IR 05000423/1986007

Insp Rept 50-423/86-07 on 860119-0214.No Violation Noted. Major Areas Inspected:Startup Program review,post-core Hot Functional Test Witnessing & Test Result Reviews & Initial Criticality

ML20140A669
Person / Time
Site:	Millstone
Issue date:	03/12/1986
From:	Eselgroth P, Prell J, Wen P NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20140A663	List: IR 05000423/1986007 ML20140A655
References
50-423-86-07, 50-423-86-7, NUDOCS 8603210090
Download: ML20140A669 (24)

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U.S. NUCLEAR REGULATORY COMMISSION

REGION I

Report No. 50-423/86-07 l Docket N License No. NPF-49 Priority --

Category --

Licensee: Northeast Nuclear Energy C P. O. Box 270 Hartford, Connecticut 06141-0270 Facility Name: Millstone Nuclear Power Station - Unit 3 Inspection At: Waterford, Connecticut Inspection Conducted: January 19 - February 14, 1986 Inspectors: om ,

// I ales A. Prell, Reactor Engineer / dage

!M C. Ma~

Peter Wen, Reactor Engineer 3 /to /sb

' date Approved ~wW [ M Pster f4elgroth,4Wief, Test Program Section, M

date OB, DRS Inspection Summary: Routine unannounced inspection conducted on January 19 -

February 14, 1986 (Report No. 50-423/86-07)

Areas Inspected: Startup program review, post core hot functional test witnessing and test result review, initial criticality and low power physics testing, power ascension program review and test witnessing, and review of licensee actions on previous inspection findings. The inspection involved 225 hours0.0026 days <br />0.0625 hours <br />3.720238e-4 weeks <br />8.56125e-5 months <br /> on site by two region based inspector Results: No violations were identifie PDR ADOCK O y3

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DETAILS 1.0 Persons Contacted

R. Bradley, Startup Engineer, NNECO

• K. Burton, Operations Supervisor, NNECO

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• J. Crockett, MP-3 Unit Superintendent, NNECO E. Fries, Startup Engineer, NNEC0 N. Hulme, Startup Engineer, NNECO

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• J. Jensen, QA Specialist, NNECO T. Kulterman, Senior Engineer, NUSCO J. Langon, Plant Engineer, NNEC0 T. Lyon, Startup Test Engineer, NNECO D. Miller, Startup Manager, NNECO D. Moure, Assistant Operations Supervisor, NNEC0 M. Pearson, Assistant Operations Supervisor, NNECO

• W. Richter, Assistant Startup Supervisor, NNECO A. Stengal, Startup Test Engineer, NNECO C. Wooten, Startup Test Engineer, NNEC0

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• D. McDaniel, Reactor Engineer, NNECO

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U.S. Nuclear Regulatory Commission

• F. Casella, Resident Inspector 1 *J. T. Shediosky, Senior Resident Inspector

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• Denotes those present at the exit meeting on February 14, 1986.

2.0 Licensee Action on Previous Inspection Findings (Closed) IFI (50-423/85-71-16) The technical manual and precautions in j SP 3622.5, relating to the Turbine Driven Auxiliary Feedwater Pumps, (TDAFWP) caution the operator to exercise the manual speed adjusting knob following turbine operation to prevent system failure. A review of the procedure related to TDAFWP operation found that they did not adequately delineate this precaution nor did they require independent verification j that the governor was reset to its standby positio The inspector verified that Revision 7 to OP 3322 requires the operator to exercise the manual speed adjusting knob and provide independent verifica-tion of valve lineu This item is close (CLOSED) UNR (50-423/85-71-17) The following items were identified related to SP 3622 and Op 3322-

! 1) OP 3322, step 4.2 refers to -TS Figure Figure 3-7 is not included in the TS.

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2) SP 3622.6 requires operation of the turbine driven AFW Pump at the minimum governor speed for 2 minutes. OP 3322 state that the AFW pump should not be operated below 1500 rpm to " ensure proper lubrication" 3) OP 3322 requires manual lubrication of the AFW pumps bearings if the pumps have not been operated within the previous 30 days. The pro-cedures did not however provide lubrication instruction The inspector verified that the licensee has deleted the reference in OP 3322 to TS Figure 3.7 and replaced it with reference to TS 3.7. The minimum governor speed setting for the turbine driven AFW pump is 1700 rpm, which is well above the 1500 rpm minimum speed stated in OP 3322. Also a standby D.C. Oil Lubrication pump would start on a low pressure signal to " ensure proper lubrication" if the rpms went below 1500 rp A letter from the manufacturer now indicates that manual lubrication is required for AFW pumps bearings if they have not been operated within the previous 40 days versus 30 days. Technical Specification surveillance requirement 4.7.1.2.1 requires each AFW pump to be demonstrated OPERABLE at least once per 31 days by developing a differential pressure of 1460 psid across each motor-driven pump and across the steam-turbine driven pump. Thus all pumps would have to be operated at least once every 39 days (31 days + 25%), which is one day short of 40 days maximum. Thus making the manual lubrication problem moo Based on the above this item is close (CLOSED) UNR (50-423/85-71-21) The precautions and restoration sections of SP 3646A.3, " Diesel Generator Interdependent Test", both listed "None",

