IR 05000382/1987021

Insp Rept 50-382/87-21 on 870914-18.Violations Noted.Major Areas Inspected:Followup of Previously Identified violations,10CFR55(e) Repts & Unresolved & Open Items

ML20236V138
Person / Time
Site:	Waterford
Issue date:	11/06/1987
From:	Haag R, Hunnicutt D, Stewart R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML20236V118	List: IR 05000382/1987021 ML20236V116 ML20236V127
References
50-382-87-21, NUDOCS 8712040104
Download: ML20236V138 (13)

Text

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APPENDIX B U.S. NUCLEAR REGULATORY COMMISSION

REGION IV

NRC Inspection Report: 50-382/87-21 License: NPF-38 Docket: 50-382 Licensee: Louisiana Power & Light Company (LP&L)

N-80 317 Baronne Street New Orleans, Louisiana 70160 Facility Name: Waterford Steam Electric Station, Unit 3 Inspection At: Taft, Louisiana Inspection Conducted: September 14-18, 1987 Inspector: $ wuruA<W D. M. Hunnicutt, Chief, Test Programs Section llbl$7 Dste'

Division of Reactor Safety

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l Q A ; } og // - 6 - 89 R. C. Stewart', Reactor Inspbctor, Materials Date and Quality Programs Section n f// N w 4&

at R. C. Haag, Reactor Inspector, Materials N k7 Date '

and Quality Programs Section Approved: f// jIes[eaf #/b/87 D. M. Hunnicutt, Chief, Test Programs Section Date Division of Reactor Safety EA236nBusjiga o i

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I Inspection Summary Inspection. Conducted September 14-18, 1987 (Report 50-382/87-21)

Areas Inspected: Routine, unannounced followup of previously identified violations, 10 CFR 50.55(e) reports, and unresolved and open item Results: Within the four areas inspected, one violation was identified (Failure to evaluate out-of-calibration test equipment in a timely manner, paragraph 3).

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l DETAILS Persons Contacted Principal Licensee Employees

• R. P. Barkhurst, Vice President, Nuclear

• N. S. Carns, Plant Manager

• G. F.'Koehler, Operations, Quality Assurance (QA)

B. G. Morrison, Licensing

• S. A. Alleman, Supervisor, Physical Health Technician

• J. E. Howard, Plant Engineer P. M. Melancon, Reactor Engineering and Performance G. M. Woodard, Nuclear Operations Support and Assessments

• J. McGaha, Operations and Maintenance

• E. Weller, 0perations, Licensing

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• D. Vinci, Maintenance

• J. J. Zabritski, Quality Assurance R. Brian, Staff Engineer

• Denotes those present at exit meetin The NRC inspectors also contacted other plant personnel, including operators, technicians, and administrative personne . Followup of Previously Identified Items (Closed) Violation (382/8501-03): Failure to Include or Address All Motor Manufacturer's Recommendations for Safety-Related Motors - The NRC inspectors reviewed Surveillance Procedure OP-904-012, Revision 1,

" Emergency Feedwater Pump Environmental Qualification Maintenance Input Check," dated November 20, 1986. This surveillance procedure addressed and closed the salient items listed in NRC Inspection Report 50-382/85-01, paragraph 3 as maintenance program item (Closed) Open item (382/8527-06): Possible Use of Preventive Maintenance (PM) Task Cards to Perform Corrective Maintenance - The NRC inspectors reviewed a sampling of PM task cards having a 3- and 5 year accomplishment periodicity that had two or more recorded accomplishments i in the first year of operation, with the following findings:

Partial accomplishments of PM tasks were entered in the licensee's maintenance information system as complete accomplishments, indicating erroneous multiple accomplishment In some cases, task cards were lost and replacement task cards issued. Subsequently, the licensee determined that both task cards were accomplished, and both were entered in the licensee's maintenance management information system.

