ML20210T457
| ML20210T457 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs, 05000000 |
| Issue date: | 08/29/1975 |
| From: | BALTIMORE GAS & ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20210T454 | List: |
| References | |
| FOIA-86-236 NUDOCS 8605300452 | |
| Download: ML20210T457 (167) | |
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i t 6 BALTIMORE GAS AND ELECTRIC COMPANY 4 CALVERT CLIFFS NUCLEAR POWER PLANT I UNIT I i i 1 1 Docket No. 50-317 License No. DPR-53 d STARTUP TEST REPORT i ! August 29, 1975 1 i l l l 1 l
. . . - . , . . . - ~ .
TABLE OF CONTENTS Section Paste
1.0 INTRODUCTION
AND .CUMMARY 1
1.1 INTRODUCTION
1 1.2
SUMMARY
2 1.2.1 INITIAL FUEL LOAD 2 1.2.2 POST CORE HOT FUNCTIONAL TESTS 2 1.2.3 INITIAL APPROACH TO CRITICALITY 2 1.2.h LOW POWER PHYSICS TESTS 3 1.2.5 ESCALATION TO POWER TESTS 3 1.2.6 EVALUATION OF STARTUP TEST PR00 RAM CRITICAL h PATH HOURS 2.0 INITIAL FUEL LOAD 7 3.0 POST CORE HOT FUNCTIONAL TESTS 18 3.1 CEDM/CEA PERFORMANCE TESTS 20 3.1.1 PURPOSE 20 3.1.2 TEST RESULTS 20 3.
1.3 CONCLUSION
S 21 3.2 REACTOR COOLANT SYSTEM FLOW TESTS 23 3.2.1 PURPOSE 23 3.2.2 TEST RESULTS 23 3.
2.3 CONCLUSION
S 23 3.3 REACTOR COOLANT SYSTEM FLOW COASTDOWN TESTS 25 3.3.1 PURPOSE 25 3.3.2 TEST RESULTS 25 3.
3.3 CONCLUSION
S 25 3.h PRIMARY AND SECONDARY VATER CHEMISTRY 27 3.k.1 PURPOSE 27 3.h.2 TEST RESULTS 27 3.h.3 CONCLUSIONS 27 3.5 PRESSURIZER EFFECTIVENESS TEST 28 3.5.1 PURPOSE 28 3.5.2 TEST RESULTS 28 3.
5.3 CONCLUSION
S 28 3.6 REACTOR COOLANT SYSTEM LEAK TEST 30 3.6.1 PURPOSE 30 3.6.2 TEST RESULTS 30 3.
6.3 CONCLUSION
S 30 h.0 INITIAL APPROACH TO CRITICALITY 31 5.O LOW POWER PHYSICS TES'"S 35 5.1 SHIELDING EFFECTIVENESS AND PLANT PE IATION LEVEL 39 MEASURDENTS i
Section P,a,Le, 5 1.1 PURPOSE 39 5.1.2 TEST RESULTS 39
51.3 CONCLUSION
S 39 EFFLUENT RADIATION MONITORS CALIBRATION h0 5.2 h0 5.2.1 PURPOSE h0 5.2.2 TEST RESULTS h0 5.
2.3 CONCLUSION
S h1 5.3 CRITICAL BORON CONCENTRATION MEASUREMENTS 5.3.1 PURPOSE h1 5.3.2 TEST RESULTS hl 5.
3.3 CONCLUSION
S hl h2 5.h TEMPERATURE COEPFICIENT OF REACTIVITY MEASUREMENTS h2 5.h.1 PUT. POSE h2 5.h.2 TEST RESULTS 5.h.3 CONCLUSIONS h2 ' hk 5.5 NON-0VERLAPPED REGULATING AND SRUTDOWN CEA GROUP WORTH MEASUREMENTS 5.5.1 PURPOSE hk 552 TEST RESULTS kk 5.5 3 CONCLUSIONS hk 5.6 OVERLAPPED REGULATING CEA GROUP WORTH MEASUREMENTS h7 5.6.1 PURPOSE h7 5.6.2 TEST RESULTS h7 5.
6.3 CONCLUSION
S h7 5.7 PRESSURE COEFPICIENT OP REACTIVITY MEASUREMENTS kB 5.7.1 PURPOSE kB 5.7.2 TEST RESULTS h8
.5.
7.3 CONCLUSION
S k8 5.8 DROPPED CEA WORTH MEASUREMENTS k9 5.8.1 PURPOSE k9 5.8.2 TEST RESULTS h9 5.
8.3 CONCLUSION
S h9 5.9 EJECTED CEA WORTH MEASUREMENTS 50 5 9.1 PURPOSE 50 5.9 2 TEST RESULTS 50 5.9.2.1 FULL POWER CEA CONFIGURATION 50 5 9 2.2 ZERO POWER CEA CONFIGURATION 50 5
9.3 CONCLUSION
S 50 5 10 STUCK CEA WORTH MEASUREMENT 52 5.10.1 PURPOSE 52 5.10.2 TEST RESULTS 52 5.
10.3 CONCLUSION
S' 52 5.11 PART LENGTH CEA GROUP MEASUREMENTS 53 5 11.1 PURPOSE 53 5 11.2 TEST RESULTS 53 5.
11.3 CONCLUSION
S 53 5 12 CRITICAL BORON CONCENTRATION AND SOLUBLE BORON WORTH Sh MEASUREMENTS 5.12.1 PURPOSE Sh 5.12.2 TEST RESULTS Sk 5.
12.3 CONCLUSION
S 55 11
Section Pagg, CH NICAL AND RADI0 CHEMICAL TESTS 57 5.13 57 5.13.1 PURPOSE 57 5.13.2 TEST RESULTS 57 5.13.2.1 BASE LINE CORROSION STUDY 58 5.13.2.2 FISSION AND ACTIVATION PRODUCT BUILDUP STUDY LITHIUM BUILLUP 59 5.13.2.3 PURIFICATION SYSTEM D. F. 59 5 13.2.L 59 5.13.2.5 VOLUME CONTROL TANK GAS 59 5.13.3 RADI0 CHEMISTRY RESULTS DURING EVAWATION AND RESOLUTION OF REACTOR VESSEL PRESSURE DIrruearlAL INCREASE INCIDENT 60 5.13.3.1 pH INCREASE TESTS 60 5 13.3.2 pH DECREASE TESTS 60 5 13.3.3 STARTUP DATA FOLLOWING REACTOR VESSEL DIF-FERENTIAL PRESSURE INVESTIGATION 513.3 k SUBSEQUENT SHUTDOWN INFORMATION 61 5 13.3-5 SUBSEQUENT STARTUP INFORMATION 61 5 13.h CONCLUSIONS 61 6.0 ESCALATION TO POWER TESTS 65 6.1 TURBINE GENERATOR STARTUP AND ATMOSPHERIC DUMP / 67 TURBINE BYPASS VALVES TEST 6.1.1 PURPOSE 67 6.1.2 TEST RESULTS 67 6.
1.3 CONCLUSION
S 68 6.2 REACTIVITY COEFFICIENT fEASUREMENTS 69 6.2.1 PURPOSE 69 6.2.2 TEST RESULTS 69 6.2.3 CONCW SIONS 69 6.3 PLANT POWER CALIBRATION 71 6.3.1 PURPOSE 71 6.3.2 TEST RESULTS 71 6.3.? CONCLUSIONS 71 6.h SHIELDING rzrrunvt.:o5S AND PLANT RADIATION LEVELS 72 6.h.1 PURPOSE 72 6.h.2 TEST RESULTS 72 6.h.3 CONCMSIONS 73 6.5 TURBINE RUNBACK / STEP LOAD CHARGE TEST 75 6.5.1 PURPOSE 75 6.5.2 TEST RESULTS 75 6.5.3 CONCESIONS 76 ) 6.6 DROPPED CEA. POWER ASYMMETRY AND AZIMUTHAL XENON 78 TRANSIElff TESTS l 6.6.1 PURPOSE 78 6.6.2 TEST RESULTS 78 6.
6.3 CONCLUSION
S 78 6.7 TRIP OF MAIN GENERATCR BREAKER 80 6.7.1 PURPOSE 80 6.7.2 TEST RESULTS 80 6.
7.3 CONCLUSION
S 80 l t iii
Sectio. Face 81 6.8 xzNON FOLLOW MEASUREET2S 81 6.8.1 PURPOSE 81 6.8.2 TEST RESULTS 82 6.
8.3 CONCLUSION
S 83 6.9 REMOTE SHUTDOWN TEST 83 6.9 1 PURPOSE 8' 6.9 2 TEST RESULTS 6.9 3 CONCLUSIONS 8h 6.10 FEEDWATER REGULATING SYSTEM TEST 8h 6.10.1 PURPOSE 8h 6.10.2 TEST RESULTS 8h 6.
10.3 CONCLUSION
S 6.H LOSS OF OFFSITE POWER WITH COASTDOWN 85 6.n.1 PURPOSE 85
- 6. H . 2 TEST RESULTS 85 6.n.3 CONCLUSIONS 86 6.12 CORE POWER DISTRIBUTIONS 87 6.12.1 PURPOSE 87 6.12.2 TEST RESULTS 87 6.12.2.1 RADIAL POWER DISTRIBUTIONS 87 6.12.2.2 AXIAL POWER DISTRIBUTIONS 88 6.12.2.3 PEAK LER 88 6.12.2.h DNBR 88 6.12.3 CONCMSIONS 88 6.13 PSUEDO FJECTED CEA POWER DISTRIBUTION ymrlMENT 90 6.13.1 PURPOSE 90
.6.13.2 TEST RESULTS 90 6.
13.3 CONCLUSION
S 91 6.1h PARTIAL LOSS OF FLOW TRIP 92 6.1h.1 PURPOSE 92 6.1h.2 TEST RESULTS 92 6.1h.3 CONCW SIONS 92 6.15 TOTAL LOSS OF FLOW / NATURAL CIRCULATION TEST 93 6.15.1 PURPOSE 93 6.15.2 TEST RESULTS 93 6.15.3 CONCWSIONS 93 6.16 CHEMISTRY AND RADIOCHD'.ISTRY TESTS AT POWER 9h 6.16.1 PURPOSE 9h 6.16.E TEST RESULTS 9h 6.16.3 CONCMSIONS 95 6.17 EFFLUENT MONITORING SYSTEM TESTS 97 6.17.1 PURPOSE 97 6.17 2 TEST RESULTS 97 6.17.3 CONCWSIONS 97 6.18 PROCESS COMPUTER MEASURED VALUES CHECK 100 6.18.1 PURPOSE 100 6.18.2 TESP RESULTS 100 6.18.3 CONCMSIONS 100 iv
Page Section UNUSUAL EVENTS 101 70 101 7.1 REACTOR VESSEL DIFFERENTIAL PRESSURE INCREASE 7.2 UNCOUPLED CEA INCIDENT 103 10h 7.3 HIGHER THAN PREDICTED CONTAINMENT RADIATION LEVELS WATER HAMMER IN MAIN FEEDWATER Lirim 106 7.h Y
l 1
1.0 INTRODUCTION
AND
SUMMARY
f 1.1 Introduction This report fulfills the requirement of Technical Specification 6.12.1 which states that a Startup Test Report vill be sulmitted to NRC within 90 days of completion of the Startup Test Program. The test program was completed on 5/30/75 The Startup Test Program was organized and administered by Baltimore Gas and Electric Company (BG&E) personnel assisted by Combustion Engineering (CE) Startup Engineers on-site and home office personnel in Windsor, Connecticut (CE, Windsor). Tite Startup Test Program consisted of several phases. The test results from each phase were reviewed by a Test Data Evaluation Group (TDEG) consisting of the BG&E Startup Test Coordinator and the CE Chief Test Engineer and others as required. Test results falling outside of acceptance criteria received an additional review by the Plant Operations and Safety Review Committee (POSRC) and were resolved prior to beginning the next test phase. The test phases were as follows: (1) Initial Fuel Load (2) Post Core Hot Functional Tests (3) Initial Approach to Criticality (h) Lov Power Physics Tests (5) Escalation to Power Tests - 20% Plateau (6) Escalation to Power Tests - 50% Plateau (7) Escalation to Power Tests - 80% Plateau (8) Escalation to Power Tests - 100% Plateau Maximus licensed reactor core power level (100%) is 2560 MWth. The Startup Test Program began at 2100 on 8/h/Th with the loading of the first fuel assembly into the reactor vessel and was ccmpleted at 1600 on 5/30/75. The test program was rather extensive and the results vill form the basis and justification for significantly reducing the required testing on Calvert Cliffs Unit 2. The design of most safety related systems on Unit 2 are very similar and generally identical to Unit 1.
Page 2 , 4 1 1 l i 1.2 Summe y 1.2.1 Initial Fuel Load Fuel loading consnenced on 8/k/Th and was completed on 8/16/Th. Approximately 25% (see Section 1.2.6) of this ' time was spent in the non-productive activities of fuel
- handling equijutent maintenance, hand manipulation of fuel assemblies, and inspection of fuel assemblies which gave momentary indications of hanging up on adjacent fuel assemblies while being lowered into the core.
None of the inspected fuel assemblies showed visible signs of damage. An evaluation by CE, Windsor, of the mechanical forces involved in the several incidents ' of fuel assembly hangup indicated that observed forces vere below those required to cause damage. Fuel handlers worked three (3) eight (8) hour shifts per day and an experienced crew could load twelve (12) to fourteen (lk) fuel assemblies per shift providing there were no maintenance problems. 1.2.2 Post Core Hot Functional Tests i Post Core Hot Functional (PCHF) Tests commenced 9/3/Th and were completed on 10/5/Th. In addition to those tests required by the FSAR, performance testing of the turbine bypass valves, auxiliary feedvater pumps, and secondary safety valves was also performed during this phase. All test results met acceptance criteria with the exception of Reactor Coolant System (RCS) flow rate for the Reactor Coolant Pump (RCP) 11A and 11B combi-nation and the performance of Control Element Drive Mechanism (CEDM) 15. RCP 11A and 11B combination yielded a flow rate slightly below that required by present Technical Specifications. That combination is presently administratively restricted from use during power oper-ation. CEDM 15 is a Part Length Control Element Assembly (CEA) in Part Length Group 2 (PLR 2). Performance testing revealed a defective upper gripper coil and/or latch in the CEDM. PLR 2 is presently fully withdrawn and admini-stratively restricted front use while the reactor is critical. 1.2.3 Initial Approach to Criticality The Initial Approach to Criticality cosmienced at 2130 on 10/5/Th and the reactor was declared critical approxi-mately forty-eight (h8) hours later. The approach was relatively uneventful with the exception of scate minor CEA Group interlock malfunctions which occurred during the CEA withdrawal sequence. A slow RCS dilution followed CEA withdrawal. Measured RCS soluble boron concentration at the time of criticality was in good agreement with that which was predicted and well within the acceptance criteria.
Page 3 l 1.2.h Low Power Physics Tests The Iov Power Physics Test (LPPT) Phase commenced on 10/7/7h and was finally completed on 12/27/7k. This ! phase was marred by tvo (2) major incidents which required approximately eight (8) veeks of non-test time to resolve. The incidents are discussed in more detail in Section 7 0. All LPPT measurements were in good agreement with predic-tions and well within acceptance limits.