which appeared inappropriate. Also the procedure did not include provi-sions for independent verification of equipment restoration following testin The reason there was no precautions or restoration section for SP 3646 is that the SP references OP 3346A for implementation. The inspector verified that OP 3346A has precautionary statements and restores the system. The SP data sheet now incorporates provisions for independent verification of system restoratio Based on the above this item is close (OPEN) UNR (50-423/85-71-22) Various problems were identified with the draft procedure SP 3646A.6, "Offsite Power Transfer Operability Test."

However, the problems identified by the inspector refer to testing of relays associated with the ESF system. Thus SP 3646A.4 " Engineered Safety Systems Integrated Test" appears to be the procedure of concer ___ _ _ ._ - _ _ _ _ _ ___ -______ _. ._

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SP 3646A.4 is still in draft for This surveillance however is only to be done during refueling operations. The initial surveillance was satis-fied by preoperational test 3-INT-2004, "ESF Without loss of Power".

Based on the above this item remains unresolved pending review of the final approved surveillance procedure SP 3646 (CLOSED) CAT Finding - Violation (50-423/85-04-03): This item pertains to the welds identified as nonconforming to the procurement specifications related to some of the installed tanks, pressure vessels, and heat ex-changers. The licensee's field quality control representatives reinspect-ed all fillet welds identified as " Undersized" for equipment nozzles and supports for the subject QA Category I tanks, pressure vessels, and heat exchanger Subsequently, the nonconformance was evaluated for actual stress loading and reworked, as required, under SWEC's QA program, to meet the ASME Code for minimum weld siz To preclude similar incidents from happening in the future, the SWEC QA department instituted a compre-hensive weld inspection training program for QA/QC personnel, including project QA inspectors. In addition, the SWEC QA division, through a memo-randum to all district managers, stressed the company's commitment to pro-per fillet weld inspection. A review of the documentation and field veri-fication supported the adequacy of the licensee's corrective action against this deficienc Based on the above, the inspector determined that the licensee's action is complete. This item is close (CLOSED) UNR (50-423/85-34-04) This item pertains to effluent monitoring sensitivity dat The inspector reviewed the K4 man summary report and de-tector calibration data for the particulate and gas radiation monitoring systems. The detector is certified to have the characteristics noted in the referenced document Calibration is based on deposited activities on a copper disc using beta efficiencies of a proportional counting syste The beta and gamma scintillation detectors used were first calibrated using the standard factory procedure KNP 18-60 revision No violations were identifie ;

This item is close .- Post Core Hot Functional Test

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3.1 Test Witnessing A various times during the inspection period, the inspectors witnessed testing in progress on a sampling basis and evaluated most portions of the Post Core Hot Functional Test. The tests witnessed included:

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3-INT-5000, Appendix 5001, Shutdown Margin;

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3-INT-5000, Appendix 5004, Rod Control Slaver Cycle and CRDM Timing;

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3-INT-5000, Appendix 5006, RCS Leak Detection;

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3-INT-5000, Appendix 5007, Pressurizer Heaters and Spray;

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3-INT-5000, Appendix 5008, Rod Drop Testing;

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3-INT-5000, Appendix 5011, Movable In-Core Detector Operation;

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3-INT-5000, Appendix 5015, Digital Rod Position Indication Operational Test;

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3-INT-5000, Appendix 5018, Rod Control Operational Test; Tests were observed for the following areas:

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Tests were conducted in accordance with the approved test procedures; change to the procedures were made in accordance with the administrative procedur Prior to performing each test, briefing with the test crew and operation personnel were conducted and the briefing was adequat Test prerequisites and initial conditions were me Operator actions were correc Summary analysis was made upon completion of each tes All test results as verified by inspector direct observations indicated that overall test acceptance criteria have been met or proper test deficiencies were documented and followed up by the license The inspector also attended some of PORC/JTG (Plant Operation Review ,

Committee / Joint Test Group) meetings which approved test procedure change '

and final test result acceptance. The inspector noted that-in all instances the PORC/JTG approval process was formal, thorough, and deliberat .2 Test Results Review Those Post Core Hot Functional Test results identified in Appendix A were reviewed by the inspector to verify that:

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test changes were approved and implemented in accordance with administrative procedures;

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changes did not impact the basic objectives of the test;

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test deficiencies and exceptions were properly identified, resolved, and resolution accepted;

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the cognizant engineering group had evaluated the test results and signified that testing demonstrated design conditions were met; and,

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test results compared with established acceptance criteria or were properly resolve Details relating to some of those test results reviewed are described belo . RCS Leak Detection (Appendix 5006)

The Millstone Unit 3 Technical Specifications (TS) require that surveillance be performed at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> using water inventory balance method to determine leakage from the reactor coolant system (RCS). The plant computer RCS leakage calculation program, SP 3J3 is designed to fulfill this requirement. Surveillance procedure SP 3601F.6, RCS Leak Test, delineates calculational steps and is used as a backup when the plant computer is not available. The purpose of this RCS Leak Detection test (Appendix 5006) was to verify that a known 1 gpm leak rate could be detected by the RCS Leakage Calculation Program SP 3J3 and surveillance procedure SP 3601 Baseline Leak Rate Test Prior to performing superimposed leak rate test, a baseline RCS leak rate was measured per program SP 3J3 and surveillance procedure SP 3601F.6 on January 22, 1986. The inspector performed independent calculations using an NRC-developed leak rate computer program NUREG-1107, "RCSLK9: Reactor Coolant System Leak Rate Determination for PWRs", to verify the licensee's calculation. These comparisons are:

Leak Rate (LR) Licensee Calculation Inspector Test Date (gpm) Pgrm 3J3 Hand Ca Ca Identified LR 0.723 0.74 0.73 (0630-1030) Unidentified LR 0.362 0.73 0.36 Based on the same set of initial and final data input, the inspector's calculation agreed closely with the licensee's program SP 3J3 calculation. The licensee's RCS leakage water inventory calculation methodology appeared to be adequat ._ _ _ _ _ .

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I However, from test results review, the inspector identified the following items were either not included or not clearly spelled out in the licensee's procedur The large variation of unidentified leak rate observed betwen the computer program SP 3J3 and surveillance procedure hand calculation (0.362 gpm vs. 0.73 gpm) was primarily due to different values of initial pressurizer level used in the calculation. Instantaneous values at beginning and end of test were used in the licensee's calculation. Since the computer point data and the hand calculation input data were taken at slightly different times, system variation and/or instrument uncertainty might have contributed to this deviated inputs. The large uncertainty of RCS leakage calculation caused by this system variation could have an adverse impact on calculation results especially when the plant is experiencing some known leakag Level changes in Containment Drains Transfer Tank (CDTT),

Primary Drains Transfer Tank (PDTT), Pressurizer Relief  !

Tank (PRT) and Accumulator Tank are counted as an identi-fied leakage. The unidentified leak rate is calculated by subtracting the identified leak rate from the total RCS leak rate. The TS limit (TS 3.4.6.2) for the unidentified leakrate is 1 gp Thus,. incorrect determination of iden-tified leak rate would affect the adequacy of the unidenti-fied leak rate calculation. The inspector noticed that the level change in these tanks, especially the PDTT, could come from various sources other than RCS. However, the surveillance procedure does not clearly indicate that leak-age from RCS can only be considered for identified leakag In addition, during the test, computer point GWS-F84 (VCT Divert flow to BRS) was observed having noise spikes which caused an erroneous computer entry of the VCT dump flo The computer program SP 3J3 requires manual adjustment of the calculated leak rate under these circumstance l These items are collectively an unresolved item pending  ;

i further review by the inspector (50-423/86-07-01). I Superimposed Leak Rate Test

At the end of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> baseline leak rate test, a known leak  !

rate of approximately 1 gpm was induced by opening valve l 3CHS-V800 (drain to auxiliary building sump). The Superim-posed Leak Rate Test was subsequently performed for 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> The licensee calculated. leak-rate results along with the inspector independently calculated results are compared below:

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I Leak Rate (LR) Licensee Calculation Inspector Test Date (gpm) Pgrm 3J3 Hand Ca Ca Identified LR 0.654 0.61 0.73 (1156-1401) Unidentified LR 1.625 1.76 1.54 ,

The change in the RCS leak test unidentified leakage is:

Licensee Calculation  :

Program 3J3 Hand Calculation Inspector Calculation 1.263 1.03 1.18 This comparison indicated that the licensee's computer program performed adequately well. The superimposed leak rate as deter-mined by the licensee's computer program was 1.263 gpm (1.625 -

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O.362) which did not meet test acceptance criteria of I gpm 9%. This discrepancy was discussed during JTG test result re-view. The relatively large uncertainty associated with this

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test was attributed to the Controlatron flow instrument reading (3CHS-FI 391) which monitored the induced flow rate. During the test, this flow meter reading was shown to fluctuate from 0.98 gpm to 1.17 gpm. The test result was subsequently accepted by the JI . Pressurizer Heaters and Spray (Appendix 5007)

Pressurizer heaters and-spray effectiveness was checked on January 21, 1986. The inspector noted the following test results:

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Pressurizer pressure response to opening of both normal spray valves (PCV 455B and PCV 455C) was within the test limits as recommended by the vendor (Westinghouse)

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Pressurizer pressure response to energization of all

pressurizer heaters was also within the test limits.

