ML20062D895
| ML20062D895 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 11/12/1990 |
| From: | TEXAS UTILITIES ELECTRIC CO. (TU ELECTRIC) |
| To: | |
| Shared Package | |
| ML20062D894 | List: |
| References | |
| NUDOCS 9011190121 | |
| Download: ML20062D895 (209) | |
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t 1 1: - i. 8 t TU ELECTRIC COMPANY' 4 COMANCHE ~ PEAK STEAM ELECTRIC S. .ON - UNIT 1~ CYCLE l'
. l' STARTUP REPORT f
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l; J o TABLE OF CONTENTS q Section Title Page q Title Page' -1 ; Table of' Cont'ents.- 2- ; List of Tables 3~ List of Figures 5
- 1. 0 - Introduction. 7 2.0 Discussion'of'the Initial Startupp Program. -10 3.0 Discussion:ofithe Initial Startup Tests 26-3.1 ~ Core Loading .28' 3.2- System' Testing After: Core Load 47 ,
and'at Various: Power Levels'
.l 3.3 ~ Physics Testing 96 ,
i 3.4 Transient Testing '123 1
-q 3.5 Instrumentation and Calibration Testing 1147J ,
_i 3.6 Deferred Preoperational Testing '194- l 4.0 References 209 q AttachmentzA Comanche Peak Steam Electric Station Unit l j Loose Parts Monitoring l System =Special' Report 1 t 4
1 LIST OF TABLES ( TABLE TITLE H9E 1.0-1 Cross Reference of FSAR Table 14.2 8 '!' and-Unit l' Cycle'l Startup Report 2.0-1 Comanche Peak Unit 1 Major-Milestones 11 l 2.0-2 Comanche Peak Operational Modes; ik 1 I I- 2.3-1 HZP Physics Testing Results 18 i 2.4-1 50% Power Flux Map Results' 120 1 l 2.5-1 75%-Power Flux Map Results 23 i l t 2.6-1 100% Power Flux Map Results 25 l l 3.0 List of Test Summaries- ' 26: it 3.2.2-1 Steam Generator Level; Control' Summary 51'- s i 3.2.5-1 RCS Chemistry Summary 61 '{
-i 3.2.5-2 Steam GeneratorfChemistry Summary -62 3.3.9-1 Measured and Inferred vs. Predicted 1 Rod' l Bank Worths- 120.
3.4.2-1 Design Load Swing Tests Summary 129 3.4.2-2 10% Load Decrease at 35%' Power Summary 130 3.4.2-3 10% Load-Increase at 35% Power Summary '131 ; Q 3.4.2-4 10% Load Decrease at 50% Power Summary 132 L 3.4.2-5 10% Load-Increase at 50% Power Summary 133 l i 3.4.2-6 10% Load. Decrease.at 100%, Power 1 Summary 134 } F 3.4.2-7 10% Load Increase at 106 4 ~ Power: Summary 135
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3.4.3-1 Trip From 100% Power: Summary 138-3.4.5-1 Large Load Reduction Tests. Summary- 144
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3.4.5-2 Large Load Reduction From 75% Power: Summary: 145 3.4.5-3 Large Load Reduction From 100% Power : Summary 146 , c k
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w LIST OF TABTRA (Continued) TABLE TITLE EAEE 3.5.1-1 Calibration of Steam Flow Transmitters 352 3.5.3-1 Process Temperature /N16-Tests vs. Plant Conditions Matrix 164 T.4-1 Nuclear Instrumentation Results Summary 168'
. 5-1 Incore/Excore Detector Calibration' Summary 176 e . 5.7-1 Startup Adjustments Summary- 184 2.5.9-1 Process Computer Algorithm Comparisons '191 3.6.1-1 Process Sampling System Summary 196 3.6.2-1 In-Place Atmospheric cleanup Filter Test Summary 2001 i
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- LIST OF FIGURES
l l FIGURE TITLE PAGE 2.0-1 ISU Program Summary 13 l t l 3.1.1-1 Unit 1 Core Loading Pattern - 30 Initial Nucleus of Assemblies . 3.1.1-2 Unit 1 Core Loading' Pattern - 31 Partial Bridge Across Core ; 3.1.1-3 Unit 1. Core Loading Pattern - 32 ' Completed Bridge Across core i 3.1.1-4 Unit 1 Core Loading Pattern - 33 ' Partial Completion of Core 3.1.1-5 Unit 1 Core Loading Pattarn - 34 { Partial Completion of Core 3.1.1-6 Unit 1 Core Loading Pattern - 35 Final Configuration 3 1.1-7 Shutdown and Control Rod Locations 36 , 3.1.1-8 Burnable Poison Rod Locations 37 3.1.3-1 ICRR vs Fuel Assemblies ' Loaded - 42 , Source Range Channels 3.1.3-2 l'CRR vs Fuel Assemblies Loaded - 43 Temporary Detector Channels 3.2.4-1 Movable Incore Detector-Path Locations 59 i 1 3.2.5-1 Lithium vs. Boron Curve 63 t 3.2.14-1 Pressure Response to Opening Both Spray' Valves 87 3.2.14-2 Pressure Response to Actuating All 9',dters 88 3.3.2-1 ICRR During RCC Bank Withdrawal 101 3.3.2-2 ICRR vs Time-During RCS Boron Dilution 102: 3.3.2-3 ICRR During RCS Baron Dilution 103-3.3.7-1 Reactivity vs. Temperature 113-i I
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n ~s. x + - an .-a x- -.. -a , - n - 1 i LIST OF FICUEEE ! (Continued) ! FIGURE TITLE PAQR i 3.3.8-1 Rod Withdrawal Limits 116 !
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3.3.9-1 Differential and Integral. Rod Worth 121 l Rod Swap Referenct Bank l 3.5.4-1 Power Range Current vs. Calorimetric Power 170 l 3.5.4-2 Intermediate Range Current vs. Calorimetric Power 171 j i 3.5.5-1 Incore/Excore Calibration - Plot of AFD vs. Time 177
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3.5.5-2 Incor:. /Excore Calibration - Control Bank D .178 Position vs. Time : 3.5.5-3 Example Axial Flux Difference Calibration 179 f
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{ 3.5.7-1 Tref, Pimp and Pressurizer Level vs. Power 185 l 3.5.7-2 Steam Pressure vs. Power 186 , 1 s 1
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i i 1.0 - INTRODUCTION ' This report describer the required testing at Comanche Peak Steam . Electric Station, Unit 1, from the preparations for loading the first fuel assembly into the reactor until the plant was placed in ! commercial operation. It satisfies the requirement of the Comanche Peak Technical Specifications that a Startup Report be submitted to .; the NRC after completion of the Startup Testing Program. Comanche Peak Steam Electric Station, located in North Central 1 Texas, utilizes a four loop Westinghouse Pressurized Water Reactor - , as the Nuclear Steam Supply System. Westinghouse Electric Corporation, Stone & Webster Engineering Corp., Gibbs & Hill, Inc. Impell Corp., Ebasco, Brown & Root, Inc. ' 1 the TU Electric ; company jointly participated in the design Ad construction of l Comanche Peak. The plant is operated by the TU Electric Company. ; The Nuclear Steam Supply System is designed for a thermal power . output of 3425 MWth (3411 MWth reactor power) . The equivalent warranted gross electrical output is 1163 MWe. Cooling for the plant is provided by the Squaw Creek Reservoir, a 135,062 acre-foot , man-made lake. Post design basis accid.ent cooling is provided by a separate 367 acre-foot Safe Shutdown coundment. Table 1.0-1 provides a cross reference between the test $mmaries in the Final Safety Analysis Report and the sections or +his report. i l l i I I
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l > Table 1.0-1 l l ; Cross Reference of FSAR Table 14.2-3 , and Unit 1 Cycle 1 Startup Report FSAR Table 14.2-3 Startup SHEET NUMBER TITLE Report Section 2 Reactor Coolant System Flow Test 3.2.8, 3.2.14 ! 3 Reactor Coolant System Flow 3.2.9 l Coastdown Test 4 Control Rod Drive Tests 3.2.11, 3.2.12, 3.5.10 5 Rod Position Indication 3.2.11 6 Reactor Trip System 3.2.13.
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8 Auxiliary Startup Instrumentation Test 3.1.2 9 Calibration of N'uclear Instrumentation 3.5.3, 3.5.4 11 Chemical Tests 3.2.5 12 Radiation Surveys- 3.2.6 1 13 Process and Effluent Radiation Monitoring Test 3.2.7 l 14 Moderator Temperature Reactivity l Coefficient 3.3.7, 3.3.8 15 Control Rod Reactivity Worths 3.3.9 16 Boron Reactivity-Worth 3.3.9, 3.3.10 17 Core Reactivity Balance 3.3.5 18 Loss of Offsite Power 3.4.1 l l 19 Rod Drop Tests 3.2.12 20 Flux Distribution Measurements 3.3.6 22 Core Performance Evaluation- 2.4, 2.5, 2.6, 3.3.6 3.5.3, 3.5.5 l 23 Unit Load Transients- 3.4.2, 3.4.3, 3.4.5 25 Remote Shutdown 3.4.4 l
i Table 1.0-1 Cross Reference of FSAR Table 14.2-3 and Unit 1 Cycle 1 Startup Report (Continued) PSAR Table 14.2-3 Startup SHEET NUMBER TITLE Report Section 28 Turbine Trip / Generator Load Rejection 3.4.3 29 Reactor Coolant Leak Test 3.2.10 31 Rod Control System Test 3 . 2 .' 12 ' 4 33 Automatic Control System Test 3.5.10 34 Incore Nuclear Instrumentation 3.2.4 i i l l l l
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3 t 2.0 - DISCUSSION OF THE INITIAL STARTUP PROGRAM The Comanche Peak Unit l' initial startup testing program consisted i of single and multi-system tests that were performed commencing ; with initial fuel loading and continuing through full power i operation. The intent of these tests is to assure that tests , deferred from the preoperational test program are performed; that the plant is safely brought to rated capacity; that plant , performance is satisfactory in terms of established design- -) criteria; and to demonstrate,' where practical, that the plant is ; capeble of withstanding anticipated transients and postulated ' accidents. These tests demonstrated overall 71 ant performance and ' I included such activities as precritical test:,ng,-low power tests, , and power ascension tests. Testing sequence documents were .- utilized for each plateau to coordinate the sequence of testing activities at that plateau. i In the subsectionc that follow, a description of the testing at each plateau is provided. The descriptions include additional details concerning special license conditions and commitments made , to the Nuclear Regulatory Commission prior to completion .of the startup testing program, where applicable. Alsc, included as a part ' i of Section 2.0 are tables and figures showing major milestones for Comanche Peak Unit 1 which occurred during the initial starttip - program and a list of operational modes as defined by the Technical Specifications. l 1 l
i I&BLE 2.0 - 1 COMANCHE PEAK UNIT 1 MAJOR MILESTONES MAJOR MILESTONES DATE 5% Power License Received 2/08/90 1 Fuel Load Started 2/09/90 Fuel Load Completed 2/14/90 Initial Criticality 4/03/90 5% License (Low Power) Tests Completed . 4/06/90 Full Power License Received 4/17/90 , 1 Entered Mode 1 4/19/90 l l Initial Synchronization to Grid 4/24/90 l 30% Power Reached 4/30/90 - 50% Power Reached 5/04/90 75% Power Reached C/27/90 100% Power Reached 7/13/90 Test Review Group Approves Startup Test Program 7/30/90 i V. P. Nuclear Operations Declares Completion of Startup Test Program and Commencement =of Commercial Operation 8/13/90 i l I l l 1 l
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i TABLE 2.0 - 2 i OPERATIONAL MODES REACTIVITY % RATED AVERAGE COOLANT i MODE CONDITION. Kaff THERMAL .. POWER *. TEMPERATURE l
- 1. POWER OPERATION 1 0.99 > 5% 1 3 50*F -l
- 2. STARTUP 2 0.99 $ 5% 1 3 50'F . I
- 3. HOT STANDBY < 0.99 ^- 1 3 50'F . [
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- 4. HOT SHUTDOWN < 0.99 0 3 50'F >T,, >2 00'F ;
- 5. COLD SHUTDOWN < 0.99 0- $ 2 0 0'F -
- 6. RETUELING** $ 0.95 0 . 5 14 0'F t
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2.1 - INITIAL FUEL IDAD SEQUENCE - ISU-001A I OBJECTIVE The Initial Load Sequence document defines the sequence of. testing j and other operations to prepare for ' and perform initial core ( loading. This test partially satisfies activities described in FSAR Section 14.2.10.1. TEST METHODOIDGY The Fuel Load Sequence Document is used to coordinate the sequence of operations associated with the initial core-. loading- program.
; This sequence includes scheduling of the individual startup tests and selected key permanent plant procedures associated with core loading. This document specifies as prerequisites which testing. i had to be completed prior to commencement of core loading, the-required status of the plant systems necessary to support core loading, and the reactor vessel status. A log is also included in the sequence document as Form ISU-001A to verify Technical Specification compliance prior to and throughout core loading.
This document also provides the criteria for stopping core loading, the criteria for emergency boration, and the actions to be followed prior to the resumption of core loading in the event loading was stopped prior to completion.
SUMMARY
OF RESULTS Initial core loading.of 193 fuel assenblien took 121 hours. Prior to the start of core loading, the condition of the reactor vessel and associated components, the reactor coolant system, instrumentation, and administrative controls were verified to be acceptable. The sequence procedure verified that reactor coolant system chemistry was properly established and maintained and l verified timely nuclear instrumentation neutron response checks. ; The procedure also ensured that a final' general fuel assembly j visual inspction was performed. Fuel loading ' operations were performed using permanent plant procedures. Results of individual tests completed during the core loading sequence are discussed in Section 3.1 of this report. Upon completion of core loading, plant systems were aligned as directed by the Shift Supervisor. l I 1 1 l l l
i 2.2 - POST CORE LOAD PRECRITICAL TEST SEOUENCE (PCLPC) - ISU-010A { OBJECTIVE ! The PCLPC Sequence Document defines the sequence of tests and operations to be performed between completion of initia) core loading and prior to initial criticality. This testing is performed in Technical Specification Modes 5, 4 and 3. j TEST METHODOL%Y This document ensures that core load testing had been successfully completed and results approved prior to continuation of the testing l program. This document schedules the. performance of procritical tests to ensure the necessary testing was completed prior ' to ; initial criticality. This procedure governs the sequence of testing through Modes 5,4 and 3. Plant operating procedures are utilized where appropriate to establish necessary plant conditions. i
SUMMARY
OF RESULTS f Results of individual tests completed during the post core load . precritical testing phase are discussed-primarily in Sections 3.2-l and 3.6 of this report. A daily log of RCS and pressurizer boron-l concentration was kept to ensure adequate shutdown margin during l testing. Boron concentration varied between 2022 ppa and 2124 ppm. l This insured that the boron concentration was greater than the 2000 ppm refueling concentration at all times. Upon completion of this ' testing phase, plant systems were aligned as directed by the Shift: Supervisor. l t f-l l l L l I _ _ _ _o
l ( ( 2.3 - INITIAL CRITICALITY & IDW POWER TEST SEOUENCE (IC & LPT) - ISU-101A OBJECTIVE The IC & LPT Sequence Document defines the sequence of tests and operations, beginning with initial criticality, which constitute the low power physics testing program. This program of low power physics testing verifies the design of the reactor by performing a series of selected measurem&nts including core flux distributions, control bank worths and moderator temperature coefficient. This test sequence partially satisfies activities described ,in FSAR Sections 14.2.10.2 and 14.2.10.3. TEST METHODOLOGY This document ensures that post core loading precritical testing has been completed and results approved prior to continuation of the testing program. Prior to commencement of. dilution to initial criticality, source range nuclear -instrumentation channels are , verified to have a signal to noise ~ ratio greater than 2 and power , range high level trip setpoints are conservatively set to 5 20% of l full power. This procedure sequences the low power physics testing into an efficient order and ensures that all required testing is performed. Surveillance Requirements for Technical Specification 3.10.3 usage are also controlled by this test. This Technical Specification Special Tert Exception permits physics testing in non-normal operating reactor controls configurations. A reactivity computer is set up using a power range NIS channel detector output to monitor core flux. This device is an analog i computer that calculates the amount of reactivity present in the l core based on the time dependence .of core flux. This device is used in core physics testing to make measurements of control rod, boron concentration and moderator temperature worths. A low power flux mapping system is used to augment the installed flux mapping system. This low power system contains very low noise bias voltage supplies and sensitive signal detection instrumentation to permit fltm mapping.below the point of adding nuclear heat, thus avoiding xenon and power stability effects on flux map results. Plant operating procedures are utilized where appropriate to establish and maintain plant conditions.
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- 2. 3 - INITIAL CRITICALTTY & IDW POWER TEST SEOUENCE (IC & I M - l ISU-101A (Continued)
SUMMARY
OF RESULTS i This sequence document obtained a full core flux map at the Hot l Zero Power, Xenon-free, All Rods Out Condition. Refer to Table 2.3-1 for a tabulation of the flux map results obtained. The low ; power flux mapping system was successfully used- to obtain this map. These map results were of sufficient quality to eliminate the need to perform a 30% power flux map. Results of individual tests i completed during the initial criticality and low power test i sequence are discussed primarily in Section 3.3 of this report. A . tabulation of key physics measurement results is also included in l Table 2.3-1. All required tests were performed. Initial criticality was achieved without incident on 4/3/90. A low' power physics testing power range . was determined and the reactivity computer was verified to be operating properly. Boron endpoint , concentration measurements were performed and data'was taken for , later comparison ' with 100% power data to.~ verify proper . power l defect. The Moderator Temperature coefficient was then-measured and found to be positive. This was not unexpected based on information in the reactor Nuclear Design Report'(WCAP-9806). 'In response to this positive' coefficient, rod withdrawal limits were i imposed using a permanent plant procedure, NUC-116. . Control rod worths were verified using the bank exchan's method (rod - swap i method). All required testina,was completed. Upon completion of.this testing phase, the plant was aligned as directed by the Shift Supervisor. l l i i
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I 1 TABLE 2.3-1 j MMSilLTit FLUX MAP RESULTS Actual Maximum Limit i Reaction Rate Error 8.62% 10% FDHN 1.605 1.643 FQ(Z) 2.5203 3.314 x K(Z) Quadrant Power 0.9828 1.0109 1.04 j Tilt Ratios 0.9828 1.0235 MISC. PHYSICS TESTING RESULTS Actual Allowed Range i All Rods Out Critical Boron (ppm) 1162.1 1096 to 1196= Reference Bank In Critical Boron (ppa) 1087.0 1012 to 11.2 ( Isothermal Temperature -0.995 -4.4 to +1.6 Coefficient (pcm/*F) Moderator Temperature +0.835 <0 (unless rod i Coef ficient(pcm/'F) withdrawal limits , are set). Reactivity Computer Error 0.81% $ 14% Source Range / Intermediate Range NIS Overlap (decades) 1.6 2 1.5
- Reference Rod Bank Worth Error -0.4% $ 110%
All Other Banks Worth Error (max.) -13.4% 5 115% Total Rod Bank Worth Error + 1.8% -10% to +7% - Differential Boron Worth (pcm/ ppm) -11.63* -9.40 to -11 18
- Refer to Test Summary 3.3.7, NUC-120, for discussion of this out of range value.
NOTE: pcm means percent millirho, equivalent to a teactivity value of 10*hhK/K l i 4 1
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i l 2.4 - 50t REACTOR' POWER TEST SEQUENCE - ISU-24OA OBJECTIVE The 50% Reactor Power Test Sequence document defines the activities which constitute the startup testing program between 0% and 50% power and at approximately 50% of rated thermal power. This test partially satisfies activities described in FSAR Table 14.2-3, Sheet 22 and Section 14.2.10.4. TEST METHOD 014GY This document ensures that the low power physics testing has been completed and the results approved prior to ' increasing power. Prior to incrassing power for this test sequence, power range high ( level trip setpoints are conservatively set to S 70% power and reactor core flux map results from a 0% power baseline map are verified acceptab.'e. The flux map results are also extrapolated to 70% power to ensure parameters indicative of DNBR and 1:near heat. rate are acceptable for power ascension to the 50% testing plateau. Plant operating procedures are utilized where appropriate to establish- plant conditions and to change reactor power. During this testing sequence following completion of 50% power testing, power is stabilized near the 30-35% and 20-25% levels to accommodate testing at those power levels.
SUMMARY
.0F RESULTS Results of individual tests completed up to and while at the 50% ; l power plateau are discussed primarily in Sections 3.2, 3.4 and 3.5 of this report. Administrative hold points-on' continued testing were observed at 10%, 20% and 30% power during the initial power I ascension to 50% power. A flux map was taken at 47.55% power with l satisfactory results as summarized in Table 2.4-1. All required testing was completed. Upon completion of this testing phase, the plant was aligned as . directed by the Shift Supervisor. a i
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I-I l TABLE 2.4-1 1 i 50% POWER PLUX MAP DERULTS i i Actual Maximum Limit [ Reaction Rate D ~ 8.854 10% l FDHN 1.4994 1.7126 i FQ(2 2.1131 4.64 x K(Z) .; Fxy - unrodded 1.6098 1.7126 Quadrant Power Tilt Ratios 1.0005 1.0100 1.02 ! 0.9840 1.0055 f t 6 9 l I f l r i e,-, . . e, + . , -+ -. , . . - , , , . . . . , . . - . ,n.. ..e.,._. - ~ . ~ . . ~ . .
l t 2.5 - 75% REACTOR POWER TEST SEOUENCE - ISU-260A I i OBJECTIVE l The 75% Reactor Power Test Sequence document defines the activities i which constitute the startup testing program during escalation from 50% to 75% power and at approximately 75% of rated thermal power. This test partially satisfies activities described by FSAR Table 14.2-3, Sheet 22 and Section 14.2.10.4. l i TEST METHOD 01DGY ' This document ensures that the 50% Reactor Power Test Sequence has been completed and the results approved prior to increasing power ; above the 50% testing. plateau. Prior to increasing power for this test sequence, power range high level trip setpoints are conservatively set to S 95% power and reactor core flux map results from a 50% power baseline map are verified acceptable. The flux map results are also extrapolated to 95% power to ensure parameters ! indicative of DNBR and linear heat rate are acceptable for power ascension to the 75% testing plateau. Plant operating . procedures are utilized where appropriate to establish plant conditions and to change reactor power.
SUMMARY
OF RESULTS Results of individual tests completed while at the 75% plateau are discussed primarily in Sections 3.2, 3.4 and 3.5 of this-report. > The extrapolation of the 50% power plateau flux map results to 95% power indicated that the Fxy peaking factor limit would be exceeded ! at 95% power. The Fxy extrapolation was acceptable for power ! levels up to 94.5%. Reactor Engineering, the TU Electric reactor core design group, evaluated this item and concluded that adequate ' l FQ(Z) margin existed because while Fxy is used as a Technical Specification Surveillance parameter to ensure adequate FQ(z) margin, this use of Fxy assumes a certain operationally varying l axial power distribution, F(z). Because the plant was in a power ascension program instead of a. load follow operating regime, F(z) values are not as large as are assumed in determination of Fxy limitations. Additional justifications for accepting the flux map t extrapolated results as sufficient for ensuring safe operation were i that the measured Fxy is typically observed to decrease with power ! increase and based on Fxy limit satisfaction up to 94.5% power which was judged to be sufficiently close to 95%. Another flux map t was taken at approximately 67% power to further confirm peaking i factor behavior. These results were satisfactory. l t j
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1 .i 2.5 - 75% REACTOR POWER TEST SEQUENCE - ISU-260A (Continued)
SUMMARY
OF RESULTS (Continued) Heater Drain system and Moisture separator Reheater 2-B Main Steam l Sample flows, temperatures and pressures were also. verified ' acceptable to close a testing item carried over from the preoperational test program. This testing was non-safety related and was not a deferred preoperational test requirement. i h flux map was taken at 77.43% power with satisfactory results as 1 summarized in Table 2.5-1. These results were of sufficient > quality such that a 90% power flux map was not required. All ; required testing was completed. Upon completion of this testing phase, the' plant was aligned as directed by the Shift Supervisor. r 1 s l l I l 1 l \ 1 1
5 l i i .i l i TABLE 2.5-1 ! I 75% POWER FLUX MAP RESULTS i Actual Navinum Limit i 1 Reaction Rate Error 5.98% 10%' I FDHN 1.4383 1.6200 j FQ(Z) 2.0659 2.9963 x K(Z) . Fxy - unrodded .1.5266 1.6200 i Quadrant Power Tilt Ratios 0.9987 1.0043 1.02 O.9904 1.0066 i
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i i 2.6 - 100% REACTOR POWER TEST SEOUENCE'- ISU-280A OILTECTIVE The 100% Reactor Power Test Sequence document defines the ! activities which constitute the startup testing program during! - escalation from 75% to 100% power and at close to,.but not more : than, 100% of rated thermal power. This test partially satisfies - activities described in FSAR Table 14.2-3, Sheet 22 and Section i 14.2.10.4. i TEST METHODOLOGY l I This document ensures that the 75% Reactor Power Test sequence has been completed and the results approved prior _to increasing power above the 75% testing plateau. Prior to increasing power above 75% - ; for this-test. sequence, reactor core flux map results from a 75%. . power baseline map are verified acceptable and the power range high level trip setpoints are set to S .109%, - their normal Technical
- Specification values. The flux map results are also extrapolated to 100% power to ensure parameters indicative of DNBR and linear heat rate are acceptable for power ascension to the 100% testing i plateau.
Plant operating procedures are utilized where ' approcriate to establish plant conditions and to change-reactor power. During ascension to the 100% plateau, power is stabilized near the 90% and 98% levels to accommodate testing at those: power levels.
SUMMARY
OF RESULTS i Results of individual tests completed during this power ascension and while at the 100% plateau are discussed primarily in Sections 3.2, 3.4 and 3.5 of this report. A flux map was taken at 99.03% power with satisfactory results, as i summarized in Table 2.6-1. All required testing was completed. 7 Upon completion of this testing phase, the plant was' aligned as i directed by the Shift Supervisor. l 1 i u l
I a i 1 TABLE 2.6-1 I 100% PO)fER FLUX MAP RESULTS Actual Maximum Limit Reaction Rate Error 6.81% 10% FDHN 1.4504- 1.553 FQ(Z) 2.0449 2.34 x. K(2) Fxy - unrodded 1.5475 1.553 Quadrant Power Tilt Ratios 1.0030 1.0067 1.02 , 0.9887 -1.0016 l
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l ! i l l 3.0 DISCUSSION OF THE INITIAL STARTUP TESTS I I l TABLE 3.0-1 ] List of Tant Su==mriaa j 4 3.1 CORE I4ADING I 3.1.1 Development and Implementation'of the Reload Fuel -I Shuffle Sequence Plan, RFO-!.6
- 3.1.2 Core . Loading Instrumentat.on and Neutron. Source checks, ISU-003A q 3.1.3 Inverse Count Rate Ratio ' Monitoring (Core Load Portion), NUC-111 ,
(Core Load 3.1.4 RCS and Secondary Coolant Chemistry Portion), ISU-006A : 3.1.5 Verification of Core' Loading Pattern, RFO-204 3.2 SYSTEM TESTING AFTER CORE I4AD AND AT VARIOUS POWER LEVELS 3.2.1 . Piping Vibration MonitoringiLISU-212A , 3.2.2 Stein Generator Level control. Test, ISU-207A . 3.2.3 Thermal Expansion, Power Ascension Phase, ISU-308A : 3.2.4 Incore Moveable Detector' System Alignment,'ISU-016A . 3.2.5 RCS and Secondary ' Coolant Chemistry (Post Core Load), ISU-006A 3.2.6 Radiation Survey Tests, ISU-208A' 3.2.7 Process and' Effluent Radiation, . Monitoring Performance Test, ISU-210A 3.2.8 Reactor Coolant Flow Measurement, ISU-023A 3.2.9 Reactor Coolant System Flou coastdown Test, ISU-024A 3.2.10 Reactor Coolant System Leakage-Rate Test,.ISU-022A l 3.2.11 Cold Control Rod Operability. Testing,-ISU-026A 3.2.12 Hot Control Rod Operability Testing, ISU-027A > 3.2.13 Reactor Trip System Tests, ISU-015A . 3.2.14 Pressurizar Spray and Heater. Capability,'ISU-021A 3.2.15 Miscellaneous. Balance of Plant Testing 3.3 PHYSICS TESTING : i 3.3.1 Inverse Count- Rate Ratio n Monitoring. (Initial ' Criticality Portion), NUC-111 . 3.3.2 Initial Criticality, NUC-106 i Determination of Core . Power Range- for. Physics
~
3.3.3 Testing, NUC-109 3.3.4 Reactivity Computer Checkout, NUC-108 , 3.3.5 Core Reactivity Balance,'NUC-205 3.3.6 Surveillance . of Core ~ Power' Distribution Factors, NUC-201 3.3.7 Zero Power Isothermal . and Moderator Temperature t Coefficient Measurements, NUC-207; I
' 3.3.8 Determination of Operating Limits to Ensure a- ,
Negative MTC, NUC-116-
\
55- e - ,.,s--,,.+ - - . .' ,.-w .-my..%,_ , . , , _ - * ,re-, ,%.-5._.-_ - t 4 - +m--
- TABLE 3.0-1 (Continued) 3.3.9 Rod Swap Measurements, NUC-120 3.3.10 Boron Endpoint Determination and Differential Boron Worth, NUC-104 3.4 TRANSIENT TESTING 3.4.1 Turbine Generator Trip With Coincident- Loss of Offsite Power, ISU-222A 3.4.2 Design Load Swing Tests, ISU-231A ,
3.4.3 Dynamic Response to Full Load Rejection and Turbine i Trip, ISU-284A 3.4.4 Remote Shutdown capability. Test, ISU-223A 3.4.5 Large Load Reduction Tests, ISU-263A 3.5 INSTRUMENTATION AND CALIBRATION TESTING 3.5.1 Calibration of Feedwater .and Steam Flow Instrumentation at Power,' ISU-202A 3.5.2 Thermal Power Measurement and Statopoint- Data Collection, ISU-224A 3.5.3 Operational Alignment of Process Temperature and N16 - Instrumentation, ISU-226A 3.5.4 operational Alignment of Nuclear Instrumentation, l ISU-204A 3.5.5 Incore/Excore Detector Calibration, NUC-203' ; Loose Parts Monitoring Baseline Data, ISU-211A' 3.5.6 l 3.5.7 Startup Adjustments of Reactor Control Systems, j ISU-020A i 3.5.8 Full Power Performance Test, ISU-281A 3.5.9 P2500 Process Computer Software Verification, ISU-019A 3.5.10 Automatic Reactor Control System. Test, ISU-203A 4 3.6 DEFERRED PREOPERATIONAL TESTING 3.6.1 Process Sampling System, ISU-028A, . . 3.6.2 In-place Atmospheric Cleanup Filter Test - Primary , Plant - ESF, EGT-751X . 3.6.3 Containment & Penetration Rooms Temperatura Survey, ISU-282A 3.6.4 Turbine Driven Auxiliary Feedwater Pump Actuation and Response Time Tests, EGT-76BA and EGT-769A 3.6.5 MSIV Isolation Response Time Tests, EGT-764A and EGT-765A. 3.6.6 Reactor Coolant System Pressure Isolation Valve- i Leakage Testing, EGT-712A 3.6.7 Condensate Reject Valve. Test, EGT-TP-90A-002 I l u j
I
-l l
1_1 CORE IDADING ) 3.1.1 - Develooment and Imolamentation of the Reload Fuel Shuffle Secuence Plan. RFO-106 , OBJECTIVES This permanent plant procedure is performed to ensure that the 1 nuclear fuel assemblies are loaded in a safe and cautious manner. j This procedure pa-tially satisfies activities described in FSAR Section 14.2.10.1. j 1 TEST METHODOI43Y The procedure is performed prior to the start of core loading to develop the detailed core loading sequence sheets.- Field-use of 1 the procedure begins following loading of the temporary core l loading instrumentation hto its initial position and determination of background count rh as for all source range and temporary nuclear instrumentation channels. The four primary source bearing assemblies and six additional' assemblies, comprising the " source i i nucleus", are loaded. Audible indication of neutron-population j changes from one of the two installed source range plant channels - . is required to be maintained in- both the control room and j containment for the duration of the core loading process. After , the source nucleus assemblies are loaded, count rate data is taken. for the nuclear channels usel in toe. core loading process (two i source range and three temporary channels) . The=first' reference value, for use in inverse count' rate ratio monitoring, is ' , determined from these counts after the appropriate background values have been subtracted. Subsequent reference values are calculated whenever core loading is. suspended _for eight hours or . longer, a temporary detector is moved, or a primary source bearing fuel assembly is moved to a different core location. i Prior to fuel load, predictions were made for comparison to actual nuclear instrumentation response, to verify that the reactor would' remain shutdown throughout _the loading process. Inverse count rate ratio monitoring is used. following each fual assembly move to ; ensure that the reactor is not approaching criticality. To ensure l reliability in the monitoring, a minimum of two of the five nuclear ~ l instrumentation channels are required to be responding to source I neutron population changes throughout core loading.- Data obtained during inverse count rate ratio monitoring is trended- and extrapolated forward to- permit -evaluation of any ' indicated .,l criticality approach. Plant procedure NUC-111 is used to perform I 1 the inverse count rate ratio measurements-and extrapolations. l l l _ ____._- __________i___._..._
i l 3.1.1 - Development and Implementation of the Reload Fuel Shuffle gaggance Plan. RFO-106 (Continued) ;
SUMMARY
OF RESULTS , Core loading was completed in a safe and cautious manner as required by the acceptance criteria of the core loading procedure. Problems encountered during the test were primarily associated with i readjustment of the source range NIS high flux at shutdown alarm bistables and actuation of the high flux at shutdown alarms. The ! alarm actuated several times, due primarily to the presence of 4 + Californium primary neutron sources and associated stronger source i to detector couplings. Due to a several year delay in actual core 1 loading, the original two primary sources that had been received , were augmented with two fresh sources to ensure a sufficiently-high ! neutron count rate for core loading and initial criticality nuclear monitoring. The high flux at shutdown alarm also actuated once in f response to source range spiking caused by high voltage switching i in the main switchyard. Source range NIS channels also lost power twice during initial core ! load. These losses of power were unrelated and not coincident. + The N31 channel power loss was caused - by an inverter breaker failure. The inverter problem resulted in a delay of greater. than 8 hours in core loading, so all neutron monitors were again , l response tested. The A32 power loss was' caused by an error in the l switching of the Solid State Protection System. Power was , immediately rectored. The fuel handling equipment performed very well, with only one ! failure. The manipulator crane (refueling machine)' gripper jammed once and was mechanically freed with vendor assistance.- It jammed in the unlatched position and not while it was gripping a fuel , assembly. All 193 fuel assemblies were loaded in the core without incident. Fuel assembly A21, however, brushed against a new fuel storage ! vault lid when withdrawn fromfits storage rack location. The slight scratch-on the. fuel assembly bottom nozzle was inspected, blended, and evaluated as acceptable with vendor assistance. There was no damage to any of the fuel pins. Refer to Figures 3.1.1-1 through 3.1.1-6 for a graphical 1 description o how core loading progressed. Refer to - Figures 3.1.1-7 and 3.1. -8 for.-inf ation on the locations of control rods and burnable , M .emblies. e v g- -.-.-,e- ,--,+r ~ ,~~ v.-, , ,- , -+,,ae e. ~.- _m_E _ - - _ _ _ _ _ _ _ _ .- _
- . _ . _ - - _ ~ p-- _- -__ -----
1 P Figure 3.1.1-l' l UNIT -1 CORE ICADING PATTERN f I INITIAL NUCLEUS OF ASSEN8 LIES : i : i SR , 31 < t N M L K. J H G F 'E D' C' 9 A R P .4 i 4 4 4- 4 4 4 1 --, , , ,4 CSO C12 C07 C50 ,,, ,, 2- ,,
+ C01 A22 C04, A10 <-
3 : ' l 4 ,,
+ ,,
5 ; C31 e i . l' S : 7 : T 8 : 9 - 10 -
-11 :
L r 12 -+ l 23 -
- C30 14 e L
iS : SR 32 A - REGION 1 C1. 6 "/o) B ' - REGION 2 C 2. 4 */o) C - REGION 3 C 3.1 "/o)
-a' - SOURCE ROD r
e
+ ---, ,.,+ , L~ -_'.-..
i l i I Figure 3'.1.1-2 ; l l UNIT 1 CONF. IAADING PATTERN l l-PARTIAL BRIDGE ACROSS CORE ,
.i SR ,
31 i i i R P N M L K J H -G F E D C' B A -i
-4 4 4 4 4 4 4 :
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+ B34 A52 B43 A01
[,, 5 : C31 A32' Bel A23 BOS e l- .; S -; B10 A20 B31 A10 ~ 7 : A15 B07 A03 91 3 - 8 : B11 AEG B53 A35 l 9 : i 10 :
..' ?
11 . 12 : !' 13 . i 14 :. C30 t 15 : t l-1 :
. SR ,
32- i A - REGION 1 C1.0 */o)
.B - REGION 2 C 2. 4 */o) e C - ~ REGION 3. C 3.1 */o) . , 7 c - SOURCE ROD i t,
r .-r-.. , .e -, . . _ . . - _ . p-,-,,m .3 + ,,y- 4 ,-
er. \ h(
- l. ,
1 .
.s r- 1 n a 5
,y 1
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Figure ~3.1.1-3 ;l 1 UNIT-1'. CORE LOADINGiPATTERN - 1l ' COMPLETED BRIDGE ACROSS. CORE u .c
. SR - ,- -
, i ,
y
- 31. ,
-9
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- t l
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. 4 11 , A27 bis 8 A2O BS2 ;
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et 14 - : C31- ..A34 C30 A19 '
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- B - REGION 12 C 2. 4 */o)- D o
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i A ; h li
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Y .r I Figure 3.1.'l-4
~
, t UNIT 1 CORE: LOADING PATTERN'.. g
' PARTIAIi COMPLETION:OF CORE {
1
;i SR- r
]
31 7 1 z i -- ; 1 R P N M L K', J3 H. -G: F E' D C- B .A' 4 c4. -4 p ,4 4 -4 i ,, ,,
+ COO C12 C07 C50 C26 ,, ,,.
2 + A22 C04 A16 C28 " C01.' .
>~a 3 : A40 B45 A42 B19 A33 ,
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,, , j i
5 A32 Bei A23 BOG A14 , 4 l 6 . B1O A26 B31 A1O B63 q 7 - A15 BO7 A03 B13 A63 BO 31- C55 C40
~e .
. 8 :
.l B1'i'- A29 B53 A35 B32 A21 B57 A24 I h .' l 9 . A47 B17 A44 B02 A56 B64 A38 C11' C24 , a
-;l 10 . B23 Ali BOin A55 B26 A04 'BOS A39 C3G 'l
,1 11 : A27 B58 'A20 B52 A59 340 'A65 C64 C ' 49 12 : B30 A12 B00 A07. Bai B27 B59 C18 13 .-+ A57 B44 A45 B25 AS4 BO8 C33 C46 ,
1
' 14 : C31 A34 C30 A19 CS2 C57 C48 .
<. -e e .
l l 15 : CS2 C10 C37 CSO d29 0 J g SR- - o '32 A - REGION 1 C1.6 M o).
- B' - REGION 2 C 2?4 Dol
>l C - REGION 3 C 3.1 bo' .
s - SOURCE ROD- li 4 1
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s Figure 3.1.1 ,; n q
' UNIT /1 CORE LOADING:PATTERNt ,f
.V a PARTIAL COMPLETION OF. CORE , ,
l q
?
1.3R ; , i
;31. ,
?
c- . . R P. N M L K' JJ H; G F- Ei ;D C ,B' Al M 4 4 -- . 4- 4 4 4. 4- , 1 ,, , y C27 C43 C60 C12 CO7 CSO C26 ,, .,, , 2 ,
,* C54 A38 C01 A22 C04 A16 C28 CO2 C63 ,,
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-* B48 A43 B34 A52 B43 A l' B'49 B59 B39 COS' ,,,- .w .;
- h. ?
5 : A51 B20 A32 B61 A23 BOQ A14 B22 A28 C25 C45 J 7 6 , Boa A41 bio A26 B31 A10 B63 A46 B14 A62
- 34 l 7 : C41 CO3 A13 B12 A53 B15 A15 BO7 AO3 B13 A63 BO6 A61 C55 C40 -{
8 , A37 B18 A36 B38 ABO B1'1 A29 B53 A35 B32 A21 B57. A24 3
.. . 1 9 C2O Cei- A58 B50 AS4 B33 A47 B17 A44 B02 A56 B64 A38 Cli. C24 L,
l- '1 0 B35 A18 B23 A11' B01. A55 B26 a04 BOS 'A3G C39 , 11 , A50 B42 A27 B58 A20 B,S2 A59 c*v A6 - C64 C49 '! 12 . BS1 A3i- B30 A12 BSO A07 B21 B27 B55 C18 13 : AOS B41 A57 B44 A45 B25 AS4 BOO C33 C46
~
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to - REGION 1~C1.6 "/o) + . B -- REGION!a LC2.4 "/o)' n-
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to - SOURCE MD 9 sm
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. Figure;3;1.1-6 UNIT 1 CORE IAADING PATTERN ,
t FINAL CONFIGURATION' . t l l SR j 31 ..; 1 L J G. E; R P N M. .K H 1 F1 -D' 'C - B' 'A . . . m t 4 .- 4 4 4: 4-1
. 4- 4 :- ;
l. 1 1 , ,- , + C27 C43 C60 C12 C07 C50 C26 2 ,,
+ C36 C32 C54 A38 C01 A22 C04 A16 C28 CO2 C03 -, ,
.i; c>
3 . CSS C51 B46 A30 B03 A40 B45 A42 B19 A33 B47 C15 C17 i- 1 i os - 4 .,,
+ C47 B28 B24 B48 A43 B34 A52 B43 AOi , 949 BSG B39 CO5 ., ,.D i ;
l 5 : COS C42 A49 B16 A51 B20 A32 Boi A23 BOS A14 B22 A28 C25 C45 6 : C38 AO6 B37 AOS B62 A41; B10 A26 B31 A10 B63 A40 B14 A62 C34 i p -7 . C41- CO3 A13 Bl2 A53 B15 A15 BO7 A03 B13 A63 B06 A61 C55 C40 ' t 8 + C22 A37 B18 A36 B38 ASO B11 'A29 B53 A35 B32 A21 B57 A24 C13 "!
