IR 05000454/1987015

Insp Repts 50-454/87-15 & 50-455/87-14 on 870310-0619.No Violations Noted.Major Areas Inspected:Previous Insp Findings & Startup Test Results Evaluation.Inadequate Test Results Review & Instrumentation Meets 10CFR2,App G

ML20235H726
Person / Time
Site:	Byron
Issue date:	07/09/1987
From:	Vandenburgh C, Wright G NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20235H705	List: IR 05000454/1987015 ML20235H702
References
50-454-87-15, 50-455-87-14, NUDOCS 8707150241
Download: ML20235H726 (14)

Text

{{#Wiki_filter:.--- __ 's 4 U.S. NUCLEAR REGULATORY COMMISSION

REGION III

Reports No. 50-454/87015; 50-455/87014 Docket Nos. 50-454; 50-455 Licenses No. NPF-37; NPF-60 Licensee: Commonwealth Edison Company Post Office Box 767 Chicago, IL 60690 Facility Name: Byron Station, Units 1 and 2 Inspection at: Byron Station, Byron, Illinois Inspection Conducted: March 10, 1987 through June 19, 198 ! Inspector: Chris A. VanDenburgh

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_O Date 1* - Approved By4 Jeo re C. Wrigty , h_ief

  . 7/9!f7
/ Test Program SectHon   Date /

Inspection Summary: Inspection on March 10, 1987 through June 19, 1987 (Reports No. 50-454/07015(DRSS); No. 50-455/87014(DRS)) Areas Inspected: Routine, announced safety inspection of licensee action on previous inspection findings (92701) and startup test results evaluation (72596, 72600, 72608 and 72616).

Results: Of the two areas inspected, one violation was identified in paragraph 3.f (Inadequate test results review and test instrumentation), however because the violation meets the requirements of 10 CFR 2, Appendix-C, Part V, " Enforcement Actions," no violation will-be issue ..

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8707150241 870709 'i DR ADOCK 0500

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I DETAILS j

       '1 1 Persons Contacted R. Querio, Plant Manager      1
*R. Pleniewicz, Production. Superintendent R. Ward, Services Superintendent I L. Sues, Assistant Superintendent, Operating
*T. Joyce, Assistant Superintendent, Technical Services
'*D. St. Clair, Assistant Superintendent, Planning
*D. Tuetken, Assistant Project Manager     i
*E. Falb, Unit 2 Testing Supervisor      1 D. Elias, Project Engineering Manager     ]

P. Donovan, Project Engineering j

*B. Pirnat, Regulatory Assurance      ]
*E. Zittle, Regulatory Assurance      l A. Chernick, Compliance Supervisor     l
*F. Hornbeak, Technical Staff Supervisor     l
*H. Erickson Jr., Maintenance Staff Assistant
*D. Berg, Nuclear Safety
*J. Snyder, Quality Assurance
*S. Altmayer, Technical Staff      j
*M. Snow, Regulatory Assurance Supervisor
* Denotes those personnel prtis'!nt at the exit interview of June 19, 198 l
       !' Licensee Action on Previous._I,nspection Fipdings (Closed) Violation (4it'85008-01(DRS)): This item concerned the testing of RC 63.32, " Reactor Coolant Systen Flow Coastdown," on Unit 1, in that i.he irispector deterntined that the Acceptance Criteria as specifien in the Final Safety Analysis Report (FSAR)

was modified in the test procedure without prior NRC approval as required by Operating License NPF-37 Condition 2.C.3.b. In a letter from J. G. Keppler dn.ed November 7,1986, Region III forwarded a synopsis of the Office of Investigations Report No. 3-85-007 which detailed the NRC investigation of the circumstances surrounding the Project Engineering Department approval of RC 63.32. The results , of this investigation were discussed in an Enforcement Conference held on December 15, 1986. As a result of these actions, a Notice of Violation and Imposition of Civil Penalty was issued on March 26, 1987. The inspector has reviewed the response to the violation dated April 24, 1987, which details the corrective actions undertaken to prevent reoccurrence and has no further concerns. Based upon the corrective actions taken this violation;is considered closed, (0 pen) Open Item (454/85008-02(DRS)): This item concerned the results of RC 63.32 on Unit 1, in that the results did not document the methodology in which the test data obtained in RC 63.32 were calculated and evaluated by Westinghouse to ensure that the test requirements were satisfied. -This item will remain open pending the development by the licensee of a letter-to-file describing the methodology used to evaluate the test' data obtained in RC 63.3 ,