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However, the total heater (Groups A through E)

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capacity (1703.69KW) was found slightly off the test acceptance criteria of 1710-1890 KW TS require that Group A and Group B heaters, which are supplied by emergency power, each has a capacity of at least 175 KW. The measured heater capacities

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for Group A and Group B heaters was 329.9 KW and 330 KW, respectively. Based on the satisfactory results for the pressurizer pressure response and TS requirements, the JTG accepted the test results.

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Constant flow was established through each pressurizer spray line to minimize thermal shock when spray actua-tion was needed. However, due to leakage of both nor-mal spray valves, the bypass valves V15 and V59 were set at 1/16 turn which caused the proportion heater (Group C) to run at about 80% demand. This exceeded the vendor recommended value of 50%. The licensee and vendor are in the process of evaluating these test identified problem . Rod Drop Testing (Appendix 50081 The control rod drop time measurement was performed in accord-ance with test procedure 3-INT-5000, Appendix 5008, Component Rod Drop Testing. A computer timing system (Auto Rod Drop Test System) was set and calibrated prior to its use in measuring the rod drop time. All 61 control rods drop times were measured at cold no flow, cold full flow, and hot full flow condition Under cold test conditions an additional three rod drop time measurements were performed for those control rods whose rod drop times fell outside the two sigma (standard deviation) limit of the drop time for all control rods. While under the hot-full flow test conditions, the fastest and slowest rods were dropped ten additional times in addition to three additional rod drop tests for those rods that fell outside the two sigma limit.

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All test results were consistent and acceptable and within the TS limit of 2.2 second .0 Low Power Physics Tests 4.1 Procedure Review Scope

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The approved test procedures listed below were reviewed for technical and administrative adequacy and to verify that test planning satisfies

, regulatory guidance and licensee commitment INT-7000, Appendix 7004, Revision 1, " Isothermal Temperature Coefficient"

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3-INT-7000, Appendix 7005, Revision 0, "RCCA or Bank Worth

Measurement"

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3-INT-7000, Appendix 7006, Revision 0, " Natural Circulation" l

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Discussion f

The procedures were examined for: management review and approval; procedure format, clarity of stated test objectives; prerequisites; i

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environmental conditions; acceptance criteria; source of acceptance criteria; references; initial conditions; attainment of test objec-tives; test performance documentation and verification; degree of

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detail for test instructions; restoration of system to normal after independent verification of critical steps or parameters, and quality control and assurance involvemen .

Findings l

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The review indicated that the procedures are consistent with regula-tory requirements, guidance, and with the licensee's commitment No discrepancies or unacceptable conditions were identified. The inspector had no further questions on these procedure .2 Test Witnessing and Data Review 4. Low Power Physics Testing At various times during the inspection period, the inspectors witnessed most portions of Low Power Physics Tests (LPPT). The tests witnessed include:

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initia! criticality;

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reactivity computer checkout;

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boron endpoint measurement;

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isothermal temperature coefficient measurement;

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rod worth measurement; and

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flux mapping.

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Low Power Physics Tests were conducted in accordance with the approved test procedure, 3-INT-7000, " Low Power Physics Tests," Revision ,

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Prior to performing each test, briefings with the test crew and operation personnel were conducted and the briefing was adequat Test prerequisites and initial conditions were met.

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Summary analysis was made upon completion of each tes The major events during LPpT were as follows:

, January 23 2200 Reactor Critical January 24 0200 Completed Hot Zero Power Testing Range ,

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Determination. The range of neutron

flux (reactivity computer picoammeter)

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was determined to be 1.6E-8 to 1.6E-7 amp ) 0420 Completed Reactivity Computer Checkout Comparisons of predicted and measured

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i reactivities based on doubling time i measurement were acceptabl j 0715 Commenced ARO Boron Endpoint Measuremen j Measured Endpoint Boron Concentration was 1570.6 ppm. The predicted value was 1566 pp l 1004 Commenced ITC (ARO) measurement l

