ML20138N206

Insp Repts 50-327/85-35 & 50-328/85-35 on 851006-1105. Violation Noted:Failure to Implement Procedures Re Reactor Trip Response Time Testing,Installation of Containment Penetration & Radiation Monitoring Testing

ML20138N206
Person / Time
Site:	Sequoyah
Issue date:	12/06/1985
From:	Jenison K, Linda Watson, Weise S NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML20138N155	List: ML20138N150 ML20138N166 ML20138N206
References
50-327-85-35, 50-328-85-35, NUDOCS 8512230406
Download: ML20138N206 (15)

See also: IR 05000327/1985035

Text

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i i. - - pn a.t r - UNITgD STATES ' g .o NUCLEAR REGU'.ATORY COMMISSION [[ - REGION 11. , . g -- -j 101 MARIETTA STREET,N.W. ~*- 2 ATLANTA, GEORGIA 30323 4. . .,. ,/ i/ % . ' Report Nos~.: 50.-327/85-35, 50-328/85-35 \\ ! Licensee: Tennessee Valley Authority - 6N11 B Vissionary Ridge Place- m "- --1101.Ma'rket Street , Chattanooga, TN 37402-2801 - Doc ket < No s'. : 50-327'and 50-328 ' License Nos.: DPR-77 and DPR-79 ' Facility Name: Sequoyah Units 1 and 2 . Inspection Conductea: Octeer 6 through November 5, 1985 Inspectors: 6Qd <W /A/05/B5 K. M. Wnisof, Senior Resident Inspector Dat'e Si'gned . G 0. nd .Ww /G-loS/A5 L. J. W4tson,gResident Irspector Dat'e Signed Accompanying. Personnel: G. . Pi Approved by: 7/ ~ II S. P. Weise,~ Section Chief DatE Signed . Division of Reactor Projects 4 Summary Scope: .This routine, announced inspection involv'ed-349 resident inspector-hours onsite in the areas of operational . safety verification including operations , -performance, . system lineups, radiation protection, . security . and housekeeping ' inspections; ' surveillance and maintenance observations; review of previous inspection findings; followup of events; review of licensee identified items; walkdown-of Engineered Safety Features;; and review of inspector followup items.

Results: One violation was identified - Failure to implement procedures'in the

areas of reactor trip response time testing (paragraph 7), installation, of, a containment penetration'(paragraph 8), radiation monitor testing (paragraph'10);

and, configuration control of a radiation monitor power source (paragraph 10). , ' . 4 % B512230406 851210 PDR- ADOCK 05000327 G PDR - -- .- - , - . - - . .- - - . ..

g . 4 . REPORT DETAILS - - 1. Licensee Employees Persons Contacted H. L. A'oercrombie, Site _ Director

• P. R. Wallace, Plant Manager
• L. M. Nobles, Operations and Engineering Superintendent
• B. M. Patterson, Maintenance Superintendent

J._M.- Anthony, Operations Group Supervisor R. W. Olson, Modifications Branch Manager ~ M. R. Sedlacik, Electrical Section Manager, Modifications Branch '

• H._D. Elkins, Instrument Maintenance Group Manager

G. B. Tiner,. Instrument Maintenance Engineer

• M. R. Harding, Engineering Group Manager
• D. C. Craven, Quality Assurance Supervisor

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• G. B. Kirk, Compliance Supervisor

, M. L. - Frye, Compliance Engineer D. H.:Tullis, Mechanical-Maintenance Group Supervisor J. H. -Sullivan, Regulatory Engineering Supervisor

• C, E. Bosley, Quality Assurance. Auditor

-Other licensee employees contacted included technicians, operators, shift engineers, security force members, engineers and maintenance personnel.