. . i i (Closed) Violation (454/85033-01(DRS)): This item concerned the ] failure to initiate a Deviation Report following the occurrence of a mechanically stuck control. rod on Unit 1 in Mode 5 on July 17, 1985, as required by the Quality Assurance Manual. The licensee provided responses to the violation in letters to J. G. Keppler i dated August 28 and September 12, 198 The' inspector has reviewed

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I the corrective actions undertaken to prevent reoccurrence.and has J no further concerns in this area. This violation is considered close , (0 pen) Unresolved (454/85035-01(DRS)): This item concerned the results evaluation of NR 52.368, " Axial Flux Calibration," 'on Unit 1, in that the inspector noted the resolution of deficiency 2C1 indicated that the test of'the Flux Deviation Circuitry was not i within the scope of the tes Further investigation determined i that the deficiency was incorrectly closed to Nuclear Work Request (NWR) 821688. A new NWR (B22551) was initiated to resolve the deficiency. Additional investigation by the licensee, provided to the inspector in a letter dated March 3, 1987, indicated that the Axial Flux Deviation Instrument (Serial No. BIP 1NY-NR8050) was incorrectly calibrated since early 198 Due to low setpoint/ reset values utilized, the Axial Flux Difference Instrument (1NY-NR8050)' Annunciator Alarm 10-804, " Power Range Lower Detector Flux _ Deviation High," was actuated almost continually and was therefore not operational because actual core tilt conditions were above the setpoint/ reset values. The instrument was properly calibrated on December 16, 1985, by the performance of Surveillance Procedure BIP 2400-33 using Temporary Change 85-1-95 The inadequate closure of deficiency 2C1 identified in Startup Test NR 52.86 resulted in the failure to implement adequate corrective actions for an identified deficient condition, This will continue to be followed as an unresolved item pending an evaluation by the licensee into the circumstances involve The inspector notes that the performance of BIP 2400-33 has been deleted from the performance of NR 52.85D on Unit 2 and is discussed as a new unresolved item in Paragraph 3.1 of this repor . Startup Test Results Evaluation The inspector reviewed the results of the below listed.startup test procedures to verify that all test changes were identified and approved in accordance with administrative procedures; all test deficiencies were appropriately resolved, reviewed by . management and retested as required; test results were evaluated by appropriate engineering personnel and specifically compared with acceptance criteria; data was properly recorded,' signed, l dated and documented as test deficiencies if. out of tolerance, i and test results were approved by appropriate personnel: l

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     .l Precritical and Low Power Test Sequence FH 32.80, " Initial Core Load Sequence" FH 32.82, " Initial Core load"
*FH 32.83, " Initial Criticality Sequence"
*IC 45.81, "Incore Flux Mapping at Low Power" IT 47.80, " Isothermal Temperature Coefficient Determination 3 (ARO)    .].

IT 47.80, " Isothermal' Temperature Coefficient' Determination' .

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IT 47.80, " Isothermal Temperature Coefficient Determination (CBC/CBD In)

*IT 47.82A, " Thermal Power Measurement'and Statepoint Data Collection (Precritical)"
*IT 47.82C, " Thermal Power Measurement and Statepoint Data Collection (0% Power)"   i NR 52.82, " Initial Criticality"   ;

NR 52.85A, " Nuclear Instrumentation System Calibration (Prior to Core Load)" NR 52.850, " Nuclear Instrumentation System Calibration (Prior to Initial Criticality)"

*NR 52.850, " Nuclear Instrumentation System Calibration (0% Power)"

RC 63.82, " Reactor Coolant Flow Coastdown" RC 63.83, " Reactor Coolant System Leak Test" RD 64.83, " Boron Endpoint Determination (SBA)" RD 64.83, " Boron Endpoint Determination (SBE)" RD 64.83, " Boron Endpoint Determination (CBA)" RD 64.83, " Boron Endpoint Determination'(CBB)"  ; i RD 64.83, " Boron Endpoint Determination (CBC)" RD 64.83, " Boron Endpoint Determination (CBD)" i RD 64.85, " Rod Droc Titre Measurement. Test" I

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l RP 68.80, " Reactor Protection System Trip Testing" ' RY 69.80, " Pressurizer Testing"

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30% Power Test Sequence d I AP 05.80, " Loss of Offsite Power" l l EM 28.80A, " Pipe Vibration (5% Power)"  !

