IR 05000327/1986032
| ML20215F481 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 09/05/1986 |
| From: | Stadler S, Wilson B NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20215F473 | List: |
| References | |
| 50-327-86-32, 50-328-86-32, NUDOCS 8610160233 | |
| Download: ML20215F481 (28) | |
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o NUCLEAR REGULATORY COMMISSION
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.[" n - REGION ll y j 101 MARIETTA STREET, >* g ATLANTA, GEORGI A 30323
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-Report Nos.: 50-327/86-32 and 50-328/86-32 Licensee: Tennessee Valley Authority
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, 6N38 A Lookout Place 1101 Market Street
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E Chattanooga, TN 37402-2801
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Docket Nos.: 50-327 and 50-328 License Nos.: DPR-77 and DPR-79 Facility Name: Sequoyah 1 and 2-Inspection' Conducted: May 19-23.and July 7-11, 1986 Conference call May 23,.1986
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Lead Inspector: h
. Stadler f@-fi Date Signed Inspectors: W. Bearden D. Falconer P. Harmon C. Vanderniet
- K.~Poertner
'**P. Moore-
- J. Brady Accompanying Personnel: B. A. Wilson Approved by hg . .b B./A. Wilson, Acting Chief
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9Datie 5'[% Signed Operational Programs Section Division of Reactor Safety
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SUMMARY Scope: This special announced inspection involved the area of. surveillance procedure revie '
Results: Cne violation was identifie * Inspected week of May 19-23 and July 7-11, 1986
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REPORT DETAILS Persons Contacted Licensee Employees
- P. R. Wallace, Plant Manager
- B. M. Patterson, Superintendent Maintenance
- R. W. Fortenberry, Engineering Supervisor
- G. B. Kirk, Compliance Supervisor
- J. M. Anthony, Office Supervisor
- H. D. Elkins, Instrument Maintenance Supervisor
- C. A. Crownover, QA Evaluator
- M. A. Cooper, Mechanical Engineer
- R. C. Birchell, Mechanical Engineer Compliance
- R. K. Gladney, Instrument Maintenance Engineering Supervisor
- L. M. Nobles, Superintendent Operations
- R. E. Thompson, Assistant Branch Chief
- W. E. Andrews, Site Quality Manager
- D. H. Tullis, Maintenance Special Projects
- C. W. LaFever, Instrument Engineer
- J. Blankenship, Manager Information Services
- K. L. Weller, System Engineer
- R. V. Pierce, Mechanical Maintenance Supervisor
,_ **H. Abercromoe, Site Director Other licensee employees contacted included engineers, technicians, operators, mechanics, and office personne NRC Resident Inspectors
- K. Jenison, Senior Resident Inspector
- L. Watson, Resident Inspector
- P. Harmon
- D. Loveless
- Attended exit interview May 23, 1986
- Attended exit interview July 11, 1986 Exit Interview -
The inspection scope "hnd findings were summarized on May 23 and July 11, 1986, with those persons indicated in paragraph 1 abov The inspector described the areas inspected and discussed in detail the inspection findings listed below. One violation was identified for failure to follow procedures in the conduct of a surveillance test, and is discussed in Paragraph No dissenting comments were received from the license The licensee did not identify as proprietary any of the materials provided to or reviewed by the inspectors during this inspectio . Licensee Action on Previous Enforcement Matters This subject was not addressed in the inspectio ,.
s Unresolved Items Unresolved items are matters cbout which more information is required to determine whether they are acceptable or may involve violations or devia-tions. Two new resolved items identified during this inspection are discussed in paragraphs 79 and.7 . Licensee's Surveillance Review Program The licensee's efforts to review and upgrade Sequoyah surveillance proce-dures have been divided into two phases; Phase I consisting primarily of a verification cross-check of technical specification surveillance require-ments to plant surveillance procedures, and Phase II consisting o detailed technical review of all plant surveillance procedures to be completed in conjunction with their two year administrative review cycle During the week of May 19, 1986, the inspectors reviewed the efforts of the Technical Specifications Review Committee (TSRC) which was established by the licensee to perform the Phase I review of the surveillance procedures at the Sequoyah facility. TSRC consisted of four members including a senior reactor cperator, an instrument engineer, and two shift technical advisor The TSRC review of the Sequoyah's surveillance instructions (SI) was completed during the months of March and April 1986, and the findings summarized in a report which was still in a draft format at the time of the inspectio The inspectors interviewed the TSRC chairman, who is an instrument engineer, regarding the SI review scope and purpose. The engineer indicated that the review was basically a cross-check to ensure that all facility Technical Specification surveillance testing requirements were covered by established sis. He further indicated that a review of completed SI data sheets was not included within the original scope of the Phase I review. The TSRC did, however, review a limited number of completed sis where there appeared to be obvious problems with test methods or result This Phase I review of Sequoyah sis appeared to be effective within the narrow scope defined. In most cases, the adequacy of an SI in meeting the intent of a specific Technical Specification surveillance testing require-ment was not determined as part of this initial review. As a result of the Phase I SI review, however, the licensee independently identified a number of instances in which Technical Specification surveillance requirements were not adequately incorporated into existing sis, or where an SI was not established to cover a particular surveillance requirement. These defici-encies resulted in the licensee's generation of several licensee event reports (LERs) defining the identified deficiencies and the planned corrective action The Technical Review Checklist established by the licensee to conduct the Phase II review of Sequoyah sis appeared to be very comprehensive and included a human factors review of the procedures and data sheets. One noted exception was the lack of requirement to ensure that equipment and a
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instruments utilized in surveillance testing were adequately labeled and in agreement with terminology in- the sis. The inspectors briefly reviewed several existing sis against this technical checklist and noted a signi-ficant number of deficiencies. The sis reviewed were not part-of the approximately 25 percent the licensee indicated had already been reviewed under Phase II. Deficiencies observed by the inspectors included . inadequate or subjective acceptance criteria, inconsistencies between equipment designa-tion in sis and actual plant labeling, signatures not required on data sheets, test equipment and calibration dates not specified, and a failure to specify actions required for failed surveillance tests or for as found data which does not meet Technical Specification requirements. The nature and number of these technical deficiencies indicate that the Phase II SI review, and the resultant corrective actions, represent a major undertaking by the license Since Phase II is scheduled to take two years to complete, consideration should be given to prioritizing the review and to the upgrad-ing of sis required to ensure safe restart and operation of the plan . Surveillance Requirement Tracking i The inspector reviewed the surveillance tracking system used at the facility. The system is a computerized management system that identifies upcoming surveillances and prints out a daily list called the Surveillance Available List. This list contains all surveillances that are scheduled for-performance on the date of the daily computer run, or previously scheduled surveillances which have not been completed. The daily computer printout is subdivided into sections for each of the licensee's organizational work groups that are responsible for specific surveillanc Each of these printout sections are sent to the respective work group on a daily basis for managerial . review and indication as to which of the surveillances are planned to be accomplished on the following day. Any modification to the printout or errors are also noted on the printout and it is returned to the Planning and Scheduling Section. The Planning and Scheduling Section then makes any changes to the surveillance printout as indicated and prepares a surveillance package for the sis that have been selected for completion on the following day. This section also prints a daily surveillance schedule based on the planned surveillances as indicated by each work group. This schedule appears to be beneficial in that each work group is aware of all planned surveillances facilitating coordination of manpower and tim Scheduling and Planning personnel indicated that the Operations Group had not responded to several requests to provide information on planned surveillances to be completed each day. As a result, operations surveil-lance tests have not been included on the daily surveillance schedule with the other work groups detracting from the intended coordination functio The surveillance tracking system printout has a comments section where the reason for delinquent surveillances can be stated. The comments section may also include additional warnings when surveillances are nearing the expira-tion of the time limits required by the Technical Specification The
