ML060680590
| ML060680590 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 05/09/1977 |
| From: | Gilleland J Tennessee Valley Authority |
| To: | Moseley N NRC/RGN-II |
| References | |
| 770512F0194, QAS 77051007, TAC MC3743, TAC MC3744, TVA-BFN-418 | |
| Download: ML060680590 (168) | |
Text
{{#Wiki_filter:ENCLOSURE 5 TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT (BFN) UNITS 2 AND 3 MAY 9, 1977 - FINAL SLM4ARY REPORT, UNIT 3 STARTUP BROWNS FERRY NUCLEAR PLANT
QAS '77 0 - 1 0 0 0 7 N11 KSy 98 1977
§12FO I I 194 Mr. Korun C. ovelaey. Director I -
U.S. I ucear Rolatory Coasai ^..tB)5. Office of Iplction anI ZiorcQ=Mat Regio It
- V-r 230 Peacbtre Street, rd., Suite 1217 tLV1 : 4:
Atlanta, Ceorgia 30303 bear Hr. Moseley: FIXA SM2MAY RPMoT - UNIT 3 STASTUP - 8WNS ERY NM= PLANT - DOCIET 10. 50-296 - OWEATEM LICL4SE DIR-68 In accordace with Brouns Ferry Technical Specificattona 6.7.1.&, us ere subCitting tue "FIal Sary Report - Unit 3 Startup - Browns rYM Nuclear Plant." Vwy truly yours TESSEE VALZ AMTEOIT J. E. Cllelend Assistant Maager ot Power SF:.JRC:HCB Encloade CC (Enclosure): Director (2) Office of Masenent Infoation and 1Progra Control U.S. Nuclear Rgulateor Comiaaion Uabington, D.C. 20555 5-10-77--PLD:CGE CC (Enclosure): Director (40) P. L. Duncan, WllD136 C-K Office of Inspection and Enforcenlnt W. M. Bivens, 5100 MCB-K U.S. Nuclear Regulatory Comicslon MEDS, E4B37 C-K \K WAChington, D.C. 20555 D.- R. 'attersod7WrUCtZD G-K A. V. Crefsuse, 401 UBB-C E. S. Fox; 716 EB-C H. J. Green, Br Ferry EJLMED FROM OClT L. 11. Hills, 303 -C F. A. Szczepabskl,, 17 UBS-C AVAILABLE COPY E. F. Thomna, 818 -C Godwin Ualliamz, Jr. 830 PR8-C BC (Enclosure): JR H.EnsEtnhm Wt Q T. D. Knight, 727 EB-C P. A. Krenkel, 268 401B-C W. E. Buist, General Electric Company
TENNESSEE VALLEY AUTHORITY Division of Power Production FINAL SUMARY REPORT UNIT 3 STARTUP BROWNS FERRY :IUCLEAR PLANT Submitted by /vfo/ F _- - S. Plan Sufre-rintenent CI//7 7 Browns Ferry Nuclear Plant Decatur, Alabama
TABLE OF CONTENTS Page 1.0 Introduction . . . . . . . . . . . . . . . . . . . . . . . a2.0 Su'ary 2.1 Significant Dates of the Startup Test Program . . . . 3 2.2 Test Completion Dates for Startup Test . . . . . . . . 3 2.3 Power Flow Map with Startup Test Conditions . . . . . 3 3.0 Results 3.1 STI 1, Chemical and Radiochemical . . . . . . . . . . 7 3.2 STI 2, Radiation Measurements .. . . . . . ..18 3.3 STI 3, Fuel Loadling . . . . . . . . . . . . . . 20 3.4 STI 4, Full Core Shutdown Margin . . . . . . . . . . . 24 3.5 STI-5, Control Rod Drives .... . . . . . . . . . . 29 3.6 STI-6, SRM Performance and Control iRod Sequence . . . 41 3.7 STI-9, Water Level Measurements . . . . . . . . . . 45 3.8 STI-10 , IRK Performance. . . . . . . . .. . a .. . 48 3.9 STI-l1 p LPBH Calibration . . . . . . . . . . . 51 3.10 STI-12 APRMI Calibration . . . . . .0 . .. . . ..52 3.11 STI-13 p Process Computer . . . . . . . . . . . . . . 55 3.12 STI 14 p RCIC . . . . . . . . . . . . . . . . , . . . 58 3.13 STI-15 I1 EPCI . . . . . . . . . . . . . . . . . . . . 61 3.14 STI-16, I Selected Process Temperat es . . . . . . . 64 3.15 STI 17, System Expansion .... . ... . . . ... 66 3.16 bs-18, Core Power Distribution . . . . . 76 3.17 S.:-19, C-:e . ... . . . . . . . ... 80 3.18 STI-20, Steam Production .. . ..... 82 3.19 STI-21, Flux Response to rods . . . . . .
....... 84 S
TABLE OF CONTENTS (Continued) 3.0 Results (Continued) Page 3.20 STI-22, Pressure Regulator . . . . . . . . . . . . . . 36 3.21 STI-23, Feedwater System . . . . . . . . . . . . . . 90 3.22 STI-24, Bypass Valves . . . . . . . . . . . . . . . . 94 3.23 STI-25, Main Steam Isolation Valves . . . . . . . . 98 3.24 STI-26, Relief Valves . . . . . . . . . . .. . . . . 101 3.25 STI-27, Turbine Trip and Generator Load Rejection . . . 1oL 3.26 STI-30, Recirculation System . . . . . . . . . . . . . . 108 3.27 STI 31, Loss of T-G and Off-Site Power . . . . . . . . . 118 3.28 STI-32, Recirculation Speed Control and Load Following 120 3.29 STI-33, Turbine Stop Valve . . .. . . . . . .. . . .. 123 3.30 STI-34, Vibration Measurements . . . . . . . . . . . . 125 3.31 STI-35, Recirculation System Flow Calibration . . . . . 126 3.32 STI-70, Reactor Water Cleanup System * . . . . . . . . . 128 3.33 STI-71, Residual Beat Removal System . . . . . . . . . 130 3.34 STI-72, DryweUl Atmosphere Cooling System . * * . 131 3.35 STI-73, Cooling Water Systems . . . . . . . 134 3.36 STI-14, Off-Gas System e ** * *... ... 136 3.37 STI-75, Reactor Scram From Outside Main Control Room - - 145
STARTUP TEST RESULTS FINAL REPOR'T BROWNS FERRY NUCLEAR PLAIT UNIT 3 Abstract The final report of the startup test program performed at Browns Ferry Nuclear Plant Unit 3 is presented In three parts: (1) Introduction, (2) Summary, and (3) Results. Results from core physics. thermal-hydraulics and system performance tests are presented such that the actual empirical values obtained are compared against expected or design values. 'Where devia-tions were noted, resolutions or corrective actions are also described. 1.0 Introduction .11 Purpose The purpose of this report is to present a concise summary and pertinent detailed results obtained inthe performance of startup tests at Browns Ferry Nuclear Plant Unit 3. Thi startup test program embraced core physice, thermal-hydraulic, electromechanical and overall system dynamic performance. 1.2 Plant Description Browns Perry Nuclear Plant Unit 3 id a single-cycle boiling water reactor designed by General Electric Company (GE) for the Tennessee.Valley Authority (TVA) and is the third of a three-unit site to be placed in service. The plant is located oan the Tennessee River in Northern Alabama. The design gross electrical output is 1098 Mve, derived from a core thermal power of 3293 MIt. 1.3 Startup Test Program Near the time of completion of plant construction, the preoperational test program begins. This period is designated as Phase I of the test program, during which testing of components, subsystems and combined systems are per-formed. These tests are not covered -Ln this report. The startup test program begins with the loading of nuclear fuel and continues through the completion of 1.OOZ power testing and the warranty run. It is composed of Phases It through V, as follows: Phase II - Open Vessel and Cold Testing Phaae III - Initial Heatup Phase IV - Power Tests Phase V - Warranty Tests
1.3 Startup Test Program (Cantinued) During this period the plant is taken to its designed full-power operating condition in a safe, controlled, gradual fashion. Extensive testing is performed under selected, controlled operating conditions to demonstrate safe, efficient performance of plant components. The startup test program began with fuel loading on July 3, 1976, and continued through completion of the warranty run and 1002 power testing. Commercial operation began on March 1, 1977. 1.4 Startup Test Description Documents such as the Operating License, Technical Specifications. Plant Operating Procedures, and equipment manuals, control operations during the plant startup test program. Two documents are supplied by GE-NED for Implementation of the startup testing of the equipment it supplies; the start-up test specification and the startup test instruction (STI). The Startup Test Specification is a document issued for review and approval by GE Management and is used for planning and scheduling tests. The basis for the chosen tests is that they are required either to demonstrate it is. safe to proceed, to demonstrate performance, or to obtain engineering data. This document defines the mini:um= test program needed for safe, efficient startup. The purpose, description, and criteria are given for each test. together with a sequential guide for performance of the tests. The Startup Test Instruction is a document written for use in the control room by qualified GE and TVA personnel. It contains sufficient pertinent information to permit such personnel to properly perform and evaluate each startup test. TVA Division of Engineering Design (DED); Division of Power Production, Plant Englieeiing Branch; and Browns Ferry engineers reviewed the GE Startup Test Specification and Startup Test Instructions; and with appro-priate revisions, specified Browns Ferry Master Hot Functional Test Instruction (IMMTI), Master Startup Test Instruction (MSTI), and Startup Test Instructions (STI's) were issued. The HSFTX and HSTI coordinated and documented all test activities from initial fuel loading to the completion of all startup tests. These instructions provided guidance for sequence of events, and control points for satisfactory test completion and review.before power ascension. The GE-supplied STI's were revised, as necessary, by TV!, engineers. These STI's were reviewed by the Plant Operations Review Committee (PJRC) anti approved by the TVA Plant Superintendent and GE Site Operations Manager.
- FINAL
SUMMARY
REPORT - NFNP UNIT 3 1.5. Startup Test Acceptance Criteria The Startup Test Instruction for each startup test contains criteria for acceptance of results of that test. There are two levels of criteria identified, where applicable, as level 1 and level 2. The level i criteria include the ialues of process variables assigned in the design of the plant and equipment. If a level 1 criterion is not satis-ffed, the plant is placed in a satisfactory hold condition until a resolution ai made. Tests compatible with this hold condition may be continued. Following resolution, applicable tests must be repeated to verify that the requirements of the level 1 criterion are satisfied. The level 2 criteria are associated with expectations in regard to performance of the system. If a level 2 criterion Is not satisfied, operating and testing plans would not necessarily be altered. Investigations of the measurements and of the analytical techniques used for the predictions would be started. By meeting the criteria, startup test results demonstrate agreement with design specifications and predictions. Startup test results were reviewed and approved by PORC and the plant superintendent and are undergoing a final review and evaluation by TVA DED. 2.0 Simmary of Test Program 2.1 Chronology of Test Program Table 2.1 presents the dates for signifircan events in the unit 3 startup test program. 2.2 Startup Test Completion Dates Table 2.2 presents a sumary of the dates of completion for all startup tests at each test condition 2.3 Power Flow Hap Figure 2. 1 presents a power flow map for Browns Ferry unit .3. showing flow control lines and the nominal positions of test conditions for the startup test program.
FIP SVHKMRY REPORT - BFNP UNIT 3 Table 2-1 Major Events of Unit 3 Startup Test Program Date Event July 3, 1976 First fuel assembly loaded. July 22, 1976 Core fully loaded to 764 fuel assemblies August 8,1976 Initial critical during STI-4, Shutdown Margin Demonstration. Also, Initial In-s equence critical same day. August 18, 1976 Full Power license received August 19, 1976 Begin initial nuclear heatup August 24; 1976 Reached rated temperature and pressure September 9, 1976 Initial generator synchronization September 12, 1976 Completion of Eeatup Test Phase October 6, 1976 Completion of 25X testing October 29, 1976 Completion of 501 testing November 12, 1976 Completion of 751 testing November 20, 1976 100X power first attaixed December 24, 1976 Completion of 1002 testing December 26, 1976 Began 300-hour warranty demonstration January 7. 1977 Completion of 300-hour warranty demonstration (1400 hours) March 1; 1977 C0MMERCIAL.OPERATION
~ -
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(1) P.,c.,mt of rated -- ar, 3223 W6t.
.(2), 1reet of tated flow. 102.5 x L0O lbo/hr.
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W CORE RECIRCULATIOI FLO'. (%of 102.5 x 10" lbs/hr) Test Condition No.! I 21 2D, ,2E
- 3A 3C 3D 3E 4A 4C 4D 4E
- 5 Ro'd Pattern V a* a a b* b, b b c* c c c
- Al
% Piiri2 Seed _ " 41 0* %'68 E 0** 41 1.68 E 0* 141 -68 1 = -
%.Powaer 15-35* t25 30-50* 40-60* * ,37 37-57* 50-70* 65-86* 'L50 55-75* 70-90* 95-100*
- Z20*
7.Core Flow - 47 HC ii70* i104*
- NC - 48* *70* 1.102* 48* %70* rl0O, _ tlOS5 a Rod Patrerni 0itained at Test Condition No. 2E A Natural Circulation b Rod Pattern obtained at Test Condition No. 3E B 20X Pump Speed c Rod Pattern Obtained at Test Condition No. 4E C Analytical low limit of master flow control
- Asterlsked values are set as initial test ( 417, speed) conditions; non-asterisked values are estimates D Contractual lower limit of flow control NC Natural Circulation (687 speed)
V Varies E Punip speed for rated flow at rated power P Nominal curve of max allowable pump speed (equipment limits other than core) G Analytical flow corresponding to max.allowable steady-state fuel channel tP FIGURE 2. 1 APPROXIMATE POWER FLOW HAP SHOWING STARTUP TEST CONDITIONS
FIMAL SUM!OARY REPORT - BFNP UNIT 3 3.0 Results
.3.1 STI-i,Chemical and Radiochemical 3.141 Purpose The principal objectives of this test are:
- 1. To secure Information on the chemistry and radio-chemistry of the reactor coolant.
- 2. To determine that the sampling equipment, procedures, and analytical techniques are adequate to supply the data required to demonstrate that the chemistry of all parts of the entire reactor system meet specifi-cations and process requirements.
- 3. Specific objectives of the test program include evaluation of fuel performance, evaluation of deminer-alizer operations by direct and indirect methods, measurement of filter performance, confirmation of condenser integrity, demonstration of proper steam separator-dryir operation, measurement and calibration of the off-gas system, and calibration of certain process instrumentation. Data for these purposes is secured from a variety of sources: plant operating records, regular routine coolant analysis, radidchemical measurements of specific nuclides, and special chemical tests. .
3.1.2 Criteria Level 1 Chemical factors defined In the technical specifica-tions must be maintained within the limdits specified. The activity of gaseous and liquid effluents must conform to the license limitations. Level 2 Water quality must be known and should remain within the guidelines of GE water quality specifications. 3.1.3 Analysis STI-1 testing was conducted at open vessel. heatup, test conditions 1, 2E, 3E, and 4E, as defined on the power flow map in section 2.3.
FINAL
SUMMARY
REPORT - BFNP UNIT 3 3.0 Results (Continued) 3.1 STI-1 Chemical and Radiochemical (Continued) 3.1.3 Analysis (Continued) Chemical tests of the primary coolant were made prior to heatup and yielded the following results: Conductivity (imho/cm @ 250) 0.32 Chloride (ppb) < 50 Turbidity (FTU) 0.15 Boron (ppb) < 50 Silica (ppb) 10 All level 2 criteria were satisfied with the exception of chloride concentrations in the condensate storage and demineralizer water storage tanks. Plant analytical procedures have a minimum chloride sensitivity of 50 ppb. GE limit for chlorides in the storage tanks is 10 ppb. GE field disposition request FDDR ER3-446, dated 8126I76,permits the acceptance of <50 ppb chloride concentration. Reported data for chloride concentration comply with this limit. No further action is required. Chemical tests of the primary coolant were made during the initial heatup. The results were: Conductivity (mho/cm @ 25°) 0.32 Turbidity (FTt) 0.46
-Chloride (ppb) < 50 Boron (Vpb) 90 Silica (ppb) 540 Throughout the startup test prmgram, chemical and radiochemical sampling and analyses were performed on a routine and special test basis. Routine surveillance of the reactor water, condensate, and feedvater,, embraced the measurement of conductivity, chloride content, turbidity, and boron content.
Testing of steam separator andi dryer performance at Browns Ferry 3 consisted of two (@ 502 amd 100% power plateaus) Injections of sodium sulphate into the reactor water to
-ncrease the sensitivity of the 14a-24 carryover measurements with the reactor cleaniup system out of service. Reactor water conductivity exceeded 2.0 umho/cm @ 250 for 33 hours from September 15 to September 19, 1976, @ 25-7 testing plateau due to placing feedwater heaters in service.
TIURL
SUMMARY
REPORT - EFW UNIT 3 3.0 Results (Continued)' 3.1 STI-l,Chemical and Radiochemical (Continued) 3.1.3 Analysis (Continued) The levels of lodines. silica, insolubles, and boron were within established limits during the startup testing. Gamma scans of primary coolant water indicated expected corrosion and activation products. Reactor water chloride concentration was within the 1 ppm technical specification maid=x- limit throughout the startup. The chloride concentration was vithin the operational technical specification limit of 0.2 ppm throughout the startup. All criteria were satisfied with the exception of condensate oxygen concentration at all power testing levels. GE fuel warranty document (22A4367), Brouns Ferry 3, sheet 9, changes the lmi4t from 14 ppb to c 2000 ppb. All oxygen values met this limit; therefore, disposition of this exception Is complete. Wo further action is required. Table STI 1-1 summarizes the results of the chemical and radiochemical testing performed during startup.
. ,A
MINAL SUIXARY REPORT - BFN'P UNIT 3 3.0 Results (Continued) 3.1 STI-1, Chemical and Radiochemical (Continued) 3.1.3 Analysis (Continued) Table STI 1-1 o5e35X 40-60X 65-85% 95-100% Power Power Power Power Sample Source and Test Date 10/11176 1113/76 11/21/76 K t 7 1970 1 2531 i 3291 We N%" Reactor Water Limit 193 612 b 847 4 1096 I. - - _______________ Conductivity. UMho/cm 1.0 0.80 0.59 0.55 0.38 Chloride, ppM 0.2 [ .05 c0.05 c0.05 ._ _05 Turbidity or insolubles, JTu lOppm 0.55 <0.075 0.13 <0.10 Iodine-131, pCi/ml 6.55 E-07 '1.47 E-06 1.24 E-05 2.15 E-05 I - - _ Iodine-133, pCi/Mi 6.52 D-06 3.52 E-05 7.37 E-05 9.87 E-05 I.1.3 I Gross Activity
-filtrate, cpm/ml, 2 brs. 2716 9852 29834 24084
-crud, cpm/ml, 2 hra. 3416 6124 3086 2374 GrcDss Activity
-filtrate, cpm/ml, 7d 57 112 217 529
-crud, cpm/ml, 7d .5 161 42.9 80 Lica, ppb 5.0 ppm 0.314 0.341 0.28 0.38 4 -
3oi ron, ppb 50 ppm <0.05 co.05 00.05 <0.05
-_ - I ..