.l ~
9 : C2O C61 - A58 B50 A64 B33 A47 B17 A44 B02 A56 B64 A38 C11 C24 10 COG AO2 B56 A17. B35 A18 B23 Al i- B01 A55 B26 A04 BOS A39 C39
}
' 11 -
C53 COO A25 BO4 A50 B42 A27 B58 A20 BS2 A59 B40 A65 C64 C49 L if 12 - C19 B54 B36 B51 A31 B30 A1'2 BSO AO7 B21' B27 BSS Cie -l A 13 : C35 C59 B2G AOO B41 A57 B44 A45 B25 A54 BO8 C33 C46 , 14 ; C16 C23 C21 AOS C31 A34 C30 A19 C62 C57 C48 ,
.g; '
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J j 1 ( f. L A Figure ~3.1.1 ~
.; j r-
. SHUTDOWN AND CONTROL ROD ' LOCATIONS l
.}
- SR , !
.31-
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-A ,, ..
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. B. ,?
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-8 : C- .E A. . D.-- A> -
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l c. Figure . 3'. l' 1-8. l 'W I BURNABLE PCISON ROD LOCATIONS l
}
.sR < ,
s 31 4 l R- P1 'N M L K J H. G- F E D C- B- ~-- A ; L. 4 4 & 4- 4 4 -4 s_ -F 1 ,, ,,
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+ 24 12 4 .8 : 8 12 24
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+,
I 8 . . 8 8 12 12' b SE -8' 4l -i t 9 . 24 8 8- 12 . 8: :8- 24 ' 10 4 iS- ,8 ,8 , 8' ,
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;i 944 BURNABLE POISON ' RODS . - 12. 5 D o BO- T
-2 a> .f
-3 NUMBER INDICATES 'THE NUMBER OF; SURNABLE POISON RODS
. r' S ~ I NDI CATES A' SEC'ONDARY : SOURCE ROD : -J P INDICATES A ' PRIMARY SOURCE RODi '
l! ;/
' Pw INDICATES A. DEPLETED' PRIMARY SOURCE ' ROD o
]
-37 2
, k
- v. ,
o j l 3
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,i 3.1.2 - CORE LOADING INSTRUMENTATION AND NEUTRON SOURCE.
CHECKS - ISU-003A OBJECTIVE I The core > loading instrumentation-test ~is: performed prior to. core: i loading to determine the proper operating and' discriminator. voltage ,: settings for the temporary core loading . instrumentation and - to; ; verify that ._both - the : temporary 'and permanent nuclear monitoring . instrument channels respond properly to a neutron source. -: The test l is also performed to verify that both tho' temporary and permanent' nuclear monitoring instrument channels respond properly to neutrons prior to resuming core' loading;following.'any eight' hour or longer' delay in loading.' This_ test satisfies _ testing 1 described by!FSAR'_ Table 14.2-3, Sheet 8'and Section-14.2.10.l. TEST METHODOLOGY , ; Following the initial installation of' the: equipment, the _ temporary o detectors,are_ positioned near a neutron source. 'Using the neutron _ 'j source,-an optimum: operating voltagetis selected for'each of.the. threa detectors-to ensure that minor fluctuationsfin detector power; rapply voltages would not adverselv ' affect detector: output. With, the individual: detector- operating Soltages selected, discriminator bias voltages are determined based ~ ont detectori characteristic , curves. Prior to core loading, ~all fiveK channels > j(two giristalled . source range and . three' temporary core " loading - channels): are neutron: responseg checked by moving L a portable
- neutron : source 1 toward !and away from each detector to verify detector response.; j)
In the event of a delay in core. loading of 8) hoursf or' greater,1 this t j test averifies proper detector neutron: response ~byJone'ofLthree . meth ac., One-method uses a portablo. neutron;sourcc moved =toward
~
and then away from.a detector to:verifyidetector response'.- .The second' method is toLuse-movement ofLan:. installed fuel assembly'to i alter neutron: flux at. a detector by , alteringyource to l detector neutronic coupling. The; third method;' uses an evaluation of counting statistics applied -to detector ' output .when.in ~ proximity to' l a fixed neutron source.- Nuclear decay'isga-random process and if .l the detector' output. exhibits statistical: ' behavior .(standard deviation, etc.); characteristic. of ai randomi process, t then' the.
. detector is judged to be responding to neutrons instead.of' 60 Hz ;or- 1 other noise.
j l o se 1
- 1 3.1.2 - CORE LOADING-INSTRUMENTATION AND NEUTRON' SOURCE CHECKS - ISU-003Al(Continued)
SUMMARY
OF RESULTS -i
~
Upon completion of' the procedure, operating voltages' were determined to be 2000 volts ' for. all' three temporary nuclear- .' instrumentation channels with d.scriminatorJbias: voltages set at 3.5 volts for all three channels. Seven4 additional detector _ tubes- l l were also tested for use as spares,.as necessary., All sevenl spare tubes also had operating voltages. of 2000' volts and.!3.5 volt - discriminator bias voltages.- .Only one spare tube was~actually ., used. One detector' tube. exhibited erratic behavior. prior to:the start of core loading and-was' replaced with a spare. , 4 Prior to core loading,.all five channels.(2 installed source ranges ! l and 3 temporary core - loading l channels)1 were --neutron response; l checked =using a. portable _ neutron source.: .The channel count ratesJ i were observed.to increase by factors of between;30-and.50,000-when. L the source'was placed nearby.the-various detectors.: i The 8 hour. delay portion cf the.-test: procedure'was executed three ; times during the core loading activity. .During -the .first-l performance, it was: observed that as the neutron source. approached l each. installed source. range L detector, ' the channel's < count rate increased accordingly, indicating that the'. detector was: responding _ l to neutrons. Also; duringL this same- ' performance,; the three-temporary channels were verified ~ using;the statistical: method. The-final two performances . used . only the ' statisticalH-method 1 for all-channels. No fuel assemblies were moved,to verify' detector n'eutron responses. ! 1 t L 1 '. f iE i-F 1 1 [x l! ?
, i.l!'p .4- ;, + __ . -. . ._._ . .o _ . . . _ , . , , m . ., - . 4 -.m;..,.-
1 3.1'.3 - INVERSE COUNT' RATE RATIO MONITORING.(Core' Load PortionF i NUC-111: . OBJECTIVE This permanent plant procedure is performed:to obtain and evaluate' I. nuclear. monitoring data during core ^ loading to' ensure that' core. jl loading' is ' done in a cautious .and' controlled' manner. This procedure satisfies activities described in FSAR'Section 14.2.10.1.: TEST METHODOLOGY I
'l Neutron count. rate data from both installed ~ - source rangelNISc ,
channels'and three temporary core load ' instrument channels'is takeni i following each ' fuel. assembly additicn. The sources"of:the'corea neutron flux ' are the four - installed Californium primary, neutron . sources with associated subcritical multiplication 1 due to Ethe -l loaded ? fuel' lattice. 'As fuel is_ loaded, thei. core Eneutron.. flux l changes due-to changes in fuel lattice ~ geometry and the addition:of uranium to the core. To determine the effect of , igle y a w embly addition on-core-reactivity, count rate dats ,er..w on fuelfassembly;is loaded;isj j compared to a reference vaAue to. ' evaluate the , effecti of the - 1 additional fuel assembly. Thisfcomparison is performed as a' ratio of the count rates to evaluate the ' 2ractional~ 3 change. If=this. ! ratio were' to be veryLlarge, it would = indicate that this . fuel-assembly addition brought the loaded fuel? lattice significantly-closer to criticality. . For,conveniencei.the' procedure evaluates the inverse of the count 1 rate ratios-(ICRR)lsuch that'an~ approach : to zero would indicate an approach:to critic'ality._LAdditionally, this procedure trends the inverse count rate ratios- and-extrapolates the trends to evaluate whether;or not additionalEfuel. , assembly loadings.would be expected to result.in_an.-approach to , criticality. . Prior to the start cf core loading, background counts are taken to allow the elimination of general background-.' radiation from the-calculations of.ICRR values.- Reference count rate datavis:taken-initially after the first ten fuel ~ assemblies.are loadedi Eight .of . these fuel assemblies are loaded together!to: constitute:ai" source' nucleus", providing a '.subcritical, multiplied iflux capable ofl being , used ai s 1 asis for meaningful comparisons. Reference, values.are 1 L redeterm md if a neutron source Lbearing fuel assembly is moved.or.
- if ' a temporary detector: is' moved, both'which cause L.a L change ' in
~
source to fuel to detector geometry. .New reference values are also .! obtained if neutron counting channel equipment. or; el~ectronics ~ " settings are changed,.to ensure that'a. valid. reference valuelfor e
. count rate . comparison is used. As - a conservative measure, new ]
reference count rate values, are . determined if core loading .is : ! delayed by 8 hours or more to ensure that any count rate changes- 1 i 1
- -_'. c
- j. i 1
i 3.1.3 - INVERSE COUNT RATE RATIO MONITORING.(Core Load Portion) NUC-111 (Continued) TEST METHODOIDGY (Continued) over time are accounted for. . Inverse count rate ratio. data taking, calculations, plotting, trend evaluation and extrapolation are~also: ) repeated hourly _ during any core. loading delay' for general- ; core monitoring. and- to aid in detection of- any: inadvertent RCS dilutions. F At all times a minimum of two selected chsnnels of instrumentation' are designated as " responding channels".. This designation is based-on source to fuel - to detectorc geometry considerations'soE as to avoid large local effects that-may not.be indicaH ve.of totalicore: behavior.- , Final reference count rate data.is;taken following;the completion:
- l. of core loading: for use as; baseline @ta to ; help verify source s o '
E range NIS signal to noise ratio.' ' s
SUMMARY
.OF RESULTS All count' rate -data- -was properlyf recorded ; 'and ICRRs were " calculated, plotted, trended;and extrapolated. -Refer to' Figures . 3.1^. 3-1 and 3.1.3-2 for. ' a : graphic Ldisplay of procedure = results ' q' during core. loading.. The inverse count rate. ratio shows that core-loading was performed in-a cautious'and; controlled manner with no-indicated unexpected approaches toward criticality. "Atino time,did -
~
j ' the extrapolated data from _a trasponding channelu indicate l that" criticality would be expected to occur with the loading of the next. , fuel assembly. Large changes in' the inverse count rate ratios from ? one core loading step ~to the next step:are-due!primarilyzto! local- 3 geometric effects when a. neutron source was: moved;ne'ar a detector - or when fuel was loaded,between a source-andta' detector!resulting V 1 in a large local cour.trate increas~e ' due y to o enhanced' neutronic e
- source to detector coupling via suberitical' multiplication. : TheseL were local ef fects observed ;on. 'onlyi one orJ two; charmels 'at-L a time- . O and is the reason five monitoring channels are;used.. -
l Monitoring . . data was properly :taken 'and 4evaluat'ed = duringi core ; L - loading delays. and reference ; ; count rates 1werel properly' recalculated. The background count rates'were sufficiently. low so a as to be nearly. negligible, alsoLan indicationi of l lown neutrcn , L detector both source channel: noise.' Final: reference! count data was taken for- . rangeLchannels. 1 1 a i t a i e t
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--e 4:
- ICRR vs.. Fuel Assemblies Loaded
- CPSES Unit 1,: Cycle 1 Source Range Channels
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lCRR~
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Numberiof Fuel: Assemblies Loaded' -- - a, Y t _ [ ) -y ;. N31 i 7 N32;
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# ilCRR vs. Fuel Assemblies Loaded.
GPSES Unit"1, Cycle 1 Fuel Load Temporary Channels , ICRR
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110 .130 '150- 170 ~ 11, 0 . Number of Fuel -Assemblies ~ L.oaded :.- j
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- s. 3.1.4 - RCS AND SECONDARY COOLANT' CHEMISTRY (Core Load Portion) -
1 ISU-006A OBJECTIVE This - test 3s performed .to :. verify: correct- and' unifonn - boron concentraticas in portions of the reactor coolant systemi(RCS) and the directly connected portions of fluid systemsEas required:for core loading - This test'is also designed-to help: ensure that the .) possibility of an ninadvertent dilution -- .of the RCS during , core loading is minimized. This test satisfinsiactivities M scribed'in i; FSAR Section 14.2.10.1. l TEST METHODOLOGY I n rior to the commencement of core loading, the'.RCS'is' sampled-and i verified to. meet specified -water chemistry? criteria.L LAs' ac prerequisite to RCS' chemistry sampling, the borated water source,: > the RCS_ loops,-Chemical and Volume l Control7 System piping, Safety Injection system piping, and_ Containment.SprayJSystem piping were o verified to'.have Laron. concentrations which would preclude'. inadvertent RCS~ dilutions. 1 Each of the RCS ~ crossover legs, o the Residual He'at Removal- (RHR) ' system,_the reactor vessel,.the Volume control Tank, the safety f Injection System ' accumulators, the boric . acid tanks, and' the Refueling Water Storage Tank: are ' sampled, and that. water is verified to contain specified-boron concentrations'. l Following the initial verification :of the chemistry iri the reactor 'i coolant system, four-samples are taken from the-reactor vessel at :) (. equidistant depths along with a sample from the operating residual heat removal train. These samples:are-thenianalyzed;for. boron-to: - i verify a uniform boron' concentration between;the RCS0and the RHR system-(within.a 30 ppm range). After;the?RCS and1RHR'is verified 1 to be at a uniform concentration, thel operating . residual heat removal train is : sampled. and ' analyzed for; boron?to verify?that the - water remains at 2 2000 ppm boron. -- Sampling l continues i every _12 1 hours until theEstartLof core' loading. With? the; start of 1 core -i loading. sampling continues on the'operatingsRHR? train every four 1 hours throughout ths core load 4.ng process'.f The~12 hour and 4 hour samples also include measurement cf RHRii~nlat; temperature forjuse. j
-in monitoring reactor ! coolant system ' temperature : for: compliance 'l p with Technical Specifications and to. ensure temperature changes do E
l
- .not adversely influence. inverse = count rate ratio. monitoring. ..The ;
p . spent . fuel pool: ~ is dry? during initial core-~ load and1the fuel l transfer. system canal: portions are not required to be borated. 3 ! l o 4 j w
.u
3.1.4 -'RCS AND SECONDARY COOLANT' CHEMISTRY-(Core Load Portion)1-ISU-006A (Continued)= IEST METHODOLOGY - (Continued) The criterion for the 4 hour-samples is'to ensure a minimum of 2000 ppm for shutdown margin and a maximum of:2150 ppm to not overly. attenuate the neutron detector signals during core:loadinty. Also, consecutive samples;are not to differ by.more than,20 ppa as a.way;. of detecting any inadvertent dilution.
- 9
SUMMARY
OF RESULTS , l
'l l During the execution of this test'which started before and' lasted-l throughout the core ~ loading process, all. acceptance criteria were :
met for each system that was sampled.- lNo corrective actions in tho' ) I core loading process were needed to meet theLacceptanceicriteria of ; this test.. Detailed results obtainedM prior : to. core load are. tabulated below:' . l Specified' Actua1L Location Ranae (com) !Value (nom) Volume control Tank 2000 - 2150: 2040, RHR Train A 2000 - 2150 -2036. RHR Trap) B 2000.-12150 '2020- ! Refueli t Water Storage Tank 2000 '-2200- .2028 Boric Ac.d Tank #1 7376
~
2 7000 Boric Acid Tank #2 '2 7000: , 7010 i Safety Injection Accumulator 1 1900 - 2100 2053. . Safety injection Accumulator 2' 1900 .2100 2064J Safety Injection Accumulator 3 1900u-L2100 c2051-Safety Injection Accumulator-4 1900? 22100- 12057 , RCS Loop 1 Crossover-Leg > ~2 0 0 0 :' 2038 ;j RCS. Loop 1 Crossover Leg > 2000. 2038' { RCS Loop 3 Orossover. Leg > 2000 2033 .I RCS Loop 4 CrosrTver Leg. >-2000. 2039 ! Reactor Vessel Surface Withinta '2041 Reactor Vessel 1/3Edown 30-ppm 2041 y Reactor Vessel 2/3 down . range- 2045L Reactor-Vessel Bottom 2045: RHR Train A: . '2048 iu RCS/RHR uniformity values.were within a 7 ppm range,.wellLwithin .
.'l the 30: ppm limit.
3 The RHR: samples prior to and.during: core: loading varied within.a range of>2037 ppm '.to 2057 ~ ppm, . . well =within . the 2000~-~2150Lppm range limits. No'two consecutive:samplesideviated'by xore than 10 ppm. This satisfied ' the <20 ppm difference 111mit. All' samples ; were f rom RHR Train - A. -. .RHR: inle+- temperatureD was . ,very steady, ! increasing: from 110 F l before .- core loadingF toi 115'F . at- the - endL of' core' loading,imore than:100 hours later. , d l 1
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3.1.5'- VERIFICATION OF CORE ~ LOADING PATTERN - RFO-204 OBJECTIVE This permanent'plantoprocedureLis performed;tolconfira thaththe L loaded core matches the design' loading pattern'and,to provide a I :- videotape record of the as-loaded core. TEST METHODOIDGY l q Using the' manipulator crane (refueling machine). television. camera. mast, an underwater TV camera-is slowly travorsed'over the entire' loaded core allowing fuel assembly.;andlinr.ortinumbers,cpositions= , j and orientations.to be observed.on a TV monitor.; This information 1 is recorded and compared against the core,: loading patternEdesign information from the fuel vendor. ;The TV signal"is also sent-to a- l video recorder so a tape record of~the as-loaded' core / pattern 11s I made. The--use of the camera within the reactor < vessel constitutes: l a core alteration, so all required Mode 6 core l alteration'related; - Technical specifications are'also: verified?by;this proceduren.tof . have been satisfied.. ;) i
SUMMARY
OF RESULTS The entire core was mapped using the underwater'TV camera. All , fuel assemblies and inserts were, found? to' be Jin their i proper ' i locations and orientations. .A videotape trecordE was made .and reviewed to ensure that it was legible.
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i o 1 3.2 SYSTEM TESTING: AFTER CORE IDAD AND AT VARIOUS POWER ' LEVELS j
.3.2.1 - PIPING VIBRATION MONITORING - ISU-212A' 1
OBJECTIVE This test demonstrates that steady a state flow induced- piping vibrations and transient response piping vibrations ' are within allowable design limits. The scope of the test is limited : to - portions of the Main Steam and Feedwater= systems ~ for transient response and the Main Steam, Feedwater and Condensate systems for' j steady state. .These are-systems which could not be1 fully, tested- - l- during ; the Preoperational . Test Program due- to- plant' conditions.- + This test partially. satisfies the? testing described by FSAR Tablei 14.2-2, Sheet 57 and Sections 3.9B.2.1.2 through-3.9B.2.1.4.= q TEST METHODOLOGY The Main Steam, Feedwater and Condensate systems are operated ~ under normal, steady state. conditions during which visual' inspections of, - C, the' piping'are conducted. The walkdowns divideJthe systems'into smaller piping subsystems between restraints.~in' order to use'thes simple beam analogy to determine deflection: limits. Portable' vibration analyzers-are also used to obtain numerical values E for i selectad vibration levels and comparisonsL areinade o between the vibration velociMes . or _ displacements and the -appropriate limits. . , Selected locations were also instrumented > for remote vibration-monitoring for safety, laccessibility and AIARA ; considerations. Based on' the outcome, vibration levels less than' the1: allowable- >
' The ' steady L state -
~
limit would satisfy the 1 Acceptance 1 Criteria. > testing of various subsystems iso performed between 3-6% power,:at
- 35% power and'at 100% power.
]
The transient responsa portion of the testialso combines dataitaken L from selected remotely instrumented portions lof Whe. Main Steam and j Feedwater systems: during theh imposed transient - with o concurrent ' , visual observations of accessible-. piping system portions.~ Portions. j of the Main Steam system are' tested in response-to'a' full power. main turbine ~. trip.with portions of the Feedwater system tested'in-response to Main Feedwater Pump trip.
SUMMARY
OF RESULTS 4 The steady state portion- of the -test generated-: only three L calculations ' where o the : levels of vibration. exceeded - the limits ? specified in theLtest; These three items were allfon Feedwater. system recirculation: piping to the condenser. All three items were- 1 evaluated'andLdispositioned by? Engineering:as' acceptable based on ,I the - small l magnitudes Lof 'the ; actual? displacements and':because the: lines.are:not in: continuous =iservice.. These. m niflow' lines have j flow.through them.onlyfintermittently. , j
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r i-l N y 3.2.1 - PIPING-VIBRATION MONITORING ' ISO-212A7(Continued).
. SUMMARL OF RESULTS (Continued))
All' owed - Measured j Pipe Line. , Val'ocity- 4 Velocity ' Location Number Iinches/sec)1 (inches /sec) ; 1A Miniflow 12FW-1-21-2002G- '50.5 2.3-1A Miniflow 12FW-1-26-2002G . 50.5: 0.95 - 1B Miniflow 2FW-1-24-2002G 50'.5 18 > ' l (;7 9 rain.Line) < L n Actual-i Allowed' Actual Displacement: Displacement Displacement i d Location ER,ti2 Ratio 1. j inches) ! q 1A~Miniflow $1.0- .2.48. 0.10' 1A Miniflow. $1.0 ~1.46: 0.04' V' y :f 1B Miniflow $1.0 - 1.'13 ' ;7.28' j (2" Drain Line) For the transient . response : portIonsc ofithe test,i there J werel no' , discrepancies: noted ' with ' regardi toithe Main Feedwater ~ Pump tript transient. There were two items noted:in connection with.the1 Full , Power Turbine Trip transient. One (instrume'nted ; anubber,' MS-1-002- 4 009-C72K (location =TR-1-MS-25),t exceeded 1 its vallowed loading j' criterion. Engineering 'evaluatedi;andL dispositioned' thisi as acceptable ' because' while1 the expected oloading e ofu 9253 lbs.. was: I exceeded by 16% .(107201lbs) , . theEsupporti had lavailable c designi margin. Even o though E the : actual 1 transient loading : exceeded! the ;; expected loading byt16%, . when the 4 transientf loading Tis ; combined' 1 with predicted-seismic loading:the total? load': change-is only an. increase of 317.-lbs, 0.'6%. ^Thisiishwell[within the~15% of-total 4 load snubber design margin 1 available. An separate Lcalculation - . indicated that? the Dpiping f stressesi;in' this1 area u were ' $16750 psi which lies welli within' -the? $21000 psi' allowable range.. Additionally,. two remote sensors, fat; locations TR-1-MS-02 and TR-1-i, a MS-03, failed; to function 1 during' the r transient. . ~ Engineering l evaluated and dispositioned this missing, data as acceptable based 1 on the data obtained from 17 other, functioning: main' steam ' system sensors. j Other than the above - noted ' items, the remaining L pipi ng' system portions all had vibration. levels'within the acceptable range.' I t 3 t 3, ., e i __..l_.m_ _ _ _____._ _ _ _ _ _ _ _ _ _ . _ . _ _ _ _2.
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e t ! .- i 3.2.2 - STEAM GENERATOR LEVEL CONTROL TEST - ISU-2Q2& : OBJECTIVE , .i This test is performed to demonstrate steam generator level control .. stability throughout power ascension. . Changing ~ ' feedwater - flow - 1 configurations and major power changes necessitate Etho'need for - l multiple performances of this test. Level control stability.of the four steam generators is demonstrated- while operating - on; the - .{ feedwater bypass control valves' and - the main; feedwater. control . , valves. 4 TEST METHODOLOGY ' i In order.to verify. level control stability while-operating on theJ bypass or main fsedwater.controltvalves,'a!5% level:deviationhis: manually established-in each. steam generator. The control system ; is 'then ' transferred to the " automatic control.. position. -
' Steam. ,
are tested- sequentially, .one at a stime,- not generators simultaneously. ,The: actual steam' generator level is monitored 1to; determine overshoot, .undershoot and whether or not level returns'to. and remains within: the allowed band; of 66.5% 12% of L narrow range level within a specified time- frame. of ' 3 -times 1 thel appropriate reset time constant.. The' bypass valves are tested at?approximately( , 5% power. The main feedwater control valves! are . tested at approximately 50%. power. In order tc, verify. level . control stability 'while transferring between the feedwater bypass control valvasiand' the' main feedwater . control valves,- steam . generator levels -are; monitored.:while S performing this transfer at approximately120% power.- 4 q At approximately 50%, 75% and.100%/ power Wdata-is taken to verify. 4 expected mainLfeedwater control valve positions, to verify proper a feedwater pump speed control operation on,its' sliding ~.p program ?[ and to verify non-excessive' feedwater header pressure-oscillations.- ! At. 75% " and 100% ' power, data is ' also' taken' to . verify proper: steady)
~
a state level control operation..
SUMMARY
OF RESULTS .I l
-1 Refer to Table 3.2.2-1 for detailed test'results.
.. a When given~a 5% level' deviation (high or. low)-', the-bypass: control-H i valves returned steam generator level to and remained)withincthel %
l programmed Elevel, 12%, within 36 minutes (30 '. minutes for steam? generator #2) with less - than 4% overshoot or undershoot, . as. This was-done"at.approximately 5% power. expected.
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i I 3.2.2 - STEAM GEg .tA6dR LEVEL CONTROL TEST - ISU-207A (Continued) , q
SUMMARY
OF RESULTS (Continued)- When given a 5% level. deviation (high or low), the main feedwater control valves returned th:~steen generatorJlevel to and remained-within the programmed level .t24 within 83.5 minutes with less than' j 4% overshoots or . undershoots as l expected. This was done J at ~' approximately 48% power. j After transferring from the feedwater bypass control val'ves to the main control valves, steam generator: -level 1 deviations were : to - return to and remain within 12.0% of the programmed level within110 minutes. This.was not satisfied-initially. . A misinstalled jumper on'a circuit board for'the Steam Generator.#2-level controls,was? corrected and all steam generators were ratested: satisfactorily at. Ll approximately 20% power. > 1
- The feedwater header pressure. oscillations: were/less than- 3%i of .
operating pressure range at:approximately 50%,e75% and 100% power.. l At approximately 75% and 100% power,.all? steam. generator steady j state levels were' verified'to remain within the expected 66.5% 124-operating band. l At approximately 50%, 75% ; and 100% power l the sliding A p program i value, used'to control main.feedwater. pump speed, was.. verified to. be within i25psig of the actual Ap lvalue ,
- At approximately 50% power, all feedwater-~ control . valves were i verified to'be within- 10% of their predicted positions.: However, at approximately -75% and 100% power, only '3 'of the 4 : ; valves -
1 satisfied theLi10%. band' limit. Feedwater! Control.Valvell-FCV-510 " indicated ' more than 10% below the predicted E position. . Valve > operator clearances were changed'and the! valve waarverified to be open the proper amount. L Two other plant. problems were noted and corrected as'asresult of performance of this test during powerLascension.: .The: sliding d p ' program value used . for controlling main' feedwater pump ' speed was.
)
adjun,ted during ! power ascension 'to " reduce). the : A p across. the feedwater control, valves'to optimize performance of thelfeedwater' system. Also, high frequency valvei motion . of c Feedwater. Control! 'i Valve 1-FCV-540,. approximately - 1/4 - ' inch in displacement, ..was - 3
, identified. -An instrumentatione: scaling change minimized - the
' ~ oscillations and further design modifications are to be implemented 1 during a subsequent outage' 1 u
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i . TABLE 3'.2'2 . , STEAM GENERATOR LEVEL CONTROL
SUMMARY
L L : BYPASS CONTROL VALVE LEVEL. f~' CONTROLLRESPONSE PERFORMED AT APPROXIMATELY 5%LPOWER
-MAX- LMAX-ACTUAL OVERSHOOT / IMUM;, IMUM LEVEL ACCEPTANCE TIME. UNDERSHOOT'OVER - =UNDER -
STEAM- . DEVI-': CRITERION; ' RESPONSE- . LIMIT-IN -SHOOT' . SHOOT' GENERATOR ATION = IN MINUTES IN MINUTES PERCENT' PERCENT ' PERCENT: I
, ~l
-d4.0 1 5%-.up- 136; 423.3- 1. 5 - .0 '
5% down.. .s36 .32.61 :<4.0? ' O. . 2 . 0" -l
- 2' .5% up _s30- '26.8- . <4.0L -2.0 ,' OL 5%-down $30 '26.6 J< 4 . 0 ' to; 2.0 lj
- m
-3 5% up .536 .
17.0- l<4. 0 L 12.0; , _ '. 0 o
-5% down' $3 6,
~
' 2 6 '. 5 . -<4.0- - 1. 0 l'. 0 '
-4 5% up .$36- 15.5 . <4.0= 1.5 0' q 5% down $36- ~22,5.
, < 4 '. 0 -- O. 12 . 0. . j LEVEL CONTROL RESPONSEJAFTERJBYPASS To-MAIN '. a
-FEEDWATERtCONTROL VALVE TRANSFER AT- ;
- APPROXIMATELY': 20%: POWER ' .
J
-ACCEPTANCE ' ACTUAL TIME,
- STEAM -CRITERION:IN sRESPONSE/IN; L!
GENERATOR MINUTES % MINUTES y a 1 i<10l 0-,
.2 ' <10: '4.4- ,
c
.3 :<10: '0:
4 <10- -
- a
..._______.______________..____________'______;__ _'O'
_i_ _______.... i FW HEADER'PRESSURElOSCILLATIONS1 I Power Plateau - Maximum Presstire Oscillation (nsic/4) L Allowed Limit- <45/3.0; % 50%/ Pump <A. , 3/.2L i : 50%/ Pump B (~ ~13/.9 75% 12/J80 j? c100% 13/.9' ' 1 i 1 3
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' TABLE 3.2.2-1'(CONTINUED)'c '
'I L-l FEEDWATER dP PROGRAM COMPARISON- ] '
Power Plateau, Maximum: O P Deviation'fnai)' s Allowed Limiti
'<25; 4 'I
- 50%/ Pump AL 15.0/ '"...
ii 50%/ Pump-B 22.9 .j L 75%~ ,
'5.9 l 100% 0.8. J l,
I U r; MAIN lFEEDWATER CONTROL VALVE: m._ ,
-LEVEL CONTROL RESPONSE AT d(
!APPROXIMATELY"48%, POWER ~ .
4
?)
.1
.,a MAX-?': MAX-A ,'<
ACTUAL) -: OVERSHOOT / IMUM: IMUM Oj LEVEL' < ACCEPTANCE TIME. - !UNDERSHOOT OVER-; ;UNDER- .
.-STEAM 'DEVI-f CRITERION iRESPONSE- LIMIT IN - SHOOTL SHOOTJ t GENERATOR ATION IN MINUTES IN MINUTES PERCEa T ' PERCENT PERCENT s
. . *v
~
5% up: ;0 1 -$83.5 ~15s6. : <4 . 01'
- 1 01 5% down' 583.5 33.8) "< 4 . 0; -0: E. 2 . 5 at a
l 2 5% up '
$83.5. ' 14'.9 <4.0L .
40.5 _ 01 , n' i 5% down, $83~.5 .
, 12.8.
;<4.0/ 0~ 0.5 l - ,
L 3 5%up. 183.5- 118' 1' . < 4 .' 0 ; -1( O! 9 g 5% down 583.5! q 17 . 3, <4.01 0: 1.5- O 4 5%fup., 5% down l583.5. 4.7-2;3' s<4.0'
<4'.0'
- 0 '.' 5 :
"'OL O' }
583'.5 >
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FEEDWATER CONTROL VALVE-POSITIONS'AT "- j VARIOUS POWER: LEVELS ' ACTUAL VALVE POSITIONS IN~4-1 Predicted' . Predicted Power Level 1-FCV-510 /-Ranae in 4" 'l-FCV-520 / RanaeJin % i 48/ Pump A 50 39-59 56.3 39-59: 48/ Pump B 50. 57'_ ,,56.3 . 39-59" ' 73- 51 54-74 6 6 .' 0 . 154-74 -;
.100 66.7- 68-88 87.5- 170-90 , g.
.Pred'icted' . . Predicted .
Power Level 1-FCV-530 / Ranae in 4' 1-FCV-540 /~Ranae'in'4'.
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48/ Pump A ,50 40-60 i '54.2 ,39-59; 4 48 C >38-58-. 39-59 54.2 - 73 -66.0 54 -66.0 54-74. 4 't 100 87.S' 68-88' '83.31 70-90.- 1 Following rework of 1-FCV-510: , . Power Level Actual Position . Predicted 'Rance' (4)- , a L 75% 68.75' 53-73 100% 75 :68-885 l 1 ' l i l -- t [
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9 3.2.3 - THERMAL EXPANSION --POWER ASCENSION PHASE -ISU-308A OBJECTIVE Thermal' expansion testing of plant systems is conducted to verify that components and piping . can - expand- ~ without restriction < of movement upon system heatup. It is: also conducted to confirm the - correct functioning of component supports,; piping ' supports e and ;! restraints.- This test covered portions of= the plant that could not 1 . be teste? during the Preoperational E Test Program- due to plant- ' conditions.- This test satisfies activities described in FSAR Table: 14.2-2, Sheets 52 and - 52a and in- FSAR Sections 3.9B.2.1.1: and 3.9B.2.1.4. TEST METHODOLOGY ; At ambient and hot' conditions, system L walkdowns l are, performed. - Both the NSSS and' selected secondary:: plant systems'are evaluated. , Piping and-components are_ visually examined andLspecific;anubber-1 positions recorded. Pipe Whip restraints are ' verified not to interfere with the piping and variable 1(spring) .hangerJmovements are recorded. Interferences a re ! identified and dispositioned by the - design engineers.. When _ necessary, system' walkdowns: are -again conducted following?theire wlutioniof interferences.' All-piping movements are evaluated.L b/- the design:1 engineers. : Selected a locations are remotely instrtimented to: measure piping movements:for AIARA, safety and accessibility reasons. -The:walkdowns and; remote i data collection are performed Lt NSSS temperatures of'approximately 7 0'F , 350*F and 557'F and 'at 'approximatelyJ 30%,- 50%, f 75% and' 100% power. < '
SUMMARY
OF RESULTS The piping and components were not'to be constrained from expanding and actual thermal expansion movementsi:could. not; vary, from ' i predicted thermal movements by morer than 25%i or - 1/4 ~ inch, whichever was greater, or reconciled by Engineering. Also,espring hanger movements were to . remain : within : their Lworking rangeX and - snubber; were not to become fully extended or retracted. During the course cf system..walkdowns,- several _ minor interferences were' s observed andL determined tof be l acceptable-as-is, or' specific corrective actions were recommendedU :All recommended corrective' 4 actions were initiated. Some portionstof the piping systemsfwere again examined and ^ measured following the removal' of' interferences. . Movement of components not within'the;125%:or 1/4: inch criterion 7 were evaluated ;byf the designz engineers :on a case-by-caseibasis. All' thermal expansion movementsiwere determinedito-be acceptable i for continued plant , operation. Remote ' movement odata iwas also-collected during plant = transient testing.-
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3.2.3 - THERMAL EXPANSION - POWER ASCENSION-PHASE --ISU-308A-- (Continued) i
SUMMARY
OF RESULTS (Continued) During the NSSS temperature increase between 70'F and 557'F three pipe whip restraints were evaluated and removed and three others-were adjusted or modified to allow free ' thermal movement of' thc pipe. These preliminary results were also evaluated by EngineeringL and - used as the basis for revision of selected predicted -- pipe - movements prior.to power ascension. In Mode 2, eight Extraction Steam-(EX) system drip pot drain lines. ) were discovered to have crushed: insulation. Thaninsulation:was removed and the system was'refloated. In Mode 1, one:EX drip pot ! drain line was found in contact with the floor.' A sma11' amount of. floor concrete- was chipped out to' provide clearance for this > 1ine. - At 30% power, the following types of items were noted and evaluated' by-Engineering:; o Thermal expansion movements in excess ofl i l/4 inch or: 25% o contact between pipe insulation 1 and ' .feedwater pipe whip restraints , o Heater Drain system piping in contact with building ' structural # steel o one bent-strut on a Steam Generator Blowdown system linel j At 50% power, the following types' of ' items were not'ed and' evaluated 1 by Engineering: 1 o Thermal expansion movements in: excess of il/4 Einch or 125% 1 o Contact between . pipe insulation _ and .feedwater --- pipe - whip j restraints . o Heater Drain system piping in contact with building ~ structural 1 steel ' o Snubber angularity discrepancies _ . _ o Two EX drip pot drain _ lines.in' contact with thel floor-o Higher than expected temperature detected on a'feedwater line - ' upstream of-a check valve. At 75% power, the -following- types of' items were 'noted and evaluated ; by_ Engineering:- 1 o Thermal expansion movements in: excess-.of 11/4' inch or t254 1' o contact between pipe ! insulation Land. feedwater pipe .w hip-restraints o EX2 drip pot drain lines 11n' contact with.the floor' i o One bent strut on the Heater-Drain system - o Same bent. strut on'the Steam GeneraterLBlowdownisystem _ l 1 i
3.2.3 - THERMAL EXPANSION - POWER ASCENSION PHASE -'ISU-308A-(Continued) SIDMARY OF .RESULTS (Continued) At 100% power, the following' types of items were noted and evaluated by Engineering o Thermal expansion movements in excess of 11/4 inch or 125% o contact between pipe insulation' and feedwater . pipe - whip . restraints o Same bent strut on the Heater Drain system o Same bent strut on the steam Generator Blowdown system o Heater Drain system piping contacts o Heater Drain system support base plates pulled away from columns o Main Feedwater system piping in contact with support steel All of the above-items were evaluated as acceptable by Engineering or have had corrective actions . initiated via design modifications or Work Orders.
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i i 3.2.4 - INCORE MOVEABLE DETECTOR SYSTEM ATTGNMENT ---ISU-016A. l
.1 OBJECTIVE l The purpose of this : procedure is' to demonstrate the proper' 1 operation of the flux mapping' system, including the leak detection- l system. In addition, top and bottom of core limits are set and the; ,
actual drive cables and detectors are installed- and verified to -l function properly. This-test satisfies activities described by FSAR Table 14.2-3, Sheets 34 and 35.- l TEST METHODOLOGY Using a dummy drive-cable, the top and bottom of core limits are- ',i~ established for normal, emergency, t4alibrate and common. modes.and for the storage mode endpoint and inmt limits by- slowly driving; the dummy detector to the top of the r., ore (or storage - position) where clutch slippage is observed. Tre position was then recorded-from the encoder display. The top limit is obtained by subtracting - two inches from the recorded position and the bottom limit is , obtained by subtracting 170. inches from the -top limit; Storage a mode insert limit is the endpoint minus 36 inches. Drive speed is ! measured by timing cable motion over a given distance to verify the i design speed of 14412 inches / minute. . The leak detection system is ; tested by filling the: drain header - with domineralized water. and ; allowing the leak detection--level switch to actuate,- thereby- j draining the water and alarming. purge system is verified ! to function properly to inject CO 2 The CO, te of less than 10 ft3 /hr at a-ra i following detector withdrawal. The - withdraw and safety . limit switches are verified to prevent the' detector from being taken'up i onto the real. All push-to-test lights are-verified. . Simulated signal transmissions to the process computer and from the incore : system are made to verify proper computer data logging from the-incore system, .j i
SUMMARY
OF RESUT/M Figuro 3.2.4-1 displays the Moveable Incore- Detector Path Locations. Proper operation of all' indicating lights were verified along with the proper operation of the leak-detection system and alarm as described in the previous'- section. One position indicating lamp failed to illuminate initially. .The' wire to'the lamp was repaired and lamp operation successfully ratested. The 1 CO purge operated properly at an 8 ft3 /hr flowrate. The dummy dekectorwassuccensfullyinsertedintoall58corelocationswith
~
proper drive : speeds -verified. All top and bottom : limits were properly established. The limit switches < were demonstrated ; operable. The simulated data transmissions verified the ability of the process computer to receive signals' from the incore flux mapping system and,the ability of ? the incore system - to supply 3 proper signals to the computer. As_a final step, the actual 1
3.2.4 - INCORE MOVEABLE DETECTOR SYSTEM AT.TGNMENT - ISU-016A! (Continued):
SUMMARY
OF RESULTS (Continued) detector cables were. installed on the drive units and a-demonstration full core flux map was taken, even'though no usable-neutron flux had yet existed in the core.- The detector cable could i not access core location B-13, even .though the dummy cable had i successfully been driven through this core 1ccation in the first- ; portion of the' test. With this path blocked, the system-still satisfied the Technical Requirements Manual minimum number - of thimbles limit of 44.' The path was accessible during the. first - portion of the test, but the detector apparently hung up at the seal table fitting when attempting the demonstration flux map using real detectors and detector cables. Repairs are planned - for: a subsequent outage'.
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Detector Drive Soeeds Distance Time Actual Speed' Allowed Drive (inches) (seconds) (inches / min) =Ranae= l i A 200.5 83.54- 144.0 142-146' B 200.9 83.52 -144'3
. 142-146 C 200.5 83.49 144.1 .142-146, ;
D 200.5 83.48 "144.1- 142-146 1 E 200.5 83.49 144.1; 142-146~ I F 200.6 83.49J 144.2 142-146 e i sa -- s ' ... .
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l' . Figure 3.2'.'4-1 , I MOVABLE INCORE DETECTOR PATH LOCATIONS - l l I R P N N L K J .H- -G F. E D: TC B. A. t i 1 ; MT MT - C B-DET DET DET 2 2 B A F T MT MT DET ' ' 3 ' A C- -D F , I DET DET DET , ; 4 ' A D E. DET DET DET DET 5 ' A D C C. 6 ' DT DET DET DET DET B F B A A= 7 ; T MT NT NT E D B 'E-8 T MT MT MT m MT MT NT F' ,E- D-C F C E .. E T MT MT 9 -A- F: F C~ DET CAL 10 A .B
- DET DET DET DET DET 11 :
B' D D C B DET DET DET 12 : E F A-DET DET DET DET E D B C , DET DET DET DET 14 : '
~B C E D DET DET 15 :
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1 3.2.5 - RCS AND SECONDABY COOLANT-CHEMISTRY fPost Core Load) -: ISU-006A OBJECTIVE j This test is performed to verify-that the water quality within the- i reactor coolant system and the -steam generators- meets the- l appropriate chemistry requirements. The~ test is performed at cold Shutdown (Mode 5), Heat-up Prior to Criticality (Mode 3), at-Criticality (Mode 2), and at approximately 30%,-50%, 75%, and 100% Power. This test' satisfies activities described by FSAR Table 14.2-3, sheet 11. TEST METHODOLOGY ] The testing is performed by obtaining - samples of ' the reactor coolant system and steam generators-from the. appropriate sample l panels. Each sample is then chemically' analyzed. The results of I these analyses are tabulated and: ' compared. to the chemistry j requirements. !