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In some cases, upon recalibration,~ Measuring and Test l ~ Equipment (M&TE) used for PMs subsequently was determined to be'

out-of-calibration and the PM task was redone.-

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• In one' case', there were two identical pieces of equipment having a 5 year PM periodicity based on TS requirements. FSAR requirements

.were interpreted by the licensee to require PM on one of these two pieces of equipment'every 18 month *

The licensee's maintenance information management system (SIMS)

should alleviate unnecessary- redundant PM in the futur (Closed) Significant Construction Deficiency (SCD) No. 63, " Procurement of Spare and-Replacement Parts" (NRC Inspection Reports 50-382/84-35, paragraph 2 and 382/85-04, paragraph 2). The licensee determined that the apparent lack of documentation of appropriate technical and quality requirements of safety-related spare and replacement parts was not reportable under the, requirements of 10 CFR Part 50.55(e). LP&L"-

contracted an engineering consulting firm to review each safety-related, and/or QA-required, purchase order issued prior to July 1, 1982. Late LP&L expanded the review to include all purchase orders issued prior to July 1, 198 The NRC inspectors. determined that LP&L evaluated e'ach documented review performed by the consultant for final resolution and disposition. The LP&L procurement review indicated that the licensee did not require removing parts or components from the plant, did not invalidate the results of any startup testing, and did not identify any breakdown in the procurement administrative control progra The licensee resolved the potential deficiencies identified by the consultan LP&L Nuclear Operations QA Group audited the procurement activities on a scheduled basis and completed an audit of the procurement activities prior to initial fuel loadin LP&L reported this problem in their first interim report (LP&L Letter No. W382-0061).

(Closed) Violation (382/8631-01): Backup Spent Fuel Pool Cooling System Was Not Accomplished in Accordance with Instruction, Procedures, and Drawings - The NRC inspectors reviewed the four pipe support discrepancies between Ebasco Specification LOU-1564-100 and as-built conditions. The licensee had reinspected and corrected the four NRC discrepancies (Reference: CIWA-030338 and QC Inspection Report 86-106 (December 18, 1986) and CIWA-032292). The design group had been directed to provide more detail and clearer directions in future design drawings for those design packages that relate to installations of support (Closed) Violation (382/8631-02): Two Examples of Failure to Control .

the use of Nonconforming Com7onents - The NRC inspectors determined that ;

the licensee had revised LP&L Procedure MM-3-054, " Administrative '

Procedure Weld Control and Documentation," Revision 5, dated March 30, 1987, to incorporate weld specification document sheets to improve dispositioning'of item l

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l (Closed) Violation (382/8602-01): The Licensee Changed Operational Modes )

of the Reactor While in Reliance on an Action Statement of a Limiting Condition for Operation (LCO) and Violated Section 3.0.4 of the TS - The l NRC inspectors determined that the licensee had identified the root causes ,

of this violation. The causes were failure to control adequately the l condition and the status of remote valve operators and failure to follow the appropriate annunciator procedures. The licensee added an instruction in the Operation Night Orders cautioning operations personnel on the shortcomings associated with reach-rods and their use for independent verification of valve lineups. Operations personnel had compiled a list of manual valves which are operated by reach-rods. Operations personnel have placed warning tags on all accessible valves to instruct plant operators to verify actual valve positions when operating these valve The necessity of treating inoperable reach-rods as plant equipment requiring maintenance and initiating a CIWA to document the work was stressed to all operations personnel. A list of manual valves that are operated with reach-rods was included in Operating Procedure OP-7-005,

" Resin Waste Management," Revision 5, dated February 10, 198 A statement of caution (paragraph 5.5.4) was added to Operating Instruction 01-10-000, " Operations Department Good Operating Practices,"

Revision 5, dated May 4, 1987. The annunciator title had been changed to identify the valves being monitored. A review was conducted by the licensee to determine if other ambiguously titled annunicators were on the boards. Two annunicator windows were identified and corrective action complete (Closed) Violation (382/8628-01): License Procedure ME-7-002, Revision 5, Maintenance Procedure Molded-Case Circuit Breakers and Thermal Relays, Did Not Contain Required Controls for Temporary Removal and Replacement of Current Limiting Fuses During Performance Testing of the Circuit Breakers, Nor Reference Procedure UNT-5-004 - The NRC inspectors determined that the licensee had revised Procedure HE-7-002 to require removal of any shorting blocks that could have been previously installed, reinstallation of fuses, and independent verification of these items. The licensee had not found any instances of shorting devices in breakers removed from service for maintenance. The licensee reviewed all of the electrical maintenance procedures which cover circuit breaker testing and found no similar deficiencie (Closed) Observation (382/86-13, paragraph 5): NRC Inspection Recommended That Procedure (NE-002-002) Be Changed to Include a lower Target Load Rate-of-Change, if Such a Lower Rate Would Not Significantly Impact Test Results - The NRC inspectors determined that LP&L had deleted references to specific values for load rate-of-change from Procedure NE-002-002, Revision 3, dated November 3, 198 (Closed) Open Item (382/8628-04): Concern That Meggering Long Runs of High Voltage Cable Would Result in Erroneous Values for Insulation Breakdown Resistance Tests - The NRC inspectors determined that the licensee's current practice is to megger high voltage cables at the circuit breakers. The licensee stated that the cable runs meggered are of