] 1030 Obtained first set of cooldown/heatup

ITC values a

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1400 Operations personnel found the RCS leakage source (Seal Filter Vent Valve) and isolated it. The inspector noted that at about 1330, a reactor operator was able to identify the RCS leakage based on control room panel indication Obtained additional set of ITC (AR0)

data. The averaged ITC's from heatup and cooldown agreed within il pcm/ The measured ITC (ARO) value of -1.03 pcm/ F agreed well with the predicted value of -1.69 pcm/ F. However, the corresponding MTC result of (+)0.92 pcm/ F exceeded the TS 3.1.1.3 a limi .l The licensee reactor engineer was fully l aware of the plant conditions and cor-responding TS 3.1.1.3 a LCO requirement l l

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l 12 l 2100 Commenced Control Bank (CB) D Rod Worth Measurement (Dilution Method)

2219 Completed CB D Rod Worth Measurement The measured CB D rod worth of 61 pcm was within the predicted range of 593 59 pc January 25 0020 Commenced Control Bank D Inserted j Measurement 0255 Measured Endpoint Boron Concentration (D IN) was 1516.7 ppm. The predicted i value was 1499 pp Commenced ITC (D IN) Measurement -

Cooldown- gp 0447 Commenced ITC (D IN) Measurement -

Heatup. The measured cooldown ITC was identical to the heatup ITC. The measured ITC value of -2.5.pcm/ F was-within the predicted range of -3.24 3 pcm/ Commenced CB C Rod Worth Measurement

(Dilution Method)

0807 Completed CB C Rod Worth Measurement

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The measured CB C rod worth of 1223 pcm was within the predicted range of 1254 125 pcm.

0830 Commenced CB C&D Inserted Measurement Measured Endpoint Boron Concentration (C&D IN) was 1384 ppm. This value was i

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within the predicted range of 1357 136 ppm.

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0935 Commenced ITC (C&D IN) Measurement-The measured ITC VALUE OF -6.07 pcm/ F

was within the predicted range of-6.52 3 pcm/ F. Also, the average ITC's from heatup and cooldown agreed within 1 pcm/* !

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1 l 1100 Commenced Dilution for Rod Worth

Measurement of CB A & B 1704 The measured Endpoint Boron Concentration (A+B+C+D IN) was 1115 ppm. This value was within the

predicted range of 1086 109 ppm.

, The measured CB B rod worth of 123 pcm was within the predicted range of 1208 121 pc The measured CB A rod worth of 121 pcm was also within the predicted range of 1239 124 pc Manually tripped the Reactor per

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TS 4.10.1.2 requirement (In preparation for All-Rods-In Testing)

2000 Started up by pulling CB B 2016 Manually tripped the reactor per Test Procedure 7000, Step 7.12.3.4, also in preparation for All-Rods-In Testin Commenced Reactor Startup

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2102 Reactor Critical with CB A at 40 steps l

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2144 Commenced Shutdown Bank E (SE) rod worth measurement i

1 2206 Completed SE rod worth measurement

i The mesaured SE rod worth of 185.7 pcm

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was within the predicted range of 188

19 pc Commenced Shutdown Bank D (SD) rod

, worth measurement j 2313 Completed SD rod worth measurement i

The measured SD rod worth of 547.8 pcm

! was within the predicted range of 526 53 pc Commenced Shutdown Bank C (SC) rod

worth measurement i

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January 26 0052 Completed SC rod worth measurement The measured SC rod worth of 679.6 pcm was within the predicted range of 655 66 pcm 0130 Measured Single Rod F-2 Rod Worth for information purpos Measured N-1 total rod worth with rod F-2 Stuck Out The measured value of 7925.7 pcm was within the predicted range of 7571 757 pc Commenced All-Rods-In With F-2 Stuck Out Boron Endpoint Measurement 0911 Completed All-Rods-In with F-2 Stuck Out Baron Endpoint Measurement The measured value of 765.2 ppm was within the predicted range of 725 73 pp Manually tripped the reactor per test procedure 7000 step 7.1 Reactor Critical 2238 Commenced control bank rod worth measurement in overlap January 27 0656 Completed control bank rod worth measurement in overlap (Boration Method)

The inspector noted that the measured total overlapping control bank rod worth of 4365.6 pcm was consistent with the previously measured sum (429 pcm) of January 28 LPPT temporarily being hold due to dif-ficulty in getting secondary side chemi-stry in specificatio l l

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I January 29 0313 Resumed LPPT. Started flux mapping with rod configuration at Rod Insertion Limi The measured incore tilt of 1.006 was less 2 than the design criterion of s 1.02 I

! 0600 Measured a single rod D-12 rod worth from Hot Zero Power Rod Insertion Limi ;

The measured value of 396.6 pcm was within

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the safety analysis assumed value of 780 pc ,