SUMMARY
OF ".ESULTS . During the executions of this test,- all- required - Acceptance criteria were adequately met for each system that was sampled. No l corrective actions in plant operation werei needed to meet the l Acceptance Criteria. On occasions, one'of the samples had to be reanalyzed or retaken because a result was:not consistent with the l others. Upon reanalysis .the sample was-'shown to be within specifications. Tables 3.2.5-1 and 3.2. 5-2 , contain; a summary of the results - for each system sampled along with the Acceptance criteria or guidelines stated within the test - 1 While not required by the test,, Pressurizer samples were also evaluated from Mode 2 through .100% power and were found to-be satisfactory when compared to the RCS - criteria.~ They are'not tabulated because no limits are specified by.the test.- i i 1
-1
TABLE 3.2.5-1 , RCS CHEMISTRY'
SUMMARY
CHEMISTRY MODE MODE MODE , PARAMETER CRITERION 5 3 2 29% 48% '76% 100% Chloride <150 ppb' 3 3 6 1 <1 4~ Fluoride <150 ppb 2 5 <1 4 3 2 6 , Dissolved Oxygen ** <100 ppb N/A 1 2 <5 <1 <1 3 , Lithium _____*** N/A N/A N/A -2.1 1.9 2.0_ 2.0 Hydrogen 25-50 . .
**** cc/kg H O N/A .N/A N/A- 26.4 _27 2 j32.9 28- ,
_______________ __2______________.___________________.______________ - Boron
- 22000 ppm 2063 'N/A N/A N/A. N/A N/A N/A-
< limit _
Gross <100/E_ of L de- 4.6 3.46 1.07 '1.51; 1.92 -i Activity Ci/ml N/A tection E-5 E-2 E-1 E-1 2-1 I Dose Equi-valent <1.0 5.3 <5.9 3.33 1.13 1.22- 1.80 I-131 Ci/ml N/A E-8 E-7 E-4. E-3 E-3 E-3
- Mode 5 test sampled RHR-instead of'the RCS, due to system 1 pressure, as allowed by the test procedure. l e
*
- When Tave >250*F
*** In accordance with Lithium vs. Boron Curve above 1MW thermal (see Figure 3.2.5-1) .- RCS boron 1 concentration was;between 400 ppm and 1200 ppm for:allnat-power test conditions-(>1 MW.
thermal)
**** When RCS >1MW-thermal Reactor Power; L
N/A = Not applicable as no criterion is specified for this' plant ' ' l condition i _. . - . - - J
l i i l TABLE 3.2.5-2 l I STEAM GENERATOR CHEMISTRY
SUMMARY
.-y CHEMISTRY MODE MODE MODE-PARAMETER CRITERION 5 3 2- 29% ~48% 76% 100%-
Cation ' Conduc- 50.8 . .. tivity* mho/cm N/A 1~. 5 0.37 0.56 0.63 0.80: 0.71 l 1 l 28.8 pH** 10.0- 9.2 9. 0 - .9.2, 9.2 9.0 '8.9 Sodium *** 520 ppb 14 51 <1. 14 17 12 6~ Chloride ***s20 ppb- 6 25. 2 7 'll 5 ______________..._______________________________________ ___8_______ j s Sulfate ***s20 ppb 14 29 <2 9' '7 18 3 'l Silica $300 ppb N/A N/A N/A- 250 220- 270. 160- , Hydrazine 275 ppm 80 N/A N/A N/A' ' N/A- N/A N/A NOTE: The recorded value is the'..value from .all 4 steam-generators having the minimal margin' tot each_ criterion. The silica criterion is not applicable in Modes 2,'3 & 5. l 1
- No limit in Mode 5, Limit is-52.0 in Modes 2 & 3 l 1
** Limit is 19.8 in-Mode 5, 29.0 in Modes-2 & 3-l
*** Limit'is $1000 in Mode 5,;s100 in. Modes 2 &.3
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q l 3.2.6 - RADIATION SURVEY TESTS - ISU-208A-OBJECTIVE The radiation survey test is. performed-to determine dose levels at specified points throughout the plant,: to verify the effectiveness, vf radiation shielding, to identify any' areas of streaming through. shield walls and to verify proper posting of radiation areas.- This i test satisfies activities described by FSAR Table 14.2-3, sheet 12. TEST METHODOLOGY Gamma and neutron radiation-dose rate values are established = by. l surveying with portable survey instrumentation in the Safeguards, Radwaste, Fuel, and Auxiliary BLildings, the Unit 1 Containment and; penetration- areas and .the plant outside' perimeter. - Neutron; radiation dose rate values are. established in the ' Unit 1 Containment and certain-penetration areas. Surveys are performed-precritical, critical at 0-5%,'40-50% and 90-100% power. . The key: 1 l results come from the-100% power execution. The lower power-data is used to verify background radiation-- values -and ' to : identify - l potential problem areas prior to reaching full-power. 1
SUMMARY
OF RESULTS , j l The effectiveness-of gamma shielding and the general-determination ~ l of dose levels were found to be adequate during performances of the test. At nominally full power, gamma . dose. rates were predominantly
<0.1 mR/hr rith 21 of 93 locations exceeding'1.0 mR/hr. Of these 21, only 2 exceeded 10 mR/hr. ; one at 12 mR/hr and one at 251 mR/hr.-
Both of these were at .the entrances to - RCS loop ' compartments. I While several points marginally exceeded their expected values, no ; dose rate exceeded the maximum allowedLlimit for that particular= - location. At nominal full power, neutron dose rates:were predominantly <0.5 mrem /hr with only 18 of 92: locations exceeding 1 mrem /hr. Of these'- ' 18, only 5 exceeded 10-mrem /hr; one eachtat 35,:40 and 50 mrem /hr and two points at 100 mrem /hr. At each of these five points above -4 10 mrem /hr, the dose rate was less than the estimated maximum.for While 23 other locations 'did - exceed: the estimated
~
each point. neutron dose rates, these-limits were not absolute requirements and-were not exceeded by more than 4 mrem /hr at:any point. Evaluation by Engine'ering of the measured dose rate values and comparison with l results from five other - similar 4 loop Westinghouse PWR plants I concluded that these results were acceptable. None' of the ? dose I rates was judged to pose extraordinary or. undue limitations on . l personnel access to plant areas during operation. l 1
l l . i' 3.2.6 - RADIATION SURVEY TESTS - ISU 208A (Continued) SW92RU2Z._EESJ&TE (Continued) One of the originally selected radiation base points on the outside of the containment dome was found to be inaccessible and was deleted based on availabil:ity of symmetrically located: points'. Nine additional radiation base points were' deleted based-on ALARA concerns during the at-power measurements. Three radiation base points were relocated due to proximity to area radiation monitor. check sources. The relocations were to nearby areas having identical expected dose rates. Containment penetration survey ~ results Jwere all- withinu allowed limits, indicating that no containment. neutron or gamma' streaming problems are evident. One penetration, indicated a gamma dose rate: of 40 mR/hr. This high dose rate.was~on the chemical;and volume l control system letdown line.from the reactor coolant: system and is indicative of the relatively high activity _ level.of the fluid in this line and is not unexpected. All measured dose rates have been: evaluated as acceptable for plant-operation. e i a i 3.2.7 - PROCESS AND EFFLUENT RADIATION MONITORING PERFORMANCE TEST-ISU-210A OBJECTIVE . This test is-performed to verify proper responses of all' process and effluent monitors and- the failed fuel monitor to- existing ! sources of radiation -(actual process or affluent fluid) . This testi .) satisfies activities described by FSAR Table 14'.2-3, sheet 13.. TEST METHODOLOGY Batch liquid monitors are tested in' either Modes 1 or- 2 when' I sufficient liquid inventory has accumulated to process. A liquid !' sample is taken and the radiochemical analysis of this sample is; compared to the radiation monitor indication. They are expected to agree to within a-factor of 2 of each other. j continuous process liquid and gaseous monitors have' samples drawn: J from adjacent sample ports and~the radiochemical analyses of these samples. are compared to the radiation monitor indication. . They are - also expected to~ agree within a factor of 2. Some monitors do not have associated sample ports. Because these monitors have~ no comparison made, the mon:. tor indication is recorded as a baseline value only. If the radiochemical result is less1than the minimum detectable activity or ,the monitor indication is less than the monitor's operational range, then the factor of 2 criterion doesn't' apply and the monitor is verified 'to be indicating a proper
~
background radiation level. i
SUMMARY
OF RESULTS j With the following listed exceptions, all batch liquid, process and i effluent radiation monitors satisfied the factor of 2 comparison criterion or had their appropriate baseline readings recorded. Several monitors' failed to satisfy: the- criterion during the. lower power executions of. this' test. The verification of . proper performance for these monitors was deferred to the full power test. The full power test is the best indicator of ' system ability to-monitor process stream and effluent radiation- under normal operating conditions. The low power tests are primarily- performed : to verify monitor backgrounds and to establish system operability' prior to ascending to. full power. T l o Spent fuel pool monitors XRE-4180, XRE-4181, XRE-4863 :and XRE-4864 were not tested because the Spent Fuel Pools were-empty. The Spent-Fuel' Pools are to be filled following completion of associated piping support work and these monitors will then be: tested. l
i L 2.7 - PROCESS AND EFRUENT.. RADIATION.MONITQRING PEREORMANCE TEST-ISU-210A (Continued) -)
SUMMARY
OF RESULTS (Ccntinued) o Monitor XRE-3230 was not' satisfactorily tested due; to ! inadequate monitor sample flow. .. The insufficient head 'I available at the Auxiliary Steam Drain' Tank loop seal did not produce sufficient flow through the monitor'and associated in-line sample cooler to clear-the low flowc alarm and permit . monitor operation. This monitor.is-not' safety related and is to be operationally verified-following resolution of the low flow problem. o Monitor 1RE-5179 was not~- satiafactorily -tested due' to inadequate monitor sample flow. LInsufficient pressure existed at the monitor's ' location -in . the . Steam Generator Blowdown-system to provide sufficient flow through the monitor to clear the low flow alarm and permit monitor, operation. This monitor is also to be operationally l verified following resolution of ; the low flow problem. o Monitor XRE-5698, on the Safeguards Building Ventilation- , 4 System, had-a failed detectorLthat cannot.be replaced until the Primary Plant HVAC system can be' isolated.- Primary Plant HVAC cannot be-isolated for this' work until'the plant is in either Mode 5 or Mode 6.- This monitor is)to be operationally. verified following detector replacement. .This; monitor.had no adjacent sample port and the monitor indication ~ was only to be lecorded as a baseline value. o Monitor 1RE-2959, on the Condenser Off Gas System, had a flooded detector chamber.which prevented'its being tested-at 100% power. The detector.and. monitor. functioned properly at 50% power and satisfied all criteria during that test portion. The monitor is' to be operationally; verified following detector replacement. The monitor was" accepted as having passed-this - test. based on the 50% power results. o Monitors XRE-5250, XRE-5253, XRE-5380,'XRE-5567A, XRE-5567B, ; XRE-5570A and XRE-5570B failed the factor.of 2 criterion at > 100% power. These results were evaluated' as 7 acceptele by -
'j Engineering ~ based on t h e ' d e s i g n '. b a s i s monitor ranges : and -
actual discriminator settings. Monitor XRE-5380 had-a-lower level discriminator setting-.of 1251 kev that screened out the 81 kev gamma. from Xe-133. that was included in the radiochemical sample results. The remaining 7 monitors were , evaluated ss acceptable because the individual radiochemical :i sample results were below the design basis operating range of the particular monitor.
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1,2.8 - REACTOR COOIANT FLOW MEASUREMENT -'ISU-023A OBJECTIVE The Reactor Coolant Flow Measurement testLis performed to determine-the Reactor' Coolant System (RCS) flowrates'for.each-of the:4 RCS loops, the total RCS- flowrate,- and' to verify proper- RCS:. flow indications. This test is performed prior to initial. criticality. (Mode 3) and during power ascension at approximately 50%, 75%i and 4 100% power. This test partially satisfies activities described by. FSAR Table 14.2-3,- sheets 2 and.2a and Technical LSpecification-3/4.2.5. 1 TEST METEMLOCY a P: or to criticality, data is obtained from the installed elbow tap ditferential pressure (d/p) instrumentation and used to-calculate - the RCS loop flowrates. Average values for pressurizer pressure, RCS narrow range -cold ' leg temperature and RCS. d/p transmitter output voltages are determined concurrently. The-temperature and-pressure readings' are used to obtain cold leg specific volumes . using Steam Tables. The d/p transmitter output voltage-readings; ! are converted - to inches of HO using .the ' known individual d/p transmitter scaling. 2 Each. loop has three flow. transmitters-from l which a d/p measurement is taken. ~The d/p readings:are'used'to ! q determine three flowrate values for each loop using an equation'for -] Reactor Coolant Cold Leg Volumetric ; Flow 1 Rate as;'a function of- j Elbow Tap d/p and specific . volume. These three: flowrates J are ; averaged to obtain the ' loop , average flowrate. ..The average ; flowrates from all four loops are summed.to obtain!the total RCS flowrato. The flou transmitters are verified to'be aligned'and calibrated by_ review cf the appropriate completed Instrumentation & Controls work , documents. RCS flow indications,. processed ~from the elbos tap d/p 1 transmitv.ers, are read from the P2500 process. computer and verified. to indicate 1 N a specified error tolerance of 1.534. -t With the plant at approximately:50%, 75%1andi100% power, data is. j taken to determine . the RCS loop flowrate. . This : data is a combination of a precision secondary plant calorimetric,.. cold leg i RCS temperature values and-N-16 Transit . Time Flow Meter (TTFM) l outputs. The TTFM is a direct flow measuring device.using gamma. j detectors mounted on the outside of..the?RCSz hot' legs.. RCS water i flowing-through~the reactor..has a portion of the' Oxygen-16' nuclei - 3 present in the HO 2 molecules activated ' to ' Nitrogen-16 by the -l neutron absorption-proton emission reaction. - This N-16 leaving the 1 l reactor has a half-life of.7.10 seconds and - emits gamma rays: of j l 6.129 and " .7.115 MeV. These gamma rays- penetrate the RCS loop l piping and are sensed by the N-16 gamma detectors. The detectors: i u
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_ . _ _ _ __ ~ __ .__ _ ___ __- _ _ _ _ _ _ __ _ _ _ _ ___ _ _ 1
- 3. 2'. 8 - REACTOR COOLANT FIhW MEASUREMENT - ISU-023A- (Continued)
TEST METHODOLOGY (Continued) are located transverse? y to RCSJ loop- flow and ' are co111 mated < to:
~
observe fluctuations in the.N-16' gamma activity as flow passes the detector. To measure RCS-loop flow, the TTFM uses two pairs of gamma detectors located approximately 21/2 feet apart, . 2 detectors , upstream and 2-downstream of each other. Loop volumetric flow is _
?
calculated by multiplying the piping inside cross-sectional area by I the fluid velocity. 'the fluid ? velocity is the known detector l upstream-downstream spacing divided by.. the fluid transit _ time l I [ between them. A statistical cross-correlation.of-the N-16= gamma. signal between upstream-downstream detector pairsLresults in.this transit time. All possible~ upstream-downstream' . detector: combinations are used to calculate transit' times, then combined to i form a mean transit time.: The: cross-correlationLdata collection, analysis and calculation of volumetric flowrate is. performed by'the. TTFM that is connected to-the N-16 detector outputs for the given loop under test.. The TTFM is moved from loop to loop sequentially , and does not measure all'.4 RCS-loop flows simultaneously. The-RCS ; hot leg volumetric flowrates from the :TTFM are converted to RCS - cold leg flows using measuredlRCS cold leg' temperature combined-with a RCS hot leg. temperature that is _ calculated from calorimetric-power, cold leg temperature-and: hot leg volumetric flowrate. -These temperatures are used to calculate hot and: cold leg ' specific 4 volumes and the ratio of' specific volumest is.used to convert. hot leg volumetric flow to cold leg volumetric flow.<
SUMMARY
OF RESULTS All values are in gallons / minute-TOTAL REQUIRED' j RCS- TOTAL i
% POWER LOOP 1 LOOP 2 LQQP_3, ~ LOOP 4 FLOWRATE FLOWRATE MODE 3 101,827 108,093 103,067 107,265 420,252 >344,520 50% 103,494 104,184 104,093 102,257 414,028' >389,700-75% 103,300 104,300 104,000 102,650 414,250 . >389,700 100% 103,331 103,928 103,932- 101,948 413,139' >389,700 L (and also L <420,000); e L The total RCS flowrate must be equal to or greater than- 344,520 gpm (90% of the Thermal Design Flow) asidetermined by elbow-tap d/p.
instruments prior to criticality. This was satisfied'in' Mode 3 r* 0% power. l L L
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- 3. 2. 8 - REACTOR COOLANT FIDW MEASUREMENT -- ISU-023A _ (Continued)
SUMMARY
OF RESULTS (Continued) At 50% and 100% power, the total RCS flowrate must be-greater than or equal _ to 389,700 gpm (101.8% of the Thermal < Design Flow) as determined by the TTFM. This was satisfied. The 50% power results also satisfied the requirements of Technical Specification 4.2.5.4 -! to have a flowrate of greater than or equal. to 389,700 gpm, as '! determined by the TTFM,- prior to exceeding 75% - power. At 100% power the flow was also verified to be less than 420,000 gpm so as >
-not to exceed ' vendor recommended NSSS mechanical- design flow ,
limits. The RCS flow elbow tap d/p transmitters were. verified to have'bden aligned for both zero and 100% flow prior to Mode 3 testing. The indicated percent RCS flows at normal RCS operating conditions' . in Mode-3 ranged from 99.9 to 100.8% which satisfied the specified 100% i 1.53% flow range. , I
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3.2.9 - REACTOR COOLANT SYSTEM FLOW COASTDOWN - ISU-024A OBJECTIVE The Reactor Coolant System Flow Coastdown test is-performed with the unit in Hot Standby (Mode 3) to verify that the measured core ' ~ flow during Reactor Coolant Pump (RCP) coastdown exceeds the , flow decay assumed in the accident analysis during flow decay. In addition, the low flow trip time delay is> verified'to be.within This test satisfies activities describedL by , acceptable limits. FSAR Table 14.2-3, sheet 3.
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TEST METHODOLOGY Strip chart recorders are connected to the '~ RCS elbow tap - d/p-transmitter outputs and- the Solid State Protection System .(SSPS) to . monitor Reactor Coolant System. flow characteristics - and - Reactor . Trip Breaker positions as a function of time. -A P-8 permissive is simulated (>48% power) to ensure that a single loop loss of--flow: results-in generation of a reactor trip signal. All 4 ~ Reactor Coolant Pumps are tripped by manual actuation. of the RCP ,l Underfrequency Trip relay. Flow and SSPS data are taken while the . RCS flow decays. All 4 Reactor Coolant pumps are verified to trip ~ j within 0.100 seconds of each other to ensure that.the ficw decay ; data corresponds to essentially a simultaneous loss of alliforcedJ RCS flow. Data from : the strip charts is then statistically , evaluated to verify acceptability of-the measured flow values and > related time-delays.
SUMMARY
OF RESULTS q The required Flow Coastdown Time ' Constant was. required to be- ' greater than or equal to 11.64 seconds...The measured value was 13.81 seconds. . The Low Flow trip time delay was required to be. less than or equal to 1.0 seconds. The measured-value was 0.976 seconds. The Reactor Coolant Pumps were also verified to trip ~ within 0.055 seconds of each other, which was well within the 0.100. ! second limit. - I 1 I
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l' 3.2.10-- REACTOR COOLANT SYSTEM u AKAGE RATE - ISU-022A~ OBJECTIVE ( l-The purpose of this procedure 'is-.to verify the Reactor Coolant- .I System (RCS) leak tightness after the system has been' closed. This ~ ; test satisfies activities described by FSAR Table 14.2-3, Sheets 29-and 30. I TEST METHODOLOGY
.1 With the plant in Hot Standby (Mode 3)' conditions,; prior to initial '
criticality, the reactor coolant system is tested to verify leak tightness. - Af ter RCS pressure .is stabilized, - a visual ~ leak. test is conducted with the. reactor pressure 1 vessel, pressurizer andL all . four reactor coolnnt loops vorified to be leak-tight. Also,. the q unidentified, identified,- and 1 controlled : leakage, rates are determined using normal ' operating ~ Technical- Specification surveillance techniques L or results - from ' OPT-303A and~ . OPT-110A. Pressure isolation valve leakage iss also verified, based oninormal' Technical Specification surveillance results'~from EGT-712A. Primary to secondary leakage: is' determined ;by measuring boron concentration of the steam generator liquid.' This calculation is-based on RCS boron concentration, steam 1 generator' boron concen - i I tration, steam generator blowdown' flowrate and time. This primary to secondary leakrate11s measured inJ gpd, gallons per-day. Under normal conditions the minimum detectable boron concentration of'0.2 ppm would result in a calculated leakage rate'of'5.76.gpd. ,i
SUMMARY
OF RESULTS I During . the visual. inspection, no pressure boundary. : leakage Ewas l observed nor was any leakage past.the Reactor-Vessel; flange seal observed from the flange seal leakoff. No-boron,was detected in the steam generators, so the conservative..'O.2 ppm :value was assumed. Leakage rate results are' tabulated lbelow: Leakage Acceptance Rate Tvoe Criterion (com) Test Results (comi Controlled 5 40 39.P Identified 1-10 0.027 Unidentified -$~l 0.*S l Pressure Isolation Valve 55 . 3.13 Primary to Secondary 5 500 gpd/ steam <5.76 gpd/ steam = generator generator l l i w -. . -
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- 3. 2.10 - REACTOR COOLANT SYSTEM TRAKAGE RATE - ISU-022A (Continued)- 1 tj
SUMMARY
OF RESULTS (Continued) 1 1 The technique used to measure Controlled Leakage with the Chemical _j and Volume control System flow control valve (1FCV-121) fully open, failed'to satisfy the $ 40 gpm criterion the-first three times it-j was attempted. This technique was identical to that contained in surveillance OPT-110A. OPT-110A was revised based on information I from the NSSS vendor and from other, similar 4 loop Westinghouse ) PWR plants.- The revised technique satisfied the 540 gpa criterion- ;l and the results from the executed OPT-110A were used to satisfy this test requirement. This test verified acceptable leak tightness of the reacter coolant a system. .i ll J I l l i _ - _ _ _ _ _ _ _ _ . _ _ _ . _ _ _ _ _ _ _ _ _ . _ _ _ . _ _ _ _ _ __ -- _ _ . _ _ _ - _ _ .--y w - o* ~ T- f
i 3.2.11 - COLD' CONTROL 10D E .RABILITY TESTING ~- ISU-026A OBJECTIVE The purpose of this test is to verify coil polarities, proper Digital Rod Position Indication 1(DRPI). system' operation, rod drop 1 timing, alarm functions,- DC Hold : Cabinet operation, and proper slave cycler timing and to perform an operational check of each control rod drive mechanism -(CRDM)- with a rod cluster control. , assembly (RCCA) attached prior to initial use of the mechanism. l This test partially satisfies activities described by FSAR Table-14.2-3, Sheets 4 and 5 and Technical Specifications 3/4.1.3.3 and 3/4.10.5. TEST METHODOLOGY This test ja performed under two plant ~conditionst Mode 5 - cold, no flow and Modes 4 and:3, full.-flow.c Proper operation of coil polarities, DRPI operation,-rod dropitiming, CRDM operation and i slave cycler timing are verifiedLunder cold, no flow conditions. The rod bottom, rod deviation, urgent and non-urgent failure alarms; and the DC Hold Cabinet are tested in Modes 3; 4 and 5. Coil polarities are verified to preclude individual magnetic fields l from the stationary gripper, movable gripper and lift coils from l interfering with one another. This test uses.a. battery to. inject low voltage pulses into the moveable gripper coil and observes the direction of current flow induced in the other two coils. Then the voltage is injected ~ into : the stationcry. gripper coil and the direction of induced current flows in the other two coils is again verified. Each of the 53 CRDM coil stacks is~ individually tested in this manner. Slave cycler timing and CRDM. operational' checks are ' performed starting with all RCCAs positioned at the core bottom. A selected ; single bank is withdrawn 50 steps to ensure the RCCAs are above-the dashpot region. Each RCCA . in - the withdrawn bank is then individually-withdrawn.5 steps and reinserted 5 steps. 'When all RCCAs in a bank have been tested,.the entire; bank'is reinserted to the bottom of the core. This is then repeated for each bank. While the individual RCCAs are being withdrawn and . inserted 5 steps, a visicorder 'is used to monitor lift coil, stationary gripper coil and movable gripper coil currents. .An optional signal from a microphone _ attached to the CRDM housing is: used to help relate actual mechanical events to the. coil currents. These visicorder traces .are evaluated to verify that the coil currents were of the proper. shape, the currents were of the proper magnitudes at the ' proper times, and to verify. - that . the events associated with -mechanism movements occur in the proper order. The
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- 3. 2.11 - COLD CONTROL ROD OPERABILITY TESTING - ISU-026A(Continued) l 1
TEST METHODOLOGY (Continued) traces for each RCCA are compared against vendor supplied criteria t.nd model traces or against the first actual trace taken, that for the RCCA at core location B-12. . Rod speeds are calculated from the-pernod of successive rod steps as being the inverse of stepping- 1 frequency, l 1 DRFI system operability is verified by monitoring DRPI -Light Em..tting Diode (LED) indications on the control board during bank-withdrawal and comparing these indicatioAs against other indications of RCCA position; .the - P2500 process computer, the demand step counters, and the pulse to analog - converter. A selected bank of RCCAs is withdrawn to 12 steps and slowly reinserted to verify when the RB (rod bottom) LED illuminates for each RCCA. The bank is then withdrawn to 231 steps, the mechanical. RCCA limit of motion. Shutdown bank withdrawals are. stopped at 18, 210, and 228 steps to record the various position indications-listed above. Control bank withdrawals are stopped every 24: steps and at 228 steps to record this data. .These periodic indication verifications are also used to demonstrate Technical Specification operability of the DRPI system per Surveillance Requirement -l ' 4.1.3.3. With a selected rod-bank fully withdrawn to 231 steps, the' DRPI data cabinets in the containment building are : de-energized. A l dedicated, personal computer based, Data Acquisition System (DAS) is hooked up to the DRPI data: cabinets and the- reactor trip breakers are then opened. As the RCCAs - drop, the slightly magnetized, individual CRDM drive' shafts, which are connected to the RCCAs, also drop through the deenergized DRPI sensing. coils'and induce a current in these coils 'which is _ proportional to : drop ' 4 velocity. As the RCCA enters' the dashpot , region of the fuel j^ assembly guide tubes, it is hydraulically braked, which also shows I up as a significant velocity-change in the induced current signal.'- l The DAS records the induced; current signals:as_a function of time from-all RCCAs in the selected bank, - a . sign'al proportional to , l stationary gripper current, and an event mark'for the' opening of-the reactor trip breakers. From this information;the rod drop time to dashpot entry can be evaluated. This. rod drop-timing testing:is - performed only one bank at a time,,but is. performed simultaneously- ~ j for all RCCAs within-a bank. The Surveillance Requirements for- ' Technical Specification Special Test Exception 3.10.5 are satisfied , within this test to allow the DRPI system to be de-energized for- 1 rod drop timing measurements. -{ l Optionally, the drop times from all RCCAs are statistically evaluated and any RCCAs with. drop times deviating more than'.two standard deviations from the mean drop time for all RCCAs may be redropped to confirm their actual performance. l l
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- 3. 2.11 - COLD CONTROL ROD OPERABILITY TESTING --ISU-026A(Continued)
TEST METHQDOLOGY (Continued) The rod deviation and dropped.-rod 1 alarms are verified by i withdrawing all shutdown RCCA banks, withdrawing Control Bank A to l 18 steps, and then moving individual RCCAs as necessary to activate the particular alarm being tested. - The rod deviation alarm is ,, I verified by deviating two RCCAs in-Control Bank A by 12 steps or more and also by partially inserting a shutdown Bank C RCCA from its fully withdrawn position.- The dropped rod alarms are verified by inserting one RCCA from Control' Bank A to near full insertion and then by inserting a' second = RCCA for : the: 22 ' rods at - bottom alarm. This alarm circuitry is such that a successful test using any RCCA verifies the alarm for all other associated RCCAs. The non-urgent failure' alarm is= tested-by removing the input power fuses for one power supply in each of;the five rod drive system's ., power cabinets and the logic _ cabinet. The cabinets are-tested-sequentially, a not simultaneously. - The urgent failure alarm is ' tested by interrupting the lift coil-firing circuit to-all~RCCAs-powered by the rod drive system power cabinet under test. - When the
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RCCAs associated with -the' cabinet . under test are ordered .to ~ withdraw, the urgent failure' alarm actuates: insresponse to the' missing lift coil current. The cabinets are tested' sequentially,- , not simultaneously, using the: permanently installed lift : coil' l disconnect switches. The logic cabinet urgent ' failure alarm is tested by removing a preselected circuit board. 4 The DC Hold Cabinet serves-as an alternate power { source-to hold.a I single group of RCCAs in a withdrawn position to allow for maintenance on the stationary gripper power circuitry - for ' that - group. A group of RCCAs consists of 2,3, or 4'. RCCAs. .The DC Hold- \ ' Cabinet is tested by switching it to hold a-group:of'4 withdrawn - RCCAs, de-energizing the normal power circuitry for'that_ group and ; verifying the RCCAs remain withdrawn. , .I l-
SUMMARY
OF RESULTE 1 Coil polarities were all verified to be correct when tested in I Mode 5. ] The current and sound traces from all 53 RCCAs were all verified ~ proper when. tested in Mode 5. The traces were all'of'the proper shape with no notable anomalies. The timing cf events and current magnitudes were all verified to be-acceptable ~ and concurred with by l
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a 3.2.11 - COLD CONTROL ROD OPERABILITY TESTING - ISU-026A(Continued) . ,
SUMMARY
OF RESULTS (Continued) vendor representatives. Actual rod speeds from evaluation of the-inverse of the period of successive rod-steps were:as follows: Expected- Actual. RCCA Bank Tyne S,gif.d ( stens/ min) Sneedfatens/ min) _} Control Bank 48- -45.5-Shutdown Bank A or B 64 62.82 L
- Shutdown Bank C,D or E 64 63.5 i
There were no tolerances on the expected speed values because these . rod speed values are to . be remeasured ' at hot (Mode.~3) RCS : conditions. These cold values . are . expected 'to differ- from the expected speed values -due to . mechanical (thermal expansion) conditions associated with the low RCS' temperature. . One. problem was noted with respect to rod speeds.- While taking trace data for i Shutdown. Bank C, the time between . rod steps was observed to be' ' significantly smaller than - expected.. Evaluation of the traces resulted in a measured rod speed of 75 steps / minute. An adjustment was made to the rod speed circuitry for that cabinet and the final ' i value following this adjustment was recorded above. The DRPI LED indications on the main control ~ board and P2500 computer were verified - to be within .14 steps J of the rod L drive system group step counter indications for all 53LRCCAs. Theiactual deviation was O steps. The pulse to analog converter-indications , were verified to be-within il step of ~ the group step fcounter , i indications for all 4 control banks. The_ actual agreement was also 3 exact, with a deviation of.0 steps. The DRPI LED indication for rod bottom (RB) indicated at or prior . to ' reaching. zero steps,. as indicated by the group step counter, during RCCA insertion.; The RB LEDs all illuminated.at-an indicated 3 steps. :This.DRPILtesting , was performed in' Mode-5. The rod deviation alarms functicned properly for an actual rod vs.. rod deviation of 12 steps and. a . shutdown - rod : at '210- steps-withdrawn. The rod bottom and 12 rods ~at bottom alarms functioned properly in response to actual RCCA insertions. 'These-alarms were tested in Modes 3 andE4. The urgent and non-urgent alarms functioned properly 'in response to failed power supplies, missing : lift coil currents and 'the missing circuit. board. These alarms were. tested.in Mode.3. t The DC Hold Cabinet was verified to hold a group.of 4 RCCAs in a-withdrawn position for 10 minutes. This verification was performed in Mode 3. 1 l l l ________.__ _ _ _ _ _ L _
1 l ! f 3.2.11 - COLD CONTROL ROD C _2RABILITY TESTING - ISU-026A(Continued)- ' e
SUMMARY
~OF RESULTS (Continued) I All rod drop times were verified to be .less than the Technical I Specification limit of 2.4. seconds. .That limitidoes not actually j apply to this test performance because the' limit is for a hot, full '
RCS flow test and these rod drops were done cold in Mode 5. The average drop time was-1.603 seconds.i:No two standard deviation , redrops were performed. They were not required at this plant ! condition. The rod drop data was taken for baseline l purposes and I to verify rod ' drop- DAS ~ operation only. : Additionally,' evaluation of the rod drop DRPI coi) current traces' verified proper operation of the dashpot decelerating devices, i The test' procedure directed the performance of the . normal' ' Instrumentation and controls = calibration procedure which was performed in Mode 3 to further confirm proper operation, of all portions of the DRPI system.
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Three miscellaneous problems were noted with respect to rod drive system operation during performance of this test - -i o Blown fuses within both rod driva system. motor-generator sets' ; generator voltage control circuitry resulted in improper phase voltages when the generator field was flashed. The fuses were replaced and the motor-generator- sets operated properly thereafter. i o Irregularities were noted with the operation of the main control board switch that closes the reactor trip breakers. Closure of the breakers is dependent on-how long the switch is held in the closed position and'sometimes also requires-two switch actuations to close the breakers. Opening.(tripping) of the reactor trip breakers is' unaffected: by this closing . problem. Operator awareness of this condition: eliminated further problems, o CRDM testing 'in Mode 4 resulted .in occasional rod misstepping due to dissimilar thermal expansions between the. CRDM and the Control Rod Drive Shaft (CRDS) which.resulted in mechanical misalignments of the CRDM gripper teeth and CRDS grooves. The rod motions were proper in both Modes 5 and 3. In both Modes 5 and 3 the tooth and groove alignments were correct but the temperature regime in Mode 4 is such that the alignments were not quite correct. The mechanical. design -is such that it - allows for proper engagement when ' cold or hot,; but not i necessarily when in between. Normally, rod motion'is demanded only when hot or cold, it is-not customary-to move rods in-Mode 4. The testing was -deferred t) Mode 3 and' was satisfactorily performed there. l
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- 3.2.12 - HOT CONTROL ROD OPERABILITY TESTING YSU-027A OBJECTIVE The purpose of this test is to verify proper DRPI system operation, rod drop timing, rod speed and direction, overlap operation, manual operation and to perform an opentional check of each CRDM with a .
RCCP attached in Mode 3-prior to ?nitial' criticality. The actual l mechanical RCCA withdrawal limit is' also verified. - This test I partially satisfies activities described by FSAR Table 14.2-3, Sheets 4, 19, 31 and 32 and Technical Specifications - 3/4.1.3.4 and- l 3/4.10.5. TEST METHODOLOGY Slave cycler timing, Control Rod Drive Mechanism (CRDM) operational' checks, measurement of the mechanical withdrawal limit, DRPI system checks and rod drop timing is performed in an-integrated fashion,. on a sequential bank by bank basis. .The selected bank is withdrawn to 228 steps, with the operation of every DRPI LED verified during. l this withdrawal with respect to the group step counter indications. l There is an LED for every 6 steps of RCCA' motion. Shutdown banks a have no LEDs to represent position between 18 and'210 steps, only a transition region (TR) LED. Each individual RCCA in the, withdrawn bank-is inserted 5 steps and then withdrawn 10_ steps, ) ending at an indicated position of 233' steps on-the-group step-counters. During these 5 and 10 step-movements, visicorder trace ; data is taken as was done in ISU-02 6A'. Refer to Test Summary 3.2.11. This-trace data is also evaluated as was done in ISU-026A with respect to rod speeds and. timing and magnitudes of coil current changes. Sound. traces are taken for'only.one CRDM per rod drive power cabinet due to microphone integrity concerns while at' normal RCS operating temperature. The_ trace data is also evaluated to verify the ' mechanical RCCA withdrawalu limit. When the CRDM drive shaft reaches its mechanical limit of travel there are-'no more grooves on the Control Rod Drive ShaftJavailable for the CRDM grippers to latch into. This shows up on,the trace asla gripper-current anomaly. The traces are evaluated near the top of travel, above 228 steps, with respect to where thisLanomaly' occurs. This mechanical limit is typically 231 steps but may_ vary from. plant.to plant. Once the mechanical withdrawal limits:have beenidetermined for all RCCAs'in a bank, that bank is dropped toLmeasure rod drop l times as was done in ISU-026A. . This' set of L rod drop times l satisfies Surveillance Requiremente for Technical Specification L 3.1.3.4 and is performed at RCS- full . flow conditions. The-Surveillance Requirements for Technical ~ Specification Special Test ! l Exception 3.10.5 are also satisfied.Within this' test. ; Any_ RCCA~~ having a drop time-deviating from the.mean drop _ time'by more?than two standard' deviations is redropped: an. additional' three times to confirm its actual performance. l
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- 3. 2.12 - HOT CONTROL ROD' OPERABILITY TESTING - ISU-027A (Continued)
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TEST METHODOLOGY (Continued) J Rod speed and direction indications;on the main control board are j verified while' withdrawing and . inserting various RCCA banks. l Control bank overlap is. verified by withdrawing the control banks in the manual' overlap mode instead of in the individual bank select-modo of operation. As-a prerequisite to this test portion, the-overlap switch settings are changed to lower, yet sequential, values. This allows verification of-overlap without the need for-complete withdrawal of the control banks. As'the control banks are withdrawn in manual - overlap, data is - recorded each time. a bank-starts or stops motion. This data ' is compared to 'the switch 1 settings. The ability of an urgent -failure alarm to block RCCA . motion 'is. tested by creating'an actual urgent failure alarm, by, interrupting > lift coil signals using permanently installed disconnect switches, and then attempting to move-the RCCAs. The urgent failure alarm is then cleared and RCCA motion is verified to have been: restored.
SUMMARY
OF RESULTS The current traces from all 53 RCCAs were verified to be proper. , The traces were of the proper shape with no notable anomalies. The timing of events and current magnitudes were all verified to be acceptable and were concurred. with . by vendor representatives. Sound trace data: from each power cabinet was:also verified to be-proper. t Actual rod speeds were verified as follows:- Expected Actual-RCCA Bank Tvoe SDeed(steos/ min) SDeed(steos/ min) Control Bank 4 Sci 2 46.2 Shutdown Bank A or B 64 12 62.9 < l Shutdown Bank C,D or E 64~i2- -63.2
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The DRPI system indications were typically.within 1 or-2 steps of the group step. counter indication.with only one group of 4 RCCAs
'.off by 3 staps at one rod position, Shutdown Bank D at 207.vs. 210 steps. This satisfied the i4 step-agreement criterion.
The mechanical withdrawal limit _was established to be 231 steps.by inspection of visicorder trace data"above 228 steps ~. This trace data was repeated for Shutdown Bank A due to legibility problems-with that portion of the original trace data for that bank. 1This-trace. data was also repeated for Control Bank A due to a visicorder paper jam. r.
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j i 3.2.12 - HOT CONTROL ROD OPERABILITY TESTING'- ISU-027A (Continued) l
SUMMARY
OF RESULTS (Continued) ., 1 Rod drop timing measurements'were made for all 53 RCCAs from the- l 231 step full mechanical withdrawal' position. All times'were less I than 2.4 seconds from decay of stationary gripper -. voltage to~ ,l d:rJpot entry. The fastest RCCA took 1.30 seconds. -The slowest . l RCUA took 1.47 seconds. The average RCCA drop time was - 1. 4 0 seconds with a standard deviation of.0.03 seconds. Only two RCCAs-were outside of the two standard deviation limits. They were - redropped 3 times each with the following'results: Dron Times (ceconds)- Dron Tvoe RCCA F-6: RCCA H-14 Original 1.30 1.47 . Redrop #1 1.33 1.43' ) Redrop #2 1.33 1. 4 2 :- , Redrop #3 1.32 1.43 ! i- These redrops also satisfied the criterion.that for each redropped RCCA, the three redrop times-shall all.be within a 0.02 seconds ' band. This rod drop tining test satisfied the Surveillance i Requiremente for Technical Specification:3;1.3.4 and Special Test Exception 3.10.5. One minor problem occurred while taking rod drop timing data. The RCCA at Core Location H-12-initially' generated a bad trace due to poor DRPI system cabinet test. probe contact. A' new test probe was used and the:RCCA-was successfully retested. The rod speed and direction indications on the main control board, > were verified to be correct. The; speed indications of either 48 or 64 steps / minute were correct. Control bank.ovarlap was verified to , occur exactly at the overlap switch settings with no deviation. t This satisfied the allowed il step-deviation criterion. l The urgent failure was generated and was verified to -inhibit manual , RCCA motion. RCCA-motion-was-restored-following clearing of the 4 alarm. f l 1 l -
i 3.2.13 - REACTOR TRIP SYSTEM TESTS - ISU-015A 9BJECTIVE The purpose of this test is to. verify proper operation of - the automatic and manual reactor trip breaker circuitry-and to verify proper operation of the reactor trip breakers prior to initial critical:ty. This procedure also tests reactor trip bypass breaker-functions and verifies proper unlatching of the control rods following opening of the reactor trip breakers.- -Thi.o test satisfies activities described by FSAR Table 14.2-3, Sheets'6 and-7. ( T2ST METHODOIDGY The Solid State Protection System (SSPS) general warning l interlocks associated with the trip breakers 'are tested - by closing- . both - i reactor trip breakers (RTBs) and one of the trip bypass breakers, - i (TBBs). The . SSPS train opposite to- the TBB thst is closed'is= placed into test and it is-verified that all three breakers _then open automatically. This sequence is repeated for the: other TBB-and SSPS train. l TBB interlocks are tested by clos!,ng one TBB'and verifying that_an attempt to close the second TBB results in the automatic opening of both TBBs. This sequence is repeated with the other TBB starting < in the closed position. ; Functional testing of RTB and TBB ' operation iis performed by closing both RTBs and one TBB. A trip signal is then: simulated on the SSPS train associated with the closed TBB. It is verifjed that the RTB' corresponding to the tripped SSPS- train opens and the other two breakers remain closed.- This sequenceais repeated for the-other TBB and SSPS train. Manual trip function is demonstrated by closing both-RTBs and one ; TBB and generating a- manual t r i p s i g n a l'. f r o m a c o n t r o l- b o a r d reactor trip' switch. All three breakers are, verified to open and the remaining TBB is closed andLverified'to onen-in response to a second actuation of the reactor: trip switch.- This1 sequence is repeated for the second main control beard' reactor trip. switch. Verification of actual control rod release following RTB opening is tested by withdrawing all 53 control rods to 12 steps and manually i initiating a trip signal using a main control board-reactor trip inserted All control rods are= verified to return to their fully switch. System. positions using the Digital . Rod ~ Position Indication
A , . . . . , , 3.2.13 - REACTOR TRIP SYSTEM TEST ISU-015A-l(Continued) HylrAARY OF REST /LTS SSPS general warning interlocks were' verified to properly result in the opening of the RTBs and TBBs. TBB interlocks were verified to properly prevent simultaneous closure of > ' both TBBs. Proper function of RTB and TBB operation was verified, demonstrating that' the TBBs permit individual RTB trip testing without resulting in an 1, actual reactor trip. The manual reactor trip. switch trip function was properly demonstrated for both main control: board reactor trip switches. All 53 control rods were verified totunlatch and fall from the 12 step position to the fully inserted position:following opening of the RTBs. Only one problem occurred during test performar.co. As discussed-- previously in the Summary of Results, for ISU-026A, 'the main control 4 board control switch often required multiple actuation to close the: i reactor trip breakers. This had no adverse, impact:on these test ,
;j results because.the trip' function of this switch did not require multiple actuation.
