SNRC-1234, Suppl 1 to Startup Rept for 851115-860215

Suppl 1 to Startup Rept for 851115-860215

ML20153E408
Person / Time
Site:	Shoreham File:Long Island Lighting Company icon.png
Issue date:	02/14/1986
From:	Leonard J, Steiger W LONG ISLAND LIGHTING CO.
To:	Harold Denton Office of Nuclear Reactor Regulation
References
SNRC-1234, NUDOCS 8602250027
Download: ML20153E408 (17)

Text

I LONG ISLAND LIGilTING COMPANY Shoreham Nuclear Power Station \

Supplement No. I to the

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STARTUP REPORT for the period November 15, 1985 to February 15, 1986

,, i '/ '/If[JU Approved:

s '115 $ ) Date W. E. Steiger Plant Manager

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TABLE OF CONTENTS 1.0 Overview 2.0 Chronology of Key Events 3.0 Startup Test Results 3.1 Control Rod Drive (STP-5 FSAR 14.1.4.8.5) 3.2 Water Level Measurement (STP-9 FSAR 14.1.4.8.6) 3.3 LPRM Calibration (STP-ll FSAR 14.1.4.8.9) 3.4 Process Computer (STP-13 FSAR 14.1.4.8.11) 3.5 RCIC System Startup Test (STP-14 FSAR 14. l~ .4.8.12) 3.6 HPCI System Startup Test (STP-15 FSAR 14.1.4.8.13) 3.7 Recirculation Flow Control System (STP-29 FSAR 14.1.4.8.26) 3.8 Reactor Water Cleanup System (STP-70 FSAR.14.1.4.8.33) 3.9 Residual lleat Removal System (STP-71 FSAR 14.1.4.8.34) 3.10 Peactor Building Closed Loop Cooling and Drywell Air Cooling (STP-37 FSAR 19.1.4.8.36) 3.11 Service Water System (STP-42 FSAR 14.1.4.8.39) 3.12 loose Parts Monitoring System (STP-814 FSAR 14.1.4.8.40) 4.0 License Conditions page 2

1.0 Ove rview This Startup Report Supplement has been written by the Long Island Lighting Company (L1LCO) for submittal to the Nuclear Regulatory Commission in compliance with Shoreham Nuclear Power Station Technical Specifications, paragraphs 6.9.1.1 through 6.9.1.3, and Regulatory Guide 1.16, Revision 4 section C.I.a.

Technical Specliications and Regulatory Guide 1.16 require that a summary report of plant startup and power escalation testing he submitted within 9 months following initial criticality. Shoreham Nuclear Power Station (SNPS) achieved initial criticality on February 15, 1985. LILCO submitted a Startup Report on November 15, 1985. The scope of that Startup Report included f uel load and initial criticality under the .001% power license, NPF-19, and l ow powe r [

testing conducted under the 5% power license, NPF-36.

Since SNPS has not completed its startup test program or commenced comme rcial power operation, additional Supplements are required on a periodic basis. This Supplecent covers the three month period from November 15, 1985 to February 15, 1986.

Shoreham has been in an outage since October. During this outage, the installed neutron sources were replaced and environmental qualifica-tion modifications were completed. The environmental qualification of plant equipment was required as a condition of the Operating License.

Water level instrumentation reference leg piping from the reactor vessel to the drywell penetration was modified due to vessel level indication problenn experienced during low power testing. A modification was made to the liigh Pressure Coolant Injection system to improve turbine startup response. Several temperature monitoring ins trume nt s in the drywell were relocated and insulation around the biological shield was modified to reduce local drywell temperatures.

Pipe supports identified during thermal expansion walkdowns as actual or potential interfere,ce problems were replaced or modified.

Installation of a coriun ring underneath the reactor vessel is in progress.