AJA - . .
.0 -
4;I_ A)~VAr a% X c _ L ..I
-pU-
'TUM SMARY UPRT - MO U1IT 3 3.0 Results (Continued) 3.1 STI-1 Chemical and Radiochemical (Continued) 3I1.3 Alysis (Continued)
Table Sm 1-1 15-35% 40-60Z 65-85% 95-1002 Power Power Power Power Ss Source and Test eal 780_I/ 6_ - 11_2_176 _Date _73g Ractor Water (Continued). .L93 542 847 1070 Chemical Analysis on filtrate, ppb
-iron __ -. ._ ._D_ 0.167 -. opper ._ XX XX 19_._74 -nickel ___ __ < o._001 -chromium x.
D_ XI 3.79 Chemical Analysis on Crud, ppb
-iron _.-8 8.95 7..1 12 4.60
-copper ii xi xx Ic 0.001
-nickel xx xI xI 0.775
-chromium X xx xx It 0.001 Spectral Analysis on major nuclides at 24 hours Filtrate Ho-99 Cr-5I Mo-99 Mo-99 Tc-99m Cu-64 Tc-99m Tc-99m Va-24 lla-Z4 Cr-51 Cr-51
. -76 Zn-69m W-187 W-187 co-Se Co-58 Zn-65 NOTE: XX symbol signlfies data- As-76 As-76 Cu-64 Cu-64 not required by the -test na-24 As-76 instruction. . . Na-24 cop.:-
FWAL SUMARY REPORT - BFNP UNIT 3 3.0 Results (Continued) 3.1 STI-1, Chemical and Radiochemical (Continued) 3.1.3 Analysis (Continued) Table Sm 1-1 (Continued) 15-35%. 40-60% 65-85% 93-1007 Power Power Power Power andI77 Dat 10pler
- Sle and Test Source KWt 780 1 70 2531 3256
_ Limit__ 193 542 847 1070 V-157 Cr-51 W-187 W-187 Crud Cr-51 Co-58 Mo-99 Mo-99 Za-69M M-54 Ho-99M Tc-99M Cu-64 Fe-59 Sb-125 Fe-59 Na-24 Co-60 Fe-59 Cr-51 Zn-65 Cs-134 Cr-51 Zn-69m As-76 Na-24 Zn-69m Co-58 Ca-137 Zr-95 1-135 Zn-65 h-54 Zu-75 As-76 Cu-64 Ma-56 Ce-141 Sb-124 As-76 Fe-S9 Zr-95
- Sb-124 la-140 Zr-97 . Mn-54 La-140 Hb-95 Co-60 Co-S8 Co-58 Mn-54 Zn-65 Cu-64 N-924 Condensate Demin. Influent Conductivity, __hocm 0.34 0.13 0.094 0.076
- Chloride, ppm- __ cOOS <0.05 <0.05 <0.05 Insoluble iron, b_. 25 <10 <25 10 Condensate Demin. Effluent Conductivity, XMho/cm 0.1 0.25(1) 0.072 0.083 0.057 (1) Heater drain problems
MINAL SUMHARY REPOIT .- BFNP UNIT 3 3.0 Results (Continued) 3.1 ATI-1, Chcal and Radiochemical (Continued) 3.1,3 Analysis (Continued) Table STI 1-1 (Continued) 1-352 40-602 -- 65-85% 95-1002 Power Power Power l Power. Date__ 1 76 101127 .1 11_13776 II. aMple Source and Test - 78101/76
- is SapeSuc adTs It 780 1770 I 2531 ! 3256
. 193 542 847 1070 Condensate Dekin. Effluent (Cont' )
- Insoluble iron, ppb 20 <10 dO 'CIO d1O 2 ab Lab Lab Lab Oxygen, ppb 14(2) 150 Anal. 100 Anal. 80 Anal. 100 Anal. Feedvater Conducti-ity, umholcm 0,10 0.46(1) 0.093 0.085s 0.072 Iron - insoluble, ppb 10 <10 10 17.64
-soluble, ppb xx 4.13 16 4.15 Nickel - Insoluble, ppb xx xx xx 0.463
-soluble,p XX xx 0.588 Copper - Insoluble, ppb _x XX _X 0.663
- soluble. ppb x DC . 0.001 Xx Crud XX Chromim - soluble, ppb XX XX XX Sol _
Off-Gas Aftivity @ SJAE;, yCi/sec. (16 gases) . <0.11 c61.6 <98 79.9 N-13 ~ SJAE, PCL/sec. 190 1450C 168j 1684 rlow rate, cfm (FR-66-111) 160.6 38 35 38 XX Symbol scinifies data not reqvire by the test instruciou. (1) Heaters placed into service. (2) Li-itts changed to C 200 ppb in GE fuel warranty docuient (22A4367),
-able I, sheet 9.
FINAL SUXQ(ARY REPORT - BFNP UNIT 3 3.0 Results (Continued). 3.1 STI-1, Chemical and Radlochemical (Continued) 3.1.3 Analysis (Continued) Table STI 1-1 (Continued) _ 15-351 I 40-602 65-85Z 95-lOOZ
. Power Power Power Power Sample Source and Test D 10 /766 - 11/3/76 , 1/2T776-I7 Mt . 780 1 1970 2531 3291 3 612 ; 847 1096
,Off-Gas (Continued) I Composition - air, cfm . 140 38 35 38 Radiolytic + .. 0. _ 0 00. 0 Delay time, min. XX Xx XX 186.6 Activity A release at stack ) 8( 1 1 tCi/sec. _) 72.S_ 128(_) 155(_) 128.7_(_) Activity Pattern __._. _ Recoil. Recoil. Recoil. Recoil. Off-Gas Monitor A 7 10 18 i6 Reading, mr/hr xX XX JCR XX XX Stack gas monitor A 10 12 18 12 Reading. cps B 10 16 18 16 XX Symbol signifies data not required by the test instruction. (1) Combined activity from units 1,.2, and 3.
FINAL
SUMMARY
REPORT - JFNP UNIT 3 3.0 Results (Continued) 3.1 STI-1. Chemical and Radiochemical (Continued) 3.1.3 Analysis (Contined) Fuel Cladding InteEritl Table STI 1-2 shows representative lodine data data obtained during the startup.
- - . Table STI 1-2 Eat-1-ted 1- (1) UCi/M UCL/m1 VCi/ml 1Ci/l/l ILCl/ml Date Time MHt Carryover (%) 1-131 1-132 1-133 1-134 1-135 10/11/76 0700 1970 _ '1.47 E-06 2.14 E-05 3.52 E-05 5.36 E-03 6.12 E-05 10114/76 2000 1693 0.3(2) _ _ .
10/25/76 0800 984 _ 6.31 E-07 9.5 E-07 3.34 E-06 4.31 E-06 6.00 E-06 11/15/76 0700 2882 - 4.96 D-06 7.0 E-05 6.43 E-05 4.02 E-04 1.11 E-04 11/21/76 1800 3275 0.22 3)_ 11/29/76 0800 2075 8.75 E-06 1.00 E-04 9.81 E-05 2.95 E-04 1.81.E-04 12/3/76 0800 3178 _ 5.48 E-06 1.14 E-04 6.32 E-05 2.30 E-04 1.32 E-04 (1)I-131 activity concentration insufficient. (2)50% power - no cleanup test (3)100% power - no cleanup test
FINAL SUMHARY REPORT - BFNP MIT 3 3.0 Results (Continued) 3.1 STI-1. Chemical and Radiochemical (Continued) 3.1.3 Analysis (Continued) Condensate
*The condensate pump discharge and condensate demineralizer effluent conductivities were only slightly high during the initial heatup through the 15-352 test conditions, however, they were within established limits throughout the remainder of startup testing. The following table, STI 1-3, shows the plant conductivity history during the startup testing.
Table STI 1-3 - Browns Ferry 3 Startup Conductivitles (umho/cm) Condensate Condensate Demineralizer Date Power Pump Combined Reactor
.(Thermal) Discharge Effluent Water 8/7176 02, No Beat 0.50 0.20 0.32 8/24/76 1, Reatup 0.15 0.10 0.3 - 0.7(3) 9/15/76 15-35% 0.34 0.185 0.30-2.20(3) 10/15/76 502 O.ll 0.07 0.50-2.40(2) 10/29/76 40-60% 0.088 0.078 0.59 11/3/76 702 0.094 0.083 0.55 11/21/76 991 (approx. 0.076 0.057 0.3 - 1.6(2)
(2) No cleanup test (3) Range of Reactor H2 0 conductivity during test period. . .. -... .... . . . ' . jS . .
FINAL
SUMMARY
REPORT - BFNP UNIT 3 3.0 Results (Continued) 3.1 STI-1. Chemical and Radiochemical (Continued) 3.1.3 nalysis (Continued) Sa§pling System Prior to startup, a root valve verification program was conducted to ensure that the origin and approximate length of sampling lines was knonm. Radwaste Both the liquid and solid radwasee systems performed satisfactorily during the startup period even though intermit-tent Inputs to the liquid system exceeded design values. Condensate and Cleanup Demineralizers The condensate demineralizers were initially placed into service in late 1975 and were subsequently used to clean water during construction and preoperational testing. Both the condensate and cleanup demineralizers performed satisfactorily during the startup period.
FINAL
SUMMARY
REPORT - BFNP UNIT 3 3.0 Results 3.2 -STI-2. Radiation Measurements 3.2.1 Purpose The purposes of this test are to:
- 1. Determine the background radiation levels in the plant environs prior to operation for base data on activity buildup.
- 2. Monitor radiation at selected power levels to assure the protection of personnel during plant operation.
3.2.2 Criteria Level 1 The radiation doses of plant origin and the -occupancy times of personnel in radiation zones shall be controlled consistent with the guidelines of-the standards for protection against radiation as outlined in TVA Radiological Control Instruction. Level 2 There are no level 2 criteria. 3.2.3 Analysis STI-2 was performed at the following unit No. 3 conditions. Table STI 2-1 Survey Conditions I. Prefuel Loading May 12, 1976 II. Core loaded, Open vessel July 23, 1976 III. Plant at 6Z power August 26, 1976 IV. Plant at 25% power September 17, 1976 (limited survey) V. Plant at 58% power October 8, 1976 VW. Plant at 762 power November 3, 1976 (limited survey) VII. Plant at 100% power November 22, 1976 VIII. Plant at 100% power-warranty run December 28, 1976 (limited survey)
FINAL SUKWRY REPORT - BEFNP MNIT 3 3.0 Results (Continued) 3.2 STI-2, Radiation Measurements (Continued) 3.2.3 Analysis (Continued) At each point gam-a and neutron measurements were made as required by the type of survey. "Limited" surveys involved a selected part of the complete surveys, with only those points of normal occupancy being measured. Exceptions to each survey were as follows: Exce tions to Surve Plant Condition Test Point Exception (See Table 1) I RB-3-38 Neutron survey not made. Inaccessible due to shield plugs not in place. RB-3-44 Neutron survey not made due to inaccessibility. (15' above floor) II 1B-3-38. Same as above III NSO EXCEPTIONS
'L 1 E XITONS B0 V HO EXCEPTIONS VI No EXCEPTIONS VII RB-3-44 Test point RB-3-44 required rezoning as per RCI-l.
VIlI NO EXCEPTIONS As noted in table STI 2-2, oniy test point RB-3-44 required rezoning to meet criteria level l. This test point ie a blank drywell penetration located in the SE quadrant at the 593' elevation in unit 3 reactor building. It is located in a normally inaccessible location 15 feet above the floor. As a result of the survey, a cage was placed around the area and proper zone posting made. This brought the zone into com-pliance with RCI-l, thus fulfilling SrI-2 requirements.
FINAL SUIMARY REPORT - EFNP UNIT 3 3.0 Results (Continued) 3.3 STI-3. Fuel Loading 3.3.1 Purpose The purpose of STI-3 Is to load fuel safely and efficiently to the full core size. 3.3.2 Criteria Level 1 The partially loaded core must be subcritical by at least 0.382 AK/K with the analytically strongest rod fully withdrawn. Level 2 Not applicable. 3&3.3 Analysis Fuel loading began with-the loading of the first fuel assembly at 1646-zbours on July 3, 1976, and was succes-sfully completed at 0136 hours on JuYl22, 1976. At that time all 764 fuel assemblies were installed, the seven operational sources were In place, and the four source range monitors (SRHts) were electronically connected and functional. Vartial core shutdown margins ware verified at designated points during the loading process and met all criteria. Prior to loading the first fuel assembly, the four fuel loading chambers (FLC) were installed in dummy blade guides at approximately 2/3 core height and wnre connected to the plant SR electronics. The signal-to-nloise ratio was verified to be >2:1 and the FLC count rate5 was >3.0 cps. 5 The rod block and scram setpoints were set at 1XO0 cps and 5x10 cps, respectively. The shorting links were removed from the circuitry, placing the FLC/SIl and rM'. electronics in the non-coincidence scram mode. The Sb-Be operational sources were installed prior to fuel loading and used throughout fuel loading to establish neutron flux. The source strength was 686 curies on the initial load date and 552 curies at completion of fuel loaiing. After completion of the loading of each control cell (2 x 2 fuel assembly array) functional ;and suberiticality checks were made by withdrawing the associated control rod. In addition, partial core subcriticality checks were made ufter the loading of 16, 64, and 144 fuel assemblies to verify that
.. . . . . . ~
- . is . - .- . ,...*,.......................,:
PINAL
SUMMARY
REPORT - BFNP UNIT 3
- 3.0 Results (Continued) 3.3 (Continued) -
3.3.3 Analysis (Continued) the partially loaded core is subcritical by at least 0.38Z hRK with the analytically strongest rod fully withdrawn. As an added assurance that fuel was being loaded safely, inverse multiplication (lIH) plots were maintained of the PLC/8RS count rates. In certain cases special interpretation of these plots was required of the nuclear engineer because of geometric effects. These geometric effects were caused by loading a fuel assembly near an operational source or FLC and were expected. The PLC's were moved as necessary to maintain the count rate >3 cps and <1105 cps. (See figure STI 3-1.) The PLC's vere removed after 360 fuel assemblies were loaded and all four SEM's were then operational. The fully loaded core was verified for fuel assembly orientation, serial number, and proper location of fuel types by lowering the water level in the reactor vessel to allow visual verification. A video-tape was also made for a permanent record. Fuel assembly locations are shown in figure 81S-3-2.
FINAL SUMHARY REPORT - BFNP UNIT 3 DATE - 3.0 Results (Continued) WE 3.3 ITI-3, Fuel BROWNS FERRY UNIT .3 S - - . S CORE POSMON MAP X % SomPWIYMV
- wI. I WA y" 1*"!
a 31 - I_ I- U-
+/-1+/-1+/-+ +1+1+-
A U b -
-I-I- U 4 U' A -
*" .7 S!1U' 3 n _
+
_-_Ir I +- +
'I F
iI*-i9-t-F
+- +- +
_ _ _ l ] l .76
-A n , -I- U I. -
+fs+U A-
+F m -+1I U A - .1+1+1+!
U 4 U A
+ .
1:+ + I + + +'rii+1+1+1 r++h 30
+ __+ +] + + +r +- + +1+
V 4- + _+_ + + 7 + +i + + + i n+ _+ + + -__t+ 4' a _+ O~~- + . - as a n-____1______Ir 3, U n__+ + + _1+_
, _+- + + __ + *+
It v U5 S-- h
- A
+ A 'USfl I
S
+
i I. - - I T+I+/-T-I-I1++'+I A 4. I 9
-j IL-
+1TT I+ + f + + I+I+T+
a U% I of . - i:, :L I 1. _ ,_
..- '. u .
111.
-151I I -I - 57.+1+ I 11 II I . ,
I1I as le Is I Ie I1 a w I II I P I I IJ I so .7 II. 1* *I .g 5 . de 7 a7 of a S a " u9 is n n aj X Iin i 4.. s7 n 1 s . I i
.4_"+
. I
# I '- SYSTM OPMATZNG CONDITIONS:
I _
."T!T-
-14"q T'i .1 ' II-44-iV.
_ 9 .H
-I j _.
it 4.4..r..u4-+ 1 ff;1;f4.4_- +N tr - - - ._ 1. Accumulator precharge PC M
-u iii. I- 1I
- .V t .74..
-s T;-
*16'
_I Z-!9 .4 41.~ 4 **4. 11~.'Jt!. L. 44 I..*I. .4*1'JII .; 1 4i 5651585 paig At 700 F. eM q-1 .t
*.i 't i4 i0_ .1 14. I41V -1 IA:. ,s.14
-. (39.9/41.2 kg/cm2 at 200 C.)
,~. -1
.,m 1+I..4t4-
*1 4 4 eT
. ,.6-.% I - -
.L ..
U -i' _ t~tw. _1-jj+
-I-I t. T1 1=
iM
. . , .4. 2. Accumulator water side
- a a
;3 2.5 I.,.
t-tuI . 1"I* j*-!. tiT4 ..-i-AA-HJ' -4.4-. .'.4- - 1510 psig, (106.3 kg/cm2 ) max. 1390 psig, (97.7 kg/cm2 ) min. U) [+/-4!~ 44 Xf -i __ 1:fr igg~; 11 4 4:. 3. Scram valve air pressure
- C.;
*M 14. 70/75 pslg. (4.9/5.30kg/cm2 )
Data applicable to single CRD
.I 2.0 scrams with charging valve closed (V-113) or full reactor scram with charging valve open.
151 *Scram time io the time from
*1 loss of voltage to scram air pilot-valves to 90% insertion (pickup of "04").
~~I~~II*~I4 4f 1.4 4j~ §i1h ~: 1'l 1.0' 0 : iv froI4 c1s1 1G00 t; '.;L ~~iv.t llpsig SCIU.:l PrltnwMtR1!MCF UlRVEF UR 10-IDEL ll;l#ZE44.'atnd VIIl§;l@/tl ':';
Figure STI 5-1
FINAL SUW(ARY REPORT - EFNP WIT 3 3.0 Results (Continued) 3.5 STI-5. Control Rod Drive System (Continued) 3.5.3 Analysis STI-5 testing was conducted at open vessel, heat up, and test conditions 1, 3E, and 4E, as defined on the power flow map in section 2.3. All the control rods met the requirements of the tests performed on them during zero-reactor-pressure testing. Position indications, rod timing, stall flows, coupling checks, and friction tests were performed on each CRD. PositIon-Indicating Check - ... The rod position information system was extensively checked and was operating propesay,.. - Rod Timing and Stall Flows
** 'The normal rod withdrawal and insert-times, together with the stall flow were measured. Some of the drives were adjusted so that their.-tines were within the above criteria.
nChk This check was performed during fuel loading whenever a rod was fully withdrawn to position 48. All rods were coupled to their drives. Friction Testing A.l of the CRD's were friction tested by continu-ously inserting them from position 48 to position 0 and photographing the insertion pressure throughout the insert process. The friction test data were acquired using a strain gauge differential pressure cell and a storage oscilloscope. Polaroid photographs of the oscilloscope traces were taken to tecord the data. All control rods passed the continuous insertion Awax. -PMin. criteria. FILMED FROM BEST AVAILABLE COP!