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lengths that do not cause a voltage drop in the cable that could prevent a

valid test of insulation breakdown resistance. LP&L completed this open

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item through Commitment Closure Verification Forms (CCVF) W3M87-0139, dated May 7, 1987, and W3M87-0104, dated April 7, 198 (Closed) Observation (382/86-06, paragraph 8): The NRC inspector recommended that LP&L consider how far above 4840 KW the Emergency Diesel Generator (EDG) can be safely run and for how long and whether a change of the EDG surveillance procedure to incorporate an upper limit on EDG output is warranted. The NRC inspectors determined that LP&L Procedure OP-903-069, Revision 3, Surveillance Procedure Intergraded EDG/ESFT, Section 4.6 stated that an emergency diesel generator load of 4400 KW was not to be exceeded for more than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> during any 24-hour period. The procedure stated that 4840 KW should not be' exceeded in any case, except as permitted in steps 8.2.6 and 8.3.6 of this procedur The NRC inspectors determined that Section 4.6 of this procedure clearly stated that the upper limit of 4840 KW is only valid for a maximum of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> when performing steps 8.2.6 and 8.3.6 of this procedure. This test at 4840 KW is scheduled to be per formed once each 18 months during the EDG 24-hour load surveillance test. This procedure permitted operation of the EDG at load rating of 4400 KW, except during this specified surveillance tes (Closed) Observation (382/84-07, paragraph IV-3.b.1): The NRC CAT inspectors had found weld defects which did not meet the acceptance criteria specified by the architect enginee The NRC inspectors determined that evaluation for engineering two of the were HVACconsidered seismic supp' worst case" welds from theort welds sele inspected weld sample. The inspected structural welds were required to be welded only with the higher strength low hydrogen E7018 electrode The licensee initiated Discrepancy Notice Site Quality Report (DN-SQ-1993)

to document the discrepant clip to embedded plate and clip to structural steel welds. The use of E6011 electrodes was prohibited only for welds to the embedded plate. QC inspection procedures required that QC verify that only E7018 low hydrogen electrodes be used to weld to an embedded plat An examination of the QC inspection records, by the NRC inspectors, for the supports affected by DN-SQ-1993 and NCR-7294 verified that E7018 electrodes were used for the attachment welds to the embedded plat Engineering analysis was performed, and the analysis determined that the welds would have been acceptable for the design load conditions, even if E6011 electrodes had been used instead of the specified E7018 electrode NCR-7294 was initiated by the licensee to document the apparent undersized weld (vendor shop drawing indicated 1/4-inch fillet weld and Drawing required 3/16-inch fillet weld) between a clip to structural steel connectio The licensee's reanalysis of this weld indicated that the as-built weld was stressed to 17 percent of its design limit. Therefore, the weld was determined to be structurally sound and met design requirements.

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- (Closed) Observation.(382/84-07, paragraph V-5): The NRC CAT inspectors had found 40 structural steel members and 19 bolted connections to be in conformance with design drawings. Four of 339 high-strength bolts ldid not

' meet the minimum torque requirements specified in ASP-IV-129. .The NRC inspectors determined that of the 4 bolts with-low torque-values, none were lower than 600 ft-lbs, whereas the required torque is 79.0 ft-lb . QC-IR-C-0062 recorded the corrective actions take The licensee records  ;

indicated that all of the other bolts in the 19 connections inspected  !

during the . field inspections were torque-tested and were satisfactor (Closed) Open Item (382/8531-01): Document' Controls for Vendor Manuals and Technical Data, Agastat Relays Used in Safety-Related Applications -

The licensee has revised Procedure ME-7-005 to add the.necessary requirements and reviewed their maintenance history of the diesel generators to verify that. qualified relays were installed as . replacement (Closed)'Open Item (382/8531-02): Review of Licensee Potentially Reportable Events (PRES) for Events That May Have Been Caused, or Adversely