• Attended exit interview

2. Exit Interview The inspection scope and findings were summarized with the Plant Manager and members of his staff on November 6, 1985. A violation with examples described in paragraphs 7, 8 and 10 was discussed. The licensee acknowledged the inspection findings and identified as proprietary a portion of- the material reviewed by the inspectors- in regard to the negative rate trip application as discussed in paragraph 10. The information in this report 'does not include that proprietary information. During the reporting -period, frequent discussions were held with the Site Director, Plant Manager and his assistants concerning inspection findings. At no time during the inspection was written material provided to the licensee by the inspector. 3. Licensee A>: tion on Previous Inspection Findings (92702) .This subject was not addressed in this inspection. 4. Unresolved Items No unresolved-items were' identified during this inspection. 4 [

. . 2 '5. ' Operational Safety Verification (71707) ' .a. Plant Tours

The inspectors observed control room operations, reviewed applicable

logs, conducted discussions with control room operators, observed shift . turnovers, and confirmed operability of instrumentation. The inspectors verified the . operability of selected emergency systems. -reviewed tagout records, verified compliance- with Technical -Specification (TS) Limiting Conditions for Operation (LCO) and verified. return to service of affected components. The inspectors verified that maintenance.. work orders had been submitted as required and that followup activities and prioritization of work was accomplished by the licensee.

Tours of the diesel generator, auxiliary, control, and turbine buildings and containment were conducted to observe plant . equipment conditions, including potential fire hazards, fluid leaks, and excessive vibrations and plant housekeeping / cleanliness conditions. The inspectors walked down accessible portions of the following safety-related systems on Unit I and Unit 2 to verify operability and proper valve alignment: Residual Heat Removal System (Units 1 and 2) Charging Pump Flowpath (Units 1_ and 2) Control Room Ventilation Chlorine Detection System (Common) Spent Fuel Pool Cooling System (Common) b. Security . During the course of the inspection, observations relative to protected and vital area security were made, including access controls, boundary integrity, search, escort, and badging. ' On November 1, 1985, the licensee declared a moderate security . degradation as a result of the actions of a security officer posted at -the entrance of -the Unit 2 containment hatch on the 690 level. Appropriate ' compensatory actions were taken and the licensee's personnel administrative process was implemented. The inspector reviewed the above incident and had no further questions. .This item will be reviewed by NRC specialist inspectors at a later date. No violations or deviations were identified, c. Radiation Protection The inspectors observed Health Physics (HP) practices and verified implementation of radiation protection control. On a regular basia, radiation work permits (RWPs) were reviewed and specific work activities were monitored to assure the activities were being conducted in accordance with applicable RWPs. Selected radiation protection +

' y- - ,. > 2 , 3 4

instruments were verified operable and calibration frequencies were

< . reviewed. 6. Engineered Safety Features Walkdown (71710) EThe , inspector verified. operability of the Component Cooling Water system (CCS) on' Units 1'and 2: by continuing a walkdown of the accessible portions of a the systems.~ -Inspection Report 327,328/85-32 documents the previous inspection.of this.' system. The following specifics were : reviewed and/or . ' observed'as_ appropriate: a. _that the licensee's system lineup procedures. matched plant drawings and the as-built configuration; p b.

that equipment _ conditions were sati sfactory and items that might

t. ' degrade ' performance were identified and evaluated (e.g. hangers and ' supports were operable, housekeeping etc, was adequate); c. ~with assistance 'from licensee personnel, the interior of-the breakers and electrical or instrumentation cabinets were inspected for debris, loose material, jumpers, evidence of rodents, etc;

d.

.that instrumentation was properly valved in and functioning and , calibration ~date's were appropriate; e. that -valves were in proper position, breaker alignment was correct, power was available, and valves were locked as required; and f. local and remote instrumentation was compared, and remote instru- ' mentation was functional. ,

No violations or deviations were identified.

' 7' -Monthly Surveillance Observations (61726) . The inspectors observed Technical Specification (TS) required surveillance testing and verified that testing'was performed in accordance with adequate procedures, that test instrumentation was calibrated, that Limiting Conditions for Operation.were met, that test results met acceptance criteria requirements -and were reviewed by personnel other that the individual directing the test, that deficiencies were identified, as appropriate, and that any deficiencies identified during the testing were properly reviewed and resolved by management personnel, and that system restoration was adequate.- For complete tests, the inspector verified that testing ' frequencies were met and tests were performed by qualified individuals. The inspectors witnessed / reviewed portions of the following surveillance test activities: SI-82.2 Functional Tests for the Radiation Monitoring System 9 --- --. .m - r..- , - , , , , , , - - . - , , , .cy ,. ,,,-,-.w-%-.,w,7m,-wg,-,,-,y %-..m- .m- -