FINAL SMOLARY REPORT - BFNP UNIT 3 3.0 Results (Continued) 3.5 STI-S, Control Rod Drive System (Continued)
3.5.3 Analysis
(Continued) Scram Testing During open vessel testing all control rods were scram tested. The average scram times fell well within technical specifications and criteria requirements. (See table STI 5-4.) Initially all rods met the level 2 criteria for individual scram times, except control rod drives 18-07.and 26-l5,. which had 90% scram insertion times of 1.914 and 1.913, respectively. The scram tests were repeated for CRD's 18-07 and 26-15 with normal accumulator pressure. The 90% scram insertion times were measured to be 1.712 and 1.752,.respectivelyl thus satisfying the level 2 criteria. From this data the four slowest control rod drives were chosen to be scrammed three times each with minimum accumulator pressure. All level 1 and 2 criteria were met for testing during the open vesael.phase.. Table STI 5-1. summarizes testing of the four slowest drives. Table STI 5-1 Four Slowest Control Rod Drives At Zero Reactor Pressure And Binfum** And Normal Accumulator Pressure
.Hiean*
Rod Location 90% Scram Time (see) 90% Scram Time (sec) Min. Accu. Press. Norm. Accu. Press. II 30-27 1.908
- 1.825 i I
18-07 2.047 1.712 II 26-15 1.974 1.752 I 14-19 1.854 1.825 i Fan of three scrams 10 psig 4I I'n 1... .
.. 7
. FINAL SUfMKAR REPORT - RFNP UNIT 3 3.0 Results (Continued) 3.5 STI-5, Control Rod Drive System (Continued) 3.5.3 Analnis (Continued)
Scram Testin& (Continued) During the initial heatup, the four slowest in-sequence CRD's were selected.for scram testing at 600 and 800 psig. Results are summarized in table STI 5-2. Level 2 criteria were not met by CRD's 34-43 and 30-07 at 600 psig and by CRD 22-S5 at 800 psis. All technical specifications and level 1 criteria were met for all.testing at 600 psig and 800 psig. At rated reactor pressure scram times were measured for all in-sequence CRD's with normal accumulator pressure.
-The selected four in-sequence CRD's were scrammed three times each with zero accumulator pressure.A.Tbe.results for rated..
pressure scram testing are summarized In table STI 5-2. The four selected.CRD's were friction tested and timed at rated pressure. All level I and 2 criteria were met for testing at rated pressure. . Table STI 5-2 Four Slowest In-Sequence Rod Scram Tests 90S Insertion Scram Time Drive Test kx Press. ax Press. Rx Press& Location Number 600 psig 800 pWg 1000 psig 1 2.87 2.77 2.34 22-55 2 2.94 3.14 2.72 3 2.84 3.06 2.80 Hean 2.88 2.99 2.62 1 2.77 2.95 2.85 26-27 2 2.81 2.84 2.76 3 2.97 2.82 2.61 Mean 2.85 2.87 2.76 1 2.89 2.86 2.72 30-07 2 3.02 2.83 2.79 3 2.81 2.82 2.77 Mean 2.91 2.84 2.76 _ __ , . 4 2.90 1 2.95 2.93 2.79 34-43 2 2.94 2.90 2.66 3 3.19 2.83 2.64 Mean 3.03 2.88 2.70 4 2.99
.5 3.01 .FILN ED FROM BE
.6 -I. 3.03
- - 5. ...--. _.- _.
I,, . . J-%A .-- v __
. .. . . r,%.v
FINAL
SUMMARY
REPORT -BF- UNIT 3 3.0 Results (Continued) 3.5 STI-5, Control Rod Drive System (Continued) 3.5.3 Aallsts (Continued) Scr!aMest31 (Continued) A reactor scram from hot-standby permitted a subsequent startup in "AS control rod sequence. This
- permitted "A" in-sequence CRD's to be scram timed at hot-standby instead of after the rod sequence control system interlocks were cleared during startup to test condition 1 as had been projected by the Master Startup Test Instruction (MSTI). The average scram times for all 185 CiD's at rated reactor pressure are summarized in table STI 5-4. Individual rod scram times are listed in table STI 5-3.
* ; . . U .
., ... ; . , ., ^ ,
FMAL SMY REPRT -S NP UNIT 3 3.0 Results (Continued) 3.5 W_-S. Control Rod Drive System (Continued) 3.5.3 Aunalvis (Ccntinued) Scram Testing (Continued) Table GTE 5-3 Isdividual Rod Scram Times Sequencea Rods - Scram Insertion T s. See. Scrat Insertion -'. Drive Press. 5 201 Z 90 Reactor Press. izoz 10 .. tocation PtIL _ 3003 9S6 0.324 0.687 .42812.475 j r.em 26-23 ij paqa 960 0.33210722 1.504 2.555
*22-03 jf 0.332 0.709 1.468 Z. 30-2 960 I 0.332 11.520 .2.611 14-;10 ~s 0.313 0.671 1.38012.403 18-23 i.564 06-19_ _9 00302 0.671 1.424 .476 -955 .318 90 1464-.563 02-31 _ 0.292 0.639 1.352 .354 10-21. I S .0.324710 01.512 2.551 i6 _0.294 0.645 9539 1:.356i .-339- 0.3620.7621.496 Z.555 59 - 9S6- 0.308 0.671 .412 .627 - .. ,
- S5 9S6 .33 7 Ab-3SJL Lb-il ac cc ~ :"UZc, Iii~syIU; WI J U LJig
£~~
A.4-7 26-SS 956 _ 0.326 .703 .4" 12.508 i 1 9 0.310j10.6711.388 2.411 2607_6 0.294 D.661 L.416 ?. 507 18-39 -z0.33810.738 0.508 2.636 l 956 7 .0308.685. .4S6 .523 .14-43 1 os q 0.294 10.682. 1.500 Ti 30-11 956 0.316 .716 .464 2.78 - *22-43 j of 1h0.310 0 hi i L-..360 55,- - 21 56_ 0.313 0.671 .420 2.459 3 8-2 [ s . . J I I0 +/- 3 27 0167 1 .4 9 0' I d . _
- :I 0.294 0.653 .364 2.402 38-4 _7 S 6 2 J210 .__2 51 06-2J -956 0.318 0.695 i.464 ? - 134-23 1 g55 110.33210.70611.492i2.546-.. . .
WU 0.316 t .Le 178 aswf littn -21 la % ran l Lfn 127IARi 06-35S 956 03 6118 2.419 Xb-7 ~ ~ 075 1 AM, 'V :9I'tlt-"- 5 06-43 958 0.318 .6S2 A . A 24-1 A4-2-23 .j Ij7CSSD I.-.579122.264 ._. 22-S9 0.0 .2 50-23 1 Dc3Jn02.178 JAIL.1 9SB 0.j313 0.6 47ZL.h 5. -S %Q 0 110-319 II Q-71 k t..].L2.5D 22 -51 9S8 0.292 0.637 i.308 kIZ74 - 42-11 1 r)C. ii? Q A32 10,71 A l .-S16 94. 30-51 9S8 0.31 -. 661. Z - U-3° 242 3 1; 0-4o.84.66i k21 .ao 958 0.324 .74 L S4iZ - t.516. :2.5.94 46-M -0.310 O.645 L.372 .419 4-43 9 . 22-19 9Q% 0.310 .679 .444 .499 46=4 10.286 L,655 L4 2 2.3.IQ 218-15 IQS;R 034 .61.440 ! 499 14-19 q9s8_ 0.332 .130 .532 .644 34-1.5 955 110.318 0.661 1,40.2,44I 10-15 -958 0.310 .671 .400 .426 38-19 955 0 10.698 2.683 22-35 58 .294 .676.6A 146 30.322 D.703 I.468 !.555
.46 _ ,69 5 V -2 .5 5 5 10-39 960 0.326 6 .440 .52 34 31, l614 i0.10 l .2Z2 2Wl . ,
95S -. "i -MA94 0nk ;L.9A2 r-,7A I k
*ja_15q A429 au InCa LAJdJ.U .4 I A141SOl2A6
;l.2-530 - --
1 955 !, X : 26-47 960 _0.318 .687 L >.474 50-A7 1 1 955 .4~ z MIS n. 7' J^tI2 Iz..450.. -
- 1. ,
FILED FROM BEST AVAILABLE COPY ... : .: -,- i ., . ..
FINAL SUMWRY REPORT - NFNP MNT 3 3.0 Results (Continued) 3.5 STU, Contro1 ibdDrive Istem (Coxntinued) 3.5.3 Analysis (Continued) Scram Testilnt (Continued) Table STI 5-3 (Continued) Inavidual Rod Scram Tines Sequence A Rods (Continued) I. Scryaq w . T." n..i i, ^A_ .......................................................................................................................... Loatlox9 42-47 Reactor Press. PoiC 955 5: q-q20.
.25 0,3; 502 i.~.
I ).4;,,11; if 34-47 95S 0 1 .62 ! 2.75_ 38-03 955 0.30 .653 1.388!2.420 M 1 95 0365 0.757 1.48012.523 50-S5 955 _0.3A 0.731 1.5632,683 50-39 955 0.31 00.71I 1.492;2.643. 46-S1 9SS 0.311 0.663 1.388!2.451' 3bS1 95 0.318 0.695 1.464.2.539 UZI T5S 0.294 0.653 1.36012.338-- 2 10. 19 1.3442.322: 7.637 L r- i 0.302 -531 I.336--2.354 58-2 9S5 0. 3050.626 1.28892.242 5831 95SS 0.35t 0.709 1.228!2.435: 54-3S 9 0.3161 0.693 1.46412.466f.i _ 58-39 9S5 0.318 0.677 1.38Ot2.291 _ 359 95S 10.308 0.679 1.41212.434, 42-5 955 0.308 0.679 1.38412.378 __: 955 0.324 0.703 1.48412.612 f-l
FIML SUMKIIRY REPORT - BFNP UNIT 3 3.0 Resalts (Continued) 3.5 ST1-5, Control Rod Drive System (Continued) 3.5.3 Analysis (Continued) Scram Testinx (Continued) Table STI 5-3 (Continued) Individual Rod Scram Times Sequence B Rods Scm Insertion T es, Sec. Scram Insartin ". Reactor Reactor Drive Press. 5% 20% 50Z 90Z Press. 5% 12o0 SOZ 50 -, Location Psix __I_ PVal I 5447 100 .329 47 2.44 34-7 7 lowoo i 0,348 0.7]03 146 2.48 ,_, 34-9 100 02 1.38 2.32 42-27 1 1000 0.326 0.681 1.40 2 62 38-7 0.324 0.68 1_3_
. 2.41 38-23 11000 i0.321!0.714 1.59 2.68 M?-1... Jnn . LAj2+/- . 2.47 50-27 1000 01 3 1.41 2.43 S4_1 I000, O 14 11 2.60 46 I 1000 ,0.31siO.674 1.44 2.52 50-35 1000 0340 1.2 2.64 38-311 1000 l 0.33410.703' 1.52 2.67 58-35 1000 0.324 0651.40 2.C 30-39t 1000 : 313 1.55 2.62i 4-1l 1000 0.313 0.679 1.46L2.51 38-39 1 1000 3 0.3 1.42 2.52 5011 MO0 0,329 10.66911.34 -2.33 50-43 l 1000 !iO3461 0.732 1.50 2.63 54_15 10, .331 0. 66611.34 2.31 142-43 1 1000 1 20.32 0.701 1.47 Y-3S 4 31 *0-.42!0.685 11Q9 2.38 I214 A2 1D0a Qia3A2LQ. 6Z11.50 .2,.
50-35 -100_ 0.366 0.74611.52 12.57 1,30-23 1000 ! q-3161 -7Q6s Ix-2 .6Z-: 58-35 1000 -0.321 0.656 1.32 2.27 ' 4-1 9 IlOne I0 326 0.ij ~l671A .2-A- -. 42-59 1000 0.313!0.642 1.31 2.27 3- I n I-1007 1 2 _0 ... . 46-55 1000 0.353 0.719 1_42 2.41 42-19 1000 n all-41-46 I010 -52 2,SL - 38-55 1000 0.332 0.711 1.51 2.63 10nn in -4A21 n xR7f i Ai ! I - AA . 30-55 .10 0326 0.703 1.44 Qj251 O. i8MG -3O 424062*3 24 8-M1lo &9k 41 12.46 .1 .. ..1 vMe. . ... 1 0000D33 0.724 1.60 2.75 O-03I 1000 4. n46
..2Z. 1JtE 2..5-02-14 lOOO p.2 0 21 56 1I8-19 3-34 1031000 (0.31 .ll.2 6 1-4o 1.3 - 2hl_
2.29L 1 2-S, o~ L 01,711 7,r]9 2 ,57 06-3L 10 1.342 0.693 1.42 .45 !00 I JAo ,3 L 58v-~5i1a2,D l46-07 02-3S 022hL 100 1 23z9 O. C691.623
- 00. 3501 0.669 1.36 2.3 34-02 1000 !o -v324p6l .49 7.56-18-59 1000 0.337 0.687 2.S 26-59 1000 0.337 0.695 .44 .47 L231looo (.321 0.767l1.5 _. 2.6L 14-55 l1000 0.362 0.738 .40 .56 2-55 1°000 0.324 0.703 .50 .57 42-ll 100lOO .313 A.677i1-.64 i. 5.6 22-07 lOOC 0.313 0.658 .40 12.45 r58-19 100i.31 0.653; 1.32 2.29~
14-07 1000 0.340 .709 .46 .57 54-23 1000 1U.313 0.653'1.15 .2. 2-_ 6-11 0-1020 Q.327 i .Z5 4 58-27 1000 I (0313t0.65311.348 2.307
-20j...32630679 .44 2.4 46-39 100o re.32 0. 64 51 1. ?V 2.44 101Q 35 0.16 47 4 2 54-39
-:.4 1000 A
(D.31310.66911.3. 2.40 0 100 10. 31310 .682 I 46 k-5 25 58-43 I.000 (0.31810.67 11.39 . 2.41 FILMED FROM BEST
FINAL SUHMARY REPORT - BFNP UNIT 3 3.0 Results (Continued) 3.5 STI-5. Control Rod Drive System (Continued) 3.5.3 Analysis (Continued) Scram Testing (Continued) Table STI 5-3 (Contlnued) Individual Rod Scram Times Sequence Z Rods (Continued) Scram Insertion Time Sec. Reactor 1 Drive Press. 5S 20W 50Z 901 Location Psif _ _ Q0-51 1000 _0.3130.642 __-5_ l-59=1 ]1002_ 0.-318 0.693 1.44 24L 0.74 1.53 1flg1 . 2&QQ . 0 *6 1.S2 2.60.
.3 7 .5 . -
18-19_.. 14-1fi 1000... Q0 0032 0.338 0.706 0.732 1.543 1.58 2.eo 12.15 I .__. &-51I 100 C .42 706 1.43 12.49 _- -19_19 100 0.353 .698 .47 .50 14-31 1000 - 0.33. .711 .50 .57s 26-3S 1000 0 .716 .63 2.77 18-35 1000 0.326 .679 1.48 2.55 lI4-7 1000 0.329 .671 L.3 2.36 14-47 1000 0 .313 .661 .39 W .4 _ 22-47 1000 _0.342 .687 .41 2.4 22-23 1000 0.350 .722 .48 .54 _ 26-27 1000 0.358 .733 Q.49 R.56 __ 14-23 1000 0.31 10.700 L.52 .67 18-27 1000 0.318 .714 _.528 _ .__ 06-23 1000 0.313 .6S3 .36 1.35 1 10-27' 1000 0.329 .693 .40 R.35 I 22-31 10 0.340 .701 .52 .61 22-39 1 0 .2 5 76744 . - 14-32 1000 0.3 . 666 .38 .43 10-43 1000 0.329 .42 .43 18-43 1000 0.342. 3 4 .56 .66 26-43 1000 0.345 p.671 .44 .51
FINAL
SUMMARY
REPORT - BFNP UNIT 3 3.0 Results (Continued) 3.5 STI-5, Control Rod Drive System (Continued) 3.5.3 Analysis (Continued) Scram Testing (Continued) Table STI 5-4 Summar of Scram Test Results Mean Reactor Accumulator Number Insertion Times (Sec.) Pressure Pressure Of Rods 5% 20% 50% 90% Tech Spec 0.375 0.90 2.0 3.5 0 Normal 185 0.286 0.511 1.007 1.664 o Hinimum 4* 0.317 0.578 1.13 1.95 600 Normal 4* 0.321 0.661 1.46 2.92 800 Normal 4* 0.350 0.768 1.66 2.90 1000 Zero 4* 0.355 0.763 1.60 2.71 1000 Normal 185 0.327 0.695 1.45 2.51
- Four slowest in-sequence rods.
The scram insertion times of the four selected in-sequence CRD's were measured in conjunction with full-core scrams per STI-75, Reactor Scram From Outside Thc Control Room, STI-27, Turbine Trip, STI-25, Main Steam Isolation Valve Full Isolation, and STI-27, Generator Load Rejection. All applicable criteria were met. The results are summarized in table STI 5-5. EIjMED Fro I3ECP AVAILA1LE CP
FINAL SUIMARY REPORT - BFNP UNIT 3 3.0 Results (Continued) 3.5 STI-5, Control Rod Drive System (Continued) 3.5.3 Analysis (Continued) Scram Testing (Continued) Table STI 5-5 Four In-Sequence Rods Scram Tests Reactor Scram Reactor Power Scram Insertion Times (sec) (%) CRD 5X 20Z 50X 9O:. Tech. Spec. Limit 0.375 0.90 2.0 3.5 STI-75 102 30-27 .343 .770 1.572 2.711 Rx Scram From 18-07 .340 .756 1.620 2.642 I Outside Control Room 26-15 .332 .732 1.3S4 2.732 14-19 .338 .780 1.624 2.805 STI-27 752 30-27 .265 .553 1.18 2.66 Turbine Trip 18-07 .265 .571 1.22 2.16 14-19 .265 .579 1.28 2.25 26-15 .265 .581 1.67 2.11 STI-25 86X MSIV Full Isolation 30-27 .324 .677 1.42 2.55 18-07 .316 .685 1.48 2.64 26-15 .324 .729 1.56 2.74 14-19 .324 .727 1.56 2.74 STI-27 98.5% 18-07 .336 .679 1:484 2.603 Generator Load 26-15 .313 .669 1.432 2.564 ReJection 14-15 .313 .722 1.556 2.758 46-07 .289 .655. 1.468 2.547 .
