ML20198A001
| ML20198A001 | |
| Person / Time | |
|---|---|
| Site: | Shoreham File:Long Island Lighting Company icon.png |
| Issue date: | 05/15/1986 |
| From: | Steiger W LONG ISLAND LIGHTING CO. |
| To: | |
| Shared Package | |
| ML20197K229 | List: |
| References | |
| NUDOCS 8605200299 | |
| Download: ML20198A001 (14) | |
Text
, ,, _
- l LONG ISLAND LIGHTING COMPANY Shoreham Nuclear Power Station Supplement No. 2 to the STARTUP REPORT for the period February 15, 1986 to May 15,1986
, Approved:
- 4 ff W. E. Steiger,Jrl/ Date Plant Manager I
B605200299 86051632 PDR ADOCK O ,;
P
TABLE OF CONTENTS 1.0 overview 2.0 . Chronology of Events 3.0 Startup Test Results 3.1 Activities Performed 3.1.1 Control Rod Drive (STP-5 FSAR 14.1.4.8.5) 3.1.2 Reactor Water Cleanup System (STP-70 FS AR 14.1.4.8.33) ~
3.1.3 Loose Parts Monitoring System (STP-814 FSAR 14.1.4.8.40) 3.2 Remaining Heatup Phase Testing 3.2.1 Control Rod Drive (STP-5 FSAR 14.1.4.8.5) 3.2.2 Water Level Measurement (STP-9 FSAR 14.1.4.8.6) 3.2.3 Process Computer (STP-13 FSAR 14.1.4.8.11) 3.2.4 RCIC System Startup Test (STP-14 FSAR 14.1.4.8.12) 3.2.5 HPCI System Startup Test (STP-15 FSAR 14.1.4.8.13) 3.2.6 Recirculation Flow Control System (STP-29 FSAR 14.1.4.8.26) 3.2.7 Reactor Water Cleanup System (STP-70 FSAR 14.1.4.8.33) 3.2.8 Residual Heat Removal System (STP-71 FSAR 14.1.4.8.34) 3.2.9 Reactor Building Closed Loop Cooling and Drywell Air Cooling (STP-37 FSAR 14.1.4.8.36) 3.2.10 Service Water System (STP-42 FSAR 14.1.4.8.39) 4.0 - License Conditions
1.0 Ove rview Startup Report Supplement No. 2 has been written by the Long Island Lighting Company (LILCO) 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 Specifications and Regulatory Guide 1.16 require that a summary report of plant startup and power escalation testing be 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 which included fuel load and initial criticality under the .001% power license, NPF-19, and low power testing con-ducted under the 5% power license, NPF-36.
Since SNPS has not completed its startup test program or commenced commercial power operation, additional Supplements are required on a periodic basis. Supplement No. I to the Startup Test Report covered the period from November 15, 1985 to February 15, 1986. It included the completion of the Neutron Source Outage and the beginning of the Reference Leg Replacement Outage. Supplement No. 2 covers the subsequent three months of Shoreham's activities, from February 15, 1986 through May 15, 1986.
Feb rua ry 15, 1986 was the twenty-eighth day of Shoreham's Reference Leg Outage. Major tasks in progress were the Corium Ring Modifi-cation, the Reactnr Pressure Vessel Reference Leg Modification and valve work on the Residual Heat Removal System. All three of these tasks were completed in March. Additional tasks scheduled through the latter part of Fbrch into April included work on the High Range Radiation Monitor Panels and the replacement of the Recirculation Pump Discharge Valve Operator Motors.
On April 17, 1986, the Reactor Mode Switch was placed in the Startup position, in preparation for startup for synchronization of the main generator to the grid and completion of Test Condition Heatup testing. The Loose Parts Fbnitor System was tested in conjunction with SRM withdrawal but then the Reactor Mode Switch was placed in Shutdown after a condenser tube sheet leakage problem was identi-fled.
As of this writing, condenser restoration work is in progress and its duration and scope are under evaluation. Plant management has scheduled a series of tests to be performed at less than 5% power.