*EM 28.808, " Pipe Vibration'(10% Power)"   y
*EM 28.808, " Pipe Vibration (30% Power)"
*EM 28.80C, " Pipe Vibration (30% Power)"

IT 47.81, " Power Coefficient Determination--(30% Power)"

*IT 47.828, " Thermal Power Measurement and Statepoint Data Collection (40% Power, cc #1)'_'
*IT~47.828, " Thermal Power Measurement and Statepoint Data Collection (40% Power, cc #2)"
*IT 47.82C, " Thermal Power Measurement and Statepoint Data Collection (30% Power)"

NR 52.850, " Nuclear Instrumentation System Calibration (30% Power)"

*RC 63.85, " Shutdown from Outside the Control Room" RD 64.84, " Automatic Reactor Control" I

50% Power Test Sequence  ! i

*EM 28.80C, " Pipe Vibration (50% Power)"   l

IT 47.81, " Power Coefficient Determination (50% Power)"

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*IT 47.82C, " Thermal Power Measurement and Statepcint Data -{

Collection (50% Power)" '

*NR 52.850, " Nuclear Instrumentation System Calibration (50% Power)"
*RD 52 86A, " Axial Flux Difference Instrumentation Calibration (50% Power)
*NR 52.87, "10% Load Swing Test-(50% Power)"

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75% Power Test Sequence

*EM 28.80C, " Pipe Vibration (75*4 Power)"
*IC 45.82C, "Incore Movable Detector and Thermocouple Mapping at Power (75% Power)"
*IT 47.81, " Power Coefficient Determination (75% Power)"
*IT 47.82D, " Thermal Power Measurement and Statepoint Data  !

Collection (75?? Power)"

*NR 52.850, " Nuclear Instrumentation System Calibration (75% Power)"
*RD 52.86A, " Axial Flux Difference Instrumentation Calibration (75% Power)
*NR 52.87, "10?4 Load Swing Test (75% Power)"  ,

l 90% Power Test Sequence

*IT 47.81, " Power Coefficient Determination (90% Power)"
*IT 47.82C, " Thermal Power Measurement and Statepoint Data Collection (90?; Power)"
*NR 52.85D, " Nuclear Instrumentation System Calibration (90% Power)" t
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Denotes those procedures for which final Project Engineering Department (PED) approval has not been received. Thi.. item i will be followed as an open item pending inspector receipt j and evaluation (455/87014-01). j

     ! During the results review of FH 32.80 performed in the Precritical Test Sequence, the inspector noted that the containment evacuation alarm was not verified operable and logged at eight hour intervals as required by procedure step 7.3.2 and Appendix C, Table C1. FSAR Table 14.2-62 requires that throughout core loading containment evacuation alarms be verified operable at least once per eight hour Deficiency 1-F-1 documents that on November 7, 1986, at 15:42 hours and on November 8, 1986, at 07:33 hours greater than eight hours had elapsed between verifications. The eight hour interval was missed by ten minutes and three minutes respectivel Initially, the deficiency was resolved by indicating that the verification had taken place but that the logging of the verification was delaye Further review by the' Project Engineering Department, however, concluded that the actual verification was delaye In a letter to the Project Engineering Department dated January 26, 1987, l the Byron Station concluded that the requirements of the FSAR i Table 14.2-62 were met because a review of the logs indicated that j an actual fuel movement was not in progress at the times when the

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containment evacuation verification was delayed. 'The inspector-believes that this conclusion is' incorrect. The'FSAR requirement' refers to the core loading procedure and'not to a. specific fuel l movemen Nevertheless, a-review of the FH 32.82, Appendix C,

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Table'C1 data indicates that although there were two instances -) in which the time between verifications minimally exceeded the eight j hour requirement, the containment-evacuation verification was-  ! performed at least once in every eight-hour period from when the  ! core loading procedure was commenced. The inspector concludes that ) this satisfies the requirement of the FSAR and-there are no further 1 concerns with this. test. .The licensee should evaluate _the testing , requirements for Braidwood Generating Station, Unit 2 to preclude similar discrepanc I b. During the results review' of FH 32.83 performed in the _ Precriti_ cal Test Sequence, the inspector noted that Test Review Board note'#9: indicates that Safety Evaluation Form-(10CFR50.59 - BAP 1310-T19) requires a clarification because the scope of the changes in the Test Change Request (TCR) do not effect requirements as defined. b Technical Specifications. There is no indication as to which TCR!is , referenced.' There are nine TCR's and one Temporary Change'in FH 32.83: .