- 1.2 5 Escalation to Power Tests A
The Escalation to Power Test (EPT) Phase began 12/27/7h. All required test measurements were performed and results reviewed and evaluated on-site by the TDEG with assis-I tance as necessary from CE, Windsor with the following exceptions: (1) A test of axial xenon oscillation dampening using 4 Part Length CEA's (PLR) was deferred until such time as a decision is made to use PLR's. Later in first cycle life, an axial xenon oscillation dam-pening test vill be performed using full length CEA's. (2) A part loop operation power distribution measurement was deferred until such time as a decision is made to operate at power using less than four (h) Reactor Coolant Pumps (RCP). Administrative restrictions prevent power operation with less than full RCS flov (h RCP's). In addition, the Reactor Protective System (RPS) vill cause a reactor trip whenever ) reactor power is greater than 10 h% and less than h RCP's are operating. (3) A performance test of the automatic CEA control l features of the Reactor Regulating System (RRS) { has been deferred until such time as a decision to I use this feature is made. The principal reason l for deferring the test was to decrease the possi- ) bility of fuel failures which could be exacerbated J by relatively rapid and near continuous CEA motion, a characteristic of automatic CEA control. Administrative restrictions prevent part loop operation and use of automatic CEA control and PLR's until such time as tests are perfomed and results reviewed and approved by the POSRC. The off-site analysis by CE, Windsor of two (2) tests, Psuedo Ejected CEA and Dropped CEA continues. A pre-liminary on-site analysis of the results of the Psuedo Ejected CEA test by the TDEG indicated that those test results were well within the power peaking acceptance criteria and on that basis, full power operation was
P:ge h continued. However, the acceptance criteria also require conpletion of a more detailed analysis by CE, Windsor. The evaluation of the dropped CEA test results by the TDEG was inconclusive. In addition, acceptance criteria require the completion of analysis by CE, Windsor. Should a dropped CEA occur during power operation, plant oper- l ating procedures specify that the reactor vill be manually tripped if at greater than 50% power. The off-site analysis vill determine the power level to which the turbine generator should runback in response to a dropped CEA. Onthat basis, the requirement to manually trip the reactor may be removed from the plant operating procedures. Results of the Shielding Survey indicated higher than ' expected radiation levels in the containment. Investigation and evaluation revealed that the primary source was gamma and neutrons streaming out of the annulus between the reactor vessel and the primary shield wall. Temporary shielding has reduced dose rates to acceptable levels (see Section 7.0 for more detail). Planning and engineering for a permanent shield design continues. Following en unplanned trip from 100% power, the feedvater rings for ooth steam generators were uncovered, as is the usual case. While in the process of regaining nonnal steam generator water levels, a water hammer was experienced in the =ain feedvater lines. Some damage to pipe supports and to motor operated stop valves resulted. Investigation and evaluation indicated that filling at a slover rate when level was in the vicinity of the feedvater ring precluded water hammer. Feedvater ring design changes are also being considered. See Section 7.0 for more detail. These and other minor equipment problems resulted in forced outages which delayed completion of EPT by about twelve (12) l veeks. Figure 1.2-1 presents an as originally planned and j as experienced power history for EPT. EPT and the Startup j Test Program were concluded on 5/30/75 Unit 1 had beca previously declared in ecumercial operation on 5/8/75. 1.2.6 Evaluation of Startup Test Program Critical Path Hours As a guide to future planning, particularly in connection with the Initial Startup of Unit 2, a detailed evaluation was made of the as experienced Unit 1 Startup Test Program critical path. The results of that analysis are susmarized in Table 1.2-1. In order to assist the reader in an evalu-ation of the significance of the various numbers in Table 1.2-1, the following explanatory remarks are made. Each chronological hour from 2100 on 8/h/Th to 1600 on 5/30/75 was investigated to determine which plant activity was controlling the critical path for completion of the Startup Test Program. Control room logs and the test
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1 P:ge 5 organization's records w the test program evolvedmaticn and the ev - erformed in increments as , Each critical path hour was as i activities: i startup testing s gned to one of four h) test and operationsmaintenance; or secondary s a ntenance. productive and non-productive pleting scheduled categories. test program med in the Productive hours no t non-productive startup test hourevolutions. Examplesprocess of of c repeating test measurements invalid; repeating test mea ously determined to be previs would be those before being interrupted by surements not fully developed and repeating originally. Examples test measurements a forced maintenance activity; vould be those required: not properly performed of non-productive operations s hour to return the plant to the lconditionerfor cond to put the plant in ced maintenance activity; the prop i resulting in loss of proper testafter a for ast previous test condition v ty; and operator error nance hours exception of post fuel loading were considered conditions. to b All mainte-e non-productive with the In asome for critical path cases hourseveral reactoractiviti vessel reassembly. and it was divided among them are provided forThe original e uled estimates of sch d . included some allowance fcomparison purposes. critical path ho l When compared with the actor minor forced mainten the non-productive, the esti . in the case of mostual every productive test phase hours, inclu 4 f a
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TABLE 1.2-1 START-UP TEST PROGRAM
SUMMARY
OF CRITICAL PATH IIOURS CRITICAL PATH HOURS START-UP TEST OPERATIONS MAINT TOTAL POWER NON- NON- NON- ORIG. RV DATE TEST PHASE LEVEL PROD. PROD. PROD. PROD. PRI. SEC. PROD. PROD. TOTAL EST. REASS. 8/Is-16/7h Fuel Load RSD (1) 21 5 20.25 108.5 80.5 46.25 - 130 1h7 277 168 - 8/16-9/3/71: RV Reass. RSD (1) - - 2h - 410 (3) - 13h - 3:34 336 - 9/3-10/5/74 PCHF HSD (2) 310 19 5 70 138.5 217.5 23 380 398.5 777.5 336 - 10/5-7/74 IAC O 41 5 - - - 8 - 41.5 8 h9 5 40 -
! 10/7-12/27/74 LPPT 0 269 108 - h2h 1089 h3 269 1661 1993 336 -
12/27/718-1/17/75 EPT 20% 11 9 171.5 66 - 109 h9 114.5 171 5 338.5 510 312 - 1/17-3/11s/75 EPT 50% 26.4 36h 315 - 72 34.5 585 5 36h 970 133h 408 - 3/14 18/3/75 EPT 80% 59.6 333 5 7 58 h5 5 - h7 391.5 99 5 h91 1:32 - h/3-5/30/75 EPT 100% 51 5 449 22 170 291 99 33h 619 746 1365 600 - t , 8/4/74-5/30/75 TOTAL 37.8 ( j, 1960 558 1:31 1161 15h3 1110 2801 h370 7171 2968 410 PERCElff 27 8 6 16 22 15 39 61 100 6 4 35 22 37 100 6 y n (1) RSD m Refueling Shutdown * (2) HSD z Hot Shutdown (3) Reactor Vessel Reassembly is figured into the total as productive critical path hours. (k) During EPT
C A L V E R T C L I F F S U N I T I ORIGINALLY PLANNED & ACTUAL POnER HISTORY FOR EPT ies , , ee $ I II
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PGge 7 a 2.0 INITIAL FUEL LOAD Fuel loading commenced on 8/h/Th and was completed on 8/16/Th. Table . 2.0-1 and Figure 2.0-1 give the fuel loading sequence for initial core , load. Figures 2.0-2 and 2.0-3 show fuel assembly core location by serial number and CEA core location by serial number respectively. Neutron count rate was monitored during core assembly on four separate detector channels. In addition to Wide Range Log Channels 1 and 3, two temporary neutron detectors were used. The temporary neutron detecters were installed prior to core load with detector A installed in core location V-7 and detector B installed in core location V-15 In Stcp 3h0 B of the loading sequence detector B was shifted to core locatica 3-15. Initial locations and movement step numbers are shown in Fic".re 2.0 k. Independent plots of inverse neutron count rate versus the number of fuel assemblies loaded were maintained to ensure that the reacter remained suberitical at all times. , 4 The fuel loading was conducted with the refueling and spent fuel pools dry and borated water in the reactor vessel to within approximately a eight (8) inches of the top of the reactor vessel outlet nozzles. This dry loading allowed access to refueling equipment for maintenance, , improved visibility of fuel assemblies during manipulation and allowed access to the reactor vessel flange during repositioning of temporary neutron detectors. l Several problems with the fuel handling equipment occurred during fuel l load. A description of these problems and their resolution are given i below: Fuel Transfer Machine and Upender i (1) On several occasions lights and interlocks on the transfer machine - carriage and upender would not function properly. Filling with t water and venting the limit switches restored the interlocks and lights to proper operation. (2) An open phase on the fuel transfer machine brakes and two faulty relays caused two separate delays. Replacement of the parts re-stcred the transfer machine to service. Spent Fuel Handling Machine The spent fuel handling machine festoened cable hung up. A shim was ' placed in a cable hanger to eliminate drag. ; Refueling Machine (1) On several occasions the hoist on the refueling machine hung up. l The problem was eventually traced to a bushing binding up. Initially ' the bushing was cleaned and extra lubricant added to minimize the problem. Two days later, the drive shaft and a bushing of different material were installed and fuel loading continued. ,
--r,- - . - , , . , - , - - - . . ,, . . , - , _ - - - . - , , - < , - - . , ..- t q
Page 8 (2) The hoist position indication malfunctioned. The hoist position read out device was replaced to correct the problem. (3) The svitch controlling the refueling machine bridge would only move the bridge in the reverse direction about half the time. Freeing up the jammed finger switch and lubricating the gears returned the machine to normal operation. Fuel Insertion Problems Fuel insertion problems vere experienced with fourteen (lh) fuel assemblies resulting in approximately lh hours of non-productive time. The insertion problems were generally an interaction between grids of the fuel assembly being inserted and grids of adjacent fuel assemblies. These problems were resolved as follows: (1) An observer was stationed at the upender to ensure that the fuel assembly orientation was parallel to the sides of the upender as it was lifted by the refueling machine. (2) After several assemblies were removed and inspected and no damage was found, adequate insertion clearance was obtained by making em11 changes to the refueling machine position index, pushing on the refueling machine spreader or pushing on the upper end fittings of adjacent fuel assemblies with a long handled pole as required. Modifications to fuel handling equipment are being made to mini =1::e these problems in the future. Except for the maintenance problems with the refueling equipment, the fuel loading was conducted in an efficient manner. l l
Paga 9 TABLE 2.0-1 FUEL ASSEMBLY LOADING SEQUENCE STEP NO. FUEL ASSE4BLY NO. CEA NO, CORE LOCATION 1 1B015 (Neutron Source) X - 11 2 1C035 - Y - 10 3 1C0h0 - Y - 12 h IC206 OJ X - 13 5 1B067 - W - 13 6 1A017 3A W - 11 7 13019 - W-9 8 1C203 OF X-9 9 1C003 - Y-8 10 1C201 - X-7 11 1A061 2D W-7 12' 1A01h 02 V-9 13 1B028 - V - 11 1h 1A026 15 V - 13 15 1A031 OK V - 15 16 1C210 - X - 15 17 1C015 - Y - ik 18 1C030 23 I - 16 19 13063 - W - 16 20 1A058 OG V - 16 21 1B025 - T - 16 22 1A0k0 12 T - 15 23 1B010 - T - 13 2h 1A055 k0 T - 11 25 13023 - T-9 26 1A027 05 T-7
i i Pase 10 ; i TABLE 2.0-1 (Cont'd) , 1 STEP No. _FGEL 1.5EEMBLY Nd, CF). NO, _ CORE LOCATION 27 13006 - T - 16 . 28 1A025 1H V-6 29 130k9 - W-6 30 100g0 10 X-6 31 1c011 - %-5 32 1C102 2C W-5 33 1B071 - V-5 3h 1A063 - T-5 35 15003 - 3-5 36 IA052 h3 s-6 37 13602 - S-7 3S 2A050 -- S-9 39 12038 - S-n l LO 1Ach9 - S - 13 41 1ET03 Surve111snee S - 1$ Holder Center Guide Tube h2 2>033 k5 s - 16 h3 1B965 - s - 17 hk 1A019 - T - 17 45 1303h - V - 17 h6 1C103 1h W - 17 47 1C009 - x - 17 k8 1A028 CE R-17 h9 13058 - R - 16 50 1A0hk 29 R - 15 51 1B008 - R - 13 52 1A0h8 24 R-U
Pcga 11 TABLE 2.0-1 (Cont'd) STEP No. , FUEL ASSDGLY N0, CEA NO. CORE LOCATION 53 13057 - R-9 Sk M0k6 2F R-7
$5 13001 - R-6 56 1A069 09 R-5 ,
57 1B037 - N-5 58 1A009 - K-6 59 1B030 - J-7 60 1A06h OD N-9 I 61 13061 - N - 11 62 1A0k5 la N - 13 63 13033 - N - 15 6h 1Ad62 a N-16 65 1B068 - N - 17 66 1A054 kl L - 17 67 13029 - L - 16 68 1Aok7 18 L - 15 69 13036 - Le 13 70 1A016 1C L - 11 71 1B077 - L-9 72 1A0h3 1A L-T 4 73 13009 - L-6 Th 1A056 3J L-5 75 1B066 - J-5 76 1A065 - J-6 77 1B032 - J-7 78 1A003 1F J-9
Page 12 TABLE 2.0-1 (Cont'd) STEP NO. M EL ASSIMBLY NO. CEA NO. CORE LOCATION 79 1B018 - J - 11 80 1A0h2 39 J - 13 81 1B00h - J - 15 82 1A038 - J - 16 83 1306h - J - 17 8h 1A066 2J G - 17 85 130h4 - G - 16 86 1A001 2A G - 15 87 13059 - G - 13 88 1A002 09 G - 11 89 1B055 - G-9 90 1A015 37 G-7 91 1BT01 Surveillance G-6 Holder Center Guide Tube 92 1A057 2B G-5 93 1B011 - F-5 9h 1A068 h2 F-6 95 1B02h - F-7 96 1A00h - F-9 97 1B0h5 - F - 11 98 1A051 - F - 13 99 1BT02 Serveillance F - 15 Holder Center Guide Tube 100 1A053 kk F - 16 101 1B051 - F - 17 102 1A060 - E - 17 103 1B040 - E - 16
- - . . ~ . . - . - - . . . - . . ~ -
Page 13 TABLE 2.0-1 (Cont'd) STEP NO. FUEL ASSD(BLY NO. CEA NO. CORE LOCATION 10h 1A029 01 E - 15 105 1B026 - E - 13 106 1A005 3K E - 11 107 1B012 - E-9 108 1A039 21 E-7 109 1B007 - E-6 110 1A010 - E-5 111 1B052 - D-5 112 1A02h 3h D-6 113 1Boh6 - D-7 11h 1A021 36 D-9 115 1B035 - D - 11 116 1A059 2E D - 13 117 1A067 OH D - 16 n8 13070 - D - 17 119 1C105 2G C - 17 120 1B031 - C - 16 121 1A030 OA C - 15 , 1 122 130h8 - C - 13 123 1A036 3B C-n j 12h 1B0h3 - C-9 125 1A008 33 C-7 126 1B016 - C-6 127 1C111 2H C-5 128 1CO32 - B-5 129 1CO28 25 B-6 130 1C211 - B-7
Paga ik TABLE 2.0-1 (Cont'd) PUEL ASSEMBLY NO. CEA NO. CORE LOCATION STEP NO. 131 1C213 1G B-9 132 13020 Neutron Source B - 11 133 1C209 38 B - 13 13h 1C205 - B - 15 135 1C03h 03 B - 16 136 1CO36 - B - 17 137 1C038 - A - 1h 138 1C01h - A - 12 139 1C002 - A - 10 1ho 1C006 - A-8 1kOB Detector B D - 15 1hi 1C060 - V - 15 1h2 1CO2h - W-h ih3 1C10h 3C V-h 1hh 1B0h7 - T-k 1h5 1A013 1D S-h ih6 130h2 - R-4 1kT 1A023 08 Nk 1h8 13050 - Lh 1h9 1A0h1 35 J-4 150 1B075 - G-h 151 1A012 11 Fk l
\
152 1B079 - Ek 153 1C106 3D Dh 15h IC017 - C-k 155 1CO22 - D-3 I
\
Pa6e 15 i TA2LE 2.0-1 { Cont'd) ] STEP NO. FUEL ASSEMBLY 50., CIA NO. CORZ 10 CATION 156 1C107 1K E-3 157 13013 - F-3 158 1A007 22 0-3 159 1B078 - J-3 160 1A016 3E 'L - 3 161 1B080 - Ne3 162 1A03h 1E R-3 163 12072 - S"3 16h IC101 ,16 T-3 165 1C021 - V-3 166 1C018 - T;- 2 167 10023 E6 S-2 168 1C20k - R-2 169 1C215 17 N-2 l 170 1E022 - L .2 171 1C203 19 J*2 172 1C202 - D-2 173 1C02T 31 F-2 174 1C033 - J-2 175 1C007 - H-1 176 1C001 - K-1 177 1C013 - M-1 1 178 1C039 - F-1 t 179 1C029 - V - 18 180 1C109 3F Y - 16 181 18069 - T e 18 182 M035 27 s e IB
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Paga 16 TABLE 2.0-1 (Cont'd) STEP NO. FUEL ASS 1HBLY NO, CEA NO. CORE LOCATION 183 13039 - R-18 1Bk 1A020 Oh N - 18 185 13056 - L - 18 186 uo37 IJ J - 18 187 1B027 - G - 18 188 3022 20 F - 18 189 11b17 a E - 18 190 10110 3G D - 18 J91 1 con 6 - c - 18 192 1C016 - D - 19 193 IQ106 2K 3:- 19 ' 3.9h 13074 - F - 19 . 155 Mon 30 0 - 19 , 196 .13062 - J,- 19 197 uo06 o 3R L - 19 198 1D005 - N - 19 199 M032 28 R - 19 i , l 200 1B053 - S - 19 . 201 1C112 13 7 , 19 ' 2C2 10025 - Y - 19 203 10012 - T - 20 i , 20h 10905 -32 S - 20 l 1 005 10212 - 3 - 20 l 2)G ICDG'l 07 1 - 20 207 ra021 - I.- 20 . 206 idaik 03 J e ac l
\
Pega 17 TABLE 2.0-1 (Cont'd) l l STEP NO. F'JEL ASS 1!NBLY NO. CEA NO. CORE LOCATION j 209 1C216 - G - 20 210 1CO31 06 F - 20 l 211 10019 - E - 20 212 1C008 - H - 21 213 1C037 - K - 21 21h 1C00h - M - 21 1 215 1C010 - P - 21 l 215B Temporary - Remove from Core Detector B 216 13041 - D - 15 216A Temporary - Remove from Core Detector A 217 1301h - V-7 4 l l l l I n ,
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E ( [ R N L J G CALVERT CLIFFS UNIT 1
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7 "- -' >- ' - - BOL, 1st CYCLE 20 202 204 205 206 207 208 209 210 211 20 D l W V , , , s,, ..,.. .. ..,,. .,.,r. ,.. .c .. ,o 1 19 202 201 200 199 198 197 196 195 194 193 192 19 f w . . , ... .... .... .... ,, ... .... .... .... ..., c 18 179 ISO 181 182 183 184 185 186 187 188 189 190 191 la l a x .... ... .... .... .... .... ... .... ... .... 17 47 48 45 44 43 48 65 66 83 84 101 102 118 119 136 17 18 19 20 21 42 49 84 67 82 85 100 103 11 7 I20 135 16 16
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I i I 6 13 24 39 52 61 10 79 88 97 106 115 123 132 I 3 2 l39 10 2'1 12 25 38 53 60 71 78 89 96 107 114 124 131 9 I 9 8 7 l '*' 113 125 130 I i 7 10 Il 217 26 37 54 59 72 77 90 95 108 7 Y ., .c , . . , A l A 6 g 6 30 29 28 27 36 55 58 73 76 91 94 109 112 126 12g l 57 74 75 92 93 110 111 127 :gg 5 5 31 32 33 34 35 56 4 142 143 144 145 146 147 148 149 150 151 152 153 154 4
. . . . . . . . . . . . . . . . . . .. ,. .. . . . ... ... .c.