FINAL SUMMRY REPORT - BFnP UNIT 3 3.0 Results 3.6 STI-6, SRK Performance and Control Rod Sequence 3.6.1 Purpose The purpose of this test is to demonstrate that the operational sources, SIM instrumentation, and rod with-drawal sequences provide adequate information to achieve criticality and to increase power in a safe and efficient manner. The effect of typical rod movements on reactor power will be determined. 3.6.2 Criteria Level 1 There must be a neutron signal-to-noise ratio of at least 2:1 on the requiredoperable SM's or fuel loading chamber prior to pulling rods. There must be a minimum count rate of 3 cps on the required operable SRK's Qr fuel loading chambers prior to pulling rods. The IM's must be on scale bei.'%,-. -ne S exceed the rod block set point. The RSCS shall be operable as specified in the technical specification 3.3.B. 3.6.3 Analysis' STI-6 testing was performed during the open vessel, initial critical and heatup phases, and at test condition 1 as defined on the power flow map in section 2.3. The operational sources were loaded in a manner consistant with STI-3 fuel loading as shown in figure STI 6-1. Prior-to pulling rods the SRK's were demonstrated to have a count rate greater than 3 cps and a signni-to-noise ratio greater than 2:1 by taking count rate 1iata with the detector fully withdrawn and fully inserted. thids data is contained in table STI 6-1. The SRM Hi Hi trips were initially set to 5 x 105 cps.
FINAL SUHCARY REPORT - BFNP UNIT 3 3.0 Results (Continued) 3.6 STI-6, SRM Performance and Control Rod Sequence (Continued) 3.6.3 Analysis (Continued) Prior to pulling rods for the initial critical Rod Sequence Control System (RSCS) was demonstrated to be operable by the performance of surveillance test SI 4.3.B.3-2. This surveillance performs a system diagnostic test and demonstrates that the RSCS will not allow j selection of out-of-sequence rods, thereby assuring compliance with technical specificatic- 3.3.B. The reactor was brought critical in rod sequence B on the 18th notch of the 29th rod (38-15) with a moderator temperature of 920 F. The period was deter-mined to be 132 seconds. The IRK's were shown to be functional, and to overlap with the Sap's. The non-coincident scram circuitry was removed from the SRM's and they were sub-mequenxtly shown not to saturate at a count rate of 7.5 x l0 cps. The reactor was hasted up from atmospheric to rated pressure by pulling control rods in sequence B. j Neutron instrumentation was monitored to insure a safe beat-up rate. The RSCS prevented out-of-sequence rod movement, thus winImizing the worth of individual rods. No anomalies were noticed and control rod sequence B performed acceptably. The reactor was heated up and brought to approx-imately 301 of rated power In sequence A. Performance of control rod sequence A was acceptable. The RSCS was verified to perform properly at 222 and 27Z of rated thermal power as evidenced by the inability to select out-of-sequence rods. The RSCS enforcement interlock cleared at 27.9% of rated thermal power. Level I criteria were met for all phases of STI-6 testing. No level 2 criteria apply.
FINAL
SUMMARY
REPORT - BFNP UNIT 3 3.0 Results (Continued) 3.6 STI-6, SRM Performance and Control Rod Sequence (Continued) 3.6.3 AnalYsis (Continued) Table STI 6-1 SRM Count Rate (cps) SRM Channel A B C I D SRM Fully Inserted 45 45 110 32 SEX Fully Withdrawn .1 .1 .1 1. Signal-to-Noise Ratio 449 449 1099 31
FINAL SUWOARY REPORT - BF"P UNIT 3 3.0 Results 3.7 STI-9, Water Level Measurements 3.7.1 brose The purposes of this test are:
- 1. To check the calibration of the various narrow and wide range indicators.
- 2. To measure the reference leg temperature and recalibrate the wide range instruments if the measured temperature is different than the value assumed during the initial calibration.
- 3. Collect plant data which can be used to investi-gate the effects of core flow, carryunder and subcooling on indicated wide range level.
3.7.2 Criteria Level 1 Not applicable Level 2 The GVMfC indicator readings on the narrow range level system should agree within + 1.5 inches of the average reading.. The wide range level indicators should agree within + 6 inches of the average reading. 3.7.3 Analysis STI-9 testing was conducted at heatup, test conditions 1 and 4E, as defined on the power flow map in section 2.3. Calibrations of the GEMAC and Yarway water level instrumentation were verified to give accurate reactor water level indication at all Times. Graphs of indicated water level versus power (flow constant) and indicated water level versus flow (power constant) were
.)lotted from data accumulated during the startup test program to obtain knowledge of the tracking perforoance of these level systems (refer to figures STI 9-1 and STI 9-2). Note that at high flows, the Yarway level was approximately 13 inches lower than the GEIAC readings due to flow velocity effects on..the Yarway vessel taps.
FILMED FROM BEST AVAILABLE COPY,
FINAL
SUMMARY
REPORT - BFNP UNIT 3 3.0 Results (Continued) 3.7 STI-9, Water Level Measurements (Continued) 3.7.3 Analysis (Continued) At test condition 4E the avegage Yarway reference column temperatures were 2650F and 256 F for columns A and B, respectively. This indicates excellent agreement with the assumed cold water calibration reference leg ter-rera-ture of 264°F. The GCE{C water level indicators read within
+ 1.5 inches of their average reading of 33.5 inches.
All wide range level indicators agreed within + 6 inches of the average reading except for 4 indicators which were one to two inches outside criteria. These 4 indi-cators were recalibrated and verified to meet criteria.
FINAL SUTARY REPORT - BEP L.IT 3 1 , Water Level - VS - Core Pnwer
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FINAL SUHMARY REPORT - BFNP UNIT 3 3.0 Results 3.8 STI-10, IMM Performance 3.8.1 Purpose The purpose of the IBM performance test is to adjust the intermediate range monitor system to obtain an optimum overlap with the SEN and APER systems. 3.8.2 Criteria Level 1 Each IRK channel must be adjusted so that over-lap with the SRH's and APEM's is assured. The IEM's must produce a scram at 120/125 (962) of full scale. Level 2 Not applicable. 3.8.3 Analysis STI-lO testing was conducted at open vessel, initial heatup, and test condition 1 levels as defined on the power flow map in section 2.3. Prior to pulling rods for the initial critical the IBM's were fully inserted and adjusted to give a scram at 120/125 of full scale per surveillance test SI 4.2.C-3. Rods were withdrawn in rod sequence B to bring the reactor critical. All the IMt's were on scale before any of the normalized SRM readings reached the operational limit of 2.0 x 107 cps. All IRK's responded to changes in neutron flux. The reactor was taken subcritical and the non-coincidence scram shorting links were removed. AU appli-cable criteria were met. During the initial heatup, the Int's wera adjusted to correspond to the reactor power level as meaasu ed by the calibrated APRM's. This verifies the IEN/APEM over-lap. Following this adjustment the IRH/SRM overlap was reverified, and surveillance test SI 4.2.C-3 was performed to verify that the TRE's will provide a scram signal at 120/125 of full scale.
FINAL SthK(AP.Y REPORT - BFMM tUNIT 3 3.0 Results (Continued) 3.8 STI-10, ITM Performance (Continued) 3.8.3 Analysis (Continued) With the reactor at test condition 1 (approt:i-mately 30%) the ILM's were adjusted in accordance with surveillance test SI 4.1.B-1 to read consistent with the APM's. All IFM's read equal to or greater than the APR{W 8 . During a subsequent reactor startup satisfactory IRM/SRX overlap was verified. All STI-l0 criteria were satisfied.
FINAL Sn.XARY REPOPRT - Br..P UNIT 3 3.0 Results 3.9 STI-l, UPPM!Calibration 3.9.1 Pur~ose The purpose of STI-l1 is to calibrate the Local Power Range Monitor (LPREM) system. 3.9.2 Criteria Level 1 The meter readings of each LPR4 chamber will be proportional to the neutron flux in the narrow-narrow water gap at the height of the chamber. Level 2 Not applicable. 3.9.3 Analysis STI-li testing was conducted at heatup, test conditions 1, 2E, 3E, and 4E levels as defined on the power flow map in section 2.3. With the reactor at hot standby, LPRW hookup a.>: response was checked in conjunction with STI-5, control rod drive scram testing. Detector 32-49C could not be verified because of upscale failure. All other LPRI's responded satisfactorily to flux changes. During operntic2 at test condition 3E it was discovered that LPRI's 56-33!A and B had their leads reversed. These two LPRM's were bypassed until their leads were correctly connected durinz the next outage. The operable LPRKI's were calibrated at test conditions 1, 2E, 3E, and 4E. This corresponds to pow-er levels of 21X, 52., 76%, and 96% of rated power, respett-ively. The Traversing Incore Probe (TIP) system inter-face with the unit 3 process computer was not operational for the initial LPRM calibration at test condition 1. A full set of tip traces were taken and the data digitized for manual input into the BUCLE offline computer program. The gain adjustment factors (GAF) were calculated by UNCLE and used to calibrate the LPRXS's to read proportional to tMe neutron flux according to surveillance test SI 4.1.3-3. A second TIP set was run and the data digitized and loaded S
FINAL SUMKLRY REPORT - B 1ThIT UP 3 3.0 Results (Continned) 3.9 STI-l1, LPRX Calibration (Continued) 3.9.3 Analveis (Continued) into BUCLE. The GAF's calculated showed 148 of the 169 operable LPRK's reading properly with 21 needing re-calibraticn. Twenty-three IPPR,'s were recalibrated a_= cording to the GAF's calculated by BL'CLE. Following t1his calibration the TIP interface with the process computer was available. .Therefore, a full tip set was loaded izzo the process computer. GAF's calculated by the process computer and BUCLE agreed within + 10%. For LPRK calibrations at test conditions 2E, 3E, and 4E the process computer was used to calculate the GAF's. The calculations at test condition 2E were verified by the offline computer program BUCLE. Agre=e-nt was within + 12. The calibrations were performed ac-cording to surveillance test SI 4.1.B-3. At all tines there were more than 14 operable LPRD1's per APRE channel. This is the minimum number required for an APart channel to be operable. There were 2, 6, and 3 LPRX's inoperable at rest conditions 2E, 3a, and 4E, respectively. The LPER's were adjusted to read proportional to the neutron flux in the narrow-narrow water gap, thereby satisfying all criteria. FILMED .FR'.M BErT AVNALABLE COPY
FINAL SUMAP.Y REPORT - BFNP UNIT 3 3.0 Results 3.10 STI-12, AERM Calibration 3.10.1 Purpose The purpose of STI-12 is to calibrate the Averaoe Power Range Monitor (APRY) System. 3.10.2 Criteria Level 1 The APRM channels must be calibrated to read equ'l to or greater than the actual core thermal power. Technical Specification end fuel warranty li-its on APRM scram and rod block shall not be exceeded. In the startup mode, all APE1 channels must produce a scram at less than or equal to 15% of rated ther.al power. Recalibration of the APRM system will not be :nece-sary from s.afety considerations if at lcast tvo MPEM :zi per RPS trip circuit have readings greater than or equal to core power. Level 2 If the above criteria are satisfied then the APRX channels will be considered to be reading accurately if t'hey do not read greater than the actual core thermal power by more than 72 of rated powzer. 3.10.3 Analysis STI-12 testing was performed at heat up, test conditions 1, 2E, 3E, and 4E levels as defined on the power flow map in section 2.3 Prior to pulling rods for the initial startup the APRP's were set to scram at < 15% and to give a control rod withdrawal block at < 12% by the performance of surveillance test SI 4.2.C-1. Initially the APER's were calibrated based on the low power heat balance calculated using the heat-up rate. The heat-up rate was measured to be approximately 700 F/hr. Gain adjustment factors were calculated for each APMX, and a
FINAL SLMXARY REPORT - BF P UNIT 3 3.0 Resu~'ts (Continued) 3.10 STI-12, APsy Calibration (Continued) 3.10.3 Analysis (Continued) the APRM'Is were then adjusted to read 4.7% of rated -;:. This value was determined, based upon the highest APERN reading with a 0.3% margin for calculation inaccuracies. At test conditions 1, 2E, 3E, and 4E, the APIT-!fs were calibrated to read equal to or greater than the actual core thermal power. The core thermal power was obtained from the process computer heat balance program (OD-3). The program was verified by the offline heat ba--e (CORPWR) and by a detailed manual heat balance. The ArR-` were recalibrated following each LPPR1 calibration. All calibrations were performed according to surveillance tec. SI 4.1.B-2. For each test condition a scram clamp was set at 20% above the nominal load line of that plateau. Immediately after en APRX calibration at test condition 4E, power was reduced to approximately 40% us1 -_ core flow and control rods and returned to the initial power level (approximately 95%). During this power ran process computer heat balances (OD-3) were run to monitcr :-:. ability of the APlUI's to track the core power level. adjustment factor's for each APPM remained less than 1.0 throughout the power ramp. All applicable criteria for STI-12 have been satisfied at each test condition. Typical results of this APM2 tracking test are shown on figure STI 12-1. II-MED FROM BESr AVAILABLe Copy
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FINAL SLARY REPORT - BFNP UNIT 3 3.0 Results I.11 STI-13, Process Computer 3.11.1 Purpose The purpose of STI-13 is to verify the peraore-ce of the process computer under plant operating conditions. 3.11.2 Criteria Level 1 Not applicable Level 2 Program OMl and Pl will be considered operaticr.' when:
- a. The MCPRcalculated by BUCLE and the process computer either:
- 1) Are in the same fuel assembly and do not ci.'er in value by more than 27;, or
- 2) For the case in which the '.CPR calculated by the process computer is in a different asse=-
bly than that calculated by BtCLE, for each assembly, the MCPR and COR calculated by the r:o methods shall agree within 2%.
- b. The maximum LHGR calculated by BUCLE and the process computer either:
- 1) Are in the same fuel assembly and do not differ in value by more than 2%, or
- 2) For the case in which the maximum LHGR cal-culated by the process co~puter is in a diV'e-er.e assembly than that calculated by BUCLE, for each assembly, the maximum LEGR and LHGR calculated the two methods shall agree within 2%.
- c. The MAPLEGR calculated by BUCLE and the process computer either:
- 1) Are in the same fuel assembly and do not di-ffer in value by more than 2%, or
- 2) For the case in which the MAPLEGR calculated bly the process computer is in a different assemblv I
FINAL SUMARYF's REPORT - BFNP tMiIT 3 3.0 Restlis (Continued) 3.11 STI-13, Process Computer (Continued) 3.11.2 Criteria (Continued) Level 2 (Continued)
- c. (Continued)
- 2) than that calculated by BUCLE, for each as!.-
bly, the MAPLHGR and X.PLPGR calculated b- tc.e two tmethods shall agree within 2%.
- d. The LPRX calibration factors calculated by the independent nrethod and the process computer agree to within 2%.
- e. The remainir.g programs will be considered oper-ational upon successful completion of static s.-_
dynamic testing. 3.11.3 Analvsis Process computer testing was conducted during 5s ::o vessel, heacup, atid tcst COudiLinOMI 1 arn 4g.. Ta&e am .: was re-initialized at 1830 on October 12, 1976, for the beginning of the dynamic testing. The dynamic system test case was completed at 51.1% power and 102.7% flow with the exception of minor testing on subsidiary programs. The manually calculates -eve balance agreed to within 0.77 of the OD-3 calculated beat balance. The offline program BUCLE and P1 were comparcd and all the thermal limits agreed to within 0.2%. Core thermal hydraulic calculations, exposure calculations, nd exposure updating were verified as being correct by ccnt£rr; with manual calculations or BUCLE. LPRM calibration factors as calculated by the process computer and BUCLE agreed l,"t-hin 1%. See table STI 13-1 for comparison of process computsr and BUCLE results.
