ML20137W619
| ML20137W619 | |
| Person / Time | |
|---|---|
| Site: | Brunswick, 05000000 |
| Issue date: | 11/30/1977 |
| From: | Poppel I GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20137W593 | List: |
| References | |
| 77NED246, NEDO-24562, NUDOCS 8512100287 | |
| Download: ML20137W619 (74) | |
Text
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NEDO-24562 77NED246 CLASSI NOVEMBER 1977 BRUNSWICK UNIT 1 STARTUP TEST RESULTS FINAL
SUMMARY
REPORT i
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NEDO-24562 Qassl l November 1977 BRUNSWICK UNIT 1 STARTUP TEST RESULTS
- FINAL
SUMMARY
REPORT
- l. D. Poppel 1 '-
Approved: - - - ' ' ~
R. C. Christianson, Manager Plant Startup and Test NUCLEAR ENERGY PROJECTS DIVISION e GENERAL ELECTRIC COMPANY SAN JOSE, CALIFORNIA s$125 G EN ER AL $ ELECTR'IC 8
DISCLAIMER OF RESPONSIBILITY This document was prepared by or for the General Electric Company. Neither the General Electric Company nor any of the contributors to this document:
A. Makes any warranty or representation, express or implied, with respect to the accuracy, completeness, or usefulness of the information containedin this docu-ment, or that the use of anyinforma00n disclosedin this documentmaynotinfringe privately owned rights; or B. AssumesanyresponsibilityforliebrHtyordamageofanykindwhichmayresultfrom the use of anyinforma00n disclosedin this document.
NEDO.24562 TABLE OF CONTENTS Page ABSTRACT. . - .. - . _ - _ 1x
- 1. INTRODUCTION .. .. .. .
. 11 1.1 Purpose . ... ., . . . . 1 1.2 olant Description 11 1.3 Startup Test Prograrn.. . .. ..1 1 1.4 Statap Test Description 11 1.5 Startup Test Acceptance Criteria. . .. ..t 2
- 2. SilMMARY OF THE TEST PROGRAM . . .. . 21 2.1 Chronology of Startup Testing- .. 2-1 2.2 Test Completion Dates for Startup Tests.. . 2-1 2.3 Power / Flow Map with Startup Test Condtions - 21
- 3.
SUMMARY
OF TEST RESULTS. .. . . - _ . 31 3.1 STI-1 Chemical and Radiochemical 1 3.2 STI-2 Radation Measurements.. . . . . 3-1 3.3 STl-3 Fuel Loading. _ .3-1 3.4 STl-4 Full Core Shutdown Margin.. . 3-2 3.5 STI-5 Control Rod Drive System.. . 3-2 3.6 STi-6 SRM Performance and Control Rod Sequence. 3-4 3.7 STi-7 Reactor Water Cleanup System. 3-7 3.8 STI-8 Residual Heat Removal System.. . .. . 3-7 3.9 STI-9 Water Level f 4easurement.. 3-8 3.10 STl-10 IRM Performance.. .3-8 3.11 STI-11 LPRM Calibration - 8 3.12 STI.12 APRM Calbration.. --3-9 3.13 STI-13 Process Computer 3-9 3.14 STl-14 RCIC System . 3-11 3.15 STl15 HPCI System.. . .3-12 3.16 STI-16 Selected Process Temperatures.. . --3 13 3.17 STI.17 System Expansion. 3-14 3.18 STl-18 Core Power Distribution.. .3-16 3.19 STI-19 Core Performance = . ;3-16 3.20 STi-20 Steam Production.. ...
17 3.21 STI-21 Flux Rasponse to Rods. 3-18 3.22 STi-22 Pressure Regulator.. 3-19 3.23 ST1-23 Feedwater System.. . . . . . . . . . . 3-19 3.24 STI-24 Turbine Valve Surveillance. . . . . .3-29 3.25 STl-25 Main Steam Isolation Valves. 3-31 3.26 STI-26 Relief Valves.. ..- . . . . . . - . . 3-33 3.27 STi-27 Turbine Tnp and Generator Load Rejection.. . . . . ;.3-36 3.28 STl-28 Shutdown From Outside the Control Room. . 43 3.29 STi-29 Flow Control.. . . . 3 43 3.30 STI.30 Recirculation System . . .. . 3-44 3.31 STi-31 Loss of Turbine Generator and Offsite Power-- -.3-47
.iii-L
i NEDO-24662 TABLE OF CONTENTS (Continued)
Page 3.32 STI-32 Recirculation M-G Set Speed Control: .3-48 3.33 STI.33 Main Turbine Stop Valve Surveillance Test--- ;3-49 3.34 STI.34 Vibration Measurements... ;3-49 3.35 STI-35 Recirculation System Flow Calbration.- 50
- 4. DISTIUSUTION . . -. . - .
. _ 4-1
.iv-
NEDO-24562 t l
LIST OF ILLUSTRATIONS Figure Title Page 2.1.1 Histogram of the Startup Test Program - November..
..2-6 2.1.2 Histogram of the Startup Test Program - December... ...
.2-7 2.1.3 Histogram of the Startup Test Program - January.- .2-8 2.1.4 Histogram of the Startup Test Program - February.. - . ~ . . ;2-9 2.1.5 Histogram of the Startup Test Program - March.. .
.2-10 2.1.6 Histogram of the Startup Test Program - April.
_2-11 2.3.1 Test Program Power / Flow Map Startup Test Conditions. -. .2-12 3.6.2.1 A Control Rod Sequence.- . . . . . .. . . .
.3-5 -
3.6.2.2 B Control Rod Sequenco.; . .3-6 3.12.3.1 APRM Tracking;;
.3-10 3.22.3.2.6 Steam Flow versus Pressure Regulator Output.. .3-21 3.22.3.2.7 Control Valve No.1 Programming.. -.-. ..
.3-22 3.22.3.2.8 Control Valve No. 2 Programming.. 3-23 3.22.3.2.9 Control Valve No. 3 Programming.. .3-24 4 3.22.3.2.10 Control Valve No. 4 Programming... . -3 25 3.23.3.2.1 CPR and Reactor Power versus Time - Loss of Feedwater Heating Transient.- .3-30
.v./.vi-
NEDO-20562 UST OF TABLES Table Title Page 2.1.1 Significant Dates of the Startup Test Program. .2-1 2.2.1 Startup Test Completion Dates and Test Report Numbers.
.2-2 3.5.1.1 Average Cold Scram Times - ;3-2 3.5.1.2 Average Rated Scram Times .
.3-2 3.5.3.1 Selected CRD Scram Times... 3-3 3.5.3.2 Zero Accumulator Pressure.. .. .-
.3-4 3.5.3.3 Normal Accumulator Pressure..
4 3.12.3.1 APRM Tracking- ..
.3-9 3.14.3.1 RCIC System Data.. .
.3-12 3.15.3.1 HPCI System Deta.-
. .-. 3-13 3.16.3.1 Selected Process Terrperatures.. . ;3-14 3.17.3.1 System Expansion Vibration Testing.,
3-15 3.18.3.1 TIP Reproducibilty Data.- .
.3-16 3.19.2.1 Summary of Core Performance Parameters..
-3 17 3.22.3.1 Final Pressure Regulator Settings-- .. .
.3-20 3.22.3.2 Pressure Regulator Backup Performance.. . . . .3-20 3.23.3.1 Level Set Point Change Testing Data- .
.3-27 3.23.3.2 Thermal Parameters During Loss of Feedwater Heating Test. ..
.3-28 3.23.3.3 Transient Thermal Parameters - Loss of Feedwater Heating Test.. .3-29 3.25.3.1 MSIV Closing Times. . .
- 3-32 3.25.3.2 MSIV Full Closure Sensitivity Study.. . . . . . . .. . . . 3-33 3.25.3.3 MSIV Full Closure Sequence of Events.. .. . . . . . .. . . 3-34
. 3.26.3.1 Relef Valve Capacities.. .. ..
. . 3-36 3.27.3.1 Turtune Trip Sensitivity Study.. . . . . . . ....... . . . . . . . . . .3-38 3.27.3.2 Turbine Trip of February 2,1977 Sequence of Events..... . . . . . .3-39
.vii.
NEDO 24562 LIST OF TABLES (Continued)
Table Title Page 3.27.3.3 Turtine Trip of February 3,1977 Sequence of Events.- 40 3.27.3.4 Load Rejection of February 28,1977 Sequence of Events-- 3-41 3.27.3.5 Generator Load Reject Sensitivity Study.. ..
. .. 3-42 3.29.3.1 Rodrculation System Stop Flow Change Testing 3-45 3.29.3.2 Large Flow Ramp Testing.- - .3-46 3.30.3.1 Recirculation Pump Trip Data.. ~
3-47 3.35.2.1 Drive Row / Core Flow Correlation = . 3-51 i
I vili-
NEDO-24562 ABSTRACT This reportconsists of a summery of the startup test programperformed at Unit 1 of the Brunssck Steam Electric Pfar;t. It ircludes results of static and dynamic reactor performance tests of the reactor and related systems within the General Electric scope of supply.
4 d
in / n.
- 1. INTRODUOTION 1.1 PURPOSE The purpose of this report is to present a brief summary of the results of al of tt's General Electric related startup tests performed at Unit 1 of the Brunswick Steam Electric Plant. The startup test program included tests of the static and dynamic performance of the reactor and related systems.
1.2 PLANT DESCRIPTION The Brunswick Steam Electnc Plant Urut 1 reactor is a single cycle boiling water reactor deogned by General Electric for the Carolina Power and Ught Company. The plant is located near Southport, North CaroEna near where the Cape Fear River enters the Atlantic Ocean. It is warranted for 2436 MWt (at which power the generator is warranted for 849 MWe).
1.3 STARTUP TEST PROGRAM The startup test program began with fuel loading on September 15,1976 and finished with the MSIV full closure on September 30,1977. The plant testing consisted of the following successive phases:
Phase 1: Pre-operational testing (not covered in this report)
Phase 2: Fuel loading and open vessel testing Phase 3: Irutial hcatup to rated temperature and pressure Phase 4: Power tests Phase 5: Warranty tests During Phase 4 testing the plant was taken to its designed full power operating cordtion in a safe controlled fashion.
Extensive testing was performed at previously specified operating cordtions to demonstrate the safe and efficient perfor-mance of plant components. The Phase 5 warranty test began with the begnning of the warranty run and successfully concluded 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> later; at Brunswick Unit 1 the warranty run was conducted before the Phase 4 power tests were completed.
1.4 STARTUP TEST DESCRIPTION Documents such as the operating Ecense, technical spoofications, plant operating procedures, and equipment manuals control reactor operations during the startup test program. Two documents are suppled by GE-NEBG for implementation of the startup testing of the equ.oment it suppies: The Startup Test Specification (22A2212) and The Startup Test Instruction (22A2229AC).
The Startup Test Specification is a document issued for review and approval by GE management and is used for planning and scheduling tests. The chosen tests are required either to demonstrate that it is safe to proceed, to demonstrate performance, or to obtain engineering data. This document defines the minimum test program needed for a safe, effk:ient startup. The purpose, description, and criteria are given for each test together with a doecription of each test condition.
The Startup Test instruction is a document wntten for use in the control room by quaufied GE personnel and for trained customer personnel to property perform and evaluate each startup test. These instructions may be expanded to include additional testing, or they may be reworded to address plant operating procedures, sutiect to the review and written approval of designated GE and customer personnel.
11
1 NEDO-24562 1.5 STARTUP TEST ACCEPTANCE CRITERIA The Startup Test Spoofication and The Startup Test instruction contain enteria for acceptance of results of that test.
There are two levels of criteria idennfied, where applicable, as Level 1 and Level 2.
The Level I criteria include the values of process variables assigned in the des #gn of the plant and equipment. If a Level I criterion is not satisfied, the plant is placed in a satisfactory HOLD condibon unti a resolution is made. Tests compatible with this HOLD condition may be contnued. Following resolution, app 6 cable tests must be repeated to verify that requirements of the Level 1 criterion are satisfied.
The Level 2 criteria are associated with expectations in regard to performance of the system. If a Level 2 cnterion is not sabsfied, operating and testing plans would not necessarily be anored. Investigabons of the measurements and of the analytical techniques used for the predctions would be started.
Safety bmits, as set forth in plant technical specifications, are riot included because there are no planned operations of teshng at such levels. By meseng the enteria, startup test results demonstrate agreement with design spoofications and predictions.
l l
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1 12
- 2.
SUMMARY
OF THE TEST PROGRAM 2.1 CHRONOLOGY OF STARTUP TESTING Some dates that were sigruficant during the test program are listed in Table 2.1.1. The completion dates of testing and the startup test report numbers at the specified test condtions are shown in Table 2.2.1. Figures 2.1.1 through 2.1.6 deplay a histogram (power / pressure versus time) of the test program bepnning with reactor startup in November 1976.
2.2 TEST COMPLETION DATES FOR STARTUP TESTS Refer to Table 2.2.1 for the test completion dates for the complete startup test program for Brunswick Unit 1.
2.3 POWER / FLOW MAP WITH STARTUP TEST CONDITIONS Refer to Figure 2.3.1 for a plot of the power and flow condtions which specify each test condtion.