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which have Safety Evaluation Forms. This' item will be followed as a open item pending the receipt and review of the Project Engineering Department test approval (455/87014-02).

c. During the results review of IT 47.81 performed in_the Low Power Test Sequence, the inspector noted the following concern ) Deficiency 1-F-1 identifies that greater _than 25E-6 amperes u s not rea,hed on all incore detectors as i required by the note preceding step 9.1.44. The-note i requires that reactor power be increased if levels below 25E-6 amperes are achieved. . The deficiency-indicates that the Onsite Project Engineering DepartmentbelievestherecordeddataJisacceptabTeand I that Westinghouse has provided a " preliminary"~ opinion , that the data is acceptabl ) Deficiency 1-S-1 identifies that repetitive step 9.2.25 on pages 48 and 49 was not signed off as required.- This step verifies that the computer constants after each pass of the incore detectors (12 passes) have not - changed. The computer constants are required to be verified to ensure that all strip chart recordings are normalized between 0 and'1.0 which is required by Acceptance Criteria ) IntheSTETestEvaluationSection12.4,["Overall Recommendation," the STE conc 1'udes that it is obvious from the number of difficulties and mechanical failures observed that the operability of the Incore System was not demonstrate > . . These items will be followed as an open item pending final approval. of the test results by the Project Engineering Department (455/87014-03).

d. During the results review of IT 47.80 (CBD In) performed in the Low Power Test Sequence, the inspector noted the following concern ) IT 47.80 (CBD In) measured the Isothermal Temperature Coefficient of Reactivity by implementing a three degree change in temperature of the reactor. coolan j The licensee's review indicates that this value is not :1 in accordance with the Final Safety Analysis Report

'(FSAR) Table.14.2-77 (Amendment 40) requirement to
. initiate an approximate.five degree change'in temperatur Westinghouse has reviewed the data obtained and' test-methodology of'IT 47.80 and concluded in letter CAW-10317 dated January. 21, 1987, that the test methodology.is in accordance with Westinghouse suggested:Startup Test Procedure CVA-SU-3.3.6, (Revision 1) Table 1, and.that the data obtained in IT 47.80 is valid. Furthermore, a recommendation was made to revise FSAR Table 14.2-77.-to require a three degree change in temperature. Action Item Request 6-87-2007 dated January 21, 1987, has.been initiated to request this chang The licensee'has discussed this change with the Nuclear Reactor Regulation (NRR) Project Manager who agreed that the clarification to the FSAR'does not represent a change to the Startup Test Program and does not need to be reported to the NRC within thirty days. The' change will be made in Amendment 48 to the FSAR following the issuance of an Operating License for Braidwood Unit This revision will be followed as an open item (455/87014-04).

2) An-additional concern identified during the result review of IT 47.80 concerned the position of Control Bank C (CBC) during the test. The actual position-of CBC was not within the range of positions recommended by Westinghouse in'their suggested Startup Test Procedure CVA-SU-3.1.4. IT 47.80 was performed with CBC positioned at 192 and 196 steps withdrawn, vice 200 steps withdrawn, as recommended. Westinghouse subsequently reviewed and accepted the test results in letter CAW-10308, Revision 1, dated January 21, 1987, and indicated that if the control rods are inserted less than 50 pcm reactivity-worth, the intent of CVA-SU-3.1.4 is met. The inspector has no further concerns since the reactivity worth of the control bank insertion was estimated to be no more than-30 pc _