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comments section of the printout appears to be inadequately utilized as several of the comments seem to be rather dated or vague in nature. This section of the printout could prove a useful tool for identifying problems related to the completion of si A separate list of delinquent surveillances is maintained by the Planning and Scheduling Section and printouts are sent out each week to the work group Once a surveillance is completed, the supervisor indicates this on the daily printout and the Planning and Scheduling Section will remove that surveil-lance from the lis Removal from the daily list does not mean the s,urveillance data have been reviewed, or that the SI package has been returned to the Planning and Scheduling Section, it simply means personnel have completed the SI. This results in the use of another computerized list titled " sis That Are Work Complete Data Not Received". The item remains on this list until the completed SI package is received by the Planning and Scheduling Section. The returning of completed SI packages can take up to six months or longe Much of this delay in the processing of completed SI packages appeared to occur while awaiting the final QA review. A licensee audit in May of 1985 (QSQ-A-0006) observed that in excess of 500 completed sis had accumulated awaiting QA review. The audit finding further noted that the SI packages were stored in an open shelf area providing no protection from damage or loss. During the conduct of this inspection over a year later, the inspectors observed approximately 300 sis stored in the same uncontrolled and unprotected manner. Although the licensee's procedures apparently do not consider these SI packages a QA related document until after the final review, the surveillance data contained in the packages is not duplicated elsewhere and would probably be unretrievable if lost, raising questions regarding system or equipment operability. The SI packages should be handled and stored as QA documents including orderly storage in locked fire resistant file cabinet In addition, the licensee should consider increasing the QA staff responsible for reviewing completed SI packages in order to expedite the reviews and reduce the backlog and storage problem Resolution of this deficiency will be tracked under inspector followup item (327,328/86-41-02). Surveillance Procedure Review The inspectors reviewed, observed or walked through portions of randomly selected surveillance related procedures to verify that plant sis were technically adequate and would accomplish Technical Specification surveil-lance requirements. Several completed surveillance procedure data packages were also reviewed to verify that surveillance testing had been accomplished at required frequencies, and that specified acceptance criteria had been met. As part of this inspection the inspectors selected several sis at
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Sequoyah that -were equivalent to those found deficient at the Watts Bar facility during previous SI inspections. The inspection findings at-Watts
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Bar are presented in NRC Inspection Reports 50-390/84-73, 50-390/85-21, 50-390/85-32, and 50-390/85-51. The inspectors reviewed these sis to determine if the deficiencies identified at Watts Bar existed at Sequoya ,
During this review, numerous examples were' identified in which - the licensee's controlled copies of Piping and Instrument Diagrams (P& ids) were found to be difficult to read -and interpre In some instances, nomenclature such as instrument numbers were illegibl Portions of the following str.veillance related procedures were reviewed by the inspection team. Those sis noted with an asterisk were equivalent to sis found deficient at Watts Ba Except as otherwise noted, the SI deficiencies at Watts Bar did not exist or had been previously corrected at Sequoya Surveillance Instruction Title SI-1 Surveillance Program SI-2 Shift Log SI-3 Daily, Weekly, and Monthly Logs SI-5 Auxiliary Feedwater Valves Position Verification
- SI-7 Electrical Power Systems: Diesel Generators
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- SI- Diesel Generator AC Electrical Power Source Operability Verification SI-26. Loss of Offsite Power With Safety Injection - D/G 1A-A CNTMT ISOL TEST SI-26. Loss of Offsite Power With Safety Injection - D/G 1B-B TEST SI-40 Centrifugal charging pump SI-4 Essential Raw Cooling Water Pumps SI-46 Component Cooling Water Pumps SI-52 Monthly Chemistry Requirements SI-54 Reactor Coolant E Determination
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SI-68 Containment Spray Valve Exercising
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Surveillance Instruction Title (Cont'd)
SI-71 Dissolved Air and Nitrogen Concentration In Upper Head Injection Water Filled Accumulator SI-9 Channel Calibration for ESF Instrumenta-tion RWST (18 Months)
- SI-100 Vital Battery Operability
- SI-102 M/M, Inspection of Diesel Generators SI-116 Quarterly Chemistry Requirements on Diesel Generator Fuel Oil SI-11 Turbine Driven Auxiliary Feedwater Pump and Valve Automatic Actuation SI-128 Residual Heat Removal Pumps SI-129 Safety Injection Pumps SI-13 RCS Controlled L'eakage
- SI-144.1/2 Testing of Control Room Emergency Ventilating (CREV) System SI-184 Periodic Calibration of Upper Head Injection System
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SI-187 Containment Inspection SI-196 Level Switch Calibration and Valve Closure Response Time Test of UHI System Hydraulic Isolation Valves SI-22 Unit 1 Load Sequence Timer Functional Test
- SI-23 Diesel Generator Battery System Weekly Inspection 51-252 Diesel Generator Interdependence Test
- SI-26 RHR Injection Flow Measurement, Pump Performance and Check Valve Test
- SI-26 BIT Cold Leg Injection Flow Balance, Pump Performance and Check Valve Test
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I ~ Surveillance Instruction Title I ~ (Cont'd)
-SI-260.31 SIS Cold Leg Injection Flow Balance, Pump p Performance and Check Valve Test l SI-273 Diesel Generator Lockout Features i
Verification l
i SI-28 Calibration of Flux Imbalance Penalty
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Function Generator l' SI-295 Diesel Fuel Receipt SI-401 Steam Generator Blowdown-Controlled Release SI-40 Radioactive Gaseous Waste Effluent -
Particulate and Iodine release rates
! from Shield and Auxiliary Building l Ventilation SI-415 Gaseous Effluent Requirements; Gross, Alpha, Noble gas, and Tritium SI-501 120 Volt AC Preferred Inventers SI-566 . Emergency Raw Cooling Water (ERCW)~ Flow i
Verification Test i-
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!- Instrument Maintenance Instruction Title
- IMI-99-CC-11.5B Off Line Channel Calibration of AT/Tavg
! Channel I l-IMI-99-RT-1 Response Time Test of AT/Tavg Channel I IMI-99-?qT Safeguards Protection System Test
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System Operating
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Instruction Title
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SOI 1 Condensate Demineralizers S0I 15 Steam Generator Blowdown
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Technical Instructions Title TI 12 Radiological Analytical Methods TI 16 Sample Points and Sampling Methods TI 37 Chemical Laboratory, Log System TI 41-68 Scaling and Setpoint Document
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The inspectors identified the following deficiencies: , SI-1, Surveillance Program
- The inspectors noted that no SI exists to implement Technical Speci-fication surveillance requirements 4.3.3.9, item Sa from Table 4.3-8, Condensate Demineralizer Regenerate Effluent Liquid Radioactive Monitoring Channel Check and Channel Calibration. This item is listed as "to be supplied" in SI- This discrepancy was discovered by the licensee during QA Audit 86-0005 conducted during April 198 The licensee submitted Licensee Event Report (LER)86-023 on June 6, 1986, that committed to add the flow indicating rotometer to SI-198,