- .___ - - __ - - . . . 4 .SI-67 . Periodic Calibration of the RPI System .The inspectors reviewed the results of reactor trip response time testing. The following procedures were reviewed:

IMI-99 ' Reactor . Protection System RT 11.6, Response Time Test of

dT/Tavg Channel II, Rack 6 IMI-99 Reactor ' Protection System RT 11.8, Response Time Test of dT/Tavg Channel 4, Rack 13 IMI-99 Reactor Protection System RT 7.14 Response Time Test of Loop 1 Steam Generator Level Channel III (L-518) (L-3-39) IMI-99 Reactor Protection System RT 7.17 Response Time Test of Loop 2 Steam Generator Level Channel III (L-528) IMI-99 Reactor Protection System RT 7.20 Response Time Test of Loop ! 3 Steam Generator Level Channel III (L-538) IMI-99 Reactor Protection System RT 7.23, Response Time Test of Loop 4 Steam Generator Level Channel III (L-548) IMI-99 Reactor Protection System RT 611A, Response Time Testing Engineered Safety Feature Actuation Slave Relay K611 The inspector observed a. portion of the performance of the response time testing for loop 1 steam generator level Channel III under procedure RT 7.14, The technician stopped the test when he could not complete step 4.4 which required that he insure that a test indicator light on the train he was testing was -lit. The test indicator light was not lit. The technician took the procedure-to his foreman for guidance. The foreman discussed the step with an instrument maintenance engineer and determined that the light would not illuminate because the reactor trip breakers were not closed. A nonintent change was requested to revise the procedure. During these discussions, the inspector observed that a piece of scratch paper with a note written on it had been inserted into RT 7.14 indicating that Step 55, which had not been performed at that point, could not be . performed because of plant conditions. Step 55 requires verification of certain block switches by confirmation that the block switch lights were -lit. The technician stated that he had been instructed to place the remark N/A (not applicable) adjacent to this step, and to continue with the test; however, the technician stopped the procedure performance prior to reaching this step. The inspector reviewed additional procedures and determined that certain steps had been marked N/A. In RT 611A, Step 5.5.6 requires that certain equipment be returned to normal position. This is an independent verifi- . cation signoff. Twenty-six of these steps were marked N/A with a note that the components were tagged under various hold orders not specified in. the , , ,----5 ,wr +- wa,y-r-. w. -4+9y. w- .-m-m-p ..eaw-- .

ft - . . 5 data-sheets. It should be noted that hold orders require independent veri- fication of return to service. One additional non-safety-related component was marked N/A with no reference to a hold order or other explanation. The procedure requires that - if a device cannot be returned to normal, the information should be entered as discrepancies on the data cover sheet. This_information had not been entered in the data sheet as a discrepancy. Additional . steps in procedures RT 7.17 and 7.23 require verification that the status and alarm lights are not lit except as allowed by Step 2 of the procedure which states that lights marked by an asterisk may be normally lit if the unit is offline. Lights had been verified to be in a status not allowed by the procedure and signed off as acceptable due to plant conditions. The licensee stated that the status and alarm light verification did not affect test performance in Modes 5 and 6 but was to assure that if the test were performed in modes 1 through 4, the reactor would not be tripped. The failure to follow procedures RT 611A, RT 7.17, and RT 7.23 constitute a violation 327,328/85-35-01. In addition, the inspector noted that the following steps in procedure RT 611A had been marked N/A in the data sheet when it appears that they .should have not been marked that way. Step 4.1.10 requires that an annunciator window for the low pressure indication from the Condensate Storage Tank to the Auxiliary Feedwater Pump (AFWP) be cleared. The pressure switch would automatically initiate Essential Raw Cooling Water (ERCW) flow to the AFWP if the pressure reached the low setpoint. The step was marked N/A with a note that H0 1073 had power off of all ERCW valves. In this case, the reason for the annuniciator window indication was clearly indicated in the data sheet. Step 5.2.1 was marked N/A. This step had a double entry for signing off one handswitch position in the data sheet. This entry was clearly a typographical error and should be corrected. 8. Monthly Maintenance Observations (62703) a. Station maintenance activities of safety-related systems and components were observed / reviewed to ascertain that they were conducted in accordance with approved procedures, regulatory guides, industry codes and standards, and in conformance with TS. The following items were considered during this review: LCOs were met while components or systems were removed from service; redundant components were operable; approvals were obtained prior to initiating the work; activities were accomplished using approved procedures and were inspected as applicable; procedures used were adequate to control the activity; troubleshooting activities were controlled and the repair 1