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, e. During the results verification of RC 63.82 performed in the Precritical Test Sequence, the inspector noted that the test results were evaluated using the revised test methodology of Westinghouse suggested Startup' Test Procedure CVA-SU-2.1.8, Revision 4, dated October 24, 1985. The. inspector additionally , notes that the Project Engineering Department performed a second I calculation to verify the flow coastdown calculations performed l in the tes Both calculations confirm that the flow coastdown is acceptable. The inspector has no further concern f. During the results review of RC 63.83 performed in the Precritical 1 Test Sequence, the inspector noted the following concern ] 1) The System Test Engineer (STE) Test Evaluation documented ) that the performance of Surveillance 2BOS 4.6.2.1,C-1, j Revision 51, used to calculate the identified and unidentified ] reactor coolant system leakage contained several mathematical ) errors. The-STE recalculated the leakages'and demonstrated " that the acceptance criteria was satisfie Further review by the Project Engineering Department determined that the values are also in error due to the incorrect interpolation of the Pressurizer Relief Tank volume. The inspector recalculated the identified and unidentified leakages and verified that the acceptance criteria was satisfie ) The Project Engineering Department (PED) review also identified that four instruments used to collect acceptance criteria ' 1 data were outside of the eighteen month calibration interva l Procedure step 7.2.1 requires a verification that the installed instruments used to collect ecceptance criteria 1 data are within an 18 month calibration interval. The I following instruments were not within this required interval and a deficiency was not initiated to evaluate the impact upon the test result RY-LI 470 Pressurizer Relief Tank Level 2RY-TI 453 Pressurizer Liquid Temperature i 2CV-LI 112 Volume Control Tank Level 1 2RE-TI 1058 Reactor Coolant Drain Tank Temperature An independent Quality Assurance Audit of RC 63.82 l (#06-86-2.63.83) identified the discrepancy and requested post-test calibration checks of the affected instrument The post-test calibrations revealed that 2 of the 4 instruments had some portion of their loop out-of-toleranc PED evaluated the results based upon these errors and determined that the data was acceptabl The two problems identified in this test (inadequate test results review and test instrumentation) are considered violations of NRC requirements, however, because the examples meet the requirements of 10 CFR 2,' Appendix C, Part V, in that

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  -they were identified by the licensee; have minimal safety l significance; were documented in the licensee's results <

review;'were not required to be reported; and corrective j actions have been taken in the form of recalculations'and ' recalibratioas; a' violation of the NRC requirements will not be issued. The inspector has no further concerns with this tes During the-results review of RD 64.'85 performed in the Precritical Test Sequence, the inspector noted the following concern ) The-Project Engineering Department (PED).' review.of the test results dete'rmined that the individual rod drop time for rod P-12 was recalculated because of an interpolation error as noted on' the strip chart' recording but that the new-value was not utilized'in the averaged rod drop time calculations because a deficiency was not initiated to identify the discrepanc Subsequent calculations by PED have verified that there-was no effect as a result of this erro ) The Westinghouse review of the test data provided'in letter MED-PCE-4510 dated February 16, 1987, indicated in comment A, that a statement used to justify TCR #1 is incorrec TCR #1 deleted the requirement to test the rod drop times at cold no-flow, cold full-flow', 'and hot no-flow conditions because ". . . Westinghouse Acceptance Criteria apply to the hot full-flow condition only, and that from previous experience they

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have found cod drop times at other test conditions fall under the hot full-flow values." Westinghouse indicated that this statement is accurate in that the Acceptance Criteria given in Technical Specification 3.1.3.4 does apply at hot full-ficw conditions'onl However, the statement that rod drop times at other test conditions fall under hot-full-flow values is not accurate since rod drop times under cold full-flow conditions are typically longer than those'under hot full-flow condition TCR #1 was implemented to delete the rod drop testing at conditions other than hot full-flow based on NRC approval to modify the Initial Test Program as documented in a letter from'V. S. Noonan to D. L. Farrar dated October.1', 198 The basis for the approval was stated in.the attached Safety-Evaluation which indicated that the licensee's justification a for the change is that the Westinghouse Acceptance Criteria l apply to the . hot full-flow condition' only, and that from l previous experience they have found rod drop times at other ; test conditions' fall under the hot full-flow values. The j NRC'found that since the justification bounds the other test i conditions, the additional testing'at conditions other than hot full-flow was not necessar , 10 i a

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This item was originally discussed in unresolved Item No. 454/84083-02 and closed in Inspection Reports No. 50-454/84087; No. 50-455/84056. As a result of the new information, further NRC evaluation of the basis I for the modification to the Initial Test Program is required. This item will be followed as an unresolved item pending this evaluation (455/87014-05).