At the direction of the Shoreham Plant Manager, the test organization reviewed the experiences of fuel loading and low power testing. Plant startup test procedures (STP's) and the station procedure for the administration of power ascencion testing (SP 12.075.01) were reviewed '

and revised to incorporate the lessons learned f rom those expe riences.

Test Condition !!catup test results have been reviewed and approved by i the Test Review Comni t tee (TRC) and the Review of Ope rations Commit tee l (ROC). Som of the station modifications implenented during the l outage affect prior test results. Conseq uently , some startup tests .

will be repeated. Plant management has scheduled a series of tests to i be pe rf o rmed a t less than 5% power. The schedule includes those tests that are identified in Chapter 14 of the FSAR to be completed in Test page 3

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7 1.0 overview - (continued)

Condition Heatup which were previously performed but whose results may have been affected by plant modifications. Certain " opportunity" tests (for example, completion of Reactor Core Isolation Cooling system controller tuning and main generator synchronization) that are not required until higher power but which can be performed at less than 5% power are also scheduled. Startup is planned for early March and the scheduled duration of testing is approximately twenty-nine i days. Af ter the completion of scheduled testing, the plant will be

ready for power escalation t.n Test Condition 1.

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2.0 Chronology of Events December 7,1984 Received low powe r license .001%.

December 17, 1984 Sources loaded in core.

Decembe r 21, 1984 Fuel loading comnenced.

January 4,1985 Partial core shutdown margin test.

Janua ry 19, 1985 Fuel loading completed.

January 25, 1985 CRD open vessel testing completed.

Februa ry 17, 1985 Initial critical and shutdown norgin test.

Completed open vessel testing.

June 4, 1985 CRD open vesNel retest completed.

July 3, 1985 5% low power license received.

July 7, 1985 Reactor critical sequence B.

Test condition lleatup.

July 7, 1985 llentup to 250*F .

System expansion perfonned.

IRM performance completed.

July 8, 1985 Heatup to 32 5'F pe rfo rex 3d .

APRM calibration.

July 9, 1985 SRV functional test pe r fo rcrd (STP-26) .

July 11, 1985 150 pstg plateau reached.

System expansion IM ent ry.

RCIC testing.

IIPCI system testing.

July 14, 1985 Reactor scram #1 on Rx level.

July 17, 1985 APRM calibration at 150 psig, itPCI testi ng.

System expansion testing.

RBCLCW performance testing.

July 18, 1985 Fenctor shutdown for RPV level instrumentation work.

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i 2.0 Chronology of Events - (continued) j July 23, 1985 Reactor critical.

July 26, 1985 Reactor shutdown for RPV level instrumentation work.

July 29, 1985 Reactor critical.

July 31, 1985 Reactor pressure to 150 psig.

lIPCI system testing.

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) August 1, 1985 350 psig plateau.

2 Drywell radiation survey.

j System expansion drywell inspection, system expansion data.

l i August 3, 1985 600 psig plateau.

System expansion.  !

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RBCLCW performance.

CRD testing.

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Augus t 5, 1985 800 psig plateau.

j System expansion.

j CRD testing.

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! Augus t 7, 1985 Rated pressure plateau.

{ System expansion testing.

l. IIPCI testing.

] RCIC testing.

! CRD testing.

j RBCLCW testing.

Water level testing.

j STP-13 testing.

j Chemical and radiochemical testing.

Radiation measurements.

LPRM testing.

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i i August 23, 1985 Reactor pressure reduced to 150 psig.

IIPCI testing.

j RCIC testing.

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August 24, 1985 Reactor shutdown af ter initial heatup.

I j August 30, 1985 Reactor critical sequence A.

August 31, 1985 Scram #2 on loss of instrument air.

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i 2.0 Chronology of Events - (continued) 4 ,

September 3, 1985 Reactor critical sequence A. Second heatup to 1

150*F.

September 4, 1985 Heatup to 250'F - 150 psig.

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System expansion testing.

i September 6,1985 Reactor pressure 350 psig.