Table 2.1.1 SIGNIFICANT DATES OF THE STARTUP TEST PROGRAM Began Fuel Loadng September 15,1976 Fuel Loadng Complete September 25,1976 initial Criticality Shutdown Margin Demonstration October 6,1976 First Heatup November 17,1976 First Turbine Roll December 2,1976 First Generator Synchronization December 4,1976 Nominal Finish of Heatup Testmg STARTREC Calibration Complete December 16,1976 Nominal Finish of Test Cond'Jon 1 Testing December 29,1976 Nominal Finish of Test Condition 2 Testing January 6,1977 Finish of Test Condtion 7 Testing January 31,1977 Nominal Finish of Test Condtion 3 Testing March 1,1977 Finish of Test Condtion 4 Testing April 6,1977 Began Warranty Run April 16,1977 Finished Warranty Run Apnl20,1977 Breakdown of Main Generator April 27,1977 Generator Outage Ends June 26,1977 Resumption of Startup Testing July 6,1977 100% Power First Reached After Outage August 20,1977 First Brunswick Rod Sequence Exchange at Power August 27,1977 MSIV Full Closure End of Test Condtion 6 Testing End of Brunswick Unit 1 Test Program September 30,1977 21
I NEDO-24562 Table 2.2.1 STARTUP TEST COMPLETION DATES AND TEST REPORT NUMBERS Open STI Vessel Hootup TCt TC2 TC3 TC4 TC5 TC6 TC7 Chemical and _
12/15/76 _ _1/24/77 1/31/77 -
3/3/77 3/9/77 -
~ ~ ~ ~
Radochemical 1-1 12 1-3 l 3/12/77 14 Radiation 9/7/76 12/2/76 12/16/76 -
1/12/77 - -
3/9/77 Measurement 2-1 ,
9/29/76 2-3 2-4 2-5 2-6 22 Fuel 9/15/76 '
to.ang l - _ - _ - ._ _ _
9/25/76 3-1 Shutdown 10/8/76 - - - - - - - -
Margn 41 CRD System 9/14/76 11/26/76 12/15/76 -
1/12/77 - - - -
9/29/76 12/13/76 12/3C/76 4/6/77 5-1 5-2 5-3 5-4 SRM-CRD 10/7/76 12/28/76 Sequence *l - - - - - - - - -
10/8/76 61 12/2/76 RWCU -
l - - - - - - -
System 12/3/76 71 RHR System - - -
1/3/77 - - - 4/28/77 -
81 82 Water Level - 12/12/76 12/16/76_ - -
4/8/77 -
3/9/77 1/31/77
~ -
9-1 9-2 94 9-3 IRM -
10/8/76 - - - - - - -
11/17/76 12/30/76 10-1 LPRM - - 12/21/76 -
1/20/77 - -
3/9/77 -
11 1 11 2 11 3 2-2
NEDO-24562 Table 2.2.1 STARTUP TEST COMPLETION DATES AND TEST REPORT NUMBERS (Continued)
Open STI Vessel Heatup TCt TC2 TC3 TC4 TC5 TC6 TC7 APRM - 11/18/76 12/21/76 1/2/77 1/27/77 -
2/23/77 3/23/77 -
12-1 12-2 12-3 12-4 12-5 12-6 4/17/77 12 7 Process -
11/17/76 12/1/76 1/26/77 - - -
3/9/77 -
Computer l 13-1 12/24/76 13-3 13 4 13-2 RCIC -
11/25/76 12/24/76 1/1/77 - - - - -
12/30/76 l 14 1 14 2 14-3 HPCI -
11/23/76 -
4/8/77 4/9/77 - - - -
15-1 15-2 15-3 Selected Process -
12/30/76 - -
1/13/77 - -
4/23/77 Temperatures l -
16-1 2/1/77 4/24/77 16-2 16-3 System - 12/1/76 - - - - -
1/2/77 Expansion l - -
12/2/76 4/24/77 17 1 17 2 Power - - - 1/3/77 1/25/77 - - 3/9/77 -
Dstnbution 18 1 18-2 18 3 18-4 Core - - 12/23/76 1/3/77 1/20/77 4/8/77 3/3/77 3/9/77 -
Performance 19-1 .92 19-3 19-6 19-4 19 5 Steam - - - - - - - 4/16/77 Production I -
4/20/77 20 1 Flux Response - - 12/15/76 1/1/77 - 4/8/77 - - -
21 1 21 2 21 J Pressure - -
~
1_2/15/76 1/3/77
~
3/1/77 4/8/77 3/1/77 3/18//7 -
Regulator 22 1 22 2 22 3 l 4/22/77 3/3/77 22 5 22 4 23
NEDO-34562 Table 2.2.1 STARTUP TEST COMPLETION DATES AND TEST REPORT NUMBERS (Continued) l Open STI Vessel Hestup TC1 TC2 TC3 TC4 TC5 TCG TC7 Feedwater System Setpoint Change - - 12/16/76 1/29/77_ - 3/2/77 3/17/77_ _
~ 23-2 '
23-1 23-3 7/8/77 7/9/77 7/9/77 4/24/77 8/26/77 23-7 23 7 23-7 23-5 23 7 Pump Trip - - - - - - -
3/12/77 -
23-4 Loss of - - - - - - -
3/13/77 -
Feedwater 23-6 Heating Turbine Valve Surveillance Bypass - - 12/16/76 1/3/77 1/13/77 - 3/2/77 3/10/77 -
Valves 24 1 24 2 24-4 24 5 24-6 Control - - 12/17/76 1/3/77_ - -
3/3/77 4/19/77; -
~ ~
Valves 24-3 24 7 MSIV - 12/2/76 12/16/76 1/2/77 1/13/77 -
3/3/77 _
~ ~ ~
Each 25 1 25 2 25-3 25 4 Valve Full - - - - - - -
9/30/77 -
Isolation 25 5 Relief - 11/24/76 12/24/76 1/27/77 - - - - -
Vaives 26 1 26-2 Load - - - 1/4/77 - - - 2/28/77 -
Rejection 27 1 27-3
- Turbine - - - - 2/3/77 - - - -
Tnp 2/21/77 27 2 l Shutdown Outside- - 12/28/76 - - - -
The Control Room 28 1 l 1/31/77 2/28/77 3/12/77 4/23/77 Flow Control - -
29 1 29 2 l 24 i
s
NEDO-24562 Table 2.2.1 STARTUP TEST COMPLETION DATES AND TEST REPORT NUGABERS (Continued)
Open STI Vessel Hestup TC1 TC2 TC3 TC4 TC8 TCG TC7 Recirc System Motor Breaker Trip - - - -
2/1/77 - - - -
30 2 4/24/77 Field Breaker Trip - - - - - - - W -
4/23/77 Two Pump Trip - - - - - - - 30-4 -
Data - -
12/22/76 1/25/77 4/8/77 1/31/77 30-1 30 1 30 6 30 5 Loss of Turbine- - - -
1/4/77 - - - - -
Generator and 31 1 Offsite Power MG Set Speed - -
12/23/76 1/29n7 1/15/77 - -
4/2(/77 -
Control 32 1 32 3 32 2 32-4 Stop Valve - - - -
1/14n7 - -
Surveillance 33-1 3/3n7-3/8/77_ _
33 2 Vibration -
11/23/76 -
Measurements 1/10/77 2/197 4/M '
1/31/77 3
34 1 34 2 7 4/24/77 34-3 Recirc Flow - - -
1/25/77 3/22/77 -
Calibration 35-1 35-2 25
1 NEDO-24562 HOLDING FOR AIR COMPRESSOR REPAIR 4
COMPLETED RSCS REPAIR AT 10:00 hr -
R 1800 hr HOLD FOR RSCS REPAIR LDING FOR MAINTENANCE TO INSTALL SV/RV1 - $
COwLETED REPAIR OF SV/RV RE ACTOR SHUTDOWN FOR SV/RV REPAIR ~
+"E" SV/RV FOUND OC GROUND IN CONTROL CIRCulT REACTOR SCRAM LOW WATER LEVEL g
FALSE LEVEL SIGNAL OPERATOR TOOK MA NUAL COflTROL _ g SV/RV ACTUATION ON E, F, C. K TESTED HPCI AND RCIC ""
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-* HOLDING FOR RFP TURBINE CONTROLS - g COMPLETED RCIC AND HPCI RUNS
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+ HOLD FOR RCIC/HPCI CONTROLLER ADJ l S
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- HOLD FOR REACTOR WATER CLEANUP O $
COND 9.3 MHO MAX _ e %
REDUCED TO ISOLATE HPCI SOR RHR V35 REPAIR I
- HOLD FOR CDD VALVE REPAIR -
S k REACTOR STARTUP 1022 hr a
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NEDO-24562 INCREASED POWER TO TC-2 (25%) HOLD FOR RECIRC CONTROLLER CAL. -+
TESTED HPCl/RCIC AT 150 psig o REACTOR CRITICAL 2045 hr a M-G SET SPEED CONTROL PROBLEM '
RHR BKR SPRING REPLACEMENT -
R EPLACE SYSTEM IN ON E CFD ~
N REPAIR LEAKING CFD VALVE OUTAGE -
R SHUTDOWN OUrflDC CONTROL ROOM TESTED RCIC -
g TESTED HPCl A
COMPLETED DSTC TESTING - E d - g o -
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DSTC TESTING BEGAN 1530 hr - m tn TURBINE FAST ACTING SOLENOlO VALVE PROBLEM PRESSURE REGULATOR RECIRCFEEDWATER g
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SET POINT CH ANGE_% y E h PRE DSTC TESTING -
- u GENERATOR ON LINE 1010 hre 12/14/76 MAIN GEN REV POWER RELAY SET TO HIGH w
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(REDUCING SETPOINT) M HPCI- NEW EGR INSTALLED AND TESTED SAT _ g j PRIMARY COMPLETED COND PUMP REVERSE ROTATON - DAMAGED IMPELLER o HPCI- EGR NOT ON SITE (LOST IN AIR SHIPMENT) -
U E GI EN BSEPll COND SYSTiM RECIRC VALVE FAILED TWICE RPV LEVEL TRANSIENT , , h PROBLEMS REACTOR CRITICAL AT 063012/11/76 -
S 61 SUFFERED FROM M LACK OF ATTENTON HPCI OIL FLUSH - o %
RECIRC SEAL DRAIN PIPING PLUGGED HPCI OIL FLUSH -
cm h d
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DRYWELL SHIM OUTAGE -
- C HPCI SECURED (WATER IN OILRESERVOIR)
SHUTDOWN FOR DRYWELL SHIM OUTAGE *
- GENER ATOR ON LINE AT 1746 hr 12/4/76 -
ADDING SNU8BERS TO HPCI STM LINES - BEGAN CRO FRICTON TESTING - -
- NO COORDINATION COULD LOCATE TEST GE AR - ROLLING MAIN TURDINE COMPLETED CRD SENSOR TESTING ROLLED MAIN TURBINE SECURED ON HIGH VfBR ATON -
- HOLLED MAIN TURBINE TO 1800 RPM WHERE NO. 2 BRG TRIPPED ON HIGH Vl8 (CRO SENSOR TESTING IN PROGRES$1HPCI CHECK CONTINUES _ n REDUCED POWER FOR DRYWELL ENTRY RCIC SU 14 COMPLETED ~
BEACTffLCBlI1qAL 1330 he
+ HOLD FOR AIR COMPRESSOR REPAIR ,
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(%) H3 mod HO13V1H 27
NEDO-24562 RECIRC MASTER LOAD FOLLOWING E DELAYED DUE TO BAD CONTROLLER COMPLETED FW CONTROL STEPS - g COMPLETED RECIRC CONTROL STEPS TOOK DATA ON EHC CONTROL INSTABILITY _
HOLD FOR STARTING HTR DR AIN PUMP RECIRC PUMP CENTR AL SYSTEM TESTING - R RFP CONTROL SYSTEM TESTING H SFCURED ONE RFP FOR CONTROL TUNING -
Q Z M ) OOK RECIRC PUMP DATA oo3
_ g
< I o AN WATER BOX STILL ISOLATED z
-j $ flP R EPRODUC181LITY "8
<mI (ROD SCR AM) SELECTED RODS SCREENED 3mu NO RWCU TESTING $$
- M3y STEADY STATE FOR Ga u.
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8m9 EHC PIPING VIBRATION 8OO Jm - N PLUGGED 80 + 12 + TUBES E5 AN CONDENSER TUBE LEAKS CONTINUES TESTED TURB CV AND SV*S ao wZ p'
REDUCED PWR TO CLOSE NO.3 CV(CYCLING) i
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- E POWER GENER ATION - g CHANGED 2 8AD COND DEMIN 7
--*- MAIN COND HAS DEVELOPED TU8E LEAKS - e
" l BEGAN DUAL RF PUMP OPERATION g
- HOLD POWER PRODUCTION e GEN ON LINE 2350 hr 3 g CRIT 8 CAL 2000 hr o SCRAM-LOSS OF DC FEED TO FW CONTROL - r-
~ CL 4 h HOLDING FOR POWER PRODUCTION _ e MAIN TUR8INE DC Oil PUMP MOTOR (BURNED UP STATON) % q
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3 D g REACTOR CRITICAL 0915 hr e W " Z SCR AM EHC FLUID LEAK
" TESTED BPV,TSV AND TOOK VIB _ ,
RECIRC FLOW CALL 8 RATION m g
-* HOLDING POWER PRODUCTON , ,
g COLD WEATHER EQq N RECIRC SYSTEM CENTH A T STS - y k PROBLEMS CONTINUED REDUCED POWER CONDENSER VACUUM b
o CIRC WATER PUMP PROBLEMS - - t
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COMPLETED OD 1/P 1 %
COMPLETED APRM/LPRM Call 8 RATION ~
COMPLE TED BPV CAPACITY TEST I d WATER IN GLAND SEAL EXHAUSTER 9 VACUUM PRO 8LEM GEN ON LINE AT 2000 hr -
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$HVIRQWN TO REPA'R LEAKS IN DRYWELL ,
COMMENCED SHUTDOWN 0030 hr 1/4/77 GEN ON LINE 1860 hr JET PUMP C All8R ATION D AT A e RE ACTOR CRITICAL 2222 hr RSCS WIRING PROBLEM _ e OUTAGE - CLEAR ENC PIPING PENETHATION (CHIP CONCRETE)
CX)MPLETED TC 2 -+ LOSS OF OFF SITE POWER (SCRAM) ,
LOAD REJECT RHR STEAM COND MODE ~ "
SCRAM CRITICAL 2346he GEN ON LINE 1337hr MSIV CLOSURE ERROR CAUSING 8PV TO OPEN (1537 hel MSIV CLOSURE. TlP REPO RCIC INJECTON, APRM CAL g lHM - APRM OVER LAP - CRD TIMING -
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29
2 REDUCED PER LOAD DISPATCHERS INSTRUCTION g
R 1
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HOLD 100% ADJUSTED , e R ECIRC M4 SET "
VOLTAGEREGULATORS TESTING SUSPENDED FOR POWER PRODUCTION RECIRC STEPS AND VOLTAGE REG INSTABILITY -
g ,
6 INCREASING RECIRC FLOW TO 1 , g 4 OF OCEAN DISCHARGE PUMP PROBLEM POWER INCREASE DELAYED BECAUSE -
@ I COMPLETED UNIT DECLARED COMMERCIAL AT 100% 0000 hr 318 77 E R 1RC STEM RUN REG FW STEPS HPCI AUTO START WITH INJECTION VALVE NOT OPENING CAUSED REA ED DISCHARGE PIPING HYDRO - REACTOR SCRAM - 1045 APRM Hi FLUXI ~
2
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7 REDUCED POWER BECAUSE OF FISH LOADING ON INTAKE SCREEN -
S q BEGIN 12 hr SOAK 9 PRECONDITIONING
- . a e
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TESTED FW CONTROL SYSTEM SETTING 1004 ROD PATTERN ~ "
COMPLETED PRESS REG TESTING - -
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2 11
l NEDO-24562 i
DO NOT EXCEED 106% LOAD LINE d
100 -
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c TC5 n
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\ \ JET PUMP NPSH
' l l - - l O
0 20 40 60 80 100 RATED CORE FLOW (%)
POWER IN PERCENT OF R ATED THERM AL POWER,2436 MWt CORE FLOW IN PERCENT OF RATED CORE REClRCULATION FLOW,77D n 106 ldfn CONSTANT PUMP SPEED LINES el NATUR AL CIRCULATION bl 20% PUMP SPEED l c) CONTR ACTUAL LOWER LIMIT OF MASTER FLOW CONTROL l di PUMP SPEED FOR PULL FLOW AT FULL POWER i
I figure 2.3.1. Test Program Power / Flow Map Startup Test Conditions.
2 12
- 3.
SUMMARY
OF TEST RESULTS 3.1 STI1 CHEMICAL AND RADIOCHEMICAL' Brunswick 1 is a heater drain (forward pumping) equipped plant; because this is a recent development in BWR's, several new procedures were developed. The above referenced report will cover these in detail, but two items are important enough to mention specifically. First, the standpipe in the deaerator tanks was removed to prevent suspended solids from settEng out; and second, the plant regularly recirculates (to the condenser and through the condensate cleanup sys'am) the heater drain water during startups. Only when no further improvement in quality is apparent does the plant bepn forward pumping. This maximizes the piping surface area that is flushed before vesselinjection and minimizes water quahty " spiking" to unacceptable levels during startup each time the heater drain pumps are turned on.
The more prosaic aspects of this test such as conformance to water quauty, technical specifications, and release limits were continually satisfied throughout the test program.
3.2 STI 2 RADIATION MEASUREMENTS 3.2.1 Level 1 Criterla Radiation doses of plant origin and the occupancy times of personnelin radiation zones was controlled in accordance with the guidelines of the standards for protection against radiation as outined in 10CFR20, USNRC General Design Criteria.
3.2.2 Discussion Radiation surveys were made prior to fuel loading, prior to initial critical, at rated temperature and pressure, Test Condtion 1, Test Condtion 3 and Test Condtion 6. The surveys included the plant general environs, all permanently mounted radiation detection instruments, and surveys of the turbine, reactor, and radwaste buildngs. All radiation levels were as expected, and all radation zones were properly denneated.
3.3 STI 3 FUEL LOADING 3.3.1 Level 1 Criteria Throughout fuel loadng the core remained subentical bf at least 0.38% .1k/k with the analytically strongest rod fully withdrawn.
3.3.2 Discuselon Fuelloading began 0100 September 15,1976. Advantage was taken of the required suberiticabty checks to morytor stroke time, indicator function, and flow checks of each control rod. Previous to fuelloadng, all rods had been scram timed.
The loEding procedure had to be modified several times because of high FLC count rates (rod blocks) (requiring FLC moves before scheduled), and low SRM count rates (requiring the FLC's to remain in core longer than scheduled).
Shutdown marpn checks were done after 16,64, and 144 fuel bundles were loaded, correspondng to September 15th, September 18th and September 19th, respectively. Five hundred and eight bundles had to be loaded before the final FLC was removed. The fully loaded core, completed 0351 September 25th, was given a quick" shutdown margin check; a more accurate check was done as part of STI 4.
A visual venfication of fuel bundle orientation and serial numbers was completed at 1630 September 25th; a video tape of the core was also made. Of the 10 days of fuelloadng, approximately 3 days were lost to seconday containment, the refueing bridge (1 1/2 days) and minor mechanical problems (1/2 day). One bundle dd not seat property the first try but was successfully seated after an inspection and interference check.
- See amo NEDE 21s44 C#wmcel and Reeochemcal Teeeng of tw srunsack Steam Elecine Plant Urve 1 sterup Test Proerem.
3-1
i l
fEDO-24562 3.4 STI 4 FULL CORE SHUTDOWN MARGIN 3.4.1 Level 1 Criteria The full core shutdown margin was successfully demonstrated to be 3.095% ak/k with the analytically strongest rod fully withdrawn.
3.4.2 Level 2 Criteria Criticatty occurred within 0.594% ak/k of the predicted critcal rod configuration.
3.4.3 Discussion Control rods were withdrawn in the normal B sequence until criticality occurred with the 38th control rod (22-23) at notch position 12; moderated temperature was 122*F. After the initia! ' riticality, rod 22-23 was pulled to notch 16 causing a positive period of 6 t seconds. From this data and appropriate rod worth curves and temperature corrections, a strongest rod out shutdown margin of 3.22% ak/k was calculated, with a projected reactivity increase of 0.125% ak/k during the initial cycle, adequate shutdown margin exists with the analytically strongest rod withdrawn at any time during the first cycle. The predicted critical was within ~0.6% Ak/k of the actual criticality (corrected for temperature and reactor period).
3.5 STI 5 CONTROL ROD DRIVE SYSTEM 3.5.1 Level 1 Critoria The withdrawal speed of each CRD was determined and adjusted where necessary to not exceed 3.6 inches per second as ind6cated by a full 12-foot stroke in a minimum of 40 seconds.
The mean scram time of all operable CRD's is summarized in Tables 3.5.1.1 and 3.5.1.2.
Table 3.5.1.1 AVERAGE SCRAM TIMES (COLD)
% inserted Time Limit 5 0.272 0.475 20 0.505 1.100 50 0.953 2.000 90 1.667 5.000 i
Table 3.5.1.2 AVERAGE SCRAM TIMES (RATED)
% inserted Time Limit 5 0.317 0.375 20 0.738 0.900 50 1.513 2.000 90 2.642 3.500 32
NEDO-24562 As indicated by the above tables the mean scam times of all rods (and the mean scram time of the 3 fastest CRD's in any 2x2 ar.ay) did not exceed the specified values.
No control rod exceeded the 7-second limit for 90% insertion.
3.5.2 Level 2 Criteria The insertion and withdrawal speed of each CRD was determined and adjusted where necessary to be 3.0 0.6 inches per second as indicated by a full 12-foot stroke in 40 to 60 seconds.
All control rod drives were within the required limit of a 15 psid variat!on in differential pressure during continuous insertion.
Scram times with normal accumulator charge fell within the time limits indicated in Figure 5.3.1 of the Startup Test instructions.