" Calibration of Compliance Instruments", SI-3, Daily, Weekly, and Monthly Logs Section 3.1.8.3, satisfies Technical Specification surveillance requirement 4.3.3.9 on daily channel checks of the condensate demin-eralizer effluent flow element. The instructions contained in. SI-3 state that' System Operating Instruction (SOI)-14 and/or S0I-15 should be used to estimate flow in the event of flow instrument inoperabilit Action statement 33 of Technical Specification 3.3.3.9 allows effluent release to continue via paths with inoperable flow instruments. provided that the flow rate is estimated every four hours. SOI-14 and/or S01-15 j contained no provisions for estimating flow through the condensate demineralizer effluent pat The inspector discussed this item with the licensee and determined that the required alternate method of determining flow is specified in SI-401 item 1.3. SI-5, Auxiliary Feedwater Valves Position Verification
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The inspectors performed a walkthrough of Surveillance Instruction SI-5 on Unit 2 and noted the following:
(1) Numerous manual isolation valves were out of the specified position; however, hold tags were in place controlling all valves except 2-67-718A, 3-875 and 3-876. Although valves 3-875 and 3-876 were not in '. heir procedurally specified positions, their
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positions were noted in the configuration log; however, the licensee was unable to identify administrative controls for valve 2-67-718 (2) Flow control valves FCV-1-15, 16, 17 and 18 are motor operated Limitorques with a standard manual hand wheel / clutch arrangemen SI-5 does not require locking the manual hand wheel to preclude manual operation of the valve at the motor operato (3) Air supplies to level control valves LCV-3-174 and LCV-3-175 are required by SI-5 to be verified; however local air regulator gauges are not provided and the auxiliary operator was unable to demonstrate how this verification was accomplishe SI-7, Electrical Power Systems: Diesel Generators During review of SI-7, Electrical Power Systems: Diesel Generators, the inspector determined that the informatien contained on page 21 had been deleted in error, and replaced with the information contained on page 23 during revision 34. Revision 34 to SI-7 had been PORC approved and was approved by the Plant Manager March 5,1986. Deleting the information contained on page 21 deleted procedural steps for starting diesel generator (D/G) 18-B using a simulated loss of offsite power signal, simulated loss of offsite power in conjunction with an ESF test signal, and safety injection test signal diesel star Review of completed SI-7 packages identified that SI-7 had been performed on diesel generator IB-B numerous times with this procedural inadequacy present indicating that the procedures are not being followed verbati The inspector identified two instances where the SI-7 data sheet indicated that D/G 1B-8 had been started by methods that ' were not procedurally present in the body of the procedure. The inspector identified this procedural discrepancy to the licensee and the licensee initiated steps to correct the error during the inspectio During the inspection effort, the inspector witnessed the performance of SI-7 on Diesel Generator 1B- The D/G was started manually which was covered in the procedure. It appeared, however, that the surveil-lance was performed using only the data sheet as a procedural guid During the performance of the surveillanca instruction the inspector did not see the operator performing the sui.7111ance refer to the body of the procedure at any tim It would apgear that this is a common practice based on the fact that SI-7 had been performed on D/G 18-B numerous times prior to the inspector reviewing the procedure, and the licensee had failed to recognize the procedural inadequacy. In fact, the diesel had been started twice by methods that had been deleted from the procedure in error during the revision proces Technical Specification 4.8.1.1.2.a.4 states: each diesel generator set shall be demonstrated operable on a staggered test basis by verifying the diesel star *,s from ambient condition and accelerates to at least 900 rpm in less than or equal to 10 seconds. The generator
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voltage and frequency shall be 6900 t 690 volts and 60 1.2 Hz within 10 seconds after the start signa The diesel generator shall be started for this test by using one of the following signals with startup on each signal verified at least once per 124 days:
(1) Manual (2) Simulated loss of offsite power (3) Simulated loss of offsite power in conjunction with an ESF actuation test signa (4) An ESF Actuation test signal This technical specification surveillance requirement is accomplished by SI-7, Electrical Power System: Diesel Generators. Review of SI-7 revealed that the diesel generators are not started by a simulated loss of power in conjunction with an ESF actuation test signal as required by T.S. 4.8.1.1.2.a. SI-7 contains a section on starting the 0/G by simulating a loss of offsite power in conjunction with a safety injection test signal, and the data sheet recognizes the simulated loss of offsite power in conjunction with a safety injection test signal diesel start signal. The section states, however, to start the Diesel by the simulated loss of offsite power starting method oj: by the Safety Injection actuation test signal starting method. The section contains a note that states:
In the D/G start circuits, there are two types of start signals, either of which will start the D/G. One signal is a safety injection signal and the other is shutdown board loss of voltage signal. There is no combination of safety injection and loss of voltage signal as suc Either of these signals (safety injection or loss of voltage) drop out relays which end up generating D/G start, and both signals simultaneous will accomplish no more than either signal independentl It appears Sequoyah incorporated the Standard Westinghouse Technical Specification surveillance requirements into their 0/G surveillance requirements during the licensing proces Newer Westinghouse plants contain circuitry to accomplish a simulated loss of offsite power in conjunction with an ESF signal diesel generator star The licensee has, at the present time, submitted a Technical Specifica-tion (TS) change to eliminate TS 4.8.1.1.2.a.4.c. Pending approval and incorporation of this change, this will be identified as an inspector followup item (327, 328/86-32-01).
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e. SI-26.1 A&B, Loss of Offsite Power with Safety Injection Technical Specification 4.8.1.1.2.d.4 requires that at least once per 18 months, during shutdown that the diesel generators be demonstrated operable by simulating a loss of offsite power by itself, and (1) Verifying de-energization of the shutdown boards and load shedding from the shutdown board (2) Vertfying the D/G starts on the auto-start signal, energizes the shutdown boards with permanently connected loads within 10 seconds, energizes the auto-connected shutdown loads through the load sequencers and operates for greater than or equal to 5 minutes while its generator is loaded with shutdown loads. After energization, the steady state voltage and frequency of the shutdown boards shall be maintained at 6900 t 690 volts and 60 t 1.2 Hz during this tes This TS surveillance requirement is supposedly accomplished on Unit 1 by SI-26.1A and 26.1B. Review of these sis revealed that this surveillance requirement was not accomplished by these sis, instead the requirement of T.S.4.8.1.1.2.d.7 was accomplished and the licensee took credit for the accomplishment of T.S.4.8.1.1.2.d.4 al so. T.S.4.8.1.1.2.d.7 demonstrates the D/G operable at least once per 18 months, during shutdown by simulating a loss of offsite power in conjunction with an ESF actuation test signal, and (a) Verifying de energization of the shutdown boards and load shedding from the shutdown board (b) Verifying the D/G ' starts from ambient conditions on the auto-start signal, energizes the shutdown boards with permanently connected loads within 10 seconds, energizes the auto-connected emergency (accident) loads through the load sequencers and operates for greater than or equal to five minutes while its generator is loaded with the emergency loads. After energization, the steady state voltage and frequency of the emergency busses shall be maintained at 6900 t 690 volts and 60 i 1.2 Hz during this tes (c) Verifying that all automatic D/G trips, except engine overspeed and generator differential, are automa'tically bypassed upon loss of voltage on the shutdown board and/or safety injection actuation signa During a loss of power in conjunction with a safety injection signal all the shutdown loads except the pressurizer heaters are sequenced onto the shutdown busses in addition to the ESF load The licensee should evaluate the requirements of TS 4.8.1.1.2. and 4.8.1.1.2.d.7 and either incorporate the requirements into the
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surveillance instructions or revise the technical specifications to recognize the actual surveillance testing conducte A subsequent review of SI 26.2A Revision 13, approved on July 9, 1986, indicated that the required testing of TS 4.8.1.1.2.d.4 is now incorporated. SI-26.1 A&B also require this revision and will be an inspector followup item (327,328/86-32-02).