] System expansion testing.

September 6,1985 Reactor scram #3 due to surveillance on level ,

1 ins t rume nt.

} Sep* ember 7, 1985 Reactor pressure 600 psig.

I System expansion testing.

J CRD testing.

I September 8, 1985 Unusual event due to Reactor level indication

{ problem.

September 10, 1985 Reactor shutdown for repair of Reactor level problem.

! Reactor critical and heatup to investigate September 11, 1985 I level problem. -

1, i September 12, 1985 Reactor scram #4 on low water level indica-tion.

Actual level did not change.

4 Level instrument problem investiga ted.

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September 18, 1985 Reactor critical sequence A.

l September 21, 1985 Reactor pressure 800 psig.

l 4 System expansion testing.

I CRD testing.

RPV level problem fixed.

September 21, 1985 Rated pressure reached.

System expansion testing.

l' Second heatup.

RCIC testing.

HPCI testing.

MSIV testing.

l Water level testing.

J CRD testing.

I RWCU testing.

Radiation testing.

l Chemist ry testing.

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l 2.0 Chronology of Events - (continued)

$ September 27, 1985 Reactor shutdown.

! Hanger placed on B reference leg. Miscellan-2 eous maintenance.

October 3,1985 Reactor critical sequence A. Third heatup to i rated pressure.

System expansion testing.

October 4, 1985 RCIC vessel injection at rated pressure.

System expansion testing.

! RWCU system testing.

i MSR relief valve testing.

October 6,1985 Initial turbine generator roll to rated speed.

i October 8, 1985 End af heatup testing.

j begin source replacenent outage.

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3.0 Test Results The plant has been shutdown during the entire period covered by this Sup plemen t. There are no new test results to report. As a result of testing performed during Test Conditions Open Vessel and li?atup (discussed in the Shoreham Startup Report, dated November 15, 1985),

some corrective actions were identified which require repeating certain tests. Some of the equipment modifications made during the current outage invalidated previous test results, and, therefore, certain other tests need be repeated. Tne following is a brief discussion of startup tests that have been scheduled for low power testing prior to power ascension to Test Condition 1.

3.1 Control Rod Drive (STP-5)

Two activities are scheduled for the next startup to complete STP-5 for Test Condition IIcatup. One activity is to obtain the individual rod scran time for rod 22-35 at rated pressure and temperature. The other activity is to perform tuning on the Control Rod Drive liydraulic system flow controller.

A measureable difference in control rod drive normal insert and withdrawal stroke times was observed for selected rods which were timed at zero reactor pressure, at inte rmediate pressures ,

and at rated reactor' pressure. Based on an evaluation of the stroke time data for the selected rods, thirty-four rods will have their normal witidrawal stroke time adjusted prior to the next startup.

3.2 Water Level Measurement (STP-9)

The problems with water level indication ceen during low power operation were attributed to condensation in the i ns t rume nt piping between the reactor vessel and the condensing chamber.

New piping has replaced the old, with improved pipe slope and shorter piping runs. All the narrow and wide range level instruments have been recalibrated. Correct response on the level indicators will be verified by lowering water level to normal prior to reassenbling the vessel. Startup test data will be collected at rated temperature and pressure to confirm that all the af fected instruments meet the criteria for agreement.

3.3 LPRM Performance (STP-ll)

The 'C' detector in the nuclear instrumentation local power range monitor (LPRM) string 20-37 did not show any response tc control rod wi thdrawal. The 'A', 'B', and ' D' de tectors in the same string responded properly. The cable connector under the vessel was repaired during the current outage. De t ect o r response will be verified in section 8.6 of STP-il, scheduled during Test Condition 1.

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3.4 Process Computer (STP-13)

Traversing Incore Probe (TIP) tubing underneath the reactor vessel was disassembled to permit control rod drive maintenance.

New core limits will be measured and progrer.med into the process computer during testing scheduled following the next startup.