W C 3 165 164 183 162 161 160 159 158 157 156 155 3 l .<., ,, ... .. i v D l 2 IS6 167 166 169 170 171 11 2 173 174 2
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E [ R N L J G s CRVERT CLIFFS UNIT 1
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/ 2' C0:0 C004 C037 C000 2i FUEL ASSEMBLY SERI Ai NUMBERS Is .u a
" " " N4 ' ^) -S i s . , . ..,.__ ,. . . . r i AND CORE LOCATIONS yo C0:2 C005 C2:2 C202 B0ri Cri4 C ri6 C03: C0iG co BOL, 1st CYCLE W , , , s, ..,,, . , . . , . . , , . ,. . . . o (9 CO25 Cl12 8053A032 9005 A006 8002 A011 8074 C10B C016 s.3 w , , , ,,,, s., ., . . . , , , , . . . , ..,,, . . . , .,., .,,,, C 18 CO29 C109 3069 A035 8039 A020 B056 A037 B027 A022 3017 C 110 CO26 #8 8
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C03B 4 C208 3087 AB 26 B010 4049 B000 AS45 8036 A042 B059 4051 B026 A059 30 4B C209 13 t3 12 C040 C014 12 f i B015 A017 0020 A055 803B ADAB B061 AOl6 B018 A602 8045 4005 8035 A036 8020 3 to C035 COS2 to 9 --
" 1- C203 0819 A014 8023 4050 B0 57 A064 8877 A003 B055 A004 B012 A021 8043 C213 2'- '2- 9 C006 e 8 MB3 7
C20l A010 3014 A027 B001 A046 8030 A043 8032 4015 8024 A039 B0 46 A000 C211 7 Y ., . . , , , , .., . . , , . , . = , . , ., . . , . . , . . , . . , .c , . . , A A 6 CO20 5048 A025 8006 A052 800 4009 B009 A065 BT01 406B B007 A024 B016 C020 6 X
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W c 3 CO21 C101 B072 4034 B000 A01B 5078 A007 B013 C107 C022 3 2 ColB C023 C204 C215 B022 C208 C202 C027 C033 2
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R N L J 6 [ 2 P M K H CALVERT CLIFFS UNIT 1
's CONTROL ELEMENT ASSEMBLY zi
=" \ -s r s 2 r E SERIAL NUMBERS AND CORE N--(' . , . .. . , . . , . to 32 07 08 06 2o o LOCATIONS W v ... ., . . . ..,,, ..-. , , . <.,., ..,., . . , . 30 BOL, 1st CYCLE 19 13 28 3H 2K i9 w . . , .., .... ..., .... ., ... , . . ... .... . . . , c is 3F 27 04 11 20 3C is x ... .,.. .... ... .... . . . ... .... ... .... o i7 14 OE 41 21 2C ir is 23 SC 45 44 GH 03 is c
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is 1 i 34 40 24 1C 09 3K 38 i 3 9 GF 02 SD 1F 36 1C 9 7 2D 05 2F 1A 37 21 33 r Y A
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E 2 R p N u L g J G CALVERT CLIFFS UNIT 1 n
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TEMPORARY FUEL LOADING l'L ANT Nonn. N-- -- -S T s 28 28 r c DETECTOR LOCATION s s 20 2o AND MOVEMENT STEP NUMBER W v . . , . . , ,
-. , . . ..,, . , s. ,, . . . . . , . o BOL, 1st CYCLE 19 19 W , , , . . , ..., ..., . . . ..., ... ..., .. .... .... C k 18 18 x .... ... .... .... .... .... .... .... .... .... .... .... . . . 8 17 I7 16 16
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GETECTOR 0 FIGURE 2.0-4
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Page 18 1 3.0 POST CORE HOT FUNCTIONAL TESTS (PCHF) Several of the tests required prior to initial criticality require installation of the fuel and all reactor internals as a prerequisite. These tests (Post Core Hot Functional Tests) are conducted after initial fuel loading and complete the prerequisites for initial criticality. A list of the required tests follows: (1) Mechanical and instrument tests on Control Element Drive Mechanisms (CEDM) and Control Element Assembly (CEA) position indicators; (2) Reactor protective trip circuit and manual scram tests; (3) Rod drop time measurements (cold and hot) at rated Reactor Coolant System (RCS) flow and with no RCS flow; (h) Final leak tests of RCS; (5) Chemistry and radiochemistry tests for water quality; (6) RCS flow determination tests; (7) Pressurizer effectiveness tests; and (8) Neutron response check of source range monitors. l All of the above tests except items (2) and (8) are described in Sections 3.1 through 3.6. Item (2), Reactor protective trip circuit and manual scram tests, was performed as an initial condition to the Initial Approach to Criticality (IAC). Item (8), Neutron response check of source range monitors, was verified as an initial condition to the IAC Procedure by comparing as observed signal levels with those observed prior to fuel loading, after fuel loading, and during PCHF and noting no significant difference other than an expected increase due to RCS heatup during PCHF. In addition, those itests or systems which required maintenance or had testing deferred from pre-core hot functional tests were tested during PCHF. This included: (1) Turbine Bypass Valves test; I (2) Checking the Turbine Generator roll to marim m RPM; (3) Taking measurements of RCS expansion; (k) RCS and Steam Generator instrument calibration checks; (5) Auxiliary Feedvater System test; (6) RCS heat loss test; (7) In-Core Detector resistance readings; and (8) Secondary Safety Valves test.
Page 19 All test results met their acceptance criteria with the exception of RCS flow rate for the Reactor Coolant Pump (RCP) 11A and 11B combi-nation, the perforinance of CEDM 15, and proper operation of the Turbine Bypass Valves. RCP 11A and 11B combination yielded a flow rate slightly below that required by present Technical Specifications. That combi-nation is presently administrative 1y restricted from use during power operation. CEIN 15 is a Part Length CEA in Part Length Group 2 (PLR 2). Performance testing revealed a defective upper gripper coil and/or latch in the CEDM. PLR 2 is presently administrative 1y restricted from use while the reactor is critical. Discussion of problems with Turbine Bypass Valves is contained in Section 6.1. l l l
~~ - . .
Page 20 3.1 CEDM/CEA Performance Tests 3.1.1 Purpose The Control Element Drive Mechanism / Control Element Assembly DEDM/CEA) Performance Tests were performed to accomplish the following four objectives. (1) To ' 2onstrate proper functioning of the CEA's and CEDM's unde. various Reactor Coolant System (RCS) temperature, pressure, and flow conditions. (2) To provide measured CEA vithdrawal, insertion and drop time data which will serve as comparison standards for future performance tests. (3) To perform a check of the position indication system and to establish proper functioning of the CEA operating and interlock lights. s (h) To verify that all Regulating and Shutdown CEA's have drop simes for 90 percent insertion in accordance with Technical Specification 3.10. 3.1.2 Test Results The CEDM/CEA Performance Tests were conducted at RCS tem-peratures and pressures of 260'F/k50 psia and 532'F/2250 psia. The tests consisted of: (1) A measurement of coil resistance. (2) A check of CEDM/CEA vithdrawal speed. (3) Timing of rod drop from full out to 90 percent insertion of all regulating and shutdown CEA's with full RCS flow (h RCP's). (k) Ten (10) additional drops of the fastest and slovest CEA's. (5) Measurement of CEA withdrawal speed and rod drop time for all regulating and shutdown CEA's with zero RCS flow. Performed at RCS temperature and pressure of 532*F/2250 Psia only. (6) A check of all CEDM/CEA position indication, operating and interlock lights. CEDM withdrawal and insertion traces were analyzed and adjustments were made to Coil Power Programmers by CE. Windsor personnel during the test to ensure acceptable
Paga 21 I operation of the CEDM's. Discrepancies in rod position ; indication were corrected by adjusting ecumputer setpoints and programming and replacing defective reed switch stacks. The results of the CEDM speed and drop time test were found to be acceptable and are as follows: 260*F/h50 psia CEDM Speed Fastest single CEA - CEDM 37 - 32.79 inches / min. Slovest single CEA - CEDM 12 - 29.89 inches / min. Fastest dual CEA - CEDM 39 - 20.ho inches / min. Slowest dual CEA - CEDM 6 - 19.15 inches / min. Drop Time to 90% Full Insertion (2 RCP's) , Slovest single CEA - CEDM 59 - 2.23 see. Fastest single CEA - CEDM 3h - 2.00 see. Slovest dual CEA - CEDM 51 - 2.03 sec. Fastest dual CEA - CEDM 7 - 1.88 sec. 532'F/2250 psia CEDM Speed i Fastest single CEA - CEDM 35 - 32.66 inches / min. Slovest single CEA - CEDM 28. 58 - 31 52 inches / min. Fastest dual CEA - CEDM 39 - 20 56 inches / min. Slovest dual CEA - CEDM 8 - 19.8k inches / min. Drop Time to 90% Full Insertion (Full Flow) Slovest single CEA - CEDM 57 - 2.36 see. Fastest single CEA - CEDM 35 - 2.10 sec. Slowest dual CEA - CDM k8 - 2.22 sec. Fastest dual CEA - CEDM 42 - 2.11 sec. Drop Time to 90% Full Insertion (Zero Flov) Slovest single CEA - CEDM 21 - 2.06 see. Fastest single CEA - CEDM 5. 35 - 1 9k sec. i Slowest dual CEA - CEDM k9. 51 - 1.88 see. Fastest dual CEA - CEDM 38 - 1 78 sec. 3.1.3 Conclusions CEDM 15 has an apparent defective upper gripper coil or latch and was not fully tested. It was placed at the upper electrical limit (fully withdrawn) after initial
Pcgs 22 testing and deenergized. Part Length CEA's (PLR) including CEDM 15 vill be left at the upper electrical limit as pre-sent operating plans do not envision use of the PLR's. Operating procedures prevent use of PLR's during power operation. All other test results were evaluated to be acceptable. l J
Paga 23 3.2 Reactor Coolant System Flow Tests ) 3.2.1 Purpose The test was conducted to tetermine Reactor Coolant System (RCS) flow rates and pressure drops around the reactor coolant loops for various Reactor Coolant Pump (RCP) combinations. 3.2.2 Test Results RCS flow measurements were taken at RCS temperatures and pressures of 260'F/450 psia and 532*F/2250 Psia using permanently installed and temporary instrumentation. At 260'F/450 psia flow data was collected for one (1) and two (2) RCP ceabinations only. Measurements at 532*F/ 2250 psia were taken for the RCP combinations listed in Table 3.2-1. 3.2.3 conclusions Flov vith two (2) RCP's in the same loop (RCP's 11A and 113) was measured to be 185.255 spn. This is 1 5% below the 49.7% of full flow specified in Technical Specification 3.1. Until such time as a part loop operation test is satisfactorily performed to determine power distributions for less than four (k) RCP operation and results are approved, administrative restrictions on plant operations prevent power operation with less than four (4) RCP's. All other RCP combinations meet acceptance criteria and ex-ceed Technical Specification requirements for minimum flow. O
Page 2h TABLE 3.2-1 ; RZACTOR COOLANT PUMP FLOW DATA (1) RCP FLOW (GPM)(3) TOTAL FLOW % of MIN.
- RCP's RUNNING 11A 11B 12A 12B (GPM) TOTAL FLOW I2) 11A, H B, 12A, 12B 96,939 99,656 10h,035 103,239 403,871 105 06 HA, H B, 12A 102,279 105.9k9 135,130 -37,950 305.k08 79.k5 11A, H B, 123 102,850 105,150 -37,150 133,150 304,000 79 08 nA, 12A, 12B 129,052 -37,200 109,5k5 108,829 310,226 80.70 nB, 12A, 123 -36,700 129,k00 109,750 108,050 310,500 80.TT 11A, HB 107,2h3 no,012 -16,750 -15,250 185,255 48.19 12A, 12B -15,250 -17,000 H3 k00 n2,250 193,hoo 50.31 i 11A, 12A 133,2ko -31,250 139,953 -32,750 209.193 Sk.k2 HAo 123 13k,600 -31,500 -30,750 138,100 210,k50 54.Tk nB,12A -33,900 13k,937 1h0,817 -33,300 208,55h 54.25 llB, 12B -33,000 133,600 -31,350 137,250 206,500 53.72 ,
11A 136,9hT -27.750 -7,000 -6,750 95,hh7 2k.83 , nB -29,500 137,222 -7,500 -5,750 9h,hT2 2k.57 12A -8,800 -10,000 1kk,056 -29,650 95,606 2k.8T 12B -8,750 -9,000 -27,850 1ko.550 9k.950 24.70 (1) RCS temperature and pressure of 532'F/2250 psia (2) Loss of Coolant Safety Analysis assumes a single RCP flow of 92,500 gInt. . Due to measurement uncertainties a flow of 96,100 gym must be measured l to assure that the 92.500 spa mini == flow used in safety Analysis is met. Total measured minimum. flow is therefore 384,400 gym. l (3) A minus sign in front of a flow value indicates reverse flow through RCP; i_e_, short circuiting core.