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- 3. 2.14 - PDMSURIZER SPRAY AND HEATER CAPARILITY - ISU-021A OIL 7ECTIVE This test is to verify pressuriser spray and heater ef fectiveness. performed In addition, the spray line bypass valves are ad$usted to maintain spray line temperature above 540'F. This test partially satisfies activities described by FSAR Table 14.2-3, Sheets 2 and 2a.
TEST METHODOIDGY In order to set the spray line bypass flows, the spray valves and spray bypass valves are closed and the line temperatures allowed to stabilize. The valves are then opened in 1/16 turn, or greater, increments until a satisfactory temperature reading is achieved. The spray line low temperature alarm is verified to actuate at 540 14 'F. To verify spray effectiveness, the heaters are manually isolated and both spray valves are placed into the full open position. Pressurizer parameters are monitored via strip chart recorders. j Thase parameters are then analyzed and plotted to verify the ' pressure transient falls within the allowable limits. Data is also taken for the response to a single spray valve opening. l To verify heater effectiveness, the spray valves are manually l isolated and the heaters are placed to the full on position. Pressurizer parameters are monitored via strip chart recorders. These parameters are then analyzed and plotted to verify the . pressure transient falls within the allowable limits. l To verify stable pressuriser pressure control ability, the spray j valves and heaters are manually operated to adjust pressurizer pressure to approximately 2200 psig. The controls are placed in l automatic and pressurizer pressure is verified to stabilize within l the normal operating band of 2235 i30 psig. A similar test is also ' performed starting at approximately 2300 psig.
SUMMARY
OF RESULTS The pressurizer spray bypass valves were properly set to ensure s that adequate s pray line temperatures exist when the spray valves are closed. Th:.s prevents excessive spray line cooldown which can cauce potentially deleterious thermal eftects on piping and 4 I components when sprays are activated. Valve IRC-8051 was set to 4 turns open and valve IRC-8052 to 1 1/2 turns open. These settings' result in spray line temperatures of 543'F, which also allows for sufficient margin above the 540'F low temperature alaria sStpoint. Testing also determined tb-t these settingc are the minimum alve positions that can maintai, lino temperatures adequately above j
3.2.14 - PDRRSURIEER SPRAY AND HRATER c1DABILITY - ISU-021A (Continued)
SUMMARY
OF RESULTS (Continued) 540'F and that, at these valve settings, pressurizer control-heater bank C could not maintain pressuriser pressure by itself, without periodic backup heater bank actuation. Periodic backup heater operation is not a safety or operability concern, only one of efficiency. Initial testing of spray bypass valve IRC-8051 could not achieve the originally expected spray line temperature of 548'F at any valve position, full open yielded 546.6'F. The plot of spray line temperature vs. valve position did indicate a significant _ slope change, a plateau, at 545'F. This plateau region is the area at' I which changes in valve position have minimal influence on spray line temperature. Investigation into the basis of the original test requirement of a 548'T minimum spray line temperature rea?alted in the change of this value to 540'F. This permits operation at 543'F with 3'F of margin to the 540'F low temperature alarm and still allows 10'F margin from the alarm setpoint to the 530'F minimum spray line temperature basis. . I The spray line low temperature alarms were verified to be set at 539.38'F and 539.56'F which satisfied the 540 1 4'F criterion.The l pressure transient resulting from the spray effectiveness test fell I within the required band, refer to Figure 3.2.14-1. The opening of 4 a single spray valve was verified to result in an average pressure decay rate of 75.4 psi / min. The opening of both spray valves. l resulted in an average rate of 88.3 psi / min. The spray effectiveness plot demonstrated that the spray valves capacity was such that the pressurizer pressure response was sufficient to respond to design plant transients but not so large as to cause excessive rates of change of pressurizer pressure and temperature. Two points of the pressure transient resulting from the heater , effectiveness tests fell outside the required band, refer to Figure i 3.2.14-2. The heater response was mostly in the low range of the allowed band indicating that the effectiveness of the heaters was marginally sufficient to support optimal response to design transients. However, sufficient heater capacity was judged to be available to adequately support design transients. The Technical Specification heater capacity requirements are more than adequately satisfied. Engineering and the NSSS vendor evaluated the data and determined the test results to be acceptable to support subsequent plant operations. The pressurizer pressure was verified to stabilize at 2235130 psig when controls were placed in automatic from starting points at approximately 2200 and 2300 psig. No sustained or diverging oscillations were noted. 3.2.14 - PDFASURIZER SPRAY AND MRATER ciDARILITY . ISU-021A (Continued)
SUMMARY
OF RESULTS (Continued) Following the completion of this test, the spray line low temperature alarm was reduced to 525 13'F .Jy a- plant design l modification. i i 1 l
l 1 Figure 3.2.14-1 l P i i i t PRESSURE RESPONSE TO OPENING BOTH SPRRY VALVES, ! i I o PRESSLRIZliR PRESSJRE N n 2200 N,' 9 ' 2175 > s, ,,, N, - w 2150 W e 2,25 N,
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i l l 3.2.15 - )(ISCEr.TANEOUS BA1ANCE OF PLANT TESTING i Q1qECTIVE > These tests are performed to verify proper performance of various l balance of plant systems and components which cannot be fully tested prior to actual power operations. The systems to be tested ; J include Main Feedwater, Steam Dump Valves, Steam Generator ' Atmospheric Relief Valves, and the Main Turbine and Generator. l These tests were performed to identify any significant problems in i the secondary plant and to perform tuning of controls to optimize plant performance. This testing is not described in the Final Safety Analysis Report and is included for information only. TEST METHODOLOGY These tests are performed at the earliest time possible (lowest ! practical power level) during-the initial startup program. Some tests are also repeated at other power plateaus, as appropriate. The linin Feedwater System Test determines if there is any significant leakage past the Feedwater Isolation and Feedwater . Isolation Bypass Valves by monitoring downstream piping temperature I before and after starting a feedwater pump. Prior to opening a ; Feedwater Isolation Valve, the purge flow through the Feedwater Isolation Bypass Valve is measured with an Ultrasonic Flowmeter. At approximately 50% power, the Auxiliary Feedwater check valve backleakages are verified to be insignificant by monitoring temperature profile of the upstream piping. At approximately 75%, . 90%, and 100% power, the feedwater split flow to the steam generator upper nozzle (steam generator auxiliary feedwater nozzle) is measured and verified to satisfy the limitations provided by the NSSS vendor. 1 The Feedwater Pressure. Oscillation test monitors the magnitude of j the continuous pressure oscillations in the feedwater piping at the inlet to the steam generator main nozzles. These pressure oscillation measurements are used to verify the fatigue capability of the steam generator preheat section structure. Piezoelectric pressure transducers installed in the feedwater line near the main j inlet nozzles provide signals to a test data acquisition system. ' This data is evaluated with a spectrum analyzer and a plot of the peak-to-peak amplitude versus frequency- is compared to the allowable pressure oscillations provided by the NSSS vendor. This data is collected for the following plant conditionst o With only auxiliary feedwater being supplied to the steam , generators i o Following transfer of Feedwater flow from the upper nozzle to the main nozzle with one main feedwater pump in service o While bringing the second main feedwater pump into service o At steady state conditions of 90% and=100% power cperation o Following a 50% step load reduction.
1 i
.h2 15 - MISCET TANEOUS RALANCE OF PLkNT TESTING (Continued) ,
i l TEST METHODOIDGY (Continued) l l { The Main Feedwater Pumn Performance Tggt verifies the Fump and pump j
- turbine trip functions,- the pump hydraulic performance in the ,
j recirculation mode, and the response of the recirculation valves. The steam Generator Atmosnheric Relief Valve Canacity Test is I performed to demonstrate the operability of the relief valves under ! hot, stemming conditions. With the plant at approximately 25% reactor power, each relief valve is manually opened and the increase in feedwater flow is used to calculate the flow capacity : of each valve. The Turbine Generator Initial Synchronization and oversneed Test performs an overspeed trip test, including an actual turbine trip, synchronizes the generator to the grid and verifies no excessive vibrations occur on the hydraulic control lines to the turbine. The turbine speed is brought up to 1800 rpm and then overspeed using the Trip Testing Lever until the Hydraulic Governor Stop. Setting causes.a trip. After the turbine has slowed to turning gear speed, the turb;.no speed is raised back to 1800 rpm and the generator is synchronized to the grid. The Electro-Hydraulic Control (EHC) hydraulic lines are monitored for. vibration during < the test. The Dynamic Automatic Steam Dumn Control test demonstrates the capability of the Steam Dump system Plant Tri:p, Load Rejection, and Steam Pressure controllers to control eather Tavg or Steam Pressure. With the steam dumps in Steam Press'.tre mode, reactor power is increased from 0% to approximately 44 and decreased back to 0% by control rod motion.. Steam pressure is verified to remain stable within the control band of 1092 i20 psig. With the reactor power at approximately 1%, a simulated trip signal (P-4) injected and Tavg elevated to approximately 560'F, the steak. dump controller is placed into automatic. While reactor power is increased from 1% , to 5% and then reduced.to 3%, the reactor temperature and steam dump valve response is monitored to verify that the Plant Trip controller maintains temperature correctly at approximately 557*F. Then, reactor power is increased to approximately 5% and " Turbine in Operation" and " Loss of Load" condition signals are simulated to cause the steam dump controller to be in Load Rejection mode. The reactor coolant average temperature and steam dump valve responses are monitored to verify the. Load Rejection controller maintains tLaperature correctly. The Steam Dumn Performance and Timina Test verifies the time ; , response and stroke length of each of the steam dump valves. With l the steam dump isolation valves closed, the steam dumps are I I modulated open with the pressure controller. The fully closed to fully open stroke lengths are measured. The valves are timed as
i j l 1
- 3. 2.15 - MISCELIANEOUS BALANCE OF PIANT TESTING (Continued)
! TEST METHODOLOGY (Continued) l ' 1 they trip closed by deenergizing each solenoid. The valves are ' also timed as they modulate closed using the pressure controller. Finally, the valves are timed as they trip open in response to the ! Tavg controller. l The Steam Dumn valves Canacity Test determines the steam flow capacity of each bank of three steam dump valves. Each bank of steam dump valves are individually opened while maintaining turbine load constant. The resultant increase in power is verified to be I approximately 10% for each bank. This increase is determined by i measuring changes in feedwater flow which is more - precise . than nuclear instrumentation at these plant conditions. 1 The Atlance-of-Plant Data Collection test is pe'. formed to gather data for the PEPSE program to calculate the Main Turbine and Secondary Plant component performances and pir.nt heat rate and for the recording of baseline secondary systems p9rformance data.- With the plant operating at steady state vonditians, data is collected by data acquisition systems, the pr.acess corputer and manually from , plant instruments for a period of one to two hours. The data is entered into the PEPSE program for calcula tions and analysis. The PEPSE program calculates Plant Heat Rate, Main Turbine Efficiency, Moisture Separator performance, Reheater performance, Main and
- Auxiliary condenser performance, Feedwater, Condensate and Heater Drains pumps performance, and Feedwater Heater performance.
SUMMARY
OF RESULTS The NSSS Vendor evaluated the Main Feedwater System Test temperature data from the feedwater piping before and after a feedwater pump start and determined the Feedwater Isolation and Feedwater Isolation Bypass Valves of all four loops were acceptably leaktight. The purge flows through the Feedwater Isolation Bypass , Valves were determined to satisfy the review criterion. Loop Purae Flow Review Criterion 1 88,378 lbm/hr 60,000 to 120,000 lbm/hr 2 93,810 lbm/hr 60,000 to 120,000 lbm/hr 3 110,914 lbm/hr 60,000 to 120,000 lbm/hr 4 86,981 lbm/hr 60,000 to 120,000 lbm/hr Prior to performance of this test, problems with the Auxiliary - Feedwater check valves leakage were well- documented and operational controls were in effect to monitor and minimize this leakage. The NSSS vendor evaluated the temperature data recorded'during the test from the auxiliary feedwater piping.and this indicated that i
l l
- 3. 2.15 - MISCEY.f ANEOUS BAI.ANCE OF PLANT TESTING (Continued)
SUMMARY
OF RESULTS (Continued) l l none of the eight check valves experienced significant leakage at f i that time. The NSSS vendor recognized the ongoing evaluations of j this issue. As discussed in TXX-90188, dated May 18, 1990, TU > Electric is planning to order check valves of a different design for this Aux:.liary Feedwater application to cover the contingency I that the replacement of the present valves becomes appropriate. i The results of the split flow measurements are as follows 'l 6 (all values are in 101bm/hr): Flow to Flow to Main Upper Nozzle Criterion Nozzle at criterion Lg.gR at 75% Power for 75% Power 100% Power. for 100% Power & 1 0.171580 0.0379 - 0.3785 3.308- S 3.39- l 2 0.198823 0.0079 - 0.3785 3.428 s 3.39 3 0..72559 0.0379 - 0.3785 3.417 s 3.39 4 0.190818 0.0379 - 0.3785 3.274 5 3.39 The 100% power measured flows to the #2 and #3 steam generator main nozzles exceeded the limitation provided by Westinghouse. Westinghouse has recommanded that operation above the specified limit may continue for the remainder of cycle.1 and that the steam . I generator preheater tubes be inrpected at the first planned outage of the steam generators. Westinghouse has recommended raising the hi!h flow alarm setpoint to 3.55 x 10' lbm/hr which is 93.9% of full flow. Nominally 10% of full flow is expected to bypass the main nozzle and flow through the upper nozzle. Evaluation of this , condition is ongoing. The Feedwater Pressure Oscillation test started in Mode 3 and j finished at the 100% power plateau. The peak to peak amplitude versus frequency plots were dell within the allcwable limit curves provided by Westinghouse.
- I The Main Feedwater Pumo Performance Test verified all the pump and pump turbine trip _ functions were satisfactory. -The pump and i turbine auxiliary equipment operated correctly and the recirculation valves operated satisfactory. During the 1B feedwater pump turbine pcrformance test, the pump inboard bearing )
overheated. Following realignment of the pump, the two-hour performance run was completed successfully. All other tests I results wore satisfactory. . l 1 l
.0
! I I j 3.2.15 - MISCET.fANEOUS BALANCE OF PLANT TESTING (Continued) {
SUMMARY
OF RESULTS (Continued) The Steam Generator Atmoscheric Relief Valve Canacity Test, Revision 0, verified that each vulve fully opened and closed under hot, steaming conditions. The measured valve flow capacities were not consistent with each other or with the specified test i criterion. Valves 1-PV-2325 and 1-PV-2328, loops #1 and #4, were i ratested in Revision 1 of the test procedure and still had insufficient flow capacity. Revision 2 of the test procedure i ratested all four valves and again the valves' capacities were + calculated to be too low, with the capacity of valve 1-PV-2326 ) significantly lower than the other three valves. Based on this data and a re-evaluation of the design basis for the steam generator atmosphoric relief valves, a design modification was made to the atmospheric relief valves to increase their stroke lengths j from 1 3/8 to 1 7/16 (+1/16,-0) inches. Valve 1-PV-2326 was also l found to need a control loop recalibration, which was performed. It was determined that the method of measuring steam flow capacity used in the test was not accurate enough and-.that measuring the valve stroke length was a more accurate mecsure of capacity. The l valves' stroke lengths were verified to be acceptable and the . valves were declared operable. The Turbine Generator Initial Synchronization and oversneed Test ' was satisfactorily performed. The turbine overspeed trip occurred i at 1984.5 rpm. The acceptance criterion was to trip between 1980 and 1998 rpm. The generator was successfully synchrenized to the grid at 1530 hours on 4/24/90. The EHC lines showed no excessive vibrations. During the initial attempt to overspeed the turbine, the SPEED REFERENCE signal to the control room was found to be 4 incorrect an( the SPEED REFERENCE card was recalibrated. . i The Dynamic Automatic Steam Dumn Control test was performed prior I to power ascension above ten percent reactor power. The Steam Pressure controlDr properly maintained steam pressure between 1072 and 1112 psig. During the Plant Trip controller and the Load Rejection controller portions of this test, Tavg was not maintained at the temperature anticipated by the procedure. However, the steam dump valve response was evaluated and still determined to be acceptable for both modes of the controller. The error was caused by the conservative gains set in the nuclear instruments, thus indicating higher reactor power than actually existed. Notes b;on reaching 30% reactor power, a secondary calorimetric was performed to correct the gain settings-on the nuclear instruments. Other problems encountered in this test included the need to add a jumper to simulate a trip condition and the need to remove the lead function of the steam dump control card to improve controller l response. l l (
. i I
- 3. 2.15 - MISCEY.Y ANEOUS RA1ANCE OF PIANT TESTING (Continued) l 1
SUMMARY
OF RESULTS (Continued) The Steam Dumn Performance and Timina Tant results were as follows:
-l Stroke solenoid. Modulate . Length Trip Time Closed Trip open
- l Valve Agn)g (in.) gigsad(sec) Tima(sec.) Time (sec.) J i
Acceptance Criteria 2 3/4 11/8 <5 $20 <3 ) 1-PV-2369A I 2 3/4 2.8 17 2.2 l 1-PV-2369B I 2 3/4 3.2 25 l '. 8 j j 1-PV-2369C I 2 3/4 2.6 18 2.5 1 1-TV-2370A II 2 3/4 2.6 15 3.2 , l 1-TV-2370B II 2 3/4 2.8 24 2.1 , J 1-TV-2370C II 2 3/4 2.8 15 2.5 j 1-TV-2370D III 2 3/4 2.6 13 2.4 - 1-TV-2370E III 2 3/4 2.6 13 2.6 1-TV-2370F III 2 11/16 2.4 12 . . 2.0 1-TV-2370G IV 2 3/4 2.4 10 2.2 1-TV-2370H IV 2 11/16 3.2 10 2.8. i 1-TV-2370J IV 2 3/4 2.6 10 2.6 i The stroke lengths for valves 1-TV-2370F and 1-TV-2370J were
- i initially too short and were readjusted to satisfy the criterion.
The modulated close times for valves 1-PV-2369B and 1-TV-2370B were too long to meet the criterion. The engineering evaluation of this data determined these times were within the manufacturer's tolerance, and the overall response of all the valves compensated , for the slightly longer times on these two valves. - These two valves are not in the same bank. These two valves were readjusted 1 to reduce their closure times following the~ completion of this-test. The trip open time of valve 1-TV-2370A . was too long to e satisfy the criterion. The engineering evaluation of this. data t determined that this time was also within the manufacturer's tolerance and the results of this-test are acceptable. The indicated Steam Duno Valves Canacity Test results were as follows: i l l Bank I = 23.055% of rated power Bank II = 21.656% of rated power Bank III = 21.859% of rated power . Bank IV = 26.962% of rated power
, . . . . ~ - , , . . ~ . . - . - , - . - - - - , . . - . . - .- - - . . - - -
. - - _ _ _ . - . - _ - - - - _. . .. .- . - - ~ - _ . - -
l l J l 1 3.2.15 - MISCEffANEOUS RALANCE OF PLANT TESTING (Continued)
SUMMARY
OF RESULTS (Continued) i The expected results were that each bank would be worth 10% i2% of ! rated reactor power. During the initial testing of Bank IV, Valve i 1-TV-2370H did not open. It was readjusted and stroked and the test repeated for Bank IV. An engineering evaluation of these i results determined that the change in power caused by opening the , i steam dump valves resulted in changes to the steam flows in too ) many other flow paths for the increased feedwater flow to be used i to accurately measure the flow equivalent of power change. It was ! concluded that this test provides a qualitative information for the flow passing capacity of the steam dump valves but not a true ; quanti.tative value and that the test results are acceptable for the ' purposes here. The operational valve capacities were later verified in the plant transient tests. , The Balance-of-Plant Data Collection test started with the plant initially at 45% power and was performed periodically throughout the remainder of the initial startup. This permanent plant procedure will also continue to be used throughout the life of the plant. Initially, the test was used as a means to identify any performance problems and excessive heat losses. Numerous steam leaks were identified and corrected. Tuning, recalibration, and modification of the design of secondary system instruments, controls and processes were identified and performed as a result of this test. Additional potential improv sents in performance were identified for further evaluation and ft.ure implementation. The following set of data is an example of the results obtained: Actual Value Parameter Desian Value on 7/24/9.0 Gross Output (MWe) 1163 1130 Turbine Power (%) 100 97.16 Condenser Vacuum ("Hg) 3.38 3.37 Feedwater Flow (1bm/hr) 15,140,015 14,951,403 Feedwater Pressure (psia) 1172 1136.9 Steam Pressure (psia) 975 1007.98 Heat Rate (BTU /KW-HR) 10,048 10,268.94 Heater Drain Flow (lbm/hr) 5,289,632 5,588,051 Feedwater Temperature ('F) 440 437.3 HP Turbine Inlet Pressure (psia) 880.28 889.09 LP Turbine Inlet Pressure (psia) 152.2 128.3 LP Turbine Inlet Temperature ('F) 510.8 524.0 S/G Blowdown Flow (GPM) 0.0 329.1 Nuclear Power (%) 100.0 98.29 l i l
1 l j ) 1 3.3 PHYSICS TESTING 3.3.1 - INVERSE COUNT RATE RATIO MONITORING. (Initial Criticality , i Portion) - NUC-111 J l OBJECTIVE , This permanent plant procedure is performed to obtain and evaluate J nuclear monitoring data during the approach to criticality to i ensure that the approach is done in a cautious and controlled manner. This procedure satisfies activities described in FSAR ! i Section 14.2.10.2. TEST METHODOLOGY . Neutron count rate data, as an indicator of core nuclear flux, from - , both installed source ' range NIS channels is taken periodically during core reactivity additions. The sources of the core neutron flux are the four installed Californium- primary neutron sources with associated subcritical multiplication due to the loaded fuel lattice. As control rods are withdrawn and boron is removed from the RCS water, the core neutron flux and source range count rates increase due to the reduction of these neutron absorbers in the , core. If the count rates were to become very large,'this would indicate that the reactor was approaching criticality. To determine the effect of a given change on core reactivity, count rate data taken after the neutron absorber decrease is compared to > a reference value to evaluate the effect of the neutron absorber decrease. This comparison is performed as a ratio of the count rates to evaluate the fractional change. If this ratio were to be very large, it would indicate that this neutron absorber decrease brought the reactor significantly closer to criticality. For convenience, the procedure evaluates the inverse of the count rate ratios (ICRR) such that an approach co zero would indicate an approach to criticality. Additionally, this procedure trends the t inverse count rate ration and extrapolates the trends to evaluate , what additional neutron absorber decrease would be expected to i result in criticality. Prior to the start or the approach to-criticality, background counts are taken to allow the verification of adequate source range channel signal to noise ratios. This data taken at nominally 557 F in Mode 3 is compared against similar data taken cold, at approximately 115'F, in Mode 6 at the end of core loading. This preliminary cold data was taken in the earlier ; performance of NUC-lll. An equation relating the cold and hot i count rates to the signal to noise ratio is supplied by the core ! designers. This equation is used to verify that a signal to noise ) ratio of at least two exists. Reference values are redetermined i just prior to the start of the dilution with control rod banks withdrawn and also to renormalize the ICRRs. When an ICRR value falls below 0.3, it is renormalized by using the latest average count rate' as the new reference value. This effectively resets the l
I l 3.3.1 - INVERSE COUNT RATE RATIO MONITORING, fInitial Critica1ity Egrtion) - NUC-111 (Continued) ; , J l TEST METHODOLOGY (Continued) I ICRR plot to a value of 1.0. Renormalization improves the I resolution of the plot and may also be done at the discretion of ; the test engineer. , Count data is taken periodically during tho' approach to criticality. Data is taken, the ICRR calculated, plotted, trended and extrapolated as a function of bank withdrawal following each incremental control rod bank withdrawal, as specified by NUC-106. l This is every 116 steps for shutdown banks and approximately every 50 steps for the control banks as withdrawn in normal overlap. Count rate data is also taken, and the ICRR calculated, plotted, trended and extrapolated during the dilution to initial criticality. The ICRR values are plotted, extrapolated and ; 4 evaluated as both a function of elapsed time during dilution and mixings and also as a function of the quantity of reactor makeup water added. The plot versus water added is-a. good indicator of core condition change as a function of the quantity of dilution-(neutron absorber removed) but contains discontinuities at points where the dilution is stopped and the RCS is allowed to mix. The plot versus time does not have mixing discontinuities. For constant rates of dilution, this plot would approximate the ICRR versus quantity of water added plot. A Chi-Squared statistical analysis of the count data is performed to verify data quality. If , the Chi-Squared value is unsatisfactory for the three data values taken, then additional data is taken to calculate an average value. Eventually, as the count rate increases when approaching criticality, only one data value is taken and no Chi-Squared ; analysis is performed.
\
SUMMARY
OF RESULTS Refer to Figures 3.3.2-1 through 3.3.2-3 for ICRR curves during the , approach to initial criticality. ; All count rate data was properly recorded and ICRRs were calculated, plotted, trended and extrapolated. The ICRRs show that the approach to criticality was performed in a cautious and controlled manner with no indicated unexpected approaches toward criticality. Large changes in the ICRR plots are due to renormalization. Monitoring data was properly taken and evaluated during RCS mixing. 1 Renormalizations and reference count rates were properly I recalculated,. The signal to noise ratios were-calculated-to be 1 52.9 for Source Range Channel N31 and 76.9 for Channel N32. These both satisfied the 22.0 criterion by wide margins.
I I i 3.3.2 - INITIAL CRITICALITY - NUC-106 j l ! OILTECTIVE This permane.it plant procedure provides a method by which initial " criticality is attained in a deliberate and controlled manner. :' This procedure is used to enter Mode 2 for the first time. The I sequence, frequency, and conditions for collection of nuclear data is specified as well as the method of analysis of this data. ! criteria for suspending the approach to criticality and for emergency boration are also specified. This procedure satisfies , activities described in FSAR Section 14.2.10.2 and Technical Specification 3/4.10.3 and 3/4.1.1.1. TEST METHODOLOGY - Initial conditions are established with the RCS at an average temperature of approximately 557'F, RCS pressure at approximately 2235 psig, RCS boron. concentration greater than 2000 ppm', and all control rod banks fully inserted. Procedures are initiated to monitor neutron flux, boron I concentration and various other plant parameters for the duration of the test. Reference counts are determined for each. source range channel per l NUC-111 . These values are used in the ICRR (Inverse Count Rate Ratio) calculations performed following reactivity additions. [ Physics Testing is declared to be in progress to permit usage of Technical Specification Special Test Exception 3.10.3 with respect to Mode 2 testing at off normal conditions and also based on the-predicted positive MTC. Shutdown banks are then withdrawn in their normal, alphabetical order. The withdrawals are made in increments of 116 steps or less and the value of the ICRR is determined after each wLthdrawal, prior to subsequent withdrawals. These ICRR values are plotted against the cumulative shutdown bank position to trend and predict by extrapolation any unexpected approach to criticality. Each bank is withdrawn to an indicated 232 steps, .the rod drive step counters are reset to the actual mechanical withdrawal limit of 231_ steps, and the bank is reinserted to 228 steps. This sets the control rods to their proper full out heights for monthly control rod repositioning to reduce localized control rod cladding wear. The Mode 2 entry checklist is verified to be completed and the control banks are then manually withdrawn in their normal overlap configuration, in nominally 50 step increments. Mode 2 is entered with the initial withdrawal of Control Bank A. Control- bank withdrawal is completed when control Bank D is positioned at 160
I I J f l 1 3.3.2 - INITIAL CRITICALTTY - NUC-101 (Continued) l TEST METHODOLOGY (Continued) steps. During control bank withdrawal, proper bank overlap and rod , insertion limit alarm functions are verified. ICRR monitoring, , plotting and extrapolation is also performed as was done for the ' 4 shutdown banks previously. 4 , The remaining reactivity insertion required to achieve criticality is made by diluting the RCS boron concentration by addition of reactor makeup water to the RCS. Periodic ICRR monitoring during the dilution is performed using NUC-111 to plot, trend, and i extrapolate predictions of expected time and quantity of water added for initial criticality. The dilution rate is initially approximately 60 gpa until the ICRR value falls below 0.3, where the dilution is then terminated and the RCS allowed to mix. The ICRR is also renormalized at this point. A dilution at approximately 30 gpm is then started and . maintained until the renormalized ICRR value again falls below 0.3 at which time the RCS dilution is again terminated to allow for mixing._ Criticality is ' achieved during this mixing time period, by Control Bank D notion )' or by small batch water additions. The core flux level is then increased to and stabilized at approximately 10 s amps using Control I Bank D anotion. CJMMARY OF RESULTS 1 The Acceptance criteria was met in that criticality was achieved l within the range of the predicted boron concentration, and the i' neutron flux level was established within specified bounds on the Intermediate Range NIS channels. The predicted critical- boron I concentration was 1139 ISO ppa and the measured value was 1153 ppm. ! The neutron flux was increased and stabilized at 9x,10*' amps and 9.5x10 amps on the Intermediate Range channels. 'This was approximately lo a amps, as required. Shutdown bank withdrawals began at 1803 hrs on 4-2-90 and were completed at 2000 hrs. Mode 1 2 was entered at 2101 hours on 4-2-90 with Technical Specification j Special Test Exception 3.10.3 invoked at 2109 hrs. The 60 gpm RCS dilution was started at 0209 hrs on 4-3-90. This initial dilution was terminated at 1448 hrs on 4-3-90 and, after mixing, a 30 gym dilution was started at 1616 hrs. The'ICRR_ plot at that time , indicated that the addition of 2200 gallons of water was necessary ! to achieve criticality. 1000 gallons of water were added by 1650 hrs and allowed to mix. Following a 37 minute mixing time, the addition of another 1200 gallons was started at 1727 hrs._ Initial criticality was achieved at 1742 hrs on 4-3-90. The final 1200 gallon dilution was terminated early, with only approximately 450 gallons actually added. The flux was-stabilized at approximately 10'8 amps at 1806 hrs on 4-3-90. t l i 1 3.3.e - INITIAL CRITICALTTY - NUC-106 (Continued)
SUMMARY
.OF RESULTS (Continued) l Three minor problems occurred during the approach to criticality, 1 none of which involved any unexpected core reactivity responses. Noise in the Power Range channel N44 signal used by the reactivity computer was traced to the N-16 circuitry Power Range Module. The noise amplitude was low enough to be ins;.gnificant with respect to this circuitry's function during normal operations but was large enough to interfere with this signal application for physics test ; measurements. The Power Range Module was de-energized and the noise was eliminated. The Rod Insertion Limit alarms did not clear when expected for control Banks C and D. These alarms were declared inoperable and were later recalibrated by Instrumentation and control. The alarms cleared at higher rod bank positions than expected which was conservative. The alarms were not needed for this criticality approach because the rod bank positions established prior to the i dilution were well above these limits. During the dilution, RCS and pressurizer boron samples deviated by more than the 150 ppm difference criterion originally used. . This was caused by mixing time delay between the RCS and pressurizer combined with sampling purge delays. The criterion was changed to
+ 200, - 50 ppm to account for the pressurizer lagging the RCS during dilution. This did not present a safety concern because.the pressurizer boron concentration was always higher than the RCS boron concentration, as would be expected.
Refer to Figures 3.3.2-1 through 3.3.2-3 for plots of ICRR versus control and shutdown bank withdrawals, reactor makeup water addition and time. i l l l
-100-
._m___._ ____ _. _ _ _ _ _ _ _ _ _ _ _ _ . _ _ . _ _ _ - .__._____ _ _ _ _ __ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ .
. . . - , - ---er -- - - " + r-'ew w e-
Figure 3.3.2-1 ICRR During RCC Bank Withdrawal CPSES Unit 1, Cycle 1 Initial Crit ICRR 1.2 1.1 1
~
.y- w.
p e'% / "2 ._
~
4 4 v - 0.9 y N O.8 g ; O.7 0.6 0.5 ,
- 0.4 0.3 0.2 0.1-0 0 100 203 300 400 500 600 .700. 800 900 1000 1100 1200 1300 1400 1500 1600 1700 Total Steps Withdrawn-N-31 - N-32
ARO Pos=228 steps, Overlap =115 steps
-10'-
u-c , 7 e, 3 e- s -W-c 9 -+ *-=v-
,w, w-ys gW. en yi,p -
f M -m=e-- .e- wver-- ~ ..w=me e eif ---mys g ,e, , ,.w. . . - -eew- = e--+ w-
Figure 3.302-2 Q ICRR vs. Time During RCC Boron Dilution CPSES Unit 1 Cycle 1 Approach to Initial Criticality ICRR 1.2 02:36 14: 55
. r-t g <- ,
! 38,sca m r, l g*% 7 a,, 0.6 1.- i 1 0.4 0.2 - 'kg di 17:4 0 0 ,. 100 200 300 400 500 600 700 800 -900 1000 1100 Time (span in minutes)
-~ N31 -B-M32~
-102-
- - - ~ ~ ~ , . . . , . _ _ . _ . . . .._,.____-_._-----~_-._-_____m_m__._.m__,___.___._.m-
Figure 3.302-3 ICRR During RCS Boron Dilution CPSES Unit 1, Cycle 1 initial Crit ICRR 1.2 i 1.1 1 - 0.9 h
% ~.
O6 05 0.4 s 0.3
=
~ 0.2 0.1'
\
.O.
0 5000- 10000 15000 20000 25000 30000 35000
' Reactor Makeup Water Added (gal) .
N-31 N - e
-103-
-- - +
. - , +_ _ _ _ _ = _ _ _ _ _ , _ . . . _ _ _ _ _ _ . _ _ _- _
_ _ _ - - - . -__ ___.v v r . . + < - - - - - ___ -z__,- - - - - - _ _ . _ . . .
.c _____--.._m . _.. .m___. --
i 3.3.3 - DETERMINATION OF CORE POWER RANGE FOR PHYSICS TESTING. l NUC-109
- OBJECTIVE I This permanent plant procedure is used to determine the power level .
(neutron flux level). at which detectable reactivity feedback effects from nuclear fuel heating occur and to establish the range' i of neutron flux in which zero . power reactivity measurements are performed to avoid interference with these feedback effects. ; TEST METHODOLOGY Initial conditions are established with the RCS at an aversgo , temperature of approximately 557'F, RCS pressure at approximatelz 2235 psig and the reactor critical with f3ux at approxnmately 13' amps on both Intermediate Range channels. Control Bank D is positioned such that approximately 40 pcm of wot.h 6 remains , available to increasa core reactivity. j Initially, the reactivity computer is stst up using the po, tar range ! channel N-44 detector, which was taken out of service. Ratctivity computer outputs of reactivity and flux along with RCS c71d leg temperature are displayed ,- strip chart recorders. The temperature input is from the tess instrumentation rs.cks. l The determination of the power ,,e for physics testing is made by withdrawing control Bank D . achieve a positive reactivity addition of 30 i 10 pcm. Reactivity and flux level are then observed to determine the point of adding nuclear heat as indicated by negative reactivity addition from the Doppler fuel temperature coefficient. RCS temperatures ara also monitored for an increase as an indication of nuclear heating. The flux is then reduced back to approximately 10'8 amps and the measurement is repeated, at least once, to confirm the value.
SUMMARY
OF RESUlTS Two measurements were performed. The reactor power level at. which detectable reactivity feedback effects from nuclear heating - occurred was determined to to 1 X 10 amps on both Intermediate Range (IR) NIS channels and 1. 3 X 10*6 amps on the reactivity computer picoammeter. These values are from the second measurement which was more refined than the first. During the first measurement, the point of adding heat is completely unknown. It is common to actually overshorst the value on this first run. However, even if overshot, the first run does yield an approximate value of the point of adding heat such that the second measurement can begin with a good prediction of where the point of adding heat lies. The
-104-
i 3.3.3 - DETERMINATION OF CORE POWER RANGE FOR PHYSICS TESTING. i NUC-109 (Continued) l
SUMMARY
OF RESULTS (Continued) , i first measurement resulted in values of 1.1 X 10*' amps (IR Channel N35), 1.3 X 10 *' amps (IR . Channel N36) and 1,65 X 10*' amps , (reactivity ccaputer picoammeter) for the point of nuclear heat ! addition. The neutron flux level range at . which zero power reactivity measurements were to be performed was determined to be 1 X 10 s to 1 X 10*T amps as indicated on the reactivity computer. The range j of neutron flux levels allowed for physics testing that was e I actually trended and used applied to the reactivity computer picoammeter. ; h I { i
-105-i
i- l l, i i 3.3.4 - REACTIVITY COMPUTER CHECKOUT - NUC-108 t OBJECTIVE i This permanent plant procedure is performed to demonstrate proper l operation of the reactivity computer through dynamic testing _using
- actual neutron flux signals and core reactivity changes. This ,
ensures that.the reactivity computer-is operating properly before i it is used to measure reactor physics parameters. 9 TEST METHODOLOGY i A reactivity increase of approximately 25 pen, as shown on the . reactivity computer strip chart, is initiated by withdrawal of Control Bank D. A stopwatch is used to measure the reactor period time. This period, P, is the time interval over which~ indicated core flux increases by a factor of e, with flux increasing on a stable period, after the dampening of initial transient effects. ! The period copes from the fo,llowing equation in terms of the-initialflux,p/,finalflux,g,andmeasuredtimeinterval,tt P = t/in - . This measured period .is used to determine the theoretical ' reactivity increase using core design ~ report predictions of reactivity as a function of reactor period. This prcJiction is given by the inhour equation using core physics constants from the- !' core design report. The predicted reactivity increase is compared to the reactivity indicated on the reactivity computer strip chart. This measurement : l may be repeated for reactivity increases of up to approximately +50 pcm. A negative reactivity insertion of up to
-20 pcm may also be optionally performed.
I'
SUMMARY
OF RESULTS Two runs were made, one each at approximately '25 and + 50 pcm. No negative reactivity insertion runs were made. The . acceptance criterion for this measurement is that the average of the absolute l '
-106- ---w.. , , . . - , ,y ww- -,w
5 I l l 3.3.4 - REACTIVITY COMPUTER CHECKOUT - NUC-108.(Contfnued) j l
SUMMARY
OF RESULTS (Continued) values of the reactivity differences be less than 14%. The results were as follows: ' Measured Predicted Indicated Approximate Reactor Reactivity Reactivity Absolute Reactivity Period Based on from Value of Insertion fnem) fseconds) Period feca) Connuter fDem) %Diffarence i 25 298.65 25.2 25.0 0.80 50 141.5 48.3 48.7 0.82 e Average % Difference 0.81 { The average difference of +0.81% satisfied the < 14% criterion. I i l l l l
-107-l l
)
l
- 3.3.5 - CORE REACTIVITY BAT.ANCE - NUC-205 )
.QRJECTIVE The purpose of this permanent plant procedure is to verify the design predictions of core reactivity during the power ascension startup terting sequence. This procedure satisfies activities 1 described by FSAR Table 14.2-3, Sheet 17. 1 TEST METHODOLOGY j This core reactivity verification is performed by comparing reactor criticality parameters at- zero power with those at full power. ;
Parameters measured include control bank positions, RCS ; temperature, RCS boron concentration, power level and core burnup. t After compensating for differences in control bank position, boron 1 concentration, reactor power and Xenon and Samarium buildup, with j respect to predicted core conditions, the actual critical e boron 1 concentration present is compared to the design prediction. This verifies the accuracy of the design predictions of core reactivity. ,
SUMMARY
OF RESULTS The Hot Zero Power, All Rods Out, Xenon and Samarium free critical boron concentration was 1162.1 ppm. The Hot Full Power, All Rods ! Out, equilibrium Xenon and Samarium critical boron concentration ; was 754.1 ppm at an average RCS temperature of 589 'F and 999 i MWD /MTU burnup. This represents a decrease of 408 ppm in boron concentration to get from. Hot Zero Power, All-Rods Out and no fission product poisons to just critical at Hot Full Power, All Rods out and equilibrium Xenon'and Samarium. For a 1000 MWD /MTU burnup, essentially the same as the 999 MWD /MTU burnup as tested, , the predicted Full power value was 743 ppm. Using the predicted just critical Hot Zero Power, All Rods.out, no fission product poison value of 1146 ppm yields a predicted difference of 1146 - 743 = 403 ppm. There is no specific criterion for this agreement, -
- but 5 ppm indicates that the core reactivity change was very close l to the design predictions. This indicated that the design predictions of Xenon and Samarium buildup' and power defect were valid.
-108-c 1
l~
l 1 l 1 I 3.3.6 - SURVEIT.fANCE OF CORE POWER DISTRIBUTION FACTORS - NUC-201 OBJECTIVE l The purpose of this permanent plant procedure is to evaluate i reactor core power distribution factors based on incore flux map results. This procedure partially satisfies activities described ( in FSAR Table 14.2-3, sheets 20-22, section 14.2.10.4 and Technical ' Specifications 3/4.2.2 and 3/4.2.3. TEST METHODOLOGY , The results from an incore flux map are processed using the CONFORM > core physics code. The axially varying heat flux hot channel' factor, FQ(Z), is evaluated by inspection of the CONFORM code , results for the maximum FQ(Z) . Values from portions of the core ' known to have relatively uncertain results are eliminated and the-remaining FQ(Z) value closest to the FQ(Z) limit is then multiplied i by 1.05 and 1.03 to account for manufacturing tolerances .and measurement uncertainties. The portions of the core eliminated are : the top and bottom.15% of the core, regions within 12% of grid straps and the region within 12% of the Control Bank D RCCA tips. This maximum FQ(Z) result is compared against the appropriate limit, 52.32/P for P>0.5 and $4.64 for P $0.5, multipined by the factor K(Z) from Technical Specification Figure 3.2-2 evaluated at the same Z core height as the FQ(Z) value. P is the fractional thermal power. The radial peaking factor, Fxy, is evaluated by inspection of the CONFORM code results for the maximum Fxy, eliminating-values from the same portions of the core as - above for FQ(Z) and again by multiplying by 1.05 and 1.03. This maximum Fxy result is compared against a power varying limit of SFxy RTP (1 + 0.2(1-P)), where P is fractional thermal power and Fxy RTP is defined in the Radial Peaking Factor Limit Report required by Technical Specification 6.9.1.6. For this initial startup, Fxy RTP was 1.55 for unrodded core portions and 1.71 for rodded core portions. The evaluations of FQ(Z) and Fxy are performed in conjunction with the appropriate sequence document and may also be used to satisfy the Surveillance Requirements for Technical Specification 3/4.2.2. The nuclear enthalpy rise hot channel factor, FDHN, is evaluated by inspection of the CONFORM code results for the maximum FDHN value and multiplying it by 1.04 for measurement uncertainty. This maximum FDHN result is compared against the limit of 51.55 (1 +
- 0. 2 (1-P) ) , where P is fractional' thermal power. This FDHN evaluation is done in conjunction with the appropriate sequence document and may also be used to satisfy the surveillance requirements for Technical Specification 3/4.2.3.
