SNRC-1357, Suppl 7 to Startup Rept for 870501-0801
| ML20237H661 | |
| Person / Time | |
|---|---|
| Site: | Shoreham File:Long Island Lighting Company icon.png |
| Issue date: | 08/11/1987 |
| From: | Leonard J, Scalice J, Steiger W LONG ISLAND LIGHTING CO. |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| References | |
| SNRC-1357, NUDOCS 8708170304 | |
| Download: ML20237H661 (17) | |
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l LONG ISLAND LIGHTING COMPANY Shoreham Nuclear Power Station Supplement No. 7 to the STARTUP REPORT
'for the period May 1, 1987 to August 1, 1987 Approved: .
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. /Bcalice / Date As stant Plant Manager Approved: P* // -f 7 W. E. - Steiget ' V Date Plant Manager g26 &
B708170304 PDR p
ADOCK870011 05000322, PM ,
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TABLE OF CONTENTS 1.0 Overview 2.0 Chronology'of Events 3.0 Startup Activities Performed / Test Results 4.0 License Conditions
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1 1.0 OVERVIEW This Startup Report Supplement has been written by the Long
[ Island Lighting Company (LILCO) for submittal to the l
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.
1 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 initini criticality on February 15, 1985, and LILCO submitted a Startup Report on November 15, 1985. The scope of that Startup Report included fuel load and initial criticality under the .001% power license, NPF-19, and low power testing conducted under the 5% power license, NPF-36.
Since SNPS has not completed its startup test program or commenced commercial power operation, additional Supple-ments are required on a periodic basis. This Supplement covers the period from May 1, 1987 to August 1, 1987.
Shoreham was shutdown from September 1986 to May 1987.
Major modifications completed during this reporting period are the Main Steam Isolation Valves (MSIVs) setpoint change from Level 2 (-38") to Level 1 (-132") reactor water level, and the standby liquid control system compliance (10 CFR 50.62) modification, which will enhance the reactor shut-down capabilities in the event of an anticipated transient without scram. In addition, we have also completed modifications to the Alternate Rod Insertion (ARI) system.
These modifications bring the ARI system into compliance with 10 CFR 50.62 (c) (3) requirements. For a description of these modifications, see LILCO letters; SNRC-1338, dated May 26, 1987, and SNRC-1344, dated June 19, 1987. The plant was started up for training criticals and 5%
operational testing on May 22, 1987. The generator was synchronized on June 4, and the plant was shut down on June j 6, 1987.
Other modifications in progress or completed during this period include work on the drywell vacuum breakers, the main circulating water systems and portions of the Colt diesel tie-in modification.
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l 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.
1 January 19, 1985 Fuel loading completed.
January 25, 1985 CRD open vessel testing completed.
February 17, 1985 Initial critical and shutdown margin test. Completed open vessel l testing. ]
l June 4, 1985 CRD open vessel retest completed. )
l 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.
July 18, 1985 Reactor Shutdown for RPV lever instrumentation work. 1 July 23, 1985 Reactor critical.
July 26, 1985 Reactor shutdown for RPV level instrumentation work.
July 29, 1985 Reactor critical.
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I 2.0 CHRONOLOGY OF EVENTS - (continued)
July 31, 1985 Reactor pressure to 150 psig.
HPCI system testing.
August 1, 1985 350 psig plateau.
Drywell radiation survey.
System expansion drywell inspection, system expansion data.
August 3, 1985 600 psig plateau.
System expansion.
RBCLCW performance.
CRD testing.
August 5, 1985 800 psig plateau.
System expansion.
CRD testing.
August 7, 1985 Rated pressure plateau.
System expansion testing.
CRD testing.
RBCLCW testing.
Water level testing.
Process computer 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 keactor critical sequence A.
August 31, 1985 Scram #2 on loss of instrument air.
September 3, 1985 Reactor critigal sequence A. Second heatup to 150 F.
September 4, 1985 Heatup to 250 - 150 psig.
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 instrument.
September 7, 1985 Reactor pressure 600 psig.
System expansion testing.
CRD testing.
i 2.0 CHRONOLOGY OF EVENTS - (continued)
September 8, 1985 Unusual event due to reactor level indication problem.
September 10, 1985 Reactor shutdown for repair of i reactor problem. l I
September 11, 1985 Reactor critical and heatup to 4 investigate level problem.
September 12, 1985 Reactor scram #4 on low water level indication. Actual level did not l' 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. l 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 3eg.