Paga 25 ; 3.3 Reactor Coolant System Flow Coastdown Tests 3.3.1 Purpose The purpose of this test was to determine the Reactor Coolant System (RCS) flow ecastdown characteristics by tripping Reactor Coolant Pumps (RCP) in various combinations. 3.3.2 Test Results . Testing was conducted at RCS conditions of 532*F/2250 psia. Recorder traces of RCS flow versus time after trip were made of each RCP combination listed in Table 3 3-1. Comparison of recorder traces for the four (h) RCP trips with predictions revealed that the measured coastdown curve was less conser-vative than prediction based on flov fractions versus time after trip. However, total RCS flow versus time after trip was more conservative than prediction. Subsequent evaluation , of measured results by CE, Windsor confirmed the acceptability , and conservatism of measured results when predictions of flow fraction versus time were revised to reflect as built RCS I flow characteristics. 3.3.3 Conclusions Measured results from the four (h) RCP trip were evaluated to be acceptable. Other results were used for evaluating the method of calculating Low Flow Trip setpoints for the Reactor Protective System (RPS).
TABLE 3.3-1 REAC'IOR C00LAffP PUMP FT,0W COASTDOWW COMBINATIONS PUMPS INITIALLY MINNING N UIN 11A, llB, 12A, 12B l 11A, llB, 12A, 12B 11A, 11B, 12A, 12B 11A l 11A, 11B, 12A, 12B 11B 11A, llB, 12A, 12B 12A 1 l 11A, llB, 12A, 12B 12B 11A, 11B, 12B - lowest total Mass Flow 12B l 11A, llB - lowest total Mass Flow llB - results in fastest RCS flow coastdown 11B, 12B - lowest total Mass Flow 12B - results in fastest RCS flow coastdown . l 1 I i s l l 2 R
Paga 27 l 3.h Pr4=7 and Secondary Water Chemistry
\
3.h.1 Purpose ) I To establish, monitor, and control primary and secondary water chemistry during plant heatup and conduct of Post j Core Hot Functional (PCEF) Tests. Baseline data to support the Low Power Physics and Escalation to Power Test phases will be obtained. ; 3.h.2 Test Results All primary and secondary water chemistry results during PCHF vere either acceptable or when acceptance criteria were not met, corrective action was instituted to achieve acceptable conditions. During baseline testing for Reactor Coolant System (RCS) particulate level and soluble corrosion products, no unusual or unexpected results were obtained. 3.h.3 Conclusions Overall the data collected indicated that good chemistry control was maintained even with the variable plant condi-tions required by the test program. I l l l l l l
Paga 28 3.5 Pressurizer Effectiveness Test 3.5 1 Purpose , The objectives of the pressurizer effectiveness test were to prove that air assisted pressurizer spray valves vill open and close fully when pressurizer pressure is 2250 psia + 15 psi, Reactor Coolant System (RCS) temperature is 5321PF and four (h) Reactor Coolant Pumps (RCP) are in operation and to determine the effectiveness of the pressurizer sprays and heaters in controlling RCS pressure during plant transients. i Acceptance criteria for t'his test was as follows: (1) Following an increase in pressure transient spray valves open fully and decrease pressurizer pressure at 82 1 20 psi / min. , (2) Following a decrease in pressure transient backup heaters ' increase pressurizer pressure at 17.5 1 3 5 psi / min. 3.5.2 Test Results The pressurizer spray valves did indicate that they would funy open and close with pressurizer pressure at 2250 psia, RCS temperature at 532'T and four (k) RCP's in operation. ; Effectiveness of the pressurizer sprays and heaters were ! determined by measuring the rate of pressurizer pressure decrease with sprays only and pressure increase with backup heaters only. Results of these tests are shown on Table 3.5-1. 3.5.3 Conclusions During runs number 1 - 6, one or both pressurizer spray valves were leaking. While it is not evident from the data, ! leaking spray valves could increase rate of pressure decrease and decrease rate of pressure increase. Run number 7 was con- ) ducted during Escalation to Power Tasting (EPI) after mainte- ; i nance on the spray valves was completed. At which time test , results satisfied the acceptance criteria. ) i I t I
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Paga 29 TABLE 3.5-1 PRESSURIZER SPRAY AND HEATER EFFECTIVENESS TEST PRESSURE DECREASE PRESSURE INCREASE . RUN (PSI / MIN) (PSI / MIN) 1 70 18 2 78 17 3 90 15 l h 7h 12 5 73 13 6 loh 16 7 72 15 i l 1 ) i i. 1 i i 6 i
Paga 30 3.6 Reactor Coolant System Leak Test 3.6.1 Purpose A leak test of the Reactor Coolant System (RCS) was performed to check for indications of abnormal leakage from the primary system. 3.6.2 Test Results The leak test of the RCS was conducted at 2300 (+25, -0) psia and covered the following areas: (1) Reactor Coolant Pump (RCP) Seal Area (2) Reactor Vessel Head Seal (3) Steam Generator Manvays (k) Control Element Drive Mechanism (CIDM) In-Cere Instrument Penetrations (5) Sten leakage on all valves (6) Pressurizer Heater Penetrations The inspection of these areas showed no abnormal leakage and the test was considered satisfactory. 3.6.3 Conclusions The RCS leak test showed that the primary system was tight after reactor vessel reassembly following fuel load and no abnormal leakage should be expected during the Startup Test Program.
Page 31 k.0 INITIAL APPROACH TO CRI*ICALITY Initial criticality was achieved on 10/7/7h at Reactor Coolant System (RCS) conditions of 260*F and h75 psia. The initial RCS boron concentration was 1990 ppm. The Initial Approach to Criticality (IAC) began by withdrawing all CEA's in specified increments with count rate data taken after each increment. Criticality was subsequently achieved by deborating the RCS to a boron concentration of 1023 pga at a rate of kh gpm. Throughout the approach to criticality, two (2) independent sets of inverse
=ulticlication plots were maintained. Two plots of inverse count rate versus RCS boron concentration were maintained during the dilution phase.
At the end of each reactivity addition, count rates were obtained from each Wide Range Log Channel (W3LC). The ratio of initial average count rate to the ceunt rate at the end of each reactivity addition was the value plotted. The CEA withdrawal sequence and intervals are shoir in Table k.0-1. The inverse count rate versus CIA position points for each WRLC are shown in Figures h.0-1 through h.0 k. The inverse count rate versus RCS dilution time in hours is shown in Figures L.0-5 through L.0-8. WRLC's 1 and 3 detectors were located closest to startup sources. The relatively greater scatter in data from WRLC's 2 and k was due to their greater distance from the startup sources. Figure k.0-9 shows the change in RCS boron concen-tration versus dilution time. During the approach to criticality, response of the VRLC's was monitored. The degree of response is recorded in Table L.0-2 and indicates that a greater than one (1) decade " overlap" existed between the proportional counters and the fission chamber cf each WPlc. After schieving initial criticality, Control Element. Assembly (CEA) Grcup 5 was used to control neutron flux. Conditions were stabili::ed at 10-k5 power and the critical data shown in Table k.0-3 was recorded ar.d compared with predicted values. In su:m:ary, initial criticality was achieved in a safe and orderly fashien. There was good agreement between the measured and predicted critical boron concentrations.
Pcga 32 TABLE h.0-1 CEA WITHDRAWAL SEQUENCE CEA POSITION POINT CEA GROUP INCHES WITHDRAWN 1 Shutdown A 67 0 2 Shutdown A Full Out 3 Shutdown B 67.0 h Shutdown B Full Out 5 Shutdown C 67 0 6 Shutdown C Full Out 7 Part Length CEA's (PLR) Full Out 8 Regulating 1 33.5 9 Regulating 1 67 0 10 Regulating 1 100.5 (28 11.25) 11 Regulating 1 136.0 (28 hh.75) 12 Regulating 2 85 5 (38 3.5) 13 Regulating 2 119 0 (38 29 5) lh Regulating 3 70 5 (28 136.0) 15 Regulating 3 10h.0 (48 1h.75) 16 Regulating h 55 5 (38 136.0) 17 Regulating h 89 0 (58 3.75) 18 Regulating h 122.5 (58 33.25) 19 Regulating 5 52.0 (h8 136.0) i
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TABLE'k.0-2 WIDE RANGE LOG CHANNEL RESPONSE (1) Signal Source 2) CHANNEL 1 CHANNEL 2 CHANNEL 3 CHANNEL h Proportional Counters and Fission Chambers CPS 2 X 103 2 X 103 2 X 103 2 X 103 Percent Power T X 10-5 7 X 10-5 9 X 10-5 8 X 10-5 Fission Chamber Only Percent Power 10-6 10-6 10-6 10-6 (1) Each Wide Range Log Channel (WRLC) signal is a combination of a signal from several proportional counters and a signal from a fission chamber. (2) At approximately 2000 counts per second (eps) increasing, proportional counters are automatically deenergized and the WRLC signal consists of the fission chamber signal only. 0 1
. -~.
Paga 3h TABLE h.0-3 INITIAL CRITICALITY DATA PARAMETER MEASURED INITIAL CONPITION RCS Te=perature (*F) 260 260 RCS Pressure (psia) hT5 h50 , RCP's Operating 11A and 11B liA and 11B WRLC 1 (% power) 2 X 10 h 1 x in h WRLC 2 (% pover) h X 10 h 1 X 10 k WRLC 3 (% power) 1 X 10 h 1 y lo-k WRLC h (% pover) 2 X 10 h 1 X 10 h CEA GROUPS (inches) PREDICTED Shutdown A 135 135 Shutdown B 135 135 . Shutdown C 135 135 Reg 1 135 135 , Reg 2 135 135 Reg 3 135 135 Reg h 13h 13h . Reg 5 h7 52 I"un 8 135 135 - RCS Boron Concentration (ppm) 1033 1050 + 100 _ 11 b _ _ , , , , . . _ , _ -_,, . - - . . + -------9 - --eer*-
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1 1 } Paga 35 l i 1 l 50 LOW POWER PHYSICS TESTS (LPPT) l The Calvert Cliffs Unit 1 initial core consists of two hundred seventeen , i (217) fuel assemblies each containing one hundred seventy-six (176) fuel l } rods / burnable poison rods and five (5) Control Element Assembly (CEA) ! ! guide tubes. Puel assemblies are divided into three (3) distinct groups ! j by enrichment, Type'A, B, and C. Twelve (12) fuel rods in all Type B i
- and several Type C fuel assemblies are replaced with burnable poison (
i rods. Table 5 0-1 tabulates this and other important core design ; characteristics. In addition to soluble boron in the Reactor Coolant System (RCS), reac-tivity control is provided by eighty-five (85) CEA's. CEA's are inserted 7 I i l into and withdrawn from the core by means of sixty-five (65) Control . 1 Element Drive Assemblies (CEDM's). Twenty (20) CEDM's are attached to { dual CEA's. Figure 5.o-1 shows the core location of the CEA's. Note i that dual CEA's have a single serial number designation corresponding to the serial number of their respective CEDM. The CEDM's are arranged into ten (10) CIA Groups. Those Groups are further defined by function. CEA Groups A, B, and C are Shutdown Groups. CEA Groups 1 through 5 are Regulating Groups. CEA Groups PLR 1 and PLR 2 are Power Shaping Groups. PLR 1 and PLR 2 may be combined into a single CEA Group designated PLR 8. Figure 5.0-2 displays the relative core location of the CEA Groups. CEA Group movement is restricted as a function of power level in order j to insure that CEA configurations unanalyzed for in the safety analysis do not occur. The uschanism for this restriction is a so-called Power l . Dependent Insertion Limit (PDIL) curve residing in the Technical Speci-ficiations. Automatic control features as well as operating instructions prevent insertion of CEA Groups into the core below this PDIL curve. l The lower the reactor power, the greater the CEA insertion allowed. i i LPPT consists primarily of the measurement cf reactivity worths of phenomena which can vary the critical condition of the core. To speed r the collection of this data, as well as to enhance its accuracy, an i analog computer which solves the kinetics equation for reactivity was ! used. Several techniques were used in conjtaiction with this reactivity r computer to measure CEA worths. The solubin boron sway technique con- [ j sisted of a continuous or slug dilution or horation of the RCS simul- t t taneous with ==a11 compensating reactivity changes in CEA position. j The reactor was kept near critical during t: sis evolution, and the reac- i l tivity computer provided a signal which could be trended and correlated ; i with CEA position as a function of time. A CEA trip technique was also i ! used in conjunction with the reactivity cosputer. The rapid change in i reactivity caused by a CEA or CEA Group trip was correlated with reac- ' tivity change detected by the reactivity craputer. j During the conduct of LPPT, several unusual events occurred. A rela- l tively rapid increase in RCS pressure drop across the reactor vessel was , noted. This necessitated cooling down, reLoving the reactor vessel head i and some internals, and thoroughly inspecti.sg the internal reactor vessel I cavity and fuel for a cause of the differectial pressure buildup. When I i
P sn 36 that situation was resolved, the reactor vessel was reassembled. However, as discovered during subsequent testing, several CEA's were not properly connected to their CEDM's. This required another cooldown and reactor vessel head removal and reassembly. A more detailed description of these events is included in Section 7 0. All rav test data was collected, reduced, and analyzed on site. In all cases, measured data met applicable acceptance criteria. CE, Windsor provided backup support for all measured data analyses and refined the analysis of several tests. 1 I l l l i I 4 4 d d 7
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P:g2 37 TABLE 5.0-1 , FIRST CYCLE CORE DESIGN CHARACTERISTICS Nuclear Characteristics Fuel Managment 3-Batch, Mixed Central Zone Average First Cycle Burnup, mwd /MfJ 15,400 U-235 Enrichment, v/o Batch A (69 assemblies) 2.05 Bsteh B (80 assemblies) 2.h5 Batch C (68 assemblies) 2.99 , H2 0/UO2 Volume Ratio, Unit Cell (Cold) 1.63 Mechanical Characteristics Fuel Assemblies No. of Fuel Rods Posion Rods Poison Rods Batch Assemblies No./Assy. No./Assy. No./ Batch A 69 176 0 0 B 80 16h 12 960 C h0 176 0 0 I C.(low Concen-tration Bhc loading) 12 16h 12 1hh C+(high concen-i tration Bhc loading) 16 16h 12 192 217 1296 ! Fuel Rod Array, square 14 x 1h Fuel Rod Pitch, inches 0 580 Spacer Grid Type Leaf Spring Material Zircaloy k Number per Assembly 8 Retention Grid Type Leaf Spring Material Inconel Number per Assembly 1
Pega 38 TABLE 5.0-1 (cont'd) Weight of Contained Uraniun, kg U Batch A 395 Batch B 368 Batch C (poisoned) 368 Batch C (unpoisoned) 395 . Outside Dimensions l Fuel Esd to Fuel Rod, inches T.980 x T.980 Fuel Rod Fuel Material (Sintered Pellets) UO2 Pellet Diameter, inches .3795 ' Pellet Dish Depth, inches 0.015 Pellet Dish Diameter, inches 0.2915 Pellet Length, inches 0.650 Pellet Density, g/cc 10.193 Pellet Theoretical Density, g/cc 10.96 Pellet Density (% theoretical) 93.0 1.1 5 Stack Height Density, g/cc 10.05h Clad Material Zircaloy k Clad ID, inches 0.3880 Clad OD, (nominal) inches 0.hk0 Clad Thickness, (nominal) inches 0.026 Diametral Gap, (cold, nominal), inches 0.0085 Active Length, inches 136.7 Burnable Poison Rod Active Length, inches 122 7 Material BkC - A1203 Pellet Diameter, inches 0 376 Clad Material Zircaloy k Clad ID, inches 0 388 Clad OD, inches 0.hk0 l Clad Thickness, (nominal) inches 0.026 Diametral Gap, (cold, nominal), inches 0.012 4 Control Element Assembly (CEA) , Full Lenath Part Length Number 77 8 i l Nuiber of Absorber Elements per Assembly 5 5 I Type Cylindrical Rods Cylindrical Rods i Clad Material Inconel 625 Inconel 625 i Clad Thickness, inches 0.0h0 0.Cho Clad OD, inches 0 9k8 0.098 4 Poison Material (1) (1) 1 Total Element Length 161 31 161 31 i f (1) Poison material is primarily BkC-A1203 Several CEA's finger's have a conbination of A1203 and BkC-A103 2 l h
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P2ga 39 l l l 5.1 Shielding Effectiveness and Plant Radiation Level Measurements l 5.1.1 Purpose l A comprehensive survey of radiation levels around the plant i in general and the biological shield in particular was per- , formed. This survey was the basis for determining later ' buildup of radioactivity in particular piping and components. It was also used to predict radiation levels at higher power
- levels and to evaluate efficiency of shielding. Design bases dose rates for the several zones are listed in FSAR Section 11.2.1.