- FINAL
SUMMARY
REPORT - BFP UNIT 3 3.0 Results (Continued) 3.11 STI-13, Process Computer (Continued) 3.11.3 Analysis (Continued) Table STI 13-1 Comparison of Process Computer and BUCLE Results Process Variable Symbol Computer RUCLE 1% Difference Critical Power Ratio MCPR 2.431 2.431 0% Linear Heat Generation Rate HLEGR 6.003 6.017 0.23% Average Planar Heat Generation Rate MAPLEGR 5.05 5.06 0.2% NOTE: The core locations of MCPR, MI.GR, and .!APLHGR limits were the same as calculated by the process computer and BUCLE. i
FINAL St%{ARY REPORT - BFNi UNIT 3 3.0 Results 3.12 STI-14, Reactor Core Isolation Cooling System 3.12.1 Purpose The purpose of this test is to verify the proper operation of the reactor care isolation cooling system over its required operating pressure range. 3.12.2 Criteria Level 1 The time from actuating signal to required f2.o-w must be less than 30 seconds at any reactor pressure bat.zea 150 psig and rated (1020 psig). With pump discharge at any pressure between 150 psig and 1220 psig, the required flow is 600 gpm. (The limit of 1220 psig includes a nominally high value of 100 psi for line losses. The measured value may be used if available.) The RCIC turbine must not trip off during startu:i. Level 2 The turbine gland seal condenser system shall we capable of preventing steam leakage to the atmosphere. The LP switch for the RCIC steam supply line high flow isolation trip shall be adjusted to actuate at 30C0. oi the maximum required steady state steam flow. For small speed or flow demand changes while ir.- jecting into the reactor vessel in either manual or auto-matic mode, the decay ratio of each recorded RCIC system variable must be less than 0.25, in order to demonstrate acceptable stability. The maximum RCIC turbine speed during Buick starcs shall be at least lO below the overspeed trip setting. 3.12.3 Analysis STI-14 testing was conducted at heatup and test condition 1 as defined by the power flow map in section 2.3 The RCIC system demonstrated under all test conditions the ability to reach rated flow in less than 30 seconds. After S
FIŽNAL SUa-tARY REPORT - BFNP UNIT 3 3.0 Resuls (Continued) 3.12 STI-14, Reactor Core Isolation Cooling System (Continued) 3.12.3 Analysis (Continued) running the rated pressure test, the system response -;.as improved by lengthening the control system ramp stroke tine from 14 to 19 seconds. After the adjustment the three test points were repeated. The results of these three tests during heatup and the cold quick start reactor vessel injection.test are presented in table STI 14-1. Required system flow of 600 gpm was reached at all test conditions and the RCIC turbine did not trip. T-e turbine gland seal condenser system prevented steam leakage. The high steam flow isolation switch trip was conservativelsv set to actuate at < 450 inches of water per the technical specifications. All process variables exhibited a decay racks of less than .25. The naxi=um RCIC turbine speed during tha quick start test has 4375 rpm which is more than 10% below the overspeed trip setting. During each test condition it was noted that the barometric condenser did not develop a sufficient vacumn. Repair work to the vacuum pump is pending arrival of parzs to improve vacuum pump performance. The RCIC high stei- Flo-. switches were found to have a required setpoint (calculated via field data) greater than the installed instrument ran-e of 500 inches of water. G.E. Design Engineering evaluated the data and calculated the setpoint to be 1064 inches of water. Final resolution to the problem is pending TVA's review. Experience has shown that after extended periods of idleness, the margin to the RCIC turbine overspeed setpoint -.y-be reduced on a cold quick start. The reason for this is that the Woodward actuator receives its oil supply from a separate sump, resulting in a starved oil supply actuator. A nodi-fication to the auxiliary oil sump in the oil supply line to the Woodward EG-R hydraulic actuator has been specified. Based on the observed system operation and the transient recordings, it was concluded that RCIC was fully operational. The final RCIC controller settings are as follows: Proportional Band: 600 Resets per Minute: 100 Ramp Time: 19 seconds Ramp Idle: -0.5 volt EGR Needle Valve; 1/2 turn ccw FILMED FROM BEST AVAILABLE COPY
FINAL St'MMAIY REPORT - ENP UNIT 3 3.0 Results (Continued) 3.12 STI-14, Reactor Core Isolation Cooling System (Continued)
- 3.12.3 Analysis (Continued)
Table STI 14-1 Results of RCIC Tests Pump Discharge Turbine i Test 2!easured Reauired Reactor Pressure Speed Ce.ntr .- Condition Flow Time Flow Time Pressure X4easured Required S.S. PeakX i f:. _ _ _ (se) pm psig _ j psig rpm rpm Reatup 600 9.75 600 3C 140 230 240 2000 20000 :- ^: _ __ ._ _ _ I Heatup 612.5 16.5 600 30 590 710 690 3300 3875i I C:oc* 612.5
-eatup 18.75 600 30 980 1120 1080 4010 43751 lCv _^
Heatup 618.0 19.75 600 30 980 1220 1220 4035 43751 I
. _ _ _ _ _ _ __ _ __ I _ i_ _ _ _ :__
1 610 20 600 30 954 1010 NIA 3900 4125i lC0 E0CZ RCIC electrical turbine trip setpoint: 4950 rpm
FINAL SLTIMARY REPORT - EFNP UNIT 3 3.0 Results 3.13 STI-15, Righ Pressure Coolant Injection System 3.13.1 Purpose The purpose of this test is to verify the proper operation of the high pressure coolant injection system over its required operating pressure range. 3.13.2 Criteria Level 1 The time from actuating signal to required flow must be less than 25 seconds at any reactor pressure bet.een 150 psig and rated. With pump discharge at any pressure between 1x3 ;sg-and 1220 psig, the flow should be at least 5000 gpm. (The limit of 1220 psig includes a nominally high value of lOC psi for line losses. The measured value nay be used, if avaiic1e.) The HPCI turbine must not trip off during starrtup. Level 2 The turbine gland seal condenser system shall be capable of preventing steam leakage to the atmosphere. The S switch for the HPCI steam supply line high flow isolation trip shall be adjusted to actuate at 225/ I the maximum required steady-state steam flow. For small speed or flow demand changes while in-jecting into the reactor vessel in either manual or automatic mode, the decay ratio for each recorder HPCI system variable must be less than 0.25, in order to demonstrate acceptable stability. The maximum HPCI turbine speed during quick starts shall be at least 10% below the overspeed trip setting. 3.13.3 Analysis STI-15 testing was conducted at heatup and test condition 2E as defined on the power flow map in section 2.3. During the heatup testing phase, the High Pressure Coolant Injection system (EPCI) took suction from and discharged to the condensate storage tank. The first test at 150 psig i
FINAL SlkO(ARY REPORT - BFNP UNIT 3 3.0 Results (Continued) 3.13 STI-15, High Pressure Coolant Injection System (Continued)
- 3.13.3 Analysis (Continued) was repeated because the test valve (73-35) was not de-energIzed. Since the valve was shut, the discharge pressure continued to climb to 1200 psig after reaching rated flou; when the valve was opened. The system performed satisfactorily during the second test at 150 psig.
The 1100 psig pump discharge pressure test was repeated due to a slow opening time on the HPCI stop valve. To increase the opening time, the ramp generator stroke time was changed from1'4 to 12 seconds, and the test was repeated successfully. Observed pump performance was wit.:- in the tolerance of the vendor ptmp performance results. The final controller settings on HPICI were as follows: Proportional Band: 600% Reset per Xinute: 100% Ramp Generator Stroke Time: 12 seconds Ramp Idle: -0.5 volts EGR Needle Valve: 112 turn Cc' The maximum time required to reach 5000 gp-i a. any reactor pressure between 150 psig and rated was < 24 seconds; the EPCI system flow was > 5000 gpm at all pressures between 150 psig and 1220 psig; and the turbine did r.ot :riD off during testing. This satisfied all level I criteria. The turbine gland seal condenser system prevented steam leakage to the atmosphere. The decay ratio for each recorded HPCI system parameter was < .25 for a 5% flow s=_? change while injecting to the vessel. Using the steady-state steam line ZP indicator readings, the calculated steam line high flow trip sett-n.,s were greater than the maximum instrument range (100 psi,) an2 greater than allowed by technical specifications. GE Engineering Design has evaluated the data and determined that the differential pressure setpoint should be 114 psi. Final resolution to the problem is pending TVA DED review. The HPCI turbine speed peaked at 4700 rpm during the vessel injection test due to an air pocket formed beneath the stop valve hydraulic oil piston. The result was that the stop valve initially spiked open and then returned to its no-ral opening ramp. An ECM to correct this problem by rerouting th oil line to the stop valve hydraulic actuator was apprcve awaits receipt of the necessary materials. i
FINAL SU1MARY REPORT - BFNP UNIT 3 3.0 Results (Continued) 3.13 STI-15, High Pressure Coolant Injection System (Continued) 3.13.3 Analysis (Continued) The final results of testing performed at each test condition is presented in table STI 15-1. Table STI 15-1 Final Results of BPCI Testing Pump Test Measured Recuired Reactor Discharge Press.l Turbir.e - Condition: Flow Tine Flow Time Pressure Actual Required Maxiran. e Date (gpm) (sec) (gpM) (sec (psig) (psig) (psig) (rpm) (r;--.) Heatup 8124/76 5062 14.5 5000 < 25 800 J 890 900 4190 35Q30 Heatup 8/28/76 5050 17.3 5000 ' 25 161 300 250 2560 2375 Heatup 8/29/76 5060 23.5 5000 c 25 4000 1100 1100 4470 3'844 Heatup 8/29/76 '5125 23.75 5000 < 25 1000 1200 1200 4500 C^00 T.C. 2E 10/17/76 5000 24 5000 < 25 950 1050 1050 4700 3SC0 T.C. 2E 10125/76 5000 24 5000 < 25 930 1030 1030 4650 3750
FINAL SUMARY REPORT - BFnP UNIT 3 3.0 Results 3.14 STI-16, Selected Process Temperatures 3.14.1 Purpose The purposes of STI-16 are:
- 1. To establish the proper setting for the low speed limiter for the recirculation pumps.
- 2. To provide assurance that the measured bottom head drain temperature corresponds to bottcm head coolant temperature during normal operations.
3.14.2 Criteria Level 1 The reactor recirculation pumps shall not be operated unless the coolant temperatures between the unpcr and lower regions of the reactor vessel are within 145? (800C). Level Z The bottom head coolant temperature as measured by the bottom drain line thermocouple should be within 50 0 F (28 0 C) of reactor coolant saturation temperature. 3.14.3 Analysis STI-16 testing was conducted at heatup and test conditions 2A, 2E, and 4A as defined on the power flc': map in section 2.3. The results for selected process temperatures for all the test conditions are presented in table STI 16-1. Note that in natural circulation the flow is insufficient to maintain the bottom drain line temperature and reactor coolant saturation tempera-ture within 50 F. Since steadv state operation without forced recirculation is not permitted by the technical specifications, except during the startup testing, this criteria does not apply to natural circulation. The difference between the bottom head drain line temperature and the reactor coolant saturation temperature was 79 F during single recirculation pump trips at test condition 4E. This does not meet the level 2 criteria and the problem will be resolved during the first refueling outage. S
FINAL SUKARY REPORT - BT? UNIT 3 3.0 Results (Continued) 3.14 STI-16, Selected Process Temperatures (Continued) 3.14.3 AnaLysis (Continued) Table STI 16-1 Summary of Temperature Behavior ( 0F) Test Condition Heatup 2A 2E 4A AE upd "B"Tine<': Pump Discharge Temp. A 530 513 528 505 500 524 i Pump Discharge Temp. B 530 513 529 505 511 513 1 Saturation Temp. 544 539.6 540 538 539 539 Rx. Bottom Head Drain Temp. S00 478 501 460 461 460 AT (Disch. - Bottom Drain) 14 35 27 45 39, 50 64, 53 AT (Sat. - Bottom Drain) 44 61.6* 39 78* 78 79
*Level 2 criteria not applicable in natural circulation.
FINAL SUT.^.!ARY REPORT - BrF\' UNIT 3 3.0 Results 3.15 STI-17. System EKpansion 3.15.1 Purpose The purposes of STI-17 are to:
- 1. Verify that the reactor drywell piping system is free and unrestrained in regard to thermal e*:-
pansion.
- 2. Verify that suspension components are functioning in the specified manner.
- 3. Provide data for calculation of stress levels in nozzles and weldments.
3.15.2 Criteria Level 1 There shall be no evidence of blocking of the displacement of any system component caused by thermal expansion cf thie systen. Hangers shall not be bottomed-out or have the spring fully stretched. Hydraulic shock and sway arrestors shall be set to within + 1 inch of the defined setting. Electrical cables shall not be fully stretched. Level 2 Displacements of instrumented points with SpeCial recording devices shall not vary from the calculated values by more than + 30 percent or + 0.5 inch, whichever is smaller. Displacements of less than 0.25 inch can be neglected, since 50 percent of this value is bordering on the accuracy of measurement. If measured displacements do not meet these Criteria, the system designer must be con-tacted to analyze the data with regard to design stresses. The trace of the instrumented points during the heatup cycle shall fall within a range of 150 percent of the calculated value from the initial cold position in the direction of the calculated value, and 50 percent of the calculated value from the initial position in the opposite direction of the calculated value.
FINAL SUMM1ARY REPORT - BFNP UNIT 3 3.0 Results (Continued) 3.15 STI-17, System Expansion (Continued) 3.15.2 Criteria (Continued) Level 2 (Continued) Bangers shall be in their operating range (batween the hot and cold settings + 10 percent). Hydraulic shock and sway arrestors shall be with-in their operating range. If the operating range is not available, verify that there is a minimum of 1" stroke Iit for the piston. Conduit connections shall remain flexible (no tig;X linear or axial junctions). 3.15.3 Analysis STI-17 testing was conducted during open vesse, heatup, and test conditions 1 and 4E as defired via t;= power flow map in section 2.3. ThIarral expansion data fo:- the reactor drywell piping system was obLulned by aculua observations and by lanyard potentiometers. In general, the drywell piping moved in the correct direction during heatup and returned to its base setting after cooldown. There was no evidence of blocking of the dis-placement of any system component caused by thermal expansion of the system at any temperature level. There were no preselected hangers found to have their springs bottomed-out or fully stretched at any temperature level. At ambient and 3 00 °F all hydraulic shock and sway arrestors were found to be within +1 inch of the defined setting; however, in all three heatups, some of the feedwater pipe movements did not satisfy level 1 criteria. A more extensive compilation of feedwater expansion data was sent to TVA's engineering design for review and the expansion was judged to be acceptable (refer to attachment number 1). The hydraulic shock ar.d sway arrestors on all other systems fell within + 1 inch of their designed setting during the three above mentioned heatups. No electrical cables were found to be fully stretched, S
FINAL SUMA1RY REPORT - BFP UNIT 3 3.0 Results (Continued) 3.15 STI-l7, System Expansion (Continued) 3.15.3 Analysis (Continued) Displacements of instrumented points with special recording devices did not vary from the calculated valtues by more than + 50% or + 0.5 inches, whichever was sma1ler. Exceptions to the criteria were resolved at the heat'?p test plateau (refer to attachment number 2). The traces of the instrumented points during the heatup cycle fell within 150% of the calculated value frc. the initial cold position in the direction of the calcui :e_ value and within 50% of the calculated value in the opposite direction. Exceptions to this criteria were specific !,Cints on the recirculation lines and the "A" and "B" feedwater l'In'z; however, the recirculation exceptions were eventually resoiv and the feedwater exceptions were cleared as the feedaater system reached rated temperature (3780 F). All hangers were found to be between their not z-n cold settings + 10 percent with the exception of one -e: hanger. This hanger was deemed acceptable after e:zhibiz:.-- correct movement at upper feedwater temperatures. All hydraulic shock and sway arrestors were within their operating range. All conduit connections remained fully flexible. Three complete heatup cycles were completed on 8/2/77, 11116/77,and 12/27/77. The comparison of these three cycles indicated that the pipe movements were apprcxi-mately the same for all three cycles. Movements that deviated slightly from calculated were deemed acceptable by piping design. Table STI 17-1 summarizes the results of the dis-placements at rated temperature for the three cycles. Attachment 3 shows the location of the instruments monitore: during the heatups. All Level I and Level II criteria have been ret fcr STI-17 testing.
FIDNAL SMU~LRY REPORT - BF;?P UN'IT 3 Table STI 17-1 Displacements at Rated Temperature _ A _ . Cycle 1 Cycle 2 C%'cle 3 x .408 .379 347 Recirc. A y .054 .033 .032 Suction z -.410 -.390 -. 350 Recirc. A X -.670 -.539 -.653 Discharge Y -.848 -.646 -.863 Z -.430 -.500 -.360 Recr. B X .096 -. 019 .090 Y -. 560 -. 297 -. 539 Suction Z -1.520 -1.190 -1.450 Recirc. B x -.907 -.869 -.900 Suction Y -.190 .065 -.159 Z -. 290 -.380 -.220 Recirc. B 2 .124 .140 .627 Discharge Y 1.030 .789 .807 Z -. 440 -. 100 -. 380 Recirc. B X -. 954 -.789 -. 807 P Y .599 .596 .627 z -1.480 -1.440 -1.470 Feedvater A X .912** 1.253 1.559 Z .512** .742 .901 Feedwater 8 X .654** .975 1.179 Z -.092** -.518 -.657
- Data taken at 2680 F.
FINAL SUMi4ARY REPORT - BFN:P LNIT 3 Table STI 17-1 (Continued) Displacements at Rated Temperature Cycle 1 Cycle 2 Cycle 3 X .536 .593 .738 tain Steam A
.102 .125 .157 Lower
.850 .71Q .910 Main Steam A x 1.674 1.335 1.975 Upper 1.192 .936 1.381 2 .200 .020 .050 I
Main Steam xX .855 .943 1.023 Lower Y .565 .518 .668 z .680 .570 .500 II I Main Steam B x 1.431 1.541 1.656 i Upper Y .966 .903 1.060 I Z .260 .100 .070 Main Steam C X 1.233 1.310 1.418 Lower Y -.806 -.866 -.877 Z .510 .520 .250 Main Steam C X 1.790 1.561 1.953 Upper Y -1.464 -1.419 -1.462 z .290 .120 -.020 Main Steam D x 1.578 1.630 1.631 rower Y .033 -. 062 .089 Z .750 .640 .740 MaIn Steam D X 1.929 2.229
- Upper Y -1.066 -1.041 -1.206 Z .140 .340 .090
- Failed potentiometer
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FINAL S=L!ARY REPORT - BFNP UNIT 3 3.0 Result Is r 3.16 STI-18., Care Power Distribution 3.16.1 Purpose The purposes of STI-18 are to:
- 1. Confirm the reproducibility of the TIP system readings.
- 2. Determine the core power distribution in three dimensions.
- 3. To determine core power symmetry.
3.16.2 Criteria Level 1 The total TIP uncertainty (including random noise and geometrical uncertainties) shall be less than 7.8T. This total TIP uncertainty will be obtained by averaging the total uncertainty for all data sets obtained. A minimum of twco data sets is sufficient for the determination of total TIP un-certainty. Hovever, if the first two data sets do not meet the above criteria, testing may be continued and up to 6 data sets obtained and compared with the criteria. If the 7.8% total TIP uncertainty criteria has not been met by the 6 sets of data, testing may continue and additional data sets be obtained provided (a) the Y.CPR limit is adjusted to reflect the TIP uncertainty determined by the 6 data sets, (b) the NWC is Informed of the adjusted 1{CPR limit, Wc) the data generated from the 6 sets of data is transmitted to the NRC, and (d) IVA's intentions for continuing to test and expand the data base is provided to NRC. If the total TIP uncertainty is reduced by taking additional sets of data to expand the data base, t~pe MCPR limit vill be adjusted accordingly until the 7.8g total TIP uncertainty is met. At this time, the MCPR limit will be returned to its original value. Level 2 Not applicable LADLE V 1I .3.16.3 Analysis TIPs sets were run at test conditions 1, 2E, 3E, and 4E to provide the process computer with proper base LPIC{
=;:I-a -- - ..
data. and to analyze the core power symmetry. Table STI 13-1 shows an axial (Z) distribution for each of eight radial (FR) rings. The core bundle power maps were inspected, and no
FIKAL SMTNARY REPORT - BMM. UNIT 3 3.0 Results (Continued) 3.16 STI-18, Core Power Distribution (Continued) a 3.16.3 Analysis (Continued) . analyzing 20 TIP traces in the common TIP channel, and the geometric uncertainty found from the analysis of TIP traces from symmetric TIP locations in accordance with the =ethocs outlined in section 7.0 of the startup test instruction. The program "TIPTWO" was written to handle the calculations. The results of the test are outline in table STI 18-2. The total noise uncertainty (ototal) was belong the allowable 7.8% at both test conditions, easily satis-fying the test criteria. Table STI 18-2 2E 3E Limit a . 2.61Z 3.99% < 7.80Z a (random) 1.26% .595% N.A. a (geometric) 2.28X 2.76? N.A.
- 73 FAL SUMHARY REPORT - EFNP U.IT 3 3.0 Results (Continued) 3.16 STI-18, Core Power Distribution (Continued) 3.16.3 Analysis (Continued) anomalies were found. Figure STI 18-1 shoes the radial power distribution (bundle powers in KW'r) for one quadrant of t'e core.