Miscellaneous maintenance.
October 3, 1985 Reactor critical sequence A. Third heatup to rated pressure.
System expansion testing.
August 13, 1986 Reactor water level measurements, rated plateau.
System expansion data, reactor building secondary walkdowns, rated plateau.
Radiochemistry water chemistry sampling data, rated plateau.
2.0 CHRONOLOGY OF EVENTS - (continued)
August 14, 1986 Scram timing of sequence A rods.
System expansion balance of plant walkdown: steam tunnel and turbine building.
Reactor cooldown for maintenance work in the drywell and sequence exchange.
August 15, 1986 Commenced reactor heatup in the B-2 sequence, reactor' critical.
Drywell entry a? rated conditions for system expansion data.
Feedwater low flow controller testing.
Scram timing of sequence B rods.
August 16, 1986 Commenced main turbine startup prerequisites.
HPCI tuning at rated conditions (CST to CST).
RBCLCW performance test at rated pressure plateau.
August 17, 1986 Main turbine roll.
Reactor building service water performance test.
August 18, 1986 RWCU blowdown and normal mode tests.
Reactor shutdown due to hurricane warnings.
August 20, 1986 Reactor startup - heatup to rated pressure plateau.
System expansion balance of plant walkdown at rated. pressure plateau.
August 21, 1986 RCIC tuning to the reactor vessel at rated pressure plateau.
RWCU bottom head drain flow calibration.
'2. 0 - CHRONOLOGY OF . EVENTS -- (continued)
August 22, 1986 CRD flow controller. tune-up.
RCIC instrumentation piping vibration test.
August 23,.1986 HPCI tuning'at rated (CST.to CST).
Reactor cooldown to 150 psig
, plateau.
Feedwater low flow controller tuning in auto during cooldown.
August 24, 1986 RCIC Vessel Injection and. Stability Demonstration at 150 psig.
Heatup to rated pressure.
Feedwater low flow controller in manual.during heatup.
HPCI tuning at rated conditions (CST-to CST).
August 25, 1986- Main turbine roll and excitation checks.
August- 26, 1986 Main turbine generator synchronization to the. grid and 24 hr. run.
August 27, 1986 Turbine generator off grid.
RCIC cold quickstart #1 (vessel injection).
RCIC transient vibration test.
Process computer TIP hot alignment.
August 28, 1986- HPCI' tuning at rated pressure plateau.
Reactor cooldownlto 150 psig plateau.
HPCI quickstart (CST to CST) and stability demonstration at 150 psig.
Reactor heatup to rated pressure plateau.
HPCI hot alignment verification.
August-29, 1986 CRD flow controller tuning.
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2.0 CHRONOLOGY OF EVENTS - (continued)
August 30, 1986 RCIC cold quickstart #2 (vessel l injection).
RCIC transient vibration test.
Loose parts monitoring data collection during RCIC vessel injection.
Reactor cooldown and shutdown.
May 22, 1987 Reactor critical.
May 26, 1987 ADS valve 18-month surveillance l May 27, 1987 APRM calibration.
i May 29, 1987 RCIC rate flow surveillance.
HPCI rated flow surveillance and controller cutput limiter .
adjustment.
May 31, 1987 RCIC functional and stability demonstration from Remote Shutdown Panel.
June 1, 1987 Power reduction due to system load demand.
June 4, 1987 Generator synchronized for 8-hour run.
June 5, 1987 HPCI endurance run.
June 6, 1987 Reactor shutdown and cooldown.
3.0 STARTUP ACTIVITIES PERFORMED / TEST RESULTS This section discusses those startup tests identified in Chapter 14 of the USAR (Table 14.1.1-1) which were performed during 5% plant operations between May 22 and June 6, 1987.
3.1 HPCI (STP-15)
A HPCI endurance run was completed on June 5, 1987.
The purpose of the endurance'run is to demonstrate that the turbine and pump lubricating and control oil temperature stabilize after extended system opera-tion, and thus, that the system is capable of operating for extended periods. This test is ,
performed with the HPCI system operating in the test I mode, from CST to CST.