f. SI-40, Centrifugal Charging Pump SI-40 Part C implements emergency core cooling systems (ECCS) pump and discharge pipe venting requirements for the centrifugal charging pump (TS 4.5.2.b.1). No acceptance criteria, i.e. steady stream of water issuing from vent or hose, to ensure pump priming is specified in either the SI or data sheet. This leaves it up to the judgement of the individual performing venting operation to determine when adequate venting is complete. The comment also applies to SI-128, Residual Heat Removal Pumps, and SI-129, Safety Injection Pumps. Implementation of an acceptance criteria will be an inspector followup item (327, 328/86-32-03).
g. SI-45.1, Essential Raw Cooling Water During the review of SI-45.1, the inspector examined all 1985 completed essential raw cooling water (ERCW) pump performance data sheet This portion of the review indicated a common practice of lining-out and initialing original data, and then recording new data on the same sheet without documenting why the changes were made. The inspector found several cases of this practice, with the most notable being the SI performed on January 25, 1985, on the KA ERCW Pump. In discussion with the Licensee the inspector questioned this practice and inquired as to procedures outlining how retaken data were to be handled. The licensee informed the inspector that no procedure was in effect, and that the rerunning of tests and the reentry of data was handled through the use of good ~ engineering practices. This response is considered inadequate as demonstrated by the heavily marked-up condition of the various data sheets the inspector reviewed. The inspector also reviewed several data sheets from SI-46, Component Cooling Water Pumps, and found the same lining-out method being used. This indicates that the problem is not isolated to the ERCW system. The failure to document SI data changes requires further review and will be identified as an unresolved item (327,328/86-32-04). SI-45.1 is intended to meet the requirements for inservice testing of ERCW Pumps as specified in the facility's Technical Specifications and FSAR. Sequoyah's Technical Specifications
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and the FSAR require inservice testing to be completed in accordance with the ASME Boiler and Pressure Vessel Code Section XI.Section XI contains requirements that must be met to insure that the ASME Class 1 ERCW pumps are in an operable condition.Section XI further provides
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a formula to be used in the calculation of the acceptable and allowable ranges of pump performances, and further stipulates the acceptance criteria for pump operability and the frequency for performing the surveillanc SI-45.1 Part A, ERCW pump performance
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data sheet, page 2 also states the acceptance criteria contained in the ASME code. One of these criteria states that if the pump data fall outside both the acceptable and allowable ranges, the pump is to be declared inoperable and, according to the ASME Code, is to be removed from servic On January 18, 1985, SI-45.1 was conducted on the KA ERCW pump and the results of this test showed the pump to be clearly inoperable. The SI was reviewed and signed by the Shift Engineer and KA ERCW pump should have been declared inoperable. The licensee however, continued to operate KA ERCW pump and no notes were made in the operator logs as to the status of the pump or the failed SI. On January 25, 1986, all the signatures on the original data sheets for KA ERCW pump were lined-out and new data was entered. The new data were the same as the original data taken on January 18, 1986 therefore, the pump was once again demonstrated inoperable according to the Licensee's procedure. The pump continued in operation and the SI data were retaken on January 25, 1986 approximately 30 minutes after the first retest. The ERCW loop flow, pump total flow and pump discharge pressure were again lined-out on the original data sheets and new readings entered. The newest data were significantly different from the other data sets, and when the calculations were completed, KA ERCW pump was found inside the allowable criteria but outside the acceptable criteria. This meant the pump was operable but had to have an increased surveillance frequenc It was only after the latest test that operator logs reflected the successful completion of SI-45.1 for KA ERCW pump. No documentation was provided on the test sheets that adequately explained the reasons for all the retesting, and why the final data were so much different from original dat During the week of July 7,1986, the inspector interviewed Operations and Engineering personnel to determine the circumstances surrounding the performance of SI-45.1 on KA ERCW pump January 25, 1985. The inspector determined that the fact that the KA ERCW pump test results were unsatisfactory on January 18, 1985 was not recognized by the Shift Engineer nor the Engineering and Test Mechanical Engineer who reviewed and signed the surveillance data sheet as reviewers. When the error was identified the surveillance was reperformed January 25, 1985. The inspector also determined that it was a licensee accepted practice for operators to throttle the discharge valve of the ERCW pumps to obtain satisfactory surveillance data when the original test results showed that the pumps did not meet acceptance criteri Throttling the discharge valve moved the flow and discharge pressure data points into the acceptable region of the pump performance curve. Review of SI-4 indicated that throttling the discharge valve was not part of the surveillance if unsatisfactory test results were initially obtaine Furthermore, the system operating instruction (501) 67.1 valve align-ment checklist requires the ERCW pump discharge valve to be in the full
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open position, not throttle Once the satisfactory test data were obtained, the pump discharge valve was reopened returning the flowrate and pump discharge pressure back to the original unsatisfactory condition Discussions with Operations and Engineering personnel indicated that the problem may be associated with the installed instrumentation used to obtain the data points, and not actually a pump performance proble The inspector identified a period from February 12-23 1985, where the KA ERCW pump was selected to auto start under accident conditions. The pump was technically inoperable during this' period based on the fact that the SI-45.1 surveillance conducted January 25, 1985, was initially unsatisfactory, and the discharge valve was subsequently throttled to obtain satisfactory data and then reopened. This failure to follow the requirements of SI-45.1 and TS 3.7.4 is identified as a violation (327, 328/86-32-05). The inspector indicated during the exit interview that the throttling and reopening of a discharge valve to obtain satisfactory test results is not an acceptable practice and based on the fact that this practice was considered by the licensee to be acceptable for all ERCW pumps, there may have been a period of time when all eight ERCW pumps were technically inoperabl The inspector reviewed the frequency of performing SI-45.1 during 1985 and found an adequate number of the surveillances had been completed to meet Technical Specifications requirement The inspector reviewed a small portion of the gauge calibration program at the facilit The program is implemented by sis, Maintenance Requests (MR), and Work Requests (WR). Each in process instrument has a data record card maintained in a file cabinet which contains the dates of instrument calibration and calibration dat These calibration data have only been maintained by the licensee as a record since 1985, and improves the reliability of the calibration program. A brief review of this gauge calibration record keeping system indicated it to be adequat h. SI-98.5, Channel Calibration for ESF Instrumentation RWST (18 months)
In order to assess the adequacy of the surveillance program for the Instrument Maintenance Section, four safety-related instruments were selected at random. Records for the selected instruments were reviewed to assure that appropriate surveillance testing had been performe This review included tracing the documentation accompanying each surveillance from scheduling through performance, data collection and various review processe .