3.5 RCIC (BTP-14)

Some controller tuning was performed with the Reactor Core Isolation Cooling (RCIC) system injecting to the reactor vessel during Test Condition Ibatup. System speed and flow response were acceptable at rated flow conditions ; however, additional controller tuning is required to increase speed loop stability at reduced flow conditions. Following completion of this additional required tuning, a series of quickstart demonstration tests will be performed. -

The RCIC turbine exhaust check valves were replaced during the current outage. Minimum turbine operating speed at which che k valve cycling does not occur will be determined with steam supply pressure to the turbine at 150 psig during the RCIC retest scheduled for March.

3.6 itPCI (STP-15)

The fligh Pressure Coolant injection (llPCI) system Woodward Governor RGSC (ramp generator and signal convertor) enclosure was relocated as part of the station's environmental qualification codifications. A " hydraulic bypass" was installed around the llPCI EGR actuator to allow control oil pressure to cycle the control valve fully open and then partially closed before the stop valve opens on system startup. This modification should significantly improve the startup response of the llPCI system. 'these two modifications require that all llPCI functional and stability demonstration tests be repeated.

A ilPCI endurance run is also scheduled for the March retest program. An endurance run was attempted in September and aborted because indicated suppression pool temperaturen approached the plant Technical Specifications limiting condition for operation. The high temperatures resulted from thermal stratification of the suppression pool during IIPCI operation.

Local to bulk suppression pool temperature differences on the order of 15'F were observed. Permanent corrective action is under evaluation. 'Ihe endurance run is scheduled with the remaining low power testing, although not required to be completed before the end of Test Condition 3.

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j 3.7 Recirculation Flow Control System (STP-29)  ;

i f The purpose of the test performed in Test Condition IIcatup was '

! to demonstrate that the initial controller settings are stable j at low power and core flow. This test demonstrated satisfactory i stability and established a new interim lower limit for operating the recirculation system motor generator sets (MG's).

The new lower limit was established to avoid the limit cycles

{ which were observed on the 'B' MG. An administrative limit of i 24% was established, and the electrical low speed limiter will

! be adjusted to 24% prior to startup. No additional l j recirculation system testing is required before the next regularly scheduled testing at Test Condition 1. [

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l 3.8 Reactor Water Cleanup System (STP-70)  ;

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Recalibration of Reactor Water Cleanup (RWCU) system flow [

transmitters and a thorough check-out of the leak detection circuit are believed to have corrected the previously reported i problem with frequent spurious system isolations.  ;

i Calibration of the bottom head drain flow indicating loop was l l not successfully completed during Test Condition lleatup. The l

! installed differential pressure transmitter was overranged with .

RWCU system flow routed entirely through the hottom head drain  !

j line. The installed transmitter is scheduled to be replaced by

! February 27, and the calibration test (STP-70, Appendix A) is j scheduled for the upcoming series of plant restart tests.

r I 3.9 Residual ifeat Removal system (STP-71) l j The pupose of this test is to demonstrate the ability of the Residuni lleat Removal (RilR) system to remove decay and residual ,

j heat from the reactor coolant system and from the suppression

! pool. A suppression pool cooling test is scheduled to be .

j performed in conjunction with a llPCI endurance run. (See {

4 Section 3.6 for a discussion of suppression pool thermal  !

stratification). (

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3.10 Reactor Building Closed Loop Cooling and Drywell Air Cooling (STP-37)

The purpose of this test is to demonstrate that with the maximum design Reactor Building Closed Loop Cooling Water (RBCLCW) supply temperature, all RBCLCW-supplied components are adequately cooled and that the Drywell Air Cooling (DAC) system maintained drywell temperatures within plant Technical Specification limits. Because modifications nade to insulation ano instrumentation in the drywell af fect the overall heat load on the system and the temperature distribution seen by the temperature nonitoring instruments, all sections of this test completed in Test Condition Heatup have been scheduled to be repeated during the next plant hentup.