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' - . w 6 record accurately reflected what actually took place; functional testing and/or calibrations were performed prior to returning components ~ or systems to service; quality control records were maintained; activities were accomplished by qualified personnel; parts and materials used were properly certified; radiological controls were implemented; QC hold points were established .where required and were observed;-fire prevention controls were implemented; outside contractor ' force activities were controlled ir accordance with the approved Quality Assurance (QA) program; and housekeeping was actively pursued. b. The inspectors reviewed the modification of feedring J-tubes in the four steam generators. The licensee had planned a modification involving replacement of the carbon steel J-tubes with Inconel J-tubes due to wall thinning in the J-tubes. Upon examination of the tubes and feedring after removal of the J-tubes, the licensee determined that the carbon steel feedring had been eroded by high velocity flow at the base of the J-tube. The modification was revised to include oversized boring of the holes in the feedring to eliminate the eroded areas and buildup of the J-tube wall with Inconel in this area to fit the larger hole. The inside diameter of the J-tube remained the same except that the entrance to the tube from the feedring was machined to a smooth rounded edge to prevent turbulance. The J-tube was welded to the feedring with Inconel weld filler metal. The inspector examined J-tubes removed from the steam generators and examined a portion of one feedring upon removal of the J-tubes. The inspector also observed an inspection of J-tube welds by the vendor's QA representative. The inspector reviewed WP 11829 which covered all the J-tube replacements in the four Unit 1 steam generators and drawings D-246-941-1 and -2. The licensee will be presenting the findings on- the feedring degradation to the Westinghouse Steam Generator Owners Group during November,1985. No violation or deviations were identified. c. The inspector observed preparation for a leak test of a containment penetration which had been replaced to meet environmental qualificaton requirements. The inspector reviewed Work Plan 11802, dated October 28, 1985, which required that the penetration be assembled and tested in accordance with a validated vendor manual. The inspector determined that the licensee had not received the updated vendor manual describing the assembly of the feed thru tubes for the penetration. The vendor manual at the work site did not describe the assembly of the feedthru tubes, but did describe the leak test requirements. The inspectors determined that the manual used had not been reviewed and validated by PORC. The penetration was installed and assembled based on verbal instructions received from the vendor at an earlier date. The vendor, subsequent to this inspection providea written instructions to the licensee which included requirements for QC hold points not

W ' . . 7 performed during -installation. Based on ' this new information, the ' licensee reworked the penetration in. accordance with the vendor's instructions and a validated vendor manual. Failure to' implement the work 1 plan for. assembly of containment electrical penetrations is a ~ further example of violation 327, 328/85-35-01. d. The replacement of electrical relays in.the 6.9KV shutdown boards was observed. The following documents were reviewed: Special Maintenance Instruction SMI-0-202-1 Maintenance Instruction'6.20 - Maintenance Request A284454 Procurement Documents (575) 5886000390, 5886000771 The maintenance appeared to be adequate and no violations or deviations were identified. -

9.

Licensee Event Report (LER) Followup (92700) 'The following LER's were reviewed and closed. The inspector verified that: reporting requirements had been met, causes had been identified, corrective actions appeared appropriate, generic applicability had been considered, the LER forms were complete, the licensee had reviewed the event, no unreviewed safety questions were involved, and violations of regulations or Technical ' Specification conditions had been identified, a. LER Unit 1 327/85021 Control Room Ventilation Isolation 327/85023 Auxiliary Building Isolation 327/85026 Failure to Obtain a. Noble Gas Sample 327/85027 Main Steam Line I:olation 327/85029 Reactor Trip on Loss of Power to Main Feedwater Pump 327/85030 Auxiliary Feedwater Initiation 327/85033 Main Control Room Isolation Due to Failure to Follow Procedure 327/85034 Diesel Generator Operability 327/85035 Emergency Diesel Generator Start While Trouble Shooting Control Power 327/85037 Main Control Room Isolation Due to Spike on Radiation Monitor 327/85038 Auxiliary Building Isolation From SFP Rad Monitor During Filter Changeout b. LER Unit 2 328/85007 Inadvertent trip 2A-A Shutdown Board Feeder Breaker 328/85008 Failuresto Complete Hourly Fire Watch . .