3.5.3 Discussion All CRD's were demonstrated to be satisfactorily coupled and proper 1y indicating their position. Fourteen CAD's required in/out timing adjustments. Five CRD's were selected from each sequence for scramming at various reactor and accumulator pressures; they were, respectively. 26-07,22-19,10-23,18-15, and 26 A sequence rods and 06-23, 18-11,18-27,30-31 and 14 B sequence rods. Tables 3.5.3.1 through 3.5.3.3 itemize some of the results from these tests.
Table 3.5.3.1 SELECTED CRD SCRAM TIMES OPEN VESSEL INSERTION TIMES (sec) (MINIMUM ACCUMULATOR PRESS.)
Rod 5% 20 % 50% 90%
10-23 0.283 0.540 1.041 1.839, 06-23 0.273 0.520 1.000 1.767 14-31 0.280 0.528 1.008 1.784 26-07 0.284 0.540 1.036 1.831 22 19 0.272 0.529 1.024 1.820 18-15 0.273 0.527 1.012 1.800 26-15 0.288 0.543 1.037 1.827 18-11 0.284 0.541 1.035 1.836 18-27 0.293 0.551 1.048 1.845 30-31 0.283 0.562 1.079 1.920 Similarly, the selected rod's friction tests and stroke time testing was satisfactory at the appropriate conditions.
All CRD's were scram timed at rated reactor conditions and the results are given in Tables 3.5.1.1 and 3.5.1.2. Finally, the CRD system was divided into nominal A and B sequence rods and the average scram times dotermined for each sequence at 5%,20% 50%, and 90% insertion; these data were later used in the analyses of the major transients.
Because Brunswick does not have an automatic control rori seram time recorder, it is impossible to recover rod scram timing data after planned scrams without using a complicated auxiliary instrumentation setup. Thus, the selected rod scram timing data could not readily be obtained from Test Condition 6 (because of thermallimits, PCIOMR's, etc.). The data in Table 3.5.3.1 reflects information obtained at Test Condition 3, the highest powered test condition possible without encroaching on any core limits. Since scram times are to a Greater extent a function of pressure rather than power, it was felt that the above exception would not adversely affect the scram time data.
3-3
l I
Table 3.5.3.2 ZERO ACCUMULATOR PRESSURE AND REACTOR AT RATED PRESSURE INSERTION TIMES (sec)
Rod 5% 20*'. 50 % 90 %
10 23 0.328 0.778 1.605 2.778 18 15 0.311 0.733 1.528 2.661 26-15 0.350 0.811 1.639 2.806 26-07 0.317 0.733 1.522 2.650 22-19 0.311 0.728 1.500 2.611 30-31 0.322 0.767 1.583 2.755 18-11 0.378 0.784 1.672 2.917 14-31 0.322 0.756 1.561 2.711 18-27 0.328 0.778 1.594 2.722 06-23 0.322 0.750 1.595 2.750 Tab!e 3.5.3.3 NORMAL ACCUMULATOR PRESSURE AND RATED REACTOR PRESSURE INSERTION TIME (sec)
Rod 5% 20 % 50 % 90 %
26-07 0.294 0.683 1.399 2.550 10-23 0.317 0.745 1.546 2.711 18-15 0.311 0.700 1.444 2.589 23-15 0.333 0.755 1.587 2.828 22-19 0.300 0.672 1.406 2.511 06-23 0.300 0.672 1.422 2.528 i4-31 0.300 0.694 1.406 2.533 18-27 0.334 0.745 1.472 2.622 18-11 0.306 0.700 1.495 2.700 30-31 0.306 0.700 1.461 2.622 3.6 STI 6 SRM PERFORMANCE AND CONTROL RCD SEQUENCE 3.6.1 Level 1 Criteria The operational sources and the SRM system were properly set up and matched so as to provide a signal to noise ratio
- of at least 2 to 1 and a minimum count rate of 3 cps.
e The IRM's were property adjusted so as to be on scale before the SRM's exceeded the rod block set point.
3.6.2 Discussion After the IRM's had demonstrated suitable overlap with the APRM's the final and satisfactory overlap of the SRM/lRM systems was demonstrated. -
The current versions of the A and a control rod withdrawal sequences are included as Figures 3.6.2.1 and 3.6.2.2. It has not yet proven necessary to change the sequences to compensate for core burnup and/or gadoEnia burnout. Many startups have proven the ability of the SRM system to efficiently and safely monitor the ascension to criticality and rated power in eithe. of these sequences; similarly the sequences themselves have proven to be safe and efficient o
3-4
NEDO-24562 51 2 1 51 5 6 5 47 3 4 3 4 3 47 12 13 13 12 43 1 2 1 2 1 2 43 7 8 9 8 7 39 4 3 4 3 4 39 12 14 15 15 14 12 35 2 1 2 1 2 1 35 5 8 10 11 to 8 5 31 4 3 4 3 4 3 4 31 13 15 16 16 15 13 27 1 2 1 2 1 2 27 6 9 11 10 11 9 6 23 3 4 3 4 3 4 3 23 13 15 16 16 15 13 19 2 1 2 1 2 1 19 5 8 10 11 10 8 5 15 3 4 3 4 3 15 12 14 15 15 14 12 11 1 2 1 2 1 2 11 7 8 9 8 7 07 4 3 4 3 4 07 12 13 13 12 m 2 1 m 5 6 5 02 06 to 14 18 22 26 30 34 38 42 46 50 02 06 10 14 18 22 26 30 34 38 42 46 50 NOTES:
t
- 1. CORE FLOW IS DUE TO 20% PUW CPEED UNTIL THE THERMAL POWER IS ABOVE 20%
- 2. COMPLETE WITHDRAWALS IN EACH COLUMN BEFORE GOING TO THE NEXT
- 3. WITHOA AW RODS IN EACH GROUP INDIVIDUALLY.
FOLLOW THE ORDER SHOWN BELOw' FOH GROUPS CONTROL ROD SEQUENCE 1,2,3, AND 4 +
ROD STEPS FOR SEQUENCE A ROD WITHOR AWAL ORDER FOR GROUPS 1,2,3, AND 4 GROUPS 1 2 3 4 5 6 7 8 9 1 48 GROUP 1 GROUP 2 GROUP 3 GROUP 4 2 48 3 #
(22,27) (30,27) (26,31) (26,23). ;
(30,19) (22,35) (1823) (3431) 4 "
(14,27) (26,15) 0 0,39) 5 48 (3827)
(30,36) (22,19) (34,23) (18,3I) 6 48 (22,43) (30,11) (42,31) (10,23) 7 #
(14,35) (38,19) (34,39) (18,15) ; 8 48 (0627) (4627) (26,47) (26,07) 9 48 (14,19) 0 8,35) (18,39) (34,15) to 24 48 (22,11) (30,43) (10,31) (42,23) 11 10 48 08,11) (14.43) (10,15) (42,39) 12 18 48 (46,19) (06,35) (18,07) (34,47) 13 4 22 40 (46,35) (06.19) (34,07) (18,47) 14 4 10 16 (38,43) (14,11) (42,15) (10,39) .
(30,51) Q2,03) (02,31) 15 6 8 10 30 40 (5023)
(06.43) (46,11) (42,47) (10,07) 16 l2 6 to 14 (06,11) (46,43) (10,47) (42,07) POWER 25 % 50% 75% 100%
(30,03) (22,51) (02,23) (50J1)
FLOW 37% 107% 104% 100%
i Figure 3.6.2.1. A Control Rod Sequence.
3-5
I NEDO-24562 51 1 2 1 51 5 5 47 3 4 3 4 47 13 14 16 14 13 43 1 2 1 2 1 43 7 8 9 9 8 7 39 3 4 3 4 3 4 39 15 17 18 17 15 35 1 2 1 2 1 2 1 35 6 10 11 11 10 6 31 4 3 4 3 4 3 31 12 16 18 17 18 16 12 27 2 1 2 1 2 1 2 27 9 11 10 10 11 9 23 3 4 3 4 3 4 23 12 16 18 17 18 16 12 19 1 2 1 2 1 2 1 19 6 to 11 11 10 6 15 4 3 4 3 4 3 15 15 17 18 17 15 11 1 2 1 2 1 11 7 8 9 9 8 7 07 4 3 4 3 07 13 14 16 14 13 m 1 2 I m 5 5 02 06 10 14 18 22 26 30 34 38 42 48 50 02 08 to 14 18 22 26 30 34 38 42 46 50 NOTES:
- 1. CORE FLOW IS DUE TO 20% PUMP FEED UNTIL THE THERMAL POWER 18 ABOVE 20%
- 2. COMPLETE WITHDRAWALS IN EACH COLUMN BEFORE GOING TO THE NEXT
- 3. WITHDRAW RODS IN EACH GROUP INDIVIOUALLY.
FOLLOW THE ORDER SHOWN BELOW FOR GROUPS CONTROL ROD SEQUENCE 1,2,3. AND 4 ROD STEPS FOR SEQUENCE 8 ROO WITHORAWAL ORDER FOR GROUPS I,2,3. AND 4 GROUPS 1 2 3 4 5 6 7 8 9 1 48 GROUP 1 GROUP 2 GROUP 3 GROUP 4 2 48 3 48 (26,27) 4 48 (18,35) 5 48 (10,27) 6 48 (18,19) 7 4g (26,11) 8 48 (34,19) (18,27) (22,23) (22,31) 9 48 (42,27) (26,19) (30,15) (1423) to 48 (34,35) (34,27) (38,23) (22,15)
(26,43) (26,35) (30,31) (30,23) 11 6 12 30 48 (10.43) (18,43) (22,39) (38,31) 12 8 18 48 (02,35) (10,35) (14,31) (30,30) 13 8 18 48 (02.19) (02,27) 106.23) (22,47) 14 8 26 48 (10,11) (10.19) (14,15) (14.39) 15 6 8 12 28 40 48 (18.03) (18,11) (22,07) (06,31) 16 6 8 18 40 (34,03) (26,03) (38,07) (06,15) 17 4 6 8 40 (42,11) (34,11) (48,15) (14J)7)
(50,19) (42,19) (46,31) (30,075 18 4 12 5 0,36) (90,27) (38,30) (38,15) POWER 25 % 50% 75 % 100%
FLOW 37% 107% 104% 100%
34 ,7 (18,51) (26,51) (0s,39) (38,47) i Figure 3.622. B Control Rod Sequence.
36
3.7 STI 7 REACTOR WATER CLEANUP SYSTEM 3.7.1 Level 2 Criteria The temperatures at the tube side outlet of the nonregenerative heat exchangers did not exceed 140* Fin any mode of RWCU operation.
The main cleanup pump available NPSH was determined to be greater than 10 feet during the hot standby mode of RWCU operation as defined in the process diagram.
The cooling water suppled to the nonregenerative heat exchangers was found to be within the flow and outlet temperature limits indicated in the process diagrams for the " normal" and " blowdown" mode of RWCU operation.
3.7.2 Discussion During the first heatups, the cleanup system was tested in the normal, blowdown and hot standby modes of operation.
Temperatures at the outlet side of the nonregenerative heat exchargers never exceeded 130*F. All flows and temperatures were within the Emits specified in the process diagrams. NPSH calculated during the hot standby mode was determined to be 69.6 ft.
Although not specifical!y connected with STl 7 testing, the cleanup pumps at Brunswick have experienced aHgnment problems because of the large temperature range over which they must operate. This had neccssitated realignment of the pumps each time the plant passed from " cold" to " hot" operation and vice versa. Presently, one of the pumps is testing a flexible coupling and results so far indicate that this ,s likely to become the permanent fix.
3.8 STI 8 RESIDUAL HEAT REMOVAL SYSTEM 3.8.1 Level 2 Criteria The RHR system was capable of operating in the steam condensing mode and shutdown cooling mode (with either one or two heat exchangers operational) at the flow rates indicated on the process diagrams.
The heat removal capabilty of each R HR heat exchanger was demonstrated to be at least 151 x 10' Btu /hr when the inlet flows and temperatures were as irsdicated on the process diagrams.
The process system variables were visually shown to have a decay ratio of less than 0.25 throughout each (level, pressure and differential pressure) controller's expected operating range.
3.8.2 Discussion Both RHR heat exchangers were placed in the steam condensing mode of operation. While in this mode of operation
' the differential pressure, pressure, and level controllers of both heat exchangers were exercised byintroducing step changes in set point; acceptable stability was demonstrated. While in the steam condensing mode the very low (42*F) service water temperatuas required that the service water flows be throttled to less than rated process diagram flows; however, rated flows were successfully demonstrated in preoperational testing. Shell side (reactor water) flow rates were determined to be 118,000 lb/hr and 160.000 lb/hr versus required flow rates of 99,000 lb/hr and 138,000 lb/hr (one and two heat exchangers operating, respectively),
it was demonstrated several times that the RHR system could be successfully placed in the shutdown cooling mode of operation. If, however, the process diagram flow rates were attempted the low delay heat load of the new core and the low service water temperatures caused the allowed cooldown rate of 100 F/hr to be exceeded. As a result, the service water flow rates had to be throttled to half of the process diagram flow rates of 8000 gpm.
Measured heat exchanger flows and temperatures were used to calculate a heat removal capacity corrected to rated conditions. The calculated value was 177 MBtu/hr (per heat exchanger) versus a required 151 MBtu/hr.
3-7
NEDO-24562 3.9 STI 9 WATER LEVEL MEASUREMENT 3.9.1 Level 2 Criteria
. The narrow range level system (GEMAC) readings were adjusted to agree with each other within 1.5 inches.
The wide range level indicators were adjusted to agree with each other within 6 inches.
3.9.2 Discussion The reactor water level measurement syste ms were checked at heatup and Test Condrtions 1. 4,7, and 6.These tests covered the widest possible range of reactcr power and flow conditions; 0 to 100% power and minimum to 100% core flow.
The various GEMAC and Yarway column temperatures were measured at several axial positions (spare analog thermocouple inputs to the process computer were used for this purpose) and the actual Yarway reference leg temperatures and Yarway ranges calculated. In each case the actual values compared quite closely with the assumed values; no recalibration was necessary.
At all test conditions the instruments satisfied or were made to satisfy the Leve! 2 criteria. The deviation o'f the average Yemay readings from the (assumed accurate) average GEMAC readings were plotted versus power and core flow. Results were as expected; core flow had much more influence on Yarway readngs than dd power. Both eccuracy and conservatism were served by using hot standby (initial) data to caibrate the Yarways.
Brunswick,like other BWR 4's had a noticeable increase in GEMAC sensed levelindication oscillation at high (100%)
core ficws and moderate (~50%) reactor powers.
3.10 STI10 BRM PERFORMANCE 3.10.1 Level 1 Criteria All IRM channels had adequate overlap with the SRM's and the APRM's. The IRM scrams occurred before 96% of full scale.
3.10.2 Discussion After the initial criticality the IRM's were shown to overlap with the SRM's by at least two decades. During the first heatup all of the IRM's were adjusted to produce adequate continuity between ranges 6 and 7. Finally after the initial APRM calibration, APRM/lRM overlap was found to be one decade. The IRM/SRM check did not have to be repeated since no adjustments were made to obtain IRM/APRM overlap. Any IRM channels inoperative during the first heatups were later readjusted to the above characteristics.
3.11 STI 11 LPRM CALIBRATION 3.11.1 Level 1 Criteria The local power range monitor (LPRM) gains were adjusted so that their meter readings were proportional to the neutron flux in the narrow-narrow water gap at the chamber elevation.
3.11.2 Discussion The location, continuity, and correct hookup of the LPRM to its respective meter and computer readout was verified by observance of the appropriate signal response during initial control rod movements in both sequences. The LPRM's were calibrated (to watts per square centimeter) at Test Condtions 1,3 and 6 using approved procedures. At Test Condition 1 the LPRM GAF's were calculated from BUCLE "TIPNEWRP" and *'P t NEWRP" options, the TIP and LPRM data were manually entered from TIP traces and meter readings. At the other test conditions, the process computer hardware and software had 3-8
NEDO-24562 been verified to be correct; the LPRM GAF's were calculated by the process computer OD-1 and P1 programs. Using either method, no problems were encountered in calibrating the TIPS or LPRMs to the given heat flux.
3.12 STI12 APRM CAL'9 RATION 3.12.1 Level 1 Critoria The APRM channels were calibrated to read equal to or greater than actual core thermal power at all test conditions.
In the startup mode all APRM channels produced a scram at less than or equal to 15% of rated power.
Technical specification and fuel warranty limits on APRM scram and rod blocks were not exceeded.
3.12.2 Level 2 Criteria The APRM channels were considered to be tracking core thermal power accurately when they agreed with the heat balance to within 6% of rated power.
3.12.3 Discussion The APRM system was caibrated after each change in the gain of the LPRM system and whenever changes in the rod pattern had a significant effect on the APRM readings. Actual core thermal power was calculated by a heat balance done by the process computer (a'ter OD-3 was proven accurate) or manually. During the first heatup, a special heat balance using the estimated heat capacity of the vessel and coolant was used. No problems were encounter 9 din adjusting the APRM system to read reactor power or higher.
The tracking ability of the APRM system was vertfied by monitoring all APRM readings through a 38% change (by flow) in rated power from Test Condition 6 and through a 24% change in rated power from Test Condition 3. All APRMs were shown to track within the 6% criteria. Data from the Test Condition 6 trial are presented in Table 3.12.3.1 and Figure 3.12.3.1.