{ 5.1.2 Results General area gamma dcse rates and all shielding point gansna and neutron dose rates were less than 0.1 mram/hr for all Survey Points outside of Unit 1 Containment. All except six of the TLD's placed outside of Unit 1 Containment indicated i dose rates attributable to natural background radiation. One TLD mounted near the boronometer showed a measurable but
; amall (0.05 mram/hr average) dose from the boronometer source, l and five TLD's mounted on the Units 1/2 barrier fence in the i turbine building showed some exposure above natural back-l ground attributable to radiography from Unit 2 construction.
i Measurable gamma and neutron dose rates were obtained at various locations within Unit 1 Containment during routine I radiological surveys throughout LPPT. It was difficult to extrapolate readings obtained into meaningful full-load dose rates due to sensitivity of measurements and the uncer-t tainties in the power measurement. However, based on TLD results and neutron survey meter results, it was estimated that at 100 percent power the total exposure rate just inside , the containment personnel hatch may be 10 - 100 res/hr and 30 - 300 rem /br near 12B Safety Injection Tank with gamma and neutron each contributing about one-half the exposure, i 5.1.3 Conclusions - The maxis.nn general area dose rates at all survey locations when extrapolated to 100 percent power were generally con-sistent with the criteria presented in FSAR, Section 11.2.1. 4 i i l I i' _. _ ~ _ ______ _ _. __ _. .. . . . _ , _ . . _ . _ _ .
Paga ho 52 Effluent Radiation Monitors Calibration 5 2.1 Purpose Gaseous and liquid effluent in process radiation monitors were initially calibrated using manufacturers detection efficiency data. During initial radioactive fluid releases, grab sample analysis is compared with in process monitor response to establish valid correlations for future reference. 5 2.2 Test Results Fluid releases to the environment were of such low radio-activity levels that collection of meaningful data for a valid comparison of monitor responses and grab sample analyses was not practical. 5.2.3 Conclusions Effluent radiation monitor calibrations were deferred until generation of higher Reactor Coolant System liquid and gaseous radioactivity levels expected during Escalation to Power Testing. b
.- - - - . - - - - . . , ,- - - - - .~_ . _ , _-. - - - - , - - - , - - - ., -v.,_. , - .- , - --_-, --v ---, ,-- .- ----
Paga h1 53 Critical Boron Concentration Measurements 5 3.1 Purpose Critical boron concentration measurements were performed at various Reactor Coolant System temperatures and pressures. The purpose of these measurements was to obtain an as mea-sured value for excess reactivity loaded in core and to pro-vide bases for verification of predicted reactivity worths. 5 3.2 Test Results Boron concentration values are averages of multiple chemical analysis measurements made during periods of stable Reactor Coolant System (RCS) boron concentration. As indicated in Table 5 3-1 below, measurement points were also independently analyzed by the reactor vendor (CE). TABLE 5.3-1 COMPARISON OF PREDICTED AND MEASURED CRITICAL BORON CONCENTRATION i
- NOMINAL RCS CEA GROUP 5 TIMPERATURE POSITION (INCHES PREDICTED MEASURED ACCEPTANCE AND PRESSURE WITHDRAVN) VALUE (tm) VALUE (em) LIMITS (etm)
- BG&E CE 260'F. h60 psia 121.5" 1062 TOTT 1060 1 100 532'F, 2250 psia 12h.5" 1078 1089 1102 1 100 533 Con.:1usions i
Results indicate that measured boron concentration are l in adequate agreement with predictions and vc11 within the acceptance criterion of i100 pga. I i i 4
1 Page h2 k
\
i ! 5.h Temperature Coefficient of Reactivity Measurements i ! 5.k.1 Purpose i The moderator temperature coefficient of reactivity can be either negative or positive, depending upon the magnitude of the Reactor Coolant Systest boron concentration. The moderator temperature coefficient cannot be measured directly i but it can be derived from a measurement of the isothermal temperature coefficient. The assumption being that the measured reactivity change due to a change in measured moderator temperature is actually a function of a like and simultaneous change in the temperature of all core compo-nents including fuel and moderator. The moderator temper- ] ature coefficient at full power condition as derived and i extrapolated from the zero power measurement of isothermal temperature coefficient must be less positive than that required by Technical Specifications. , j 5.k.2 Test Results ) Isothermal temperature coefficient measurements were con- ; i ducted at several different Reactor Coolant System temper- j atures and boron concentrations. Measured values for each , condition are the result of averaging data from several , segments of the heatup and cooldown phases of the measurement. i Throughout the measurements, reactor power was maintained below the point of adding nuclear heat to minimize the con-t fusing effect of doppler feedback. Reactor Coolant System l ramp temperature changes were affected by proper positioning of turbine bypass or atmospheric dump valves. ; 1
, Table 5.k-1 summarizes the results of the measurements and !
l comparisons with predicted valves. Agreement between measured l j and predicted values improves with increase in Reactor Coolant i System temperature and shows good agreement at 532'F. i j Extrapolation of the zero power all rods out isothermal tem-j perature coefficient repulted in a full power moderator i coefficent of -0.26x10**Ak/k/*F. Technical Specification , l 3.10.I.1specifiesthatthemoderagortemperaturecoefficient shall not be greater than +0.2x10* Ak/k/*F at full power. l 1 J l i I 5.4.3 conclusions !1 j Forallcases,themeasuredvaluesofisothermaltemperaturg coefficient are within the acceptance criterion of +0 5210- { Ak/k/'F of the predicted value. The extrapolated value of the sero power isothermal temperature coefficient to the full l i power condition is less positive than the limit specified in Technical Specifications and is therefore acceptable. ] i k i
~ _
l l TABLE 5.k-1 SIDWARY OF ISOTHFRMAL TEMPERATURE COEFFICIENT MEASUREMENTS NOMINAL RCS B030M ISOTHERMAL TEMPERATURE COEFFICIENT RCS TEMPERATURE CONCENTRATION CEA CROUP POSITION (x10-k A k/k/*F) AND PRESSURE (ptsa) (INCHES WITHDRAWN) MEASURED PREDICTED
- 260*F, h60 psia 1022 5@0" +0.125 -0.030 l k@0 to 36" (3) 260*F to 360*F. 1071 (1) 5068" +0.211 +0.1ko l k60 psia l
l 360*F to k50*F, 108k (2) 5@68" +0.2k0 +0.230 1100 psia l 532*F 2250 psia 1087 58105" +0.257 +0.300 532*F, 2250 psia 949 580" -0.289 -0.260 k80" 3818" 532*F 2250 psia T20 1-5@0" C@l0 to 30" (3) -0.809 -0.680 (1) @ RCS tearperature cf 310*F (2) @ RCS teamperature of kO5'F (3) CEA Group position was changed during measurement, however, data reduction took account of this movement. > 5 l % 5 l
1 Pass kk 4 55 Non-Overlapped Regulating and Shutdown CEA Group Worth Measurements j
)
i 5.5.1 Purpose l i During reactor operations, nearly all excess reactivity is held down by soluble boron concentration in the Reactor ) Coolant System and burnable poison shim rods in the fuel L
! assemblies. Additional hold down and reactivity control is
- provided by moveable Control Element Assemblies (CEA).
These CEA's are arrayed in sysunetrical groups about the
- core (see Figure 5 0-1). The number of CEA's in each Regulating and Shutdown CEA Group and the function of that group is described in Table 5 5-1. The CEA Group worths vere measured in a non-overlapped mode over the full I range of their movement and at various Reactor Coolant
] System temperatures. ] 5.5.2 Test Results l All CEA Group reactivity worths were measured using a lf soluble boron swap method, either dilution or boration, j to maintain criticality while inserting or withdrawing j CEA Groups in increments. The reactivity trace generated by this evolution was reduced to obtain the relationship , between CEA Group positions from full in to full out and ! integral reactivity worth at those positions. i I For Shutdown CEA Group A, integral worth was measured using i the soluble boron swap method in combination with a group trip method. The ccmbination of methods allows total
- integral worth of Group A to be determined from extra-1 polation of measured data without decreasing shutdown ;
margin below the Technical Specification limit. i The integral worths of CEA Groups k and 5 were measured I at a Reactor Coolant System temperature of 260*F. The l integral worths of all Shutdown and Regulating CEA Groups l were measured at 532*F. These results are compared with ! predicted valves in Table 5 5-2. In addition, the integral j reactivity worth curves developed at 532*F for all Shut- , down and Regulating CEA Groups are displayed in Figures : 5.5-1 through 5.5-8. i i j 553 Conclusions l l 1 The measured CEA Group integral reactivity worths are in ; , good agreement with predicted values and are well within ) acceptance limits. i 1 l t i I
I TABLE 5.5-1 - REACTIVITY CONTROL FUNCTION OF CEA GROUPS CEA GROUP NUMBER NUMBER OF CEA's CONTROL FUNCTION A 16 Safety B 8 Safety C 16 Safety l 1 8 Power Regulating l 2 8 Power Regulating l 3 8 Power Regulating h h Power Regulating 5 9 Power Regulating l l PLR 1 h Axial Power Shaping PLR 2 h Axial Power Shaping 1 r u
TABLE 5.5-2 COMPARISON OF MEASURED AND PREDICTED CEA CROUP INTEGRAL REACTIVITY WORTHS I. Reactor Coolant System Temperature at 260'F ACCEPTANCE CEA GROUP NUMBER OF CEA'S MEASURED WORTH (% Ak/k) PREDICTED WORTH (I Ak/k) LIMITS (5 Ak/k) k k 0.135 0.120 1 0.05 5 9 0.h53 0.h30 1 0.11 II. Reactor Coolant System T eperature at 532*F ACCEPTANCE CEA GROUP NUMBER OF CEA'S MEASURED WORTH (% Ak/k) PREDICTED WORTH (% Ak/k) LIMITS (5 Ak/k) A 16 3.327 3.200 1 0.85 D 8 0.997 0.990 1 0.220 c 16 1.303 1.h00 1 0.350 1 8 0 953 1. oho 1 0.260 2 8 0.778 0.Tho 1 0.190 3 8 0.932 0.900 1 0.230 k k 0.353 0.350 1 0.090 5 9 0.552 0.5ho 1 0.1h0 Total TT 9.195 9 160 W
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Page h7 5.6 Overlapped Regulating CEA Group Worth Measurements 5.6.1 Purpose Reactor Power level may be controlled by sequential inser-tion or withdrawal of Regulating Control Element Assemblies (CEA). Percent of overlap is selected so as to insure a relatively constant insertion rate of positive or negative reactivity over the full range of CEA Group movement. Technical Specificat.ons allow CEA Group insertion as a function of reactor power level. Maximum insertion occurs at zero power and amounts to approximately 1.5%Ak/k negative reactivity inserted due to CEA's. The integral reactivity worth curve for Regulating CEA Groups 5, h and 3 in an overlapped mode was measured. Maximum allowed insertion at zero power being at approximately 60 inches withdrawal on CEA Group 3. This measurement was only made at a Reactor Coolant System (RCS) temperature of 532 F. Principal purpose of the mes-surement being to develop an integral vorth curve for use in making estimated critical condition calculations prior to a reactor startup. Reactor startup's are performed at a nominal RCS temperature of 532'F. 5.6.2 Test Results The overlapped integral reactivity worth of CEA Groups 5 h, and 3 was measured using a soluble boren swap method to maintain criticality while sequentially inserting CEA Groups in increments. The reactivity trace developed by this CEA Group movement was reduced to obtain the relation-ship between CEA Group positions and integral reactivity worth at those positions. Figure 5.6-1 displays the over-lapped integral reactivity worth curve for CEA Groups 5, h, and 3 from all rods out to less than 60 inches withdrawn on CEA Group 3. 5.6.3 Conclusions The Overlapped Integral CEA Worth curve derived from this measurement has been adequate for use as an operational tool. F 1 _ _ _ - _ .- _. - - - - - _ .. . _ _ - .
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P2gn 48 5.7 Pressure Coefficient of Reactivity Measurements 5 7.1 Purpose The pressure coefficient of reactivity can be either negative or positive depending upon the magnitude of the Reactor Coolant System baron concentration. While =all and relatively insignificant, the measured value of the pressure coefficient of reactivity is of interest. 5 7.2 Test Results The pressure coefficient of reactivity was measured at two different Reactor Coolant Systen (RCS) temperatures over two different pressure ranges. The results are pre-sent in Table 5.7-1 below. TABLE 5.7-1 MEASURED PRESSURE COEFFICIEiTS OF REACTIVITY NOMINAL RCS RCS PRESSURE PRESSURE COEFFICIEIT OF TD@ERATURE('F) RANGE (psie) REACTIVITY (X10-7Ak/k/ psi) 360 h60 to n00 k.81 h50 n oo to 2250 -5 03 5.7.3 Conclusions Results indicate that the pressure coefficient of rese-tivity is relatively insignificant and was several orders of magnitude smaller than the temperature coef-ficient of reactivity. l i 1 )
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l Page h9 5.8 Dropped CEA Worth Measurements 5.8.1 Purpose A dropped CEA under power operation conditions vill reduce reactor power level and distort the core power distribution. The reactivity worth of the most reactive CEA is measured from a full power CEA configuration in order to verify safety analysis. 5.8.2 Test Results The dropped CEA integral reactivity worth measurement was performed simultaneously with a check of core sy= metry. The integral reactivity worth of each CEA was measured using a CEA swap technique. This measurement was compared with that of all symmetric CEA's in its CEA Group in order to detect any unexpected core asymmetry. No significant asymmetry was noted and thereby gave additional assurance of proper assembly and core loadings. The most worthy dropped CEA from a full power CEA configu-ration was CEA B-8. Its measured integral reactivity worth is compared with predicted worth in Table 5.8-1 below. TABLE 5.8-1 COM5'ARISON OF PREDICTED AND MEASURED DROPPED CEA INTEGRAL REACTIVITY WORTHS REGULATING CEA GROUP POSITION PREDICTED WORTH ACCEPTANCE NOMINAL RCS TEMPERATURE (INCHES WITH- MEASURED WORTH (GROUP B CEA) LIMITS AND PRESSURE DRAWN) (B-8) (%Ak/h) (%Ak/k) (%Ak/k) 532 F, 2250 psia 1-k 8 AR0(1) 0.151 0.lho +0.050 5 8 110" (1) ARO a All Rods Out 5.8.3 conclusions l The integral reactivity worth of the most reactive dropped 1 CEA front a full power CEA configuration was determined to l be slightly more than predicted and was well within the acceptance criterion.