1046 Started flux mapping with rod configuration ct Rod Insertion Limit with D-12 fully out to simulate ejected ro '

The incore program picked up this asymmetric

rod configuration with calculated maximum l Fg (Z) = 6.4757 which was less than FSAR as-

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sumed value of 1 January 30 0630 Started flux mapping with rod configuration

, of Control Bank D.In i The measured incore tilt (1.023) was within i

FSAR/ Safety Review Criteria of 1.04, however, I

was slightly.in excess of the design limit of j

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1.0 The licensee reactor engineer evaluat-ed this situation and attributed the cause to the Xenon left from the previous D-12 ejected rod worth / flux mapping tes I A six pass symmetric thimble map taken at i approximately 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> later showed almost

0 tilt. This confirmed the licensee reactor

engineer's evaluation.

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103*, Started flux mapping with All-Rods-Out rod

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configuration.

j The measured incore tilt was the same as the Control Bank D In case.

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The measured FAH in both Control Bank D In

, and ARO cases were acceptabl Completed flux mapping and LPPT.

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4. Low Power Physics Test Results Review

The inspector independently verified that the predicted i values and acceptance criteria were obtained from "The

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Nuclear Design and Core Physics Characteristics of the

Mills., tone Generating Station Unit 3 Cycle 1," WCAP-10791, Revijion 1. The LPPT results are summarized below:

. Endpoint Boron Rod Worth ITC

, (ppm) (pcm) (pcm/ F)

Test Con Mea Predicted Mea Predicted Mea Predicted t

ARO 157 .03 -1.69 l D IN 151 .5 593 -2.50 -3.24 D+C IN 1384 1357 1223 1254 -6.07 -6.52 D+C+B IN - -

123 D+C+B+A IN 1115 1086 121 '

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54 S *b E D IN

D+C+B+A+

SE *b D+bC IN - - -

67 N-1 Rod 76 .7 7571 - -

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i Worth with F-2 Stuck Out Control - -

436 Banks Over-lapping (Boration Method)

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39 i Worth from RIL All measured values were close to and within the analytically predicted ranges with the exception of D-12 ejected rod worth

! measurements from RIL. In this case the measured rod worth of 39 pcm was within the safety analysis assumed value of 780 pc I Although the licensee did not measure the Shutdown Banks A and B rod i

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worth in fuel vendor recommended rod configuration (Table A.4, WCAP-10791), instead an N-1 configuration with Rod F-2 stuck out was

, performed. This configuration provided a direct comparison of

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measured rod worth (7925.7 pcm) against the value assumed in the FSAR analysis (6814 pcm). The measured rod worth confirmed the F5AR j analysis valu I

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Following the discovery of the positive MTC value at ARO.HZP condi-tions, the licensee took correct actions including establishing ad-

ministrative restrictions on Rod Withdrawal Limits and Boron Concen- ,

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i tration Limits. The special report on this subject was submitted to

NRC Region I (Letter from W. D. Romberg (NNECO) to T. E. Murley (NRC),

dated February 3,1986) in accordanc e with TS 3.1.1.3 requiremen f i The inspector verified that the operation personnel were aware of i

.I these restrictions and that the associated curve was located in the ;

control room curve log and was being use :

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The inspector reviewed the low power flux maps and noted that the pre- !

dicted power distribution was generally agreed with the predicted !

values. No unacceptable power distributions were identifie I r

4. Conclusion -

1' I Low Power Physics Test was accomplished in accordanc( with !

approved procedures, data were acceptable, and test ot iectives '

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Licensee performance during approach to criticality and

ubse- f quent LPPT was deliberate, and carefully controlled. T5.,urveil- l lance requirements associated with the special test exceptions !

during LPPT were correctly addressed in the controlling pi ace- :

dure 3-INT-7000, " Lower Power Physics Tests," and adequately per- !

formed. Licensee management was responsive to inspector obser- !

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vations. Problems ~ identified during the test such as positive !

MTC value at all-rods-out condition were disseminated to the :

appropriate groups and corrective actions were implemente !

l 5.0 Power Ascension Tests i t

j 5.1 Procedure Review l

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The approved test procedures listed in Attachment B were reviewed for technical and administrative adequacy and to verify that test

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j planning satisfies regulatory guidance and licensee commitment j d