Data sheets accompanying the surveillance instructions are used to record the as-found and the as-left instrument settings. Double verification of the data is used for those procedures that involve safety-related equipment. When the data sheet information is trans-cribed onto instrument history cards, no formal second review or
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verification is used to assure that the data are correctly transferre In one instance, incorrect calibration data were found on one of the cards of the selected instruments. This error involved LS-63-51A, a level switch that initiates ECCS changeover from reactor water storage tank (RWST) suction to containment recirculation sump suction following a loss of coolant accident (LOCA) and depletion of the RWST inventor The calibration data listed on the data sheet of the surveillance instruction, SI-98.5, were incorrectly transcribed to the card. The acceptable setpoint of the level switch is 0.2095 volts, with an allowed deviation of plus or minus .002 volts. The as-found setpoint of 0.2092 volts was correctly entered onto the card, but the as-left value was listed as 0.2139, a value clearly in excess of the allowable range. The error occurred when the transcriber entered the as-found reset value of 0.2139 from the line directly below. Although the official (SI data sheet) information was correct, transcribing errors should be prevented by standardizing the data sheet / card formats, or by establishing a review of the transfer proces While attempting to resolve the error on the card, the inspector was unable to locate the completed SI data package. The completion date was listed on the card as November 30, 1985, but the completed SI data package was apparently awaiting QA review and was not readily retrievable as addressed in Section i. SI-137.3, Measurement of the Controlled Leakage to the Reactor Coolant Pump Seals The inspector walked through this surveillance instruction with a licensed operator and had some concerns regarding the purpose and technical adequacy of the procedure. Specifically, the procedure is intended to meet the requirements of TS 4.4.6.2.1.c, which verifies .
that the controlled leakage to the reactor coolant pump seals is less than 40 gpm at an RCS pressure of 2235 psig. The basis for the specification is to ensure that in the event of a LOCA, the safety injection flow will not be less than assumed in the accident analysis, The Boron Injection Tank (BIT) flow rate minus the seal inspection flow rate is within acceptable limit Since the Sequoyah Technical Specification is consistent with Standard Westinghouse Technical Specification, the inspector contacted NRR to determine why an apparent ECCS related specification is contained in the RCS leakage section. Their explanation is that it is primarily for convenience in that the seal injection flow rate is factored into the calculation of RCS leakag With respect to the technical adequacy, SI-137.3 requires that the charging flow control valve, FCV-62-93, be fully opened and seal flow determined from each of the pum$ injection flow paths. There is an additional valve, FCV-62-89 in the normal charging header that provides a back pressure on the seal injection heade The surveillance
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instruction does not account for any changes in the position of this valve. For example, if subsequent to conducting SI-137.3, FCV-62-89 was throttled down by the operator, the seal injection flow rate could increase,-possibly in excess of the allowable limits set by the S In response to the inspectors concerns, TVA provided a letter from Westinghouse dated May 29, 1986. This was in response to a previous}
TVA request for re-evaluation of the ECCS Appendix K analysis tof'
reflect a seal injection flow rate of 10.0 gpm per pump. Westinghouse analyzed the seal water injection line resistances and flow values assuming a seal flow of 10.0 gpm per pump with a pressure drop of 100 psig and a total RCP seal flow rate of 74 gpm at pump runout condi-tions. They concluded that the ECCS analysis and Technical Specifica-tions do not require re-evaluatio However, this does not fully answer the inspectors concerns regarding the adequacy of the surveill-ance tes These concerns are:
(1) Should administrative or other restrictions be placed on FCV-63-89 such that test data are not later invalidated by changing the position of this valve?
(2) Following a safety injection signal, the normal charging path is isolated and the safety injection path through the BIT is opene Has TVA shown that this flow path offers less resistance than the charging flow path and therefore will not result in forcing more water through the seal injection line?
In a telecon with a licensee's engineering representative on August 8, 1986, the inspector was informed that if a minimum pressure difference of 100 psig is maintained between charging pressure (as measured by PT-62-92A) and RCS pressure, the position of FCV-62-89 is not relevan i This normal operating pressure difference will satisfy both concerns abov The licensee, however, was unsure if a procedural step requiring the operator to verify this pressure difference will be added to the S This item will therefore be carried as an laspector Followup Item (327, 328/86-32-06).
J. SI-144.1 and SI-144.2, Testing of Control Room Emergency Ventilation (CREV) System SI-144.1 requires testing control room isolation and the transfer of the CREV system to the recirculation mode on a safety injection signa SI-144.2 tests the same evolution on a high radiation signa Surveillance testing of the chlorine and smoke initiation signals are limited to alarms only. The actual CREV isolation and transfer to the recirculation mode is not verified. The lack of testing of isolation and recirculation on a chlorine signal was identified as an inspector followup item (327,328/85-46-06) during an inspection conducted in December 198 Sequoyah's FSAR Sections 6.4.1.2 and 6.4.2.1 assumed that the control room ventilation will isolate within ten seconds and transfer to the emergency mode on a toxic hazard. The toxic hazard
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utilized for the assumption was the rupture of a 600 pound chlorine tank underneath the control room ventilation intak Westinghouse Standard Technical Specification Section 4.7.7.e.5 requires that once per 18 months verification be performed to ensure that on a high chlorine (toxic gas signal), the CREV system automatically switches to the recirculation mode within 15 second CFR 50, Appendix A, Criterion 19 requires that a control room shall'be provided from which actions can be taken to operate the nuclear power unit safely under normal conditions and to maintain it in a safe condition under accident condition The licensee has not, as part of their surveillance testing program, separately tested the ability of a chlorine signal to isolate the CREV system within ten seconds as assumed in the FSAR, or the ability of a chlorine or smoke (toxic gas) signal to switch CREV to the recircula-tion mode as required in Westinghouse Standard Technical Specifica-tion In addition, Sequoyah Technical Specification 4.7.8. requires testing the initiation of the CREV system only under safety injection and high radiation signals. The Technical Specifications do not address chlorine or smoke signal testin In response to this concern, the licensee indicated they have drafted a request to the NRC to delete the chlorine initiation mode (/ rom the CREV system and to amend applicable sections of the FSAR. The basis for the request is that the chlorine on site has been replaced by a non-toxic chemical and that there are no chlorine barges or railroad tank cars passing within five miles of the facility. A similar request had been previously submitted by the Watts Bar facilit If this request is not granted, the licensee should revise the applicable surveillance instructions to require testing with chlorine and smoke signals to include timing of CREV isolation (5 ten seconds) and switching of CREV to the circulation l mode (515 seconds). In addition, a Technical Specification change
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request should be submitted to include surveillance testing require-ments for the chlorine and smoke signals. Pending resolution of this request to delete chlorine testing of CREV, IFI (327, 328/85-46-06) is l
closed, and the item will be carried as unresolved (327, 328/86-32-07).
I i SI-260.2 BIT Cold Leg Injection F. low Balance, Pump Performance and l Check Valve Test j SI 260.2 balances Boron Injection Tank (BIT) flows, strokes applicable check valves, and evaluates Centrifugal Charging Pump (CCP) performance to ensure that Technical Specification and minimum ECCS criteria are satisfie During the review of SI-260.2, the inspector noted that the CCP minimum ECCS head capacity curve provided in SI-260.2 was less conservative than the CCP minimum ECCS head capacity curve in FSAR figure 6.3.2-14 at flow rates greater than 450 gpm. This discrepancy was the result of the licensee submitting an incorrect head capacity curve as amendment three to the Segouyah FSAR on April 11, 198 .