Table 3.12.3.1 APRM TRACKING Highest APRM Lowest APRM
% Core Flow Actual % Power - Reading Reading 98.3 96.0 96.7 94.81 77.5 83.9 85.2 80.4 54.9 70.3 71.8 70.7 49.9 66.6 68.2 67.6 36.8 58.3 61.4 60.5 3.13 STI 13 PROCESS COMPUTER 3.13.1 Level 2 Criteria -
The process computer dynamic and static systems test cases were successfully completed; all the computer programs are operational.
Programs 00-1 and P-1 were found to calculate MCPR's and LPRM GAF's to an accuracy nf better than 2% relative to an independent method (BUCLE).
3-9
+6%
/
1G) -
l
-6%
O 80 -
O E
=
h g m - O E
2 m
w I =
g -
< A k
E sx
- b 00 - O A HIGHEST APRM O - LowEsr APRu l
u -
l l
l I I I , I I SO 70 80 90 100 ACTUAL THERMAk POWER (%) '
i l
l Figure 3.12.3.*, APRM Tracking.
3-10 1
NEDO44562 3.13.2 Discussion The dynamic system test case was completed before Test Condition 2 of the startup test program was completed. No significant problems were encountered with the process computer hardware or software, but the TIP machine / process computer interface did delay the checkout of the major nuclear programs. All of the NSS,S programs were checked out and proven operational. The process computer is capable of accurately determining plant power, thermal imits, and accumulat-ing core exposure.
At several test conditions after the DSTC was completed the process computer's OD-1 and P-1 programs were checked by comparison with the off line program BUCLE. The level 2 cnteria was always satisfied. In most cases the comparisons of CPR, core maximum peaking factor, maximum linear heat generation rate, maximum average planar heat generation rate, and LPRM GAF's were within 0.5% in value and agreed as to location.
During the startup test program several minor problems were encountered and corrected. The computer analog signal taps were rearranged in the pla nt cabinets to measure only the voltage drop across the sensing resistors (and not the sensing resistor plus wiring). This made a dfference of ~0.5% in the sensed values. A new core flow / drive flow correlation consistent with final cahbration results was entered. (See ST1-35.) Finally, the data bank was changed (after mechanically Emiting core flow to 102%) to allow the process computer to calculate CPR using the most " liberal" K, Emit curve. This corresponds to the 102.5% core flow limit rather than the master auto limit used t'oroughout the test program.
3.14 STI 14 RCIC SYSTEM 3.14.1 Level 1 Criteria The reactor core isolation cooEng system (RCIC) was able to deEver rated flow,400 gpm, in less than 30 seconds against any reactor pressure between 150 psig and rated (including Ene losses).
The RCIC turbine did not trip on overspeed during manual or auto starts. The peak turbine speed in either mode exhibited satisfactory rpm margin to the overspeed trip.
3.14.2 Level 2 Criteria The RCIC turbine gland seat condenser system was capaule of preventing steam leakage to the atmosphere.
The RCIC steam supply line high flow isolation switches have been adjusted to actuate at 300% of the maximum required steady state flow (with reactor pressure near rated).
The RCIC system controllers and mechanical components have been adjusted to provide RCIC system decay ratios of 0.25 or less, during small speed or flow command changes.
I l The RCIC turbine margin to the overspeed trip (for manual and auto starts) is at least 10% of the trip value.
3.14.3 Discussion i
The RCIC system was tested by injection from the condensate storage tank (CST) to the CST at 150 psig 550 psig and rated reactor pressures while simu*ating a pump d scharge pressure 100 psi above reactor pressure. A special 1220 psig pump dscharge pressure test was also performed. Additionally, an actual vessel injection was done at Test Condition 2.
Each of these tests were monitored for peak turbine rpm, time to rated flow, and decay ratio res,ponse to small step changes in speed and flow. At appropriate times RCIC steam supply line flow data were taken. All Level 1 and 2 criteria were satisfied; the turbine was stable at all reactor conditions.
RCIC pump performance was satisfactory and as predicted by the vendor.
The steam supply high flow switch set points were calculated by San Jose; the trip set points are 422 inches of water and have been transmitted to the appropriate CP&L personnel.
3-11 l
L
l NEDO-24562 l
Final setpoint data and measured values are summarized in Table 3.14.3.1.
Table 3.14.3.1 RCIC SYSTEM DATA 400 gpm Une Losses to Vessel 40 psi Time to Rated Flow - Vessel injection 26.6 see Peak Turbine Speed - Vessel Injection 4560 rpm Flow Controller Proportional Band 600 %
Resets / Minute 100 Ramp Generator Ramp Time 22 sec Idle Voltage -2.0 volts Woodward Controller Gain 5 Stablity 5 3.15 STI 15 HPCI SYSTEM 3.15.1 Level 1 Criteria The average HPCI pump dscharge flow was equal to or greater than 4250 gpm after 25 seconds had elapsed from initiation on auto starts at any reactor pressure between 150 psig and rated (except the 1220 psig HPCI pump discharge pressure test).
With HPCI pump discharge pressure at any value between 150 psig and 1120 psig, the required flow is 4250 gpm (the limit of 1120 psig includes a conservatively high value of 100 rsi for ine losses. The measured value may be used if available).
The HPCI turbine dd not trip on overspeed during auto or manual starts. The peak turbine speed, in any auto or manual start, exhibited a satisfactory rpm margin to avoid a possible high speed trip.
3.15.2 Level 2 Criterla The turbine gland seal condenser system was capable of preventing leakage to the atmosphere.
The efferential pressure switches for the HPCI steam supply line high flow isolation trip were adjusted to actuate at 300% of the maximum required steady state flow (with the reactor pressure near rated).
For small speed or flow command changes in either manual or automatic mode, the decay ratio of each recorded HPCI system variable was less than 0.25 in order to demonstrate acceptable stabilty (except the 1220 psig HPCI pump discharge pressure test).
The margins to avoid the overspeed trip were at least 10% of t'ie trip value.
3.15.3 Discussion The HPCI system was tested by injection from the condensate storage tank (CST) to the CST at 150 psig,550 psig, cnd rsted reactor pressures while simulating a pump discharge pressure 100 psi above reactor pressure. A special 1220 pcig pump discharge pressure test was also performed. Additionally, an actual vessel injection was done at Test Condtion 3.
Each of tnese tests we's monitored for peak turbine rpm, time to rated flow, and decay ratio response to small step changos irt speed and flow. At appropriate times HPCI steam supply line flow data were taken. All Level 1 and Level 2 criteria 3-12
r NEDO-24562 were satisfied; the turbine was stable at all reactor conditions. HPCI pump performance was satisfactory and as predcted by the vendor.
Several problems were encountered in setting up the HPCI turbine. A constant speed offset in the relationship between flow controlier output and turbine rpm was introduced by a Woodward EGR actuator with an incorrect " null" (offset) voltage. Some confusion also existed in setting the turbine high speed stop to be compatible with the pacific pump.
Additionally, four HPCI injections were unsuccessfully run before the proper compromises were made between turbine speed overshoot, injection time, and stability. The Woodward components on the H PCI turbine were completely changed out and many adjustments of the turbine's hydraulic system were required.
The steam supply high flow switch set points were calculated by San Jose; the trip set points are 230 inches of water and have been transmitted to the appropriate CP&L personnel.
i Final set point data and measured values are summarized in Table 3.15.3.1.
Table 3.15.3.1 HPCI SYSTEM DATA 4250 gpm Une Losses to Vessel 87 psi Time to Rated Flow - VesselInjection 22.3 see Peak Turbine Speed - Vessel injection 4350 rpm Flow Controller Proportional Band 400 %
Resets / Minute 20 l Ramp Generator Ramp Time 12 seconds idle Voltage -0.5 volt Woodward Controller Gain 5 Stability 5 4
3.16 STI 16 SELECTED PROCESS TEMPERATURES 3.16.1 Level 1 Criteria The reactor recirculation pumps were not started nor was a natural circulation startup attempted unless the coolant temperature dfference between the steam dome and vessel bottom head drain was less than 145'F.
The recirculation pump in the idle loop was not staited unless the temperature of the coolant within the idle and operating recirculation loops was within 50*F of each other.
3.16.2 Level 2 Criteria During operation of both recirculation pumps at rated core flow, the bottom head coolant temperature as measured by the bottom drain line thermocouple was within 30"F of the recirculation loop temperatures.
3.16.3 Discussion During heatup testing the reactor recirculation pump low speed stops were determined. One of the criteria for a permissible low speed set point is that temperature stratification of the vessel be avoided. At 20% recirculation pump speed, the following data was recorded:
20% RECIRCULATION PUMP SPEED TEMPER ATURE DATA Average R ' circulation Pump Inlet Temperature 528.8'F M Y.-
Saturation 7emperature 533.0 F 2 7S 3 'c Bottom Dranline Temperature 519.O*F 2704 %
3-13
NEDO-24562 All cnteria were easily satisfied and the low speed electrical stops were set to 20% speed and the mechanical stops were set to 19% speed.
At Test Condtions 3 and 6 the recirculation pumps were tripped as part of startup testing; a two pump trip was also performed at Test Condition 6. Temperature data for these inps and 100% core flow operation is presentedin Table 3.16.3.1.
M temperature criteria were met. and it should be specifically noted that even in natural circulation only a 55' difference existed betwe6.. the bottom drain and saturation temperatures.
Table 3.16.3.1 SELECTED PROCESS TEMPERATURES Test Condition Test Condition 3 6 Both Pumps "B" Pump "A" Pump Both Pumps Running Tripped Tripped Tripped Saturation Temperature 548.8 539 543 540 Bottom Drain Temperature 545 500 500 485 A Loop Temperature -
7.7 2 -
A (Saturation - Bottom Drain) 3.8 39 43 55 A (Loop - Bottom Drain) 19.5 - - -
All Temperatures in 'F 3.17 STI 17 SYSTEM EXPANSION 3.17.1 Level 1 Criteria
, There was no evidence of blocking of the dsplacement of any system component caused by thermal expansion.
Electrical cables were not fully stretched. Flow induced or continuous (steady state) vibration range dsplacement measurements for the recirculation system dd not exceed 0.035 inch and for the main steam system dd not exceed 0.065 inch (mean to peak).
The measured range of displacemerf. for vibration of the recirculation system due ta recirculaton pump trip (s) did not exceed 0.06 inch (mean to peak).
The measured range of dsplacements in the main steam lines for relief valve operation were less than the calculated Caplacements.
3.17.2 Level 2 Criteria At the steady state condition the displacements of instrumented (with displacement measuring devices) points dd not vary from the calculated values by more than 50% or 20.5 inch (except as described below)1.e., this criterion was satisfied if the resultant of the displacement along the $ree mutually perpendicular axes meets either the 250% or the 0.5 inch difference. Displacements of less than 0.25 inch could be neg.ected since 50% of this value is bordering upon the accuracy of the measurement. If the measured displacements did not meet these criteria, the piping design engineer was contacted to reanalyze the data with regard to design stresses.
During the heatup cycle. the trace of the instrumented points fell within a range of 150. of the calculated value from the initial cold position in the direction of the calculated value and 50% of the calculated value from the initial position in the opposite direction of the calculated value.
Hangars were always within their operating range (between hot and cold settings).
3 14
NEDO-24562 Vibration did not reach 80% of the applicable Level 1 cnteria.
The acceleration at any measured location dd not excead 5 g's, based on a sinusoidal vibration mode.
3.17.3 Discussion STI 17 testing was conducted during plant heatups, transients, and major trips. Thermal expansion data were obtained by lanyard potentiometers and actual observation at rated and intermediato temperatures. Before the first heatup, the drywell was inspected and all possible expansion problems were corrected. investigation of actual heatup data and visual inspection indicated that, in general, the drywell piping moved in the correct direction and returned to its base setting after cooldown.
Several instrumented points fell slightly short (in displacement) of meeting Level 2 criteria. These points were specifically identified in the heatup data roturned to San Jose. Further San Jose examination of these points venfied that the resulting stresses were acceptable.
Two hangars were found to be outside their operating range and were readjusted to comply with the Level 2 criteria.
Thus, system thormal expansion met or was resolved to meet all Level 1 and 2 criteria. Later heatups confirmed that the system expansion data were repeated and that thermal motion was thus unimpeded.
System expansion vibration data were taken for all the transient tests indicated in Table 3.17.3.1. All data were returned to San Jose for analysis and found acceptable; all Level 1 and 2 cnteria were met. In general, discernible vibration levels were seen only on relief valve lifts and major trips (that resulted in scrams). MSIV full closure and Test Condition 3 turbine trip data taking were waived because of previous relief valve testing or because a more severe test would be performed later.
Table 3.17.3.1 SYSTEM EXPANSION VIBRATION TESTING Test Test Being Pipe (s) to be Condition Performed Monitored Power % Flow % Date 2 MSIV Steam Lines 41 35 1-2-77 3 MSIV Steam Unes 63 99 1-13 77 6 Load Reject Steam & Recire.Unes 100 96.5 4-1-77 3 1 Recire. Pump Trip Recire. Lines 68 100 2-1-77 80 70 3-10-77 5 MSIV Steam Lines 87 79 3-11-77 6 Feedpump Taip Feedwater Lines 98 96 3-12-77 6 Full isol. (MSIV) Steam Unes Waived 5 Stop Valve Test Steam Unes 64 57 3-3-77 6 2 Recire. Pump Trip Recire. Unes 97 99 4-23-77 7 Recire. Cavitation Test Recire. Unes 20 94 1-31 77 6 1 Recirc. Pump Trip Recire. Unos 96 100 4-24-77 Rated Relief Valve Temp. & Pres. Test Steam Lines 9 29 2-27 77 3 Turbine Trip Steam and 69 100 2-12-77 Recirc. Lines 3-15
d NEDO-24562 3.18 STI 18 CORE POWER DISTRIBUTION 3.18.1 Level 2 Criteria The results of the TIP reproducibilty test indcate that the overall standard deviation of the segment averaged TIP values (BASE distribution from the process computer) of the central nodes (nodes 5-22) was less than 7.8%.
3.18.2 Discussion Complete sets of TIP scans and appropriate process compver output were obtained at Test Condtions 2,3. and 6. An offline computer program was written to analyze the random, geometrical and total noise of the TIP system.
The random noise data were obtained by the analysis of multiple traces (and BASE output edits) of the same (common channel) location for each TIP machine. The data from the OD-2 outputs were corrected for machine normalization constants (LPRM effects).
The total nc!se data were obtained by the analysis of diagonally symmetric TIP locations (and 8 ASE output edits) from e complete OD-1. All data were taken while the reactor was operating in a diagonally and rotationally symmetric rod pattem; non-central nodes were not considered. All Level 2 criteria were satisfied.
Table 3.18.3.1 summarizes the results of the above investigation.
Table 3.18.3.1 TIP REPRODUCIBILITY DATA Uncertainty Test Condition Random Geometric Total '
2 1.99% 4.68 % 5.086 %
3 1.33 % 2.905 % 3.19%
6 1.52 % 2.05 % 2.55 %
The genera lly lower values of uncertainty at the higher power levels reflect the increasingly better setup of the TIP machines with time and the higher neutron flux levels. After Test Condition 3, the TIP machine flux ampifiers were calibrated (in W/cm') to the average heat flux in the appropriate narrow-narrow channels.
Both symmetnc TIP location traces (after TIP machine calibration) and process computer bundle power edts were investigated over a wide range of reactor powers. The results confirm that the reactor core does operate with an appropriately symmetrical (rotational or mirror) power dstribution given a symmetric rod pattern.
3.19 STI 19 CORE PERFORMANCE 3.19.1 Level 1 Criteria The maximum linear heat generation rate (MLHG R) during steady state conditions did not exceed the allowable heat flux as specified in the plant technical specifications.
The steady state minimum entical power ratio (MCPR) was maintained greater than or equal to the allowable MCPR as specified in the plant technical specifications.
The maximum average planar linear heat generation rate (MAPLHGR) did not exceed the bmits given in the plant technical specifications.
3-16
NtDO-24562 Steady state reactor power was imited to full rated maximum values on or below the design flow control L 3.19.2 Discuselon Core performance evaluations were performed at each (powered) test condition and whenever it was judged necessary during routine operations. Process computer calculations, which were periodically checked by off-line BUCLE calculations, were used throughout the test program. The resulting core parameters at the various test conditions are summarized in Table 3.19.2.1.
Table 3.19.2.1
SUMMARY
OF CORE PERFORMANCE PARAMETERS Rated Power = 2436 MWt Rated Core Flow = 77 M1b/hr Test Condition Power Flow MCPR CMPF LHGR MALHGR Limit (At Rated Power) 100% 100 % (1.28) (2.43) 13.4 -
1 18.9 19.4 3.574 2.306 2.412 2.02 2 36.7 34.7 2.547 2.032 4.154 3.48 3 63.0 95.4 2.013 2.477 8.616 7.25 4 36.5 23.3 2.113 2.389 4.18 4.00 5 63.4 55.0 1.665 2.412 8.442 7.18 6 98.1 99.6 1.329 2.374 12.85 10.90 No problems were encountered in kcaping the plant within its licensed thermal Emits;in fact, even PCIOMR limitations dd not evidence themselve s until 75%-85% reactor power (depending upon Xenon conditions). Later in the test program, the Bailey mechanical and electrical stops were adjusted and the master limiter set to 100% core flow. Consistent with this, the process computer data bank was changed to allow the CPR calculation to take advantage of the 102.5% core flow K, CPR lirnt curve.
3.20 STI 20 STEAM PRODUCTION 3.20.1 Level 1 Cnteria The NSSS parameters determined by normal operating procedures were within the appropriate license restrictions.