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] i An endurance run'was initially attempted in Sep- 1 tember, 1985. The test was stopped when the suppres- !
sion pool temperature began approaching the Limiting j L Condition for Operation specified in section l 3.6.2.1.a.2.a. The Nuclear Engineering Department )
evaluated the test data and determined that thermal '
stratification occurs in the suppression pool during HPCI operation and that the arithmetic average of the installed suppression pool temperature elements does not accurately represent average bulk pool tempern-ture. PATP staff reviewed the station startup tect i procedure (STP-15) against the GE startup test specification. The STP required that an endurance ?
run be initiated from a cold quickstart; the GE test specification requires that an endurance run be initiated from any automatic initiation, i.e., from either a " cold" or a " hot" quickstart. STP-15 was revised to allow initiating the endurance run from any quickstart.
During 5% power operation between May 26 and June 5, 1987, the HPCI system was operated and tested under station operating and surveillance procedurec, and startup test procedure STP-15.
The HPCI operability and rated flew surveillance test was performed on May 29, 1987. The required system flow of 4,250 gpm was not achieved with the pump dis-charge pressure established at the required 1,203 psi above suction pressure. The control room controller output limiter war adjusted upward. This limiter, or clamp, does not affect the dynamic response of the control system, but could affect the margin to overspeed. The surveillance test was successfully repeated after the clamp was adjusted.
On June 3, 1987, HPCI was started manually under STP-15, CST-to-CST, to establish the required throttle position for the test return valve in order to achieve a pump discharge pressure 100 psi greater than reactor pressure. With the system operating in the test mode, the flow controller was adjusted up and down, in both manual and automatic control, at rated flow and at low flow (about 2,200 gpm), in order to demonstrate controller stability. Decay ratios of 0.58 to 0.76 were observed for flow changes at low flow with the controller in automatic, failing the Level 2 acceptance criterion for decay ratio.
The test return valve throttle position was estab-lished at rated flow, and the valve was not re-posi-tioned for the low flow step changes. This appears to have reduced system stability somewhat at low flow. The response of the HPCI control system at rated fl:ow is similar to the response observed following system tuning in 1986, and the decay ratios at rated flow meet the acceptance criterion.
Before.the system was shutdown for an automatic
' initiation and endurance run, dynamic tests were performed on the test return throttle valve (MOV-037) and the' test-return isolation valve (MOV-038).- A description of these dynamic tests is given in NRC inspection report 50-322/87-13.
Following a review of-the test return valve design i specificati~nso and consultations with the valve vendor, the test return throttle valve' operator's
. motor controller was modified, and.HPCI testing was rescheduled. Long-term corrective actions.to address the root cause:of test return valve problems are still under engineering evaluation.
On June 5, the HPCI system was manually started to j establish the. desired throttle valve position and to dynamically test the; test return 1 valves. Both valves ,
were commanded to close simultaneously with the j system at rated flow and the throttle valve in its throttled position. Both valves' closed and the system was shutdown to cool down the suppression pool. The system was'then automatically initiated under STP-15 for an endurance run. HPCI oil-tempera-tures stabilized after about 40 minutes into the run.
The system was shutdown about 50 minutes.into the run,whenindgcatedsuppressionpooltemperature reached 100.5 F. The test data are still under review; but the preliminary conclusion is that the endurance run was successful.
HPCI. test results have been evaluated against the acceptance criteria as follows:
- 1. The average HPCI pump discharge flow shall be equal to or greater than 4250 gpm after 25 seconds have elapsed from initiation on cold quickstarts at any reactor pressure between 150 psig and rated.
The ability to achieve 4,250 gpm was successfully demonstrated, but the June 5 quickstart initia-tion was not recorded, so the time to rated flow could not be evaluated. The time to rated flow has been successfully demonstrated during previous testing, so no action is required.
- 2. The.HPCI turbine shall not trip or isolate during auto or manual starts.
I The turbine did not trip or isolate on any occasion.
- 3. The turbine gland seal condenser system shall be capable of preventing steam leakage to the atmosphere in excess of allowable releases. For
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the purposes of this test, less than or equal to 0.1 times the maximum permissible concentration (MPC) of airborne activity in the vicinity of the HPCI unit is considered allowable releases.
Airborne samples showed activities less than 0.1 times MPC. One test exception report (TER) was 1 issued for a missed sample during the endurance run. Another airborne sample taken during the manual start and run prior to the endurance run i showed airborne less than 0.1 times MPC, so no I further action is required.