____ __ _ _ _ __ _ ~ , . . 8 , '10. Event Followup-(93702, 62703, 61726) On October 1,1985,- the inspectors received a copy of a Westinghouse a. Technical . Bulletin which dealt with- negative and positive flux rate reactor trip setpoint calibration methodology. Discussions were held with'several levels of plant management including cognizant engineers and technicians. As a result of this process the following documents > .were reviewed: Westinghouse Technical Bulletin NSID-TB-85-13 Westinghouse Technical Manual N2M-2-1-X, Nuclear Instrument-System Surveillance Instruction (SI) E0, Power Range Nuclear Flux Channel Calibration and Functional Test Standard Practice SQA26 Attachment 4, Operating Experience Review Recommended Action Sheet Instrument Maintenance Instruction (IMI) 92-PRM-CAL, NIS Power Range Standard Practice SQA26 Attachment 3, Experience Review Evaluation Form Standard Practice SQA26 Attachment 2, from Supervisor, Regulatory Engineering to Supervisor, Instrument Maintenance and Lead Instrument Engineer (D. Elkins, R. Gladney) Standard Practice SQA26 Attachment 1, Operating Experience Review Screening Sheet TVA memo McGriff to Brimer, Sullivan of September 3,1985 TVA memo Gibbs to Wilson copy to Sauer of July 17, 1985 (Note: This is a Watts Bar site memo) Technical Specification Change Request 85-122 Management A: tion Tracking System (MATS) Assignment Sheet dated June 26, 1985 TVA memo McCloud to McGriff dated June-25, 1985 Precautions, Limitations and Setpoints for Sequoyah Nuclear Plant NRR memo, T. Dunning to Dunenfeld, Westinghouse Neutron Flux Rate Setpoints Sequoyah Nuclear Plant Startup Test 9.5 Evaluation Report WCAP-10297-P-A Westinghouse Dropped Rod Methodology for Negative Flux Rate Trip Plants Technical Specification Table 2.2-1 The SNP management and staff were knowledgable of the WTB about the second week in July. The WTB discussed the alignment procedure for the Nuclear Instrumentation system power range positive and negative rate trip bistables and explained that some plants had misinterpretated the procedure as outlined in the Westinghouse Nuclear Instrumentation System Technical manual. The electrical circuit addressed by both the Nuclear Instrumentation Technical manual and the WTB consists of an upper and lower detector whose signals are indicated on nuclear instrument (NI) meters 301 and 302. The signals are added together and averaged through a level and averaging circuit. A resulting adjusted signal is then read on full percent power meter 303. . _ . _

p 9 . 9 The adjusted electrical signal passes through two subsections of the power range rate and delay circuit (NM311), resulting in a potential difference on a downstream operational amplifier. The output of the

operational amplifier is fed to the input of a bistable which will trip

when a given input value-is reached, resulting in a reactor trip. The- Westinghouse Nuclear Instrumentation (WNI) Technical Manual described the process used to calibrate this power range rate and delay ' circuit to ensure that bistable NC301 has the proper TS reactor trip setpoint. Surveillance Instruction SI 80 was reviewed and appeared to conform with what was indicated in the WNI Technical Manual. Performance of the steps described in the WNI Technical Manual and SI 80 resulted in a stepped potential difference of 3% (negative rate trip) and 5*4 (positive rate trip) being applied to the operational amplifier in the rate and delay circuit. The Westinghouse Technical Bulletin stated that the power range detector A test signal is used to create a step signal which is the input to the power range rate and delay circuit (NM311) and that the detector A test signal should be set numerically equivalent to the value of percent full power change given~ in the plant Precautions, Limitations and Setpoints(PLS) document. For Sequoyah, the PLS document disagrees with the Technical Specifications, and the licensee used the Technical Specification values. The WTB also stated that due to possible misinterpretation of the Nuclear Instrumentation System manual, plants may have doubled the Detector A test signal in order to compensate for the summing the level amplifier. Additionally, the WTB ~ requires maintenance personnel to set the detector A test signal in power units or percent of full power detector current, to the value given in the PLS document for tne percent full power change for the rate trip. For example, if the PLS document requires a rate trip on a 5% change of full power, then the detector A test signal should be set to 5 power units or 5% of detector A full power current. The difference between the Westinghouse Technical Bulletin (WTB) and the Westinghouse Nuclear Instrument Technical Manual is in the amplitude of the potential applied. The WTB requires that the amplitude be read on the meter after the leveling circuit. The initial TVA management review determined that the bulletin could not be complied with because the operational amplifier input would have to be set to 1.5*. and that this value would not allow sufficient margin from' normally present nuclear flux circuit noise (approximately l*s). The licensee interpreted that the trip value of the TS should be equal to the magnitude of the detector input since this is consistent with standard TS trip setpoint methodology.