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Apparently, the licansee had requested that Westinghouse reanalyze CCP ECCS performance based on the reduced flow rates due to reduced CCP performance identified in 198 The results of Westinghouse's reanalysis were transmitted to TVA on March 16, 1984 and revised the minimum CCP ECCS head capacity curve. SI-260.2 was subsequently revised to incorporate the new minimum CCP ECCS head capacity curve; however, an incorrect minimum CCP ECCS head capacity curve was submitted as amendment three to the FSA . SI-260.3, SIS Cold Leg Injection Flow Balance, Pump Performance and Check Valve Test SI-260.3 balances safety injection flow, strokes applicable check valves, and evaluates Safety Injection Pump performance to ensure that Technical Specifications and FSAR criteria are satisfie During the review of SI-260.3, the inspector noted a discrepancy between the Safety Injection Pump minimum ECCS head capacity curve SI-260.3 and the minimum ECCS head capacity curve in FSAR figure 6.3.2-16. Apparently, the head capacity curve in SI-260.3 had not been revised to reflect a Westinghouse reanalysis of Safety Injection pump ECCS performance transmitted to TVA on March 16, 198 The failure of the licensee to provide a timely revision to SI-26 did not appear technically significant, in that the original head capacity curve in SI-260.3 was more conservative than that based on the Westinghouse reanalysis; however, further review of the FSAR's minimum ECCS head capacity curve revealed that the head capacity curve submitted as amendment three to the FSAR did not cover the full range of Safety Injection Pump ECCS flow rates. Although the head capacity curve contained in the March 16, 1984 Westinghouse reanalysis included pump flow rates from 0 - 650 gpm, the head capacity curve submitted as amendment three to the FSAR only included flow rates from 285 to 648 gp Apparently, the engineer responsible for submitting the curves in the FSAR amendment used the raw data supplied in the Westinghouse report to generate the curves, as opposed to using the actual curves supplied by Westinghouse. The licensee plans to submit the correct curves in the next amendment to the FSAR. Until submitted and approved, this item will be identified as inspector followup item (327, 328/86-32-08).
m. SI-501, 120V AC Preferred Inverters The 120 volt AC preferred inverter surveillance testing is not required by technical specifications, but is listed by the licensee as equipment affecting Technical Specification equipment. SI-501 was reviewed by the inspector utilizing the licensees' SI technical review checklist which was established for the Phase II SI review. In addition, the inspector " walked through" the surveillance test with an operato . .
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Although, the inspector did not perform an in-depth technical review utilizing the technical review checklist, SI-501 clearly did not meet several of the checklist requirements. It should be noted that this SI had not yet been evaluated against this checklist by the licensee. SI review checklist items which appeared not to be met adequately by SI-501 included:
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Item 4: The frequency was not clearly defined but was listed as
"should" be performed every six month Item 9: The SI did not require verification that the other inverters are in the nontripped condition prior to testin Item 10: The SI did not notify the operator which local or control room annunciators will ligh Item 12: The SI did not require collecting "as left" data but only that the output voltage and amps are " normal".
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Item 13: The SI did not clearly define the acceptance criteria or what should occur if the acceptance criteria are not met. Section 4.0 defines acceptance criteria as the inverter operates satisfactory, and the frequency does not drift for >30 seconds without synchronizing potentia " Satisfactory" operation for two hours is not defined, i.e., how far can the voltage and frequency dri f t and still be considered satisfactor Satis-factory transfer is also not defined, nor is the amount the frequency could change' in 30 seconds without being considered " drift". Step 9 of the SI data sheet requires the operator to initial that the inverter output voltage and amps are normal, but does not define the band which is considered normal for each paramete Steps 14 and 15 require the operator to log the inverter output voltage, amps, and frequency after one hour and two . hours on DC, but does r.ot define what would be considered acceptable changes in these parameters during the two-hour test perio Item 20: The permanent plant equipment including voltmeters,
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ammeters, and switches were not identified by TVA l numbers in the S In addition, the terminology
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utilized in the SI to identify this equipment was in
- some cases different than the actual labels on the l equipment. This difference can lead to confusion and I operating errors. The licensee should consider adding a line item to the SI Review Checklist to ensure that the l procedure terminology and plant labeling agree and that
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Item 25: The allowable values are not clearly labeled in most areas of SI-501 and are stated simply as " satisfactory".
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Item 28: The information in the SI text did not match the corresponding item in the data sheet. The SI-501 text consists of a single page containing short paragraphs on ,
scope, precautions, and acceptance criteria. Section 3 l contains no actual instructions and refers the SI performer to the data shee Item 34: The SI did not require double verification of proper equipment restoration even though the inverters are identified by the licensee as affecting Technical Specification equipmen The walkthrough of SI-501 also indicated that the procedure was deficien The responsibility for performing this SI had recently been redesignated to the Assistant Shift Enginee Thus, although the operator performing the SI was not technically responsible, he was qualified by virtue of his completion of the prerequisite electrical Step III training. He appeared confused at times as to which instru-ment to read or which switch to operate. This confusion appeared to be directly attributable to a lack of consistency or specificity between procedural terminology and component labeling. The SI requires record-ing 250V AC battery charger amps but does not specify on what panel or what instrument. The 250V AC charger ammeters are not permanently labeled by name or TVA number, but do have hand written labeling beside the meters. The SI also requires the operator to manipulate the "480V AC supply to the inverter". The breaker's numerical designation is not specified nor the panel on which it is located. The operator went to
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4 several locations before deciding on a breaker labeled "480V AC input",
a different designation than that in the SI. SI-501 also requires the operator to manipulate 120V AC sychronization switches on distribution panels but does not indicate that these panels are located in the 120 volt vital battery room. Resolution of the identified deficiencies in SI-501 will be an inspector followup item (327, 328/86-32-09). SI-566, ERCW Flow Verification Test The inspector reviewed the data from an SI-566 ERCW flow test completed on May 21, 1985. Sectiori 5.1 of the procedure requires that all differential pressure (DP) gauges used in the test be calibrated and listed on data sheet 2. Several days af ter the successful completion of the surveillance test, pressure differential gauge 83618 was tested and determined to be out of calibration. This differential pressure gauge was utilized to verify adequate ERCW flow to the SIS pump room cooler and oil cooler 2B under Unit 1 in hot standby and Unit 2 under LOCA conditions. The calibration error was large enough to invalidate the DP and flow obtained, but the licensee was reluctant to rerun the entire surveillance due to the considerable time and effort required to reestablish test condition Engineering judgement was utilized to
