation i

.3 Due to Personnel Error

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259/88-08 ! adby Gas Treatment

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1/20/88 Rt.ative Humidity Heaters Have Not Been Tested In l

Accordance With Technical Specifications Due to Inadequate Procedures 260/87-01 1/6/87 High Pressure Coolant Injection Valve Operator With Improperly Sized Worm Gear 260/S7-05 7/25/87 Safety Relief Valve

'- Setpoints Exceeded During Laboratory Testing

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LER N Date Event (cont'd)

260/87-09 9/11/87 Unplanned Reactor Water Cleanup Isolation Caused' by Electrical Short Circuit 296/86-08 9/5/86 Shorted Generator Coil Reduces RHR Capability LER 84-13 described various problems discovered with Limitorque valve actuators. This is the same concern tracked.under Inspector Followup Item 259/85-15-08 in paragraph 4 of this report and the LER is therefore being close '

LER 87-02 described the deficiency related to overstressing certain portions of drywell platform steel during a seismic event. The condition existed as a result of poor configuration controi .,uch that the as-built condition and as-described configuration differed substantiall The subsequent analysis of the as-built configuration revealed the over-stressed areas. This problem is being tracked in the Browns Ferry Nuclear Performance Plan under Section 3.8 and as an NRC Unresolved Item <

(259/260/296/86-14-03) pending resolution for each unit prior to each respective unit rettart. The LER is therefore being closed to consolidate tracking item The setpoints of degraded voltage relays on the 4 KV shutdown boards drif ted down 3 percent in 6 months after installation (LER 259/82-13).

The drif t was caused by initial aging and by variations in ambient temperature and supply voltag The licensee replaced the relays with a more stable type of rela The interim measure for possible block wall failure during a tornado is to have selected interior doors olocked open in the event of a tornado (LER 259/84-30). These doors are identified in EPIP-18, Tornado Emergency Procedure. The long te-m resolution is being tracked under I.E.Bulletin 80-11, Masonry Wall Desig The licensee Engineering and Design group discovered during a schematic review of the reactor protection system that nine wires were not run in conduit as required (LER 259/84-37). The wires were installed in conduit on all three unit During a walkdown inspection the generator end of a low pressure coolant injection (LPCI) motor generator (MG) set was found to be overheated (LER 259/85-04). The problem was determined to be a diode failure caused by vibration of the rectifier rin The coil clamps were also found degrade The MG sets were upgraded to new metal diode discs and coil clamp _ _ _ _ _. _ _ . _ _ . - . _ _ . . , _ ,_ _ _ _ ___. _ _ -,

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Because switchyard problems lead to reactor scrams (LER 259/85-23) a periodic retraining ~ course was developed and administered to all operations personne Caution signs were mounted in the capacitor switchyard to instruct operators on the proper method of switchin Walkdowns were performed on the control air system to repair any leakag The operators did not know that moving the radiation monitor, on the refuel floor, operating switch to Zero would cause various containment

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isolations (LER 259/87-28). The radiation monitoring system operating instruction and alarm response procedure were revised to provide a clearer unierstanding of instrument respons A critique of this event was pro /ided to all operations personne Four emergency equipment cooling water pumps were inadvertently started due to personnel error (LER 259/88-02). While returning a -wire to its normal position it was allowed to momentarily contact an adjacent terminal. The craft personnel involved were cautioned to exercise extreme care when wo. king with energized circuits. All instrument maintenance personnel were provided a critique on the even A potential transformer fuse compartment was opened for inspection causing the deenergization of a shutdown board and the start of a diesel generator and an EECW pump (LER 259/88-05). The personnel involved were counseled on the even A critique of the event was provided to maintenance, modifications, operations, and engineering groups. The walkdown procedure was upgraded to include a planning revie During a programmatic upgrade of surveillance instructions it was discovered that the surveillance instruction which tests the standby gas treatment relative humidity heaters did not full test the heaters as required by technical specifications for ANSI-N510-1975 (LER 259/88-08).