Page 50 5.9 Ejected CEA Worth Measurements 5.9.1 Purpose Technical Specifications state that the maximum reactivity worth of any one CEA in the core shall not be more than 0.38%ak/k at full power at beginning of life. In addition, the maximum worth of any one CEA in the core shall not be greater than 1.06%dk/k at hot zero power at beginning of l life. A pseudo ejected CEA reactivity worth measurement was made under full power and zero power CEA configurations in order to verify safety analysis calculations related to the hypothetical CEA Ejection Incident. 5.9.2 Test Results 5 9.2.1 Full Power CEA Configuration. Under full power conditions, CEA Group 5 may be inserted to approxi-mately 79 inches withdrawn. All other CEA Groups are at all rods out configuration. With respect to reactivity worth, CEA Group 5 consists of three (3) different types of CEA's. The integral reac-tivity worth of each of these three (3) CEA types with a predicted CEA vorth typical of others of its same symmetrical core position was measured. The measurement technique was one of boration of the first CEA to full out followed by a swap between succeeding CEA's to measure integral reactivity worth of each CEA from approximately 79 inches withdrawn to full out. The most reactive CEA was 5-1. Its measured results are compared with pre-dicted in Table 5 9-1. 5 9 2.2 Zero Power CEA Configuration. Under zero power conditions CEA Groups 5 and h may be inserted to full in with CEA Group 3 at approximately 60 inches withdrawn. All other CEA Groups are at All Rods Out configuration. Certain CEA's from all three (3) inserted CEA Groups were selected and their reactivity worths measured in accordance with the method out-lined in Section 5 9 2.1 above. In this instant, there were five (5) different CEA types with respect to reactivity worth. The most reactive CEA vas 5-36. Its measured results are compared with predicted in Table 5 9-1. 593 Conclusions l The measured values for both pseudo ejected CEA conditions are less than the predicted reactivity worths and in addition, are well within the acceptance criteria. Measured values are also approximately an order of magnitude less than those assumed in performance of the Safety Analyses. 1 1 l i
i TABLE 5 9-1 COMPARISON OF MEASURED AND PREDICTED PSEUDO EJECTED CFA REACTIVITY WORTIIS I. FULL POWER CEA CONFIGURATION NOMINAL RCS TDIPERATURE REGULATING CEA GROUP MEASURED WORTH PREDICTED WORTH ACCEPTANCE CEA NUMBER AND PRESSURE POSITION (INCHES WITIIDRAWN) (% 4k/k) (% ak/k) LIMIT (%4 k/k) 5-1 532*F, 2250 psia 1 h @ ARO (1) 0.030 0.050 -+ 0.050 5 @ 79" II. ZERO POUEI: CEA CONFIGURATION NOMINAL RCS TEMPERATURE REGULATING C,1 GROUP MEASURED WORTH PREDICTED WORTl! ACCEPTANCE CEA NUMBER AND PRESSURE POSITION (IN 'HES WITHDRAWN) (% d k/k) (% 4 k/k) LIMIT (% 4k/k) 5-36 532*F, 2250 psia 1-2 @ ARO (1) 0.11h 0.120 -
+ 0.050 3 @ 54" h-5 9 0" (1) ARO m All Rods Out 8
ic
i l Page 52 5.10 Stuck CEA Worth Measurement 5.10.1 Purpose Technical Specifications state that available shutdown margin shall not be less than 2.h hk/k with the highest worth CEA stuck out whenever the reactor is critical. The reactivity worth of the most reactive stuck CEA vas measured in order to verify the validity of predicted stuck CEA reactivity worths. 5.10.2 Test Results A Group A CEA was predicted to be the most worthy stuck CEA. CEA's from Group A (A kh) and from Group B (B-9) were selected for measurement of stuck CEA reactivity worth. The measurement technique consisted of CEA Groups A and B combined trips with and without the " stuck" CEA stuck in the full out position. The reactivity data from these measurements was then used in combination with data from the CEA Group A reactivity worth measurement described in Section 5.5.2 to extrapolate an integral reac-tivity worth for CEA's A kh and B-9. This preliminary on-site evaluation resulted in a measured maximitt stuck CEA reactivity worth of no greater than 1,98%Ak/h. The worth of CEA B-9 being slightly greater than that for A kh. A subsequent off-site analysis of the measured data by CE, Windsor has resulted in a final stuck CEA reactivity vorth of 1.70%ak/k. 5 10.3 Conclusions The preliminary on-site analysis of measured data indicated that CEA B-9 was the most reactive " stuck" CEA. Its reactivity worth being within the bounds of ! the acceptance criteria of 1.58 + 0.kokk/k. Subsequent off-site analysis of data by CE, Windsor has verified l this on-site conclusion. ! l l l l l l l 1 l
Paga 53 5.11 Part Length CEA Group Measurements 5.11.1 Purpose i l The purpose of these measurements was to collect Part Length Control Element Assembly (PLR) Group reactivity worth data. 5.11.2 Test Results The reactivity worth of PLR 1 (see Figure 5.0-2) was j, measured over its full travel from full out to full in for two different Regulating CEA Group configurations. A soluble boron svsp with PLR 1 technique was used to generate reactivity worth versus CEA position information. Figure 5.11-1 displays the integral reactivity worth curve of PLR 1 for a 50% (Power Dependent Insertion Limit) PDIL Regulating CEA Configuration (i.e., CEA Group 5 9 0 inches and Group h 9 76 inches). Figure 5.11-2 displays similar infor=ation for a Full PDIL Regulating CEA Configuration (i.e., CEA Group 5 @ 79 inches). The reactivity worths of PLR 2 and a com-bination of PLR 1 and PLR 2 referred to as PLR 8 vere not measured. t 5 11.3 Conclusions Measured PLR 1 integral vorth curve data was cc11ected.
)
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-0.15 0 10 20 30 40 50 IO 70 00 90 100 110 120 130 140 CEA WITHDRAWAL (INCHES)
FIGURE 5.11-1
CALVERT CLIFFS UNIT 1 BOL, 1st CYCLE, 532*F, 2250 PSIA INTEGRAL PARTLENGTH CEA GROUP WORTH FULL POWER - PDIL 0.25 _ _ . _. ..
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FIGURE 5.11-2
Pega Sh 5.12 Critical Boron Concentration and Soluble Baron Worth Measurements 5.12.1 Purpose Soluble boron in the form of dissolved boric acid in the Reactor Coolant System provides variable reactivity control over the life of a core. It can supplement the reactivity control provided by CEA Groups. However, its principal function is to compensate for burnup of excess reactivity as core depletion proceeds. The critical boron concentra-tion for various CEA configurations was measured in order to develop a relationship for determination of the soluble boron reactivity worth. CEA Group hold down values were also measured and are presented in Section 5.5. 5.12.2 Test Results CEA Group integral reactivity vorths were measured using a soluble boron swap technique. In addition, the soluble boron concentration at the end point of seversl of those CEA configurations was also measured. Soluble boron sa=ples were independently analyzed by BG&E and by CE, Windsor. A cc=parison of measured with predicted critical boron concentrations for these several CEA configurations is presented in Table 5.12-1. A relctionship between reactivity change, CEA configuration, and critical boron concentration was developed over a range of Reactor Coolant System (RCS) boron concentrations. An average soluble boron verth was developed fr=n that data and is co= pared with the predicted values in Table 5.12-2 below. TABLE 5.12-2 COMPARISON OF MEASURED AND PREDICTED SOLUBLE BORON WORTH j NOMINAL RCS RANGE OF MEASURED PREDICTED ACCEPTANCE TEMPERATURE RCS SOLUBLE SOLUBLE BORON SOLUBLE BORON LIMITS AND PRESSURE BORON CONC. (1) WORTH (pix:r/%ak/k) WORTH (pum/%6k/k) (pTun/%4k/k) l BG&E CE BG&E CE 532*F, 2250 psia 628-1087 555-1102 77.3 77 83 + 15 (1) CEA configurations for BG&E and CE ranges are not the same.
Paga 55 5.12.3 conclusions The agreement between measured and predicted critical boron concentrations and between measured and predicted soluble boron worths are adequate and well within the acceptance criterion. t J l l a }
TABLE 512-1 COMPARISON OF MEASURED AND PREDICTED CRITICAL BORON CONCENTRATIONS FOR VARIOUS CEA CONFIGURATIONS NOMINAL RCS CEA GROUPS MEASURED PREDICTED TEMPERATURE INSERTED CRITICAL BORON CRITICAL BORON ACCEPTANCE AND PRESSURE (INCHES WITHDRAWN) CONC.(ppm) CONC. (ppm) LIMITS (ppm) BGLE CE 260*F, h60 psia ARO (1) 10h8 1060 1062 1 100 260*F, 460 psia 5 h @ 0" 1022 - 1023 1 100 532*F, 2250 psia ARO (1) 1089 1102 1078 1 100 532 F, 2250 psia 5-3 8 0" 9h9 -- 929 1 100 532 F, 2250 psia 5-1 8 0" T20 726 665 (2) g too C @ 0" B 8 120" (1) ARO a All Rods Out (2) Predicted Critical Boron concentration for CEA Groups 5, h, 3, 2,1, and C fully inserted and CEA Groups A and B fully withdram. N R M
i l Page 57
)
I l 5.13 Chenical and Radiochemical Tests 5.13.1 Purpose Chemical and radiochemical analysis of the Reactor Coolant System (RCS) and the steam side of the steam generators can give clues as to the metallurgical condition of critical system components including fuel. The purpose of these tests was to determine baseline corrosion data, fission product activity levels and buildup, tramp uranium contri-butions to activity levels, and early detection of failed fuel. 5.13.2 Test Results 5.13.2.1 Baseline Corrosion Study. A number of crud samples were taken during Low Power Physics Test (LPPT). Most of these vere taken as a result of investigations into the cause of the reactor vessel differential pressure situation. The crud samples were analyzed by emission spectroscopy and X-ray diffraction. Those results are summarized in Table 5 13-1. The following observations were made: (1) Weight of particulates (ppm) - In most cases all crud samples were less than
.01 ppm insoluble.
(2) Microscopic Examination of Particulates - Specific examination of crud samples col-lected 10/23/Th and 11/5/7h showed greater than 99% crud-type material, and less than 1% fibers and resin fines. I (3) Soluble corrosion products - During LPPT, the following -4== and minimum values were observed: Max Min Fe, ppm .17 less than .01 Cr, p p less than .02 less than .02 Ni, pp .3h less than .04 Cu, pin .0h less than .01
Paga 58 5.13.2.2 Fission and Activiation Product Buildup Study. Iodine - 133, Cs - 137 and Xe - 135 were the only fission products pcsitively identified during LPFr. The maximum and minimum values of the following were observed: Max liin Xe-135,pei/ml 3.07 x 10-7 3.07 x 10-7 I-133,fei/ml 2.8h x 10-7 2.05 x 10-8 Cs-137,,vei/nl 2.31 x 10-6 2.01 x 10-8 Gross Beta Gamma,p ci/nl 1.h0 x 10 h 1.25 x 10-7 Total activity,,vei/ml 1.10 x 10-h h.73 x 10-7 It is felt that the source of Cs - 137 may be from the conta=inated equi; stent supplied by CE for the core inspection rather than from the fuel. The reactor was not critical when Cs - 137 was first observed. Other activation products ident'.fied in liquid or crud samples include: F-18 Cr-51 Zr-95 Na-2h Ni-65 Zr-97 Mn-56 Np-239 Co-60 Co-58 Mo-99 Cu-64 Ar h1 W-187 Generally Mn-56 and Ar-h1 vere the largest contri-butors to total activity during LPPT in liquid samples. In crud samples Mn-56, W-187 and Co-58 were generally the largest contributers to crud activity. Since only sporadic occurrences of I-131 were observed in liquid samples, no effort was made to determine an I-131/I-133 ratio value. The generally lov activities observed and poor counting statistics obtained would make any ratio inaccu ate at best. i
Pega 59 5 13.2.3 Lithium Buildup. During LPPT lithium buildup was not appreciable. Typically, Li was less than .1 ppm throughout LPPT vhen not added deliberately. Lithium was added to the RCS on 11/9/Th to bring RCS concentration to 2.0 pp for pH control. CE recommended the higher lithium concentration as a change in chemistry control of the RCS. 5.13.2.h Purification System D.F. (1) Crud samples from filter inlet and outlet were collected when crud levels were greater than .01 pp. The D.F. was approximately
- 5. crud levels were typically less than
.01 ppm from filter inlet and outlet, and it was not possible to obtain really meaningful filter D.F. data during LPPT samples.
(2) Due to the limited operation of the ion exchangers, sample analysis of Ni, Fe, and Cr on effluent were not performed. (3) During LPPT, activity leveh in the RCS vere near minimum detectable. Due to the minimum activity and only limited operation of the Chemical and Volume Control System (CVCS) ion exchangers, gamma scans and gross beta-gamma analysis were not taken. Any data taken under these conditions could not have been considered meaningful. 5 13.2 5 Volume Control Tank (VCT) Gas. Due to the fact that fiscion gases were at or below their mini-mum detectable activity, no buildup of fission gases were observed in the VCT and gamma scans were not conducted weekly. Samples of reactor coolant collected during LPPT showed no fission gas. 5.13.3 Radiochemistry Results during Evaluation and Resolution of Reactor Vessel Pressure Differential Increase Incident. During the conduct of LPPT, a relatively rapid increase in RCS pressure drop across the reactor vessel was noted
- (see Section 7.0). As a result of that observation and its aftermath, the reactor vessel was disassembled and reassembled twice. Additional radiochemistry evolutions were required during this period and their results are summarized as follows
i ! l 4 l l
Paga 60 5.13.3.1 pH Increase Tests. (1) A==onia addition. At 13h5 on 11/8/Th approximately 3.25 gallons of a 28% NH3 solution was added to the RCS to raise NH3 concentration. pH rose to about 7.3, conductivity to about 130 ppm and ammonia to about 25 ppe. Suspended solids collected every 15 minutes during the test all showed less than 10 ppb. (2) Lithium addition. At 1630 on 11/9/7h approxi=ately 3 kilograms of ILi OH-H2O was added to the RCS. This vns done in order to raise the high temperature pH about 1 pH unit into the alkaline region. Room te=perature pH vas essentially unaffected by the addition. Conductivity shovel a slight increase and lithium reached 2.0 ppm. Suspended solids collected every 15 minutes during the test all showed less than 10 ppb. 5 13.3.2 pH Decrease Tests. Following a preservice rinse to the Waste Processin; S: sten (WPS) to insure that chlorides would not be expelled fre= the resin, Purification Ion Exchanger 12 was placed in service at 1710 , on ll/1C/Th. Lithium was re=oved frc= the RCS at a rate dependent on the purification half-life. pH showed a slow decrease to that of boric acid alone. By the end of the test, following shut-down and feed and bleed operations, pH fell to about 5.7. This pH vas about the same as the pH before testing began. Suspended solids collected every 15 minutes continued to show less than 10 ppb. 5 13.3.3 Startup Data Following Reactor Vessel Differential Pressure Investigation. Prior to hydrazine addition, suspended solids were measured routinely to be about 70 ppb with a high value measured at 125 ppb. Following the addition of hydrazine, suspended solids fell to less than 10 ppb and remained there with the exception of one sample at 50 ppb during the subsequent heatup. Oxygen levels dropped to within specification after 12 hours following start of hydrazine addition.