Table STI 18-1 95
- R - Z Power Distribution __,_
LYL 1 2 3 4 5 6 7 8 AVG. Core Top 12 0.215 0.282 0.267 0.252 0.252 0.263 0.232 0.156 0.223 11 0.384 0.543 0.513 0.474 0.409 0.515 0.467 0.307 0.434 10 0.546 0.778 0.731 0.666 0.571 0.737 0.681 0.449 0.622 9 0.675 0.982 0.920 0.836 0.719 0.9371 0.869 0.571 ,7SI 8 0.834 1.224 1.14? 1.041 0.901I1.1 1S04 O.7265 O.993 7 0.969 1.419 1.332 1.212 1.052 1.382 1.296 0.8591 1.164 6 1.106 1.643 1.542 1.400 1.217 1.599 1.519 1.0071 1.354 5 1.139 1.681 1.581 1.417 1.238 1.636 1.564 1.027j 1.384 4 1.132 1.653 1.572 1.401 1.222 1.612 1.545 1.002i 1.363 3 1.182 1.737 1.540 1.513 1.328 1.669 1.516 1.006 1.397 2 1.148 1.736 1.427 1.565 1.388 1.654 1.404 0.940 1.361 Core Bottom 1 0.665 1.190 0.982 1.084 0.949 1.142 0.943 0.568 0.912 AVG. 0.833 1.239 1.129 1.072 0.937 1.194 1.0951 0.718! 1.000 At test conditions 2E and 3E, additional TIP traces were run to verify that the TIP signal uncertainty was baeou the aUowable criteria. The random noise (orn) was found by
FINAL SUINARY REPORT - IFNP UNIT 3 31 7=
.- r- - - U
- 2. 3.72 14.35 4.98 15.02 [4.72 3.13 13.07 4.61 15.05 ,.. 4 29
" -U q. . U Z.61 3.73 1 4.37 4.95 5.01 5.06 4.61 4.46 4.83 5.06 5.05 5.08 5.19 p5.22 27-~
2.48 3.79 4.33 5.07 5.30 4.81 4.44 4.36 4.68 4.84 4.61 4.51 4.86 *5.C3 25 - . mm 2.45 3.70 4.09 4.75 24.7 4. 3.03 .99 4 .33 4 45 3.06 3.02 4.53 04.93 3* 0 a +/- 23-2.37 3.49 3. 9342 4.7 4.3 3.29 2.98 4.27 4.48 3.0 2.5 4.4 !5.s,; J* . 21-I427 _ 7 w 7 l 2.16 3.33 3.98 4.44 4.80 S.10 4.81 4.38 4.57 44.86 4.47 4.62 t5 t 19 1.63 2.92 3.54 4.38 4.7 2 5.14 45.05.08 4.54 4.22 14.30 4.64
- 4. -
1 4.77 _ _ - l _ 2.32 3.25 4.09 4.54 4.97 4.f6 e4.7 5 5 .05 4.38 4.29 '.47 15
'44 4 2.40 3.46 3.99 4.58 4.65 4.85 5.15 4.86 3.2 3.07 i4.49 '4.42 13 -..... ' - u~m. :
3.56, 4417 4.63 1500 -5.17 5.18 4.83 4.57 4.83 14.B4 99 - h.r I.99 3.64 4.10 4.80 4.98 4 4.98 4.98 5.03 '5.38 I- A - - I _ . . ., s I p.
%01 2.73 3.581 436' 4.65 4.69 4.56 4.54 14.83 e I Ib I 07 1.94 2. 3.33 2.54 I3 3.64 4.08 3.97 4.15 4.41 44 3
[ j2.98 31,414 3.56 3 81 3 87 329 1.6 2 10 .92 2.55 2.61 2.621 el Im 00 02 I II.1 ok 06 08 10 I I I 11 I I 1 12 1i4 16 18 20 22 24 26 Fixure ST 18-1 S . Bumdle Power (SWt) Map at 95Z Power
- FINAL SU.%ARY REPORT - BFlW UNIT 3 3.0 Results 3.17 STI-19, Core Performance 3.17.1 Purpose The purposes of STI-19 are:
- 1. To evaluate the core thermal power.
- 2. To evaluate the following core performance parameters:
Maximum Linear Heat Generation Rate (MLHGR) Minimum Critical Power Ratio (.!CPR) axt.mum Average Planar Linear Heat Generation kate CHA.LHGR). 3.17.2 Criteria Level 1 The maximum linear heat generation rate (L1GP.) of any rod during steady-state conditions shall not exceed the limit specified by the technical specifications. Steady-state reactor power shall be limited to 3293 Slt and values on or below the design flow control line (defined as 3440 Mt with core flow of at least 102.5 x 100 lb/hr.) The minimum critical power ratio (HCPR) shall not exceed the limits specified by the technical specifi-cations. The maximum average planar linear heat gene-ration rate (NAPLEGR) shall not exceed the limits of the technical specifications. Level 2 Not applicable. 3.17.3 Analysis STI-19 testing was performed at test conditions 1, 2A, 2D, 3E, 3C, 3D, 3E, 4A, 4C, 4D, and 4E as defined on the power flow map as shown in section 2.3. The core performance parameters; linear heat generation rate (LHGR), core thermal power (CTP), minimum 0
FINAL SUMHLARY REPORT - BFNP UNIT 3 3.0 Results (Continued) 3.17 STI-19, Core Performance (Continued) 3.17.3 Analysis (Continued) critical power ratio (XCPR), and max-Imum average planar linear heat generation rate (MAPLHGR), were monitored at each test plateau of the startup test program. Table STI 19-1 contains a summary of these core parameters compared.to the criteria limit. All calculations were performed using the plant process computer. Core thermal power calculation of the process computer was verified using an offline computer program (CORPMR), and a detailed manual heat balance. Core performance parameters (LEGR. XICPR, MAPLHGR) calculated by the process computer were verified by the offline program BUCLE. All calculations agreed within the re-quired 2I. All test criteria have been satisfied. Table STI 19-1 Core Performance Parameters Test Core LEGR MCPR MAPLHGR Condition Power (MHtW Value Limit Value Limit Value Limit 1 768 3.82 <13.3 3.493 >1.514 3.20 cll.I 2L 783 4.275 <13.36 3.133 >1.572 3.53 <11.15 2D 1544 5.963 <13.26 2.486 >1.328 4.94 2E 1689 6.19 <13.27 2.428 >1.27 5.20 '11.13 3C 1536 6.98i <13.36 2.178 >1.445 5.80 <11:14 3D 2136 8.92 <13.35 1.805 >1.315 7.49 <11.15 3E 2502 9.56 <13.275 1.659 >1.270 8.03 '11.15 1 4A 1329 5.427 '13.36 1.9605 >1.566 4.50 <11. 20 4C 1902 7.625 <13.24 1.6789 >1.454 6.40 '11.19 i 4D 2309 10.35 <13.35 1.665 >1.311 8.75 <11.21! 4E 3173 12.26 <13.35 1.4259 >1.270 10.36 <11.221 b I
FINAL SIMMARY REPORT - BFWP UNIT 3 3.0 Results 3.18 STI-20, Steam Production 3.18.1 Purpose The purpose of STI-20 is to demonstrate that the Nuclear Steam Supply System (NSSS) is providing sufficient steam to satisfy all appropriate warranties. 3.18.2 Criteria Level 1 The NSSS parameters as determined by using normal operating procedures shall be within the appro-priated license restrictions. The appropriate warranty requirements, as out-lined here, shall be satisfied. The nuclear steam supply system shall be capable of supplying ste&.1, of not less than 99.7X quality at . pressure of 985 psia at the second isolaticn valve. 1a-> system shall supply a maximum continuous steam flow cu-- put of 13,422,000 pounds per hour contingent upon tc!e feedwater flow being 13, 372,000 pounds per hours at 3780 F., and CRD flow being 50,000 pounds per hour at ZOQO A. Level 2 Not applicable. 3.18.3 Analysis Warranted plant conditions were attained on December 26, 1976, and the start of the warranty dem.on-stration was officially declared at 2230 hours. The war-ranty demonstration was officially declared completed on January 8, 1977, at 1400 hours after 303.5 hours of oper-ation. The 300-hour warranty run was interrupted twice for routine weekly control valve surveillance testing for a total of 3.5 hours. This time was not included in the 300-hour accumulation. Reactor power was raised as close as possible to its rated value of 3293 )Mt, such that during the warranty demonstration the average reactor power was 99.51Z. Hence, 4.9ffi for the two 2-hour runs it was necessary to extrapolate the plant conditions to the conditions of the contract. During a
FMNAL S"MIARY REPORT - BEnP UVNT 3 3.0 Results (Continued) 3.18 STI-20. Steam Production (Continued) 3.18.3 Analysis (Continued) the 4-hour precision test runs the average main stear. '1ot:, adjusted to contract conditions, was 13.4155 x 106 lb/hr. Uncertainty calculations determined that the uncertai.:; in measured feedwater flow (parameter which mainly affects steam flow) was + 0.02745 x 106 lb/hr. This made the un-certainty in steam flow calculations to be 13.4155 +
.02745 x 106 lb/hr and the contract specification of 13.422 x 106 lb/hr was satisfied.
All core performance parameters were within lihirs throughout the 300 hours. The following table is a su.ar. of the two hour precision test runs and the average of th-. process computer data accumulated for the 300-hour duratic-,. Table STI 20-1 Parameter Rated Run 1 Run 2 300 hr. Aye. Main Steam Flow 13.422 lab/hr 13.234 13.266 13.236 - Feedwater Flow -13.372 X1b/hr 13.195 13.228 13.2;4 CRD Flow .050 Mib/hr .039 .038 .036 Recirc Pump PWR 10.52 XW 8.803 8.24 10.04 Rx Water Cleanup Loss 4.3 PW 2.061 0.0 2.53 Fixed Loss 0.6 KW 1.0 1.0 1.0 Reactor Thermal PWR 3293 MWt 3271 3281 3277 Feedwater Temperature 3780 F 372.5 371.4 373.15 Reactor Dome Pressure 1020 PSIA **1019 **1019 *1032 Steam Quality @ 2nd MSIV 99.72 DRY 99.84 99.86 N/A Steam Pressure 2zLd 4SIV 985 PSIA 995 995.8 N/A Steam Flow ( Contract Conditions 13.422 MNb/hr 13.411 13.420 NJA
*Station Instrument **Test Dead-Weight Gauge 11 a,
8.
FINAL SL'.-ARY REPORT - BFN I 3 UwrT 3.0 Results 3.19 STI-21, Flux Response to Pods 3.19.1 Purpose The purpose of STI-21 is to demonstrate the stability of the core local power-reactivity feedback mechanism with regard to small perturbations in reactivity caused by rod movement. 3.19.2 Criteria Level 1 The decay ratio must be less than 1.0 for each process variable that exhibits oscillatory response to control rod movement. Level 2 The decay ratio must be less than or equal to 0.25 for each process variable that exhibits oscillatory response to control rod movement when the plant is operatin; above the lover limit of the master flow controller. 3.19.3 Analysis STI-21 testing was conducted at test conditions 1, 2E, 3E, 4A, and 4E as defined on the power flow map in section 2.3. At each test condition the stability of the core power-reactivity feedback mechanism was tested by checking tle local and macroscopic effects of control rod movement. The selected rod was moved near a location of limiting core thermal conditions. A nearby LPMY was used to monitor local power changes. Overall plant and core conditions were monitored by STAR TREC. Only local power as monitored by the LPIP and local heat flux responded to the control rod movement. The LPM reading and local heat flux moved promptly to a new reading following the control rod movement and exhibited negligible oscillatory characteristics. Table STI 21-1 summarizes the results. All test criteria were met. .
...... ..... .. .. s .
FINAL SLMhINURY REPORT - BFNP UNIT 3 3.0 Results (Continued) 3.19 STI-21, Flux Response to Rods (Continued) 3.19.3 Analysis (Continued) Table STI 21-1 Resvonse To Control Rod Movement Sun-arv Peak Peak Highest Test Rod Rod LPRM LPR'I Heat Deca: Condition Moved Movement Monitored Change Flux Change Ratic" I 1 50-35 48
- 40 48 - 33A 6.4% 6.4% <.25 40 . 48 6.4% 6.4% c.25 2E 42-43 48 - 44 40 - 41A 9% 4% <.25 44
- 48 7% 4% <.25 3E 50-19 48. 44 48 - 17A 5% 4% <.25 44
- 48 51 4% <.25 4A 26-15 48 - 40 24 - 17A 9.6% 7.2% <.25 40
- 48 10.0% 7.2% .25 4E 50-15 48
- 40 48 - 17A 19.2% 13.9% <.25 40
- 48 16.8% 13.9% c.25 S
I
FINAL SUtTfrXRY REPORT - BF.P UNIT 3 3.0 Results 3.20 STI-22, Pressure Regulator 3.20.1 Purpose The purposes of STI-22 are:
- 1. To determine the optimum settings for the pressure control loop by analysis of the transients induced in the reactor pressure control system by means of the pressure regulators.
- 2. To demonstrate the take-over capability of the ban':-
up pressure regulator upon failure of 'the coitrollir.-7 pressure regulator and to set spacing between the set points at an appropriate value.
- 3. To demonstrate smooth pressure control transition between control valves and bypass valveE when reactor steam generation exceeds steam used by the turbine.
3.20.2 Criteria Level I The decay ratio must be less than 1.0 for each process variable that exhibits oscillatory response to pres-sure regulator changes. Level 2 In all tests except the simulate failure of the operating pressure regulator, the decay ratio is expecte! la be 0.25 for each process variable that exhibits oscillatory response to pressure regulator changes when the plant is operating above the lower limit setting of the Master Flowa Controller. Pressure control system deadband, delay, etc., shall be small enough that steady-state limit cycles, if any, shall produce turbine steam flow variations no larger than + 0.5% of rated steam flow. Optimum gain values for the pressure control loop shall be determined in order to give the fastest return from the transient condition to the steady-state condition within the limits of the above criteria.
FINAL SM.HARY REPORT - FNP UNIT 3 3.0 Results (Continued) 3.20 STI-22. Pressure Regulator (Continued) 3.20.2 Criteria (Continued) Level 2 (Continued) During the simulated failure of the controlling pressure regulator, if the setpoint of the backup pressure regulator is optimuly set, the backup regulator shall control the transient such that the peak neutron flux and/or peak vessel pressure remain below the scram settings by 7.5% and 10 psi respectively. Maintain a plot of the peak variable values versus power. Following a + 10 psi (0.7 kg/cm2 ) pressure settoint change, the time between the setpoint change and the occurrence of the pressure peak shall be 10 seconds or less. 3.20.3 Analysis STI-22 testing was conducted at test conditicns 1, 2E, 3E, AA, 4C, 4D, and 4E as defined on the power ico -n.- in section 2.3 The Electrohydraulic Control (EEC) svste-controller setting was adjusted to provsde 'or stabili v-the pressure control loop. The backup capability of each pres-sure regulator was demonstrated via simulated failure of Zhe controlling regulator. Final adjustments of the EHC system was completed at test condition 3E with implementation of the following settings: The EEC system pressure regulator settings were:
"A" Lag Pot CRS) 2.4 turns ('1 - 5 seconds)
"A" Lead Pot (R6) 4.6 turns ( I' - 2 seconds)
"B" Lag Pot CR3) 2.4 turns (Y - 5 seconds)
"B" Lead Pot (R4) 4.0 turns (7' - 2 seconds)
The ERC system steam line resonance compensator settings were:
"A" Notch Center 3.63 turns "A" Notch Depth 2.00 turns "A' Notch Width 1.67 turns "A"l Small Lag 1.47 turns
'B"Notch Center 3.63 turns "B" Notch Depth 2.00 turns "B" Notch Width 1.67 turns "B" Small Lag 1.47 turns
FINAL SUOARY REPORT - BFNP UNIT 3 3.0 Results (Continued) 3.20 STI-22, Pressure Regulator (Continued) 3.20.3 Analysis (Continued) ERC bias adjustments: Regulator Potentiometer 3.33 turns (30 psi) Pressure setpoint bias (3 psi separation) 4.24 turns Speed regulator 7.39 turns Intercept valve bias 10.00 turns Bypass valve opening bias 1.20 turns Recirculation flow signal limit 7.72 turns 3 psi separation between the regulators was established for normal operation Table STI 22-1 summarizes the results of the pr-ssure regulator setpoint changes. A smooth pressure transitio-. between control valves and bypass valves was demonstrated during the setpoint changes. the simulated failure test of the pressure regulator was conducted with a 2 to 4 psi bias between regulators :- orfer to minimize the neutron flux margin to scram to < 7.5%. A 5 differential had been generally recominended in the past '; . the plugged bottom core plate caused greater sensed neutron :?iux peaking. In order to minimize the neutron flux peaking urinal the backup regulator event, a setpoint differential of 1 to 4 psi has been recommended by General Electric and accepted by TVA, Division of Engineering Design. The current operat+/-.Cg setpoint differential is 3. psi. With this setpoint pressure regulator testing satisfied all level 1 and 2 criteria.
I . Table STI 22-1 I.. Pressure Repul.ntnr Peaponse Summary (Recirculation In Hatter Manual hIe) II Test Condition 1 2E ly lZ i ' Step Input -102 -10 oo02 -102 -10: 4102 -102 l+10 -10 4102 -lox +1O: -102 l+10% -101 +.eo i .: j ,.. Reguator (AIR) A A a R A A I A A a _a A A _ a
- -Y - - Ulv--1wV - InTrTv--- --- I _ $
j'." Valves (CYIDPY) C.V. Tnepnto e.v. 502 C.V. 50. C.v. Inepnt. C.V. 50% C.v, 5n0 C.v. 50s lncrnt. 50: Initial Dome rresas. 950 957 945 957 951 941 947 938 954 944 952 940 990 930 99b 984 I Final r.e Press. 941 946 955 947 940 950 938 946 943 957 943 951 980 996 983 998 Prcss. Peak (1) 2.0 2.7 2.0 2.8 5.0 3.5 3.0 2.8 4.0 6.0 2,i 7.0 4.0 7.0 5.1r 8.0 BLihOst Rat latay 4.25 4.25 4.25 c.25 c.25 c.25 c.25 '.25 c.25 (.25 C.23 2.z5 q.25 *.25 c.2S .25 Parameter (2) A APM AiRH ARM APM APRME r j -r r A Artn rjn rw A.4_ _ (1) Level 2 criteria limit to 10 seconds. (2)Level 2 criterla Is 0.25.
; r
'i ' i I II
FINAL
SUMMARY
REPORT - BFNP UNIT 3 3.0 Results 3.21 STI-23; Feedwater System 3.21.1 Purpose The purposes of STI-23 are:
- 1. To adjust the feedizater control system for acceptab-le reactor water level control.