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determine that increasing the ERCW flow to the SIS pump room cooler and oil cooler 28 would provide enough " cushion" to cover the out of tolerance range of the gauge and result in a successful test and would not adversely affect other safety related ERCW cooling load On May 30, 1985, a procedure change to SI-566 was written and PORC approved to allow changing the throttle valve position to add two GPM flow to the SIS room cooler and oil cooler 28. It was noted by the inspector that the revised data sheet utilized on May 30 documented a differential pressure of 8.0 inches and a flow of 10.9 GPM before the flow adjustment, and 11.1 inches and 12.9 GPM following the adjustmen These numbers were much lower than the differential pressure of 8 inches and flow of 35.7 GPM listed as acceptable on the original data sheet. The licensee indicated that the numbers were different because the " loss of dam" conditions simulated during the full surveillance test were not simulated for the revised data sheet and flow verification. This item was initially identified by the inspector as an unresolved item pending determination if the revised data sheet was adequate or whether the surveillance should be ~ rerun. The major concern was whether the engineering evaluation was adequate, since the calculations were apparently undocumented, and also the validity of the revised data sheet completed under conditions different from the original test, i.e., no loss of dam. At the exit on Friday, May 23, 1986, the licensee informed the inspector that a subsequent recalibration had indicated that the original calibration which indicated the differential gauge to be out of tolerance had been in error and provided a supporting lette In a subsequent telephone conversation the licensee retracted the statment made at the exi During the week of July 7,1986, the inspector reviewed the engineering determination and subsequent decision to increase the flow by 2 GPM through the safety injection pump room and oil coolers as opposed to performing the entire SI-566 ove SI-566, Step 4.1 contains the following precaution: " Care must be taken not to adjust a throttle valve that has been set. Should it be necessary to adjust a previously set valve all previous system balances in which flow through this branch was measured must be repeated." The only documented engineering evaluation is contained in the guage out of calibration report. This evaluation only documented the fact that increasing the flow through the safety injection pump room and oil coolers by 2 GPM at normal operating conditions would meet the acceptance criteria for these components under test condition No engineering evaluation is documented for how this ficw increase would affect the other compo-nents, particularly the RHR pump room cooler and the core spray pump room cooler. In addition, no attempt was made in the performance of the temporary change to SI-566 to determine if increasing the flow through the safety injection pump room cooler and oil cooler had any effect on the other components in the branch. Discussions with the engineers that performed the surveillance instruction and the temporary change to increase the flowrate by 2 GPM, determined that they
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considered this small increase in flowrate to these components would not affect the performance of the other component This failure to adequately verify that the ERC'd system met the requirements of SI-566 is identified as an inspector concern. The licensee should review the process by which unsatisfactory tests are accepted using engineering evaluation IMI-99-CC-11.58, Off Line Calibration of AT/Tavg Channel I, Instrument Maintenance Instruction (IMI)-99-CC-11.5B accomplishes the 18 month refueling frequency channel calibration of the Overpower Delta Temperature (0PDT) and the Overtemperature Delta Temperature (0 TDT)
, reactor trip channels as required by Technical Specification Surveil-lance Requirement (4.3.2.1.1, Table 4.3-2). A review of this instruc-tion indicated that the licensee does not perform a calibration verification of the dynamic functional response of the OPDT and OTDT channels as detailed belo Technical Specification Table 2.2-1, Reactor Trip System, Instrumentation Trip Setpoints specifies the OPDT and 0 TDT reactor trip setpoints as follows:
- 3 Overpower AT ( 1 ) s ATo (K. - K, ( )( 1 ) T -K'[T ( 1 ) - T"] - f (AI))
1+2S 1 + + + Overtemperature AT ( 1 ) s ATo (K -K 2 (1 + T2 S) [T(
i 1 )-T'] + K (P-P') - f (AI))
3 i 1+t 2S 1+t 2S 1+ Where:
AT. = Indicated AT at Rated Thermal Power T = Average Temperature ( F)
T' < 578.2 F T" E !adicated Tavg P = Pressurizer Pressure (Psig)
P' = 2235 Psig t = Time Constant (seconds)
K1 s 1.15 K2 = 0.011 K3 = 0.00055 K. s 1.087 K. = 0.02/*F or increasing average temperature and 0 for decreasing average temperature K. = 0.0011 for T>T" and K. = 0 for TsT" f( AI) = function of delta flux S = Laplace transform operator (Sec 1) -
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The dynamic terms in the OPDT and 0 TDT trip equations compensate for inherent instrument delays and piping lags between the reactor and the loop temperature sensor (PJ0s). Lead / lag and rate / lag compensations are utilized to accomplish these corrections; however, due to the licensee's testing method, these dynamic responses are not verifie The method utilized in IMI-99-CC-11.58 performs only a calibration verification of the static terms in the above equation Performing only static calibration verification effectively reduces, the OTDT and OPDT equations as follows:
Overpower AT 5 AT, (Ks - K.(T-T") - f2 (AI))
Overtemperature AT s AT, (K i -K 2 (T - T') + K3 (P-P')-f (AI))
t Specifically, the licensee does not verify that K, = 0.02/ F for increasing average temperature and 0F for decreasing average temperature. Additionally, this calibration method does not verify the response of components utilized for the lag ( 1 ), rate lag ( TS )~and lead lag ( 1+ts)
(1+t) (1+tS) (1+toS).
Technical Specification Surveillance Requirements 4.3.1.1.1, Table 4.3-1 requires that the OPDT and 0 TDT reactor trip system instrumentation channels shall be demonstrated operable by the performance of a channel calibration at least once per 18 month Technical Specification Definition 1.3 defines a CHANNEL CALIBRATION as the adjustment, as necessary, of the channel output such that it responds with the necessary range and accuracy to known values of the parameter which the channel monitors. The CHANNEL CALIBRATION shall encompass the entire channel including the sensor and alarm and/or trip functions, and shall include the CHANNEL FUNCTIONAL TEST. The CHANNEL CALIBRATION may be performed by any series of sequential, overlapping or total channel steps such that the entire channel is calibrate Contrary to the above requirements, the licensee's CHANNEL CALIBRATION method specified by IMI-99-CC 11.5B for the OPDT and 0 TDT reactor trip channels does not ensure that the channels respond with the necessary range and accuracy; in that the procedure method does not include a dynamic response of the OPDT and OTDT channel The licensee's test engineer discussed with the inspector a rough draft revision for IMI-99-CC-11.58. The revision appears to adequately address the dynamic response of the OPDT and 0 TDT channels. Pending a review of the revision to this procedure, this is identified as inspector followup item (327, 328/86-32-10).
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p. IMI-99-RT-11.5, Response Time Test of AT/Tavg Channel I Instrument Maintenance Instruction (IMf)-99-RT-11.5 performs the response time testing of the Overtemperature Delta Temperature (0 TDT),
trip and the Low-Low Average Temperature (Tavg) Safeguards actuation as required by Technical Specifications Surveillance Requirements 4.3.1.1.3, Table 3.3-2 and 4.3.2.1.3, and Table 3.3-4, respectively. A review of this instruction indicated that IMI-99-RT-11.5 does not
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perform a complete response time test of these safety functions as detailed below:
0 TDT has five channel sensors: Nuclear flux lower ion charr5er, nuclear flux upper ton chamber, pressurizer pressure, cold leg temperature and hot leg - temperature. The response time test method in IMI-99-RT-1 consists of maintaining constant inputs from the nuclear flux lower ion chamber, nuclear flux upper ion chamber, pressurizer pressure, and cold leg temperature and varying hot leg temperature input in a step function which results in the actuation of the OTDT tri Response times are not obtained from when the other four channel sensors exceed the trip set poin The temperature portion of the low-low Tavg coincident with steam line flow-high safeguards actuation has two channel sensors: Hot _ leg temperature and. cold leg temperature. The response time test method in IMI-99-RT-11.5 consists of maintaining the T cold input constant and varying T hot in a step function which results in actuation of Low-Low Tavg bistable. A response time is not obtained from the cold leg temperature channel senso The licensee stated that the low-low Tavg response time test method is technically adequath because there are no components in the cold leg temperature sensor loop of the Tavg circuit which could fail in a manner resulting in a slower channel response. The licensee did concur that the OTDT response time test is p_otentially inadequate due to the lead-lag components in this circuit and that this problem had been previously identified by I&C engineers and was under evaluatio Technical Specification Definition 1.24 states that REACTOR TRIP FYSTEM RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its trip setpoint at the channel sensor until loss of stationary gripper coil voltag Contrary to the above, the licensee's response time test method does not time the interval from when all monitored parameters of OTDT and low-low Tavg exceed their trip setpoint at the channel sensor. As a result of this deficiency, the longest channel response time may not be ,
compared to the technical specification acceptance criteri Discussions with a test engineer and review of a draft revision to IMI-99-RT-11.5 revealed that the licensee understands the concern and is taking steps to address i Specifically, the instruction will