. Affected, by Licensee Maintenance Practices - Addition of incorrect hydraulic fluid to safety-related Feedwater Isolation Valve FW-184B. The licensee has~ incorporated proper hydraulic fluid identification uses in the controlled Plant Lubrication ~ Manua (Closed) Open Item (382/8531-02): Failure of B Charging Pump, GE AKR-50 Circuit Breaker to Make Contact On One Phase Resulting in Failure of the Pump Motor to Start .The licensee has revised Procedures ME-3-330 and ME-4-145 to include contact adjusting screw prevailing torque and to add l LOCTITE 220 if necessary. All motor-operated breakers and Train B breakers have been checked. Train A_ breakers will'be checked during the next scheduled outag The NRC inspectors brought to the attention of the' licensee several minor errors'in Procedures ME-3-330 and ME-4-145 that were caused from changes and revisions. It.was suggested that greater attention to details in the preparation and review cycle of changes and revisions to procedures would -

reduce the number of changes required to correct error (0 pen) Open Item (382/8527-07): Design Life of Safety-Related i Capacitors - The NRC inspectors discussed this item with licensee 1 personnel. The licensee had not completed the review and evaluation of this ite (0 pen) Violation (382/8531-03): Failure to Evaluate Adequately and to Report a Defect to NRC - The licensee took exception to, and denied, this violation by letter on March 21, 1986. NRC Region IV forwarded a letter to the Quality Assurance, Vendor, and Technical Center Programs office for l evaluation by that office on October 8, 1986. This violation remains open l '

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pending receipt of information from the Quality Assurance, Vendor, and Technical Center Programs office, NRR, to complete the evaluation of this violatio No violations or deviations were identified during the reviews of the above listed previously identified item (Closed). Unresolved Item (382/8612-01): (NOTE: Unresolved Item 382/8612-01 is a typographical error - should be: 382/8712-01)

System for handling M&TE nonconformance could present a potential for an inordinate time delay in identifying an out-of-tolerance conditio Additional information, provided by the licensee during this inspection, identifies this matter as a procedure violation. This item has been upgraded from an unresolved item to a violation (see paragraph 3).

3. Calibration Program Activities - Followup on Unresolved Item (382/8712-01)

During a previous review of the licensees' calibration and control of M&TE program activities, the NRC inspector made a random selection of five nonconformance documents (CIWA's) relating to M&TE. devices found to be out-of-calibration after being returned to the onsite metrology laboratory for recalibration. The CIWA identified the component, nonconforming condition, and the previous users. The CIWA was then forwarded to the previous user department (s) for record-search (where used) and evaluatio LP&L Procedure UNT-5-009, " Administrative Procedure Disposition of M&TE Nonconformances," dated October 12, 1986,' paragraph 5.5 states, "The M&TE nonconformance evaluation CIWA must be completed by the department within 30 days from the date of identification recorded on the originator section of the M&TE nonconformance evaluatio The 30-day limit may be exceeded in cases where retest or recalibration of equipment is necessary to evaluate the nonconformance."

As noted in NRC Inspection Report 50-382/87-12, dated July 16, 1987,

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paragraph 3.b, three of the five CIWAs randomly selected by the NRC inspectors could not be located by the cognizant licensee representatives by the end of the NRC inspection period. Furthermore, discrepancies were apparent in the computerized tracking system (MTS) identified during the records searc During this inspection, the licensee provided the NRC inspector copies of the three CIWAs previously unobtainable during the June 15-19, 1987, NRC inspection period. In reviewing the documents, the NRC inspector observed that none of the three out-of-calibration devices required a retest or recalibration of plant equipment; however, all three exceeded the 30-day evaluation time limitation as follows:

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Department (s) QA CIWA Orignator Date Evaluation Date Closure 031565- March 9, 1987 September 17, 1987 March 17, 1987 031566 March 9, 1987 May 22, 1987 May 29, 1987 031600 March 30, 1987 May 22, 1987 May 29, 1987 10 CFR 50, Appendix B, Criterion XVI, " Corrective Action," states,

" Measures shall be established to assure that condition: adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformances are promptly identified and corrected. In the case of significant conditions adverse to quality, the measures shall assure that the cause of the condition is determined and corrective action taken to preclude repetition. The identification of the significant condition adverse to quality, the cause of the condition, and the corrective action taken shall be documented and 1 reported to appropriate levels of management." '

The licensee's failure to evaluate promptly the M&TE devices found to be out-of-calibration is contrary to the requirements established in LP&L Procedure UNT-5-009, " Administrative Procedure Disposition of M&TE Nonconformances," dated October 12, 1986, paragraph 5.5 and constitutes an apparent violation (382/8721-01) of the above Appendix B requiremen It should be noted that during the exit meeting held on September 18, 1987, this matter was discussed in sufficient detail to highlight the underlying concerns of the NRC inspector:

Notwithstanding, the expressed concern by the licensee that the document control system is currently undergoing a transition from the MTS computerized system to the SIMS system and may have contributed to the record search discrepancies, the NRC inspector reiterated the concern that with the limited sample of records selected there is an appearance that the present system of handling M&TE nonconformances presented a potential for an inordinate time delay in identifying an out-of-calibration conditior approaching the operability limitations i prescribed by Technical Specification LCO's or other safety related i equipmen Further classification is required of LP&L Procedure UNT-5-009, paragraph 5.5 in that, based on the discussions at the exit meeting held September 18, 1987, the difference between the nonconformance

" evaluation time" period and final " closeout" should be differentiate . Control Rod Worth Measurements for Pressurized Water Reactors - The NRC inspectors reviewed the licensee's procedure for measuring differential and integral control rod worth. The procedure met the licensee's i commitments to verify and measure control rod bank reactivity worths,

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required by the TS requirements.

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I The NRC inspectors reviewed the licensee's records for control rod bank l worth measurements that were performed to verify that core performance was j consistent with engineering, design, and safety analyses. The post-fuel load startup testing commenced on February 2, 1987, with performance of precritical test Low power physics testing began on February 4, 1987, at 10:19 a.m. , when initial criticality was achieved, following the refueling outage (Cycle 2), and was completed on February 6, 1987. Power ascension testing commenced on February 6, 1987. The intermediate testing plateau-(68 percent of full power) was attained on February 10, 198 Power was restricted to less than 70 percent of full power until an emergency TS change was received by the licensee from NRR on February 13, 1987, due to a projected positive moderator temperature coefficient (MTC).

The TS change permitted Special Test Exception 3.10.2 to be invoked while the unit was in Mode 1 (greater than 5 percent of full power). Reactor power was increased on February 13, 1987, to 84 percent of full power and remained at 84 percent until completion of the MTC measurement on February 20, 1987. The power level of the unit was raised to full power (100 percent) on February 21, 1987. Testing at 100 percent of full power was completed on February 27, 1987. All control rod bank worth measurements were conducted in accordance with appropriate licensee procedures, according to records reviewe The NRC inspectors reviewed the control rod bank worth measurement calculations and the reactivity worth values. The calculations and reactivity worth values were within acceptance criteria requirement The NRC inspectors determined the following by record reviews'and discussions with licensee personnel: Plant conditions were maintained as specified in the procedure Changes to plant conditions were approved by appropriate licensee personnel prior to the change being initiate Reactor coolant and pressurizer boron concentration samples were taken as required during testing. Boron samples were taken at 15-minute intervals during specified portions of reactivity worth measurements to assure accurate, current data on boron concentrations. Boron concentrations in the samples were determined by a chemical titration procedure; titrations were completed by qualified laboratory personne Calculations and reactivity graph information of control rod bank reactivity worths were accurat Values reported for control rod bank worths and for boron concentrations were within the acceptance criteri Control element assembly (CEA) (control rod) trip tests were performed to verify that the elapsed time between initiation of the CEA (control rod bank) trip and 90 percent insertion of each CEA was within the TS requirements. The measured CEA drop time for each full

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length CEA from fully withdrawn-to 90 percent insertion was less.than l 1 the-TS' requirement.of.less than 3.0 seconds.

. The Critical' Boron Concentration (CBC) was obtained for all rods out

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and for:a partially rodded configuration (Group B' inserted). The

' actual-CBC measurements were compared to predictions to verify design, fabrication, and proper loading of the cor The measured CBC values.for both measurement conditions were within 30 ppm boron

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when predicted and measured values were compared.~ -A value of less than 100 ppm difference between predicted and measured values was acceptabl 'g . The reactivity worths of various CEA groups (A, B, 6+3,5+4,2+1)

were measured to verify calculations of available shutdown margi The measured values were in good agreement with the predicted value The CEA worths were measured using the CEA exchange technique, consisting of measuring the worth of a " Reference Group." Group B was selected as the reference. group. The measurement was through use of standard boration/ dilution techniques of the reactor coolant system, then exchanging Group B with the other groups to measure the reactivity worths. Each CEA was verified to be coupled to its respective. extension shaf The power ascension was monitored by an off-line NSSS performance and data processing computer algorithm. The computer code was periodically used during power ascension to monitor performance related to the processed plant data. . The measured RCS f. low. rate was used to calculate' flow calibration factor A. fuel symmetry test was performed to verify that no detectable fuel misloadings' existed. Specified power distribution parameters were obtained and compared to predictions to verify the acceptability of the measured power distribution. All acceptance criteria were met by evaluating the following:

(1) 68 percent power plateau radial power distribution (2) 68 percent power plateau axial power distribution (3) 100 percent power plateau radial power distribution (4) 100 percent power plateau axial power distribution (5) All rods out peaking ' factor results The licensee performed radial peaking factor and CEA shadowing factor verification by measurement and' calculation, using fixed incore detector and excore detector data from selected CEA configuration The NRC inspector's review and evaluations determined that the Waterford 3 Cycle 2 core was properly installed and tested. The measured and calculated values were consistent with predicted values and met commitments and TS requirement _ _ _ _ _ _ _ . _ - _ _ _ _

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No violations or deviations were identified in this area of the inspectio . Isothermal and Moderator Temperature Coefficient Determinations The NRC inspectors reviewed the procedure and data for the isothermal temperature coefficient (ITC) measurements at the Essentially All Rods Out (EAR 0) configuration and at a configuration with the Group B control rod bank partially withdrawn to verify core physics parameters. The moderator temperature coefficient (MTC) was calculated by the licensee from the measured ITC at the EAR 0 condition as required by the TS. The ITC measured value agreed within limits with the predicted values for EAR 0 and Group B partially withdrawn. The measured EAR 0 MTC value (0.505 E-4 dRho/degF) was found to be slightly above TS zero power limits. The licensee placed administrative limits on the RCS boron concentration to assure that the MTC was maintained with the LC0 limit The NRC inspectors reviewed the data for the ITC measurements at approximately 82 percent of full power. The predicted fuel temperature coefficient was subtracted from the measured ITC to determine the MT The results of the ITC and MTC measurements at 82 percent of full power agreed closely with the predicted values and met acceptance criteri The projection of MTC at 70 percent of full power was calculated to be-0.04 E-4 dRho/degF, which is in compliance with the T The temperature shadowing factor test was performed using ex-core detectors and RCS cold leg data. The measured values were within the acceptance criteria bound The NRC inspector's review and evaluation of data and calculations determined that the tests were conducted as required by the procedures and that the values were within acceptance criteria stated in the procedures and T No violations or deviations were identified in this area of the inspectio . Procedure Review The NRC inspectors reviewed the following procedures: NE-2-060, Revision 0, "Startup Test Procedure Isothermal Temperature Coefficient," dated Oc.tober 9,1986. This procedure provided

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instructions for determining the isothermal temperature coefficient of the Waterford 3 reactor core during low power physics testin The moderator temperature coefficient derived in this procedure was utilized to satisfy TS Surveillance Requirement 4.1.1.3. NE-2-020, Revision 0, "Startup Test Procedure CEA Insertion Time Measurement," dated January 19, 1987. This procedure was used to verify that the 90 percent insertion time of all full-length (control

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l element assembly) CEAs was equal or less than 3.0 second Records indicated that at the start of Cycle 2 this test utilized CEA timing software that would allow the simultaneous measurement of all CEA drop times. The procedure provided that with all CEAs withdrawn criticality monitoring was required and provided for the following:

(1) Determine that the 90 percent insertion time of each full-length CEA from the time power is removed from the holding coil met TS requirements stated in paragraph 4.1. (2) Perform a functional test of each CEA reed switch position transmitter in accordance with TS 4.1. NE-2-040, Revision 0, "Startup Test Procedure CEA Group Worth and CEA Coupling Check," dated November 12,198b. This procedure is used to measure CEA Group worth using the CEA Exchange Technique. The procedure is used to verify that all CEAs are coupled to the (control element drive mechanism) CEDM NE-2-050, Revision 0, "Startup Test Procedure Critical Boron Concentration Verification," dated October 12, 1986. This procedure is used to measure the critical boron concentration for all control rods in the "out" position and in a rodded configuration. This procedure also is used to calculate the inverse boron wort j No violations or deviations were identified during the review of the above procedure . Exit Meeting The NRC inspectors conducted an exit meeting on September 18,-1987, with the licensee personnel denoted in paragraph 1. The NRC senior resident inspector also attended. At this meeting, the scope and findings of the inspection were summarize ____ _ _ _