The surveillance instruction was revised to fully incorporate the testing described in ANSI-N510-197 The high pressure coolant injection valve operator for 2FCV73-2 (LER 260/87-01) was regeared to a 60 to 1 gear ratio as required as on March 3, 198 The Unit 2 main steam safety relief valve setpoints were exceeded during laboratory testing (LER 260/87-05). The relief valves were refurbished, recertified, and placed in the syste An unplanned reactor water cleanup system isolation was caused by a blown fuse when a lead from test equipment slipped out of the connection and created a short circuit ..nich blew the fuse (LER 260/87-09). The craft personnel involved were counseled on the need for increased caution when working on energized equipment. A note was added to the maintenance procedure cor.cerning short circuit ._ _ _ , . _ _ - - _ _ . . . _

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A Unit i low pressure coolant injection (LPCI) motor generator (MG) set tripped from & shorted generator coil (LER 296/86-08). It was determined that the MG set tripped due to a random failur e of the generator stator winding. The damaged generator was removed and sent to the manufacturer for repair . Restart Test Program The inspector attended RTP status meetings, reviewed RTP test procedures, observed RTP Tests and associated tests performances, reviewed RTP Test results and attended selected Restart Operations Center (War Room) and Joint Test Group meetings. The following are the RTP activities and associated activities monitored and status of testing during this reporting period: Restart Test Status (1) RTP-023, Residual Heat Removal Service Water (RHRSW)

The north header outage of the system was completed and the  ;

south header was taken out of service for Dresser Coupling l replacement at the intake structur Section 5.9, of the test i procedure requiring verification of proper operation of the l

RHRSW pump room sump pumps was complete The level switches I were calibrated and the sump pumps performed adequatel However, the licensee was not able to complete Section 5.5 of the procedure to demonstrate operability of the standby coolant valve, 2-FSV-23-56. The reason was that the Parts Request (PR)

88-1435 for this valve was moved between various organizations such as Mechanical Maintenance (MM), System Engineering (SYS ENG) and Department of Nuclear Engineering (DNE). This PR made several back and forth trips until it finally ended up with DN DNE will issue a Design Change Notice H 0161A to replace valve 2-FSV-23-56 with new valv (2) RTP-030, Diesel Generator and Reactor Building Ventilation (DG &

RX BLDG VENT)

Section 5.8, Diesel Generator Room "3C" Exhaust Fans, was successfully complete Additional testing of the system depends on the closecut of various Maintenance Requests, hold order released and interfaces with other test such as RTP-075, Core Spray and Surveillance (3) RTP-31A, Control Building Heating Ventilation and Air Condi-tioning (CONT BLDG HVAC)

During this reporting period the RTP Group decided to split RTP-31 into 31A and 318 in order to perform the overall system test more effectivel A will contain the Control Room

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Emergency Ventilation power supply modification and those items required to support the Loss of Power / Loss of Coolant Accident series of test (4) RTP-031B, Control Building Heating Ventilation and Air Condi-tioning (CONT BLC3 HVAC)

This system will have the control room habitability issue, testing of flow rates and balancing, the control room pres-surization test and the duct work modifications to verify. No LOP /LOCA items are in this system RTP Tes (5) RTP-57-2, 120 Volt Distribution System (120 V DIST)

This test was recently approved for testing by the Joint Test Group (JTG) consequently no actual testing has been accomplishe (6) RTP-57-4, 480 Volt Distribution System (480 V DIST)

Due to the extended testing of the diesel generators (DG) the RTP Group decided to test the load shed of the individual circuits rather than test them as a whole. A total of 118 l individual circuits were identified to be tested and a total of i 102 were teste Of the remaining 16, twelve (12) will be i tested in conjunction with the remainina DG testing. A total of four circuits require material and cannot be tested with the DG (7) RTP-57-5, 4160 Volt Distribution System (4.16 KV DIST)

The additional testing of this system is dependent on the  !

Special Tests being performed on all eight (8). Diesel Generators (DG). As the DGs become available for Load Acceptance Tests the various sections of this test will also be performe See RTP-082 for add tional interfacin d

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F i (8) RTP-57-7, 250 Vol t DC Shutdown . Batteries (250 VDC S/D BATT. ).

The system testing was essentially completed with one signifi- i cant test exception (T~; still outstanding which involved the  !