The nuclear steam supply system was determined to be capable of supplying steam of better than 99.7% quality at a pressure of 985 psia at the second isolation valve in an amount consistent with the final feedwater temperature and control rod drive flow as given by the formula:
=
8284 Wsruu g_ + Wcno (Mib/hr) 3.20.2 Discuselon The plant was brought to an indicated power of 100% and allowed to stablize for one day. The 100 hour0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> warranty run was divided into two 50 hour5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br /> periods, and a special 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> data taking session was run during each period.
s 3-17
NEDO-24562 The averaged readings of each parameter from each period were combined into a single set of parameters to be used in determining reactor power and steam flow. Carryover was determined from STI 1 Test Condtion 6 testing.
The results of these calculations were:
Reactor Power 100.34 %
Steam Flow 10.492 Mlb/hr Carryover 0.0023 %
Outboard Isolation Valve Pressure 952.4 psig Recently caibrated plant instrumentatien was used to measure most parameters maidng up the power calculation; feedwater nozzle orfferential pressure and MSIV pressure were determined using speciaNy installed instruments.
Because of several measurement problems and accuracy considerations, an error analysis was performed on the steam flow and outboard MSIV pressure calculatens. Additionally, the calculated values had to be ocn octed to rated (100%
power) conchelons. When the above corrections were taken into account, the results indicate that aN the criteria are satisfied.
AN of the liconeed thermal limits apphcable to Brunswick Unit 1 were maintained throughout the warranty run.
3.21 STI 21 FLUX RESPONSE TO RODS (CORE POWER-VOID MODE RESPONSE) 3.21.1 Level 1 Criterte The decay ratio was less than 1.0 for each process variable that exhibited an osciNatory response to control rod moton.
321.2 Level 2 Critoria The decay ratio was less than 0.25 for each total core process variable that exhibited an osciNatory response to control rod motion when operating above the lower Emit of the master flow controller.
The decay ratio was less than 0.5 for each locaEzad process variable (LPRM) that exhibited osciNatory response to control rod moton when operating above the lower tirrut of the master flow controller.
3.21.3 N=
- At Test Conditions 1. 2, and 4 (natura circulation) a control rod and an adjacent LPRM were selected for movement cnd monitoring. The test involved moving the control rod one or more notches past the LPRM while simultaneously monitoring the response of the selected LPRM and other process variables. Over the ranges of reactor power and flow given by the above test condibons, the Level 1 and 2 critoria were satisfied. In fact, aN of the tests the LPRM signal stabilized without any observable oecellatory motion. Additionally, the groes core signals (pressure, core flow, APRM, etc.)did not display any oeculatory behavior that could be attnbuted to rod moten.
It should be noted that STI 21 testing scheduled for Test Condtion 6 was not performed because of utitty test acheduhng problems and the loss of STARTREC after the April 1977 generator breakdown.
The loss of data from this test condtion was not considered serious because of the excellent plant response to the same test in natural circulaswn, a potentially more unstable position on the power / flow map. Also, many informal observa-tions of LPRM and core response to control rod motion at high power (including Test Condtion 6) have demonstrated the plant's abitty to satisfy the Level 1 and 2 criteria of Startup Test 21.
3-18
NEDO-24562 3.22 STI 22 PRESSURE REGULATOR 3.22.1 Level 1 Criteria The decay ratio was less than 1.0 for each process variable that exhibited oscillatory response to pressure regulator changes.
3.22.2 Level 2 Criteria The decay ratio of any oscitlatory variable was 50.25 when operating above the minimum speed for the master manual recirculation system mode. Below th;s speed, the decay ratio was 50.50 with the recommendation that each control system be adjusted to meet so.25 unless there is an identifiable performance loss at higher power levels.
Pressure control system deadband, delay, etc., was small enough that steady state hmit cycles, if any, produced turbine steam flow variations no larger than 0.5% of rated flow (except as discussed below).
The response time from set point input until pressure peak was within 20 seconds in the recirculation system manual mode.
The normal difference between regulator set points was small enough that the peak neutron and thermal flux and/or peak vessel pressure remained below the scram settings by 7.5% and 10 psi, respectively.
3.22.3 Discussion The pressure regulators were tested by the introduction of 10 psi step changes at Test Conditions 1 through 6.
Various condtions of load limiting (bypass valves, control valves or both in :ontrol of the transient) and recirculation system control modes were used. Additionally, the backup capability of each pressure regulator was demonstrated via simulated failure of the controlling regulator. Actual plant test data indicated that a 5 psi bias between regulator set points was necessary for maintenance of adequate scram margins should a regulator fail.
After reviewing Tes: Condtion 3 results, the settings of both regulators were changed. Additionally, a second steam line resonance compensator card (in each regulator) was found to be necessary to reduce a 5 Hz oscillation (thought to be from the bypass piping - Brunswick 1 has a different bypass configuration than Brunswick 2). The new, final settings are desenbod in Table 3.22.3.1 and were used for the remainder of the test program. These settings allowed the regulators to successfully meet all cnteria for the remainder of the test program and for the retesting of the lower power test conditions.
Times to pressure peaks were typically 5 seconds for set point changes and 7 seconds for simulated failures, but in all cases always less than 10 seconds. Decay ratios were less than 0.25 at all test conditions and scram ovoidance margins were generous. Table 3.22.3.2 summarizes several test cordtions of the simulated regulator failure data.
During many of the ascensions to 100% power EHC data were taken to determine the steam flow to steam flow demand linearity and to determine how well the control valves were following their partial are programming. The data from one power ascension is presented in Figures 3.22.3.2.6 through 3.22.3.2.10; inearity is within limits and the control valves were found to be accurately following their programming.
The pressure regulator (possibly) has one remaining unresolved prob,3m. Oscillations have appeared at approxi-mately 735 MWe corresponding to the 30% open position of the r. umber 3 control valve; their magnitude is about 15 MWe peabio-peak and the cause has not been identified. The utiity is investigating the phenomenon and LSTG and NED have been notified.
3.23 STI 23 FEEDWATER SYSTEM 3.23.1 Level 1 Criteria The decay ratio was less than 1.0 for each process variable that exhibited oscillatory response to feedwater system changes.
3-19
Table 3.22.3.1 FINAL PRESSURE REGULATOR SETTINGS
- Pressure Regulator Pressure Regulation 3.35
) A Regulator Lag (R6) 2.3 Lead (RS) 3.5 B Regulator i
1 Lag (R4) 2.6 Lead (R3) 5.6 Steam Line Resonance Compensator Cards Card Location A42 A46 A38 ASS
- Notch Center 5.23 5.23 0.99 1.02 Notch Depth 2.0 2.0 2.0 2.02 t- Notch Width 1.655 1.655 1.77 1.77 l Small Lag 2.12 2.12 14.21 14.32 Bias F'otentiometer A Regulator in Control 5.0 B Regulator in Control 7.1 2
(5 psi Difference Between Regulators)
All units given are potentiometer turns Table 3.22.3.2 PRESSURE REGULATOR SACKUP PERFORGAANCE Test Condition 1 2 3 5 5 5 6 8 MA MA MA Failure Rectre blode Ltd LM RARI MM PS = 500 PS-1200 IAM PS=1200 Type Peak Pressure 940 947.3 963 970 970 971 1010.7 1006.5 B-+A (Umit 1025 poig) 945 942.4 967 968 969 968.5 1011 1013 A-.B Peak Neutron Flux % 26.5 49.2 74.2 76.7 77.7 77.7 101.8 100.0 B-+A i
26.5 56.7 79.1 73.8 74.8 75.7 102.2 105.0 A-B !
Peak Heat Flux % 20.7 40 6 64.4 66.0 68.9 67.0 97.9 97.0 B-. A 21.2 42.0 64.9 65.5 68.0 65.5 97.3 98.0 A-+B l
Heat Flux Umst 51.5 54.4 103.5 80.1 85.0 79.1 108.2 109 5 B-* A ;
. 51.5 54.5 103.5 80.1 82.1 82.1 106.2 109.5 A-B i
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NEDO-24562 Adequate fuel thermal limit margns were maintained following the trip of one feedwater heater; i.e., all applicable reactor and fuel Econse requirements were met throughout the transient.
The maximum feedwater temperature decrease due to a single failure was 5100*F.
The increase in simulated heat flux dd not exceed the predicted Level 2 criteria by more than 2% The predcted value was based on the actual test values of feedwater temperature change and power level in addition, the peak heat flux did not rise above the thermal power monitor scram set point.
3.23.2 Level 2 Criteria The decay ratio of any oscillatory variable was 50.25 when operating above the minimum speed for the master manual recirculation system mode. Below this speed, the decay ratio was 50.50, with the recommendation that each control system be adjusted to meet 50.25 unless there is an 6dentifiable performance loss at higher power levels.
A scram did not occur because of the trip of one of the operating feedwater pump turbines. There was greater than a 3 inch water level margin to avoid a scram for a feedwater pump turbine trip initiated from 100% reactor power.
The increase in simulated heat flux did not exceed the predicted value referenced to the actual feedwater temperature change and power level. Instrument uncertainty may be appled to the observed heat flux, but this uncertainty cannot exceed the design value for the standard plant instrumentation.
3.23.3 Discussion During Test Conditions 1 through 6 of the startup test program a 5-inch step change in level set point was imposed upon the water level control system. Single element mode, three element modes and master manual and master auto recirc system modes were tested. At low powers (Test Conditions 1 and 2) no attempt was made to optimize the feedwater :ystem with only 1 pump operating. The initial setup produced a very stable (but slow) system. After both pumps were operated for the first time, the system was optimized concentrating on stable two pump operation at low powers. Later the low power test condtions were retested. Opbmization of the level control system included the testing and setup of each feedwater turbine speed control system (signal converter).
At higher reactor powers, Test Conditions 5 and 6, level set point change results indicated that the feedwater system was not as responsive as it could be. These observations were accompanied by operator complaints of "stuggishness."
Accordingly the level control system was optimized a second time, sacrificing some stabihty at lower powers (but still maintaining decay ratios of 50.25) but gaining " responsiveness"(shortening the time to level peaks). Before the reoptimiza-tion, reactor level would drop with large, f ast increasing power changes; afterwards, this problem had been greatty mitigated.
After these final adjustments level set point change testing was repeated at all affected test conditions (to maintain a consistent set of test results at all power levels). Final level control system settings were as follows:
FINAL FEEDWATER CONTROL SYSTEM SETTINGS Proportional Band 82.5 %
Resets Per Minute 1 Mismatch Gain 48 inches Lag Time Constant 1 Second Lead Time Constant 5 Seconds Table 3.23.3.1 summarizes the results of level set point change testing with the final feedwater control system settngs.
On March 12,1977 while operating at 97% power and flow in the master manual recirc system mode and 3 element feedwater control system mode, one of the two operating feedwater pump turbines ("A" Pump) was tripped. The recirc system runback began 7 seconds after the pump trip and reduced reactor power to 60% in approximately 33 seconds. By 24 3-26
NEDO-24562 Table 3.23.3.1 LEVEL $ETPOINT CHANGE TESTING DATA Time to Recirc Control Step Decay Level Peak Test Condition Mode Mode Direction Ratios (Seconde) 1 LM 1 Element Up < 0. 25 31 Down <0.25 34 3 Element Up <0.25 30 Down <0.25 31 2 LM 1 Element Up <0.25 34 Down < 0.25 30 3 Element Up <0.25 34 Down < 0.25 34 3 MM 1 Element Up <0.25 21 Down <0.25 26 3 Element Up < 0.25 30 Down <0.25 28
- 5 MM 1 Element Up <0.25 10 Down <0.25 15 3 Element Up <0.25 19 Down <0.25 23 5 MA 1 Element Up <0.25 12 (PB-500) Down <0.25 16 j 3 Element Up <0.25 25 i Down < 0.25 25 l
6 MM 1 Element Up <0.25 18 Down <0.25 20 3 Element Up < 0. 25 31 Down <0.25 30 6 MA 1 Element Up <0.25 20 (PB - 1200) Down <0.25 19 3 Element Up <0 25 29 Down < 0. 25 31 6 MA 1 Element Up <0 2P 21 (PB = 500) Down < 0.25 20 3 Element Up < 0.25 31 Down <0 25 34 NOTES: LM - Local Manual MM = Master Manual MA = Master Automatic 3 27 l
_ . . _ _ _ _ _ . . _ .m . _ _ _ . _ _
NEDO-24562 seconds into the transient, the water level had dropped from its normal level of 36 inches to a minimum of 21.5 inches (maintaining a 6.5 inch scram margin).
Reactor level recovery and later power changes from subcooing were smooth and wel behaved. Control system response was nonoscillatory, and the overal plant response was exceHent.
Brunswick Unit 1 was the first plant to perform the "new" loss of feedwater hea6ng test that was modfied in response to new NRC concerns about feedwater temperature changes and plant modeuing. For the Brunswick plant the worst single IClure in the feedwater system (that can produce colder feedwater) was not the tripping of the last stages of feedwater heaters but rather the open failure of the valve in the line which bypasses both strings of fourth and fifth stage feedwater heeting. The former failure produced feedwater temperature changes of ~30*F; the latter failure was predcted to produce a change of ~65'F. Both failures are slow, taking place over 3 to 5 minutes.
The NED transient group produced prodcted power changes versu s initial power and varying feedwater temperature changes. Utitty (CP&L) heat balance computer codes and actual plant data was used to predict the temperature change that would result from t,ypassing the heater strings; i.e., how would the feedwater flow spit between the bypass Ine and ash-operating foodwater heaters. All of the data and prodctions were used to determine an acceptable initial power from which to begin the test without violating PCIOMR's or plant safety Emits.
On March 13,1977 the feedwater heater bypass valve (1-F10 V120) was opened with the reactor at 73% power and 99% flow. The valve was completely open in 2 minutes and steady state condtions reached in 9 minutes. The transient was smooth and wen behaved, and aN thermal Emits on the core were continuously satisfied. The increase in simulated heat flux was 111% (of initial)instead of a predicted value of 113%; au criteria were satisfied. Foodwater temperature dropped 63*F.
Before and after the transeent computer program P1 was run; during the transient reactor power, core flow, subcoohng, pressure and LPRM data were taken at approximately 1 minute intervals. The initial and final data were used to
" benchmark" the off line computer program "BUCLE." Transient etata were used as inputs to the BUCLE program "PINEWRP." It was thus possible to obtain initial and final core thermal limits (which agreed closely with the process computer) as well as transent thermallimits. As expected the CPR smoothly decreased between its initial and final value; future plants should not find it necessary to obtain transient data. Before and after transient data are included in Table 3.23.3.2. Data generated during the transient are included in Table 3.23.3.3. The react graphed in Figure 3.23.3.
, Table 3.23.3.2 THERIAAL PARAGAETERS DURING LOSS OF FEEDWATER HEATING TEST FROGA 1861 TO 1889 313 77 1661 1869 Parameter Before Peak % Change CMWt 1774.0 1961 7.26 DHS (Blu/H>m) 15.40 22.0 30.1 MCHFR 4.19 3.64 13.1 MFLPO 0.676 0.777 13.0 CMPF 2.255 2.355 4.2 TFW (*F) 392.0 329.0 16.1 CPR 1.753 1.631 6.9 (.1 CPR = 0.122)
MAPLHGR 7.75 8.90 12.9 Pressure 990.65 992.20 0.15 The above data were generated by the Off-Line Process Computer Program BUCLE using inputs from the Plant Process Computer Programs 0D-8 (LPRM Console Readngs) and 00-3 (reactor heat balance).
3-28
NEDO-24562 Table 3.23.3.3 TRANSIENT THERdAL PARAMETERS LOSS OF FEEDWATER HEATING TEST Pro-Transient Post-Transient Bench Mark Transient Bench Mark 1539 pm BUCLE Cases 1659 pm P1 BUCLE 1651 1652 1654 1656 1656 BUCLE P1 Power (MWt) 1778 1778 1774 1851 1914 1940 1946 1951 1951 Subcooling 15.46 15.48 15.40 18.71 21.54 21.85 22.07 22.04 22.04 Core Flow 76/J8 76.38 76.18 76.21 76.43 76.85 76.29 76.57 76.57 Core Peaking 2.210 2.207 2.255 2.262 2.339 2.331 2.355 2.355 2.353 CMFLPD 0.664 0.663 0.676 0.706 0.757 0.764 0.775 0.777 0.776 Location 27-8-4 27-8-4 27-8-4 27-8-4 27-8-4 29-8 4 27-8-4 27-8-4 29-8 4 CPR 1.7514 1.753 1.756 1.703 1.672 1.648 1.633 1.631 1.6 7 2 Location 9-28 9-28 9-28 9-28 9-26 9-28 9-26 9-26 9-28 MAPLHGR 7.61 7.60 7.75 8.11 8.67 8.76 8.88 8.90 8.89 Location 7-26-4 7-26-4 7-26-4 7 26-4 27 8-4 7-24-4 27-8-4 27-8-4 7244 3.24 STI 24 TURBINE VALVE SURVEILLANCE 3.24.1 Level 1 Critoria The decay ratio was less than 1.0 for each process variable that exhibited oscillatory response to bypass valve changes.
3.24.2 Level 2 Critoria The reactor did not scram because of the test. Peak neutron flux was at least 7.5% below the scram trip setting (flow biased thermal pawer monitor and 120% scram clamp). Peak vessel pressure remained at least 10 psi below the high pressure scram.
The reactor did not isolate because of the test. Peak steam flow in each line remained 10% below the high flow isolation trip setting. Additionally vessel pressure remained at least 25 psi above the steam line low pressure isolation.
, 3.24.3 Discussion The turbine bypass valves were tested at Test Conditions 1,2,3,5 and 6. The successfully completed bypass valve test program demonstrated that the EHC system had adequate capabihty to respond to abrupt changes in steam flow.