The schedule includes those tests that are identified in Chapter 14 of the FSAR to be completed in Test 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 req uired until higher power but which can be performed at less than 5% power are also scheduled. The scheduled duration of testing is approxi-mately thirty-five days. Af ter the completion of scheduled testing, the plant will be ready for power escalation to Test Condition 1.
2.0 Chronology of Events December 7, 1984 Received low power license .001%.
December 17, 1984 Sources loaded in core.
December 21, 1984 Fuel loading commenced.
January 4, 1985 Partial core shutdown margin test.
January 19, 1985 Fuel loading completed.
January 25, 1985 CRD open vessel teeting completed.
February 17, 1985- Initial critical and shutdown margin test.
Completed open vessel testing.
June 4, 1985 CRD open vessel retest completed.
July 3, 1985 5% low power license received.
July 7,1985 Reactor critical sequence B.
Test condition Heatup.
July 7, 1985 Heatup to 250*F System expansion performed.
IRM performance completed.
July 8, 1985_ Heatup to 325'F performed.
APRM calibration.
July 9, 1985 SRV functional test performed (STP-26).
July 11,1985 150 psig plateau reached.
System expansion DW entry.
RCIC testing.
HPCI system testing.
July 14, 1985 Reactor scram #1 on Rx level.
July 17, 1985 APRM calibration at 150 psig.
HPCI testing.
System expansion testing.
RBCLCW performance testing.
I
- July 18, 1985 Reactor shutdown for RPV level instrumentation
- work.
l- July 23, 1985 Reactor critical.
July 26, 1985 Reactor shutdown for RPV level instrumentation wo rk.
July 29, 1985 Reactor critical.
July 31,1985 Reactor pressure to 150 psig.
HPCI system testing.
L:
r-2.0 Chronology of Events - (continued)
Augus t 1, 1985 350 psig plateau.
Drywell radiation survey.
System expansion drywell inspection, system ex-pansion data.
August 3, 1985 600 psig plateau.
System expansion.
RBCLCW performance.
CRD testing.
August 5, 1985 800 psig plateau.
System expansion.
ORD testing.
August 7, 1985 Rated pressure plateau.
System expansion testing.
HPCI testing.
RCIC testing.
CRD testing.
RBCLCW testing.
Water level testing.
STP-13 testing.
Chemical and radiochemical testing.
Radiation measurements.
LPRM testing.
August 23, 1985 Reactor pressure reduced to 150 psig.
HPCI testing.
RCIC testing.
August 24, 1985 Reactor shutdown after initial heatup.
August 30, 1985 Reactor critical sequence A.
August 31, 1985 Scram #2 on loss of instrument air.
September 3,1985 Reactor critical sequence A. Second heatup to 150*F.
September 4, 1985 Heatup to 250*F - 150 psig.
System expansion testing.
September 6,1985 Reactor pressure 350 psig.
System expansion testing.
September 6, 1985 Reactor scram #3 due to surveillance on level ins t rument .
September 7, 1985 Reactor pressure 600 psig.
System expansion testing.
CRD testing.
September 8, 1985 Unusual event due to Reactor level indication problem.
2.0 Chronology of Events - (continued)
September 10, 1985 Reactor shutdown for repair of reactor level problem.
September 11, 1985 Reactor critical and heatup to investigate level problem.
- Sept ember 12, 1985 Reactor scram #4 on low water level indication. ,
Actual level did not change.
Level instrument problem investigated.
September 18, 1985 Reactor critical sequence A.
September 21, 1985 Reactor pressure 800 psig.
System expansion testing.
CRD testing.
RPV level problem fixed.
September 21, 1985 Rated pressure reached.
System expansion testing.
Second heatup.
RCIC testing.
HPCI testing. ,
MSIV testing.
Water level testing.
CRD testing.
RWCU testing.
Radiation testing.
Chemistry testing.
September 27, 1985 Reactor shutdown.
Hanger placed on B reference leg. Mi scellan-eous maintenance.
October 3,1985 Reactor critical sequence A. Third heatup to rated pressure.