- 2. To demonstrate stable reactor response to subcooling changes.
- 3. To demonstrate the capability of the automatic core flow runback feature to prevent low water level scram following the trip of one feedwater pump.
3.21.2 Criteria Level 1 The decay ratio must be less than 1.0 for each process variable that exhibits oscillatory response to fee:- water system changes. Level 2 The decay ratio is expected to be less than or equal to 0.25 for each process variable that exhibits oscIll:a=- response to feedwater system changes when the plant is operatinS above the lower limit of the master flow controller. Following a 3-inch (7.5 cm) level set-point step adjustment in three-element control, the tine from set-p!;_ step change until the water level peak occurs shall be less than 35 seconds without excessive feedwater swings (chargaes in feedwater flow greater than 25% of rated flow.) The automatic recirc-flow runback feature shall prevent a scram from low water level following a trip of or.e of the operating feedvater pumps. The water level margin to scram should be greater than 3 inches for a pump trip from the 100X power condition. With the condensate system operating normally, the maximum turbine speed limit shall prevent pump damage due to cavitation S i
FINAL SMAtRY REPORT - BFNP UNIT 3 3.0 Results (Continued) 3.21 STI-23. Feedwater System (Continued) 3.21.3 Analysis STI-23 testing was conducted at test conditions 1, 2E, 3E, 4C, 4D, and 4E as defined on the power flow map in section 2.3. Step changes of + 3 inches were made at each test condition listed above with the feedwater system in both the single and three element mode of control. Response of the feedwater system during the transients is summarized in table STI 23-1. At test condition 1 the time from initiation of the setpoint change to reaching the level peak was greater than the criterion of 35 seconds. No attempt was made to optimize system response at that power level, because only one feed pump was in operation. During all subsequent testing with three feed pumps in operation the level peak was reached wi-in the required 35 seconds, thus satisfying the criterion. hiring level setpoint change testirE at all lev'ls, the decay ratio was less than 0.25 for all process variables exhibiting response to the changes. Therefore, all criteria applicable to level setpoint change testing were met. During testing at test condition 2E, all three feed pumps were in operation. Final system optimization was, there-fore, performed at this level. The final settings on the level controller were: Proportional Band - 200% Reset - 1 repeati-inute. The mismatch gain was set for a 36-inch corrected level for 1002 mismatch of rated feedwater flow and steam flow. The lead-lag unit was set for a lag time constant of 5 seconds, and a lead time constant of 1 second. From test condition 4E, with all three feedwater pumps operating and the feedwater controller in the 3-element mode, one feedwater pump was tripped to test the automatic recirculation pump run back feature. The time from pump trip until the minimum reactor water level was reached was 27 seconds. Theininimum reactor water level reached was 22.5 inches, which is well above the scram setpoint of 11 inches. The feedwater and recirculation-systems responded satisfactorily to the feedwater pump triq, and all criteria were satisfied.
.1o
FINAL S1.ThIARY REPORT - BFNP UNIT 3 3.0 Results (Continued) 3.21 STI-23, Feedwater System 3.21.3 Analysis (Continued) Table STI 23-1 LEVEL SETPOONT CHANGES Test Controlling Control Setpoint Tire To Max. Decay Condition Level Mode Change Peak Level Ratio _ (in.) (sec.) -- _-- 1 1 element + 3" 66 <.25 1 1 element - 31 76 <.25 1 3 element + 31 64 <.25 1 3 element - 3_, 71.25 <.25 2E 3 element + 3" 30. <.25 2E 3 element - 3" 30. <.25 3E A 3 element + 3" 30.5 <.25 3E A 3 element - 3". 31 <.25 3E B 3 element + 3" 31 <.25
. 3E B .3 element - 3" 28.5 c.25 3E A 1 element + 3" 25. <.25 3E A 1 element - 3" 28 <.25 3E . B 1 element + 3" 26.5 <.25 3E B 1 element - 3 25.5 <.25 4C B 3 element _ 3 34.5 <.25 4C B 3 element + 3" 34. <.25 4C A 3 element - 3" 34. <.25 4C A 3 element + 3" 33 <.25 4C A 1 element - 3" 44. c.25 4C A I element + 3" . 35 <.25 4C B 1 element - 3" 32 <.25 4C B 1 element + 3" 42 <.25 4D B 3 element -3" 32 <.25 4D B 3 element +3" 32 <.25 4D A 3 element _ 3 31 <.25 4D A 3.element + 3" 34.5 <.25 4D A . 1element - 3"1 21 <.25 4D A 1 element + 3" 30 <.25 4D B 1 element - 3 30 <.25 4D B. 1 element + 3"1 31 <.25
. . I
FINAL SLWMARY REPORT - BFnP UNIT 3 3.0 Results (Continued) 3.21 STI-23, Feedwater System (Continued) 3.21.3 Analysis (Continued) Table STI 23-1 LEVEL SETPOMN CHANGES (Continued) Test Controlling Control Setpoint Time To Max. Decav Condition Level Mode Change Peak Level Ratio (in.) f_. (see.) _ 4E A 3 element - 3"F 30 c.25 4E A 3 element + 3" 32 <.25 4E B 3 element - 3" 31 <.25 4E B 3 element + 3" 32 <.25 4E B I element - 3 18 c.25 4E B 1 element + 3" 21 <.25 4E A 1 element - 3 21 <.25 4E A 1 element +3T 31 < .25 t
FINAL SUL'UARY REPORT - BFNP UNIT 3 3.0 Results 3.22 STI-24, Bypass Valves 3.22.1 Purpose The purposes of STI-24 are:
- 1. To demonstrate the ability of the pressure regulator to minimize the reactor pressure disturbance during an abrupt change in reactor steam flow.
- 2. To demonstrate that a bypass valve can be tested for proper functioning at rated power without causing a high flux scram.
-3.22.2 Criteria Level 1 The decay ratio oust be less than 1.0 for each process variable that exhibits oscillatory response to bypass valve changes.
Level 2 The decay ratio is expected to be less than or equal to 0.25 for each process variable that exhibits oscillatory response to bypass valve changes when the plant is operating above the lower limit setting of the Master Flow Controller. To avoid approaching steam line low pressure isolation, the maximum pressure decrease at the turbine islet during valve opening shall not exceed 50 psi (3.5 kg/cm2 ). System pressure shall reach a steady-state value within 25 seconds after the bypass valve has been opened or closed. The regulator shall limit the pressure disturbance during valve reclosure so that a margin of at least 7.5X shall be maintained below flux scram. 3.22.3 Analysis Bypass valve testing was conducted at test conditions 1, 2A, 2E, 3E, 4A, 4C, 4D, and 4E as defined in the power flow map in section 2.3. The successfully completed bypass valve test program demonstrated that the EEC system had adequate capability to respond to abrupt changes in steam flow.
FINAL
SUMMARY
REPORT - ZFhP UNIT 3 3.0 Results (Continued) 3.22 STI-24. 7
- Bypass
. Valves (Continued) 3.22.3 Analysis (Continued)
For test purposes, the bypass valve opening time was adjusted so that the valve would open In as s'Cot a time as possible. Since it is not possible to have 3'oth fast opening and closing times, the valves were adjusced for a fast opening tine of approximately 3.0 seconds and a slower closing time of approximately 16 seconds. Table STI 24-1 contains a summary of the by ass valve test transient data from all test conditions. !+'n--ss valve testing at all test conditions listed in the table satisfied all test acceptance criteria. Throughout the startup test program, data were taken to extrapolate for the minimum flux margin to scram when operating at 100% rated power. The graph containing all points is shown in figure STI 24-1. Each test netter results which shoved this margin to be approximately 18.3¶ of rated power, which satisfies the level 2 criteria.
-. 4*
FINAL
SUMMARY
REPORT - BFNP UNIT 3 3.0 Results (Continued) 3.22 STI-24, Bypass Valves (Continued) 3.22.3 Analysis (Continued) Table STI 24-1 Bypass Valve Transient Data Suwmmary e!-I~t
,Condi~tlon S
.t CLlmlt 1 2A 2E 3E AA 4C 4D 4E Parameter _ _
1120 MWt 823 MWt 1811 MWt 2637 MIt 1322 MWt 1873 MWt 2387 MWt 3239 M-t Thermal Power P A 34.0% 25.02 55.0% 80% 40.2% 56.9% 72.5% 98.3% 51.0 Mlb/hr 26.7 Mib/hr 106.6 Mib/hr 104.2 Mib/hr 29.1 Nlb/hr 47.6 Mlb/hr 74.4 Mlb/hr 99.0 Hlb/br Total -Cre Flow NA 49.8% 26.0% 104.0% 101.6% 28.4% 46.4% 72.6% 96.6% Date NA 10/24/76 10/28/76 10/11/76 11/3/76 11/26/76 11/27/76 11/28/76 11/23/76 D Marimum Time to S.S. Pressure (sec) c25 16.0 11.0 16.0 19.0 11.2 16.0 0.0 18.0 Margin to Flux Scram (M) >7.5 15.8 10.8 31.1 15.0 13.0 20.26 13.08 18.29 Scram Setpoint (X) NA 51.8 35.3 86.1 95.0 54.0 80.26 88.28 115.99 Decay Ratio <.25 0.0 _25 0.0 0.0 O _ _ 0.0 0.0 .0 0.0 Initial Dome Pressure (psig) NA 988.0 946.0 979.0 970.0 964.3 960.0 975.0 998.0 Change in Dome Pressure (psig)g 2 2 2 2 1 2 . 0 1 Opening Time of Bypass Valve (sec) w3.0 u3.0 3.-0 -3.0 =3.0 3.52 3.70 3.76 3.28
FINAL SUNMARY REPORT - DF"P UNIT 3 3.0 _Results (Continued) 3.22 STI-24, Bypass Valves (Continued) 3.22.3 Analysis (Continued) 11, X rig
- IOU -T~i o ,' till.
Iii S~;-1 I.! I I i. III3ZI2) 1,1,, 8 E 0 0 707 Percent RAtIdPoe Peakjur :;Is MV24-1 a Bypt~~~s~NV v. :.n!t
- !:....'*2 IDestrI zI r- l2n: Vl tmx Naly 1 ai Ist Fx ramtIt I 0M.OP'W.-wr
FINAL SUM!HARY REPORT - BFNP UNIT 3 3.0 Results 3.23 STI-25. Main Steam Line Isolation Valves 3.23.1 Purpose The purposes of STI-25 are:
- 1. To functionally check the Hain Steam Line IsolatiLn Valves (HSIMs) for proper operation at selected power levels.
- 2. To deternine reactor transient behavior dpring and following simultaneous full closure of all YsSIVs, and following full closure of one valve.
- 3. To determine Isolation valve closure time.
- 4. To determine the maximum power at which a single valve may be closed without a reactor scram.
3.23.2 Criteria Level 1 - HSIV closure time must be greater than 3 and less than 5 seconds. The initial transient rise in vessel dome pressure occurring within 20 seconds of the main steam isolation valve trip initiation shall not be greater than 150 psi, and the transient rise in simulated heat flux shall not exceed 107. Level 2 The initial transient peak in vessel dome pressure occurring within 20 seconds following initiation of the 'ISIV closure and the transient peak in simulated surface heat flux shall not be more limiting than the predicted transients in
-the Transient Analysis Design Report (lOO psi and no heat flux.
Increase.) During full closure of individual valves, pressure nust be 20 psi (1.4 kg/cm2 ) below scram, neutron flux must be 10? below scram, and steam flow in individual lines must be 10% below the isolation trip setting. 3.23.3 Analysis STI-25 testing was conducted at heatup, test conditions 2E, 4E, and 4E levels as defined on the power flow map In section 2.3.
*- I. .
FINBALSS RY REPORT - BFNP UNIT 3 3.0 Results (Continued) - 3.23 STI-25, Main Steam Line Isolation Valves (Continued) 3.23.3 Analysis (Continued) Main Steam Isolation Valves C(SIV) were indis.vieal1-i closed at heatup, test conditions 2E and 4D. Closi.f.g tif:5as are sumarized in table STI 25-1. Data taken at each p va was analyzed to ensure that individual closures could be performed at the next plateau of higher power. Closure times at all levels of testing were between the require 3-5 seconds. Slow closure to the 90% open position for each MSIV was satisfactorily performed at heatup and test corditeixs 2E and 4D. During all HSIV closures transient behavior o+/- significant reactor and plant parameters were monitored by STARTREC.. For ali parameters performance during the tran.,ent met level 1 and 2 criteria. Transient behavior iL su--arized in table STI 25-2. On December 3, 1976, a simultaneous full closure of all HSIV's was initiated from 96.5% of rated core thercal power. Reactor transient behavior and MSIV closure ti-.3 were recorded by STARTREC. Closure times were within t:-.e required 3--5 seconds. During the initial 20 seconds aftr :-.:e scram the peak dome pressure rise was 84 psi. No incrras .L simulated heat flux was measured. All level 1 and 2 criteria were satisfied.
-100-FINAL S?!'%ARY REPORT - EFNP UNIT 3 3.0 Results (Continued) 3.23 S$I-25, Main Steam tine Isolacion Valves (Continued) 3.23.3 Analysis (Continued)
Table STI 25-1 Y MSIV Closure Times Closure Time (sec.)* T. C. 4D l _SIV Number MSIV Wumber Reatup T.C. 2E T.C. 4O FCV-1-14 (1LA) 3.47 3.39 3.481 FCV-1-15 (C2A) 3.09 2.99* 3.069 FCV-1-26 (ILB) 3.30 3.70 3.296 FCV-1-27 (2!B) 3.50 3.50 3.60S FCV-1-37. (O.C) 3.50 3.60 3.605 FCV-1-38 (2!C) 4.20 4.60 4.223 FCV-1-51 (1.D) 3.40 3.30 3.193 FCV-1-52 (2D) 3.30 3.20 3.193
- Times are for 0 - 97% closure.
** Closure time for 0 - OOX was 3.08 sec.
Table STI 25-2
- Transient Behavior During DlSI Closure Parameter _Heatup T.C. 2E T.C. 4D Dome Pressure (psig)
Scram Setpoint 1055 1055 1055 Peak Value No Change 990 1005.5 Margin to Scram 65 49.5 APIM Heat flux (%) .. Scram Setpoint 1SX 70% 91.7% Peak Value No Change 48% 80.5S Margin to Scram 22% 11.2% Iadividual Steam Line F1ow AbXh) ScramcSetj..oint .. 4.69 4.69 Peak Value - No Change 2.0 3.20 Margin to Scram 2.69 1.49 S t
-101-FINAL SU.rARY REPORT - BFNP UNIT 3 3.0 Result:s 3.24 STI-26, Relief Valves 3.24.1 Purpose The purposes of this test are:
- 1. To verify the proper operation of the primary system relief valves.
- 2. To determine the capacity and response charac:er-istics of the relief valves.
- 3. To verify the proper seating of the relief valves following operation.
- 4. To verify that the discharge piping is not blocked.
3.24.2 Criteria Level 1 There should be positive indication of steam dis-charge during the manual actuation of each valve. The sum total of capacity measurements from the 11 relief valves shall be equal to or greater than 8.83 x 1j6 lb/hr + 2Z corrected for an Inlet pressure of 1112 psig. Level 2 Relief valve leakage shall be low enough that tha temperature measured by the thermocouples in the discharge side of the valves returns to within 100 F. (5.60 C) of the temperature recorded before the valve was opened. Each in-dividual relief valve shall have a minimum capacity of 720,000 lb/hr corrected to an inlet pressure of 1112 psig. The pressure regulator must satisfactorily control the reactor transient and close the control valves or bypass valves by an amount equivalent to the relief valve discharge. The transient recorder signatures for each valve must be analyzed for relative system response comparison. 3.24._' Analysis STI-26 testing was conducted at heatup, test con-ditions 1 and 3E. The bypass valve calibration phase of STI-26 was performed in test condition 1 testing. A least-squares fit was made to the data to relate the bypass valve capacity to the relief valve capacity. During TC 1 relief valve testing, the feedwater flow decreased by approximataly
-102-FINAL SUM=RX REPORT - BFNP UNIT 3 3.0 Results (Continued) 3.24 STI-26, Relief Valves (Continued) 3.24.3 Analysis (Continued) 3.9 Nlb/hr, reactor pressure dropped by 6 psig, stean flog decreased by approximately .75 Xlb/hr, and APIX A decreased by 32 when the valve was opened.
Table STI 26-1 represents a summary of all the pertinent data obtained during relief valve testing. At l relief valves met steam discharge, capacity, and reseatinS criteria at all levels of testing. The pressure regulator satisfactorily controlled the pressure transient when the relief valves were opened.
-103-FINAL SLUMARY REPORT - BFNP UNIT 3 3.0 Res.1ts (Continued) 3.24 STI-26, Relief 'Valves (Continued) 3.24.3 Analysis (Continued) 6 Electric TVA Table STI 26-1 Summary of Relief Valve Data Corrected Capacity Klblhr Time For Temp.
I Relief valve ;:er-nocouple Tcz:.at l T Return to TC 3E Relief Relief Tes clest within 10OF Initial Fin-1 Valve i,7o. Valve No. Condition 1 Condition 3E (sec.) OF 1-4 A .8212 .8385 1.25 208 213 1-5 3 .8301 .8734 1.50 220 230 1-18 C .8301 .8734 1.00 221 2.9 1-19 D .8186 .8297 1.00 195 203 1-22 E .8036 .8122 1.00 174 184 1-23 F .7965 .8473 1.62 181 190 1-30* G .8770 .8821 1.00 222 220 1-31* . .8780. .8909 2.20 261 271 1-34 J .8231 .8647 .75 208 217 1-41 K .8372 .7598 1.53 225 235 1-42 L .8328 .7949 1.00 269 276 Total Total Capacity Ilblbr - I. 9.1483 I 9.27
*Crosby Relief Valves Capacity Limit Individual Capacity: .720 Mib/hr Total Capacity: 8.83 Mlb/hr
-104-FINAL SLIMARY REPORT - BFNP UNIT 3 3.0 Results 3.25 STI-27, Turbine Trip and Generator Load Rejection 3.25.1 Purpose The purpose of STI-27 is to demonstrate the response of the reactor and its control systers to protective trips iA.
the turbine and generator. 3.25.2 Criteria Level 1 The Initial transient rise in vessel dome pressure occurring within 10 seconds of the turbine/generator try initiation shall not be greater than 150 psi and the transient rise in simulated heat flux shall not exceed 10 percent. The turbine stop valves must begin to close before
.the control valves for the turbine trip. The turbine control valves must begin to close before the stop valves during the generator load rejection.