"B" charger filter capacitor-resistor network. The TE is documented on CAQR BFP 880163. The 250 VDC S/D Batt, system restart test procedure was compiled, initially reviewed by the RTP Group and forwarded to DNE for revie A total of 14 TEs were identified and 3 CAQRs were written as a result of the testing. The CAQR's that were dispositioned involve the ripple voltage test, Test Acceptance Criteria 6.2 and Steps 5.2.26 through 5.2.32 of the RTP procedure. By correspondence, the ripple criteria was changed from .5% to 1%, thus the CAQR was no longer applicabl . - .. - . . _- -

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(9) RTP-065, Standby Gas Treatment (SGTS)

Testing was virtually completed during this reporting perio One significant TE involving the stack effect .was observed by

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the inspector and a CAQR was generated to address this issu (10) RTP-067, Emergency Equipment Cooling Water (EECW)

The north header outage and replacement of Dresser couplings was completed and the south header was taken out of service for replacement of Dresser couplings. The system was impacted by DCN 3549 issuance which addresses the thermowell issue. Testing will continue as equipment, especially chillers, become availabl (11) RTP-070, Reactor Building Closed Cooling Water (RBCCW)

JTG released this system for partial testing in order to support the LOP /LOCA Test. The testing involved Section 5.1, -Operation f rom the Control Room, which included subsections 5.1.1 through 5.1.13, Section 5.2, operation outside the control room, which included subsections 5.2.7 through 5.2.18 and Section 5.4, which verified the start of RBCCW Pump B auto-start following a LOCA with diesel generator voltage available when RBCCW Pump A fails to star (12) RTP-075, Core Spray (CS)

System testing was impacted by ECN 3018 to replace the breakers for Flow Control valves 2-FCV-75-23, 25, 51 and 5 These valves are involved with the GE Loss of Coolant Accident Valve Time Study. The valves were modified; however, the electrical power breakers were no (13) RTP-082, Standby Diesel Generators (STDBY DG)

Special Test (ST) 8806 was attempted involving 1B DG, However, excessive vibration plus the need to replace the DG blower (supercharger) impacted the completion of this special tes The inspector was informed by the RTP group that the load acceptance t(st was being rewritten from an Surveillance Instruction (SI) to a S Upon review of the RTP procedures SDSP 12.1 and 12.2, it is not clear as to the use of ST's to meet the RTP Test requirements. This item is identified as Inspector Followup Item 'IFI) 259,260,296/88-10-05, Performing Specia Tests to Meet RTP Test Requirement I

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b. Design Testing Observations  :

(1) Standby Gas Treatment System (SGTs) Smoke Generation Test On April 12, 1988, the RTF Test Director for System 65, SGTS, attempted to generate a smoke test in order to visually observe the direction of air movement in the vent duct work. This activity was observed by a QC representativ This smoke generation was required by RTP Test Instruction 2-BFN-RTP-065, Section 5.12, Off-Gas Stack Ef fect, Step 5.12.8.1. Specifi-cally, the smoke blowing test was an attempt to verify the existence of a draft caused by the SGTS trains forcing air up the stack and veri fy . presence of negative pressure in the off gas cubicle vent ductwor The smoke blowing was not successful even af ter several attempts due to the fact that a positive pressure was present in the vent ductwork. This was verified when the smoke, rather than being sucked up into the vent duct work, blew away. The inspector accompanied the system engineer and the test director on a walkdown of the off gas system and off gas building ventilation ductwork because it ties into the off gas cubicle vent ductwork inside the stac Two fans were observed in the off gas (System-066) building with one running and exhausting air from the building; the other fan was in standby condition. The test director and system engineer indicated that the exhaust fans in the off gas building should have been addressed in the RTP. The inspector noted the RTP test director was not the original author of RTP-065. Howe';e r ,

this is an example of an RTP test not adequately scoped prior to performance. Upon further evaluation it appears that when power is lost and stack effect is called upon to function, the effect ,

would not only place a negative pressure on the off gas cubicles l located in the basement of the stack, but it also place a <

negative pressure inside the off gas building as well. Further followup and review of the as-built ventilation installation in the off gas building indicated an additional three (3) exhaust fans located in panels within the off gas buildin These 1 exhaust f ans in the building and in the panels were turned off I and the smoke blowing was attempted again without success. This time it was determined that the positive pressure was being caused by air flowing back down the stack, therefore no stack effect existed. This is identified as Inspector Followup Item 259,260,296/88-10-02, Lack of Stack Effect for Anticipated Air i Circulation using Smoke Mediu I (2) RTP and Operations Interface (a) While the inspector was observing the performance of RTP-065 SGTS, Section 5.2, Secondary Containment Draw Down With Two SGTS Trains, the RTP Test Director was informed by the Assistant Shift Operations (AS0) Supervisor that the