3) The Westinghouse review of the test data provided in , letter MED-PCE-4510 dated February 16, 1987, indicated j in comment C, that the acceptance criteria for rod ' drops initially found to be outside the upper and lower , two sigma limit of the initial drop times for all rods l is typically a band of 0.020 seconds for additional ! drops. This acceptance criteria was not provided in RD 64.85; however, the rod drop times were obtained and provided to Westinghouse for evaluation and subsequently accepted. The Project Engineering Department has indicated . in a letter to both the Byron and Braidwood Generating i Stations dated April 1,1987, that the addition of acceptance !

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criteria will be considered for future testin The inclusion of the new acceptance criteria will be followed as an open item (455/87014-06).

4) The Westinghouse review of the test data provided in letter MED-PCE-4510 dated February 16, 1987, indicated in comment D, that FSAR Table 14.2-66 required that the i drop traces confirm proper operation of the decelerating devices and that RD 64.85 did not address this in the i acceptance criteria of Section 4.0. In Attachment F to J Westinghouse letter CAW-10392 dated February 20, 1987, the Westinghouse Oversight Committee recommended that the rod drop traces be reviewed to verify the consistency of the dashpot region (i.e., the decelerating devices) and that this review be documented. The PED approval letter and associated attachments provided to the Byron Station on April 1, 1987, do not provide documentation of the review nor resolution of the concern. Because this item was identified late in the inspection period, this item will be followed as a unresolved item pending review and evaluation by the licensee (455/87014-07).

h. During the results review of RY 69.80 performed in the Low Power Test Sequence, the inspector noted that the total pressurizer heater output was less than expected and that the rate of pressurizer pressure increase was not within the test acceptance criteri j The System Test Engineer (STE) concluded in the Test Evaluation that the pressurizer heaters did not meet their acceptance criteri Review by the Project Engineering Department and Westinghouse concluded that deviations in the anticipated pressure response are acceptable and do not present a safety concern because credit is not

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taken for pressurizer heater performance to mitigate the response from any plant transient by the safety analysis. Westinghouse confirmed in letter CAW-10295, Revision 1,-dated January 7,=1987 that the FSAR Chapter 14 Test Requirements (Table 14.2-71) have been satisfied. The licensee's Project Engineering. Department concluded in a Safety Evaluation performed January 8, 1987, and documented in a letter to the Byron Station dated January 23, 1987, that NRC notification of the deviation in pressurizer heater performance ~was not required because the variations were in the the test acceptance criteria and not in the FSAR Acceptance Criteria. The inspector reviewed the referenced documentation and has no further concern During the results review of AP 05.80 performed in the 30% Power Test Sequence, the inspector noted that the strip chart data to , substantiate Acceptance Criteria 4;5 was twenty seconds short of i the required thirty minute demonstration of the operability of at

'least one safe shutdown train. -The missing twenty seconds was the   i result of personnel error. A subsequent review of the operating logs indicated that the equipment remained operable for the   ;

required period. Additionally, the inspector noted that the Reactor Coolant System (RCS) T-ave was not maintained above the minimum value allowed by Acceptance Criteria 4.6.1. Evaluations by Westinghouse and the licensee indicated that the excessive cooldown of the primary was a direct result of low decay heat, the absence ' of pump heat addition and -the auxiliary feedwater flow. .. Revisions to prevent the recurrence of these deficiencies were forwarded to the Braidwood Generating Station by a letter dated April 8, 198 The inspector verified that these deficiencies are acceptable, that , the FSAR Table 14 requirements have been satisfied and that NRC 1 notification of a deviatior, from the Startup Test Program is not - require j

       ! During the results verification'of EM 28.80A performed in the 30% Power Level Sequence, the inspector noted that a conflict between FSAR Section 3.9.2.1 and Safety Evaluation Report (SER)

Section 3.9.2.1 requirements for allowable piping stress values t was identified. The discrepancy was identified in a letter from P. Donavan to D. Elias dated March 9, 1987. The letter additionally indicated that NRC Inspection Report 50-454/86036 had accepted the 1 FSAR values as correct. The inspector was not'able to verify this  ! statement and will follow this item as an open item pending further l review and evaluation (455/87014-08). j 1 During the results review of'IT 47.82B performed.in the 30% Power Test Sequence, the inspector noted that controlled copy (cc) #1 of the test procedure was rejected by-the Test Review Board and reperformed as cc #2 on March 4, 1987, due to the 2A.and 2C Steam Generator feedwater flow differential pressure data reading abnormally l high as documented in deficiency 1-C. The test was reperformed using 1 Barton differential pressure transmitters vice Valadyne transmitters I with satisfactory results. The inspector has no further concern I