System expansion testing.
October 4,1985 RCIC vessel injection at rated pressure.
System expansict testing.
RWCU system testing.
MSR relief valve testing.
October 6,1985 Initial turbine generator roll to rated speed.
October 8,1985 End of heatup testing.
Begin source replacement outage.
April 18, 1986 Reactor mode switch to startup for generator syne and completion of Test Condition Heatup.
April 24, 1986 Loose Parts Monitor System Test.
Reactor Mode Switch to Shutdown due to condenser leakage problem.
+
I
3.0 Test Results 3.1 Although the plant has been shut down for a majority of the period covered by the supplement, some startup related activities were performed during the period in which the Reactor Mode Switch was in Startup. These activities are discussed below.
3.1.1 A measurable difference in control rod drive normal insert and withdrawal stroke time was observed during previous startup testing for selected rods which were timed at zero reactor pressure, at intermediate pressures, and at rated reactor pressure. Based on an evaluation of the stroke time data for the selected rods , thirty-one rods had their normel withdrawal stroke times adjusted.
3.1.2 As previously reported in Supplement No. 1, the Reactor Water Cleanup (RWCU) bottom head drain dif ferential pressure transmitter was overranged with RWCU System flow routed entirely through the bottom head drain line. The installed transmitter and its corresponding indicator were replaced in April.
Calibration of the transmitter is scheduled for the next testing period.
3.1.3 Loose Parts Monitnring System Startup Test (STP-814)
The section of this test that was completed demon-strates that SRM movement does not af fect the LPMS system or cause a LPMS system alarm.
3.1.3.1 Acceptance Criteria Level 1 1.1 None Level 2 2.1 The LPMS system alert setpoint moni-tor voltmeter functions properly.
2.2 The LPMS acurately detects the simu-lated impacts.
2.3 The automatic alarm and recording functions actuate when the LPMS Impact simulator is used to generate simu-lated impacts.
2.4 The final LPM system sensitivity (from Section 8.8) meets the 0.5 f t-lb im-pact detection criteria from Regula-tory Guide 1.133 in the presence of the measured background noise levels.
~
't 3.1.3.1 Level 2 - (continued) f 2.5 The results of this test must be transmitted to SWEC Engineering within 14 days after final warranty testing.
3.1.3.2 Test Method /Results The first section of STP-814 which checks 1 the LPMS response to SRM movement was con-i ducted April 24, 1986.- This is the first check of the LPMS response to various plant maneuvers to ensure the system will ade-quately detect loose part impacts while minimizing alarms during normal plant maneuve rs.
l The system was first -ri fi ad to be opera-ting correctly, then . ' aensitivi-
, ty was set. The four SRM's w.. ..rtially i withdawn from the core, then fully with-drawn, then partially inserted, and then l fully inserted. No system alarms were re-l ceived during the maneuver.
i.
The LPMS response to simulated impacts (Acceptance Criteria 2.2 and 2.3) was sat-isfactory, and the alert setpoint volt-meter functioned properly (Acceptance Cri-i, teria 2.1)..
3.1.3.3 Corrective Actions /Open Items 4 There are no Test Exception Reports associ-ated with the STP-814 testing completed to date.
The remainder of STP-814 will be performed when various other plant maneuvers are scheduled during the startup test program.
l l
~ -. -. -. -- . - . , .- . . . , _ _ . , _ _ ..-,_ - - _ - . - ,
3.2 A brief discussion of the remaining testing scheduled for Test Condition Heatup foll ws.
3.2.1 Control Rod Drive (STP-5)
Two activities are scheduled for the next startup to complete STP-5 for Test Condition Heatup. One activity is to obtain the individual rod scram time for rod 22-35 at rated pressure and temperature. The other activity is to perform tuning on the Control Rod Drive Hydraulic System flow controller.
3.2.2 Water Level Measurement (STP-9)
The problems with water level indication seen during low power operation were attributed to condensation in the insrument piping between the reactor vessel and the condensing chamber. New piping has replaced the old, with improved pipe slope and shorter piping runs.