-109-
I 1 1 3.3.6 - SURVEIf.fANCE OF CORE POWER DISTRIBUTION FACTORS - NUC-201 l (Continued)
SUMMARY
OF RESULTS This procedure was performed as necessary to support ' sequence document performances throughout power ascension. The required ) power plateau maps were those flux maps taken at approximately 0%, ! 50%, 75% and 100% power. The results were recorded in the test ! summaries in Section 2 of this report. All performances were satisfactory. Some performances were also used to. satisfy l Technical Specification requirements as well as criteria contained' in the sequence documents. i N e 1 1
-110-
1 i ( )
- 3. 3.7 - ZERO POWER ISOTHERMAL AND MODERATOR TEMPERATURE COEFFICIENT MEASUREMENTS - NUC-207 OBJECTIVE This permanent plant procedure is performed. to determine the i Isothermal Temperature Coefficient (ITC) of reactivity and to i derive, from this, the Moderator Temperature coefficient (MTC) of reactivity at the beginning of core 1; fe. This procedure satisfies activities described by FSAR Table 14.2-3, Sheet 14 and Technical i specification 3/4.1.1.3. l TEST METHODOLOGY The ITC is determined by measuring the_ change in reactivity ' induced by changing the temperature of the moderator, cladding and fuel and.
dividing by the temperature change. The MTC is obtained by analytically removing a precalculated Doppler broadening coefficient factor from the ITC. value to eliminate the fuel temperature change portion of the-ITC. , A voltage signal proportional to core reactivity is obtained from the reactivity computer output and a voltage signal proportional tt RCS cold leg temperature is obtained from the process instrumentation racke. These signals are input to an X-Y plotter such that the slope of ".he X-Y plot corresponds to the ITC, change in core reactivity pee unit change in RCS temperature. RCS temperature is slowly changed by manipulation of the rate of heat removal from the RCS by the secondary plant. The resulting reactivity as a function of the varying temperature is plotted and evaluated. By changing the RCS temperature slowly, the fuel, cladding and moderator temperatures all change at the same rate, nearly isothermal, with minimal temperature gradients. This measurement result must be adjusted to eliminate the effect of Doppler resonance peak broadening in the fuel to yield the effect of the moderator alone. The effect of.the cladding is negligible in this analysis. The slow RCS temperature change permits the fuel temperature to change uniformly, isothermal, without the heat transfer that would result in a non-linear temperature profile across the fuel pellets. Because cf this, the fuel temperature at a given time is essentially the sama as the RCS temperature. This allows a feel type and enrichment specific calculation ' of the Doppler broadeteing effect to be performed for the temperature regime at which the test is executed. Variable fuel temperature distributions would render analysis of the isothermal temperature coefficient impossible. The Doppler broadening coefficient of
-1.83 pcm/'F is subtracted from the ITC value to result in the MTC.
The measured ITC values are evaluated,to verify they.are within il pcm/'F of each other, to demonstrate data consistency, and the
-111-
j
- 3. 3.7 - ZERO POWER ISOTHERMAL AND MODERATOR TEMPERATURE COEFFICIENT MEASUREMENTS - NUC-202'(Continued)
TEST _MFJRQDQLQGX (Continued) average ITC value 'is verifiedJ to ' be within 13 pcm/'F of the-predicted ITC value of -:-1.4 pcm/'F. - The MTC:is verified-to be so pcm/'F or, if not, Rod Withdrawal: Limits, using NUC-116, are imposed to ' ensure that the MTC is maintained $0 by. operational ! controls. This test is performed from Hot Zero Power conditions,-nominally 5 57'F , starting with a Reactor Coolant Systen1 (RCS) cooldown of approximately 3'F at a - rate of approximately 10'F/hr. Af ter a - stabilization period at"this-lower temperature,.an-RCS heatup is then initiated for an approximate- 3'F increase, also a a' rate-of-approximately 10'F/hr. A, plot of reactivity vs. temperature is ! made for both the cooldown and heatup-portions of the test. The j cooldown and heatup are performed at the All Rods'out (Control Bank j D > 200 steps). control rod configuration. Multiple 1cooldown-and ! heatup cycles may be performed,'if required for data consistency. l
SUMMARY
OF RESQLTS, The cooldown resulted in an ITC of.-0.89 pcm/*F. The .heatup - { resulted in an ITC of -1.10pcm/'F. Only'one cooldown and heutup cycle was performed. The average :ITC was -0.995 pcm/*F. The ilt l ( pcm/'F criterion between .the ccoldown and- heatup values was i satisfied as they differed- by only 0.21 pcm/'F. The'13 pcm/'F. criterion between average measured ITC and the prediction of -1.4 pcm/'F was satisfied as they differed -by? only 0.405 pcm/'F. The calculated MTC value of -0.995 -(-1.83) = +0.835 pcm/*F did not satisfy the S > 0 pcm criterion so Rod Withdrawal Limits were calculated and impose d /'Fusing NUC-116. The calculated MTC value was l within the measurement tolerance of the design value and the- l positive value was not-unexpected. Control Bank D was at >200 steps.during this test performance with- { all other control rods fully withdrawn. Core neutron -flux was maintained below the point'of nuclear heat addition _to preclude nuclear heating feedback effects from' invalidating test results. Refer to Figure 3.3.7-1 for the plot of reactivity vs. ter erature'. l
-112-i e
y _
- s f , ,
t K 4 L
<,1 oroe.v 1 Reactivi,ty vs.< 2amparat:ure q .. . . .
7.~.__. f j' ...- .l_ . g J.t, J.... . .i. . :)
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j . l i j , I 'i t l
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u . l (h 5 t . , . ga8 .o b Oqh
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j j' Tcy% ,4 .3 'F1 Reactivity g
. 1 a rai y ,.
pe= er
. _ h_
l( i g ; Ny.U..
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l 'lI
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3 i i
-113-i 1 i a
l. l l i ! 3.3.8 - DETERMINATION OF OPERATING LIMITS TO' ENSURE'A-NEGATIVE MTC.
- NUC-116 OBJECTIVE -l 4
This permanent plant proceduto-is performed to establish operating: ! limits, also . called- Rod Withdrawal Limits, to ensure that the 1 Moderator Temperature Coefficient (MTC) remains negative. It:is : only performed if the All. Rods Out MTC value measured in NUC-207 is 1 found to be positive.' This procedure satisfiest activities I described by FSAR Table'14.2-3, Sheet 14. l a TEST METHODOLOGY .j Boron,-a neutron absorber, is dissolved 'in the RCS water, which' ) also serves as the reactor moderator. As moderator temperature-increases, the moderator becomes less dense and is less efficient-at_ slowing.down fission neutrons.: The effect is to add negative reactivity and cause the reactor neutron flux to decrease. But, as. the moderator density decreases so does the density of the soluble. .l boron. - This boron density-decrease reduces the number of parasitic > neutron captures by the boron and -effectively adds positive reactivity- which causes the reactor : neutron flux . to increase.- i These conflicting effects can either cancel each other out or the' balance will shift one way or:the other.. As boron concentration increases, the net effect 'is a positive reactivity addition with an increase in moderator temperature. For reactor stability' considerations,- the MTC is restricted to non-positive values .only, such that .an- increase in moderator temperature results in addition of negative reactivity which tends to shut the reactor down. 'If the measured MTC is positive, measures must be taken to limit it to negative values. This 2n done-by limiting the maximum moderator boron concentration to va'. 1 which ensure a. negative,MTC. This can'be achieved _by
- operat. a control rods partially inserted (i.eE ' limiting their ,
withdrawal) such that the boron concentration -is reduced' to maintain criticality. This procedure generates.a-family of curves which relate permitted Control Bank D. positions as functions of l reactor power level. and boron concentration. The - curves are l j generated using reactor core designer supplied methodology and are
- based on the actual misasured MTC, actual All Rods Out endpoint boron concentration-and design predictions of control bank worth, i boron concentrations and MTCs.
]
i l -
-114-l
3.3.8 - DETERMINATION OF OPERATING LIMITS ~TO ENSURE A NEGATIVE-MTC
- NUC-1 H (Continued)
SUMMAR'/ OF RESULTS The MTC in NUC-207 wa s - + - 0. 8 3 5 pcm/'F so performance of this procedure was required.--Rod Withdrawal-Limit curves were properly generated and implemented, refer to - Figure - 3. 3. 9-1. Note. that-D- is allowed to be withdrawn further as power is'
-t control Bank increased. This is due to the normal- reduction of the boron.
concentration to compensate for power defect as power is increased. l l I I
-115-I
Figure 3.3.8-1 Rod Withdrawal Limits Control Bank D Position (steps) 0% .10% 2d % -2 0%40% 6 3% Power
<, <, <, i, i, 4, see see see -
W****eptense 4
\
=
"" \ \ (
\ \ \ \ \ \
x xx x x x a y w g y x <s xx <x xx xs m s s Nx <i sx-3 v v x-w~m
~ ~ ;
. . . -e x
x x x x e
\ w\ N_
xx\ V \_ woe wu so e sees wee - noe ime sue see ._noe == inn noe me.
- Boron Concentration '(ppm) 5
-116- ,
-- _..-------.-_____-~=_:* .
. _ - - _ _ - _ _ _ _ - _ - - _ _ = - = - -_ -
3.3.9 - ROD SWAP MEASUREMENT 'NUC-1201 ; OBJECTIVE f This permanent plant procedure isf performed' to verify that ' the
- ' differential and, integral worth of individual control rod banks- !
- agree with the_ design predictions made in the core design report..
l This procedure also measures differential boron worth. This , procedure partially' satisfies activities described by-FSAR Table! , 14.2-3, Sheets 15 and:16. TEST METHODOLOGY , The rod swap method, al'so - called the bank exchange method, of . determining bank worth only directly measures one bank, this' bank is designated the reference bank,. and infers the worth of the other banks based on the reference - bank worth.- The reactor core designers selected Shutdown Bank B_as'the-reference bank for this-initial fuel cycle. Starting with the' reactor. critical and control Bank D partially inserted to adjust core neutron flux, Control Bank D is fully withdrawn. The reference bank,. Shutdown Bank B, is then immediately inserted to restore; the. core to the just critical condition. An endpoint measurement -is made for the currently inserted worth of the reference bank- by fully withdrawing it, measuring the core reactivity change with the reactivity computer, 1 4 and reinserting the bank to the just' critical- position.~ Next, an i RCS dilution at a rate of approximately 25 gpm is started. This is; ; ( roughly equivalent to-a 300 'pcm/ hour reactivityEaddition rate. This positive reactivity addition 'is . compensated by periodic insertions of the reference ~ bank in -10 to -20 pcm-increments in order to maintain the just critical ; reactivity : condition. The . individual worth of these incremental insertions is' measured with l the reactivity computer. The dilution is . suspended with the reference bank near fully inserted.-.?Another endpoint.neasurement'- is made for the last portion of reference bank worth-by inserting . the reference bank to core bottom ad measuring this worth with the reactivity comput6r. The sum of the incremental-wortha and the two-endpoint worths.is the integral. worth:of the reference bank. 'The differential worth is calculated las the incremental: worth' divided by the number of rod steps moved . to' result in -- that reactivity _ change. Both of these worth values.are recorded.and plotted. With the reference bank nearly - fully cinserted and all other rod banks fully withdrawn, the integral worth.of the other rod banks are individually verified by comparing their relative worths with l respect to that of the just. measured reference bank.~- A selected I test bank is inserted to result in approximately 20 pcm of negative core reactivity. The reference bank,is then immediately withdrawn to result in . approximately 20 pcm of positive - core ' reactivity. This process is repeated until the' test bank is- fully inserted and the reference bank-is-adjusted to a just' critical position. The
-117-(
1 3.3.9 - ROD = SWAP ~ MEASUREMENT - NUC-120'(Continued)' TEST METHODOLOGY _(Continued). worth of_ tha t test bank is thenHinferredL as beingr equal to the fractional portion of the reference. bank worth;that was withdrawn to compensate for the. test--bank's1 insertion. ..The test and reference banks are returned to their. initial positions,: reference bank in and the test bank fully withdrawn, in the reverse-order ofL steps used for the measurement. The~same process is. repeated for each remaining bank until all' banks have been , exchanged, or. swapped, against the reference bank. Following the exchange.of all banks against the reference bank,. core conditions are restored in one of two ways. If Rod Withdrawal Limits a'r e not to be imposed as-a result' of the MTC measurement, an RCS boration is: started and the reference bank? is withdrawn to compensate for this negative reactivity: addition until1it is fully-withdrawn. If Rod Withdrawal Limits are to - be imposed', : the ~ reference bank is exchanged against Control Banks D and C until' the reference bank is fully withdrawn. Differential boron worth is obtained by dividing th's total worth ~of the reference bank by the difference of . the two endpoint boron a concentrations to yield pcm/ ppm. [
SUMMARY
OF RESULTS The reference bank measured and predicted worth were to differ by I no more than 17%. -They differed by only -0.4%. The remaining test banks measurci and predicted worths were to. differ by_ no more than- i 10% or 100 pcm, whichever was greater. .Three banks, Shutdown - Banks A and C and Control Bank A, had differences of, greater than : 10%. However, the difference did not exceed =1100 pcm for any of l these banks. The other banks, which met ' the 110% criterion, ) automatically met the i100. pcm criterion because no ibank ihad a l predicted worth _of more than-1000 pcm. The percent error between measured and predicted l total worth was 1.8%. This satisfied the criterion that the measured' total worth be within i7% of the predicted total worth. The best individual bank agreement was found in the measurement: of. Shutdown : Bank B which was within -0.4% of its predicted value. The_ worst agreement j was found in the measurement of Control Bank A which was within
-13. 4 % - of its predicted value- . - A ' summary, of the! bank worth measurements and the predicted values appears in Table 3.3.9-1 and Figure 3.3.9-1 is a plot of the integral and differential reference ,
' bank worths. 1 (1
-118-l l'
l
._1_ _ = _ _ _ _
I
~'
3.3.9 - ROD SWAP MEASUREMENT - NUC-120-(Continued)
SUMMARY
OF RESULTS (Continued) Rod Withdrawal Limits had been imposed by NUC-116, 'so ShutdcWn; Bank -- B - was swapped against - Control Banks C and D , to rectore core conditions. No problems were encountered during. field performance of this test. However, two items- occurred pertaining to the calculations performed. The methodology used in the calculations of' inferred-bank worths differed from the original Westinghouse methodology. The Westinghouse method uses the average of the initial'and final inserted reference bank positions in the inferred worth calculation to account' . for. any not core reactivity drift due to outside influences over the course - of the rod swap measurement.' NUC-120 l used only the initial reference bank position in this calculation. l The inferred worths were recalculated using the Westinghouse method and documented'in accordance with plant procedures. The changes to the resulting values were so small as to have~no adverse impact onL , test results acceptability. These' recalculated' values were those 1 tabulated in Table 3.3.9-1. The other item pertains- to the differential boron . worth calculation. The-calculated value from core physics. measurements was -11.63 pcm/ ppm. The expected value was -10.44 pcm/ ppm 110%. However, this -10.44 pcm/ ppm value was not specifically calculated i for this core configuration by the reactor designer. It - was l calculated based on . predictions of the All - Rods Out (ARO) and i l Reference Bank In coron concentrations and the reference bank worth l contained in the core design. report. The Shutdown Bank B In boron concentration of 1062 ppm was cal-culated using the rod' swap model which is different;from the model used to generate the CPSES Unit 1, Cycle ~1 boron endpoints (ARO"
- critical boron concentration of 1146 ppm) .- The rod swap method H gave an ARO critical boron concentration-of.1139 ppm. Based upon j use of a consistent model, the predicted differential-boron worth would be
Rod Swap Model: Predicted Reference Bank In I l Boron Endpoint 1062 ppm Rod Swap Model Predicted ARO' Boron Endpoint 1139 -ppm Predicted Reference Bank Worth 877.1 pcm Revised Predicted 877.1 pcm Differential Worth = -------------------- = -11.39 pcm/ppe 1062 ppm - 1139 ppm Using this consistent model prediction, the revised predicted differential worth acceptance range would have been -11.39 pcm/ ppm l 110%. The measured value differs from this revised prediction by l only 2.1%. l
-119-1
n p i
. i.
TABM: 3.3'.9 ~ 5 1-' . ,.
~i Measured and Inferred Versus Predi h m jf MEASURED / I
= INFERRED PREDICTED- ABSOLUTE '
a' BANK WORTH (pcm) WORTRIpsal DIFFERENCEfpcm) % Differengg i
.i Shutdown A -591.6. 524.C' 6 6 '. 8 ' 12.7 -;
Shutdown B* 873.3' 877;1' -3.8' -0.4' 1 Shutdown C 468.4 425.2- !43.2 10.2 1 Shutdown D 461.3' 425 . 2 36'.1 8. 5- 'i Shutdown E. 459.1 '487.9 -28.8' - - 5 ~. 9 -
)
Control A 301.9 '348.5 -46.6! .-13.4-Control B 816.4 767.2 49.2! 6.4. ,l Control C 824.1- 853.1 -29.0- -3.4 q Control D 662.7 654.4 8.3~ M r Total 5458.8 5363.4 +95.4 +1.8- !
-i 1
- Shutdown Bank B was the reference' bank'and.has a measured worth. j All other banks have inferred worths.:
1 l
< t i
1 1
,T
-120- ,
u - 8
Figure 3.3.C-1 Differential and Integral Rod' Worth Rod Swap Reference Bank, Shutdown Bank B Integral Rod Worth (pcm) ' Differential Rod Worth (pcm/ step) 1000 10-900 9
~N *
800 8
)
- 700 ; 7:
.600 /\ L .x 6-500 / NL 'N; .
.c [\
.sg y .;
400 .. 4: j .~ k
~
~
_ ;, . 300. _
.3 _ . 3- --
- h '
~{K 200. .2
-100
, \ '. \ '
11
.O O O 20 40 60 80 10 0 . 120 140- 16 0 118 0 200 220 240 Steps Withdrawn 4
121-us3 y - ,n tiyrg.i,-- 1 .- ,-b , we c~wer--- -+w *r'- e F w *~-.o u, *Nh-*w
- _ _ _ _- m. gr+ew-? m 2._W'"u- m gp-.mc__ 2nV __.a___J__ _ . -._mz"-. .a
1 3'.3.10 - BORON ENDPOINT DETERMINATION AND' DIFFERENTIAL BOPON WORTH
- NUC-104 I OBJECTIVE This pearmanent plant procedure-is performed'to determine rod' worth at the. extreme ends of rod bank travel, at the _ near. fully withdrawn- .
or near fully inserted positions. : In addition, the just' critical, All- Rods .Out - ( ARO) , . . Reactor . Coolant . System' (RCS) - boron . concentration is determined. This procedure! partially _ satisfies activities described by FSAR Table 14.2-3, Sheet 16. TEST METHODOLOGY -l The test starts with RCS temperaturesi andf boron concentration-verified stable and all control rods. withdrawn except for Control Bank D, which is controlling flux attthatjust critical 1 condition.- With no more than approximately, 50 pcm of' Control' Bank D worth inserted, Control Bank DLis: then fully withdrawn to reach the - desired ARO endpoint configuration while, neutron flux, reactivity,: RCS temperature and pressurizer' level-are monitored on strip chart recorders. When the reactivity trace stabilizes, Control Bank D~ is' repositioned' to re-establish Lthe ~ initial! flux , level' and core reactivity. This process is repeated at least two more times. = ' The-endpoint boron concentration:is obtained by. dividing the measured reactivity change.due to Control Bank D withdrawal by the design prediction for differential boron worth at this particular rod bank configuration.- This converts,the measured!pcm of reactivity worth' to ppm of equivalent - boron worth. This : boron -worth : value is combined with the actual measured boron concentration to yield the boron concentration that would exist' with j Control: Bank D ' fully withdrawn and the reactor just critical. . This is ' a' ' calculated method which replaces the alternative of' 'actually ~ adding small quantities of boron to the RCS untiloall of the: control rods'are fully withdrawn with the core at the just critical; condition. Differential boron worth lover a particular , bank is obtained by dividing the total integral worth.of the. selected rodibank by the difference in the endpoint boron concentrations, on'e for the. bank fully withdrawn and one-.for the bank fully inserted. This results in a value of pcm/ ppm and is.always negative.
SUMMARY
OF RESULTS The All nods Out just critical RCS boron concentration'was found.to be 1162.1 ppm which satisfied the acceptance criterion of 1146 ISO ppm. Differential boron worth over the reference bank was calculated in- ; NUC-120 using.the above technique.
-122- ,
_ :_ z _ _ _ _ _ _ _ . _ _ _ _ . - - - _ _ _ _ _ _ _ - _ _ - - - . . .- - - -
- <, ~
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- )
3.4 -' TRANSIENT TESTING
._3.4.1 - TURBINE GENERATOR TR POWER - ISU-222A ]
[. OBJECTIVE L This test is performed to - verify the~ plant's L ability . to safely sustain a turbine-generator trip with.no-offsite power available. for at least thirty minutes. This test satisfies L activities I described by FSAR Table 14.2-3, Sheet 18. j TEST METHODOLOGY The test is initiated with. Unit 1 in Mode 1, at greater than 10% reactor power, with the main generator output at approximately 130
~
MWe. All normal 6.9 kV electrical buses are initially energized by . the Unit Auxiliary Transformer (1UT). Their alternate . source,. j Startup Transformer 1ST, is- locked out, . preventing ' a , designed automatic bus' transfer to this backup supply. Class 1E Safeguards-Busses 1EA2'and 1EA1 are initially aligned to Startup Transformerf , XST2, with theP; alternate source, Startup i Transformer - XST1, ! locked-out. The turbine is manually tripped, the; feeder breaker to non-class 1E bus XA1 is opened, and the , offsite power . feeder. breakers to the Class 1E-busses are opened. This resultssin an immediate loss of Class 1E . AC power and a -loss of Unit 1' non-Class-
~
1E AC power when the main generator tripsr approximately- 11.5 seconds after the turbine trip. The 11.5 seconds is;due to normal protective relaying time delays. Plant conditions are monitored to ensure that the~ standby emergency diesel generators start and re-energize the safeguards buses and ' that plant equipment functions properly =to-stabilize the. Reactor ^ Coolant System: (RCS) in a Mode 3, hot " standby condition and l maintain it in that condition for. at least , 30 minutes.- 'Also ! monitored is the ability of the Steam Generator Atmospheric Relief Valves'to control steam line pressures below 1185 psig for at least 30 minutes.
SUMMARY
OF RESULTS The alignment of Unit 1 power supply breakers;was c'ompleted,.the , turbine was manually . tripped and - the appropriate Lfeeder breakers- > l were opened. Both standby emergency. diesel generators started'and powered the safeguards buses. The safeguards sequencers both ! loaded the required plant equipment onto'the' safeguards buses at' . the proper times. Stabilization of and recovery from this event = was performed in. accordance with the permanent = plant emergency operating and abnormal operating procedures. l l I
-123-1 I
l 2 t 3.4.1 - TURBINE GENERATOR TRIP WITH COINCIDENT IDSS' OF OFFSITM-POWER - ISU-222A (Continued): 4
SUMMARY
OF RESULTS (Continued): The' safeguards buses were energized and plant equipment functioned properly- to stabilize -and maintain theTRCS "in a safe- shutdown - condition. This condition was maintained for 31 minutes satisfying _ the .>30 minute criterion. The non-Class.'lE buses remained de energized > for the duration of this test. RCS hot leg, _ cold leg and core exit thermocouple readings 1 were _ verified- to stabilize following the. initiation of the transient indicating that natural circulation cooling was established.- RCS subcooling was verified. to be greater than 60*F. Even though the Main Steam. Isolation. Valves.were closed early in the event, in-accordance with the permanent plant procedure used.to l stabilize the plant, main steam line pressures never. exceeded that 11251 psig setpoint where the' Steam Generator Atmospheric; Relief __ .I Valves begin to open. Therefore,. the: limit of 1185 psig !for l maximum main ' steam..line_ pressure was never exceeded. The atmospheric relief valves were opened manually to initiate natural' circulation cooling as prescribed by the permanent plant procedure used for plant recovery. The following is a summary of indicated plant conditions during the. event: Item Maximum Value Minimum Value-RCS Cold Leg Temp. 55 8 . 5'F 55 0. 7'F RCS Hot ~ Leg Temp. . 5 69. 3*F _ 561.1*F Pressurizer Pressure 2296 psig 2202.2'psig Auctioneered High 573F 571 'F Core Exit Temp. Main Steamline Press. 1119.6.psig .1036.7 psig a During the transient, only one unexpected,-events occurred. Pressurizer Power Operated Relief Valvet (PORV) 1-PCV-455ALopened ' for less'than 5 seconds. While' actual pressurizer pressure never reached the-normal 2335 psia PORV setpoint, this valve actuated in i response to a compensated pressure: signal which includes - ' an - integral signal component. The pressurizer pressure was above the 2235 psig nominal-pressure setpoint~for a'long enough time: period. to allow the- integrated ~ signal' to grow' large enough to open- the- ! PORV. This PORV actuation did not ; disrupt the RCs cooling.by i natural circulation. A number of the Test Data Acquisition System channels originally called to be monitored were unavailable for use during'_ test - performance. None of: the unavailable channels were required' to verify test acceptance criteria.
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1 l l
3.4.2 - DESIGN LOAD SWING TESTS - ISU-231A' OBJECTIVE This test is performed to demonstrate:the dynamic response of the Reactor Coolant = System (RCS)- and the Rod Control.- System- to - automatically-bring the plant to steady state conditions following a rapid -10% reduction in turbine ' load, and then to a - rapid 10% increase in turbine load. This test partially satisfies activities . described by FSAR Table 14.2-3, Sheets 23 and 24. l TEST METHODOLOGY With pl . .c conditions stable at approximately 35%, 50% or 100% power, a.10% load decrease is manually initiated from the turbine-generator Electro-Hydraulic Controls (EHC) at a rate, of approximately 200% power / minute. Plant parameters are allowed to stabilize, and after stabilization, a 10% load increase is manually 1 initiated. Plant parameters are again allowed to stabilize. The load decrease is performed by manually reducing: the turbine-generator load limit setpoint to a value approximately 10% in power below the initial load reference operating. power level. .The load. increase is performed by manually raising the load limit setpoint' back above the original load reference operating power level. This allows the load to increase back to its original value at the start of~the-test. The load limit setpoint adjustment; occurs at a rate of approximately 200% power / minute and is performed _by main control board manual push button operation of. a motor. driven ' potentiometer that is set to move at that rate. These push. buttons are permanent plant control features and the related circuitry- is closely associated with the built-in turbine-generator runback. circuits. In fact,-prior to initiation of the 10% load increase, the runback circuits are temporarily bypassed by actuating at switch inside the. EHC cabinets. This is done to. disable load increase. inhibiting , circuitry that is activated by the 10% load decrease via the-shared runback circuit portions. The 10%. power load' changes:are nominal values and are actually specified:to be 10% i2%-in magnitude. The 10% load change may result in reactor power changes of-greater than 10% power due to relatively low plant efficiency at lower power levels. During the course of the test, strip chart recordings and Test Data. 4 Acquisition System (TDAS) recordings of-key plant parameters are
~
taken so that plant response can be analyzed.a .The principal parameters monitored included RCS.Tavg, Tcold, Tref, pressurizer 7 pressure and level, steam generator pressures and levels, steam and feedwater flows,: control rod positions and speed, 0TN16 and OPN16 -i setpoints, reactor power, feedwater pump . speed and discharge pressure,.N16 power, safety and relief valve positions, and steam dump valve positions. ; I
-125-l 4
l'
- 3. 4. 2 - DESIGN _ LOAD SWING TESTS - ISU-231A (Continued)
(
SUMMARY
OF REuULTS
-l The first test: performed was at the 50% power plateau from approximately 47.5% reactor power. The second-test: execution was l
from approximately'34% reactor power and the final test execution i was from approximately- 100%. reactor _. power. -All three test executions satisfied the following criteria: o The load decreases and-increases did'not cause the reactor to trip nor the turbine to trip. ; o Safety injection-did not initiate. ., l o The steam generator safety or atmospheric relief valves and . pressurizer safety or power oparated relief valves did not l lift during any of the load swings, o Nuclear power over/undershoot was-less than.3%.- o No manual intervention was. required'to bring plant cotiditions-to steady state. I o Plant variables returned to steady state; conditions-without sustained or diverging oscillations.
~
The first test performance, from approximately 50%' power,.resulted in two ratests. The magnitude of the first load decrease was'only 7% power, which did'not adequatelyLapproximate a 10% load change. j Plant initial conditions were restored.and Retest #1 was performed-to repeat the load decrease. The L Retest -' #1 load decrease was approximately 12% power 'in magnitude c and equilibrium Tavg ' was reached approximately 17 minutes following initiation of the load _ ! decrease. The subsequent _ load increase _was blocked lby the turbine-generator control circuitry after an increase of only 20'MWe. The- - load was-reduced by this 20 MWe, the plant allowed-to stabilize, the runback circuits were bypassed and the load' increase-repeated, as Retest #2. The Retest #2 load increase was approximately 11.8% power in magnitude and equilibrium- Tavg -was reached: in approximately 12 minutes. The. interference- of - the runback circuitry with respect to the load increase.was unknown at the time. the test was written. The procedure was-. changed to, bypass the runback circuits by actuation of a permanently installed switch inside the EHC cabinets. This procedure. change was made prior to the performance of Retest #2. The second test' performance, from approximately 35% power, did not result in any retesting. The load decrease was approximately 9.5% power in magnitude and equilibrium' Tavg was reached in approximately 15 minutes. The load increase was also approximately 9.5% power in magnitude and equilibrium Tavg was also reached in approximately 15 minutes.
-126-
- - - - - . - - _ . . .- . - . - ~ .
I 3.4.2 - DESIGN LOAD SWING TESTS - ISU-231A'(Continued) i
SUMMARY
OF RESULTS (Continued)' j l: The final test performance, from approximately 100% power, also did' not-result in any ratesting. The load decrease was approximately- . 9% power in magnitude and equilibrium Tavg Lwas- reached in ! approximately 8' minutes. The load-increase was also approximately 9% power in magnitude and equilibrium Tavg was reached in approximately - 9 minutes. Ten' minutes of Axialo Flux Difference , , (AFD) penalty time was accrued following the load decrease due to l AFD deviating outside of its target band. . This short deviation was ( an expected occurrence, caused-by the deep. insertion.of. Control: .r Bank D in response to the load' decrease. Normally,. power changes f are made using RCS boron concentration changes, leaving Control- , Bank D nearly fully withdrawn. The AFD is permitted - by the-Technical Specifications to be outside- of11ts~ target band for_ up to-one hour without subsequent action, when below 90% reactor power. Numerical acceptance-and review criteria are summarized on Table 3 3 3.4.2-1. Test performances were generally)without incident with : the following items noted: ! During the tests performed at approximately 35%.and 50% power, the ? first bank of. three steam dump, or turbine bypass,- valves-momentarily opened as a result of.the rapid load-decrease. The . steam dumps are armed by a 10% or greater' decrease in turbine e impulse chamber pressure. The turbine impulse chamber pressure is proportional to turbine-generator power. The 50% power test:had a load decrease of approximately 12% which allowed _ the steam dumps to l arm.. While the 35% power test had a: load decrease .of only e.pproximately 9.5%, the initial impulse pressure change was greater than 10% due to a .. minor overshoot ._ in turbine governor _ valve positions in response to the rapid load decrease. This also allowed.the steam dumps to arm. .The minor overshoot is.a normal i result of a rapid load-change and damps out very quickly. Once armed, the steam . dump ' valves modulate to force _ average RCS-temperature (Tavg) to within 5'F of the load varying reference RCS l average' temperature (Tref). In both test performances, Tavg was l: only marginally more than 5'F greater than Tref. This is why.only_ l three of the twelve steam dump valves modulated.open for a short time, loss.then one minute. These steam dump, actuation'did not invalidate the test results because the rod' control' system properly responded to and stabilized the load decrease transient. The steam-dumps. aided the rod control system, as they are designed to do, during the_ initial portion' of :the transient. These test = s performances both had reactor power changes in-excess of110%,-11% ; at 35% power and 14% at 50% power. The rod _ control system is designed to respond to absorb a nominal '10% power change.1 The assistance _ of ~ the steam ' dumps absorb the excess - over 10%' was proper. ; l -127 : ('
,= ~~ . + . . . . _
c . . , . - , m,.. ,-.
1 3.442 - DESIGN LOAD SWING TESTS - ISU-231Af(Continued) i L
SUMMARY
OF RESULTS (Continued)- . I I 2 During the'first test' performance, at'approximately 50% power, the , recorded values ' of initial, finali andT transient: steam header. l pressures resulted in an indicated excessive undershoot of 82 psig. . I l Investigation of'the data disclosed.that because the initial-and final values were recorded from a main control board indication and the transient values from temperature data logging: instrumentation,- the slight bias between the two' data'souroes was'enough to induce I excessive error into-the results.- When the results from a single- ) data source, a strip chart recorder, were evaluated, they showed i the true undershoot to be only 65 psig,~ which met the <70 - psig . , criterion. The discrepancy between the : two data sources. of -I approximately 10 psig was minimal, representing'only 0.774 of the l 1300 psig. span of the- instrument iloop,- well within normal : tolerances. I During the final test performance,.at approxic tely'100% power, it; was noted that more time could have been allowed between the load decrease _and the subsequent ~ increase to ' achieve greater stability : ! However, the actual time was-judged to have been acceptable based I on examination of the convergence,of monitored parameters. Refer to Tables 3.4.2-2 through . 3.4.2-7)for1 additional- detailed - data, i l l l l
-128-V *
, . ----- - . _ _ _ _ rs
o 1 TABLE 3'.4.2 - DESIGN IDAD SWING- TESTS
SUMMARY
Nuclear
' Power- Pressurizer- ~
Power Load Over/Under- Allowed ' Pressure- -Allowed Plateau (%) Swina shoot (4) ' Limit (4) swinatomia) -Limit (esia)
'35 Decrease 2 <3 +0,-5 <i100 Increase 1.5 <3 +0,-5 <i100. q 50 Decrease- 2 -<3 +8,-27 <i100 -!
Increase 2 <3 i+10,-35' - <i10 0 - l 100 Decrease O- <3 +12,-3dL <i100 Increase O <3 40,-20 <i100: ) i I
.l
'J Stieam Header--
Steam Gen. Pressure Over/ , Power Load- Level Allowed. Undershoot Allowed- l Plateau (%) Swina Swina(%) Limit (%)- (nsia) Limit (nsia)- 35 Decrease +4,-3 $ i10 - 39 $70' ' Increase +6,-2 $ 110 '52 - $70 50 Decrease +3,-1 5 i10 21 $7 0 - ,
. Increase +7,-3 $ 110 -65 $70 >
200 Decrease - +7,-3 $ i10- 42' $70 Increase +4,-1 $ 110 - 13l. '570-Power Load Tavg Over/ Allowed' Plateau (%) Swina Undershoot ('F)
~
Limit ('F) 35 Decrease < 1.0 $2.0 Increase < 1.0 $2.0 50- Decrease O '- '$2.0 Increase 2.0 $2.0 100 Decrease 2.0 52.0 Increase 0 $2.0-l l l
--129 .
s
- TABLE ' 3 . 4 ~. 2-2 ' 4 10% LOAD DECREASE AT 35% POWER-
SUMMARY
INITIAL CONDITION FINAL CONDITION' Generator Load (MWe)= 315< 205 Nuclear Power (%) L34: 23= Tavg Auctioneered' ('F) 567_; 562-Tref ('F) - .566- 562.5-N-16 Power (%) 354 25 I OPN16 Setpoint (%) '112: 112- ' OTN16-Setpoint (%) -1141 116'
- i Pressurizer Pressure (psig) 2235c 2235 i
Pressurizer Level (%) '36 31 Steam Generator Level Loop l'(%) 65 =65- .f Steam Generator Level Loop 2 (%)- 66J 66 -! Steam Generator Level Loop'3 (%) 65 65 q Steam Generator Level Loop 4 (%) 65- 65- 3 Steam Header Pressure (psig) =1058: 1045 ] Steam Flow Loop 1 -(pound's/ hour) 1.0E6- 0 .~ 7 E 6 Steam Flow Loop 2"(pounds / hour). l'.0E6: 0.7E6- ; Steam Flow Loop 3 (pounds / hour) 1.0E6; '0.75E6; _ Steam Flow Loop 4 (pounds / hour) O.9E6 0.6E6 - i Feedwater Flow Loop 1 (pounds / hour) 1.1E6: 0.75E6 i Feedwater-Flow' Loop 2 (pounds / hour) 1.1E6 0.8E6-- ] Feedwater Flow Loop'3 (pounds / hour) l'.1E6' O.8E6 i Feedwater Flow Loop 4 (pounds / hour) 1.1E6 .0.8E6~ i Feedwater Temperature Loop 1 '('F) 350' 330-Feedwater Temperature Loop 2 (*F) 350- ^330- ! Feedwater Temperature Loop 3 ('F) 350- 330 j Feedwater Temperature Loop 4 ('F) 350 330 J l Feed Pump Discharge Pressure (psig) 1175- 1162 i j Control Bank.D Position (steps) 166- 136.5 ! Control Bank C Position (steps) 227! 227 l . 4 Feedwater Pump 1-A Speed (rpm) 4090. 3900 ,
.]
-130-1 ,
L , t l TABLE 3'.4.2-3' i 10%. LOAD INCREASE AT-35%' POWER.
SUMMARY
INITIAL CONDITION FINAL CQNDITION Generator Load (MWe)' 205 .315 Nuclear Power (%) 23 34.5. Tavg Auctioneered ('F)' .562 565 Tref ('F) 562.5 566, N-16 Power-(%) 25 33.5 OPN16 Setpoint (%) 112 112-t OTN16 Setpoint (%) 116- -116-
~
Pressurizer Pressure (psig) '2235 2230l Pressurizer Level (%) . 31 34' Steaia Generator Level, Loop 1 (%) 65' i 65 , Steam Gencrator.LevelELoop:2:(%) '66 ~66~ Steam Generator Level Loop.3 (%): '65- 65 Steam Generator Level Loop 4 (%)- 65 ,66 Steam Header Pressure.(psig) 1045 -1058 Steam Flow Loop 1 (pounds / hour) 0.7E6 1.1E6' j' Steam Flow Loop 2 - (pounds / hour) 0.7E6 11'1E6 Steam Flow Loop. 3 (pounds / hour). ;0.75E6- 1.05E6 Steam Flow Loop 4 (pounds / hour) 10.6E6 .0.9E6 Fe6dwater Flow Loop 1 (pounds / hour) 0.75E6- '1.1E6 1 Feedwater Flow Loop 2 (pounds / hour) :0~8E6
. 1.15E6- :
Feedwater Flow. Loop 3 .(pounds / hour) O'8E6
. 1.15E6 l Feedwater Flow Loop 4 -(pounds / hour) 0.8E6 '1.15E6' Feedwater Temperature-Loop 1 ('F) 330 3501 L Feedwater Temperature Loop.2 ('F) 330 350:
l Feedwater-Temperature Loop 3 ('F) 330 350 Feedwater Temperature Loop 4 ('F) 330 350 q Feed Pump Discharge Pressure (psig) 1162 1175 Control Bank D Position'(steps)- 136.5 173.5; Control? Bank C Position-(steps)' 227' 227 l Feodwater. Pump 1-A Speed (rpm) 3900 .4030
-131-l 1
l
.- .. .- . .. . , . ~
'l TABLE ' 3'. 4 . 2 -4 ; l l
10% LOAD DECREASE:AT'50%5 POWER:
SUMMARY
' ( 1 INITIAL CONDITION FINAL.JO.NDITION l
' Generator Load (NWe)' 455 315 Nuclear Powet' (%) 47.5 33.5 Tavg Auctioneered ~('F) 569 565~
Tref ('F) 570 566 N-16 Power (%)- 50- 39 l 1 OPN16 Setpoint (%) 112 112-OTN16 Setroint (%) 116 118 Pressurizar Pressure (psig) 2232 1 2240: Pressuriter Level (%) 39 34-66 66 Steam Generator Level Loop 1-(%) Steam Generator Level Loop 2 (%) '66- 67 Steam G>3nerator Level Loop 3 (%) 65 65 Steam Generator Level Loop 4-(%) 65 65' Steam Header Pressure -(psig)' 1035 -1060 ~ Steam Flow Loop 1_ (pounds / hour) ,1.40E6- 1.05E6 Stea'n Flow Loop 2 (pounds / hour) :1.50E6- 1.15E6 Steam Flow Loop 3 (pounds / hour) 1.45E6: 1.15E6 Stenta Flow Loop 4 (pounds / hour) 1.~35E6 1.05E6-Feedwater Flow Loop 1 (pounds / hour). 1.55E6: -1.20E6 l Feedwater Flow' Loop 2 -(pounds / hour) 1.55E6 1;25E6 a Feedwater Flow' Loop 3 (pounds / hour) 1.60E6 -1.25E6 , FeedwaterLFlow Loop 4 (pounds / hour). 1.'65E6 1.30E6 Feedwater Temperature Loop 1- ('F) 372 345-Feedwater Temperature Loop;2 ('F) '374' L348 Feedwater Temperature Loop 3 ('F) 372 345 Feedwater Temperature Loop 4 ('F)' 372 346 . Feed Pump Discharge Pressure (psig)
- 1160 1170 1
Control' Bank D Posi. tion (steps) .182 ,144 l Control Bank.C Pos.. tion (steps) 227- 227 Feodwater. Pump 1-A Speed - (rpm) 4175 14170
-132-L
. _ . - . _ . _ , .I
f TABLE ~3.4.2-5 10%. LOAD INCREASE AT'5'0%-POWER
SUMMARY
INITIAL CONDITION FINAL CONDITION { L < Generator-Load (MWe) 315 452 l
?