- 4. The speed and flow control loops shall be adjusted so that the decay ratio of any HPCI system variable is less than 0.25 for small speed i or flow command changes in either manual or (
automatic mode. l For all tests at rated flow and rated pressure, decay ratios were less than 0.25. Decay ratios of 0.58 to 0.76 were observed for flow changes at low flow with the controller in automatic. The ,
test return valve throttle' position was established at rated flow, and the valve was not re-positioned for the low flow step changes.
This appears to have reduced system stability somewhat at low flow. The response of the HPCI control system at rated flow is similar to the response observed following system tuning in 1986, and no controller adjustments affecting the dynamic response of the controller have been made since that time. Since the rated flow response is consistent with the response recorded in 1986, and since controller response was satisfactory at that time, no further action is required.
- 5. In order to provide an overspeed and isolation trip avoidance margin, the transient start first peak shall not come closer than 15% (of rated speed) to the overspeed trip, and subsequent speed peaks shall not be greater than 5% above rated speed.
For automatic initiations, the first speed peak should not exceed 4,400 rpm and the second speed peak should not exceed 4,200 rpm. The quickstart on June 5 showed a first speed peak of 2,180 rpm and a second speed peak of 3,940 rpm.
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4.0 LICENSE CONDITIONS j 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 core thermal power.
Status: During the period covered by this Supplement, the reactor did not exceed 5%
rated core thermal power.
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. Any deviation from these documents is required to be reported to the NRC and be 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 plant has adhered to the provisions of the Fire Protection Program. Any non-compliance or missed fire watches shall be the subject of a Licensee Event Report.
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 USAR will be reported per the license condition.
4.5 Condition
The initial 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.
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- 4. 6- Condition: Within thirty (30) days after planti startup following'the first refueling y outage, the licensee shall comply with
! items 1, 2, and 3 of I.E. Bulletin No.
79-26, Revision 1, " Boron Lots From BWR Control Blades", and submit a written response on item.3.
Status: This condition is not yet applicable.-
4.7 Condition
The provisions of the NUREG-0737 action plan described in the SER, Supplement 1
.and 4, shall be followed.
4.7.1 Status
The qualifications of the required seven backup STA's have been submitted to the Commission, all have been approved.
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 thr. provision remain under review and shall be11 mole- ~
mented prior to the completion of j the 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 85-030. This requirement is being tracked by Region I as 50-322/85-04-19.
4.7.4 Status
The modifications required to imple-ment the emergency response capa-bilities, as require by Attachment 1 !
to the license, are in progress.
4.8 Condition
Prior to November 30, 1985, all electrical equipment shall be qualified.
Status: Station modifications to environmentally qualify the required electrical equipment are 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.
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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 steam condensing mode of operation except as a last resort 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 refueling outage.
Status: Modifications to satisfy this condition will be completed by the end of the first refueling outage.
4.12 Condition: The provisions of Attachment 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.
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 power, radiation monitoring panels 1D11*PNL-117A, B and radiation monitoring pumps 1D11*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 administratively prohibited from use.
4.15 condition: The plant shall have on-shift advisors as required by Attachment 2 of the license.
Status: The plant currently has sufficient numbers of qualified on-shift advisors to satisfy this condition.
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- A i o4.16 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.
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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. In a Ju?.y'15, 1987 NRC (A. Thadani) letter to the BWROG ,
f T. Fickens) , the NRC approved the' UWROG I methodology in an attached safety evalu-ation report. Therefore, the LILCO ]
I review is in progress.
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! ggE,cf j LONG ISLAND LIGHTING COM PANY SHOREHAM NUCLEAR POWER STATION lsemammuwame md P.O. DOX G* 9, NORTH COUNTRY RO AD + W ADING RIVER, N.Y.11792 l
I JOHN D. LEONARD, JR.
VICE PRESIDENT . NUCLE AR OPE R ATIO'4S AUG 131987 SNRC-1357 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555 Supplement No. 7 to the Startup Test Report Shoreham Nuclear Power Station Docket No. 50-322 Gentlemen:
Attached is Supplement No. 7 to the Startup Test Report, prepared by the Long Island Lighting Company in compliance with Shoreham Nuclear Power Station's Technical Specifications paragraphs 6.9.1.1 through 6.9.1.3 and Regulatory Guide 1.16, Revision 4, Section C.I.a.
This report covers the period from May 1, 1987 to August 1, 1987.
It includes a description of the measured operating condition values or characteristics obtained during this test period, 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, dskrwf J. D. Leonard, Jr.
President - N'a ar Operations TD:ck Attachment h cc: R. Lo W. Russell I {
C. Warren
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