Pega 61 Fe and Ni were analyzed routinely. Maximum and minimum values were: Max Min Fe, ppm .080 .020 Ni, ppm .335 . oho 5.13.3.h Subsequent Shutdown Information. Shortly after criticality, it became apparent that several of the CEA's were uncoupled and a shutdown was inevitable. Since H2 overpressure had been established, the shutdown involved degassing the RCS to 5 cc/Kg. The H2 supply to the VCT was isolated and to hasten the degassing process the pressurizer was vented through the NSSS sample sink. The RCS H2 concentration shoved only gradual decrease (VCT vapor still about 60% H2) until the VCT was purged with N2 and H2 concentration in VCT vapor decreased to about 9% H2 At 1315 on 12/12/74 the RCS H2 was found to be h 8 cc/Kg allowing the RCS ta be opened to atmosphere. 5.13.3.5 Subsequent Startup Information. The startup to complete LPPT was uneventful. It should be noted that 2.2 Kg of Li OH-H2 O was added to the RCS with a corresponding rise in pH (G 25*C) to about 6.7. 5 13.h C lusions Chemical and radiochemical data was reviewed throughout LPPT and all results were either acceptable or, when j appropriate, corrective action was instituted to achieve acceptable conditions. Little meaningful data was obtained on the performance of CVCS ion exchangers and filters due to infrequent operation of the ion i exchangers and low levels of suspended solids and radioactivity. j I Lithium exceeded its acceptance limits due to intentional ' additions of TLi for pH control. The acceptance limit ) had been previously changed to allow the RCS to operate on 7Li pH control. "'he new limit being 2 ppm instead of 0.5 ppm. l
.._~ _. - - . _. .- ._. - ._ ____- - - . _ _ - _ - . - . . - - - -
j TABLE 5.11-1 SPECIAL CRUD SAMPLE RESULTS DATE TIME TTPE OF SAMPLE ANALYSIS RESULTS AND REMARKS 11/5 2330 50-liter crud h6 ppb - microscopic exam showed greater than 99% crud i 11/6 1531 100-liter crud less than 10 ppb - first showed significant Cu - 6h on ganana scan
, 11/7 0430 h62-liter crud less than 10 ppb
- Fe, Ni, Cu, Pb (major) 1' B, Zn, Cu (minor) i
! 11/8 0700 577.5 liter crud less than 10 ppb l 11/9 1000 630-liter crud major - Cu large - Fe, Ni, Na (questionable) minor - C1, Pb, Ca, Mg, Al, Ag, Ti, Si, B 11/11 2035 30-liter crud less than 10 ppb - CE analysis (2 filter papers) Metal micro grams /30 liter Metal micro grams /30 liter Fe 2.h II) Zn 1.2 1.1 (2) 1.0 Cu h.T (1) Pb 26.h ) 3.9 (2) less than 2.0 .o I & ; (1) Top l (2) Bottom
4 TABLE 5 13-1 (cont'd) DATE TIME TYPE OF SAMPLE ANALYSIS RESULTS AND REMARKS 11/11 2035 continued Metal micro grams /30 liter Metal micro grams /30 liter Ni non-detectable II) Ci non-detectable (1) non-detectable non-detectable (2 Al not completed CE remarks - unusual color would expect gray to black; Cu and Pb very unusual 11/13 1800 fil Purifi. Filter 1. Yellowish soft wax - grease Deposits 2. white material - 2 types elastic - organic polymer particulate - titanium dioxide (TiO 2)
- 3. black particulate Zr 70%
Ni, Ci, Fe - less than 30% total Ti - about 5% S102 - less than 1% m Cu - trace $e Visual inspection of filter indicated greater than 99% black @ particulate. The grease and white material were scattered at random. (1) Top (2) Bottom
TABLE 5.13-1 (cont'd) DATE TIME TYPE OF SAMPLE ANALYSIS RESULTS AND REMARKS 11/15 0900 Residue from hot leg Sample collected from drain line off of hot leg sample line sample point upstream of pneumatic isolation valve in containment - 28 mg collected. Analysis Zr (major) Fe, Cr, Ni (large) 11/19 Unknown Swipes from upper Fe301s - Fe2O3 50-50% guide structure Fe (major) Ni, Ti, Zr, Cr (large) Cu (trace) 11/20 0000 Swipes from fuel End fitting - Cr (major) assembly Fe, Ni, Ti (large) ' s Zr (minor)
., Cu (trace)
Fuel pins - Zr (major) s Fe, Ni, Ti, Cr (large)
?
T 9
Page 65 6.0 ESCALATION TO POWER TESTS (EPT) The Escalation to Power Tests were conducted to detern.ine as-built plant characteristics during steady state and transient operations from 0% to 100% power and to demonstrate, with reasonable assurance, that the plant is capable of withstanding the accidents and transients analyzed in the FSAR. Tests requiring steady state power vere performed at major pla-teau's of 20%, 50%, 80% and 100% power. Several minor tests were per-formed at 30%, h0%, 60%,70% and 90% power. j During EPT, two unusual events occurred which interrupted or delayed testing until they were temporarily or permanently resolved. During the 20% power test plateau, radiation measurements taken outside Unit 1 containment structure equipment hatch indicated that dose rates vould exceed the 0.5 mrem /hr limitation specified in the FSAR for the outside surface of the containment structure. In addition, higher than anti-cipated radiation levels were indicated at a number of locations inside the containment, where access on an infrequent basis is required during i operation. Analysis indicated that the high radiation levels were pri-i marily coming from the annulus between the reactor vessel flange and the primary shield vall. Temporary shielding was installed over the annulus and outside the equipment batch to reduce radiation levels. A detailed discussion of this problem is given in Section 7.3. Late j in EPT, an unplanned trip from 1005 power occurred which resulted in uncovering of the feedring in both steam generators. During the refilling of the steam generators, a water hammer occurred in the main j feedvater piping resulting in damage to the feedvater regulating valves.
- Subsequently, operating procedures were revised, such that, between
-85 inches and -30 inches, auxiliary feedvater is utilized to restore water level at a marimum rate of 1.2 inches per minute. This casualty f rerulted fn a seven (7) day delay in the test program. A more detailed i description of the water hanner incident is included in Section T.k.
Three EPT program tests covered belov vere deferred with approval of the Nuclear Regulatory Commission. (1) A test of axial xenon oscillation dampening usin6 Part Length CEA's (PLR) was deferred until such time as a decision is made to use PLR's. Later in first cycle life, an axial zenon oscillation dam-pening test vill be performed using full length CEA's. (2) A part loop operation power distribution measurement was deferred until such time as a decision is made to operate at power using i less than four (k) Reactor Coolant Pumps (RCP). Administrative l restrictions prevent power operation with less than full RCS flov l (k RCP's). In addition, the Reactor Protective System (RPS) vi p l cause a reactor trip whenever reactor power is greater than 10-*% and less than k RCP's are operating. ! (3) A performance test of the automatic CEA control features of the l Reactor Regulating System (RRS) has been deferred until such time ar a decision to use this feature is made. The principal reason
Pega 66 A i for deferring the test was to decrease the possibility of fuel failures which could be exacerbated by relatively rapid and near continuous CEA motion, a characteristic of automatic CEA control. Administrative restrictions prevent part loop operation and use of auto-matic CEA c.cntrol and PLR's until such time as tests are performed and results reviewed and approved by.the POSRC. Tha off-site anslysis by CE, Windsor of two (2) tests, Pseudo Ejected CEA and Dropped CEA continues. A preliminary o6. site analysis of the results of the Psuedo Kiected CEA test indicated "that they were well within acceptance criteria. A detailed on-site evaluation could not be accomplished as the In-Core Analysis (INCA) program is not capable of analy::ing asymetric power distributions. i i I , c , 4
, _ . _ _ _ _ _ _ _ , . . - . . - - , - - -- --~ . . - -r -
l PcEa 67 l l l i 6.1 Turbine Generator Startup and Atmospheric Dump / Turbine Bypass Valves Test l l 1 6.1.1 Purpose The test was designed to accomplish several objectives: l (1) To verify the proper operation of the turbine bypass / f i atmospheric dump valves control . system- ) (2) To verify the proper operation of turbine controls, generator controls and support systems; (3) To bring the turbine generator to about 10% power; and (h) To transfer feedvater control to autcmatic. J 6.1.2 Test Results i The test of the turbine bypass / atmospheric dump valves 1 was conducted concurrently with preparations for startup 1 of the turbine generator. The turbine bypass / atmospheric l dump talves were used to maintain Reactor Coolant System I (RCS) temperature at 532*F. The steam dump valves l operated normally during this test. All the turbine ' bypass valves were meche,nically limited to about half (1/2) ) stroke. 1-CV-39hh and 1-CV-39h6 vere isolated from their automatic signal and operated in a - 1 local upon deter- , mining that bypass control became unstable and flow through these valves insignificant when operated in automatic or manual from the control room. Evaluation of the turbine bypass problems continued throughout EPT. In addition I to instability problems, the valves would leak past the seat and required manual isolation to prevent unwanted cooldown. Evaluation has indicated that a possible solution vill be the installation of hydraulic dampeners j on all valve operators and installing new plugs made of ! h20 C steel. The hydraulic dampeners would increase the j stiffness of the valve to operate and dampen out valve ; motion in the unstable region. The original plugs were hko C material which was brittle and was fracturing. The l change to the more ductile h20 C material plugs should eliminate the valve cracking problem. The corrections ) for the turbine bypass valve were only partially installed by the end of the Startup Test Program. l l Testing of the turbine generator control circuits prior to initial operation consisted of the completion of Surveillance Test Procedures. These procedures covered the following: (1) Test of opening and closing circuits on main stop valves, combined intercept valves and bleeder trip test valves; 1 I
Paga 68 (21 Checking the phase a:ngle of vibration; (3) liydraulic Thrust Wear Detector tect; (k) i4sst.er Trip Solenoid Valve test; (5) Lachp Speed Control Amplifier test; (6) Iutomr. tic start of emergency beari2g oil pump; (7) Automa^.ic start of hydraulic p!ap (.MIC hydraulic power unit); and (8) 011 'frip test. No .w.jor probleat were Cetected during t'ae testing of tu;rbine gecerater comrols. 6.1.3 Conc 1c.stons No maaor problems yere encountered durias; startup of the turbine 6enerator. Problems with the t1.trbine bypass M1ves continued thronghovt TPf. Corrective measures are peM iaQy cJaplata. Pinsi ete1VAtion of the problems vill be completd after t.estirig the valves upon completion of
. the maintenance setivity.
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Pye 69 6.2 Reactivity Coefffeient Measurements ) 6.2.1 Purpose A test commonly referred to as a Variable T avg Test was conducted to determine the Power Coefficient and the Iso-thermal Temperature Coefficient (ITC). 6.2.2 Test Results Variable T avg Tests were conducted at each major power plateau (20%, 50%, 80% and 100%) with the Control Element Assemblies (CEA) inserted to approximately 100 inches on CEA Group 5. The test was conducted with the Power Coef-ficient and the Isothermal Temperature Coefficient (ITC) as separate tests. During the ITC test, AT power was held constant and Reactor Coolant System (RCS) T cold was varied. T cold was decreased 10*F below original temperature, con-ditions stabilized, data recorded and temperature increased - to 5'F above origin G temperature, conditions stabilized, and data recorded. This cycle was repeated twice with the exception that T cold was returned to the original temper-ature at the end of the final cycle. The Power Coefficient Tes't was conducted by holding T cold constant and decreasing grosa electrical power by approxi-mately 3h MWe at a rate of 15/ min, conditions stabilized, data recorded. Then gross electrical power was increased to the original power level at 1%/ min, conditions stabilized and data recorded. This cycle was repeated two (2) more times. The final power coefficient and ITC values were , < the average value of the runs conducted. The measured and ! predicted values for the temperature and power coefficient i for each power plateau are shown in Table 6.2-1. J 6.2.3 Conclusions The measured values for Isothermal Temperature Coefficient compared well with predicted values. The Power Coefficient was predicted to decrease in value with increasing power level. It was found that the measured Power Coefficient was reasonably constant with increasing power. However, results were all within acceptance limits. i l l 1
,.o--~ .,_--n
i TABLE 6.2-1 COMPARISON OF MEASURED AND PREDICTED ISOTHERMAL TEMPERATURE AND POWER C0FFICIENTS i 1 NOMINAL ISOTHERMAL TEMPERATURE COEFFICIENT ( Ak/k/'F) POWER COEFFICIENT ( Ak/k/% power) REACTOR POWER (%) PREDICTED MEASURED PREDICTED MEASURED
+0.05x10 h _1,99 1,oxio h -1.3hx10-4 20 +0.02 i 0.5x10-0 50 -0.17 i 0.5x10 b -0.11x10-4 -1.k2 1 1.ox10 b -1.ohx10 4 80 -0.18x10-b -1.10 1 1.ox10 h -1.04x10-4
-0.28 1 0.5x10 k 100 -0.37 1 0.5x10-4 -0.21x10 h -1.02 1 1.ox10 k -0 99x10 4
?
2
l Page 71 6.3 Plant Power Calibration 6.3.1 Purpose The purpose of the test was to: (1) Determine core thermal power by means of a secondary plant heat balance. (2) Adjust the Power Range Safety Che.nnels and AT Power Reference Calculators to agree with the thermal energy balance caluctions. (3) Perform when necessary a calibration of the Safety and/or l Control Power Range Channels followed by a calibration check using a current source standard. 6.3.2 Test Results Secondary Calorimetries using hand caluelations were con- l ducted at the 20, 50 and 80% test plateaus. The calori- l metrics were used to calibrate nuclear instrumentation and to verify the plant computer core thermal power calculations.
]
The Power Range Safety Channels and AT Power Reference Cal-culators were adjusted to agree within 0 5% of the Secondary Calorimetric calculations. These adjustments were performed at the 20, h0, 50, 70, 80, 90 and 100% power test plateaus. Initial calibration of the Power Range Safety Channels was conducted during the 20% test plateau using the Keithley pico-ammeters as a standard for subchannel calibration. Adjustment of both the Power Range Safety and Control Chan-nels was completed at the 50, 80 and 100% test plateaus. 6.3.3 Conclusions Hand calculations of core thermal power pointed out some minor discrepancies in the computer calculations. After correction of the deficiences, computer calculations proved to be reliable and accurate. Calibration of the Power Range Safety and Control Subchannels was acceerplished acceptably at each major test plateau. The intent of the Ex-Core Nuclear Instrument Calibration was to adjust nuclear power, AT power and the calorimetric to within 0.5% of each other. Due to noise on the AT power channels, nuclear power and AT power did vary by 0 7% while both were within 0.5% of the calori-metric. An investigation to determine the source of noise on the dT power channels and eliminate it is under way. Plant operating procedures contain instructions for hand calculations in case of computer failure and/or to verify computer calculations.