3.11 Service Watet System (STP-42)

The purpose of this test is to demonstrate that the service water system adequately cools its normal service heat loads.

This test, although scheduled for Test Condition Heatup, was not performed before the plant shut down for the current outage.

This test is now scheduled for the next startup and low power testing.

3.12 Loose Parts Monitoring System (STP-814)

The purpose of this test is to perform the initial set-up of the Loose Parts Monitoring System (LPMS), to adjust the system's sensitivity and alarn setpoints, and to demonstrate the ability to detect impacts as required by Regulatory Guide 1.133.

Plant personnel were unable to successf ully complete the initial sensitivity adjustment. The system vendor was called to the site to troubleshoot and repair the system. This test is scheduled to be performed concurrently with neutron nonitoring system detector drive-in and drive-out operations during the next startup.

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i 4.0 LICENSE CONDITIONS j

i Technical Specifications require that the Startup Report discuss the i license conditions which affect plant startup and power escaintion

) testing. The specific license conditions are delineated in paragraph i 2.C of the Shoreham Orcrating License (NPF-36 of July 3, 1985). Each condition is sumrurized and its statun, as it applies to the completed portion of the test program, is provided below.

t j 4.1 Condition: The vnximum core thetual pot er shall not excerd 51 1 rated core thermal power.

! Status: During the period of this report the reactor was j maintained in the refueling mode.

4.2 Condition

The plant shall be operated in accordance with the

'acchnical Specifications and the Environnental Protection Pla n .

I Status: The lew Power Test Program has be en conducted in j accordance with Technical Specifications and the Environmental i Protection Plan. In a very few circumstances, the provisions

] of Technical Specifications were exceeded. Each circuantance

] has been previounty reported to the NRC and has been the subject of a Licensen Event Report (LER).

I j 4.3 Condition: The plant shall ruintain the fire protection

{ program as described in the Fire llazards Analysin Report and i in the FSAR.

j Status: The provisions of the Fire Protect ion Program have j been adhered to, except for a single minsed fire watch in a j law Pressure Coolant Injection (LPCI) rotor-generator net i room. This fire watch was minned because the door latch was b roke n. A sira complete description will he provided in n Licensee Event Report which in currently in preparation.

< 4.4 Condition: Changes to the initial tent program nhall bc

) reported within one month.

I j Statum: All changes to the test progran an dencribed by J Chapter 14 of the FSAR will be reported to the NRC prior to their implementation.

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, 4.5 Condition: The initial inservice inspection shall be

! developed and implemented before the first refueling outage.

Status: This condition is not af fected by the Low Power Test Program. Development of the inservice inspection program is I

in progress.

4.6 Condition

Control rods shall be tested for boron loss after the first refueling outage.

Status: This condition is not yet applicable.

4.7 Condition

The provisions of the NUREC-0737 action plan described in the SER, supplements 1 and 4, shall be followed.

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4.7.1 Status

To date, the qualifications of five of the required seven backup STA's have been submitted to and approved by the Commission. The remaining two positions are unfilled. LILCO is currently preparing a letter that will submit the resume of a qualified candidate for the sixth position.

4.7.2 Status

The requirement to mark control room indicators with operating limits, and with trip and alarm setpoint values is not yet impl ement ed. The requirements of the provision rennin under review and shall be impleernted prior to the completion of this startup test program.

4.7.3 Status

Modifications to the post accident sample f acility, which will enable sampling using the modified core damage procedure, are in progress as noted in Section 3.21 of IE Inspection Heport 85-038. This requirement is to be tracked by Region I as 50-322/85-04-19.

4.7.4 Status

The modifications required to impicment the emergency response capabilities, as required by Attachment 1 to the license, are in progress.