Nuclear Power (%) 36.51 48 I Tavg Auctioneered- (*F) 568 571 , Tref ('F) 5671 571-. N-16 Power.(%) ;38 49- .[ OPN16 Setpoint (%) . 112 <112 OTN16 Setpoint (%) 114 114 l Pressurizer Pressure (psig) -2230. 2235 E Pressurizer Level (%) 391 43 q Steam Generator Level Loop 15(%) '66 66 Steam Generator Level Loop 2.(%) .- 6 7 67' Steam Generator Level Loop 3-~(%) 65 -65 i Steam Generator. Level Loop 4' (%):
. 66- 66 l
L Steam Header' Pressure (psig) -10601 1055 Steam Flow Loop 1 (pounds / hour) 1.07E6 1.40E6= J Steam Flow Loop'2 (pounds / hour)- 1.12E6 l'.50E6-Steam Flow Loop 3-(pounds / hour). :1.10E6- .1.45E6 q Steam Flow Loop-4 (pounds / hour) 1.00E6 '1.35E6 1
-l Feedwater- Flow Loop .'. -(pounds / hour)- 1.15E6. 1~.50E6-Feedwater Flow Loop 2 - (pounds / hour). 1.25E6 11.70E6 Feedwater Flow Loop 3 (pounds / hour) 1.20E6 1.65E6-Feedwater Flow: Loop 4 (pounds / hour) 1.2SE6- 1.60E6 l Feedwater Temperature Loop 1 -('F) 350 .370 L Feedwater Temperature Loop 2 ('F): 350 374.
l Feedwater-Temperature' Loop 3 ('F) 350 372 L Feedwater Temperature Loop 4 ('F) .350 372 l-L Feed Pump Discharge: Pressure (psig)- 1175L '1180' j Control Bank D Position (steps) ~139' 173.5 Control Bank C Position;(steps) 227 227 l l l Feedwater Pump'l-A' Speed (rpm) 4200 4240
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1 l
1 i
;i f
' TABLE 3~.4.2-6 , ';
i 10% IDAD DECREASE AT 1004 POWER-~
SUMMARY
- 4 INITIAL CONDITION FINAL CONDITION i
Generator Load (MWe) ' 1135 .1030- ,l
'1 Nuclear Power.(%) 100 -88:
l Tavg Auctioneered (*F) 15 88 586~ ' Traf ('F) 588 586, N-16 Power (%) 105 93; OPN16 Satpoint (%)_ 111- .112 , OTN16 Setpoint (%)- 110 112L e -l l Pressurizer Pressure (psig) 2250: 2240-Pressurizer Level.(4) 63' 58' Steam Generator Level- Loop 1 (%) G6 ? 66 Steam Generator Level Loop 2.(%) 66 66 ! Steam Generator Level Loop 3_.(%) 62 .63 - Steam Generator Level Loop 4 (%) 65 L66 Steam Header Pressure:(psig) :9861 :982 Steam Flow Loop-1 (pounds / hour). 3.7E6 13'.4E6-Steam Flow Loop 2 (pounds / hour) 3.8E6 3.4E6 -l, Steam Flow Loop 3 (pounds / hour) 3.6E6L 3.3E6i Steam Flow Loop 4 .(pounds / hour) -3.75E6- 3'.35E6' j Feedwater Flow Loop-1,(pounds / hour) 3.8E6 3.4E6 > Feedwater Flow Loop-2 (pounds / hour) .3.8E6 -3.4E6 Feedwater Flow Loop 3 (pounds / hour) 3.7E6. 3.4E6 Feedwater Flow Loop 4 -(pounds / hour) 3.75E6: 3.35E6 Feedwater Temperature Loop 1 ('F) 440' 430-Feedwater Temperature ' Loop 2 ('F) 440 430 Feedwater Temperature Loop 3 ('F) 440 430-Feedwater Temperature Loop 4 ('F) 440- 430' Feed Pump Discharge Pressure =(psig) 1144 1134 Control Bank D Position (steps) 188.5 '153: Control Bank C Position (steps) 2253 225-Feedwater Pump 1-A Speed (rpm) -4803- 4658
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. _. .n . _- J
i a TABL2 3.4.2-7 1 10% IAAD INCREASE AT.-100%" POWER
SUMMARY
INITIAL CONDITION FINAL CONDTTION Generator Load (MWe) 1030 ' 1135- p ; Nuclear Power (%) 88 100 Tavg Auctioneered. (*F)- 586: - 588: t Tref ('F) 586r- ' 588 ; N-16 Power (4)- 93. 105 , OPN16 Setpoint (%) 112- ' 112: OTN16 Setpoint (%) 112 = 110. 4 Pressurizer Pressure-(psig)- .2240 - 2260-Pressurizer Level-(%) 58: 60- , Steam Generator Level Loop 1 (%)- 66 67 Steam Generator Level Loop 2: (%) .66 67 Steam Generator Level Loop 3-(%) 63 63 , Steam Generator Level Loop 4 (%) 66 :66 Steam Header Pressure (psig)- '
'1020 . 984 Steam Flow Loop 1 -(pounds / hour) 13.4E6_ .3.7E6 Steam Flowf Loop 2 (po "is/ hour) 3.4E6 :3.8E6 Steam Flow Loop-3 (poun s/ hour)' 3.3E6 3.6E6:
Steam Flow Loop 4 '(pounds / hour) 13435E6L
- 3.75E6. I
! Feedwater Flow Loop 1 (pounds / hour) 3.4E6- 3.8E6 o ! Feedwater-Flow Loop 2 (pounds / hour) -3.4E6 3.8E6. . Feedwater Flow Loop 3 (pounds / hour) :3.4E61 >
. 3.7E6-Feedwater Flow Loop 4-(pounds / hour) 3.35E6: . 3.75E6 Feedwater Temperature Loop.1 ("F) 430 440- ,
Feedwater Temperature Loop 2 - ('F) 430 : 440' l Feedwater Temperature Loop 3 ("F) 430 440 l Feedwater Temperature Loop 4 ("F) 430 435 l-1 Feed Pump Discharge Pressure.(psig) 1163 1140 Control Bank D Position (steps)' 153 187.5-Control Bank-C. Position (steps) 225 225 Feedwater Pump 1-A Speed (rpm) 4658: 4840 l
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i l b l
- i. ~
lt 3.4.3 - DYNAMIC RESPONSE TO A FULL LOAD REJECTION AND TURBINE' . TRIP - ISU-284A OBJECTIVE This test is performed to verify the ability'of the primary and' secondary plant and the plant automatic control systems to sustain a generator trip.from full power and to bring the plant _to stable, conditions following the transient. . The N-16 instrumentation ~ 1 response time is also determined. LThis' test satisfies activities-- l described.by FSAR Table 14.2-3, Sheets 23, 24 and 28. l TEST METHODOLOGY From a stable plant power offapproximatelyL100%, a generatorLtrip- ; is initiated by opening both 'of the main generator output breakers. : This - directly causes a turbine trip and a reactor trip. The- :! operators follow the permanent plant Emergency Operating Procedures. to bring the plant to stable- conditions.: The z data trendingiis. terminated when Tavg is stabilized at approximately 557*FL(no-load . Tavg). n I
SUMMARY
OF RESULTS The generator output breakers were opened at 0930 hours and the RCS-temperatures stabilized at 0932' hours. All of the following test' , acceptance criteria were met:- o The pressurizer and steam generator : safety; valves . did not lift. o Safety injection did not initiate. o All control'and shutdown rods c released . and - dropped ~ to the fully inserted position. o The plant was stabilized in Mode 3. - o The steam dump valves modulated closed in the proper sequence. a o Feedwater isolation occurred immediately following_the-plant trip. This was prior to reaching the' no-load average RCS. i temperature of 557'F, as expected. o Average RCS temperature stabilized at 558 F. This satisfied 0 the >5531 review criteria. The response time of _ the N-16 instrumentation was 2.0 seconds. This satisfied the time response requirement of $2~.17 seconds. Nuclear flux dropped to less-than-15% power in 1.6 seconds. _This satisfied the $2 second response requirement.
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. ~- -- --
)
i 1 3.4.3 - DYNAMIC RESPONSE TO'A FULL LOAD REJECTION AND TURBINE j TRIP - ISU-284A (Continued)
SUMMARY
OF RESULTS (Continued) Narrow range steam generator levels, remained 121.6%. This satisfled the review. criterion 'that levels may ' drop out of span (<0%) but should return to' span;(>0%). Pressurizer level remained -'224.92%.- This satisfied the minimum level of 220% review criterion. Pressurizar pressure remained 21977 psig. _ This satisfied the minimum pressure of 21950 psig-review criterion'. 1 Refer to Table 3.4.3-1 for detailed test results, r l i 1 s I i i i
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4
=!
TABLE 3.4.3-1
' TRIP PROM 100% POWER.
SUMMARY
INITIAL CONDITION FINAL CONDITION Generator Load (MWe) 1160? O Nuclear Power (%) -100 0 ! Tavg Auctioneered = ('F) 587- 557 f Tref (*F) '589 557 N-16 Power (%) -101' 2' OPN16 Setpoint (%) -112 112I P OTN16'Setpoint-(%)- ;108 .107-
. Pressurizer Pressure (psig), :22500 2265- i Pressurizer Level (%) 59 25 ,
Steam Generator Level' Loop 1 (%) .65 40 Steam Generator Level Loop 2 (%) 16 6' '39 Steam Generator Level Loop-3 (%) L63 36' Steam Generator Level Loop 4.(%) 65 35 Steam Header Pressure (psig)
'l 9611 1070- I
't Steam. Flow Loop 1 (pounds / hour) 3.9E6 'O.4E6 Steam Flow Loop 2 (pounds / hour) 13.9E6. 0.2E6 -1l Steam Flow Loop 3 (pounds / hour) 3.7E6 'O.3E6 1 Steam Flow Loop 4 (pounds / hour) 3.8E6; 0.0E6
.)
Feedwater Flow Loop 1 (pounds / hour) ,3.9E6i 0.25E6- 1 Feedwater Flow Loop 2 (pounds / hour) 3.9E6 .0.25E6 Feedwater Flow Loop 3 (pounds / hour) 3.8E6 0.3E6 Feedwater Flow Loop 4 (pounds / hour) 3.8E6 0.05E6 Feedwater Temperature Loop 1 '('F) 445 440 _ ! Feedwater Temperature Loop 2- ('F)- -445 440 Feedwater Temperature Loop 31(*F)' 445: 440 Feedwater Temperature Loop. 4. ('F) 445 .435 i Feed Pump Discharge Pressure-(psig) -1120 510 j I Control Bank D Position (steps) 201.5 0 l Control Bank C Position (steps) 225 O Feedwater Pump 1-A Speed (rpm)- 4870 2
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3.4.4 - REMOTE SHUTbOWN' CAPABILITY' TEST - ISU-223A OBJECTIVE This test verifies that the unit'can be taken from-approximately 20% reactor power to Hot Standby conditions 1from outside the control room with a minimum shift crew. The potential:to safely cool the unit to' cold shutdown conditions from outside the control i room is also- demonstrated. This . test. satisfies activities I described by FSAR Table 14.2-3, Sheets 25~and 26 and Regulatory Guida 1.68.2. The cool down to 350'F. and remote switch over to Resiaual Heat-Removal System > cooling was performed as part-of the preoperational' test program. TEST METHODOIDGY i From the condition of greater.than 10% generator load and less than s 25% reactor power, the reactor is manually . tripped locally from the reactor. trip breakers. Utilizing abnormal' operating procedure ABN- -t 905A, Loss of Control Room Habitability, the minimum shift crew f establishes control of the reactor plant and stabilizes itfin Mode ; 3 from the Remote Shu+ iown Panel .(RSP) .' From the Mode 3 stabilized' q condition, a controlled cool down of at. least 50'F is initiated' q c from the RSP to demonstrate ~ cool down capability. - Upon completion, control of the plant-is transferred back_to the Main ControlLRoom. J A standby operations crew remains in. tho ' Wain Control ; Room throughout the test to assume control, if-needed.
SUMMARY
OF RESULTS The reactor was locally tripped ~ at 0910 hrs' from 20% reactor power and 150 MWe generator load. The' minimum shift crew used ABN-905A to establish a stable, hot standby-(Mode 3)' condition.by 1025.-hrs, j The Reactor. Coolant System (RCS). was' stable at approximately 558'F : as indicated in the control room and 540'F as' indicated at the RSP. d The RSP temperature indications come' from strap-on RTDs -instead of RTDs in thermowells actually immersed the process _ fluid _ .and are less accurate. They do, however, indicate trendslin an acceptable- , manner. Following 40 minutes of data taking to ensure stableiRCS conditions. a controlled cooldown was started'in accordance with ABN-905A. This stable 40. minutes satisfied- the requirement to j maintain a stable, hot standby condition'for'at least 30-minutes. ~l The cool down was performed from 1107 hrs. to 1304 hrs. The cooldown .was 50'F as indicated by: the RSP strap-on RTDs,. approximately 53'F based on the changer in steam generator steam pressures . indicated at the RSP,' and . 52 F to. 60 F from various main control room temperature indications. This satisfied the greater. than or equal to 50*F cooldown requirement. The'cooldown rate did not exceed Technical Specification limits at any-time. i
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a - 3.4.4 - REMOTE BMUTDOWN c1DABILITY TEST - IBtt-223A (Continued)
SUMMARY
OF RESULTS (Continued) All transfers of contro? to and from the RSP were properly done and the RSP equipment was verified to operate properly with the following minor exceptions: o The failure of the #1 Motor Driven Auxiliary Feedwater Pump run light to illuminate had no adverse impact on test performance because alternate indications existed on the RSP' to verify pump start (e.g. flows to steam generatois and pump suctiori and discharge pressures) . o The failure of auxiliary feedwater flow indicator 1-FI-2466D to indicate flow for up to 48 minutes had no adverse impact on test performance because alternate indication of auxiliary feedwater flow to the #4 steam generator existed on the RSP for Train B, and based on verification of #4 steam generator level indications at the RSP. , o A problem with the #4 Main Steam Isolation valve dual indication was associated with the valve's stem mounted limit-switch, not the RSP indicator, and had no adverse impact on test results based on auxiliary operator verification of salve closure and limit switch operation. The valves were verified closed prior to the cooldown and were not required to be closed to establish or maintain stability following the reactor trip. o The sticking meter movement on pressurizar level indicator 1-LI-460B had no adverse impact on test results because lightly tapping the meter bezel resulted in the proper indication and also based on the alternate 1-LI-459B indie9tions. 1-LI-460B was not used to take cooldown data. i The above four items did not compromise the overall performance of ; the remote shutdown panel instrumentation and controls. L l i
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- 3. 4. 5 - LARGE IDAD REDUCTION TESTE - ISU-263A QN ECTIVE This test is performed to demonstrate the dynamic response of plant systems to automatically bring the plant to steady state conditions following a rapid 50% reduction in turbine load, and then to stabilize conditions at the reduced load. This test partially satisfies activities described by FSAR Table 14.2-3, Sheets 23 ap.1 24.
TEST _ METHOD 0laGY With plant conditions stable at approximately 75% or 100% power, a 50% load decrease is manually initiated from the turbine-generator Electro-Hydraulic Controls (EHC) at a rate of approximately 200% power / minute and plant parameters are allowed to stwbilize. The load decrease is performed by manually reducing the turbine-generator load limit setpoint to a value approximately 50% in power below the initial load reference operating power level. The load limit setpoint adjustment occurs at a rate of approximately 2004 power / minute and is performed by main control board manual pushbutton operation of a motor driven potentiometer that is set to move at that rate. These pushbuttons are permanent plant control features and the related circuitry is closely associated with the built-in turbine-generator runback circuits. The 50% power load change is a nominal value and is actually specified to be 50% 12% 1 in magnitude. The 50% load change may result in - reactor power changes of less than 50% power due to relatively low plant efficiency at lower power levels. During the course of the test, strip chart recordings and Test Data Acquisition system (TDAs) recordings of key plant parameters are taken so that plant response can be analyzed. The principal parameters monitored included RCS Tavg, Tcold, Tref, pressurizer pressure and level, steam generator pressures and levels, steam and j feedwater flows, control rod positions and speed, OTN16 and OPN16 ; setpoints, reactor power, feedwater pump speed and discharge < pressure, N-10 power, safety and relief valve positions, and steam dump valve positions.
SUMMARY
OF RESULTS The first test was performed at the 75% power plateau from ! approximately 77% reactor power. The final test execution was from ' approximately 100% reactor power. Both test executions satisfied-the following criteria o The load decreases did not cause the reactor to trip nor the i turbine to trip. i o safety injection did not initiate.
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1 J
.]
o
i l I i 1
,t j 3.4.5 f N E fnAD REDUCTION TERTS - 15U-2631 (Continued) l l
SUMMARY
OF RESULU (Continued) ) o The steam generator safety and pressuriser safety valves did not lift during either of the load reductions. l o No manual intervention was required to bring plant conditions l to steady state. o Plant variables returned to steady state conditions without sustained or diverging oscillations. l l l o Steam dump valves did not repeatedly cycle from open to closed , j position, although open position modulation did occur. l l Numerical review criteria are summarized on Table 3.4.5-1. -The 100% power test performance - was without significant incident. However, two items were noted with respect to the 75% power test performances l 1 During the first test performance, from 77% reactor power, the I turbine-generator load was reduced by approximately 570 MWe, 49% of I full power. However, due to an urgent failure in. the rod drive J system approximately 2 minutes into the event, the control rods t stopped their insertion. The effect of this failure was to limit the initial reactor power decrease to only approximately 28% of full power instead of a change of 49% to . match the turbine- j l generator load change. This power mismatch was absorbed by the ; i steam dump valves (turbine bypass valves) which stayed open for 48 - ! minuter, instead of a more typical 5-8 minute duration.- The open steam dumps created a " false" steam load not related to-actual turbine-generator load such that the reactor, the steam generators ' e and the feedwater system were not challenged to respond to the full 49% power change. The response of plant equipment, instrumentation l and control circuitry to the imposed transient was excellent, even : with the abrupt truncation of the control rod insertion. Plant process parameters behaved properly with no sustained or diverging oscillations. The principal reason for performing this test from approximately 75% power was to verify proper plant response to this type of transient prior to performing it from 100% power. If the , plant control systems vere not tuned properly for- the test from J 100% power, a much higher potential for a reactor trip would exist. This test from 77% power adequately demonstrated proper control 1 system interactions and performance such that reperformance of this test from approximately 75% power was not -)ustified.
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'I
I i f 3. 4. 5 - 1ARGE IDAD REbUCTION TESTS - ISU-263A (Continued) , l j
SUMMARY
OF RESULTS (Continued) l Also in the test from approximately 75% power, steam dump valve 1- , ' l TV-2370H failed to fully close. The problem was not related to the steam dump control circuitry. The circuitry demanded the valve to l close, but the valve remained approximately 25% open. This valve ;
- was later repaired and was demonstrated to not " hang up" on closure.
i ! Refer to Tables 3.4.5-1 through 3.4.5-3 for additional detailed data. ; l 1 i l 4 l l l
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i i TABLE 3.4.5-1 ) i Larae Imad maduction Testa sn---rv 1 i 1 Peak Auc- Auctioneered Power tioneered Expected Tavg Expected , Undershoot f'F) Resnonse (*F) l P1ateauf4) Tava (*F) Responsef'F) 75 2 <7 0 <3 i 100 2 <7 0 <3 , i ; r Auctioneered Tavg Expected Peak to Valley ('F) Response ('F) Pressurizer . , Power During / After During,/ After Pressure Expected Plateauft) Steam Dumn Steam Dumn SwinaInsin) Resnonsefnsia) ; 75 2 1 . <3 <5 +15,-80 +100,-160 i 100 2 0 <3 -<5 +25,-75 +100,-160 l Steam Duration Steam ' Generator of Max. Expected Dump Expected Power Level Expected Rod Speed Time ' Duration Time Plateauft) 3,winaft) Resnonsaft) (seconds) (seconds) ,fainutes) (minutes) , 75 +8.65,-9.15 <il5 48 approx.<30 3 <8 100 +8.3,-10.6 <il5 45 approx.<30 5 <8 t b F
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l j d , TABLE 3.4.5-2 j IARGE IDAD REDUCTION FROM 75% POWER
SUMMARY
j INITIAL CONDITION FINAL CONDITION ' Generator Load (MWe) 830 260 Nuclear Power (%) 77 33 Tavg Auctioneered (*F) 581 565 J Tref ('F) 581 565 1 N-16 Power (%) 80 33 OPN16 Setpoint (%) 112 112 l OTN16 Setpoint (4) 113 120 Pressurizer Pressure (psig) 2240 2250 ( 1 Pressurizer Level (%) 51 40 Steam Generator Level Loop 1 (%) 65 66 Steam Generator Level Loop 2 (* ) 66 66 , Steam Generator Level Loop 3 ,) 63 63 3 Steam Generator Level Loop 4 .) 65 66 l Steam Header Pressure (psig)' 1003 1030' f Steam Flow Loop 1 (pounds / hour) 2.85E6 1.2E6 Steam Flow Loop 2 (pounds / hour) 2.8E6 1.2E6 Steam Flow Loop 3 (pounds / hour) 2.8E6 1.15E6 , Steam Flow Loop 4 (pounds / hour) 2.8E6 1.2E6 Feedwater Flow' Loop 1 (pounds / hour) 2.8E6 1.7E6 i Feedwater Flow Loop 2 (pounds / hour) 2.85E6 1.7E6 i Feedwater Flow Loop 3 (pounds / hour) 2.9E6 1.65E6 I Feedwater Flow Loop 4 (pounds / hour) 2.8E6 1.6E6 1 Feedwater Temperature Loop 1 ('F) 416 350 1 Feedwater Temperature Loop 2 ('F) 418 350 Feedwater Temperature Loop 3 ('F) 418 350 Feedwater Temperature Loop 4 ('F) 416 350 l Feed Pump Discharge Pressure (psig) 1188 -1170 l Control Bank D Position (steps) 208.5 110.5 Control Bank C Position (steps) 226 225 Feedwater Pump 1-A Speed (rpm) 4646 4065 ) 4 Turbine Impulse Pressure (psig) 650 245 l 1
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{ _ _ _. _.. _ . . _ . . _ . _ - _ __ __
, I I I TABLE 3.4.5-3 IARGE LOAD REDUCTION FROM 100% POWER
SUMMARY
I INITIAL CONDITION FINAL CONDITION Generator Load (NWe) 1130 595 I Nuclear Power (%) 100 50 Tavg Auctioneered ('F) 588 570 1 Tref ('F) 589 572 i N-16 Power (%) 101 62 i . l OPN16 Setpoint (%) 110 111 j OTN16 Setpoint (4) 110 120 Pressurizer Pressure (psig) 2250 2245 i Pressurizer Level -(%) 61 36 Steam Generator Level Loop 1 (%) 68 67 Steam Generator Level Loop 2 (%) 68 67 l Steam Generator Level Loop 3 (%) 65 65 Steam Generator Level Loop 4 (%) 68 65 Steam Header Pressure (psig) 975 1018 ; , i j Steam Flow Loop 1 (pounds / hour) 3.7E6 1.80E6 i Steam Flow Loop 2 (pounds / hour) 3.8E6 1.00E6 Steam Flow Loop 3 (pounds / hour) 3.7E6 1.75E6 Steam Flow Loop 4 (pounds / hour) 3.75E6 1.85E6 i Feedwater Flow Loop 1 (pounds / hour) 3.85E6 1.75E6 Feedwater Flow Loop 2 (pounds / hour) 3.8E6 1.75E6 Feedwater Flow Loop 3 (pounds / hour) 3.75E6 1.75E6 Feedwater Flow Loop 4 (pounds / hour) 3.75E6 1.70E6 Feedwater Temperature Loop 1 ('F) 439 352 l Feedwater Temperature Loop 2 ('F) 440 353 Feedwater Temperature Icop 3 ('F) 441 353 Feedwater Temperature Loop 4 ('F) 438 352 Feed Pump Discharge Pressure (psig) 1133 1130 - Control Bank D Position (steps) 195 90 i Control Bank C Position (steps) 225 205 ;
)
Feedwater Pump 1-A Speed.(rpm) 4848 3928 Turbine Impulse Pressure (psig) 880' 405 I
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. - . . , . . . . . . . , . , _ -. .....s.. _ ,_ _ . , . . .- _, _ , . . . . . . . . . , _ , _ _ . . . . . . . . -
l < i !
- 3.5 INSTRUMENTATION AND CALIBRATION TESTING l 4 I
- 3. 5.1 - CALIBRATION OF FEEDWATER AND STEAM FIDW INSTRUMENTATION AT POWER - ISU-202A OIL 7ECTIVE l i .
The purpose of this test is to verify the calibration of Feedwater 1 ! (FW) flow and Steam Flow (SF) instrumentation at each of the major , reactor power levels. Calibration of the Feedwater flow 1 instrumentation, because the characteristics of the Main Feedwater , ! flow venturi are well known, is relatively straightforward. The ! j differential pressure d/p transmitters are zero and span checked,
- and the downstream electronics conversion cards are shown to be in '
calibration. Steam flow is determined by measurement of steam generator pressure and the differential pressure developed by the i flow of steam across the steam generator steam exit nozzle and ; associated piping to a downstream point on the main steamline. The r j major task is that of determining both the zero and the span of the
- steam flow differential pressure transmitters, because neither of these quantities are known precisely prior to actual power.
operations. TEST. METHODOLOGY The test is comprised of three related activities. First, at hot, zero power, zero flow conditions, both FW and SF flow transmitters
- are verified to be at or adjusted to be at zero output. provicions
- are made for high accuracy test d/p transmitters to be installed in parallel with the permanent plant FW d/p cells. As power is increased, the permanent plant instrumentation is verified to be I
within specified calibration tolerances with respect to the test d/p transmitters. Finally, the as-built spans of the SF measurement system are determined by requiring that the indicated SFs agree with~the simultaneously measured FW flows. i With the plant in Mode 3 at hot, zero power conditions,'an average - 1 , reactor coolant temperature of approximately 557'F and steam l generator pressures of approximately 1100 psig, the SF flow transmitters should have a zero output. Because the condensing pot for the low pressure side of the SF dp cell is at an elevation approximately three feet above the high pressure condensing pot, 4 zero offset has to be incorporated into the transmitter calibrat!.on i to account for this' static head difference. Due to differences in , the thermal expansion of structures and piping to which these. ' ce densing pots are attached, and also due to high pressure static l' shift effects on the actual transmitters, the pre-test estimate of ; the zero offset usually has to be modified. The zero flow check of the FW flow transmitters is accomplished at' low power operations " (<10% reactor power) with the feedwater header pressurized by a Main Feedwater pump. Because the FW d/p cell' instrument taps are '
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_r - _ . . , . , _. i
l i ! I l \ j 3. 5.1 - CALIBRATION OF FEEDWATER AND STEAM VIDW INSTRUMENTATION i
- &T POWER - E -202A (Continued) ]
) TEST METHODOIDGY (Continued) I installed in horizontal feedwater piping, the zero flow check is j performed primarily to account for high pressure static shift of i the transmitters. At higher power levels up to and including 100%, data is taken d which ultimately is used to determine the full span of the SF d/p i transmitters as well as verifying the calibration of the FW flow l j instrumentation. With steam generator blowdown secured and the plant stable, a high accuracy measurement of FW flow is obtained. 1 The most important parameters are the FW venturi differential i pressures,and these are obtained using precision, temporary test d/p cells which have both pre-test and post-test calibrations
! performed to minimize errors due to instrument drift or zero and span shifts. The output of the permanent plant d/p cells are +
compared to these values and adjustments are made to the permanent ] plant d/p cells,- if necessary. Feedwater pressures and temperatures are also required for the determination of FW flow. l As stated previously, only two parameters are used to measure SF, steam ganerator pressure and SF differential pressure. Unlike the FW permanent plant instrumentation, SF instrumentation has an allowance for density compensation based on steam generator , pressure. Setting SF equal to FW flow, and extrapolating to full 3 scale flows, allows the corresponding full span of the SF transmitter to be established. Recognizing that the most accurate , extrapolation is that using data from higher power levels, the SF transmitters are not respanned until the 75% power data has been obtained. During steady operations at the 1004 power plateau, a final set of data is taken and used to verify the calibration of both the SF and FW flow instrumentation. Adjustments and i recalibration of the instrumentation is then undertaken, if found necessary.
@MMARY OF RESULTS
' l Refer to Table 3.5.1-1 for detailed test results. In Mode 3 at an RCS temperature of approximately 557'F and a ) corresponding steam generator pressure of approximately 1092 psig, the output of the SF transmitters at the NLP card was required to be 0.001, (+0.024,-0.000) volts at this zero flow condition. The as-found data ranged from -0.009 to +0.079 volts for these eight SF transmitters and all eight were recalibrated. With the Main Feedwater header pressurized by a Main Feedwater pump, and the equalizing valve on the FW flow transmitter manifold
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_ _ _ _ _ _ _ _ _ - - - . -- - - --. - 1
h 1 - CALTBRATION OF FEEDWATER AND STEAM FIDW INSTRUMENTATION AT POWER - ISU-202A (Continued)
SUMMARY
OF RESULTS (Continued) open, the NLP card output voltage is also required to be 0.001 (+0.024,~0.000) volts. The collected data showed a range of -0.288 to +0.0034 volts, with only one of the eight transmitters satisfying the Acceptance criterion. The remaining seven out-of-tolerance transmitters were recalibrated. The zero flow check of the SF and FW flow transaf tters was therefore shown to be necessary and all sixteen transmitters were shown to be properly calibrated at zero flow conditions prior to ascending to s:.gnificant power levels . At-power testing was performed, with ratesting as required, at the 30%, 50%, 75% and 100% plateaus. Primary and secondary plant' systems were first shottn to be stable by trending the following parameters reactor power, average reactor coolant temperature, pressuriter pressure, steam generator levels, and SF and FW flows. After demonstrating adequate plant stability, the Test Dsta Acquisition System (TDAS) was used to collect the required data both from permanent plant and high accuracy _ test equipment. The 30% and 50% data indicated that all eight of the SF transmitters would eventually have to be recalibrated. The SF/FW flow mismatches, however, were. not so large tnat they adversely impacted plant operations. SF/FW flow mismatches serve as inputs to the steam generator level control system. The required level of agreement between SF and FW flows was ,<5% of full flow, with the differences ranging from 3.3% to 14.4% at 30% power and 6.7% to 16.6% at 50% power. Comparison of the permanent plant FW flow instrumentation with test instrumentation showed that only one transmitter,1FT-540, required adjustment at the 30% power plateau. After recalibration of this d/p transmitter a retest of this instrument was performed with satisfactory results. At the completion of the 50% plateau testing, all FW flow loops were i verified to be calibrated adequately and it was expected that the d SF transmitters would have to be respanned after obtaining and evaluating the 75% power data. Test data acquired at the 75% plateau, when combined with the lower power data, was used to derive the SF transmitter scaling shown on Table 3.5.1-1. The table also includes the scalings that were installed prior to power ascension testing. ~ These were based on comparisons to other similar plants and best estimate engineering calculations. The SF transmitters were recalibrated to these new scalings and a - retest at the 75% power level showed that the required 5% agreement level between indicated and calculated SF was achieved for all eight transmitters. This ' first retest at this power level showed that FW flow transmitters 1FT-530 and 1FT-541
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__.__ _ _ _ . _ _ . _ _ _ _ _ _ _ __ _ .._ ~__ _ __ 1 3.5.1 - CALIBRATION OF FEEDWATER AND STEAM FIDW INSTRUMENTATION ! AT POWER - ISU-202A (Continued) 1
SUMMARY
OF RESULTS (Continued) . required recalibration. After completion of these recalibrations, a second retest showed that while FW flow ~ indications were satisfactory, the 0.5% agreement level between permanent plant and test equipment differential pressures was not satisfied for 1FT-520 , and 1FT-530 (differences of 0.73% and 1.65%, respectively). The Test Review Group, after consideration of these results, decided i that the magnitudes of out-of-calibration conditions were such that plant operations would not be adversely affected. Based on this and because at least one more complete set of data was to be ! collected at the 100% plateau, these two transmitters were not l J recalibrated at that time. 4 The initial set of data at 100% power indicated that three FW flow transmitters failed the Acceptance criterion requiring no larger than a 0.5% difference in measured differential pressure between J the permanent plant and test instrumentation. In addition to 1FT-520 and 1FT-530, previously identified at 75% power to be out-of-tolerance, 1FT-521 indicated a 0.7% difference. Additionally, 1FT-520 and 1FT-530 did not satisfy the required 1% agreement level , between indicated and calculated flows ( differences of 1.5% and > l 1.4%, respectively). After recalibration of these instruments the ! next retest yielded data which demonstrated that all eight FW flow , transmitters and their associated circuitry were within the : required levels of calibration. ,
~
At 100% power the indicated and calculated values for SF were required to be different by no more than it. The first set of data at this plateau showed that all SF transmitters except 1FT-543, with 3.5% difference, satisfied this requirement. A calibration ! check of this instrument was performed and it was found to be out 1 of calibration. Af ter recalibration using the new scaling shown on Table 3.5.1-1, the 1FT-543 retest data yielded acceptable results. However, in the retest 1FT-532 was found to have now failed the 1% agreement level,'having a 1.1% difference. A detailed examination of all components in the SF instrument loops was undertaken at this I time to determine why SF instruz::nnts or loops were, at different times, apparently drifting out of calibration. Although it had , been noted that several of the steam generator pressure transmitters had been out of tolerance to varying degrees, the impact of this on indicated SF had not been fully evaluated.- The calibration of these particular instruments is not directly evaluated by this procedure. As noted earlier, steam generator I I pressure is used to compensate for variations in steam density by the permanent plant instrumentation. Steam generator pressure values from test -instrumentation were then ~ used to avoid this , potential error and provided the best evaluation of the i
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j i , i i
- 3. 5.1 - CALTBRATION OF FEEDWATER AND STEAM FIDW INSTRUMENTATION '
j AT POWER - ISU-202A (Continued) SUMMARLOLRES11LTS (Continued) as-built calibration for the SF transmitters. This steam generator pressure compensation calcult'c ion was performed for the eight SF loops with the result that all scalings recommended at the 75% power plateau were found to be satisfactory with the. exception of 1FT-543. As shown on Table 3.5.1-1 the recommended scaling is now
~44 to 313 inches of water column (inWC) with a ' span of 357 inWC.
It should be noted that although this is a significant change, the impact on indicated flow is only 2%. A Work Request was written to recalibrate 1FT-543 to correspond to this scaling. When the work has been completed, ratesting of 1FT-543 is to be accomplished using permanent plant procedure PPT-P1-5001, " Calibration of Feedwater and Steam Flow Instrumentation at Power", which uses the same basic methodology as this test. The last portion of testing performed by this procedure was to verify the SF transmitters zero flow outputs upon return to Mode 3 following the plant trip from 100% power test. Seven of the eight transmitters were found to be out-of-tolerance, ranging from
-0.003 to +0.055 volts versus the required range of 0.001 (+0.024, j
-0.000) volt. The instruments were re-zerood and ratesting showed that all as-left values satisfied the Acceptance criterion l
tolerance. a 4 In summary, the testing was viewed as having been successful with . the overall final results being: I o The FW flow transmitters were left within their tightly specified calibration range j i o The downstream FW flow instrumentation was shown to be within 1 calibration o The SF calibrations established in this test were acceptable, I I with only one transmitter requiring a new scaling. Performance of the permanent plant procedure PPT-P1-5001 may be used in the future to finalize and/or enhance the scaling of the SF transmitters over tima. l l
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- l 1 l
l i Table 3.5.1-1 calibration of steam Flow Transmitters j
! All values are given in units of inWC.
ORIGINAL SCALING 75% POWER SCALING Transmitter Rance Scan Ranae Snan 1FT-512 -29 to 447.6 476.6 -27 to 334 361 1FT-512 -45 to'431.6 476.6 -41 to 320 361 1FT-522 -30 to 427.8 457.8 -27 to 370 397 1FT-523 -28 to 429.8. 457.8 -26 to 371 397 , 1FT-532 -30 to 489.2 519.2 -28 to 420 448 1 1FT-533 -31 to 488.2 519.2 -28 to 420 448 1 1FT-542 -30 to 432.6 462.6 -29 to 344 373 1FT-543 -46 to 416.6 462.6 -44 to'329 373
- I
~
- Evaluation of the 100% data indicated that the best estimate range for 1FT-543 was -44 to 313 inWC with'a span of 357 inWC.
4 b l l l l l -152- I o l , ,. _ - - - . . _ - _ . _ _ . _ _ _ . _ . . . . . _ _ _ _ _ _ _ _ . . _ . . . . _ . . _
i ! 3. 5.2 - THERMAL POWER MEASUREMENT AND STATEPOINT DATA COLLECTION - ISU-224A OILTECTIVE , This test is performed to determine reactor thermal power by a l secondary plant calorimetric measurement and to collect control and ! protection instrumentation data at steady state power levels ' (statopoints). TEST METHODOLOGY l From stable plant conditions, statopoint data is collected and calorimetric power measurements made at the approximate 0%, 30%, 50%, 75%, 90% and 100% power levels. Calorimetric data includes feedwater temperature, main feedwater flow venturi differential j pressures, steam pressures, atmospheric pressure, and steam' ' generator blowdown flows. Statopoint data is taken from the main control board indicators, from the P2500 Process computer, and from the Test Data Acquisition System (TDAS). Data recorded includes N-16 powers, RCS flows, RCS temperatures, pressurizer level and pressure, nuclear instrumentation outputs, steam pressures, main generator output, ' steam generator levels, feedwater flows, steam flows, and feedwater pressures and temperatures. The recorded values are compared against each other to ensure consistency. ' Four data sets are taken within an approximate 20 minute time period for each of the parameters to assure good quality of data. '
SUMMARY
OF RESULTS Reactor thermal power was determined based on calorimetric measurements at the 30%, 50%, 75%, 90% and 100% power testing ! regimes. Calorimetric measurements do not apply to the Mode 3 test at 0% power. I statepoint data was taken at all power levels from all available channels. Certain P2500 process computer and TDAS channels were unavailable at some power levels due to software reconfiguration. and hardware installation status. There were no- absolute requirements that all channels be compared at all power levels. There were sufficient channels available at all power levels to adequately verify proper display:of plant conditions. Throughout the various performances of this test, a number of main control board indicators and associated P2500 process computer and TDAS channels were noted to differ by notable amounts. There were no specified agreement criteria in this test. All items were evaluated and satisfactorily dispositioned as being within calibration tolerances or corrected by'recalibration. 4
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l I i l i 3.5.3 - OPERATIONAL ALIGNMENT OF PROCESS TEMPERATURE AND N16 ( INSTRUMENTATION - ISU-2264 . OILTECTIVE ; This test is performed to align the N16 and Tavg process instrumentation, to verify the linearity of the . N16 and . Tavg ( instrumentation, to detern:.no the optimum voltage setting for the i 4 N16 detector High Voltage power supplies, and to determine the N16 ;
- detector currents at various power levels. This. test partially )
i satisfies activities described by FSAR Table 14.2-3, Sheets 9, 10 , and 22. l l TEST. METHODOLOGY ; At Comanche Peak, the Reactor Coolant System (RCS) Resistance Temperature Detector (RTD) manifold hot leg temperature (Thot) and ! i cold leg temperature (Tcold) measurement instrumentation has been replaced by a Nitrogen-16 (N16) power monitor and an in-line Tcold , RTD. The N16 power monitor measures the thermal power of the : reactor by detecting the amount of N16 present in the coolant. The concentration of N16 in the coolant is directly proportional to the fission rate in the core and is detected by measuring the high energy gamma flux from the N16 decay which penetrates the walls of the hot leg piping. The fast response in-line Tcold RTD is in a thin wall thernowell installed in the cold leg piping. The process control system uses these inputs to generate a Tavg signal which is used for input to Rod Control, pressurizer level control, and steam dump control. An N16 Power signal is also generated which inputs to the Reactor Protection System for Overtemperature and Overpower reactor trips. This test is a collection of eleven different tests of the N16, Tcold and Tavg process instrument loops which are performed throughout the startup program from Mode 3 through 100% power. . Refer to m a ble 3.5.3-1 for a matrix of'which tests are performed at each plant condition. The DETERMINATION / SETTING OF N16 DETECTOR HIGH VOLTAGE test is , performed at approximately 50% reactor power. The N26 gamma I detectors are tested one loop at a time. The current output from the N16 gamma detectors is measured by a picommmeter while the high l voltage power supply output voltage is adjusted from 300 volts-to 1200 volts. This data is plotted to determine the plateau region l i of the curve, the region. of minimum out put current change for a given voltage change. The power supply .s then set to a value in this plateau region of the curve, nominally 800 volts. , 1 The N16 CURRENT MEASUREMENT test is performed in Mode 3 and at the ! 30%, 50%, 75%, and 100% power plateaus'.. The input voltage and ) output voltage of each N16 power monitor module is recorded 1
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) 3.5.3 - OPERATIONAY. ATIGNMENT OF PROCESS TEMPERATURE AND N16 ! l INSTRUMENTATION - ISU-226A (Continued) ] TEST METHODOIDGY (Continued) simultaneously with the reactor thermal power ' from a precision secondary calorimetric. At zero power (Mode 3) the power monitor 1 module output is verified to be zero 10.03 volts. The RCS COLD LEG TEMPERATURE CHECKS are performed at every power plateau and in Mode 3. The active cold leg RTD temperature, as measured at the output of the NRA card in the process control rack, ) is compared to its associated spare RTD temperature, measured as J resistance from the RTD. These temperatures are required to agree j i within 1.W for each RCS loop. The VEFIFICATION OF TAVG CIRCUITRY test is performed at every power plate,u and in Mode 3. The process control system Tavg signal is l comp =M4 to a calculated value generated by the following equations ! Tavg = Tcold + (K9) (N16 Power) ) 1 where Tcold is the cold leg temperature from the active RTD l circuitry, N16 is a power signal generated by. the N16 process loop, and K9 is a constant equal to 1/2 the full powet tetaperature ; dif ference of hot leg to cold leg. The calculated Tavg is verified to be within 0.5'F of the Tavg signal. The NEUTRON STREAMING DETERMINATION test is performed at the 75% power plateau. .The N16 gamma detectors in the RCS hot legs monitor ! gamma rays from the decay of N16 in the RCS water. Additionally, gamraa rays streaming directly from the upper portion of the reactor core and secondary gamma rays generated by streaming neutrons add to the N16 power signal. This contribution to the N16 power signal ' i comes primarily from the top region of the core. The signals from the top two detectors in each nuclear instrumentation power range channel arr used to compensate the associated N16. power signal. During the Incore/Excore Detector Calibration test, an axial Xenon ' transient is initiated causing the neutron flux to shift axially in the core. The following data is taken during the transient: N16 power monitor output, the output from each of the top two power < range detectors and precision calorimetric power ' measurements. Using the relationship Q = Ag (VUN16) + A2 (VA ) + A3 (Ys ) where Q is calorimetric power, V"N16 is the N16 Power monitor output, V, is the top power range detector output and V, is the next to top power range detector output, the constants A,, A, and A3 are determined by linear regression analysis. The neutron streaming compensation gains are then calculated and used to calibrate the process channels to negate the core streaming effects.
-155--
, _ . , ._r,.,, . _ . _ . , _ , , , , . , . , , .
,,,,m. , _ . , . , ,
i j I l
- l
; 3.5.3 - OPERATIONAT.Af.TGNMENT OF PROCE j j INSTRUMENTATION - ISU-226A (Continued) ,
- 1 TEST METHODOLDGY (Continued)
The FULL-POWER DELTA-T (K-9) DETERMINATION test is performed at the
- 75% power plateau. The purpose of this test is to determine the 100% power hot leg to cold leg temperature difference by extrapolation of data collected at the previous power levels. Then a new K-9 constant (1/2 of.the full power temperature difference) is calculated and used to recalibrate the N16 Tavg circuits.