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1. During the results review of IT 52.85D-performed in the 30% Power Test Sequence, the inspector noted that Major Test Change Request . .I (TCR) #2 dated January 22, 1987, deleted Test Section 9.10, " Deviation Alarm Circuitry Check," and Test Objective 3.4, which required verification of the proper settings of the flux deviation alarm FSAR Table 14.2-68 (Amendment 40) required that the alarm and protective functions of the Nuclear-Instrumentation System be verifie Section 9.10 required that Surveillance Test Procedure BIP 2400-33, " Calibration of Nuclear Instrumentation System Flux Deviation," be performed and a copy attached in Section 14,

   " Attachments". The justification for TCR #2 stated that the deviation circuitry was acceptably tested during Preoperational Test 2.52.60 and that the circuit was calibrated in May and November 1986 by the Instrument Mechanic Department. The justification infers that.the deletion of Section 9.10 is acceptable because IT 52.850 required the Instrument Mechanic Department to calibrate the drawer which had been accomplished-prior to the performance of IT 52.85 IT 52.85D was performed on December'27,'1986. Section 7.4,
   " Calibration," of IT 52.85D indicated that the deviation-instrument drawer (2NY-8050) was calibrated one year prior-to the performance of IT 52.85D on December 16, 1985. The-licensee's Project Engineering Department _ approved IT 52.85D      !

in a letter dated June 10, 1987. Because this item was identified late in the inspection period, this item will be followed as an unresolved item pending the licensee's review and evaluation (455/87014-09), m. During the results review of RC 63.85 performed in the 30% Power Test Sequence, the inspector noted that Test Change Request (TCR) #3 changed the minimum shift crew composition to increase the number o Auxiliary Operators from two to four. The justification for the TCR indicated that most recent Auxiliary Operator requirement per Byron Station Appendix R audit of 1986 is four Auxiliary Operators. This change conflicts with the Technical Specification-Requirements of Table 6.2-1, " Minimum Shift Crew Composition," which only requires three Auxiliary Operators. Further discussion between the licensee and the NRC staff indicated that there may be some discrepancy between the minimum shift crew composition required to be demonstrated in RC 63.85, the number of people required for a fire brigade, the minimum shift crew specified in Technical Specification Table 6.2-1,. and the minimum shif t crew requirements of Byron Administrativ Procedure BAP 320-1, " Minimum Shift Manning." In order to clarify the requirements, the licensee is evaluating the minimum manning required to operate both units in the event that the control room must be evacuated. This item will be followed as an unresolved item = pending the receipt and review of this evaluation (455/87014-10).

n. During the results review of IT 47.81 performed in the 75% Power Test Sequence, the inspector noted that deficiency 4-A documented 1 that the load swings performed were less than 47 MW as required by the procedure. The Test Review Board concluded that the data

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obtained was adequate to properly calculate the power coefficient l of reactivity. This item will be followed as an open item pending l Project Engineering approval of the the test results (455/87014-11). j One violation for which no enforcement action will be taken was identified in paragraph . Open Items  ! Open items are matters which have been discussed with the licensee, which will be reviewed further by the inspector, and which involve some action on the part of the NRC or licensee or both. Open items disclosed during l the inspection are discussed in Paragraphs 3, 3.b, 3.c.3, 3.d.1, 3.g.3, j 3.j, ! 4 Unresolved Items I Unresolved items are matters about which more information is required in order to ascertain whether they are acceptable items, violations, q or deviations. Unresolved items disclosed during the inspection are  ; discussed in Paragraph 3.g.2, 3.g.4, 3.1, . Exit Interview (30703) l The inspector met with licensee representatives denoted in Paragraph 1 at the conclusion of the inspection on June 19, 1987. The inspector summarized the purpose and scope of the inspection and the finding j The inspectors also discussed the likely informational content of the 1 inspection report with regard to documents or processes reviewed by the inspectors during the inspectio The licensee did not identify any such documents / processes as proprietar i

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