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test would have to be terminated and all co_mpon en t s -

returned to normal status. Subsection 5.2.7 had already been completed which required the lif ting of leads in control room panels in order to lock in the ' isolation of secondary containment signal and a complete restoration had to be performed on this subsection. The ASO supervisor was asked why Section 5.2 had to be terminated and the answer involved a safety question with Operations Instructions 01-65, Standby Gas Treatment System, which concerned the stack dilution fan The position taken by the ASO supervisar appeared to be' that the RTP test could not override an 01, even though in this case the RTP Test did not reference the 01-65 or require the dilution fans to be operable or operatin The Operations Superintendent was informed of this conflict. The situation was resolved and the RTP test section was completed that da (b) It was brought to the attention of the inspector that while RTP-030, Diesel Generator Building and Reactor Building Ventilation System (DG & RX BLDG VENT), Sections through 5.9, DG Rooms Exhaust Fans, were being performed a request was made of the Shift Operations Supervisor (SOS)

that a specific Diesel Generator monthly SI be performed in conjunction with the specific RTP Test Section. This request was apparently granted. However, later in the shift the RTP Test director noted that the SI was in 1 progress and the SOS had not notified the Test Director of 1 the change in plan The RTP missed an opportunity to l perform a specific test section and would have to wait another month to perform the affected test section, i l

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In observing the RTP and operations personnel interface during RTP testing, the inspector noted that there was a lack of understanding of the Restart Test Program by the incoming Senior Reactor Operator This is identified as an Inspector Followup Item (IFI) 259,- i 260,296/88-10-03, lack of understanding of the Restart Test Program by ta shift senior personne . Personal Dosimetry (83724)

The inspector observed reading of TLDs during the quarterly processing on April 1, 1988. The Browns Ferry program has been judged acceptable by the National Voluntary Laboratory Accreditation Program (NVLAP) per National Bureau of Standards (NBS) criteria. During this review, the inspector spot-checked various requirements of DSIL 12, Operation of the Automatic Panasonic Model UD-710 Reader, reviewed trend charts which visualize display equipment performance history, and verified operator reading files were current for the individuals reading TLD Adequate management -

involvement was apparent throughout the program as evidenced by the onsite presence of the corporate manage *

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11. Fuel Reconstitution (60710)

On March 23, 1988, the licensee submitted a proposed fuel inspection and reconstitution program. The purpose of the program was to improve fuel reliability with a goal of zero fuel failures during the next operating cycl The process .is to be controlled as a special test per 10 CFR 50.59. The licensee's safety evaluations concluded that no Unreviewed Safety Questions exists; however, operability questions with non-seismically qualified secondary containment penetrations and an unanalyzed in-leakage question with the Control Room Emergency Ventilation (CREV)

system necessitated a concurrence by the NRC. On April 11, 1988, the NRC found the proposal acceptable. In preparation for the reconstitution activity, the following activities were inspected: SI 4.7.C, Secondary Containment Integrity, was witnesse The upgraded maintenance instruction for the refueling platform was reviewed (MMI-34), Reviewed selected CAQRs on the required systems such as the CREV bypass flow problem (CAQR No. BFP 870591) and lack of a calculation to document the safety limit and instrument setpoint for the control room isolation function (CAQR No. BFP 870876). Reviewed the Temporary Alteration Control Form (TACF) file k_ viewed the PORC meeting minutes and attended PORC meetings where the special test was discusse Reviewed the f ailure evaluation on the Refuel Floor overhead crane .

bolt I Reviewed the detailed step-by-step fuel reconstitution instructions and i.,terviewed contractor personnel regarding actions taken or planneo to minimize the potential for repeat occurrences of individual rod drops and loss of loose part Reviewed, on a sampling basis, outstanding Maintenance Requests (MR's) on the required system Discussed with Radiological Controls personnel the augmented Rad Con coverage and the ALARA prepla Reviewed the accidental criticality analysis and the revised accidental criticality analysi Reviewed the calcu tion which supported the safety evaluation conclusion that the C.sdV system was not needed to protect the operator from overexposure in the unlikely event of a fuel handling acciden l l

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l Reviewed Surveillance Instruction for the required system which are only required to be performed once.per operating cycle to ensure each had been completed within the last 18 months.