- . y .. . 10 A TS change request was processed through the Plant Operations Review - Committee (PORC) on. September 11, 1985 to implement the standard TS trip setpoint valves. It stated that implementing the calibration method stated in the WTB would significantly increase the chances of inadvertant trip actuations caused by nuclear noise using the current TS valves. . -The licensee requested Westinghouse to perform a study and determine whether the WTB applied to Sequoyah. The TS change request was- submitted to the NRC by letter dated October 22, 1985. Westinghouse's response addressed the conservatism of the current Sequoyah TS compared to the PLS values and performed some calculations on the power range rate and- delay circuit. Although Westinghouse calculations were provided for several cases, the results appeared to be only conditionally acceptable. Conversations were held between NRC Region II and the licensee and NRR personnel. The licensee's interpretation on setpoint methodology for testing was consistent with TS intent. In light of the WTB, the NRC determined that the TS were in error and that this issue appeared to be generic. Resolution of this TS issue prior to the startup is an Inspector Followup Item 327, 328/85-35-02. On October 30, 1985, The inspector witnessed a surveillance test which verified the Digital Rate Circuit time constant on Power Range Monitor channels N-41 and N-43. The work was requested under MR A-539515 and A-539516 for channels N-41 and N-43, respectively. The technicians utilized Instrument Maintenance Instruction, IMI-92-PRM-CAL steps 5.2.9.12 through 5.2.9.14 to perform the test and IMI-134 to record the data. The test was conducted by inputing a negative three percent change in power level and upon reaching the desired level determining the decay time to reach 37% of the initial value, f.e., one time constant. The test determined that the time constant for N-41 was 1.31 seconds for for N-43 was 1.30 seconds. The time constant is required per TS table 2.2-1 to be greater than 1 second. No violations or deviations were identified. b. On October 26, 1985, the licensee discovered a leak in the reactor cavity liner. The cavity was drained and a nozzle cover was repaired and reseated and the cavity was refilled. On October 28, 1985, the licensee discovered that the cavity liner was again leaking. The leakage was going to the keyway sump-under the vessel and through the number 2 cold leg penetration to the containment sump. The licensee evaluated the leakage, which remained steady at approximately I gpm through the end of the report period, and determined that the liner itself was probably the source of the leakage. The inspector reviewed the procedure for failure of the reactor cavity seal, Abnormal Operating Instruction, A0I-290 and discussed the leak rate with the cognizant engineer. The licensee stated that based on the present indications that refueling operations would proceed with close monitoring of the leakage. At the end of refueling the licensee will drain the reactor cavity and repair the leak.