I First, full power cold leg temperature is extrapolated using Tcold I data from previous power levels. Then full power volumetric ! enthalpy (density compensated specific enthalpy) is extrapolated i using the volumetric enthalpy frota RCS flow measurements performed i at previuus power levels. The full power hot leg enthalpy and temperature are calculated based on these extrapolated values and l the ASME Steam Tables. From this, the full power hot leg to cold 1 l leg temperature difference and K-9 constant are calculated. 1 . The FULL POWER DELTA-T (K-9) VERIFICATION test is performed at the 100% power plateau. Cold leg and hot leg temperatures are determined from the RCS flow test procedure. (hot leg temperature is determined by iteration of TTFM measurement of RCS flow, secondary precision calorimetric power, and Tcold). Then this temperature 1 difference is compared to a temperature difference equivalent to l twice the current value of K-9. Any loop which differs by more , than 1% is recalibrated using the new X-9 value determined from the actual full power cold leg and hot leg temperatures. The N16 POWER CHECK - K8 ADJUSTMENT test is performed at every power plateau and in Mode 3. The N16 power signal for each loop is compared to precision secondary calortmetric power and adjustments i are made to make them match. Below 75% power the gain.of the N16 power monitor module itself is adjusted. . After the neutron streaming gains have been determined at 75% power (as described in I a previous paragraph), adjustments are made to the K8 constant I instead of the power monitor module. At zero power (Mode 3), the j N16 power signal is verified to be zero 10.05 volts. ' The TEh'ERATURE DECALIBRATION DATA test is performed at the 50% power plateau. This test is performed to determine the sensitivity of N16 power measurements and nuclear instrumentation power range l measurements to changes in RCS temperature. The Automatic Reactor L Control System test changes the RCS - average temperature to 5'F above and 5'F below the normal average temperature,- with reactor l power held constant. At those plant conditions, this test collects l the following data: N16 power, RCS temperatures, Nuclear l Instrumentation outputs, and calorimetric power. i
~156- j l
l
_ . _ _ _ _ . _ _ . _ _ _ _ _ _ . _ . _ . _ . _ . ._.._ __ _ _ ~ . _. - _ . . _ _._ a I I i
- 3. 5. 3 - OPERATIONAL ALTGNMENT OF PROCERE TEMPERATURE AND N16 j INSTRUMENTATION - ISU-226A (Continued) !
TEST METHODOLDGY (Continued) The N16 POWER LINEARITY CHECKS are performed after the N16 power instrumentation has been adjusted at full power and during the power ascension following a plant trip. The power output from each 1 loop is plotted against calorimetric power and evaluated for ! linearity by the NSSS vendor. The N16 ELECTRICAL ZERO DETERMINATION test is performed at least
- four hours after a reactor shutdown from 100% power. Input and )
1 output voltages of each power monitor module are recorded. The
, output voltages are verified or adjusted to be sero 10.033 volts. i ! This ensures proper compensation for the background gamma flux.
This startup test also collects N16 data from the following transient tests: Full Load Rejection and Turbine Trip, Design Load Swings, Large Load Reduction, and Turbine / Generator Trip with
- Coincident Loss of Offsite Power. This data is evaluated by the !
NSSS vendor for proper response of N16 during transients. J
SUMMARY
OF RESULTS The DETERMINATION / SETTING of N16 DETECTOR HIGH VOLTAGE test rosults , indicated that the N16 detector current is independent of the power supply high voltage setting between 300 and 1200 volts. The high voltage power supplies were set to 800 volts. The N16 CURRENT MEASUREMENT test results weret Power Plateau Mode 3 30% '50% 50% Retest #2 Calorimetric Power (%) 0 29.21 50.14 48,15 Loop 1 input (volts) 0.0002 -0.251 -0.4306 -0.415 i Loop 1 output (volts)_0.0005 .2.046 3.2645 3.153 ) Loop 2 input (volts) 0.0003 -0.244 -0.4182 -0.404 j Loop 2 output (volts) 0.00004 2.082 3.3085 3.190 Loop 3 input (volts) 0.00025 -0.240 -0.4127 -0.397 Loop 3 output (volts) 0.0008 2.081 3.2929- 3.169 Loop 4 input (volts) 0.00003 -0.242 -0.4203 -0.405 Loop 4 output (volts) 0.00035 2.067 3.3151 3.193 At zero power, the criterion is 0.0 10.03 volts
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I c , 3.5.3 - OPERATION 11,if,Ttit j INSTRUMENTATION - li.l-226A (Continued) i
SUMMARY
OF RESULTS (Continued) j l N16 CURRENT MEASUREMENT test results (continued) j Power Plateau 2.5.1 1.0.01 Calorimetric Power (%) 76.19 99.33 j Loop 1 input (volts) -0.668 -0.86 Loop 1 output (volts) 5.222 6.77 , Loop 2 input (volts) -0.645. -0.84 ) Loop 2 output (volts) 5.262 6.86 Loop 3 input (volts) -0.633 -0.83 Loop 3 outpnt (volts) 5.267 6.86 Loop 4 input (volts) -0.638 -0.83 Loop 4 output (volts) 5.201 6.76 The loop inputs were all within 0.03 volts of zero in Mode 3. The ratest at 50% power was performed due to a discrepancy in'the calorimetric power measurements. All of the data collected was acceptable. The RCS COLD LEG TEMPERATURE CHECKS test results were: Epwer Plateau Mode 3 304 30% Reteet#1 50% Nuclear Power (%) 0 29.5 28.1 47.5 Loop 1 Active ('F) 557.57 555.56 556.24 558.34 Teold Spare ('F) 557.19 555.79 556.43 558.77 ! 4 Tcold ('F) +0.38 -0.23 -0.19 -0.43 Loop 2 Active ('F) 557.13 555.34 555.92 558.48 Teold Spare ('F) 556.93 554.82 555.41 557.63 l ATcold ('F) +0.20 +0.52 +0.51 +0.85 Loop 3 Active ('F) 557.36 555.53 556.10 558.51 Tcold Spare ('F) 557.43 556.00 556.55 559.17 A Tcold ('F) -0.07 -0.47 -0.45 -0.66 l Loop 4 Active ('F) 557.53 555.36 556.01 558.23 Tcold Spare ('F) 556.83 554.14 554.83 556.81 ATcold ('F) +0.70 +1.22 +1.18 +1.42
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3.5.3 - OPERATIONAL ALIGNMENT OF PROCESS TEMPERATURE AND N16 INSTRUMENTATION - ISU-226A (Continued) I
SUMMARY
OF RESULTS (Continued) 50% 100% ' Power Plateau Retest 42 754 100% Retest 43 Nuclear power (%) 48% 76% 100% 97% Loop 1 Active ('F) 559.62 560.16 561.42 559.10 ; Tcold Spare ('F) 559.95 560.64 562.05 WR 559.76 i A Tcold ('F) -0.33 -0.48 -0.63 -0.66 Loop 2 Active ('F) 559.72 560.17 561.02 558.95 Tcold Spare ('F) 558.87 559.10 559.79 WR 559.96 ATcold ('F) +0.85 +1.07 +1.23 -1.01 Loop 3 Active ('F) 559.85 560.09 561.08 558.68 < Tcold Spare ('F) 560.45 560.88 562.14 WR 559.56 ! A Tcold ('F) -0.60 -0.79 -1.06 -0.88 ' Loop 4 Active (*F) 559.63- 560.00 561.22 _558.96 4 Tcold Spare ('F) 558.24 558.44 559.43 WR 559.18 ATcold ('F) +1.39 +1.56 +1.79 -0.22 The values listed as WR for Retest #3 were from the Wide Range Tcold RTDs. The acceptance criterion of this test was that the ATcold for each loop shall be 1.2'F or less. Retest #1 was performed at 30% due to the failure of loop 4 to meet this criterion and because the RCS temperature had changed 1. 2'F during the test. Retest #2 was performed at 50% due to another failure of loop 4 and a discrepe'cy. in the calorimetric measurement. The NSSS vendor recommended that power ascension continue to 100% power to gather more data since the 1.2'F criterion only applied at 100% power. At 100% power loops 2 and 4 failed the criterion. The NSSS vendor determined the measured difference was due to a physical temperature divergence in the cold leg piping (Cold Leg Streaming) . Since the purpose of the 1 1.2'F criterion is to verify the NRA card is properly adjusted, the wide range RTD which has the same installation orientation as the active RTD was used for comparison in Retest #3'at'100% power. These results were satisfactory. The NSSS Vendor has reviewed the data and believes that the active RTDs provide the most accurate , measurement of bulk coolant temperature in the cold leg piping. ! The Cold Leg Streaming issue is still under evaluation by the NSSS 1 Vendor. l 1 l l l l
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l
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. ~ - - . . _ . , .i-.. , ,,.+4 _ m m_ __ -2._a __ _ v-- twe - --'
l 3.5.3 - OPERATIONAL ALIGNMENT OF PROCESS TgttP7.:RATURE AND.N16 : INSTRUMENTATION - ISU-226A (Continued) SIDRER'. .QF_RESULTS (Continued) : The VERIFICATION OF TAVG CIRCUITRY test results were Taver Error (*F) l Power Plateau Imon i Loon 2 Imop 3 Imon i criterion l Mode 3 -0.11 +0.07 -0.21 +0.07 <10.5 l Mode 3 Retest #1 N/A N/A -0.04 N/A- <10.5 30% -0.11 -0.07 -0.12 +0.08 <10.5 : 50% -0.02 -0.01 -0.15 0.00 <10.5 , 75% -0.141 -0.085 -0.161 -0.069 <10.5 ! 100% -0.498 +0.06 -0.27 0.0 <10.5 100% Retest #2 -0.26 -0.09 -0.46 0.0 <i0.5 Retest #1 in Mode 3 was required by' recalibration of. loop 3. Retest #1 at 100% was to be performed due to circuit calibrations
- for new K9 values identified in the FULL POWER ' DELTA T (X9)
VERIFICATION test. Due to problems with test instrumentation, the data obtained was indeterminate and ratest - #2 at 100V power was performed. All loops calculated Tavg within the required 0.5'F eccuracy. The NEUTRON STREAMING DETERMINATION test results were , Item Loon 1 Loon 2 Loon 3 Loon 4' A, 14.9855 14.7652 14.8892 15.0800 A2 -0.8191 -0.4350 -0.8046 -0.8749 A3 0 0 0 0 , G, -0.0547 -0.0295 -0.0540 -0.0580 G, 0 0 0 0 where G, = A 2/A1 nuclear, and G A next to top = instrumentation/A i , the gains for.the top and detector signal l compensation. These values of gains were acceptable and used to calibrate the neutron streaming compensation circuits. The FULL-POWER DELTA-T (K-9) DETERMINATION test results were Extranolated Values Looo 1 Loon 2 Loon 3 Loon 4 Tcold ('F) 559.71 559.78 559.66 559.55 Hot Leg Enthalpy 634.76 635.10 634.12 635.95 (BTU /lbm) Thot ('F) 614.59 614.81 614.17 615.37 DELTA-T ('F) 54.88 55.03 54.51 55.82 K-9 ('F) 27.44 27.52 27.26 27.91 l
-160-y --w.. ...3 ,e - . . _ . , . , _ , , ,, _ _ _ . . . _ _ . _,_ . _ . . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ , _ . _ . _ _ _ , , . . . _ . ,
l 1 [ i 3.5.3 - OPERATIONAL ALTGNMENT OF PROCESS TEMPERATURE AND N16 l INSTRUMENTATION - ISU-226A (Continued) ! i .: I
SUMMARY
OF RESULTS (Continued) l The K-9 constant used for preliminary calibration - of N16 Tavg circuitry was 28.1'F. Since these extrapolated values were within j
- 1. 0'F of 2 8.1'F, no adjustments to K-9 were made at 75% power. l The FULL-POWER DELTA-T (K-9) VERIFICATION test results were as I follows where the % Error = ((Full Power AT - (2xK9))/(2xK9))x100 l and the New K9 value is equal to the Full Power A T/2: !
l Measured Valuaa Loon 1 Loon 2 Loon 3 Loon 4 i Tcold ('F) 560.96 560.58 561.18 560.54 ' Thot ('F) 615.49 616.04 615.20 616.27 Full Power AT('F) 54.89 55.66 54.34 55.97 2 x K9 ('F) 56.20 56.20 56.20 56.20 J Error (%) -2.33 -0.96 -3.31 -0.41 j New K9 ('F) 27.45 27.83 27.17 N/A NOTE: Tcold and Thot are measured values at 99%, thT is . extrapolated to 100.0% power. The error for loops 1 and 3 exceeded the allowed 14 and were recalibrated with new K-9 values. Loop 2 was also recalibrated with a new K-9 value since it was very close to the allowed 1%. The loop 4 value was left at 28.1'F. The VERIFICATION OF TAVG CIRCUITRY Retest #2 at 100% was performed following the adjustments to K-9 and was satisfactory. The N16 POWER CHECK - K8 ADJUSTMENT test results were ; For zero power, the N16 output voltage must be -0.05 volts to +0.05 volts Test Loon 1 Loon 2 Loon 3 Loon 4 Mode 3 (volts) 0.01752 0.00951- 0.07667- 0.01082 Mode 3 Retest #1(volts) N/A N/A 0.00723 N/A For at-power measurements, if N16 power differs from calorimetric -> by 1% or more, the N16 gain is to be adjusted.-(Listed Values are calorimetric power (4) - N16 power (%)) l Power Plateau LooD 1 Loon 2 Loon 3 Loon 4 l 30% -0.48 -0.63 -0.60 -0.64 l 50% 2.21 1.95 2.13 2.07 9 50% Retest 62 1.78 1.59 1.84 1.59 -l (before adjustment) , 50% Retest #2 0.02 0.01 0.16 0.16 f (after adjustment) 75% -0.28- -0.37 -0.25 +0.44 100% - 0.46 -0.16 -0.91 -0.01
-161-e I
y ,=- , w + - , - -cwe ,,..w~n , , , . g=wg y
. _-- _ - ~ ... . . _ _ - - . ~ . - . . . - .- - -. - . .-- - .
I 3e5.3 - OPERATIONAL ALIGNMENT OF PROCESS TEMPERATURE AND N16 INSTRUMENTATION - ISU-226A (Continued) l
SUMMARY
OF RESULTS (Continued) i Retest #1 in Mode 3 was performed due to recalibration of loop 3 to correct an out of specification voltage. Following recalibration all voltages were within 0.05 volts. of zero. At- the 50% power plateau, a discrepancy in the calorimetric data resulted in Retest
#2. The results of Retest $2 were all greater.than the required )
l maximum 1% difference between the calorimetric and N16 power. All. ) l four of the N16 power monitor modules gains were readjusted and I following readjustment, all results were within 1%. -The results'of , testing at 75% and 100% power were within it and no further 3 adjustments were made to the power monitor modules and none were , made to K8 constants. 4 The TEMPERATURE DECALIBRATION DATA test'results were Tavg Tavg DELTA Item At Normal Tava ij'E .fE +5 to -5'F Calorimetric , Power (%) 48.19 48.21 48.62 -0.41 N16 Power (4) 48.02 48.51 48.05 0.46 Nuclear Power (%) 47.36 49.16 46.24 2.92 Tcold ('F) 558.81 564.03 553.54 10.49 Thot ('F) 588.32 593.51 583.35 10.16 j Tavg ('F) 572.47 578.08 567.28 10.80 l l The test results indicate that N16 power measurements are less sensitive to temperature changes.than Nuclear Instrumentation power range measurements. The small change of 0.46% power /10'F shows ' that the N16 instrumentation is sufficiently insensitive to temperature changes to not require compensation. circuitry. The N16 POWER LINEARITY CHECKS were performed following the large load reduction test. Reactor power was reduced to 30% and then returned to 100% with hold points at 504 and 75% power to collect N16 power data. The NSSS vendor evaluated the data and determined it to be acceptable. The N16 ELECTRICAL ZERO DETERMINATION test results were Loop Inout(volts) Outout(volts) As Left Outout(volts) , 1 0.000 0.004 0.004 2 0.000 0.003 0.003-3 0.000 0.004 0.004 4 0.000 0.003 0.003 l 1
-162-
) l 3.5.3 - OPERATIONAL AtTCNMENT OF PROCEEE TEMPERATURE AND N16 INSTRUMENTATION - ISU-226A (Continued)
SUMMARY
OF RESULTS (Continued) l The N16 Electrical Zero Determination was performed approximately four and one-half hours after the Full-Load Rejection and Turbine Trip from-100% power. The output voltages were within the required-10.033 volts of zero tolerance and no adjustments were made. The N16 Transient Response Data was collected and provided to the NSSS vendor. The evaluation determined that the N16 system responded properly to the transients. I i I 1
-163-b
TABLE 3.5.3-1 PROCESS TEMPERATURE /N16 TESTS VS. PLANT CONDITIONS NATRIX MODE 3 30% 50% 754 100% DETERMINAT. ION / SETTING X OF N16 DETECTOR HiGH VOLTAGE N16 CURRENT MEASUREMENTS X X X X X Retest 2 RCS COLD LEG TEMPERATURE X' X X X X CHECKS Ratestl Retest 2 Retest 3 VERIFICATION OF TAVG X X X- X X CIRCUITRY Ratesti Retest 2-NEUTRON STREAMING X DETERMINATION FULL POWER DELTA-T(K-9) X DETERMINATION FULL POWER DELTA-T(K-9) X VERIFICATION N16 POWER CHECK - K8 X X X -X X ADJUSTMENT Retesti Retest 2 TEMPERATURE X DECALIBRATION DATA N16 POWER LINEARITY Performed- during Power Ascension CHECKS following 100% power testing N16 ELECTRICAL ZERO Performed at zero power following DETERMINATION 100% power testing N16 TRANSIENT N16 Data collected during: RESPONSE DATA Full Load Rejection and Turbine Trip Design Load Swings Large Load Reduction Turbine / Generator Trip With Coincident Loss of Offsite' Power
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I I
\
- 3. 5. 4 - OPERATIONAL ALTGNMENT OF NUCMAR INETRUMENTATION - ISU-204 A t
OBJECTIVE This test is performed to verify that the excore Nuclear ' Instrumentation System (NIS) functions per design. This test satisfies activities described by FSAR Table 14.2-3, Sheets 9 and , 10. TEST METHODOLOGY selected parameters are evaluated, monitored, and determined during 4 various testing phases. Source Range (SR) Prior to and atRange to Intermediate the time (IR) of Initialoverlap channel criticality,is data taken to verify how much overlap exists between them. Data is recorded simultaneously from both SR and both IR channels as the reactor neutron flux increases during the approach to criticality. This data permits the calculation of whether or. not tho ' IR channels r begin to indicate at a sufficiently low flux level such that adequate margin exists to be able to doenergize the SR channels prior to reaching the SR reactor trip setpoint. Data is also recorded simultaneously from both IR and all four Power Range (PR) channels. This data, when combined with similar data at full power, permits the calculation of IR and PR channel overlap. During power escalation, at approximately 30%, 50%, 75% and 98% power, the % power outputs from all four PR channels are aligned to be within 11% of reactor thermal power (calorimetric power).- The 98% power execution provides additional assurance that the PR channels are properly calibrated so as not to exceed 100% power ;' when power is increased from 98% to 100%. Additional data is
- recorded for use in the IR and PR overlap calculations. The measured PR channel detector currents are extrapolated to 120%
power, the full instrument span, for use as needed for. Quadrant Power Tilt Ratio (QPTR) calculations. The measured PR channel detector currents are also plotted as a function of calorimetric . power for use in verification of detector linearity. At approximately full power, final data of the type taken during t power ascension is recorded. The PR channels are aligned to be. within 11% of calorimetric power. The full power currents are combined with those during power ascension to verify PR detector linearity. The full power IR and PR current data is evaluated to , demonstrate adequate IR and PR overlap, such that adequate margin exists to be able to block the IR reactor trip. The IR data is extrapolated to 100% power and this calculated 100% power value is used to compute the IR high level rod stop and high level trip setpoints and reset values.
-165-l . . _ _ _ _ .._ _ _. ____ _ _ _ . _ _ -. _ -. ___ J
2 5. 4 - OPERATIONAL ALTGNMENT OF NUCMAR INSTRUMENTATION - ISU-204A (Continued) TEST METHODOLOGY (Continued) After Shutdown from Power Operations of at least 800 MWD /MTU, the operating high voltages and discriminator bias voltages for the SR channels and the compensating voltages for the IR channels are determined and set. l Prior to core loading the SR channel high voltages and discriminator bias voltages were set using neutron sources to produce detector currents. This test has these voltages readjusted to properly correspond to actual reactor neutron and gamma spectra producing the detector currents. The IR detector consists of two concentric detector volumes, one sensitive to neutrons and gamma rays and one sensitive only to gammas. The current outputs from the two detector volumes are placed in opposition to one another, i.e., bucked against each~ other, such that the current signal components that are proportional to gammas cancel and the net current corresponds only to neutrons. To compensate for size, geometry and efficiency ' differences between the two detector volumes a bias current is also applied between the two volt ses. The IR compensating voltages, l which provide these proper bias currents, are initially set to -40 volts to ensure complete elimination of the gamma signal, even at the price of losing a portion- of the neutron signal. This ensures ' that the channel output is forced-low enough following a reactor trip to permit the SR chanr.els to automatically reenergize. If improperly set, the large gamma signal present following a trip could cause the IR channel output to remain abnormally high for an extended period of time and prevent automatic reenergization of the SR channels. The compensating voltages are set using actual , reactor neutron and gamma spectra as detector inputs ~ to ensure proper screening of the gamma signal.
SUMMARY
OF RESULTS -j A minimum overlap of 1.5 decades was observed on all SR/IR and IR/PR channel combinations. Specifically, the _ ovsrlaps for the four SR/IR channel combinations were observed to be more than 1.57 decades and the overlaps for all eight IR/PR channel combinations were observed to be more than 1.97 decades. Refer to Table 3.5.4-1 for detailed results. The high voltages for the SR channels were set to -1880 VDC. The discriminator bias voltages for the SR channels compensating were voltages set to -0.500 VDC for the IR (N31) andwere channels -0.599setVDC (N32). VDC at -23.551 The (N35) and -10.94 VDC (N36) with a core burnup of approximately 1400 ' ; MWD /MTU.
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- 3. 5. 4 - OPERATIONAL ALTCNMENT OF NUcmR INSTRUMENTATION - ISU-204A (Continued)
SUMMARY
OF RESULTS (Continued) The PR channel outputs were either found to be or were aligned to be within ilt of calorimetric power at all power plateaus and at full power. 120% power detector currents were calculated but were not actually used for QPTR calculation. The PR channels were verified to demonstrate acceptable linearity of outputs. Refer to Figure 3.5.4-1 for a plot of channel N41' summed top and bottom detector currents as a function of calorimetric power. Similar plots were made for the other three channels. Refer to Table 3.5.4-1 for detailed results. The 100% power IR channel outputs were calculated and the IR rod stop and trip setpoints and reset values were calculated. Refer to Table 3.5.4-1 for detailed results. Refer to Figure 3.5.4-2 for a plot of average IR detector current as a function of calorimetric power.
- Only one significant problem occurred during test performance. In the 50% power plateau test execution, it was discovered that the calorimetric power value was in error due to instrument tubing leaks associated with the precision feedwater flow instrumentation.
This precision instrumentation, which was installed specifically for calorimetric measurements, was separate from the permanently installed flow instruments. The leaks were repaired and the 50% power plateau test portion was repeated. Only the final 50% power plateau results are included in Table 3.5.4-1. The 30% power results were also slightly affected by this problem, but the uses of this data did not merit repetition of this data at 30% power. i
-167-l
L 1 Table 3.5~.4 Nuclear Instrumentation Results Summarv^ Testina Plateau-g-. g. .m- g, g Calorimetric Power (%) 29.31 .48.04 77.59i , 97.7 99.61 IR N35 output-(amps) 1.SE-4 2.2E-4 3l. 2E-4 4.SE-4 :4.2E-4 IR N36 output (amps) 1.5E 2.3E-4 3.3E-4 4.8E 4.8E PR N41 summed. current amps)- 155 .248 376: 480' 486 PR N42 sunned current- amps)- _198 320- 484 606 618 i PR N43 summed current- amps) 172 . .278- 421 531 540 d PR N44 summed current amps) 172.1. 280~ 424 534- =544 ' ' QPR N41 Power (%) -29.8 OPR 46.5 '78.0 ' 9 7 . '4 100.0-N42 Power (%) 29.8; '46.0 78.0' ' 96. 5' -100.1: OPR N43 Power (%) 30.0- 46.5 '98
*PR 78.0 100=0 1
N44 Power (%) 30.0 46.5: ;78.0' 97.6- '100.0
- Values recorded are the as found values.. The as left values were either the same as the as found values - or adjusted to be r within -11% of )
calorimetric power. I i
.i I
i I h
-168-t u
Table'3.5.4-1 Nuclear Instrumentation Results gn===rv (Continued) Source Range vs. Intermediate-Range _ overlaps Channels Ovarian (decadas) N31 vs. N35 1.57' a N31 vs. N36 1.57: N32=vs.-N35 1.6 N32 vs. N36 1.6 Intermediate Range vs. Power Range Overlaps; Channels Overlan (decademi- 1 i N35 vs. N41 1.97 mj N35 vs. N42 ;.54-2 Ti N35 vs. N43 2,08 ' N35 vs. N44 2.23: 4' N36 vs. N41 1.981 N36 vs. N42 2.51 N36 vs. N43 2.09: N36 vs. N44 , 2.27' - i i Intermediate. Range Currents Channel g g ;! Full Power Current (amps) -4.22E-4 4'.82E-4 d High Level Trip Setpoint (amps) 1.06E-4 1.21E High Level Trip Technical Specification Limit (amps)_ 1.33E-4 -1.52E-4 j High Level Rod Stop'(amps) 0.84E-4' 0;96E-4. NOTE: ~ The IR High Level' Trip Setpoint-is the current equi'/alent' to.525% of full power,.the< Rod'Stop.is at'$20% of_ full = power and the Technical Specification Limitt im> at is31'.5% of full power. The reset values are nominally calculated , to be 1/2-of:the actuation values. t 4 h
>k
-169- ,
i r
? ,
A-- 'i
l l Figure 3.5.4-1. Power Range Current-vs. Calorimetric Power-l
- 600 '
O' oPOWER- RANG 5 450
,/
h400 - w 350
/g '
x m 300 v 250 o-m 200
/
1 E-D 150 o/ ' w .I e 100 / E m '
/ 1 '
50 .-- -
/
$ 0 .
l Q- 0 10 20- 30 40 50- 60: 70 .80- 90~ 100 3 CALORIMETRIC POWER (D 1 l i
?
a 1
-170-4 .m_
1 q-i li
; Figure 3;5.4 . l Intermediate Range Current vs.. Calorimetric '.)wer i j i ? 5.0 :l oI NTERMEDIRTE RANGE > @4.5 m 4.0 o.
1 53.5
* /
g _
$3.0 7 5 /
g2.5 g w 2.0 e f 2 *' - 1.5 w :i r jio0.5 y q E0.0 >
~
0 10 20 30 40 = 50 60- 70. 80 90 100-CRLORIMETRIC POWER C D <! L l I l
-171- ,
.I {
i i
- 3. 5. 5 - INCORE/EXCORE DETECTOR CAT TBRATION - NUC-2 03 OBJECTIVE q This permanent plant procedure is performed to assure that a1 11near- A relationship exists between the excoren neutron currents and the -
9 incore Axial Flux Difference (AFD) .. Once established, this excore-current /AFD relationship is used to perform various calibrations of l the-excore_ channels ~, the OTN16 AFD inputs,= Axial-' Flux Differencet . indications and plant process computer- inputs. This; procedure-partially satisfies activities described - by - FSAR Table .14. 2-3, Sheet 22-and Technical Specification 3/4.3.1.1. TEST METHODOI/lft1 > For the 50% power execution of this procedure, a base case' full F core flux map is taken at stable core conditions.- A small reactor T coolant dilution is made and Control Bank D is. inserted' 15 to '25: 1 steps to compensate for this reactivity change,.-with reactor power- i held constant. The effect is to push neutron flux toward the < bottom of the-core which makes AFD more negative. . With AFD more =l negative than the base ' case, a quarter core flux map is taken. The-reactor coolant is then borated to' restore Control Bank-D'to its -; original position. The small magnitude..(approximately; 5% AFD) and short duration (approximately 1 hour) of this AFD change does not result in any significant residual Xenon: transient. effects on the core. The following data. is- taken' during_ both ' flux maps:. calorimetric power, excore _ nuclear d e t e c t o r - c u r r e n t s - a n d .' m a i n , control board and P2500 process computer AFD and Axial' Offset-indications. The flux map axial power' distribution ~1(top' half of core vs. lower half of core) results,are. combined with;the other-data to compute the proper calibration' constants; The incore flux map results are assumed to represent'the; correct ' axial power distribution and are used as' the basis for all . AFD indications. However, the. axialL power distribution, AFD, indications are based on outputs from the excorei power o range t detectors. Due to changes in-core radial' power distributions the flux that the excore detectors see may not accurately. represent the actual core averaged conditions. To .compensatec for this, the calibrations of the excore detector based- AFD indications are based . on incore flux map results. This procedure plots' actual-excore detector outputs as a function of' incore flux map Aq. :4q is the-relative top vs. bottom'incore power distribution while AFD is the excore detector measured top vs. bottom. core power distribution.- l When the channels are calibrated, Aq = - AFD.- The resulting: plots allow calculation of the excore currents that would be expected to be present if the core were to be at selected incore Aq values. These selected incore Aq values would be the calibration points. These values are supplied to Instrumentation _and controls:for use
-172-b I
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r
~
3.5.5 - INCORE/EXCORE DETECTOR CALIBRATION - NUC-203 (Continued)- TEST METHODOLOGY (Continued) with their normal plant calibration procedures. - They inject the j specified current signals into the power range circuitry inputs.and- i adjust the outputs and indications to correspond to theiselected' incore' A q. The direct relationship between P2500 process computer Axial- offset and incore flux map Axial offset .is calculated ~ as the slope ;of th e curve for the plot of the P2500 values as alfunction~of the incoro , i flux map _ values. These slopes are input. to the. P2500 process computer as conversion constants to convert the excore computer l inputs to correspond to incore Aq values used for reactor! J monitoring. , The results of the excore current vs.' incore AFD plots are . also: used to calculate the full span (120% power)' currents'that would' l exist at the 0% AFD condition for use in Quadrant Power Tilt Ratio j (QPTR) calculations. For the 75% power execution of' this - procedure,; ~ an , axial , Xenon > i oscillation is created by.a significant insertion of Control-Bank D (up to 40 steps) in response to a reactor-coolant d lution,
- holding this inserted position for-approximately two hours.{and then-:
borating the reactor. coolant to restore Control - Bank' D to D its starting position. While Control. Bank D was deeply ' inserted, Xenon is preferentially depleted and Iodine preferentially produced in the lower half of the core. When Control Bank ~D is withdrawn the neutron flux shifts toward the top of the core over time as the Iodine decays to Xenon in the_ lower half < of' the core. : While the flux moves upward,.with the corresponding. positive change in AFD, numerous quarter core flux maps are- taken.1 Full _ core flux maps'are taken prior to the control Bank D insertion and with Control Bank _ , D at its maximum inserted position. 'The same. plant data is-taken- 1 during the flux maps as was done; cat 50%' power._ The 'same calculations are also performed'asJat';50%-power.; The 75%1 power results are generally expected toibe more accurate,than those at < 50% power due to a reduction of;'the adverse.: temperature redistribution effects with increasing core . . delta ; temperature. Following completion of data acquisition, the axial Xenon transient
- is suppressed using permanent plant procedure NUC-118, " Xenon '
oscillation Dampening". For the 100% power execution of this procedure, a_ full. core flux- - map is performed at stable core conditions and the same. plant data is taken as done at lower power levels. . A comparison is made L '! L
-173- t
. . . - , ,. , - , , - - - ,,,-+--e..y9
- . - _ . - . - -- - .. .. ., -- .. ~ . .
i
-j r
i 3.5.5 - INCORE/EXCORE DETECTOR CALTBRATION
- NUC-203 (Continued)
TEST METHODOIAGY (Continued) l between the indicated-AFD values =and the-incore flux map'AFD. If-l the comparison'is satisfactory, no adjustments are made to the- AFD circuitry. If .the comparison is ' not satisfactory, : the AFD' . , circuitry would be recalibrated based on a-combination of 75% power and the 100% power data or by generation ofia much smaller axial Xenon transient at 100% power with data acquisition performed as at 75%-power, i
SUMMARY
OF RESULTS i- -1 The excore detector. data and incore. flux map results were -' successfully used to calculate the calibration parameters 7for the OTN16 inputs, P2500 process. computer- inputs and Axial Flux' Difference indications at both ' the 150% ' : and 75% = reactor power > > levels. The 100% power -results were satisfactory.'with no recalibrations required. During the 50% power test, performed at approximately-47% power, two retests were performed. . The' Aq values . for the full core flux map and. the quarter core flux map.were not sufficiently far enough ' apart to yield reliable results. A third flux map,was taken.as Retest #1 and used in conjunction ~ with the' first map, the full core i map. The full core flux map had an-indicated extrapolated-incore : Aq of -3.09%, the first quarter core map hadx-5.6%. and the Retest i
#1 quarter core map -25.4%. Retest > '#2 did ' not ' involve the !
acquisition of new data but was performed to repeat'P2500 process computer input calculations due to a software methodology change. The calculations were revised to correspond to..the new methodology before Retest #2 was performed. This:50% power. test' performance ensured that the AFD circuitry was - properly Lcalibrated prior to exceeding 50% power, where Technicalispecification 3/4.2~.1 first applies. During the 75% power test, performed 'at approximately 77.5% power,- l ten flux maps were obtained ranging from an incore Aq~ of,-22.6% to ! +9.5%. No retests were performed. This test performance was also. used to satisfy Technical' Specification Surveillance-Requirement 4.3.1.7. During the 100% power test, the differences.between:the excore AFD values and the incore flux map . Aq were all less than the allowed L maximum of 3%.and, when' statistically combined using the' square L root of the sum of the squares method, the difference was less than I the . allowed maximum. of- 8%. Therefore,. ~ no instrumentation adjustments were necessary. No retests were performed. .-This test performance was also used to satisfy. Technical Specification Surveillance Requirement 4.3.1.1.2a. 1
-174-a
-- ,r , #, r +-9 "N* f f* '*'- t
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l l
.1 3.5.5 - INCORE/EXCORE DETECTOR CAYTBRATION - NUC-203 (Continued)
'1
SUMMARY
OF RESULTS (Continued)- l Refer to Table 3.5.5-1 for detailed test- results. . ~ Refer to Figures 3.5.5-1 through 3.5.5-3 for example. plots of AFD and Control Bank C position during the 75% power ' axial Xenon transient and for - example results for-Power Range Channel N41. ' t I 1 l 1 s k p I b l . l
-175- ..I l l l
1
. - . . - . . - . - ~ . . . . .- - . . . . ,
1 i
- Table 3.5.5 -Incore/Excore Detector Calibration Su===ry- ]
i (Incorefjlg valuefin %)-
~
50% Power. ) Slope- 0 ' of Incore-NIS Upper Detector Lower Detector 'vs. Excore Channel Current (Atamns) r Current (LAmanni Axial Offset r-N-41 (2.90384 x Incore A q) (-2.1716'x Incore &q) 1.0080'-
+310.42 +326.56 N-42 (4.0111 x Incore A q) (-2.3289 x Incore dq)L 1.0317:
+433.01 +388.18 N-43 (3.2495 x Incore Aq)f(-2.3233 x Incore Aq) 1.0166 j
+354.36=
+352.14 y N-44 (3.3056 it Incore d q)) (-2.3481 x .Incore Aq)' 1.0095' '
S
+354.58 +358.07' l 75% Power (Incore dkq value in %); Slope of Incore NIS Upper Detector. Lower Detector v,. s Excore-Channel Currentfu anna)- Current ( ta mans) Axial' Offset-I I N-41 (2. 3463 x Incore l- A q)' (-1.9417 ' x c Incore Aq) 1.1248
+304.78-
+283.63 ,
N (3. 2688 x -Incore ' A q) (-2.0720L x Incore A q) 1.1682-
+397.03 +362.811 N-43 (2.6572 x Incore A q) (-2.0290 x Incoredhq) -1.1414
+322.10 +331'44 . :
N-44 (2.7179 x Incore A q). (-2.0759 x :Incore 4q)' 11.1270 i
+325.71 +335.08 100% Powgr NIS Channel Excore AFD (%)- Incore Ouu(%) Difference (%)
i N41 -7.56 ' -8.535 0.975 L N42 -7.14 - -8,535 1.395 N43 -7.31 -8.535 1.225 d N44 -7.35 -8.535 1.185 1
. Square root of;the sum of ! the 2.409%-
- Squares '
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1 , . , Figure-3.5.5-1_<,. Incore/Excore Calibration - Plot of AFD vs. Time l 1 1 e j l l l 10 . . - 8 PLOT 3F 9FO VS TI ME- i o a 4 [i 2 -- - - - - - ---
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u .j i l Figure 3.5.'5 Incore/Excore Calibration:- Control Bank'D Positionsys.; Time . i 1 l l 1 1 i ! 230 l 225 l 220 cc,tini scre :OP:miti on; 4' [- 215 ! 210 1 F' ' ' n 205 if 200 : w En'196 4 . w , l g 190 w , p 185 ( 1 c \ ' 180
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tr 180 155 il50 . . 0 2 4 8 10 12 14 16 18 . 204 22 241 26' .l HOLRS I! i l l t
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Figure 3;5;5-3 .. Example Axial'. Flux Difforence:-Calibrationi l Channel N41.. i t l I' e 1 i c] INCORE: DELTR Q:VS. EXCORE CURRENTS R 360 h 'a- q BOTTCMEETECTOR lCFIAMEL W 41 6- {340 x N 0 sa' N >
-u
- 320 N ,
I 6-3 > v 5 % O- , g N ,,.o ' l H 300 -a s N W s.a-o' E-8280 ' ,
./01 D. i o . TOP'DETECTOF
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- INCORE OELTR' G' % 1 I
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l L 3.5.6 - LOOSE PARTS MONITORING BASELINE DATA ISU-211A- ! l OIkTECTIVE -l L This test is performed to gather' noise frequency response:dataJin-l Modes 5 and 3 and at approximately 0%, 50%, 754 and 100%' reactor l power. This data is used-as-a reference-baseline for-analyzing-suspected loose parts in the NSSS and to verify proper _ alarm levels - and noise filter settings. TEST METHODOLOGY At each specified plant condition, a background noise recordingLof. each of the 20 loose parts accelerometer channels is made using the-permanently installed recording equipment. The recordedLdataHis' i then played back through an oscilloscope to verify adequate signal ! l quality.
SUMMARY
OF RESULTS A comprehensive summary of the results of' Loose Parts Monitoring-System testing ~ including preoperational ' impact ' testing l results, . filter settings and. background noise . spectra" is-..provided~_in At.tachment A pursuant to Regulatory Guide..l.133. requirements _for ' loose parts monitoring system testing. Baseline. data was satisfactorily collected in this test' to provide a baseline for each of the 20 accelerometer = channels.- ~ Testing:was. performed without incident except for - several' minor problems. as , follows:
.l During the initial test performance.in Mode 5',cthe-output signals- il were distorted and erratic. The recorder? heads were cleaned,and.
demagnetized and the data was retaken. Also, the' alarms associated . with module LPM-8. would not clear. The: sensor?cablerand line
~
driver associated with LPM-8 were replaced:and the - channel was successfully ratested. l At 30% power, one of the two installed recorders'wasLnoted to'be making excessive noise. .The recorder was.Lrepaired and testing. resumed. It was noted that the output from-accelerometer.#4~was - sporadically erratic. The - output . was not ~ erratic' during ' the recording of the output signal from tl41s accelerometer. The: cable connectors-for-this arcelerometer were cleaned and' tightened ~and' the problem did not recur. q i -180- __________I_. ___________________1______._____._______.___.m - W
i J L 3.5.7 - STARTUP AELTUSTMENTS OF REACTOR CONTROL SYSTEMS - ISU-020A : OBJECTIVE , This test is performed to determine tho' average: RCS temperature' (Tavg) value which results in establishment of:the design steam
]
l pressure at full load within the temperature limits for the maximum. allowable Tavg.- This is accomplished by_ making: adjustments to the; reference Tavg (Tref) program and rescaling the turbine impulse pressure instrumentation, as necessary.- Pressurk.or level Lis also verified to correspond to the proper programmed value as a: function j of power. l TEST METHODOLOGY "i At approximately: 30%, 50%, 75%, 90%'and 100%l power, plant-data is taken for use in evaluation and-extrapolation of the Tref program and turbine impulse pressure (Pimp) program value. This data ) consists of calorimetric power, _Tavg values, pressurizer level. and- . level setpoints, Tref, steam pressures, . turbine impulse pressures,- , main generator electrical output and'feedwater flows. = Data'is also- t taken to verify proper response of- the pre'ssurizer. level' controll . program as a function of power. A change in Tavg results in a change to average ' steam generator . saturation temperature which directly ~ affects-_- steam generator saturation pressure. The. rod- control system automatically-functions to maintain Tavg at,.or very close to, Tref. The-value-
~
of Tref increases as a programmed function'of power. In order to optimize steam pressure,-the test. evaluates Tavg, Tref and steam generator ~ pressure and calculates :what change in Tref would be necessary to alter Tavg by : the proper. amount -to result in the optimal steam generator pressure.- The optimal steam generator pressure is assumed to be the> full power design l pressure of 1000 psia. There is an upper limit' on . Tref f of . 589.2*F, the highest Tref value assumed in the Laccident analyses. Full. power? Tref is l initially set to 589.2'F. The . power input to : the Tref progrom comes from turbine impulse pressure. Pimp increases linearly with turbine-generator output and the predicted. values'may; require-rescaling to correspond to
- actual impulse pressures. Pimp' data is.taken and' extrapolated to l full power' for comparison- with the vendor ' supplied Pimp l
predictions. Any-significant deviations in the Pimp program would !, have to be. corrected by recalibration of-the. Pimp channel or taken. . I into account in the calculation'of1the new Tref program. These Tref and Pisp extrapolations are made at'75%, 90% and;100% power only.