i Made weekly and in most cases daily tours of the refuel floo Reviewed outstanding LERs and NRC open items to determine the need for expedited closecu Reviewed the licensee's System Operability Evaluation Reports which provide the necessary justification for considering the required systems operable in light of any partially completed modifications, outstanding MRs, adverse findings of the Design Baseline Verification Program, Restart Test Program, and other conditions adverse to quality. Additional emphasis was placed on any special operational constraints or compensatory measures required in order to consider the systems as operable in light of any outstanding problem Examples of this type of activity are the temporary installation of jumpers to prevent CREV load shed in the event of offsite power loss, jumpering out auto-start of the Residual Heat Removal (RHR) pump, and ;

tagging closed certain Emergency Equipment Cooling Water (EECW) '

valve One significant problem was detected with the licensee's program submittal of March 23, 198 In the Control Room Emergency Ventilation System (CREV) section of Enclosure 1, the licensee stated that "a calculation has been performed to evaluate the effects of the increased unfiltered I

inleakage." The results of this calculation were used in the submittal to conclude that the CREV system was not needed to mitigate an accident during the fuel reconstitution activitie The inspectors review of the calculation (ND-N0079-88013) showed that the calculation was not approved and issued for use until two days af ter the licensee's submittal . The information in the submittal was therefore based upon a draft calculation without explicitly identifying it as suc In fact, some of the second party checks of the calculation were not even performed until March 25, 1988 (two days af ter the licensing submittal). This issue was discussed with Site Licene.ing and Corporate Licensing personnel who assured the ;

inspector that the policy is not to use draft information in NRC l submittals and that a verification program is in development that should minimize the potential for recurrence of this proble The inspector reviewed PMP 0602.01, Management of TVA's Interface With the Nuclear j Regulatory Commission, which details the responsibilities and requirements l for the preparation of licensing documents. This procedure was found to i lack any requirement or guidance on the use of draf t information in the preparation of licensing submittals. ibis deficiency appears to be a violation of Technical Specifications 6.8.1.1.j. which requires that administrative procedures be established that control technical and cross-disciplinary review (Violation 260/88-10-04). '

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Unit 2 Fuel reconstitution activities began on April 28, 1988.

t 1 Exit Interview (30703) .

The inspection scope and findings were summarized on May 2, 1988, with the Plant Manager and/or Superintendents and other members of his staff. New t items identified: Inspector Followup Item (260/88-10-01) Main Steam Tunnel Blowout Fanel deficiencies. Paragraph Inspector Followup Item (259,260,296/88-10-02) Lack of Stack Effect for Standby Gas Treatment Syste Paragraph 9.b (1). Inspector Followup Item (259,260,296/88-30-03) Restart Test Program Awareness by Senior Personnel on Shif Paragraph 9.b.(2). Violation (260/88-10-04) Failure to Establish Procedures to Adequately Control Technical and Cross Disciplinary Review as Required by T.S. 6.8.1.1.J. Paragraph 1 Inspector Followup Item (259,260,296/88-10-05) Performing Special I Tests to Meet RTP Test Requirement Paragraph 9.a.(13). l

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The licensee acknowledged the findings and took no exception The !

licensee identified certain material associated with the fuel reconstitution procedures (e.g., visual examination standards, drawings for special tools, etc.) as being General Electric Proprietary Information. On May 25, 1988, the NRC notified TVA that Item d. above had been upgraded from an inspector followup. item to a v'olatio I i

13. Acronyms and Abbreviations l ASO - Assistant Shift Supervisor i BFN - Browns Ferry' Nuclear l CAQR - Condition Adverse to Quality Report CREV - Control Room Emergency Ventilation CS - Core Spray CSSC - Critical Structures, Systems, and Components DG - Diesel Generator DNE - Department of Nuclear Engineering ECN - Engineering Change Notice EECW - Emergency Equipment Cooling Water EPI - Electrical Preventive Maintenance Instruction EQ - Equipment Qualification FI - Flow Indicator HPCI - High Pressure Coolant Injection HVAC - Heating Ventilation and Air Conditioning

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LER - Licensee Event Report LOP /LOCA - Loss of Power / Loss of Coolant Accident LPCI - Low Pressure Coolant Injection MG - Motor Generator MR - Maintenance Request 01 - Operating Instructions PORC - Plant Operations Review Committee QA - Quality Assurance RBCCW - Reactor Building Closed Cooling Water SGTS - Standby Gas Treatment System RHR - Residual Heat Removal RHRSW - Residual Heat Removal Service Water RPS - Reactor Protection System RTP - Restart Test Program RWCU - Reactor Water Cleanup SDSP - Site Director Standard Practice SI - Surveillance Instruction SOS - Shift Operations Supervisor TR - Temperature Recorder TS - Technical Specifications

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