Because of the slow opening and closing times of the bypass valves (5 to 10 seconds),little or no heat flux or neutron flux spiking was observed even at Test Condition 6. Scram avoidance margins were easily maintained and. as would be expected from a slow transient, all of the observed oscillatory responses had decay ratios less than 0.25.
The 'our turbine control valves were tested at Test Conditons 1,2,5 and 6. The latter two test conditions were actually part of a series of tests along the 100% load line wherein the observed heat flux, neutron flux, and pressure spikes were extrapolated to the next higher powered test. The highest power at which the valves were actually tested was 95%; however, the extrapolations of the abcve measured parameters indicate that (not considering PCIOMR's)it is possible to safely test all four control valves at 100% power. Taking PCIOMR's into account, assuming a valid 100% power envelope and considering the magnitude of th6 obsorved heat flux spikes, a power level of 95% should be considered an operational maximum until preconditioning restrictions are removed.
3-29
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NEDO-24562 Before the startup program began, the control valve test circuitry was adjusted to maximize valve closing time; as a result valve reopening (which cannot be controlled) produced a greater transient than closing - although in a conservative drection. Addtionally, Brunswick 1 added circuitry to the EHC system that minimized control valve number 3 testing transients (by biasing valve number 4 less full closed). It was recommended to CP&L that these modifications be made permanent. In all cases of valve testing. decay ratios were less than 0.25 and the regulator behaved in a smooth, stable manner; differential steam line flow and pressure was not a problem in control valve testing.
3.25 STI 25 MAIN STEAM ISOLATION VALVES 3.25.1 Level 1 Criterla The MSIV closure times were found to be within the limits provided in the technical specifications.
For the full MSIV closura from full power, assuming no equipment failures and applying appropriate parametric corrections, predicted analytical results based on beginning-of-cycle design basis analysis were used as the basis to which the actual transient was compared. The following table specifies the upper limits of these criteria during the first 30 seconds following initiation at the Indicated conditions.
Initial Conditions Criteria Dome increase in incrosse in Power Pressure Heat Flux Dome Pressure
(%) (pols) (%) (psi) 100 1020 2 145 The feedwater system settings prevented flooding of the steam lines.
A reactor scram fimited the severity of the neutron flux and simulated fuel surface heat flux transients within thermal limits during the MSIV full closure.
3.25.2 Level 2 Criteria For the full MSIV closure from full power, assuming no eouipment failures and applying appropriate parametric corrections, predicted analytical results based on beginning-of-cycle design basis analysis were used as the basis to which the actual transient was compared. The following table specifies the upper limits of these criteria during the first 30 seconds following initiation at the indicated conditions.
Initial Conditions Criterie Dome incrosse in increase in Power Pressure Heat Flux Dome Pressure
(%) (pele) (%) (psi) 100 l 1020 0 120 The relief valves reciosed properly (without leakage) following the MSIV full closure.
During tfse tall MSIV cfosure, RCIC started without manual assistance and operated without isolating when initiated from an automatic initiaton signal, except as dscussed below.
The reactor did not scram from individual MSIV closure; the peak neutron flux was at least 7.5% below the trip setting.
The peak heat flux was at least 5% below the trip setting. The peak vessel pressure remained at least 10 psi below the high pressure scram seteng.
3-31
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The reactor did not isolate from an individual MSIV closure; the peak steam flow on each line remained 10% below the high steam flow isolation tnp setting.
3.25.3 Discussion All 8 MSIVs were tested for technical specification closing times at heatup, Test Conditions 1,2,3,5 and during the MSIV full closure. The measured times are presented as Table 3.25.3.1; all closing times were found or odjusted to be within , ,
criteria. The ability of each valve to be functionally tested by a partial (100% to 90% open) closure wes. also demonstrated.
4 Table 3.25.3.1 MSIV CLOSING TIMES ,
(sec)
Test Condition i MSsV Full MSIV Heatup 1 2 3 5 Closore F022 A 4.40 4.4 4.63 4.4 ;4.63 4.35 B 4.30 4.4 4.53 4.53 4.63 , 4.12 C 4.53 4.7 4.98 4.64 4.54 4 28 D 4.19 4.3 4.13 4.20 4.41 4.16 F028 A 4.19 4.2 3.96 3.97 4.09 4.08 8 4.13 4.2 4.09 4.09 4.44 4.35 C 4.53 4.5 4.31 4.55 4.31 4.37 D 4.19 4.1 3.86 3.86 4.20 4.21 Additionally individual MSIV closure data were taken along the 100% load line to determine the maximum reactor power at which the test can safely be performed. For each test the observed steam line flow, reactor pressure, heat flux and neutron flux spikes were extrapolated to the next higher power. The highest power at which an MSIV was actually closed was 87%; however, extrapolations of the above measured parameters indcate that it is possible to safely snut an MSIV un to 90%
power (pressure spike margin to scram becomes limiting).
The above extrapolations do not take PCIOMR's into account, but assuming a valid 100% power envelope they would be less limiting than the pressure margin to scram (but more limiting than the heat flux margin to scram). With the pressure scram raised, PCIOMR's would limit the power at which an MSIV could be shut to c. bout 93%; the other parameters would bepn limiting at about 96% power. In any case, Brunswick Unit 1's technical specifications requm this test to be run at no higher than 75% power.
On September 30,1977 Brunswick 1 performed the MSIV full closure test from 100% power /97% flow. Brunswick 1 was the first plant to attempt verification of transient modelling methods as well as demonstrating the reactor's response to a very severe transient. The reactor responses of interest (for modelling) were the pressure rise and heat flux increase that could possibly result from simultaneous closure of the MSIVs during the first 30 seconds of the transient. In advance of the actual test, pressure and heat flux rises were calculated for about 30 simulated cases which varied (one at a time) one of 9 independent variables. The resulting data were used to determine the sensitivity of the transient pressure and heat flux rises to the input variables. The 9 independent variables, their range (for the sensitivity study) and their nominal values are presented in Table 3.25.3.2. A 110 psi pressure and 0% heat flux rise was calculated for the base case using nominal nuciear characteristics; 120 psi is calculated for the same case using conservative nuclear characteristics.
The test was initisted by simulating high steam line radiation; a reactor scram (caused by MSIVs less than 90% open) occurred seven seconds later but before any pressure increase. The scram, in fact, allowed pressure to decrease ~ 10 psi by 3.6 seconds into the transient, but full closure of the MSIVs (4.3 seconds) rapid;y started pressure increasing. An initial 3 32
NEDO-24562 Table 3.25.3.2 MSIV FULL CLOSURE SENSITIVITY STUDY Ncminal Value Range Variable 100 % (85% to 100%) Initial Power 1020 psia (1010 to 1030) Initial Dome Pressure 1095 psia (1085 to 1106) Relief Valve Setpoint Pressure 85% NBR (68% to 102%) Rehef Valve Capacity 0.4 sec (0.2 to 0.6) Relief Valve Opening Delay 0.1 sec (0.05 to 0.2) Relief Valve Opening Time Constant 678 (0% to 25%) Scram Speed 270 msec (100 to 270) Scram Delay 3 sec (3 to 5) MSIV Closure Speed This results in a 110 psi pressure and 0% heat flux rise using " nominal" nuclear characteristics.
pressure peak of 1074 psig was reached by 6.6 seconds, almost simultaneously with the water level negative peak (caused by collapsing voids). Auto initiation of the HPCI and RCIC turbines, coupled with the s0ll very large feedwater flow, caused reactor pressure to level out for about 6 seconds. A second reactor pressure peak of 1096 psig was reached (by 19 seconds) and relief values A and C lifted to reduce pressure to ~1043 psig.
The reactor feedpumps were not tripped until well over a minute into the transient (they remained runting on the stored energy of the steam in the crossaround/ reheater systems), but the feedwater control system responded correctly in recovering water level. Also, as sensed level crossed its set point feedwater flow was reduced to zero. The relief valves successfully reclosed and vessel level was normal at 26 seconds but increasing rapidly because of the successful HPCI injection. The cold HPCI water and running turbine h 3ld pressure almost level until the HPCI turbine was tripped (at 87 seconds). The relief valves dd not lift again until over 2 minutes into the transient; no problems were encountered in equalizing and opening the MSIV's and getting back on the pressure regulator for a normal shutdown. A sequence of events is included as Table 3.25.3.3.
After using the sensitivity studies and inputting actual measured or calculated plant parameters existing at the time of the trip, a corrected pressure and heat flux rise of 96.4 psi and 0% was calculated for the Brunswick 1 trip. The correspondng measured values were 93 psi and 0% satisfying all Level 1 and 2 criteria. The reactor scram reduced power long before the inch vessel level tripped the recirculation pumps, thus maintaining the reactor core within thermal Emits.
During the transient the RCIC turbine auto initiated on level but tripped before it could bepn significant pumping. The trip was unexplained but it was not a high steam flow isolation;it was felt that the probable cause was the RCIC system's recent flooding. CP&L operations was notified and the RCIC turbine was successfully run at a later time, indicating the tripping problem was of a temporary nature.
3.26 STI 26 RELIEF VALVES 3.26.1 Level 1 Criteria The sum total of the capacity rrmasurements of all 11 relief valves was greater than 9.181 Mlb/hr corrected to an inlet pressure of 1112 psig 3.26.2 Level 2 Criteria Each relief valve had an individual measured capacity of greater than 0.7512 Mlb/hr corrected to an inlet pressure of 1112 psig.
3-33
l NEDO-24562 Table 3.25.3.3 MSIV FULL CLOSURE SEQUENCE OF EVENTS Time (sec) Events
-15 Fuse A718-F2B Pulled (Channel B Trip) 0 Fuse A718-F2A Pulled (Total isolation Initiation) 0.694 Reactor Scram (MSIVs at 90% Open)
~ 0.8 APRM A & Heat Flux Start Decrease
~ 1.5 Steam Flow & Reactor Pressure Start Decrease
~3.0 Vessel Level Scram (12 inches) 3.6 Reactor Pressure Negative Peak 995 peig
~4.3 All MSIVs Shut Steam, Flow -0.0 5.68 RCIC Initiation 5.77 HPCI initiation
~6 Recire Pump Trip 6.6 1st Pressure Peak (1073.6 psig)
~7 Vessel Level Negative Peak (FW Flow Still 8.8 Mlb/hr) 8.1 RCIC Turbine Start
~12 Reactor Pressure Starts to increase 13.3 RCIC Turbine Speed Zero (Turbine Tripped) 3-34
NEDO-24562 Table 3.25.3.3 RASIV FULL CLOSURE SEQUENCE OF EVENTS (Continued)
Time (sec) Events 18.8 2nd Reactor Pressure Peak (1096 psig)
(FW Flow 8.6 Mb/hr)
(Relef Valve Temperature Recorders indicate Valves A and C Open) 19.1 HPCI Turbine Start 21.2 HPCI at 3600 rpm
~24 Vessel Level at 25 inches 26.1 Vessel Negative Pressure Peak (1043 poig)
(Vessel Level 36 inches)
(FW Flow 9.7 Mb/hr) 27.6 HPCI at 4700 rpm (FW Flow 8.6 Mlb/hr) 32.2 HPCI Stable at 4400 rpm (FW Flow 1.6 Mlb/hr)
(Vessel Pressure 1052 psig)
~42 Vessel Level at 49 in.
(Vessel Pressure 1052 psig) 60 HPCI at 4400 rpm (FW Flow 0 Mlb/hr)
(Vessel Pressure 1042 psig)
~70.3 Isolation Roset 76.3 A FW Pump Turbine Trip I
initial Power 99.9 %
Dome Pressure 1003 psig FW Flow 10.48 Mib/hr Water Level 38 inches i
a 3-35
NEDO-24562 l The pressure regulator satisf actorily controlled the ieactor transient by closing the turbine control or bypass valves by an amount equivalent to the relief valve steam flow. The characteristic Iesponse signatures for each valve did not show any significant differences. l The relief valve leakage was low enough that temperature measured by the thermocouples in the dscharge side of the valves fell to within 10*F of the temperature recorded before the valve was opened. All thermocouples operated property.
3.26.3 Discusslen The relief valves were functionally tested at 250 psig reactor pressure during heatup and at rated (950 psig) reactor pressure at Test Condition 1. All the valves were individually opened for about 10 seconds and successfully shut while recording the plant response (specifically pressure) on STARTREC. Discharge thermocouple response and pressure
, regulator response was acceptable.
A review of the reactor pressure responsa to valve opening indicated similar relief valve timing and flow behavior.
Two of the relef valves had been instrumented with ultrasonic probes to determine timing response. but due to instrumentation problems and the limited number of relief valve cyc'es the amount of data obtained dd not fairly represent valve response. Therefore, the manufacturer's data were deemed an acceptable substitute.
Relief valve capacity was determined by recording bypass valve position response during the functional testing and subsequently calibrating the bypass valve positions versus an accurately measured steam flow (feedwater flow plus CRD flow). The calculated relief valve flows were corrected for reactor power and pressure, also measured during bypass and relief valve opening, and finally corrected to 1112 psig inlet pressure. A summary of relief valve capacities appears in Table 3.26.3.1.
Table 3.26.3.1 RELIEF VALVE CAPACITIES Relief Valve (Mlb/hr Flow at 1112 psig)
A 1.02 B 1.01 C 1.04 D 1.00 E 1.04 F 1.03 G 1.04 H 1.03 J 1.03 K 1.00 L 1.00 Total 11.24 Mlb/hr 3.27 STI 27 TURBINE TRIP AND GENERATOR LOAD REJECTION 3.27.1 Level 1 Criteria High Power Trips:
Predcted analytical results based on begnning of cycle design basis analysis, assuming no equipment failure and applying appropriate parametric corrections, were used as the bases to which actual transient results are compared. The 3 36
NEDO-24562 following table specifies the appropriate upper limits for these critoria during the first 30 seconds following initia6on at the indicated condtions.
Initial Conditions C.; ;-
Dome incrosse incresee in Test Power Pressure in Heat Flux Dome Pressure Transient Condition (%) (psia) (%) (pel) .
Turbine Trip 3 75 988 2 105 Generator Breaker Trip 6 100 1020 2 131 The feedwater system settings prevented flooding of the steam unes.
3.27.2 Level 2 Criteria Predicted analytical results based on beginning of cycle design basis analysis assuming no aquipment failure and applying appropriate parametric corrections. were used as the bases to which actual transient results are compared. The following table specifies the appropriate upper limits for these criteria during the first 30 seconds following initiamon at the indcated condtions.
Initial Conditions Criteria Dome incrosse Increses in Test Power Pressure in Heat Flum Dome Pressure Transient Condition (%) (psia) (%) (Pel)
Turbine Trip ' 3 75 988 0 80 Generator Breaker Trip 6 100 1020 0 106, Proseure regulator settings were such that the bypass valves regened pressure control before the turbine inlet pressure reached the low pressure main steem line isolation set point.
Feedwater control system settings helped prevent low level initiation (-32 inches) of the HPCI system and main steem Ene isointon (-32 iThes) for as long as feedwater flow remained available.
Bypees valve quick opening began by 0.1 second after the start of stop valve dosure. and 80% o' the bypass valve matkm was complete within another 0.2 second.
3.27.3 Discueelon A generator load rejection within bypass valve capacity was performed on January 4,1977 by opening the main genes ator output breakers at 20.3% reactor power. The control valves closed from 7.3% open in ~0.7 seconds. The bypass valves began to move 0.1 second after the control valves and reached a final value of 70% in about 3 seconds. No increase in heat flux or reactor pressure was observed. the APRM spikes were about 1%. The maximum turbine speed was 1920 rpm and within 36 seconds it returned to normal.
The reactor handled the transient very smoothly with no danger of an isolation or a scram. The level and pressure control systems were quite stable and maintained their process parameters within their normal range.
3 37
NEDO-24362 Brunswick Unit 1 expertenced two inadvertent turbine trips; one from 72.5% power /100% flow on February 21,1977 and the other from 70.0% power /100% flow on Feruary 3,1977. Both turbine trips were acceptable for use as the planned startup testing transient and, where applicabie, both trips met all Level 1 and 2 cnteria. Brunswick Unit I was the first plant to l attempt venfication of transient modeiling methods as well as demonstrating the reactor's response to turbine trips and load rejections (discussod later). The reactor responses of interest (for modelling) were the pressure rise and heat flux increase that could possibly result during the first 30 seconds of the transient. In advance of the actual test, pressure and heat flux rises were calculated or about 20 simulated cases which varied 6 independent variables. The resulting data were used to determine the sensitivity of the transient pressure and heat flux rises to the input variables. The 6 independent variables, their range (for the sensitivity study) and their nominal values are presented as Table 3.27.3.1. A 75 psi pressure and 0% heat flux l rise was calculated for the base case using nominal nuclear characteristics; 80 psi is calculated for tha same case using i
conservative nuclear characteristics.
i Table 3.27.3.1 TUR8INE TRIP SENSITIVITY STUDY l Nominal Value Range Variable j
75 % (65% to 85%) Initial Power 988 psia (988 to 1000) Initial Dome Pressure 26 % (20% to 35%) Bypass Valve Capacity 1
0.1 sec (0.05 to 0.2) Bypass Valve Delay Time 678 (0 to 25%) Scram Speed
! 270 msec (100 to 270) Scram Delay This results in a 75 psi pressure and 0% heat flux rise using "nominar nuclear characteristics.
, The February 21,1977 turbine trip was accidentally initiated from a false high water level (58 inches) signal from a l
mistakenly drained reference leg. The resulting turbine inp was in all respects identical to a planned trip except that the feedpump, HPCI, and RCIC turbines were tripped with the main turbine; a sequence of significant events is included as Table j 327.32.
i The bypass valves were quick enough to satisfy the bme to first motion and opening time criteria; however, because the reactor isolated on low level (because of the feedpump turbine trip) the feedwater and pressure control systems could not j be fairly tested. HPCI and RCIC dd successfully auto initiate and inject.