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- 3. 5. 7 - STARTUP htLTUSTMENTS OF REACTOR CONTROL' SYSTEMS - ISU-020A' i (Continued),
TEST METHODOLOGY (Continued), 1 l Actual pressurizer ' levels are compared to' calculated program levels R l based on the actual power levels. - The level control. setpoint values are thus also verified to be proper. - This comparison: is ' made at 75%, 90% and 100% power only.- , o l Refer to Figure 3. 5. 7 for . example plots of- Tref, ~ Pimp L and ] Pressurizer Level as functions of. power _and Figure 3.5.7-2 for an 'j example plot of steam pressure vs'. power. ,
SUMMARY
OF RESULTS At the 75% power. plateau, Tref was extrapolated.to.a full-power value of 587.12'F which was verified to be below the design maximum , of 589. 2'F. . Steam generator pressure-was extrapolated toL1017.S' psia which ~ did not match the~ design. range of 110001110 ? psia.- , Turbine impulse chamber pressure was'extrapolatedito;865: psia at 1 100% power. This compared favorably- with the = 880L psia : vendor ' supplied prediction. . I.t 90% power, Tref extrapolated - - to 587. 24*F. 'This was again .* verified to be- below the -design. maximum of ' 589. 2'F. : Steam generator pressure was extrapolated to!.1016.3 psia-which was also- -
- - not within the design range 'of 1000 ilotpsia. Turbine impulse -
chamber pressure was extrapolated to 893 psig at 100%-power.: This i l also compared favorably 1 with the 880 -psia- vendor- supplied prediction. .j At the 100% power plateau,. Tref extrapolatedt to 587. 3'F. This was again verified to be below the, design ' maximum; of - 589.2'F; - Steam generator pressure was extrapolated to 1015.0;paia which.was not a within the range of 1000 i10 psia.- Therefore, an-adjustment of , approximately -1.9'F to the Tref. program value was required to red.uce the steam generator. pressures by approximately 15 psi. The-full power extrapolated turbine: impulse pressure-of 907 psia.'.was higher than the 880 psia vendor - supplied; prediction. This~ 3% difference did not adversely affect the- Tref program use of. the , Pimp signal because the output' clips. at= 100%'~ Pimp power. i
~
Therefore, even though Pimp _ power would oindicate 'as c103% when actual power was at 100%, the. Tref output would be the 100% value. This 3% error was considered . insignificant and' the full power impulse pressure adjustments were not-performed, j I
'1
-182- 1
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3.5.7 - STARTUP AELTUSTMENTS OF REACTOR CONTROL SYSTEMS- - ISU-020A (Continued)-.
SUMMARY
OF RESULTS (Continued). l Pressurizer level was found to deviate from the calculated program l value by -0.2% at 75% power, by -0.61% at 90% power and by -1.07%
- .at 100% power. . This satisfied the $ 13% ' allowed' deviation criterion.
Refer to-Table 3.5.7-1 for detailed results. 2l l l ! I t i l
-)
-183- ,f I
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.Startum Adiustments Su===ry i
i 101 101 ni 101: 10.01 1
. 1 Calorimetric Power (%)- 29.30 48.14 76.57 89.'36- . 99.77 I Tref ('F) 564.1. -572.9: 581.2; 586.2- 589.5) ]
-l Tavg ('F) 564.3 572~21 . '581.37 584.5 t 588.6; J Pressurizar. Level _(%) 32.90- 41'96' 151.64"
. 55.50-l60.46 Calculated . . . .
t 1 Pressurizer Level (%) N/A .N/A- -51'.44 l54.89 : 59.39: ) Actual Pressurizar < Level setpoint (%) R33.04 42.01- 50.89! -56.05; -60.40_ -
-)
Average impulse . . . _ l pressure (psia) 198.1 ~438.9 66263 .798.'0: 905.0 Average steam generator ' pressure (psia) 1068.0 1070.0 -1044.4' 1020.7 1615.2" ,, 3 100% Extrapolated steam . . generator pressure (psia) N/A :N/A 1017.5 . '1016.3' 1015.2' q Gross Electric Output (MWe) 230 1520 921. 1128~ 1155L l l 1
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Figuro 3.5.7-1 + iPROGRAMMEDEPRESSURIZER LEVEL TURBINE IMPULSE PRESSURE im iwo 99 vxm mwn emm 999 m s. 8. /
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l 1 1 Figure 3.5.7-2 ' Steam Pressure vs. Power ,
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-186-4
i i c 3.5.8 - FULL POWER' PERFORMANCE ~ TEST - ISU-281AV l l OBJECTIVE This test is performed to. demonstrate - the ; reliability of the . a i Nuclear Steam Supply System '(NSSS) to maintain its warranted output; 1 of 3425 MWth (+0, -5%) for 100 hours without a load reduction or i plant trip resulting from an NSSS malfunction,:to demonstrate the ability of the plant to generate - 1150 MWe - (+0, -5%). for .100 consecutive hours and to demonstrate theLability of--the NSSS to develop 3425 MWth (+0, -1%) at a steam generator pressure of 1990 psia. TEST METHODOIDGY The test is initiated with the plant operatingt within 'ii%)of' its rated NSSS output as determined l by power range: . nuclear. instrumentation which is calibrated to, correspond to calorimetric j power and at a steam pressure of 2990 psia., Plant conditions are then stabilized at their' design values for 35 to 65 hours'with stability verifi~ed by calorimetric data' acquisition. -Power is l verified to be or is increased to be within 1% of the 3425 MWth design -NSSS output and a four hour performance measurement is, initiated to verify. actual NSSS power;by-collecting calorimetric data every 5 minutes and computing; power-hourly. The remaining < hours of the 100 hour, minimum duration.are then completed._
SUMMARY
OF RESULTS I i The 100 hour run was. started at 2100 hours ~on 7-17-90 'and completed 3 at 0100 hours on< 7-22-90. . However,- plantL conditions were .! maintained and the. test was continued until 0900'nours'on 7-23-90. This resulted in a total' documented- run of - 132' hours.- The first 48 hours of the run occurred prior to the formal start of-the test. ], ( Credit was taken for these 48 hours based on Reactor Operator logs and data acquisition system records. The minimum and maximum hourly NSSS power measurements were 99.11% :and 100.58% of the' rated ; power of 3425 MWth over the entire 132 hour duration.- - There was no
- load reduction or plant trip.during the'132 hour duration.
A net electrical output of between 1093.4 and 1108.6 MWe (95.08% to 96.40% of 1150 MWe) was demonstratedH over the - entire 132 hour duration. This satisfied'the.1150 MWe (+0, -5%) criterion. The average output 'over the 100 hour duration was 1102.6 MWe, 95.88% of- -) 1150 MWe. The four hourly calorimetrics ' demonstrated NSSS - output to be between 3415.9 and 3420.3 MWth, corresponding to a range of 99.73% to 99.86% power, at a steam generator outlet-pressure of'2990.42 1 psia. This satisfied the 3425 MWth ' (+0, -1%) at 1990 - psia L criterion.
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l
- l I
1 1 3.5.9 - P2500 PROCESS COMPUTER' SOFTWARE VERIFICATION - ISU-019A~ _ OBJECTIVE j The P2500 Process Computer Software Verification . test 'is pedorned' to verify that the process' computer receives, correct inputs from selected-process-variables in thecfield and to validate selected- 1 performance calculations performed'by the process computer. This j verification and validation is performed by: comparing-the output j from the process computer to permanentiplant instrumentation and 1 the Test Data ~ Acquisition System (TDAS) computer output. LI TEST METHODOLOGY At plant power levels- of 0%, 50%, 75% and 100%,- selected analog land 1 digital plant' parameters monitored ' byt the: ~ P2500 Westinghouse, q process computer; are' compared to Main : ControlL Board instruments ! and/or outputs from the TDAS'ccmputer tol ensure they agree within. I specified tolerances. The parameters selected were those judged'to be most important to the operators in monitoring plantLconditions and evaluating equipment performance. The tolerances are based on the accuracy of the instrumentationi and' associated instrument-loops. The following five programs arel' thef performance calculations .; verifled by.this test: The Total Thermal Power algorithm used byj the process computer calculates the thermal output.of .each of:the' four steam; generators as well as a value for total secondary. calorimetric power. The results of these calculations are = compared = to .the . precision . calorimetric performed by the~TDAS computer; y
-(
I The Heater Differential algorithm performed by the process computer calculates the temperature difference: across each :of L the twelve < feedwater heaters based . on inputs fromi thermocouples in the l feedwater system. 'The TDAS.also. takes' temperature readingarfrom J the same locations -and calculates' the -. temperature- dif ferences'. These temperature difference' values are then compared.. The Percent Turbine Power algorithm performed J by - the process i computer calculates turbine electrical tloadE based lon a linear I function conversion of the turbine impulse pressure. The' output'is provided to a digital meter onsthe~ Main control Board. Data ' is'- recorded in this test to correlate turbine impulse- pressure L to j reactor thermal power' and -ton correlate . calorimetric. power to generator megawatts. The ' data is provided to Engineering for determination of the proper calibration' constants to be used in the P2500 process computer for this linear function conversion.- The Calibration check of Power Range ENuclear Channels algorithm performed by the process computer compares a calculated average
-l88- I
- . - - - . -~ - . - .- - -- - - -.- _- .-
i i I- l i. V . . L5 9 - P2500 PROCESS COMPUTER SOFTWARE' VERIFICATION' JISU-019A (continued) - 1 l TEST METHODOIDGY (Continued) l Nuclear Instrumentation System (NIS) power range power: output with . ; the reactor thermal power calculated in therTotal. Thermal: Power l algorithm. This comparison is provided to appraise the operator of a possible drift' in the power range channel- outputs.. .The calorimetric power calculated by the TDAS'is-used in this. test toL l evaluate the comparison made by the P2500 process computer. 1 l The Quadrant Power Tilt Ratio algorithm-performed by.the-process computer calculates' upper, lower, and average ~ radial'fluxLtilts on a quadrant basis using inputs:from.the NIS power range detectors. . The results of this calculation are compared to the l manual. calculations . performed in accordance with the permanent' plant - surveillance procedure.
SUMMARY
OF RESULTS , During the 0% power level' test, all of the: digital' inputs to the P2500-were verified correct. Approximately one. third of the analog signals could not be ' verified at this t power ilevel for c various reasons including instruments out' of service',.-instruments off-scale. , due to plant conditions, TDAS- non-availability, instruments out of ' calibration, and incorrect scaling or processing by e the , P2500 process . computer. Instruments were calibrated ; as necessary and ' ' i retesting was performed at higher power levels. / During the' 50% power J1evel test,. nine of the computer addresses failed the instrument correlation. - Three required modification to - the computer. database to correct _the engineering ranges,- three - failed due to .the inaccuracies of the Main control Board indicators ' at the low end of their scales, two could not be tested: due 3 to unavailability of the TDAS. channels, and one had a wiring' error?in the field. The ~ remainder of the o inputs 1 passedL- the - instrument correlation. ) During the 75% power level test, ten .of . the : computer addresses-failed the instrument-correlation. .Seven needed'recalibration of the input devices, two could not be tested'due to unavailability .of l L the TDAS. channel's, and.one needed a modification.to,the1 computer database-to correct the~ engineering range.s The' remainder of;the-inputs passed the instrument correlation.- During the 100% power level. test, four.of the computer' addresses- . failed the instrument correlation. Three needed'recalibration of the input clevices and one needed a modification: to the _ computer I database to correct the engineering range. .The remainder of'the
-inputs passed the instrument correlation. '
l.
-189- j i
_ . . -.. ., ., ,__,,,,I
I J l 3.5.9 - P2500 PROCESS COMPUTER SOFTWARE VERIFICATION - ISU-019A~ (Continued)
SUMMARY
OF RESULTS (Continued). The comparisons of the process computer's performance calculations l at the various power levels are listed in-Table 3.5.9-1. . The .i values of % Mean Deviation listed'are calculated from the following equation: 4 Mean Deviation = connarison Value - P2500 value X 100
, Comparison.Value + P2500-Value -
l The recorded results values have been rounded off. The % Mean. ! Deviation values were calculated--using the-not rounded raw values. , During the 50% power test, the', Total Thermal Power algorithm was satisfactory. The Heater Difference algorithm wasynot successful because the feedwater= heater thermocouples were-incorrectly wired i to the process: computer. The thermocouple wiring was corrected and ; the process computer. was detormined 1 to have J been J calculating. correctly but had been receiving-incorrect inputs. . The, Power Range; I Nuclear Channels Calibration algorithm was successfully' tested at ; 50% power. The data was.. collected for the.PercentLTurbine Power.
- Meter for transmittalLto Engineering.
During the 75% power test, the Total Thermal- Power' algorithm was satisfactory. The Heater Difference algorithm was not successful because the feedwater heater system was not in the normal operating . ', configuration and one channel of TDAS-was outnof service.; This testing was deferred'to a higher power level.. .ThaiPower Range ; Nuclear Channels Calibration algorithm was successfully tested at: a 75% power. The Quadrant Power Tilt Ratio algorithm was tested.for l the first time at 75% power and was satisfactory; Additional data- - was collected for the Percent Turbine Power Meter for transmittal to Engineering. 3 During the 100% power test, the Total Thermd Power elgorithm passed the review criterion. However. a difrerence of 98.7 MWth existed between the P2500 process comy:ute2.'and the TDAS- precision calorimetric. The major contributor co this c differencc.~ ;was an error in the feedwater temperature input-to theLP2500 program that resulted . -in : readings that were . approximately 15'F - high .~ at 100% power. 'This was caused by errors'in the-linear approximation.of- > the thermocouple curve used in the analog - circuits. The Heater l-Difference algorithm was successfully verified at 100% power. The, temperature differences across feedwater heaters l 2A and 6B were not within the review criteria but were accurate to within 24,- and also 2 F, which was judged acceptable. The Power Range Nuclear Channels Calibration and Quadrant Power Tilt Ratio . algorithms were successfully verified. The data for the ~ Percent -Turbine Power Meter was transmitted to Engineering for calculation of the calibration constants.
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i 1 l I TABLE'3.5.9-1 .. PROCESS COMPUTER AIt,ORITHM COMPARISONS TOTAL THERMAL POWER The Review Criterion for % Mean Deviationsisfs2% Power P2500 Process % Mean - Level Comouter(MWth) TDAS(MWth) Deviation 50% 1621.8 1627.0 l0.16 75% 2628.8 2662.1- 0.63 100% 3318.7 3417.4 -1.45-HEATER DIFFERENCE i The Review Criterion for % Mean Deviationiis $1%l 100% Power Temnerature Diffarences: Feedwater P2500 ProcessL '
. .%'Mean '
Heater Comouter('F) TDAS,(PF) Deviation-1A 41.7 41.3 't 0.45' ! 1B 41.5- 42.0 0.59 2A 42.8 -44.1 1.6 2B 41.7 42.2: 0.67 3A 81.1 80.3 0.44'- 3B 82.4 81.9' =0.29 4A 54.0 53.9' O.07; 4B 54.6 54.3' O.28 5A 69.0 69.'O .0.05 - 5B 69.4 69 ~. 6 L 0.11 l 6A 38.0 38.1 0.14 5 6B 23.7 24'3! 1.1 NOTES: 1) Thermocouples were incorrectly wired'for 50% power. test ' and the data was inconclusive.
- 2) The: Feedwater 2 Heater' System ' was not -in its normal- .
configuration. at' 75% power,and the. data was not representative of actual plant conditions.
- 3) The % pean deviations . for Feedwater Heaters 2A and 6B exceeded the 1% Review Criterion'for 100% power. It was determined that the process L computer: calculations for I I
such low temperature differences was: acceptable.
- 4) The temperature difference values recorded-~were rounded. d off. The!%:Mean Deviation values-were calculated based I on not rounded values.-
-191-i L _ - - __ - . . _ -- , , - , - - - ,r--
l L , 1
. TABLE 3 '. 5. 9-1" PROCESS COMPUTER ALGORITHM COMPARISONS (Continued)- 7 l
i CALIBRATION CHECK OF POWER RANGE NUCLEAR CHANNELS , 1 l The Review Criterion .is % Mean Deviation ~<2% QB ; Actua'l -% Power Difference < ,2.5% (for 50%,'75% results)!and 4 Power Difference d
<1.5%-(for 100% results)
Power Level P2500 (NIS) TDAS Calorimetric Mean Deviation-
,50% 46.8% 47.74-- 1% ,
'75% 77.7% 78.05%- 0.22% < l 100% 99.7% 99.93%- 0.11% i OUADRANT POWER TILT RATIO ;
J The Review Criterion for % Mean Deviation ils $2% < 75% Power P2500 Process Hand: %'Mean-UDoer Radial Tilt Computer . Calculation Deviation N41 0.995 -0.998 0.15-N42 1.002 0.996 0.30 N43 1.001 0.996, 0.25-N44 1.002 1.006- 0.20 ? P2500 Process Hand % Mean j Lower Radial Tilt Computer .Calculationi ,: Deviation 4 N41~ 0.999 0.997- 0.10 N42 'O.996 0.994. 0.10 , N43 1.003 :1.008 0.'25 1 N44 1.002 0.997, 0.25 :l 1 100% Power < P2500-Process Hand % Mean a Uocer Radial Tilt Comouter Calculationf Deviation N41 1.005 11.0017- 10.16 N42 0.995 0.9952 0.01 ) L N43 0.999 0.9990 0.0 N44 1.001 1.0041- -0.15 P2500 Process ' Hand 4 Mean < Lower Radial Tilt Computer Calculation Deviation l N41 1.007 1.0034 0.18 l N42 0.991 0.9896- 0.07 N43 1.001 1.0081 0.35 N44 1.002 0'.9989' O.15
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L l; T , I
_ _ ___ _ _ . . _ __ _ _ _ . . . _ _ _~ _ _ _ _ _ _ __ i H 1 3.5.10 - AUTOMATIC REACTOR CONTROL SYSTEM TEST ISU-203A-OBJECTIVE l l This procedure is performed to demonstrate.the capability.of:the L automatic reactor control system to maintain Reactor Coolant System
- average temperature (Tavg) within an acceptable tolerance about the-
!. reference Tavg (Tref) under steady state and transient conditions. Tref is the programmed Tavg setpoint.as:a' function of power. This procedure satisfies activities described . by, FSAR Table 14.2-3,' Sheets 4 and 33.
*l TEST METHODOLOGY l
With reactor power stabilized at approximately 50%-and Tavg matched-. ! to Tref, rod control is placed in automatic to monitor Tavg-for D oscillations. After approximately ten: minutes, TavgLisimanually
'I increased to approximately. S'T- higher 1 than.g Tref: by: manual.
withdrawal of Control Bank D.' Rod- control' is ( then, placed.in' automatic and Tavg is allowed to return to and< stabilize within-approximately 11.5'F of Tref by. automatically controlled. Control. Bank D notion.- After-Tavg.has-stabilized, rod control is again' placed in manual to' decrease Tavg to- approximately 5'F Elower than Tref by manual insertion of Control Bank D.- Rod control.is then-again placed back in automatic and Tavg again;allowe'd'to return to and stabilize within approximately 11.5'F of Tref. 1Various plant? , parameters and instrumentation signals within the automatic reactor" l control loops are monitored on strip chart: recorders'during these; temperature transients. Values recorded are Tavg, Tref,. nuclear. , flux, turbine power, turbine impulse pressure,: steam'- header: pressure, pressurizer pressure and rod control' mismatch.and error signals.
SUMMARY
OF RESULTS During steady state uperation, it' was found thati Tavg- was maintained within il.5*F of Tref with no problems. :When:Tavg was increased by 5'F, it took approximately(79 seconds >to return Tavg I to within 11.5 F of Tref. When Tavg was decreased;byj5'F,11t: took approximately 68< seconds to return Tavg to within: 1. 5'F of Tref. l
-193-
1 Id DEFERRED PREOPERATIONAL TESTING , 3.6.1 - PROCESS SAMPLING SYSTEM - ISU-028A-OBJECTIVE ;
'i The Process Sampling System test is performed to demonstrate-the capability of the sampling system to provide-liquid and gas samples j
j through the - correct flow' paths- from; the = primary and secondary- j systems, to demonstrate the adequacy oftplant sampling procedures j and to verify sample- line holdup times.- The test tverifies j acceptable flow rates at design temperatures and pressures' and- l verifies the operability of automatic-on-line analyzers:and sample: coolers. This test satisfies: activities described by'FSAR= Table. q 14.2-2, Sheets 6 and 6a, and the deferred preoperational testing lin _{ System Test Matrix 1-2200. '
)
TEST METHODOLOGY j
-i The operability of the sampling: system;is ' demonstrated by. obtaining-samples from the primary and= secondary systems and: measuring-the; j sample flow rates, pressures, - and . temperatures. > The on-lins--
analyzers are compared to grab sample-analysesi Plant chemistry procedures are used- to obtain- samples .and are thus : verified'- adequate. The reactor coolant hot leg sample lines.are purged.at i the maximum flow rate and -verified to..be delayed at least 60 seconds inside the missile' barrier.
SUMMARY
OF RESULTS-Refer to Table 3.6.1-1 for detailed test results. Safety Injection Accumulators #2 and-:#3 i sample flow rates - were initially too low. The sample lines wsrefflushed and retested with i satisfactory-results. Reactor Coolant Hot Leg- Loops #1 and.#4 sample flow . rates .were initially too high, resulting in holdupitimes inside.the missile barrier of less than 60 seconds. A new valve control rod,Tcut to a length of 3 1/2 inches, was installed in drag valveL1PS-0252 to , provide sufficient flow resistance.- The final-flow rates,were O'.8 gpm in the purge mode and 0.73- and ' O.72 .'gpm .in the ' grab sample mode, respectively. These correlate to hold 'up . times- of ' 64.4 -seconds in the purge mode and 70.6 and 71.6' seconds in the' grab sample mode, respectively. ; The sample line from the pressurizer steam space was initially , found to be blocked. A faulty quick-disconnect fitting was ' repaired and the flow rate was determined to be satisfactory.
-194-
-l
'3.6.1 - PROCESS SAMPLING SYSTEM - ISU-028A'(Continued)
SUMMARY
-OF RESULTS (Continued) Upon completion of- this test, the . flowrates,_-pressures and temperatures were satisfactory. . Initial.. test .results indicated that the original acceptance criteria were' too restrictive. . These criteria were changed and - FSAR Amendment ; . 79 incorporated the-modified criteria. The plant chemistry sampling procedures were used to sample. each of the sample points and were demonstrated to be: satisfactory. The Steam Generator Blowdown analyzers weretvarified' operable by: comparison with analyzed grab samples for specific' conductivity,' cation conductivity and sodium. ' i 1 i i
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e TABLE 146.1-1 PROCESS BAMPLING SYSTEM SLDD8ARY GRAB SAMPLE MODE CONDITIONED SAMPLE POU[I FLONRATE(GPM) PRESSURE (PS1G) TEMPERATURE ('F) ACTUAL JtEO ' D. ACTUAL BEQ % ACTUAL REQ'D.
#1 SG Blowdown 0.63 0.4-1.0 51.5. 5-80 95 51:5
#2 SG Blowdown 0.69 0.4-1.0 52 5-80 105 5115
#3 SG Blowdown 0.69 0.4-1.0 51 5-80 114' s115
#4 SG Blowdown 0.79 0.4-1.0 49 5-80 90 s115 Downstream SG Blowdown Cation Domin 0.32 0.05-1.0 9 5-10 80 N/A Downstream SG Blowdown Mixed Bed Domin 0.29 0.05-1.0 9 5-10 80 N/A CVCS Letdown Downstream of Demineralizer 0.91 s1.0 N/A N/A N/A N/A CVCS Letdown Upstream of Demineralizer 0.77 51.0 N/A N/A N/A N/A RHR Train A 0.87 0.15-1.0 40 5-80 95 5115 RHR Train B 0.50 0.15-1.0 30 5 70 5115 S.' Accum #1 0.81 0.15-1.0 49 5-80 83 SI Accum #2 5115 0.78 0.15-1.0 60- 5-80 84 5115 SI Accum #3 0.75 0.15-1.0 54 5-80 85 5115 OI Accum #4 0.85 0.15-1.0 58 5-80 84 5115 Pzr Steam Space 0.45 0.15-1.0 62 5-80 95 5115' RCS Hot Leg #1 0.73 0.15-1.0 74 5-80 99 5115 RCS Hot Leg #4 0.72 0.15-?. 0 75 5-80 103 5115 Pzr Liquid 0.48 0.15-. 0 58 5-80 95 5115 SFP Domin #1 Inlet 0.76 0.75-1.0 N/A N/A N/A N/A SFP Domin #1 Outlet 0.77 0.75-1.", N/A N,*A : N/A N/A SFP Denin #2 Inlet 0.81 0.75-1.0 N/A- N/A N/A N/A SFP Demin #2 Outlet 0.77 0.75-1.0 N/A N/A' N/A N/A
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( Tintz 3. 6.1-1 (Continued) PROCESS SAMPLING SYSTEM
SUMMARY
_[ (Continued) CONTINUOUS PURG5 MODE CONDITIONED-SAMPLE POINT FIhWRATE(GPN) PRESSURE (PSIG) TEMPERATURE ('F) ACTUAL REQ 8 L ACTUAL RED'D. ACTUAL ML
#1 SG Blowdown 0.48 0.4-1.0 51 5-80 102 5115 42 SG Blowdown 0.64 0.4-1.0 55 5-80 101 5115 #3 SG Blowdown ,0.59 0.4-1.0 55 5-80 100. s115 #4 SG Blowdown 0.64 0.4-1.0 54.5 5-80 90 $115 Dswnstream SG Blowdown Cation Domin 0.05 51.0 9.5 5 N/A N/A Downstream SG Blowdown Mixed Bed Domin 0.05 $1.0 9.5 5-10 N/A N/A CVCS Letdown Downstream of Domineralizer 0.55 51.0 N/A N/A N/A N/A CVCS Letdown Upstream of Demineralizer 0.4 $1.0 N/A N/A N/A N/A RHR Train A 0.95 0.75-1.0 59 5-80 92 $115 '
RHR Train B 0.9 0.75-1.0 54 5-80 80 -5115 SI Accum #1 0.8 0.75-1.0 58 5-80 83 5115 SI Accum $2 0.8 0.75-1.0 45 5-80 84 5115 SI Accum #3 0.75 0.75-1.0 45 5-80 83 5115 SI Accum #4 0.9 0.75-1.0 45 5-80 84 $115 Pzr Steam Space 2.0 0.75-1.0 64 5-80 97 $115 RCS Ilot Leg #1 0.8 0.75-1.0 70 5-80 95.3 5115 RCS Hot Leg #4 0.8 0.75-1.0 66 5-80 103 - $115 Pzr Liquid 0.9 0.75-1.0 50 5-80 110 $115 i I 1
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- 3. 6. 2 - IN-PIACE ATMORPMERIC cf mUP FILTER TEST -
PRIMARY PIANT - ESP ~ EGT-7M OMECTIVE The In-place Atmospheric cleanup Filter Test is performed to demonstrate proper operation and integrity of the Primary Plant Ventilation ESF filtration units, including the High Efficiency-Charcoal Absorbers (HECA), High Etficiency Particulate Air (HEPA) filters and unit heaters. This test satisfies activities described by FSAR Table 14.2-2, Sheet 29 and the deferred preoperational testing in System Test Matrix 1-2400. TEST METHODOLOGY With one ESF train in service, air flow through each filtration unit is determined by traverse air velocity measurements or from in-line flow elements and compared to Technical specification limits. The total pressure drop across the filter housing is measured with a manometer and is also compared to Technical specification limits. The power input to each unit heater is determined by measurements of current and voltage. The rate of heat added to the air by each heater is determined by measurements of upstream and downstream wet and dry bulb temperatures and air flow measurements. The ratio of heat output to power input then determines the heater's efficiency. The penetration and bypass leakage is determined for each HEPA filtration unit by injection of dioctyl phthalate (DOP) aerosol and measuring the concentration upstream and downstream of each filter. There are two HEPA filters per unit, one referred to as the upstream filter and one referred to as the downstream filter. Each HECA filter unit is leak tested by injection of R-11 refrigerant as a tracer gas and measuring the upstream and downstream concentrations.
SUMMARY
OF RESULTS Refer to Table 3.6.2-1 for detailed test results. This test was performed in conjunction with a separate test procedure which verified filter air flow distribution and air-aerosol mixing uniformity. The initial testing of heater CPX-VAFUPK-01 resulted in a power calculation of 94.9 KW which failed the acceptance criterion of 100 15 KW. The heater was ratested over a longer duration with satisfactory results. All heaters dissipated 100 15 KW.
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- 3.6.2 - IN-PLACE ATMOEPMraTC crmDP FILTER TEST - !
l PRIMARY PIANT . ESF - EGT-751X (Continued) !
SUMMARY
.OF RESULTS (Continued) All primary plant ESF filtration units were tested and satisfied i the air flow requirement of 15,000 cfm 110% with a pressure drop ; across the combined HEPA and HECA filters of less than 8.5 inches < water gauge. ( i : The refrigerant gas (R-11) penetration and bypass leakage of each ' HECA bank was less than the required maximum of 1.0% at rated flow. The DOP penetration and bypass leakage of each HEPA bank was less [ than the required maximum of 1.0% at rated flow. ; Balancing of the HVAC air flow distributions was performed prior to ! this test to ensure adequate flow rates to the areas served by these HVAC systems. , 1 l l l f
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TARf2 3.6.2-1 IN-PLACE ATMOSPHERIC CLEANUP FILTER TEST StDORARY
)
4 l , q Required CPX- CPX- CPX- CPX- l Jtem Performance VAFUPK-01 VAFUPK-02 VAFDPE-15 VAFUPK-16 I i Air 13500-16500 15915 16216 15251 14928 I Flow (cfa) at <8.5"WC at 6.2"WC at 6.3"WC at 6.1"WC at 6.3"WC H3ater Power ) Dissipation (KW) 95-105 96.04 97.6 96.5 99 i Upstream HEPA
- Filter Pene-tration & Bypass ,
Leakage (%) <1.0 <0.025 <0.05 <0.05 <0.05 ; I Downstream HEPA ' Filter Pene- , tration & Bypass i Leakage (%) <1.0 <0.1 <0.05 <0.10 <0.05 l I HECA Leakage (%) <1.0 0.042 0.02 0.01 <0.01
+
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3.6.3 - CONTAINMENT & PENETRATION..R000$S TEMPERATURE.BURVEY , l - ISU-282A r i ! i OIL 7ECTIVE The Containment and Penetration Rooms Temperature Survey. is ; performed to verify that the Reactor Coolant pipe penetrations, air supply to Reactor Vessel Supports, Neutron Detector Well discharge
- air, containment air, Steam Generator compartment air, . pressurizer room air, CRDM shroud air, CRDM platform area air, and Feedwater 4
and Main Steam penetration rooma are maintained at or below their design temperatures when the RCS is at normal operating temperature ; and also when the RCS is at nominal full power conditions. This. l test satisfies activities described in FSAR Section 9.4. A, and the deferred preoperational testing in System Test Matrix l-3600. i TEST METHODOLOGY The concrete temperature around each Reactor Coolant System (RCS)- pipe penetration is measured with a thermocouple-when the RCS is at ' normal operating temperature in Mode 3. Temperdtures are recorded from permanent plant instrumentation for Neutron Detector Well ; exhaust air, CRDM shroud exhaust air and containment air. Local readings using thermocouples or resistance temperature detectors are recorded for containment areas, Pressurizer room, Feedwater and Main Steam penetration areas, both inside. and outside containment,. and the Reactor Vessel Support supply air. The same measurements are repeated with the reactor in operation at approximately 100% power. SUMMARLOF. RESULTS After the required plant conditions were verified:to have existed for a minimum of 24 hours, three sets of measurements were taken, ! each net at least two hours apart. The highest reading of each parameter was then compared to the acceptance : criterion. All , temperatures were within the acceptance criteria of the' test at both 100% reactor power and when in Mode 3.
\
TEST RESULTS Criterion Mode 3 100% Power i ! Concrete temperatures in each' RCS Pipe Penetration are less 14 0. 9'F 166.0'F than or equal to 200'F j containment average air temperature 92'F 105'F is less than or equal to 120'F
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i I 3.6.3 - CONTAINMENT & PENETRATION ROOMS TEMPERATURE SURVEY l 1 - Isu-282A (continued) ] 2 SUMNARY OF RESULTS (Continued)
- TEST RESULTS !
Criterion Mode 3 1004 Power Steam Generator compartment air 9 8. 2'F . 0 6. 4 'F temperatures are less than or equal to 120'F Pressuriser room temperature is 99.0'F 108.0'F less than or equal to 120'F In containment, Main Steam and 8 5. 2'F 105. 6'F Feedwater penetration area temp- . eratures are less than or equal to 120'F ' Outside containment, Main Steam 97. 2'F 10 0. 4'F and Feedwater penetration room temp- > eratures are less than or equal ! to 104'F , 1 Neutron Detector Well and reactor 14 5'F 14 3*F > vessel support area exhaust air temperature is less than or equal to 150'F CRDM Shroud Exhaust air temper- 13 3'F 131'F ature is less than or equal to 163'F : CRDM Platform area temperature 104.7'F 108. 3'F is less than or equal to 140'F .; r l Reactor Vessel Support supply air 8 3 . 4'F N/A temperatures are less than or l equal to 90'F. (Mode 3 only) _j i
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l I i 3. 6. 4 - TURBINE DRIVEN AUXILTARY FRRtafATER PUMP ACTUATION AND JtESPONSE TIME TESTS - EGT-768A and EGT-769A OIL 7ECTIVE ) The Turbine-Driven Auxiliary Feedwater Pump (TDAFP) test is performed to demonstrate the capability to deliver flow to the steam generators within the acceptable . time after an initiating signal. The main steam header isolation valves are stroked and ! verified to open within the required time. These valves supply steam to the TDAFP turbine to drive the pump. This test satisfies i activities described by FSAR Table 14.2-2, sheet 51 and the , deferred preoperational testing in system Test Matrix 1-3700. i ' l TEST METHODOLOGY The TDAFP is lined up to recirculate back to the condensate storage . Tank with its discharge isolated from the steam generators. The l pump is started by simulation of an Auxiliary Feedwater Actuation- l signal from the Train A Solid State Protection System'and the , response time is measured from the time of relay actuation to when ' the pump flow exceeds the minimum design flow of'860 gpm. The pump is shut down and placed in standby. The pump is then restarted'by < simulation of an Auxiliary Feedwater Actuation signal from the l Train B Solid State Protection System and the response time is l again measured from the time of relay actuation to when the pump flow exceeds 860 gpm. This pump response time is required to be , less than or equal to 58.0 seconds. The stroke open time of the Main steam Header Isolation Valves are recorded and varified to be between 9.0 and 11.0 seconds. , EUMMARY_OF_RESULTS . The TDAFP was started from the Train A Auxiliary Feedwater Actuation signal (Relay K641) and the time to reach a pump flow of ; greater than 860 gpm was 26.2 seconds. This satisfied the acceptance criterion of 58.0 seconds or less. The stroke open time for Main Steam Header Isolation Valva 1-HV-2452-1 was'4.0 seconds which did not satisfy the review criterion of 9.0 to 11.0 seconds. The actuator on the valve was readjusted and the valve was ratested ; resulting in a stroke open time of 9.02 seconds. The TDAFP was restarted from the Train B Auxiliary Feedwater Actuation signal (Relay K641) and the time to' reach a pump flow of greater than 860 gpm was 29.2 seconds. This satisfied the acceptance criterion of 58.0 seconds or less. The stroke time for the Main Steam Header Isolation Valve 1-HV-2452-2 was.5.9 seconds which did not satisfy the review criterion of 9.0 to 11.0 seconds. l The actuator on the valve was readjusted and the valve was ratested resulting in a stroke open time of 10.4 seconds. l
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I I i 3.6.4 - TURBINE DRIVEN AUXILIARY FEEDWATER PUMP ACTUATION AND f RESPQHEE_ TIME _ TESTS - EGT-76BA and EGT-769A (Continued) i t
SUMMARY
OF RESULTS (Continued) The.TDAFP was also ratested using the Train B Auxiliary Feedwater j Actuation signal to determine the impact on the pump response time ; of changing the steam header isolation valve stroke open time. The i pump response time was 22.84 seconds after the adjustment was Psde !' to valve 1-HV-2452-2. Thus, the steam header isolation v. Ave stroke time adjustments did not adversely af fect overall TDAFP ? response time. I f i t i l I
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- 3. 6. 5 - MSIV ISOLATION DMPONSE TIME TESTS - ECT-764 A and EGT-765A I l
l OBJECTIVE I The MSIV Isolation Response Time Tests are performed to demonstrate ; that the Main Steam Isolation Valves (MSIVs)- . close within the : maximum allowed time upon initiation of a close signal from the. l Solid State Protection System. This test satisfies activities j described by FSAR Table 14.2-2, Sheets 50 and 50a and the deferred i i preoperational testing in System Test Matrix 1-6400. 1 TEST METHODOLOGY This test is performed in Mode 3 or in Mode 4 above 300'F. Strip chart recorders are connected to the MSIV position indication j circuits and to a test switch which has an input to theLTrain A- 1 Solid State Protection System that can actuate slave relay K627. i The MSIVs are then closed by operation.of the test switch and the 1 response times from the test switch actuation to the MSIV fully l closed indications are determined from the recorder traces. q The test is then repeated with actuation of the Train B Solid State Protection System K627 slave relay. l l
SUMMARY
OF RESULTS ! The Main Steam Isolation Valves were response time tested frc,m the u Train A Solid State Protection System. The recorded closure times 1 were as follows: Valve Number Closure Time
, 1-HV-2333A 4.04 seconds 1-HV-2334A 4.48iseconds 1-HV-2335A 4.44 seconds 1-HV-2336A 3.92 seconds l
The valves were then responso time tested from Train B Solid StatG ! Protection System and the recorded closure times were as follows: { valve Number 21.gsure Time 1-HV-2333A 3.78 seconds 1-HV-2334A 3.62 seconds , 1-HV-2335A 3.82. seconds J 1-HV-2336A 4.68. seconds The valves all satisfied the maximum allowed closure time criterion of 5.0 seconds.
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r 1 ) 3.6.6 - REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVE LEAKAGE TESTING - EGT-712A OBJECTIVE . The Reactor Coolant System Pressure Isolation Valve Leakage Testing is performed to demonstrate that the leakage past these valves is within the limits required by CPSES Technical Specification 3.4.5.2.f. The leakage tests of the Train A RHR Hot Leg Injection Valve 1-8841A, and the four RHR Cold Leg Injection Valves 1-8818A, 1-88188, 1-8818C, and 1-8818D satisfy-the deferred preoperational testing described in System Test Matrix 1-5700. TEST METHODOLOGY With the plant in Mode 5 and . the Train A RHR and SI Hot Lag Injection flowpaths not in use, a valve lineup is established to route all leakage from Valve 1-8841A through-the SI test header. Either actual RCS pressure or a temporary hydrostatic pressure pump i is connected to apply pressure against valve 1-8841A from the RCS side. The leakage is measured by flow through the SI test header flowmeter and then mathematically converted to the leakage that would exist at the normal RCS pressure of'2235 psig. , With the plant in Mode 5 and the applicable loop of RHR and SI Cold Leg Injection flowpaths not in use, a valve lineup is established to route all leakage through the RHR Cold Leg Injection valve under test to the SI test header. Either actual RCS pressure or a temporary hydrostatic pressure pump is connected to apply pressure against the valve from tho'RCS side. The leakage is measured by ! flow through the SI test header flowmeter and is then also i i converted to leakage that would exist at the normal RCS pressure of l 2235 psig.
SUMMARY
OF RESULTS Each of the RCS Pressure Isolation Valves (check valves) were verified operable by forward flow prior to the leak test. The Train A RHR Hot Leg Injection talve 1-8841 A.was tested in Mode 5 with a hydrostatic pressure pump supplying a test pressure of 230 psig. The recorded SI test header flow (leakage flow) was 0.0 gpm. ! The leakage, converted to the leak rate that would exist at 2235 l psig, was also 0.0 gpm. This satisfied the leakage flow acceptance l criterion for this valve of 3.0 gpm or less when at 2235 psig.
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- 3. 6. 6 - REACTOR COOIANT SYSTEM PRESSURE ISOIATION VA4VE LEAKAGE l TESTING - EGT-712A (Continued)
I
SUMMARY
OF RESULTS (Continued) l The RHR Cold Leg Injection valves, 1-8818A, 1-8818B, 1-8818C and ; 1-8818D, were individually tested in Mode 5. The test results were ,
- as follows
- l Measured Flow valve Test Pressure Measured Flow Converted to 223 50 sic 1-8818A 225 psig 0.0 gpa 0.0 gpa 1-8818B 340 psig 0.1 gym 0.26 gpa '
1-8818C 290 psig 0.0 gpa 0.0 gpe 1-8818D 290 psig O.0.gpa 0. 0. gpa These valves also satisfied the leakage rate acceptance criterion for these valves of 3.0 gpa or less when at 2235 psig. Valve + 1-8818B was tested using actual RCS pressure. The other three , valves were tested using a temporary hydrostatic pressure pump. 1 i 9 i l I l t
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l 3.6.7 - CONDENFATE REJECT VALVE TEST - ECT-TP-90A-002 OIL 7ECTIVE The Condensate Reject Valve test is performed to demonstrate that ! condensate reject and makeup isolation valves 1-HV-2484 and 1-HV- ; 2485 are capable of stroking to the fully open-and closed positions , under dynamic ' operational conditions. This test satisfies ! activities described by the deferred preoperational testing in System Test Matrix 1-9505.
-)
TEST METHODOLOGY ) With the plant in Modes 4, 5, or '6 .and condenser vacuum j established, makeup flow from the Condensate storage Tank (CST) to ; the condenser hotwell is established. Each of the isolation valves is then closed and reopened. Then, condensate reject. flow from the i condenser hotwell to the CST is established and each of the 1 isolation vr.1ves is again individue.11y closed and reopened.
SUMMARY
OF RESULTS ! With the plant in Mode 5, condenser vacuum established, and with condensate makeup flow to the hotwell, valve 1-HV-2484 failed to fully close from the handswitch operation. The valve's limit r switches and torque switches were readjusted. The condensate i reject isolation valves were then retected and both properly opened i and closed under makeup flow and also under reject flow conditions. l l I l
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l { 1 4.0 - REFERENCES
- 1) Comanche Peak Steam Electric Station Final Safety Analysis !
Report ;
- 2) Regulatory Guide 1.68, Revision 2 l
- 3) Regulatory Guide 1.68.2, Revision 1 ,
- 4) Regulatory Guide 1.133, Revision 1 >
- 5) Comanche Peak Technical Specifications >
- 6) Comanche Peak operating License NPF-28
- 7) Comanche Peak Operating License NPF-87 *
- 8) WCAP-9806, Rev. 2, The Nucisar Design and Core Physics ,
Characteristics of the Comanche Peak Unit 1 Nuclear Power ' Plant, Cycle 1
- 9) Westinghouse NSSS Startup Manual
- 10) Letter, N.R. Metcalf, Westinghouse Electric Corporation, to B. W. Coss, TU Electric, " Comanche Peak Unit 1 Cycle 1 Boron Worth", 90TB -G-007, April 12, 1990. !
r t t u l t l -209-r _ . _ _ .}}