4.8 Conditions prior to November 30, 1985 all electrical equipment shall be qualified.

Status: LILCO sought and was granted an extension of the tiovembu 30, 1985 deadline by the Commission. Station inodifications to environmentally qualify the required electrical equipment are now complete.

4.9 Condition

The remote shutdown system shall be improved prior to the first startup following the first refimling outage.

Status: The modifications will be implemnt ed as required.

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4.10 Condition: The RilR systeta may not be operated in the steam

'c~ond'e~nsing modo except under emergency conditions.

Status: The station procedures have been modified to preclude Ehe steam condensing mode of operation except as a last resort when all other methods of core and containment cooling have failed.

4.11 Condition: TVo containment isointion barriers in series will j be installed by the end of the firat refuel outage. t i

Status: Methods to satisfy this condition will be completed l by the end of the first refuel outage. ,

4.12 Condition: The provisions of Appendix 3 to the itcence shall '

be satisfied as they apply to the TDI diesel generators.

.S t a t us : Station procedures and maintenance schedules have been modified to include the required TDI diesel generator tests and inspections. Currently, new procedures are being written to segregate the inspections required by Attachment 3 from the regular technical specification surveillances and preventive maintenance procedures.

4.13 Condition: The results of the independent design review shall be incorporated prior to exceeding 5 percent power.

Status: Complete.

4.14 Conditions a.) Prior to exceeding five percent power, radiation monitoring panclu IDil*pHL-ll 7A. H and radiaton monitoring pumps 1D11*P!26, 134 shall be qualified. b.) The invessel storage racks shall be qualified prior to use.

Status a.) The required qualification for the radiation monitoring equipment has been completed. b.) The invassel storage racks have been administrativo1y prohibited from use.

4.15 Condition: 1hc plant shall have on-shif t advicors as required by Attachment 2 of the license.

Status: 1he plant currently has sufficient numbers of qualified on-shift advisors to satisfy this condition.

4.16 Condition: 11m Emergency Core Cooling Systemn perf ormance shall be reanalyzed for the second cycle and beyond, utilizing models that cerount for burnup gas pressure and local oxidation and which are approved by the NRC.

Statues 1his condition will be completed by the required tist e'.~~

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4.17 condition: The licensee shall implement the response to Generic Letter 83-28 on schedule.

Status: All provisions stipulated in SKRC-1013,1116,1184, and 1217 are complete except for the provision that requires the Safety Parameter Display System (SPDS) to be operable prior to the first restart after the first refueling and the review of Technical Specification testing intervals. This review will be done as committed to in SNRC-Il84.

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N'%25G16M@btMiPC Tete.CrL*

gggl t.w m w m ;x# aaa#wrj LONG ISLAND LIGHTING COMPANY SHOREHAM NUCLEAR POWER STATION P.O. BOX 618. NORTH COUNTRY ROAD + WADING RIVER. N.Y.11792 JOHN O. LEON AR D. JR.

ViCE PaESIDENT NUCLEAR OPERAf t0NS February 15, 1986 SNRC-1234 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555 Supplement No. I to the Startup Test Report Shoreham Nuclear Power Station Docket No. 50-322

Dear Mr. Denton:

Attached is Supplement No. 1 to the Startup Report written by the Long Island Lighting Company in compliance with Shoreham Nuclear Power Station Technical Specifications, paragraph 6.9.1.1 through 6.9.1.3, and Regulatory Guide 1.16, Revision 4, section C.1.a.

This report addresses each of the startup tests identified in Chapter 14 of the FSAR to be perforned in the test conditions of Open Vessel and Heatup. It includes a description of the measured operating condition values or characteristics obtained during the test program period between November 15, 1985 to February 15, 1936, with a comparison of these values to the acceptance criteria, together with a description of any corrective actions required to obtain satisfactory operation.

If there are any questions, please contact this office.

Very truly yours,

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D. Leonard, Jr.

Vice President-Nucle [ar I)perations

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.Ng G fG ck Attachment cca J. A. Derry I\