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consist of a complete test of each lead / lag circuit for all inputs with the.most conservative value used to meet the TS requirement. This test will be performed once every three cycles for each channel. Pending review of an approved revision of IMI-99-RT-11.5, this will be an inspector followup item (327, 328/86-32-11). TI 41-68, Scaling and Setpoint Document Technical Instruction (TI) 41-68 provides an approved document pertaining to scaling and setpoint information' for plant instrumenta-tion. During the review of selected scaling information for Overpower Delta Temperature and Overtemperature Delta Temperature instrumenta-tion, the inspector noted the following discrepancies in TI 41-68 scaling data sheets:
(1) RCS Loop 1 Overpower AT setpoint calculator (TY-411M)
The scaling data sheet indicated that the K. constant should equal 38.93 ma which corresponds to 108.5% rated power. IMI-99-CC-11.5B calibrates this constant to 38.99 ma. corresponding to 108.7%
rated power. A review of Technical Specification Table 2.2-1 indicated the K. = 1.087 (108.7%) or (38.99 ma) is the correct scaling facto ,
(2) RSC Loop 1 Overtemperature AT Setpoint Calculator (TY-411L)
The scaling data sheet indicated that Ki should equal 40.4 ma which corresponds to 114% rated power. IMI-99-CC-11.58 cali' orates this constant to 115% rated powe A review of Technical Specification Table 2.2-1 indicates that K = 1.15 (115%) or (4 ma) is the correct scaling facto (3) RCS Loop 1 Overtemperature AT, f t (AQ)(NY-411A)
The scaling data sheet scales NY-411A f (AQ) as follows:
t (a) for AQ between -30 and +4 percent f (AQ)=0 t
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(b) for each percent that the magnitude of AQ exceeds -30 percent of f (AQ) is increased by 0.89 percent of rated powe t (c) for each percent that the magnitude of AQ exceeds +4 percent, f (AQ) is increased by 0.8 percent of rated powe t Contrary to the above scaling information, SI-282.1, Calibration of Flux Imbalance Penalty Function Generator (FAQ), Revision 6 calibrates f (AQ) pursuant to Technical Specification Table 2.2-1 t
as follows:
(a) for AQ between -29 and +5 percent f (AQ)=0 t
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(b) for each percent that the magnitude of AQ exceeds -29 percent, f (AQ) is increased by 1.50 percent of rated powe (c) for each percent that the magnitude for AQ exceeds +5 percent, f t (AQ) is increased by 0.86 percent of rated power The scaling document data for items a. and b. above are conservative with respect to the Technical Specifications; however, portions of the scaling data in item c are clearly non-conservative with respect to Technical Specifications Table 2.2- Apparently, these scaling data sheets have not been revised since 1981. Although in each case above, the applicable calibration procedures provided accurate calibration information, TI-41-68 is an approved plant procedure and should receive timely and accurate revisio TI-41-68, Section 2, applicability, further emphasizes this need by stating that "This document represents the present status of scaling and setpoint information on plant instrumentation. It will be revised as soon as reasonably achievable to represent any approved changes in the precautions limitations and set point document, FSAR or the channel accuracy document."
A generic deficiency in the revision of TI-41 was identified by the licensees' QA staff as noted in corrective action report (SQ-CAR-85-08-013). The QA staff identified range and/or setpoints information for five instruments in TI-41 that did not agree with calibration tabulation The corrective action, to be completed by October 1986, included a complete review of all CSSC systems comparing TI-41, 47B601 series drawings and calibration cards. The inspectors consider their level of review adequate to correct the problems identified above with TI-41-6 Until completed this item will be identified as an inspector follow-up item (327, 328/86-32-12).
8. NRC/TVA Conference Call On June 23, 1986, a conference call was held between NRC Region II and TVA regarding concerns and potential enforcement items identified during this inspection of Sequoyah Surveillance Instructions. The licensee provided:the Region with a status update on the following items: Thi D/G testing requirements of Technical Specification 4.8.1.1.2..d.4 were not incorporated into SI-26.1A and SI-26.18 (Section 7e of Report). The licensee concurred with this finding and indicated that the requirements of T.S. 4.8.1.1.2. would be incorporated into SI-26.1A and SI-26.1 During the inspection conducted July 7-11, 1986, the inspector confirmed this testing requirement was incorporated and this item is close .
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b. The D/G start circuitry design at Sequoyah does not facilitate the testing required by T.S.4.8.1.1.2.a. (Section 7.d of this Report). The licensee concurred with this finding and plans to submit a T.S. change request. This will continue to be carried as an inspector followup item until the T.S. change is implemente c. The control room emergency ventilation system isolation (10 seconds) and realignment to the recirculation mode (15 seconds)
were not being tested on a chlorine signal in SI-144, and was not required in T.S.4.7.8.d.2 (Section 7.j of the Report). The licensee indicated that the chlorine at Sequoyah had been replaced by a non-toxic chemical and that an FSAR change request would be submitted. This will continue to be carried as an unresolved item pending resolution of the relief reques d. SI-137.3 for measurement of controlled leakage to the reactor coolant pump seals did not address the positioning of FCV-62-89 which controls backpressure on the seal injection heade (Section 7.i . of this report). The licensee indicated that a Westinghouse response to a previous TVA request for a re-evalua-tion of the ECCS Appendix K analysis determined that it would not be necessary to isolate this charging line provided a minimum 100 psid was maintained between the charging header and the RC Pending review of a procedure change to SI-137.3 to incorporate verification of the 100 psid, this will be carried as an inspector followup ite IMI-99-CC-11.58, Off Line Calibration of AT/TAVG Channel I, did not require calibration verification of the dynamic functional reference of the OPDT and 0 TDT channels. (Section 7.o of the report). The licensee indicated that this same issue was raised during an inspection at Watts Bar (390-85-32) and that the SI would be revised to incorporate this testin This will be carried as an inspector followup item pending resolutio IMI-99-RT-11.5, Response Time Test of AT/TAVG Channel I did not complete response time test for all channel sensors. (Section of this report). The licensee indicated that response checking all channel sensors is good engineering practice and intends to revise the procedure This will be carried as an inspector
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followup item pending revision of the test procedur SI deficiencies noted such as inadequate acceptance criteria and labeling per the Sequoyah SI Technical Review Checklist (Section 7.m of this report). The licensee indicated that the SI technical deficiencies would be resolved under Phase II of the SI revie Resolution of the technical deficiencies noted in SI-501 will be an inspector followup ite Violation for failure to declare ERCW pump inoperable and to remove from service following failed surveillance test (Section 7.g of this report). The licensee provided the following additional background information:
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On January 18, 1985, the KA ERCW pump failed the differential pressure acceptance criteria during a surveillance test but was not declared inoperable
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On January 25, 1985, the KA ERCW pump was retested and once again failed the surveillance. The operators throttled the discharge valve to meet the acceptance criteria and then reopened the valve. The SI data package was modified with data crossed-out with no accompanying explanatio Surveillance test data must be reviewed within 96 hour0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> i. Unresolved item for using out of calibration differential pressure gauge for ERCW flow test (Section 7.n of this report). The licensee indicated that the letter that was provided at the exit interview on May 23, 1986, was applicable to another case where this instrument was utilized. The letter had indicated that the determination that the differential gauge was out of calibration was in error, J. An apparent reluctance to utilize procedures during the conduct of surveillance tests (Section 7.d of the report)'. The licensee indicated that the SI data sheets would be upgraded to. include all applicable instructions, precautions, notes, et This would enable the operator to follow the procedure verbatim while completing the data sheet without flipping back and forth between the procedure and the data sheet, Completed SI packages were not being properly stored or controlled while awaiting QA review (Section 6 of this report). The licensee indicated they believed that they had " tightened up" on this practice following the QA audit finding in May of 198 . The controlled copies of piping and instrumentation diagrams (P &
ID's) were found to be difficult to read and interpret (Section 7 of this report). The licensee acknowledged a problem with P&ID legibility in the past. They committed to provide acceptable p &
ID's in the control room prior to startup of the first unit. This will be tracked under inspector followup item (327, 328/86-37-07).
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