Page 72 6.h Shielding Effectiveness and Plant Radiation Levels 6.h.1 Purpose The test was conducted to accomplish the following four objectives: (1) Determine background radiation levels prior to plant startup. (2) Evaluate the adequacy of plant radiation shielding. (3) Determine radiation levels at varying power levels throughout the plant. (h) Datermine radioactivity buildup in specified piping and components. 6.h.2 Test Results Radiation surveys were conducted prior to initial criti-cality and at 0%, 50%, 80% and 100% power. After initial criticality, all measurements taken in and adjacent to the containment structure included neutron and gamma readings. During the initial phase of the power range testing program, radiation measurements taken outside Unit 1 containment equipment and personnel hatches indi-cated that dose rates vould exceed the 0.5 mrem /hr limi-tation specified in the acceptance criteria for the outside surface of the containment structure. In addition, higher than anticipated radiation levels were indicated at a number of locations inside the contairment, where access on an infrequent basis is required during operation. Analysis of the measured radiation levels and a review of the existing shield arrangement indicated that the high radiation levels were caused by neutron and gamma streaming out of: (1) the annulus between the reactor vessel flange and the primary shield valls (2) the annulus around the Reactor Coolant System (RCS) piping where it penetrates the primary shield vall and (3) to a lesser extent, an access opening through the lower part of the primary shield. Neutron and gamma dose rates were measured at 20% power and extrapolated to 100% power. Neutron dose rates at the 69' elevation ranged from 1000 aren/hr to more than 25000 arem/hr; gamma dose rates ranged from 250 aren/hr to 10000 aren/hr. The highest dose rates occurred at the north end of the refueling pool and over 12B Reactor Coolant Pump (RCP) removal hatch. At the 45' elevation, neutron dose rates ranged from 75 to 1500 aren/hr. Gemma dose rates ranged from 50 to 300 mrem /hr. The dose rates outside the equip-ment hatch were hoo arem/hr from neutrons and 150 mren/hr
Page 73 from gam as. At the 10' elevation, dose rates of 5000 mrem /hr from neutrons and 3000 mrem /hr from gama were found. j To reduce the radiation levels, temporary shielding was installed in the following locations: (1) Temporary neutron shielding consisting of bagged crystalline boric acid and polyethylene sheets were stacked on support grating specially erected to span the annular gap between the reactor vessel flange and the pri: nary shield. (2) A shield consisting of concrete blocks was fitted inside the access hole located at the bottom of the primary shield to a thickness of h8 inches. (3) Polyethylene shielding, approximately 12 inches thick was added to RCP hatches 11A, llB and 12A. Two layers, approximately 12 inches of boric acid bags were placed on 12B hatch. (h) A 2h-inch concrete block shield was installed outside of equipment hatch between the hatch and the rolling metal door reduced the total gama and neutron dose rate to less than 0 5 mrem /hr outside the equi;xnent hatch. (5) Shielding was also installed outside the emergency personnel hatch. After shielding installation, a neutron radiation survey at approximately k% power showed that a dose rate reduc-tion factor for the 69' elevation was from 30 to 100. The dose rate reduction factor for the h5' elevation was 5 to h0. The neutron dose rate reduction factor for the 10' elevation was 2 to 10. The gama dose rate reduction factor for the 69' elevation was-10 to 15 On the h5' elevation, the sama dose rates were less than 28 mrem /hr. On the 10' elevation the gaanna dose rates near the containment wall were reduced by fac-tors of 1.2 to 8, whereas the dose rates near the primary shield did not change significantly. 6.4.3 conclusions The totaldose rate at the h5' elevation inside containment is now less than 100 aren/hr, which is believed to be an acceptable level on a permanent basis to allow access during power operation. A further reduction by a factor of 5 to 10 in dose rates at the 69' olevation would be necessary to reach a nominal target level of 100 mrea/hr. Design of
Page 7h i permanent shielding for the reactor vessel annulus and < RCP motor hatches is underway. This shielding is expected to bring radiation levels down to or below 100 mrem /hr in normally accessible areas. I b I T l 4 i + ] k 1 o I i N
Page 75 6.5 Turbine Runback / Step Load Change Test 6.5.1 Purpose The objectives of the test were to determine the following: (1) Plant power levels at which TRAC I (negative startup rate), TRAC II (dropped CEA reed switch), and appli-cable turbine-Senerator undercurrent relays arm and disarm. (2) Time between receipt of turbine runback signal and arrival st regaired power. (3) That the Feedvater Regulating System maintains steam generator levels above the reactor pretrip setpoints and below the turbine trip setpoint. 6.5.2 Test Results The turbine runback was initiated from a load limiting condi-tion by placing a temporary jumper across one of the turbine runback control element assembly reed switch contacts. This caused a turbine runback to the point at which the runback circuit disarmed. The temporary jumper remained in place until turbine control using the load limiter had been estab-lished. A turbine runback was attempted on 5/6/75. This runback test was unsuccessful because the turbine generator undercurrent relays were improperly adjusted. The relays were set for 70% of turbine full load instead of the equivalent of 70% of 2611 Wth as specified in the Safety Analysis. The current setting for 70% of turbine full load exceeded the actual current value at 1828 Wth and the runback circuit was not armed. The relays were adjusted to the equivalent of 70% of 2611 Wth and the test was rerun on 5/18/75. The arming and disarming setpoints as determined by slowly varying reactor power are shown in Table 6.5-1. The runback I was initiated with reactor power about 715 by jumpering the dropped CEA reed switch for CEA 5-1. A runback resulted to 1574 w th or 61.5% reactor power and occurred at a rate of 0.75% power /sec. The turbine reached 61.5% reactor power at 27 seconds after runback was initiated. During the runback, steam generator levels varied between +12 inches and -16 inches. Variations of plant parameters for a 35 minute period following initiation of turbine runback are shown in Figures 6.5-1 through 6.5-5. The RCS Letdown Excess Flow Check Valve shut at two (2) minutes and five (5) seconds after initiation of the runback.
Paga 76 6.5.3 Conclusions The test adequately demonstrated the performance of the turbine runback circuitry. The steam generator levels remained in a relatively small band centered around a normal level of
- 0 inches. However, the runback rate was slower than specified in the acceptance criterion. As the turbine runback is a scheme for recovering from a dropped CEA incident, CE, Windsor will consider the as measured runback rate in their analysis of the dropped CEA test results (see Section 6.6) . /
l I l 1 /
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Page 78 . I 6.6 Dropped CEA. Power Asymmetry and Azimuthal Xenon Transient Tests ! 6.6.1 Purpose The test was designed to accomplish the following objectives:
- (1) Measure the asymmetric power distribution resulting from the drop of the most worthy Control Elenent Assembly (CEA).
(2) Determine the ability of the excore detectors to sense the drop of the most remote full length CEA. (3) Determine the dampening characteristics of an azimuthal xenon oscillation. 6.6.2 Test Results a The dropped CEA test consisted of one static and two dynamic rod drop tests. The static test was conducted to observe the static power tilt and resulting azimuthal xenon oscillations. The static rod drop measurement consisted of diluting CEA B-8, a dual shutdown CEA, to its full in position while holding
- AT power constant. Power tilt indication at the start of the test was 0.8%. The maximum power tilt occurred after CEA B-8 had been inserted approximately b hours, and was 16.5%.
CEA B-8 was then withdrawn and azimuthal xenon oscillations continued until dampened with time. See Figure 5.0-1 for relative locations of CEA's and excore detectors. Analysis of incore detector signals prior to and as close as possible
- to just after the static drop test indicated a greater than 12% change on at leact three (3) incore rodium detectors.
Two dynamic rod drop tests were conducted. CEA 5-58, a low reactivity worth, single CEA on the core periphery was used for the first drop and the center core CEA 5-1 was used for i the second drop. i When CEA 5-58 was dropped, excore nuclear instrumentation near the rod (Power Range Regulating Channel (PRRC) X) showed a drop of about 75 power, while instrumentation about 90 degrees away (PRRC Y) showed only about 1 5% power reduction. Power Range Safety Channels (PRSC) C and D showed dropped rod indi-cation with CEA 5-58. 1 i When CEA 5-1 was dropped, all four (k) PRSC's indicated a dropped rod almost inusediately and PRRC's X and Y indicated a decrease of 9% and 8% power respectively. 6.6.3 Conclusions Test results indicate that a dropped CEA would be detected and azimuthal power oscillations would converge. 4
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Paga 79 A preliminary on-site analysis of incore detector data indi-cated that the acceptance criterion for the change in incore detector signals before and after the static CEA drop had been exceeded. The situation could not be adequately resolved on-site. Consequently, administrative restrictions currently prevent operation above 50% reactor power with a dropped CEA. Another acceptance criterion was that CE, Windsor complete a i more detailed analysis of the data to determine maximum allowable power level with a dropped CEA. That analysis is in ' progress. The Safety Analysis reported in the FSAR was con-ducted at 100% power and assumed conditions other than those at which this test was conducted. f l 4 l l 1 j
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4 Page 80 6.7 Trip of Main Generator Breaker 6.7.1 Purpose The test was designed to evaluate system reliability during a loss of generator load from nominal 100% power. i 6.7.2 Test Results The trip was initiated by opening the generator output 500 KV breaker 552-23 from the main control panel. Reactor Coolant System (RCS) temperature and pressure indications were nominal for a reactor trip until about one minute after the trip. At that point, turbine bypass and atmospheric dump valve controllers called for zero demand and temperature continued to decrease.
! It was then realized that turbine bypass valves 39ho and 39hh
! vere not fully closed and/or vere leaking. Manual isolation i valves ahead of valves 39ho and 39hh were closed to terminate the cooldown, while retaining turbine bypass capability. How-
- ever, a shaft key was missing from one valve handwheel making the isolation extremely lengthy for 39h0. During the cooldown RCS temperature reached a minimum of h65'F, pressurizer pres-sure a minimum of 1200 psia, and steam generator pressure
; a minimum of 520 psig. As a result of the RCS depressurization, safety injection was automatically initiate'. Due to a leaky valve connecting the safety injection with the containment spray system several gallons of safety injection water were released into the containment. Reduction of secondary system pressure to 520 pais resulted in a Steam Generator Isolation Signal (SGIS) automatically shutting the main steam isolation ! valves. With the shutting of the main steam isolation valves, j the cooldown was terminated and a return to Hot Shutdown commenced. RCS temperature, pressurizer pressure and steam 1
generator level indication was back on scale or approaching
! Hot Shutdown conditions at the end of one hour from time of trip. Variations in plant parameters for one hour after the I trip are shown in Figures 6.7-1 and 6.7-2.
6.7.3 Conclusions Evaluation of data indicates that the lov levels in the pressurizer and steen generators did not cause any damage j to plant systems and the plant was safely returned to Hot
! Shutdown. The failure of the turbine bypass valves to close l~
vas due to cracked plugs in the valves and the valves have been repaired. The turbine bypass valves have continued to be a problem throughout the test program. Discussion of 1 design changes under consideration is contained in Section 6.1. 1 i
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Page 81 6.8 Xenon Follow Measurements 6.8.1 Purpose The purpose of this test was to obtain transient test data i at several test plateaus for the purpose of evaluating the { Shape Annealing Factor (SAF) for each Power Range Safety ! Channel and to evaluate an induced free xenon oscillation performed at the 80% test plateau. l 6.8.2 Test Results During the 80% test plateau, axial oscillations were induced i in the core. These oscillations were monitored by the plant i computer In Core Analysis (INCA) program and the Axial Shape j Index (ASI) calculated by the Reactor Protective System (RPS). 1 All axial oscillations were convergent and the minimum DNBR was 6.13 and the =v4-= measured Linear Heat Rate (LHR) was l 9.h7 kw/ft. When corrected for uncertainities, the maximum LKd was less than the acceptance criterion of ik.9 kw/ft. DNBR acceptance limit was 1.3. The SAF for each Power Range Safety Channel was also measured.
- The SAF corrects the detector signal to account for the dis-1 tance from the detector to the reactor core and corrects for i
the signal received by the upper detector from neutrons generated in the bottom of the core and the signal received by the lower detector from neutrons generated in the upper ,l part of the core. The SAF is determined by plotting ASI (INCA) versus the ASI (EXT) as read from the Reactor Pro-tective System (RPS) during a xenon oscillation with all CEA's
, full out. The slope of the line resulting from this plot is the Shape Annealing Factor. ASI is Axial Shape Index, a i ratio of the difference in power generated in the lower and j upper halves of the core to total core power. Table 6.8-1
- summarises the comparisons between measured and predicted l
! results. As measured SAF's were incorporated into RPS set- ! points. 1 TABLE 6.8-1 COMPARISON OF MEAstfRED AND PREDICTED SHAPE AlUtIALING FACTORS f t
- RPS CHANNEL N m fJRED PREDICTED l
l A 2.18 1.kT i j B 2.05 1.kT i
; C 1.60 1.kT D 1.88 1.kT i
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Paga 82 6.8.3 conclusions s The induced xenon oscillation test proved that induced xenon oscillations were self dampening to a stable power distri-bution and that resultant DNBR and LHR vere both well within acceptance limits. The ASI's calculated by the RPS as well as INCA was continu-ously observed and evaluated during EPT. Observation verified the adequacy of SAF's previously determined from nessured data. I i 4 1 I 1 ! I i 1
Pcga 83 6.9 Remote Shutdown Test 6.9.1 Purpose The purpose of the test was to demonstrate that the plant responds properly to a generator trip and that it can be safely maintained in the Hot Shutdown condition fran outside the contrcl room. 6.9.2 Test Results The generator trip was initiated from 50% power by simulating a generator fault at the Unit Protection Panel in the Cable Spreading Room. An emergency crew was used to monitor the control panels and to provide emergency control should the need have arisen to abort the test. The regular shift crev left the control room after verifying the trip and vent to the local panels to complete the test. RCS temperature was decreased to 527*F, held momentarily, then increased and maintained at approximately 537'F by local operation of the turbine bypass valves. Steam generator levels were maintained by local control of the Auxiliary Feedvater System. Variations of plant parameters for sixty (60) seconds following the trip are shown in Figures 6.9-1, 6.9-2A and 6.9-2B. 6.9 3 Conclusions The ability to safely shutdown and maintain the plant in Hot Shutdown from local panels was successfully demonstrated. Both the trip and subsequent cooldown progressed smoothly with no equipment damage.
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t Page 8h 6.10 Feedvater Regulating System Test 6.10.1 Purpose The purpose of this test was to monitor the operation of the Feedvater Regulatig System during power changes. 6.10.2 Test Results Operations of the Feedvater Regulating System was monitored throu6hout EFT during reactor scrams; rod drops; steady state operations; ramp load changes; and the Turbine Runback / Step Load Change test. During the rod drop test, it was noted that the steam generator levels increased to approximately 20 inches when turbine load was reduced about 10% to control Reactor Coolant. System (RCS) tersperature. Investigation showed that feedvater;flov 1;as not decreasing as fast as steam flow resulting in increasing steam generator levels. The setting of the proportional bands on the reedvater regulator valve controller was changed to increase response time. Monitcring of the system during scram,s and ramp , load changes after the above modification indicated satis- l I factory operation of the system. During ramp changes of between 0 5%/ min 1Eo 1.05/r.in, the Feedvater Regulating Systen maintained steam generator-levels "ith + 10 inches of zero (normal) level. 6.10.3 Conclusions Operation of the Feedvater Regulating System was acce'ptable-or when' discrepancies were noted, adjustments were made to provide acceptable results. r s 4 A a a
P gn 85 6.11 Loss of Off-Site Power with Coastdown 6.11.1 Purpose l The test was designed to accomplish the following objectives: (1) To evaluate plant systems reliability during a total loss of AC power. j (2) To demonstrate correct operation of the turbine generator assisted Reactor Coolant Pump (RCP) coastdown circuitry. i (3) To obtain Reactor Coolant System (RCS) flow data during ! the coastdown. 6.11.2 Test Results In order to defeat the automatic load transfer to live buses in case of a generator trip, Unit 1 house loads were trans-ferred to the Unit 1 side of the plant output ring bus and the ring bus was split. When the outgoing 500 KV breakers were tripped during the test, Unit 1 was isolated from the grid and depended on internal sources for electrice.1 power such as station batteries, diesel generators, and turbine generator assisted RCP coastdown. i The plant was manually tripped from 20% power using the trip pushbuttons on the Reactor Protective System Panels. Upon tripping of the turbine and generator, the electrically - assisted coastdown of the RCP's did not occur as scheduled. During turbine generator (TG) assisted coastdown, the RCP's and other plant equipnent are powered by the generator until coast down is terminated by either less than 80% generator output voltage or at 20 seconds following the trip whichever occurs first. TG assisted coastdown did not occur since the generator field breaker tripped inueedi- ' ately due to a misvired relay. The viring of the relay has been corrected to allow TG assisted coastdown following a l total loss of off-site power. After loss of off-site power, diesel generators 11 and 12 started immediately and diesel generator 11 loaded within 10 seconds after the trip. From a post trip time of 2 seconds until approximately 10 seconds, test instrtamenta- , tion recordings and observation indicated that the muriliary ' boiler, instrument air compressor, atmospheric duarp con- a troller, turbine bypass valve controller and some control , room instrumentation was lost. 1 Part of the indication, the instrument air compressor and i the controller output si6nals were restored when diesel ' generator 11 picked up 11 h K7 bus. The remainder of the I O
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