_ _ _ _ - - _ _ - F . 11 No violations or deviations were identified. c. On October 10, 1985, the licensee conducted Surveillance Instruction, SI-82.2 as part of a post modification test to restore radiation monitors 2-RM-90-106B and -112B to service. Work plan 11793 had been written to incorperate changes descriosd in the Engineering Change Notice 5198 and Field Change Request 3785. The maintenance consisted of a modification to an electrical ground point. The Instrument Technician placed switch HS-90-136A in the block position on Unit 2 and then inserted a test signal into the Unit I circuit in error. This action resulted in a containment ventilation isolation. Failure to adequately implement SI-82.2 is a further example of violation 327,328/85-35-01, d. On October 31, 1985, while transferring start bus.1B from normal to alternate supply, the alternate breaker faiied to latch. This resulted in a loss of power to the 1A Shutdown Board and a start signal to the diesel generators. Two of the diesel generators started; the other two diesel generators were out of service for maintenance. The licensee attributed the failure of the alternate breaker to mechanical binding at the end of travel resulting in the failure to latch. The breaker was subsequently relatched; however, the licensee stated that main- tenance would be performed on the breaker to investigate the problem. In conjunction with this failure, a "B" Train Auxiliary Building Isolation occurred due to loss of power to spent fuel pool monitor 0-RM-90-103. This monitor is required to operate to prevent a release of radioactive material from the Auxiliary Building in the event of a fuel handling accident in the spent fuel pool. As identified on drawing PL J281-53 the monitor should have been powered from a Train B power source inside the radiation monitor cabinet. The licensee determined that the monitor was plugged into a nonessential power source. The inspector reviewed MR A-530620, which the licensee identified as the latest maintenance involving unplugging of the power source. The MR required maintenance to be done in accordance with Instrument Maintenance Instruction IMI-134, Configuration Centrol of Instrument Maintenance Activities. This procedure required the use of a configuration control sheet to assure that equipment was returned to its proper orientation. The requirements for use of the configuration control sheet were not properly followed in that the sheet did not identify the specific plug mold from which the monitor was unplugged and returned. Failure to implement configuration control procedures is a.further example of violation 327, 328/85-35-01, 11. Inspector Followup Items (92701) Based on inspection activities in the affected functional areas the following items were determined to require no additional specific followup -

,- . .. 12 and are closed. Discussions were held with the licensee with regard to the tinieliness of corrective actions. 83-23-04 (units 1 and 2) 84-11-03 (unit 1) 12. Review of Part 21 Reports (36100) a. The inspector reviewed a 10 CFR Part 21 report, provided to the NRC in a letter dated March 13, 1984, on Brown Bovari Corporation Type ITE-27N undervoltage sensing relays. Correction . of the design deficiency required replacement of a 100 kilchm resistor with a 200 kilohm resister on fourteen relays provided to Sequoyah. The inspector reviewed MRs A-082428, A-082427, A-082426 and A-082424 which replaced 12 of the resistors on the subject relays which are utilized for undervoltage protection on the 6.9 KV shutdown boards. The inspector randomly selected six of the relays and verified replacement of the resistors. Two additional relays maintained at replacement parts were also verified to be modified. This item, identified as 327, 328/P21-85-03 is closed. b. The inspector reviewed a 10 CFR Part 21 report, provided to the NRC on June 15, 1984, on the use of Crawford Fitting Company Swagelock fittings. Crawford Fitting Company determined that this issue was not of safety concern as documented in their November 16, 1984 letter to the NRC. This item, identified as 327,328/P21-85-02, is closed. Note that vendor recommendations on the use of Swagelock fittings was reviewed in Inspection Report 327/85-27, 328/85-28 and an Inspector Followup Item was left open regarding the licensee's evaluation of high pressure seal fitting adequacy. 13. Refueling Activities (60710) Unit 1 began removing fuel from the reactor for the Cycle 4 fuel load on October 23, 1985. Reload of the core was in progress at the end of this inspection report period. The inspector observed preparations for refueling, fuel handling operations in containment and in the spent fuel pool, movement of thimble plugs and rod cluster control assemblies in the spent fuel pit, and other ongoing activities associated with the rifueling. The inspector verified that_ selected Technical Specification requirements were met, that appropriate procedures were being utilized, that containment integrity was being maintained, that housekeeping and control of materials entering containment was adequate and that staffing was in accordance with the Technical Specification requirements. The following documer,:s were reviewed: Fuel Handling Instruction FHI-5, RCC Change Fixture Fuel Handling Instruction FH'.-6, Preparation for Refueling Fuel Handling Instruction FHI-7, Refueling Operation -