Usmg the sensitivity studies and inputting actual measured plant parameters existing at the time of the inp, a corrected
! pressure and heat flux rise of 58.4 psi and 0% was calculated for this Brunswick 1 turbine trip. The corresponding measured values were $1.1 psi and (essentially) 0%
The February 3,1977 turbine trip was a legitimate trip caused by low condenser vacuum, caused in turn by the loss of the circulating water pumps. Since the feedpump turbines did not (immedately)Inp, the resulting transient was closer to the Intent of the planned turbine inp. The sequence of events is included as Table 3.27.3.3.
The bypass valves were again quick enough to satisfy their time to first motion and speed criteria, thus enabling them to regain pressure control before a low pressure isolation was reached. The feedwater control system prevented steam line 4
flooding (by reducing flow to zero) and low water level (by increasing flow to maximum). The lowest level reached was approximately -8 inches, the reactor did not isolate nor did HPCI or RCIC initiate.
By using the sensitivity studies a corrected pressure and heat flux rise of 54 3 psi and 0% was calculated for this trip.
The corresponding measured values were 49.6 psi and (essentially) 0% Both turbine trips met all Level 1 and 2 cnteria; neither tnp opened a relief valve and no neutron flux spikes were observed.
i 1
3-38
J NEDO-24562 Table 3.27.3.2 FEBRUARY 21,1977 TURSONE TRIP SEQUENCE OF EVENTS Time (sec) Event 0 Turbine, Feedpump, HPCI & RCIC Trip 0.100 Stop Valve No. 21st Motion, Heat Flux Peak 0.160 '3ypass Valve 1st Motion, Reactor Scram 0.200 Stop Valve No. 2 Shut 0.260 Control Valve 1st Motion 0.340 Main Generator CB's Open, Bypass Valve No.1 Full Open, Total Bypass Valve Position 83.5% Open Control Valve Fast Closure Signal 0.720 Control Valves Shut 1.320 Turbine Speed Peak (101.6% ' speed)
~3.0 0-inch Water Level 3.460 Reactor Dome Pressure Peak (1026.69 psig) 6.0.!O HPCI and RCIC Auto initiate on Level 6.920 Totalisolation Signal
~7 APRM's 0%, Heat Flux 49%
12.40 Bypass Valves in Pressure Control
'(Valves Closing)
Initial Power 72.5 %
Dome Pressure 975 psig FW Flow 7.25 Mlb/hr Water Level 37 inches On February 28,1977 a generator load rejection was performed from 100% power and 97% flow by opening the main generator output breakers. The sequence of events is included in Table 3.27.3.4. As for the turbine trips a parametric study was done in advance of the transient wherein about 30 simulated cases - which varied (one at a time) one of 10lndependent variables -were modelled. The resulting data were used to determine the sensitivity of the transient pressure and heat flux rises to the input variables. The 10 independent variables, their range (for the sensitivity study) and their nominal values are presented as Table 3.27.3.5. A 95 psi pressure and 0% heat flux rise was calculated for the base caso using nominal nuclear characteristics: 106 psi is calculated for the same case using conservative nuclear characteristics.
3 39
l NEDO-24562 Table 3.27.3.3 FEBRUARY 3,1977 TUR88NE TRIP SEQUENCE OF EVENTS -
Time (sec) F. vents 0 Turbine Trip (from Low Vacuum) 0.100 Stop Valve No. 21st Motion 0.160 Bypass Valve 1st Motion, Reactor Scram 0.200 Stop Valve No. 2 Shut, Control Valve 1st Motion 0.340 Control Valve Fast Closure Signal 0.720 Control Valves Shut 0.820 Turbine Speed Peak (100.8% Speed)
~3.0 APRM's 3%, Heat Flux 57%
3.480 Reactor Dome Pressure Peak (1013.69 peig) 4.25 Negative Level Peak (-8 inches) 15.660 Bypass Valves in Pressure Control initial Power 70.0 %
Dome Pressure 964 psig FW Flow 6.62 Mlb/hr Water Level 35 inches J
Because the trip was above 30% power (and not within the capacity of the bypass system) the reactor scrammed at 35 miniseconds on control valve low oil pressure. The bypass valves were quick enough to satisfy their first moton and speed criteria and to show that the pressure regulator had regained pressure control before the low pressure isolation was reached.
Feedwater flow was available throughout the transient, and the control system acted to prevent a low level isolation and steam Ene floodng. The lowest water level reached was -30 inches, and HPCI and RCIC dd not auto Initiate.
i The turbine speed peak was 104% and was reached 1 second into the transient. Reactor pressure peaked at 1094 pog,2.9 seconds into the transient, and the three relief valves (C, D and E) which opened to relieve pressure successfully 1 reciosed.
1 Using the sonstivity studes a corrected pressure and heat flux rise of 90.0 psi and 0% respectively, was calculated for !
this load rejection. The correspondng measured values were 86.3 psi and 0%; addtionally no APRM spikes were recorded. 3 The load reject test satisfied all Level 1 and 2 criteria.
3-40
NEDO-24562 Table 3.27.3.4 FESRUARY 28,1977 LOAD REJECTION SEQUENCE OF EVENTS Time (sec) Event 0.0 Main Generator Breaker Trip, CV Fast Closure initiated 0.035 CV Rrst Motion. Sctsm 0.050 BPV First Motion 0.075 SV No. 2 Arst Motion 0.165 SV No. 2 Full Shut 0.235 BPV 80% Open 0.305 BPV 90% Open 0.395 CV Ful' Shut 0.505 BPV Full Open 0.950 Turbine Speed Peak (1879.58 rpm) 2.0 APRM 12% Heat Flux 84%
2.885 Peak Dome Pressure (1093.97 psig) 3.25 VesselLevel0 inches 6.25 Vessel Level Peak -30 inches APRM 0% Heat Flux 56%
21.5 Dome Pressure Negative Peak (912 psig)
Water Level 20 inches 22.5 Bypass Valves Shut initial Power 100 %
Dome Pressure 1007.6 psig FW Flow 10.48 Mlb/hr Water Level 38 inches 3-41
- - . - = - - - _ _ _ - - - _ - _ - - ~ . . - - ~ _ - . - . _ - _ . . - . - . . -
Table 3.27.3.5 GENERATOR LOAD REJECT SENSITfWITY STUDY Nominal Value Range Verlebte 100 % ~ (85% to 100%) Initial Power 1020 pela (1010 to 1G50) Initial Dome Prosaure 85 % (68% to 102%) Relief Valve Capacity
- 0.1 sec (0.05 to 0.2) Reuel Valve Time Constant 1095 pele (1085 to 1106) Rotet Valve Setpoint 0.4 sec (0.2 to 0.6) Rotel Velve Doley Time 26 % (20% to 35%) Bypees Velve Cepecity 0.1 sec (0.05 to 0.2) Bypees Valve Delay 678 (0% to 25%) Scram Speed 270 meec (100 to 270) Scram Doley i TNs res a in a 95 pel pressure and 0% heet flus rise using " nominal" nucteer cherecterletics 3.38 STI 30 SNUTDOWN FROtl OUTSIDE THE CONTROL ROOld
}
3.25.1 - Level 2 Critorie During the simulated control room evaluatson, the reactor was brought to the point where cooldown was initated and ,
under control, the reactor veneel pressure and water level were controlled using 7 'n.c.; and controis outside the control i
room.
3.38.3 Dioeuselon l '
The reactor was scrammed and leolated from ~20% power by tripping both RPS M O set circuit breakers Ade<pate j vesselievel wee mentained by the comeling feedpumpe. At no time did veneel water level approach the HPCl/RCIC system's auto irubation level (-30 inches). Reactor pressure chd not rise to the rouel valves' preneure set point; as water level toes the i operator manually opened 2 retof valves (one at a eme) for 10 seconde. Since the pressure peak was only 1006 peig, the velve openings were more of a fvnctional check of the shutdown panel rather then a pressure retet function. '
At 840 peig and 66-inch reactor pressure and level, a successful cooldown having been started, the test wee terminated 19 minutes after irstiaton.
3.29 STI 29 PLOW CONTROL f i
3.39.1 Level 1 Criterte !
The decay rato wee less than 1.0 for each process verlable that exhibited oscNietary response to recirculaton flow
. changes.
3.39.3 Level 2 Celterte l The decay rato of any osciliatory verlebte was less than 0.25 when operating above the minimum speed for the mester manuel recirculaton modo, except as docussed below. Below tNo speed, the decay rabo was less than 0.50 with the recommendeWon that sech controlloop be adjusted to meet 50.25 unises there is an idonellable performance lose at higher l Power low *ie.
l The auto load following range along the full power rod Sne was et least 35% of rated power, except as docussed l below.
3 42
NEDO-24502 in both the master manual and master auto modes, along the power flow line on which the load following specificabon was demonstrated, a flow change which would cause a 65% to 100% power change on the 100% load line was accomplished in 60 seconds.
Plus or minus 10% and 220% power step changes were performed within 40 seconds.
A reactor scram did not occur due to the flow controf maneuvers. The neutron flux margin was a 7.5% a nd the heat flux margin e5.0%. All the power maneuver rates were extrapolated to those that would occur along the 100% rated rod line to obtain their true effect.
Flow control system deadband, delay, etc., was small enough that steady state Emit cycles (if any) produced turbine steam flow variations no larger than 10.5% of rated steam flow. '
3.29.3 Discussion The recirculation flow controf system was tested (after STi-32 optimizats ) it Test Conditions 1,2,3,5, and 6 in the master automatic (precondhoning and nonprecond boning modes) and master manual modes of operation (the reactor power and flow were actually somewhat higher at Test Condtions 2 and 5 to allow operation on the master recirculation controller). Small flow and power changes and large ramp changes in power were used to demonstrate acceptable system operation; the data from all power changes were examined for flux scram margins, stabilty, and responsiveness. Results of system testing were used to determine controller settings for precondtioning and nonprecondihoning modes of system operation. Additional test data were used to set the various limiters in the recirculation system.
Results of recirculation system step flow change testing are presented in Table 3.29.3.1. Test Condtion number 1 teshng was not included in Table 3 29.3.1 (because it cannot be performed oa the r aster recirculaton controller) but the local manual tests passed all criteria. Because the high APRM decay ratios at Test Condcion 2 did not repeat at Test Condition 5 (a potentially more unstable position on the power flow map) and because a recirculation sistem noise problem was thought to have existed during Test Condition 2 testing, the tests were not repeated. Although within criteria, the step change times in Table 3.29.3.1 are somewhat deceptive; times from 10% to 90% of the step were typically 810 seconds with the remaining time spent in system settling. The long power change times in the precond.tioning mode (PB - 1200%) are, of course, dehberate.
Remaining flow testing consisted of large ( ~ 50% flow) flow ramps beg nning from 100% core flow at Test Condhons 3 and 6. The ramps from Test Condtion 3 (70% load line) had to be used to demonstrate ultimate plant performance along the 100% load line (by extrapolation) because of PCIOMR restrictons. After several trials a proportional band of 500% was chosen for the nonpreconditioning mcde of master auto operation; this produced ramp times within enteria, flux margins, and the required stabilty. Both master auto (PB = 500%) and master manual modes are capable of producing a 35% power change along the 100% load Ine within one minute. A proportional ba nd of 1200% was chosen for the precondationing mode of master auto operation. The lower gain provided much slower ramp response, typically,12% power / minute on the 100%
load line, which is well within the PCIOMR required 15%/ minute.
The master recirculation controller limiter settings needed to produce the (approximately) 35% power change along the 100% load Nne corresponded to 41% and 89% recirculation pump speeds. Specifically,89% recirculabon pump speed results in 100% core flow at 100% power; the Bailey scoop tube positioner electrical and mechanical stops were set to pump speeds corresponding to 101% and 102% core flow. Several of the Test Condition 8 large power ramps fell 1% or 2% short of meeting the required 35% power change. Although there was enough decay ratio margin to lower the minimum master recirculation controller speed, this would have required extensive rotestmg. Because of the small failure margin, the utilty's desire not to retest and because the NSSS contractual requirement is approximately 35% power,it was decided to accept the test results as sabsfactory.
Data from the large power ramp tests are summarized in Table 3 29.3 2. The reactor response to all flow induced power changes was stable, predictable (because of the cam shape), well beh,ved, nonoscillatory, very responsive and had adequate flux margins to scram. Final controller settings are indicated below; the controller was left in the preconditioning mode.
3 43
NEDO-24562 FINAL MASTER RECIRC CONTROLLER SETTINGS Proportional Band 500 % Nonpreconstioning Mode 1200 % Preconditioning Mode flosets/ Minute 10 Rate 0 (Minimum)
Limiter Settings Master Controller High 89% Speed (100% Core Flow at 100% Power)
Low 41% Speed Belley (High Speed)
Mechanical 102% Core Flow (at 100% Power)
Liectrical 101% Core Flow (at 100% Power) 3.30 STI 30 REClRCULATION SYSTEM 3.30.1 Level 1 Criterie Not appicable.
3.30.2 Level 2 Crtseria The single recirculation pump trips did not result in a Ngh water level turbine trip. The level margin was at least 3.0 inches at Test Condition 6.
The reactor dd not scram during the recirculation pump restart. The scram avoidance margins were 7.5% for neutron flux and 5% for thermal flux.
3.30.3 Discueelon STl 30 testing consisted of a cavitation search, recirculation pump trips, and data accumulation; they are dscussed separately below.
Data from the recirculation pump trips is presented in Table 3.30.3.1. In all cases reactor power, pressure and core flow decreased as expected. The feedwater system was able to mitigate the resulting level swell and maintain adequate 3 44
T NEDO-24562 Table 3.29.3.1 RECIRC SYSTEM STEP FLOW CHANGE TESTING Test Condition 2 3 4
Negative Positive Nogettve Poeottve MM MA MA MM MA MA MM MA MA MM MA MA PS.500 PS -1200 PM = 500 PS-1200 PS=500 PS-1200 PS=S00 PS=1200 Decay Ratios Core Flow <0 25 s0 25 <0 25 <0 25 <0 25 <0 25 ( 0 25 so 25 < 0 25 < 0 25 < 0 25 < 0 25 A Dnve Flow <0 25 <0 25 < 0 25 < 0 25 < 0 25 <0 25 <0 25 < 0 25 ( 0 25 <.0 25 < 0 25 < 0 25 B Drtve Flow ' <0 25 <0.25 <0 25 < 0 25 <0 25 < 0 25 < 0 25 < 0 25 <0 25 < 0 25
- 0 25 (0 25 APRM 0.5 0 465 =.0 25 0 44 0 418 m 0 25 < 0 25 < 0 25 < 0 25 <0 25 <0 25 <0 25 Trip Margine (%)
Heat Flus 34 36 33 5 38 34 5 37 45 41.5 39 48 39 di APRM 70 74 75 58 52 59 57 61.5 53 50 35 45 Steady State Umst Cycle <0.5% <05% <05% < 0 5% < 0 5% <05% (05% <05% <05% (05% <05% <05%
Step S2e(% Power) 92 86 8.1 93 7.9 95 67 7.2 7.1 78 7.4 81
- Time to Complete Trarwent (sec) 38 22 22 32 32 18 20 32 66 20 36 82 5 8
, Negative Positive Nogettve Poeotive MM MA MA MM MA MA MM MA MA MM MA MA PS = 500 PS = 1200 PM = 500 PS = 1200 PS=500 PS =1200 PS = See PS = 1280 Decay Ration Core Flow <0 25 ( 0 25 < 0 25 (0 25 ~ 0 25 - 0 25 <0 25 < 0 25 < 0 25 < 0 25 < 0 25 ( 0 25 A Dnve Flow <0 25 < 0 25 <0 25 (0 25
< 0 25 0 25 < 0 25 < 0 25 < 0 25 < 0 25 < 0 25 < 0 25 i D Drive Flow < 0 25 (0 25
- 0 25 (0 25 <0 25 <0 25 <. 0.2 5 <0 25 <0 25 < 0 25 <0 25 < 0 25 APRM ( 0 25 < 0 25 <0 25 <0 25 < 0 25 ( 0 25 < 0 25 < 0 25 <0 25 < 0 25 <025 < 0 25 i Trip Mergine (%)
Heat Flus 24 20 7 18 9 19 1 16 8 19 3 20 1 20 20 4 16 5 16 1 19 2 APRM 65 47.7 472 40 33 8 46 6 19 2 2 20 4 15 7 13 5 18 6 l
l Steady State bmit l Cycle
<05% <05% <05% <0 5% 0 5% <05% <05% <05% <05% = 0 5% <0 5% <05%
' Sep Sie (% Power) 85 85 84 92 87 91 10 9 11.5 10 4 12 5 17 8 178 Time to Complete Transient (sec) 32 40 48 18 40 68 38 40 120 38 34 54 1
3-45
Table 3.23.3.2 LARGE FLOW RAMP TESTING Test Condition 3 8 Neseew Pocahre Nessem pooneve uns asa un nas ma esa su asa esa aus esa esa re-see Pe.tsee ps.soo Pe-tase Ps. sos pe.lses ps.ses pe .tsee Decay Recoe Core Flow < 0 25 ( 0.25 < 0 25 < 0 25 <025 < 0 25 = 0 25 <02S < 0 25 < 0 25 '
< 0 2S A Drwe Flow < 0 2S <025 < 0 25 = 0 25 = 0 25 = 0 25 < 0 25 <025 <0 25 < 0 25 -
( 0 25 B Drtve rlow < 0 25 <025 ( 0 25 = 0 25 < 0 25 <0.25 < 0 25 < 0 25 (02S < 0 25 - <0 2S APRM < 0 25 < 0 25 < 0 25 ( 0 25 < 0 25
- 0 25 <025 < 0 25 < 0 25 <0 25 - < 0 25 Tap Marone Heat Flus - - - 40 40 42 7 - - - 15 -
12 APRM - - - 39 3e 41.3 - - - 19 - 20 DNorental Power (%) 23.3 25 4 23 3 25 4 24 1 23 3 34 7 32 0 31 7 35 6 - 30 s Flow (%) 44 e 49 5 de 1 49 0 49 0 49 0 $3 0 $3.3 $2 4 $4 0 - et s Time to Cornpiece Flow Change toec) 3e $4 264 56 44 440 $2 46 400 210* - 1So Time to Cornpiste Power Change (secl 36 46 264 50 4J 480 50 44 400 210* - 1S4
'PCIOMM RootWone margins to the high level trip (58 inches). Because of the flow biased scram and rod block, reactor power was reduced w6th control rods shortly after the 2 recirculation pump trip.