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Discussion The procedures were examined for: management review and approval; procedure format, clarity of stated test objectives; prerequisities; environmental conditions; acceptance criteria; source of acceptance criteria; references; initial conditions; attainment of test objec-tives; test performance documentation and verification; degree of detail for test instructions; restoration of system to normal after testing; identification of test personnel; evaluation of test data, independent verification of critical steps or parameters, and quality control and assurance involvemen Findings The review indicated that the procedures are consistent with regula-tory requirements, guidance, and with the licensee's commitment No discrepancies or unacceptable conditions were identified. Ques-tions raised by the inspector to the licensee relating to the proce-dures were satisfactorily answere It was noted that the licensee is planning on seeking from the NRC full credit for two R.G.I.68 required tests: the ejected rod test at above 10% full power and the dropped rod test at 50% full power. This request will be based on the fact of these tests have been adequately demonstrated at prototype plants similar to MP-3. In a letter to NEU from Westinghouse dated January 20, 1985, Westinghouse concurred with NEU and stated that "The cost in time and the probability of subse-quent undesirable power distributions resulting from these tests lead to the conclusion that they should be dropped from the Millstone Nuclear Power Station Unit 3 test program." Westinghouse supported this statement with a table showing the number of operating plants similar in design to MP-3 which had successfully completed the tes lhe licensee committed in the FSAR Table 14.2-2 item number 30, to perform these two tests and will therefore be seeking a change to the FSAR from NRC allowing deletion of the This is an unresolved item (50-423/86-07-02).

5.2 Test Witnessing Portions of the following Power Ascension Test and the entirety of the Natural Circulation Test were witnessed by the inspectors. This included the initial test preparation and test restoration as well as witnessing the actual tes INT-7000, Appendix 7006, " Natural Circulation" Revision 0

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3-INT-8000, Appendix 8018, " Automatic Steam Generator Level Control", Revision 0, Step 7.1 - Testing at Low Powe INT-8000, Appendix 8013, " Steam Dump control", Revision 0

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These tests were witnessed for the following attributes: Appropriate procedure revision was available and in use by all '

crew member i Minimum crew requirements were me i All test prerequisites and initial conditions were met and/or  !

! those which were waived were reviewed / approved in accordance  ;

j with procedure / technical specification (TS) requirement <

I Test equipment required by the procedure was calibrated and in

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service.

.1 I Test data equipment required by the procedure was calibrated to a common time bas . Test was performed as required by a technically adequate

procedur . Crew actions aopeared to be correct and timely during the performance of the tes Coordination was adequat . Quick summary analysis was made to assure proper plant response i to the tes . All data was collected for final analysis by the proper

, personne . Overall test acceptance criteria were me . The licensee's preliminary test evaluation was consistent with inspector's observatio i 12. Adherence to TS requirements was maintained for those tests which affect TS LCO The following discussion pertains to these test . Natural Circulation Test (Appendix 7006)

It took approximately ten minutes to establish natural circulation after the reactor coolant pumps had been trippe Natural circulation took place when AT reached approximately.

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e,, . -p. - - e . g - - - - , , , - - - - - - .,..,_p .-,.,,,,,.yc,_ ..c. , , . . . , - . w- , , - , . . , ,r ..9--9

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37 F (TH0T ~ 594 F and TCOLD ~ 557 F). There appeared to be close correlation between the incore thermocouple readinos and T

H0 The majority of incore thermocouple readings ranged between 596 F and 602 After the RCS had stabilized on natural circulation, the letdown / charging system was initiated to see what affect it had on cooling down the primary coolant. It was discovered that the RCS began cooling down at a rate of approximately 1 F every four minutes. Questions raised by the inspector concerning the results of the test were satisfactorily answered by the license . Steam Generator Water Level Control Test (Appendix 8018)

This test was first performed on February 10, 1986. During the test, I&C personnel inadvertently shorted the steam generator level control system circuits which caused all feedwater regula-ting bypass valves to clos This resulted in a S/G low level reactor trip. The test was resumed on February 11, 1986. The inspector noted that a lesson learned from an earlier reactor trip which occurred on February 7,1986, was incorporated. Auc-tioneered high nuclear flux was fed to the bypass control cir-cuits to position the bypass valves during low power automatic S/G level control. Following the February 7, 1986 reactor trip, the bypass valve response was determined to be too fast. Cur-rently a gain adjustment factor of 2.44% valve opening /1% NI power was set in lieu of 8% valve opening /1% NI powe Preliminary test results indicated that the automatic S/G level control was stable. No unacceptable conditions were identifie . Steam Dump Test (Appendix 8013)

The steam dump test was performed in accordance with procedure Appendix 8013, " Steam Dump Control," Revision 0. This test was originally planned to be performed at 0% power. Test Change No. 3 modified the test procedure to be performed at 15% power level. There appeared, however, to be a lack of proper review for Test Change No. 3. As a result, during initial performance of Step 7.4, Plant Trip Controller Response Test, the procedure was found to difficult to follow. The test engineer quickly identified the problem and corrected the situation through Test Change No. 6. which was further reviewed by the PORC/JTG. Upon PORC/JTG approval, the test was resumed and completed without inciden . . -. . - . . - - ._ .- _ . - . ._. . - - . .