All of the narrow and wide range level instruments have been recalibrated. Startup test data will be collected at rated *emperature and pressure to confirm that all the affected instruments meet the criteria for agreement.
3.2.3 Process Computer (STP-13)
Traversing Incore Probe (TIP) tubing underneath the reactor vessel was disassembled to permit control rod drive maintenance. New core limite will be measured and programmed into the process computer during testing scheduled following the next startup.
3.2.4 RCIC (STP-14)
Some controller tuning was performed with the Reactor Core Isolation Cooling (RCIC) system injecting to the reactor vessel during Test Condition Heatup. 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 chesk valves were replaced during the outage. Minimum turbine operating speed at which check valve cycling does not occur will be determined with steam supply pressure to the turbine at 150 psig during the RCIC retes t.
~
3.2.5 HPCI (STP-15)
The High Pressure Coolant Injection (HPCI) system Woodward Governor RGSC (ramp generator and signal convertor) enclosure was relocated as part of the station's environmental qualification modifications.
A " hydraulic bypass" was installed around the HPCI EGR actuator to allow control oil pressure to cycle the control valve fully open then partially closed before the stop valve opens on system starts. This modifica-tion will significantly improve the startup response of the HPCI system. These two modifications require that all HPCI functional and stability demonstration tests be repeated.
A HPCI endurance run is also scheduled for the next retest program. An endurance run was attempted in September and aborted because indicated suppression pool temperatures approached the plant Technical Specifications limiting conditions for operation. The high temperatures resulted from thermal stratification of the suppression pool during HPCI operation. Local to bulk suppression pool temperature differences on
.. the order 'of 15'F were observed. Permanent corrective action is under evaluation. The endurance tun is scheduled with the remaining low power testing, although not required to be completed before the end of Test Condition 3.
3.2.6 Recirculation Flow Control System (STP-29)
The purpose of the test performed in Test Condition Heatup'was to demonstrate that the initial controller settings are stable at low power and core flow. This test demonstrated satisfactory stability and estab-lished a new interim lower limit for opera;ing the recirculation system motor-generator sets (MG's). The new lower limit was established to avoid the lindt cycles which were observed on the 'B' MG. An adminis-trative limit of 24% was established, and the electri-cal low speed limiter has been adjusted to 24%. No additional recirculaton system testing is required before the next regularly scheduled testing at Test Condition 1.
3.2.7 Reactor Water Cleanup System (STP-70)
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 problem with frequent spurious system isolations.
Calibration of the bottom head drain flow indicating loop (STF-70, Appendix A) is scheduled for the upcoming series of plant restart tests.
3.2.8 Residual Heat Removal System (STP-71)
The_ purpose of this test is to demonstrate the ability of the Residual Heat Removal (RHR) system to remove decay and residual heat from the reactor coolant system and from the suppression pool. A suppression pool cooling test is scheduled to be performed in con-junction with the HPCI endurance run. (See Section 3.2.5 for a discussion of suppression pool thermal stratification).
3.2.9 Reactor Building Closed Loop Cooling and Drywell Air Cooling (STP-37)
The purpose of this is to demonstrate that with the maximum design Reactor Building Closed Loop Cooling l Water (RBCLCW) supply temperature, all RBCLCW-supplied l components are adequately cooled and that the Drywell Air Cool,ing (DAC) system maintains drywell tempera-tures within -plant Technical Specification limits.
Because modifications made to insulation and instru- .
mentation in the drywll affect the overall heat load on the system and the temperature distribution seen by the temperature monitoring instruments, all sections
, of this test completed in Test Condition Heatup have been' scheduled to be repeated during the next plant '
heatup.
3.2.10 Service Water 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 neutron source outage. This test is now scheduled for the next startup and low power testing.
l t
4.0 LICENSE CONDITIONS Technical Specifications require that the Startup Report discuss the license conditions which affect plant startup and power escalation testing. The specific license conditions are delineated in paragraph 2.C of the Shoreham Operating License (NPF-36 of July 3, 1985). Each condition is summarized and its status, as it applies to the completed portion of the test program, is provided below.