Data from the redrculation pump trips is presented in Table 3.30.3.1 in all cases reactor power, pressure, and core flow decreased as expected. The feedwater system was able to metigate the resulting level swell and maentain adequate margins to the high level trip (58 inches). Because of the flow biased acram and rod block, reactor power was reduced with control rods shortly after the 2 recirculation pump trip.
Operation a'ter the recirculation pump (s) were tripped was stable and well behaved, even in natural drculation. The pump restarts were equally uneventful and all mair.tained adequate flux margins to scram. Vibration and performance data wero taken during the pump trtpe and later rnaneuvering necessary to restart them and subsequent retum to irwtlal power. AH Level 2 cnteria were met.
A cavitation search was made from Test Condtion 3 and 95% flow by inserting control rods in the inverse order of the operatonal sequence. The resulting power reduction co.1tnued unbl the feedwater flow imit that Irvtlales a recirculation l pump rt.nback was reached (A and B redre pump " flow Imit"); since the scoop tubes had been locked before the teet was
! started an actual runback could not occur.
l Because the runback signal was received at 25% power (21% feedwater flow) and the nominal set point is 20%
l feedflow, power was further reduced Io 20% (17% feedflow). No signs of cavttabon were observed in the jet pumpe or recirc j pumps at any power level during the test.
l Recirculation system performance data were tsken at Test Condtion 1 along the 60% and 100% flow controlines and l at fest Condtion 4 (natural circulation). Performance of the system was satisfactory at all condttons.
l I
! 3 46 l l
l l
NEDO-24562 Table 3.30.3.1 RECIRCULATION PUMP TRIP DATA Level Neutron / Thermal Recirc Level Swell (inches)/ Trip Flux Mergin Pump inittel Type of Time into Tranelent Mergin to Scram Tripped Power / Flow Trip (sec) (inches) (Pump Restart)
Motor No Spikes B 62 %/100 % Breaker 6.5/56 19 Observed Field A 96 %/100 % Breaker 11/43 14 55%/7%
Motor No Spikes /12.5% A Pump Both 97%/99 % Breaker 8/50 17.5 No Spikes /14.3% B Pump 3.31 STI 31 LOSS OF TURetNE GENERATOR AND OFFSITE POWER 3.31.1 Level 1 Criteria HPCI and/or RCIC systems were not required to mainten the reactor vessel water level above the initiation level of the LPCI, core spray, and automatic depressurization systems.
All safety systerrs, including the reactor protection system RCIC, HPCI, and the desel generators functioned property or were made to function property without manual assistance The reactor protection system prevented violation of neutron flux and simulated fuel surface heat flux thermal power imitations.
3.31.2 Level 2 Criteria Reactor water level, neutron flux, heat flux, and presaure signals were not unespleinably worse than the predctions in the FSAR and plant transeent safety analysis report.
The normal reactor cooling systems were able to mentain adequate suppression pool water temperature, maintain adequate drywell cooling, and to prevent actuation of the automatic depressurization system.
3.31,3 Discuselon The test was performed at 32% reactor power by manuelty operating a (dfferential) relay to simulate a generator fault.
The plant s#tch gear had been previously arranged to simulate failure of the busses that the automatic s#tching circuitry would fast transfer to.
The main turtsne tripped and the reactor scrammed andisolated as expected. Reactor water level dd not drop to the HPCl/RCIC automatic initiation set point. No increase was noted in the heat or neutron flux signals, nor was the transiont behavior of the level, flux and pressure signals worse than the transsent described in the FSAR. Suppression pool and drywc!!
temperatures were maintained at adequate levels throughout the transient and resulting cooldown.
Five minutes after the isolation, reactor pres sure hadincreased to the point where three reRef valve s cpened; a fourth vane was manually opened to control reactor pressure. The RCIC and HPCI systems were manually started and used to successfully control vessel Nivel and pressure.
3 47
As a result of the simulated failure of all a-c power, all desel generators were expected to start, but only number 1 and number 2 were expected to synch and foad into the emergency busses. Diesels number 3 and number 4 started (and were l not expected to load) but Inpped off after approximately 30 seconds due to low oil pressure. A plant modfication was generated, approved, and installed to prevent this cond tion from repeating. The loss of off site power was again simulated in the same manner (but with the reactor shutdown); all four diesels performed sat:sfactorily.
3.32 STI 32 RECIRCULATION M-G SET SPEED CONTROL 3.32.1 Level 1 Criteria The decay ratios of each process variable that exhibted an oscillatory response to recirculation f.i O set speed changes were less than 1.0.
3.32.2 Level 2 Criteria The decay ratios of e.ch process variable that exhibited an oscillatory response to recirculation M-O set speed changes were less than 0.25 when operating above the lower speed Brnter of the master recirculation controiler and less than 0.5 when operating below the lower speed limiter, The recirculation speed control systems were adjusted to meet the requirements of the Level 2 open loop perfct-mance enteria of the startup transient test specification.
After a 10% speed demand step from the low end of the speed range, the time from the step demand to the generator peak speed was less than 25 seconds.
Steady state limit cycles wore less than 0.5% as measured by the turbine steam flow.
Step inputs between 90% and 100% speed were adjusted so that 10% of the demanded change was reached within 2 seconds, and the time between 10% and 90% of the demanded change was less than 5 seconds, except as docussed below.
3.32.3 Discussion The recirculation M O set speed control loops were not really set up until the plant first reached 100% flow at Test Condition 3. Enough data wero taken to enable property shaped cams to be cut for each M O set; compar6 son of the speed and speed demand meters over the operat ng ranges of interest indicate a good shape was obtained. It was also necessary to readjust the Bailey positioner linkage to ensure that an appropriately large portion of the cam's surface was used.
Nter the above setup was accompished, the M O set controllers were adjusted for best response at approximately 5 points evenly spaced in core flow (--20% to 90% speed). Over the speed ranges of most plant operation (30% to 90%), the recirculation system was adjusted to meet all Level 1 and 2 cnteria.
Some difficutty was encountered at high speeds (90% +) both t:ecause the recirculation M O set speeds had to be delibere tely mismatched to avoid exceeding the 100% core flow limit and because of voltage regulator instability problems.
The volta 0e regulator problems will eventually be resolved by correcting an MV/l converter noise problem and by new vultage regulator controller settings. This checkout also discovered a tachometer calibration problem. The test schedule did not allow time to retest at 100% recirculation M O set speeds; the data obtained are from ~94% speed.
In any case, Ihis is not an operational problem since the Daeley mechanical and efectrical stops have been set at ~ 92%
arvi -91% speed (102% and 101% core flow at 100% power). Over the speed ranges of interest, the M O sets are very stable Cnd responsive and meet the Level 1 and 2 enterta. Final controller settings are:
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NEDO-24562 A M-G Set S M-G Set Proportional Band 200 200 Resets / min 15 16 Rate 0 0 3.33 STI 33 MAIN TURSINE STOP VALVE SURVEILLANCE TEST 3.33.1 Level 2 Criteria The reactor did not scram because of these tests. The peak neutron flux was maintained at least 7.5% below the scram tnp settings. Peak reactor pressure remained at least 10 psi below the Ngh pressure scram setting. Peak heat flux was maintained at least 5% below the scram trip setting.
The reactor dd not isolate because of these tests. The peak steam flow in each steam kne remained at least 10%
below the Ngh flow isolation trip setang.
3.33.2 Discuselon The turbine stop valves were tested at Teet Condition 3 (63% power) and later were tested along the 100% load Ine at increasingly Ngher powers. Each stop valve was individually closed wNie monitoring steam Ine flow, neutron and heat flux and reactor pressure. The Nghest value of each parameter (for any stop valve closure) was then plotted versus reactor power and extrapolated to the next anticipated powor level. If the scram avoidance margins sabafled the Level 2 criteria,Ihe test was repeated at that Ngher power level.
In tNs manner the Nghest allowable power level at wNch a single stop valve could be closed and maintain prudent seram avoidance margins was determined to be - 93% (Emited by reactor pressure). Considering ihe results of other BWR/4 leshng, ins is a low value. However, it is perhaps bsased by the lower than normai reactor pressure (for the power level) and the 1035 peig (versus 1045 poig) pressure scram set point. Further testing could result in a higher permissable power level, but in any case the observed heat flux spekes would quickly limit the Nghest allowable power fetel to -95%96% pow because of current PCIOMR restrictions. Even tNs value assumes a valid 100% envelope and a test in the same rod at wNch the envelope was stored.
3.M STI M VISRATION MEASUREMENTS 3.M.1 Level 1 Criteria The peak stress intensity ed not exceed 10,000 psi (single amphtude) when any (vesselinternals) component was deformed in a manner corresponding to one of its normal or natural modes. TNs is the low stress limit which is suitable for sustained vibrason in the reactor environment for the design hfe of the reactor components.
3.M.2 Level 2 Criteria The peak stesis intensity did not exceed 80% of the Level 1 cnteria.
3.M.3 Discueolon Before fuel loading, RTD's, strain gauges, and acceleromoters had been installed in or on selected reactor vessel internals. During heatup, through generator synchronifation, and at selected points to 100% power, feedwater sparger tempere.ure data were obteined. Although there are no criteria associated with this test, a subsequent analysis indcated that the spargers' blend radil are well protected against thermal cycling.
3>48
NEDO-24562 During the single motor breaker trip at 70% power and during the subsequent recovery along an intermediate load hne vibrabon data were obtained. Similarly, during the single field breaker tnp and two recirculabon pump motor breaker trip and subsequent recovery along the 100% load hoe more data were obtained.
Finally, at selected power levels, feedwater sparger vibration data were obtained. All of the above data were determined to have satisfied the t evel 1 and 2 cnteria after a quick site analysis, and later, after a more detailed analysis by Component Engineenng in San Jose. It is expected that a topical report on Brunswick Unit 1 vibration testing will be issued some time after the startup testing period.
Aside from the normal testing. the vesselintemal vibrabon sensors were used to merutor for the onset of cavitation dunng STI 30 testng. This is not required by the test program but no cavitabon was detected in any case.
3.35 STI 35 RECIRCULATION SYSTEM FLOW Call 8 RAT 10N 3.35.1 Level 2 Critoria The jet pump flow instrumentanon was adpstod such that the jet pump total flow recorder provides the correct core flow indication at rated condtions.
The APRM/ ROM flow-bias instrumentabon was adjusted to function property at rated condtions.
3.35.2 Discuselon The core and dnve flow instrumentation was cabbrated at a nominal 100% core flow at Test Condtions 3 and 6. Rated drive flow (that ftow which gives 100% core flow at 100% power) was determined to be 43,174 gpm (each loop). This flow was obtained at ~89% recirculation pump speed and 52.16 ft of differential pressure across the drive flow nozzles. The flow convertera were adjusted Io produce 8 vofts (100%) total flow output into the flow biased scram and rod block circuitry with the above mentioned A and D loop inputs.
It should be noted that Brunswick Unit l's drive flows have been recahbrated t055 kgpm (from 70 kgpm) full scale flow The flow converter gains, plant meters and recorders, nozzle ,iP caibrabons and process computer calibrations were all adjusted to accommoote this change. The change seems desirable in that Crunswick is not Econsed to use its equalzer vahre and thus will not have need of the higher measurement capabilty.
The San Jose tirie share computer program "JR PUMP" was also used to calculate total core flow. Three sets of single tap and double tap jet pump diff(rential pressures (from 4 minute MAXMIN averaged STARTREC recordings) were used to generate 3 "JR PUMP
- runs. The 3 results (loop and core flows) were averaged, and the results used to adjust the loop flow proportional amphfiers and total flow proportional amphfier.
The computer output was also used to determine the double tap let pump offerental pressure transmitter full scale pressures (35.33 psi). The average M ratio was determined to be 1.34 as compared to the design value of 1.26. Jet pump riser and nozzle plugryno data were all acceptable. The CP&L l&C shop was notfled of those electronic components whose input / output relationships indcated that an adjustment was needed.
In order to use the most advantageous CPR calculabon in the process computer, each Bailey positioner mechanecal stop was set to a speed correspond ng io 102% core flow and each electrical stop set at a speed correspond ng io 101% core flow. The master flow crsntroller bmiter was set to dsallow any pump speed above 100% core flow.
Finalty. Table 3 35 2.1 itemizes the final drtvo flow / core flow correlation that was input to the process computor; this has allowed a WTFLAO - 2 (WTSUB within 5% of WT) to be obtained at any drive flow and with the reactor anywhere between the 70% and 100% load lines.
M0
NEDO-24662 TeWe 3.38J.1 '
DANE PLOW / CORE PLOW COfMIELATION WD path) 8 12 16 21 27 33 M8ue (Mh/M 24.se 33.12 40.4e 50 40 63.72 77.3o Mh 2.12 1.76 1.53 1.4 1.36 1.33 .
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- 4. DISTRIBUTION Name M/C J. Armenta (5) . . . . . . . . . . . . .. ....881 P. K. Bingham .. . . . - . . . . . . . . . . . . 3 94 A. P. Dray.. . . .. . . . . . . . . . . . . . . . ... .. 600 F. R. Channon... . . . . . . . . . . . , ,, ..,174 R. C. Christianson... . .. . . . . . . . ... .. ... .. 88 3 D. W. Defendorfor . . . . . . .. . . . . . . .. .... . .. . 8 83 D. C. Otmoro .. . . . . . . . . . . . ......150 E. C. Eckert.. . .... . . . . . .. . ... . . ... 7 63 N. L Folmus.. ........ . . . . ........................882 D. L. Rscher.. .. . . . ....... . . . . . . . .. .. . 733 International Uconsing c/o R. R. Aoof (5).. . . . . . . . . . . . . . .. . . . . . , . . . . 12 6 A. M. Jensen.. . . . .. . . .... ...195.#*
- F.D. Judge.... ... . . . . .. . . . . . . . . . . . . . . . . .. .. ... 150 J. D. Lambert.. . . .. .. . . . . . . . . . . .. . ..180 D. E. Lawlor., . . . . . .. . . . . .. . . . . . . . . . .. . . .. . . . . 18 4 S. L Mather., . . . . . . . . . . . . . . . . . . . . . . . . . . . ... . . 143 R. J. Pickenng (3) .. . .. .. . .........................171
- 1. D. Poppel . .. ... . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . 8 8 4 E. J. Romosberg .. . . . . . . . . . . . . . . . . . . . . . . . .. .. 885 G. C. Ross .. . .. . . . . . . . . . . . . . . . . . . . . .. .. . .. . ,8 93 G. J. Scatena . . .. . . . . . . . . . . . . . .. .. .. .... 763 S. l. Schreiner . .. ... .. . . . . . . . . . . . . . . . . . .. . 174 R. G. Sorenka . . . . . . . . .. . . . . . . . . . . . . . . . . . . . ..158 J.J.Sheehan.. ... ,. .. .... .. , . .... ......884 A. R. Smith . .. . . . . . . . . .. .. . . . . . . . . . . . . . . . . . . . . . . . .. 385 E. P. Stroupo... . . . . . .. . . . . . . . . . . . . . . . . . . . .. .146 R. D. Wilhams.. .. . . . ...........................197 P. J. Zimmerman (10) .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . .884 R. R. Zrubek.. . .. ... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ....884 NED Ubrary (5).. .. . .. . . ............................328 VNC Ubrary (2).. . . . . ... . .... . . . . . . . . . . . . ..V01 41/4.')
NUCLEAR ENERGY DIVISIONS
- GENERAL ELECTRIC COMPANY SAN JOSE, CALIFOMNIA 96t28 GENER AL $ ELECTRIC TECHNICAL INFORMATICN EXCHANGE l flTLE PAGE AUTHOR SUOJECT TIE NUMSER 77Nrh9dB
- 1. D. Poppel Startup Test Results DATE g,,
TITLE GECLASS !
Brunswick Unit 1 Startup Test Results i Final Summary Report GOVERNMENT CLASS REPRODUC18LE COPY PlLED AT TECHNICAL NUMSERDEPAGES SUPPORT SERVICES, MAUO SAN JOSE, CALIPORNI A 96128 (Mail Code 211)
SUMMARY
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This report is a summary of the results of all General Electric related startup tests performed at Unit 1 of the Brunswick Steam Electric Plant.
By cutting out this rectangle and folding in half, the above information can be fitted into a standard card flee.
DOCUMtNT NUM98M NEDO 24r02t INPORMATION PREPARE D POn Nuclear Energy Projects Division NCTION Mimitun Test Desion & Analys s t BUILDINO ANO MOOM NUM8EM 1887 2818 M AIL CODE 884 L
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