. - .__ __ _ _ _ _ _ - _ _ . . .. A 13 Fuel Handling Instruction FHI-13, Burnable Poison Rod Assembly Handling Tool Fuel Handling Instruction FHI-14, Thimble Plug Handling Tool Fuel. Handling Instruction FHI-17, Rod Cluster Control Change Tool Administrative Instruction AI-26, Prevention of Foreign Material in the Primary System - Restart Test Instruction (RTI)-2, Core Loading Technical Instruction (TI)-1, SNM Control and Accountability System No violations or deviations were identified. ' 14. Inspection Plan for Followup of Sequoyah Nonconformance Report A staff review was conducted, by a team of NRR technical reviewers and Region II personnel,.of the management processes involved in the resolution of Nonconformance Report (NCR) SQNNEB 8501 and its associated Failure Evaluation Engineering -Report (FEER). Attendant to this staff review, selected NCRs and FEERs were collected for additional evaluation. As a result of this additional review several cases were identified where potential safety questions were raised. Safety Evaluations were made by the staff for each safety question and required inspection effort was identified in a staff memo (Verrelli et al to Denton) dated August 9, 1985. An ' inspection plan for followup of the Sequoyah NCR open concerns was established by Region II in a staff memo (Weise to Walker) dated September 23, 1985, that identified several items which required resident inspector followup. The status of those items which required resident inspector followup is indicated below: a. NCR SQN CEB 8406 involved two air clean up units that were not welded to their steel supports in accordance with TVA. drawing 48N726. The welds were later upgraded to the requirements of drawing 48N726 under Maintenance Request A236959. The welding discrepancy was an undersized weld which was later determined to have been a temporary fit-up weld that should have been replaced with a permanent weld after installation. The licensee inspected all applicable welds in the mechanical equipment room and identified no other welds which were undersized. These particular welds, because of their temporary nature, did not have strike numbers or other means with which to identify the crew that performed the welds. The licensee's corrective action appeared to be adequate in this instance, and this item is closed, b. NCR SQN EEB 8406 involved some Class 1E 480 volt switchgear breakers and motor control center molded case circuit breakers which could be subjected to fault currents beyond their design capability. A FEER was issued by the licensee identifying this condition as a Category III. A Category III indicates that a component is unable to perform its required design function unless corrective modifications are made. Subsequently a safety evaluation was performed and found that the condition did not impact the safety of the plant and that no operational limitations were required. As a result of staff review it - - - -

p O . ~14 was determined that certain aspects of the FEER were deficient and the licensee committed to revise the NCR. The inspector obtained a copy of -the revised NCR 'and transmitted it to the appropriate Region II personnel. -In addition, it appeared that the original NCR was written before a calculated load study was completed and there was. no i statistical validity for the assumptions made in the FEER. As a result of the~ revised NCR, this item was reduced. in condition to a Category I, acceptable for all modes of operation and design conditions. For the . purpose of this. inspection, this item is considered closed. c. NCR SQN NEB 8407 involved eight Class IE radiation monitors which had been miswired or had their identification tags interchanged. This item was the subject of ' Region II enforcement action (327,327/84-38). The . licensee's response to this enforcement action was reviewed by the inspector. A team inspection is planned to address the NRC order EA 85-49 which will include a review of the licensee's NCR corrective ' actions. After the team inspection is complete the inspector. will review the licensee's corrective actions for the previous violation. For the purposes of this review plan, this item is closed. d. NCR SQN NEB 8408 involved a relative humidity control component which could fail as a result of high radiation during a reactor accident. The licensee's resolution to this issue was to allow the relative humidity heater to energize when the fan starts and reaches full speed. The relative humidity control component would be used for alarm purposes only. A ' review - of the adequacy of .TS surveillance was conducted by reviewing Surveillance Instructions SI-141 and -142 and . Technical Instruction TI-9. The surveillances conducted on the Emergency Gas Treatment System appear to be adequate.. This issue is closed. e. NCR SQN EEB 8412 involved Bettis Actuators with potential deficiencies. This issue was resolved in Inspection Report 327,3P8/85-26. f. NCR SQN NEB 8413 involved a discrepancy between the as found spent fuel pool alignment and that alignment described in the FSAR. A review of the reportablity aspects of this issue was conducted, and the issue was determined not to be reportable. An update was made on the most recent FSAR amendment submittal by the licensee to reflect current spent fuel . pool alignment. A review of the established makeup sources and applicable procedures, System Operating Instructions 501-70.1 and -78.1 and Abnormal Operating Instruction A0I-15, was conducted. The procedures and system alignments appear to be adequate and in compliance with TS. This issue is closed. _. }}