4.1 Condition
The maximum core thermal power shall not exceed 5% rated cores thermal power.
Status: During the period of this report the reactor was 33'intained suberitical.
4.2 Condition
The plant shall be operated in accordance with the Technical Specifications and the Environmental Protection .
Plan.
Status: The Low Power Test Program has been conducted in accordance with Technical Specifications and the Environmental Protection Plan. In a ve ry few circumstances, the provisions of Technical Specifications were exceeded. Each circumstance has been previously reported to the NRC and has been the subject of a License Event Report (LER).
., 4.3 ;" Condition: The plant shall maintain the fire protection program as described in the Fire Hazards Analysis Report and in the FSAR. ,
Status: The~ provisions of the Fire Protection Program have been adhered to. Any missed fire watches shall be the subject of a Licensee Event Report. During the period covered by this supplement two such LER's were generated.
4.4 Condition
Changes to the initial test program shall be reported in one month.
Status: All changes to the test program as described in Chapter 14 of the FSAR will be reported to the NRC prior to their implementation.
4.5 Condition
The intial inservice inspection shall be developed and implemented before the first refueling outage.
Status: This condition is not affected by the Low Power Test Program. Development of the inservice inspection program is in progress.
4.6 Condition
Control rods shall be tested for boron lo s af ter the first refueling outage.
Status: This condition is not yet applicable.
3 - -
, - - - , - - - - - , , , ,,,- -,y
4.7 Condition
The provisions of tha NUREG-0737 action plan described in the SER, supplements 1 and 4, shall be followed.
4.7.1 Status
To date, the qualifications of six of the required seven backup STA's have been submitted to the Commission, all of which have currently been approved.
LILCO is preparing a letter that will submit the resume of a qualified candidate for the seventh 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 implemented. The requirements of the provision remain under review and shall be implemented prior to the completion of this startup test program.
4.7.3 Status
Modifications to the post accident sample facility, which will enable sampling using the '
modified core damage procedure, are in progress as noted in Section 3.21 of IE Inspection Report 85-038.
This requirement is tracked by Region I as 50-322/85-04-19.
4.7.4 Status
The modifications required to implement the emergency response capabilities, as required by Attachment I to the license, are in progress.
F
[ -- 4.8 Condition: Prict to November. 30, 1985 all electrical equipment shall be qualified.
Status: LILCO sought and was granted an extension of the November 30, 1985 deadline by the Commission. Station modifications 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 refueling outage.
Status: The modifications will be implemented as required.
4.10 Condition: The RHR system may not be operated in the steam condensing mode except under emergency conditions.
Status: The station procedures have been modified to preclude the stees condensing mode of operation except as a last rescrt when all other methods of core and containment cooling have failed.
4.11 Condition: Two containment isolation barriers in series will be installed by the end of the first refuel outage.
Status: Methods to satisfy this condition will be completed by the end of the first refuel outage.
_ - - _ _ = _ _ -
4.12 Condition: The provisions of Appendix 3 to the license shall be satisfied as they apply to the TDI diesel generators.
Status: 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: The requirements of this condition have been met.
4.14 Condition: a.) Prior to exceeding five percent powe r ,
radiation monitoring panels ID11*PNL-ll7A, B and radiation monitoring pumps 1Dll*P126, 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 invessel storage racks have been administrative 1y prohibited frog use.
4.15 Condition: The plant shall have on-shif t advisors as required by Attachment 2 of the license.
Status: The plant currently has sufficient numbers of qualified on-shif t advisors to satisfy this condition.
4.16 Condition: The Emergency Core Cooling Systems performance shall be reanalyzed for the second cycle and beyond, utilizing models that account for burnup gas pressure and local oxidation and which are approved by the NRC.
Status: This condition will be completed by the required date.
4.17 Condition: The licensee shall implement the response to Generic Letter 83-28 on schedule.
Status: All provisions stipulated in SNRC-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-1184.