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Preliminary test results indicated that the test objectives

} Were met and test data were acceptable.

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l 6.0 Independent Calculation

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j During the review of Post Core Hot Functional Test, Appendix 5009 - Pre-i critical RCS Flow Measurement, the inspector verified that the calcula-

! tions and math used to convert percentage of flow to inches of water (AP)

J and to calculated flow rates were correct.

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! Also, as described in Section 3.2.1 detailed RCS leak rate calculation j was performed using NRC-developed leak rate comouter program NUREG-1107, j "RCSLK9: Reactor Coolant System Leak Rate Determination for PWRs", to

verify the licensee's calculatio .0 Quality Assurance / Quality Control i

The site QA/QC organization, manpower and planned activties to cover the startup test program was described in the NRC inspection report 35-5 During this inspection period, the inspectors noted that QC inspectors were actively following startup program tests. Appropriate surveillance reports were issued.

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i No unacceatable conditions were identified.

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j 8.0 Plant Tours *

The inspe: tor made several tours of the facility during the course of the i inspectio1. This included tours of the turbine building, control building,

and contral room. A review of the work in progress, security, cleanliress i and housekeeping was made.

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9.0 Exit Meeting

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An exit meeting was held on February 14, 1986 to discuss the inspection j

scope and findings, as detailed in this report (see paragraph 1.0 for attendees).

At no time was written material given to the license l The inspector determined that no proprietary information was utilized ,

during this inspection.

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APPENDIX A POST CORE HOT FUNCTIONAL TEST RESULTS REVIEWED

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Appendix 5004, Rod Control Slave Cycler and CRDM Timing

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Appendix 5006, RCS Leak Detection

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Appendix 5007, Pressurizer Heaters and Spray

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Appendix 5008, Rod Drop Testing

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Appendix 5009, Precritical RCS Flow Measurement

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Appendix 5011, Moveable In-Core Detector Operation

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Appendix 5018, Rod Control Operational Test

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APPENDIX B POWER ASCENSION TEST PROCEDURES REVIEWED i

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3-INT-8000, Power Ascension Test, Revision 0

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3-INT-8000, Appendix 8001, Calorimetric, Revision 0

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3-INT-8000, Appendix 8002, Nuclear Instrumentation Operational Alignment Verification, Revision 0

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3-INT-8000, Appendix 8003, Calibration of Steam Flow and Feedwater Flow

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Instrumentation at Power, Revision 0

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3-INT-8000, Appendix 8004, Operational Alignment of Process Temperature Instrumentation, Revision 0

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3-INT-8000, Appendix 8005, Reactor and Turbine Control, Revision 0 i

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3-INT-8000, Appendix 8007, Radiation Monitoring System, Revision 0

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3-INT-8000, Appendix 8008, Ventilation System Operability, Revision 0

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3-INT-8000, Appendix 8009, Plant Chemistry, Revision 0

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3-INT-8000, Appendix 8010, Neutron Shield Tank Cooling Test, Revision 0

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3-INT-8000, Appendix 8011, Containment Penetration Temperature Monitoring, Revision 0

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3-INT-8000, Appendix 8013, Steam Oump Control, Revision 0

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3-INT-8000, Appendix 8015, RCS Flow Measurement, Revision 0

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3-INT-8000, Appendix 8016, Turbine Overspeed, Revision 0~

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3-INT-8000, Appendix 8017, Automatic Reactor Control, Revision 0

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3-INT-8000, Appendix 8018, Automatic Steam Generator Level Control, Revision 0

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3-INT-8000, Appendix 8019, Turbine Plant Component Cooling Water System I

Balancing, Revision 0 j --

3-INT-8000, Appendix 8020, Power Coefficient, Revision 0

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3-INT-8000, Appendix 8022, 10% Load Swing Tests, Revision 0 i

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3-INT-8000, Appendix 8023, Reactor Trip / Shutdown Outside Control Room, Revision 0

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3-INT-8000, Appendix 8026, Large Load Reduction, Revision 0

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3-INT-8000, Appendix 8029, Pipe Fluid Transient Vibration Testing, Revision 0

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3-INT-8000, Appendix 8030, Loss of Power (20% Power), Revision 0

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3-INT-8000, Appendix 8031, Reactor Coolant System Boron Measurement, Revision 0

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3-INT-8000, Appendix 8034, Thermal Expansion Restraint, Revision 0

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3-INT-8000, Appendix 8035, Loose Parts Monitoring System, Revision 0

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3-INT-8000, Appendix 8037, Main Steam Isolation Valve Closure Test, Revision 0

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