Following fast closure of the turbine stop a2,r control valves, a reactor scram shall occur it the turbine first stage pressure is greater than 154 psig. Feedwater systems must prevent flooding of the stear-line following the transients. Level 2 The initial transient rise in vessel dome pressure occurring within 10 seconds of the turbine/generator trip initiation and the transient rise in simulated surface heat fly: shall not be more limiting than the predicted transient presen.:e.- in the Transient Analysis Design Report (100 psi and no heat flux increase.) The pressure regulator must prevent a low pressure reactor isolation. The wide range level sensing system and the feed-water controller must prevent a low level initiation of the EPCI and NSIV's as long as feedwater flow remains available. The trip scram function for higher power levels must meet RPS specifications.
-105-FixAL SUmAnRY REPORT - BFNP UN'IT 3 3.0 Results (Continued) 3.25 STI-27, Turbine Trip and Generator Load Rejection (Continued) 3.25.2 Criteria (Continued)
Level 2 (Continued) The load rejection within bypass capacity must nct cause a scram. For the.case of turbine trip at 75-percent pos-.er, the measured transient parameters will be conpared with the predicted values. If any paraneter is significantly different from the predicted values the test will be repeated at 100-percent power. 3.2.5.3 Analysis STI-27 was performed at test conditions 1, 3E, and 4E as defined on the power flow map in section 2.3. A generator load rejection within bypass valve capacity was performed by opening the main transformer breakers at 24.5% power. The control valves closed in axprcx-imately 0.5 seconds after the main generator breaker Was ope.nui. The bypass valves opened to 85% of total capacity, APES A in-creased by approximately 1%, the control valves decreased ironi 14 to 0% open, and feedwater flow decreased by 0.1 Nib/hr. The wide range level sensing system and the feedwater controller prevented a low level initiation of HPCI and MSIV's. The turbine trip test was performed at 75.3%. po-w;er. The reactor is ediately scrammed, initiated by the 10%?stop valve closure condition. The peak reactor dome pressure was 1044 psig after 4.0 seconds, well below the 1080 psig relief valve setpoint. A low-low water level reactor isolation occur-red. As resolution to this problem, the following feedwater controller system changes will be made:
- 1. The low level isolation setpoint will be lowered.
- 2. Installation will be made of an automatic level set-polut setdawn and a high level feedwater pump trip.
All reactor protection systems functioned as expected. The pressure rise was less than the predicted and the projected 100X power case. The following table summarizes the significant events during the test. S - . , A~. . ..
-106-FINAL SUMA-RY REPORT - BFNP V.1IT 3 3.0 Results (Continued) 3.25 STI-27, Turbine Trip and Generator Loed Rejection (Continued) 3.25.3 Analysis (Continued)
Table STI 27-1 Time (sec.) Event 0.0 APPM A - 76.51; Dome pressure - 965 psig; Feedwater. flow - 9.8 1lb/hr; water level - 36 inches; Vain turbine trip. 0.2 Stop valves closed. 0.3 Control valves closed; reactor scram. 1.7 APRH A - 17%. 4.0 Feedwater flow - 8.4 n11b/hr; water level - 0 inch. i 4.6 Reactor isolation on low water level; dome press - 1040 psig. 9.0 Feedwater flow - 19.4 Mlb/hr. 12.0 Simulated thermal power - 02; feedwater flow - 8.0 lab/hr. f, . The generator load rejection test was performed a:: 98.7% power by opening the main transformer breakers. Due to the failure of the time delay relay in the power/load unbalance circuit, a control valve fast closure did not occur. This resulted in a turbine stop valve trip due to turbine over-speed. The resulting transient on the turbine was more severe than a control valve .fast closure transient because the turbine overapeed reached '%v 113% compared to approximately 1053 for a control valve trip. The transient on the reactor is co=Par-8ble to that resulting from a control valve fast closure. '. increase in LPRK's, APM1Rs, or simulated heat flux were noted after the trip. As noted in the turbine trip test, a low water level isolation occurred. The first pressure pea'k ccur-red at 4.43 seconds with a maximum reactor dome pressure or 1085 psig, and the second at 25.63 seconds at 1101 psig, due to S
-107-FINAL SUWARY REPORT - BBP. UNIT 3 3.0 Results (Continued) 3.25 STI-27, Turbine Trip and Generator Load Rejection (Continued) 3.25.3 AnalYsis (Continued) the low reactor water level isolation. Relief va'Aves D and F opened in both cases to reduce the reactor pressure to less than 1075 psig. The feedwater controller system changes discussed previously should enhance the post-scram recoverability and prevent low water level isolations.
The time delay relay that prevented a control valve fast closure was repaired and a special test was per-formed to demonstrate its operability. The following table summarizes the significant events of the test. Tabel STI 27-2 Time (sec.) Event 0.0 .PRI1 A -- 98'3%; Dome pressure - 1000 p-ig; water level - 33 inches; Main transformer breakers opened. 0.020 Initiates control valve fast closure. 0.120 C.V. begin to close as turbine overspeeds. 1.6 Water level - 38.1 inch. 1.63 Turbine stop valve trip; reactor scram. 2.00 Water level - -63 inches; APRM A - 65%. 4.0 APRK k - 0X. 4.43 Dome pressure - 1085 psig; D and F relief valves open. 6.4 Water level - 32 Inches; Low water level isolation. 6.63 Dotme pressure - 1077 psig; Water level - 20 inches. 25.63 Dame pressure - 1101 psig; D and F relief valves open; Water level 31.1 inches. 29.63 Dome pressure - 1070 psig.
-108-FINAL SUMIARY REPORT - EFNP UNIT 3
- 3. Results 3.26 STI-30, Recirculation System 3.26.1 Purpose The purposes of STI-30 are:
- 1. To verify that the feedvater control system can satisfactorily control water level with-out a resulting turbine trip/scram, and to obtain actual pump speed/flow coastdown data.
- 2. To verify recirculation pump startup under pressurized reactor conditions..
- 3. To obtain recirculation system performance data.
- 4. To verify that no recirculation system cavi-tation will occur in the operable region of the power-flow map.
- 5. To provide the opportunity to obtain flow induced vibration data.
- 6. To evaluate the recirculation flow and power level transient following trips of one or both of the recirculation pumps.
3.26.2 Criteria Level 1 Not applicable Level 2 The power and flow coastdowns are expected to agree with pre-calculated power and flow coastdown rates. The plant shall not scram as a result of a high level turbine trip. 3.26.3 Analysis STI-30 testing was performed at test conditions 2A, 2E, 3E, 4A, and 4E as defined on the power flow map in section 2.3. Recirculation system performance data was taken on the 50X flow control line at various combinations of pump speeds as specified by section 6.3 of STI-30, and
-109-FINAL SLMIARY REPORT - BTh'P UNIT 3 3.0 Res-ults (Continued) 3.26 STI-30, Recirculation System (Continued) 3.26.3 Analysis (Continued) at each end of the 75% and 100% flow control lines.
Performance of the system was satisfactory at all conditions. A test for cavitation in the recirculation system was performed from % 502 power by inserting control rods in the reverse order of rod sequence "A" until the feedwater flow limit that initiates a recir-culation pump runback was reached. The recirculation pump runback circuitry was disconnected during the test to prevent an actual runback from occurring. Power was reduced to 22.3% (736 HWOt) of ratgd, which corre-sponds to feedwater flow of 2.61 X 10 lb/hr. The recirculgtion pump runback setpoint is set at 2.7 X 10 lb/hr. No signs of cavitation were seen in the jet pumps or recirculation pumps at any power level during the test. A single pump trip was performed at " 50% core thermal power and 100% flow by opening the genera-- tor field breaker on pump "A". Single pump trips and simultaneous 2 pump trips were performed at 502 and 100 core thermal power and 100% flow by tripping the drive motors. Transient traces were taken by STARTREC of significant plant and recirculation system parameters. Figures STI 30-1 through STI 30-7 compare plant para-meters as recorded by STARMIREC with predicted behavior for the first 10 seconds of analyzed trips. Except for "A" recirculation pump drive flow signal, all parameters agreed closely or were conserva-tively compared to predicted behavior for analyzed transients. "A' pump drive flow did not decay off as expected. Analysis of loop jet pump flow and total core flow indicated that "A" pump was actually performing as predicted, and that "A" and "B" pumps reacted in substantially the same manner during the transients. It was therefore felt that the difference in drive flow signals was in the flow measurement circuitry. Circuit repairs have been completed. All level 2 criteria have therefore been met.
-110-FMNAL SULIMARY REPORT - BFNP UNIT 3 3.0 Results (Continued) 3.26 STI-30, Recirculation System (Continued) 3.26.3 Analysis (Continued)
Following pump trips at 50% po,'er testing, each recirculation pump was tested for its ability to restart under pressurized conditions. Significant system parameters were recorded by STARTREC during the restart. No difficulties were encountered and each pump performed as expected.
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-118-FINAL SUMfNARY REPORT - BF-N IUMIT 3 3.0 Results 3.27 STI-31, Loss of T-G and Offsite Power 3.27.1 Purpose The purpose of STI-31 was to investigate the --
reactor transient performance during the loss of the rain generator and all offsite power and to demonstrate tLe acceptable performance of the station electrical suppiv system during the loss of the main generator and all off-site pawer. 3.27.2 Criteria Level 1 The initial transient rise in vessel dowe pressure occurring within 10 seconds of turbine/generator trip act'-r. when initiated simultaneously with loss of offsite power -:.n performed at 25-percent power shall not exceed 150 psi an_ the simulated heat flux rise shall not exceed 10 percent. All safety systems, such as the RPS, diesel-generators, and the RCIC and HPCI, must function properly w-4thout m&an.l ass4stance. Level 2 The initial transient rise in vessel dome pressure occurring within 10 seconds of turbine/generator trip shal1 not be greater than 75 psi, and there shall be no significant increase in simulated heat flux. Normal reactor cooling water systems should be able to maintain adequate suppression pool water temperarure, adequate drywell cooling, and prevent actuation of the auto-depressurization system. 3.27.3 Analysis STI-31 testing was conducted at test condition 1 as defined in the power flow map in section 2.3. Prior to the test, the plant electrical system was aligned so that the only source of power to the unit 3 auxiliaries was the unit 3 stats.. service transformer. The loss of offsite power test was per-formed by tripping the unit 3 generator negative phase sequence relay 346X and opening breaker 1405 on September 27, 157$. Water level dropped to -9.0 inches below the bottom of the dryer separators. Without intervention, auto initiation of
- f. less
-119-3.0 Results (Continued) 3.27 STI-31, Loss of T-G and Offsite Power (Continued) 3.27.3 Analysis (Continued)
BPCI and RCIC would have occurred at -31.5 inches. Approximately 5 minutes after the trip, RCIC was manuall' initiated to demonstrate operability. All diesel-generacors came on-line after approximately 6.44 seconds. At apprex-imately 18 seconds the reactor was manually scramed. The scram function of the RPS was verified to operate properly by indication of AUTO scram at approximately 24 seconds due to low water levels During the test, RPS XG set A continued running and MG set B's load breaker did not trip. !Zormally, the HG set motor input contactor will be opened in approximately 3 seconds; then the flywheel will carry the RPS bus loads until the frequency drops to 54.2 hertz at which tine the breaker will trip. Investigation of MG set A and MG set 3 found that the time delay relays were improperly set to trip at 6.5 and 5.2 seconds and the output load breakers Cere incorrectly set. Both MG sets tine delay relays were adjusted to drop out in approximately 3.0 seconds and the load breakers were correctly reset so that they would e to an underfrequency trip signal. The initial transient rise in vessel pressure occurring within 10 seconds of the turbine/generator trio was measured to be 3 psi. No rise in simulated heat flu:. was observed. Normal cooling water systems maintained satisfa.tcry suppression pool and drywell temperatures and prevented actuation of the auto-depressurization system. After t-e above corrections were made to the RPS-M!G sets, all level 1 and 2 criteria were considered satisfied. FILMED FROM B3ST AVAILABLE COP6 9
-120-FINAL SUthiMMY REPORT - BhP? MNIT 3 3.0 Results 3.28 STI-32, Recirculation Speed Control and Load Following 3.28.1 Purpose The purposes of STI-32 are:
- 1. To doternine correct gain for optimum performana:e of individual recirculation loops.
- 2. To determine that the recirculation loops are correctly set up for desired speed range and for acceptable variations in loop gain.
- 3. To demonstrate plant response to changes in recirculation flow.
3.28.2 Criteria Level 1 The decay ratio must be less than 1.0 for each vrcvcc:. variable that exhibits oscillatory response to flow controa changes. Level 2 The decay ratio should be less than 0.25 for any process variable that exhibits oscillatory response to 10." speed change inputs in local or master manual modes. Steady state limit cycles, if any exist, must rot cause turbine steam flow to vary in excess of + 0.5% rated steam flow as measured by the gross generator electrical power output. Following a 10% speed demand step from the low end of the master manual flow control range, the time fron the step demand until the speed peak occurs shall be less thant 25 seconds. 3.28.3 Analysis STI-32 testing was conducted at test conditions 1, 2D, 2E, 3C, 3D, 3E, 4C, 4D, and 4E, as defined on the power flow map in section 2.3. Prior to power operation, the recirculation system controllers were set up for stable operation. The initial settings were: proportional band = 500Z; resets/min. = 23. At test condition 1 the settings were changed to give a slightly faster response with negligible overshoot. The new settings were: proportional band = 225%; resets/min. = 12.
-121-FINAL SUk{ARY REPORT - BFNP UNIT 3 3.0 Results 3.28 STI-32, Recirculation Speed Control and Load Following (Continued) 3.28.3 Analysis (Continued)
Further optimization of system controls resulted in final settings as summarized below: Controller A: P.B. = 500%. 22 resets/min. Controller B: P.B. - 200%, 9 resets/min. Master Controller : P.R. - 80%, .9 resets/min. To determine system response, + 10 speed changes were performed on each pump individually, arnd with the pumps in the master-manual mode of control. Speed change testing was conducted at each test condition as required by section 6.1 of the test instruction. For all speed changes the decay ratio of all effected parameters was less than a.23. No steam flow variations caused by steady state limit cycles were observed. For speed changes performed at the lover end of the master manual flow control range, the maxinum tine from the step demand to the speed peak was 24 seconds. A2.1 level 1 and level 2 testing criteria have been met. Gain curves were obtained for each pump at test condition 2E. The curves were very nearly linear for bot'b pumps; therefore, no cam cutting or linkage adjustment e.-as necessary. The gain curve is shown in figure STI 32-1. The mechanical stops of the recircupation pumps were set at a point corresponding to 105% core flow at te3t condition 4E. The electrical stops were set just below this. The load following range limiter was set for 44% pump speed on the low end and 105% core flow on the high end. FILMED FROM BEST AVAILABLE COPx
-122-FINAL SMH._ APS' REPORT - BF.W rt.;'t 3 I
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-123-FINAL SThINARY REPORT - EFNh UNIT 3 3.0 Results 3.29 STI-33, Main Turbine Stop Valve Surveillance Test 3.29.1 Purpose The purpose of this test is to demonstrate acceptable procedures for daily stop valve surveillance testing at a power level as high as possible without producing a reactor scram.
3.29.2 Criteria Level 1 Not applicable Level 2
*Peak neutron flux must be at least 7.5% below the scram trip setting. Peak vessel pressure must remain ar least 10 psi below the high pressure scram setting.
Peak steam flow in the main steam lines must run-in 10% below the high flow isolation trip setting. 3.29.3 Analysis STI-33 testing was successfully conducted at test conditions 1, 2E, 3E, and 4E as per the power flow nap in section 2.3. Turbine stop valves were closed individual-Y at selected power levels. Due to the turbine bypass header, most of the pressure peaking effect was dampened, produciz? negligible perturbations in the reactor. STI-33 demonstraced that the stop valve surveillance test may be satisfactcrily performed at full power. The following table summarizes all the pertinent results from the stop valve surveillance test. All test criteria were met.
-124-3.0 Results (Continued) 3.29 STI-33, Main Turbilie Sto, *Valve Surveillance Test (Continuee) 3.29.3 Analysis (Continued)
Table STI 33-1 i . dltion 2E 3E 4E Limit Parameter ~'_ Date 9/19/76 1018/76 11/3/76 11/23/76 ': Reactor Power 75%nVt=22.8% 1799M,7t=54.6% 270Q.Wt=82.1Z 3214011t=97% Reactor Pressure 956 psig 950.6 psfg 987 psig 997 psig Peak Neutron Flux 25.4% 57.2% 84.5% 98% Margin to Scram 10.5% 12.5Z 10.5% 22% 1 > 7.5 Peak Vessel Press Margin to Limit 98.5 psi 95 psi 90.4 psi 56.2 psi > 10 psi; Peak Steam Line Flow Margin to Limit 110.5% 92.15% 55% 34%Z 10. C
-125-FINAL SMiNARY REPORT - BFNP UNIT 3 3.0 Results 3.30 STI-34, Vibration Measurements 3.30.1 Purpose The purpose of STI-34 is to obtain vibration measurements on various reactor components to demonstrate the mechanical Integrity of the system to flow induced vibration and to check the validity and accuracy of tic analytical vibration model.
3.30.2 Criteria Level 1 The vibration criteria, used to judge the results of the vibration measurements, is the precalculated vibra-tion amplitude at each sensor when the maxicum stress in any one of the internal's structures or components equals 10,000 psi including stress concentration factors. This stress represents approximately one half the stress li-tit given in ASME Code Section III for 40-year life. Because of their complexity, the criteria are not presented here but will be administered on site bv the vibration test engineer conducting the test. (See section 8 of the startup test instruction for more detail.) Level 2 Not applicable 3.30.3 Analysis STI-34 testing was conducted at heatup and test conditions 1, 2D, 2E, 2A, 3C, 3D, 3E, 4C, 4D, 4E, and 4A as per the power flow map found in section 2.3. Vibration data was taken in conjunction with the recirculation ptm~p trips and with the pumps at different speeds. Review of the data by the General Electric vibration specialist Indicates that the vibration amplitudes are well within criteria limits.
-126-FINAL SUMNARY REPORT - BFNP MNIT 3 3.0 Results 3.31 STI-35 Recirculation System Plow Calibration 3.31.1 PuERose The purpose of STI-35 is to perform a cormplete calibration of the installed recirculation system flow instrumentation.
3.31.2 Criteria Level 1 Not applicable. Level 2 Jet punp flow instrumentation shall be adjusted such that the jet pump total flow recorder will provide a correct core flow indication at rated conditions. The APRN/RBMI flow-bias instrumentation shall be adjusted to function properly at rated conditions. 3.31.3 Analysis STI-35 testing was conducted at the open vessel test plateau and test conditions 2E, 3E, and 4E as deiir.ed by the power-flow map in section 2.3. Prior to power testing, the recirculation flow nozzle transmitters were calibrated for a 0 to 29.4 psi span and an off-set of .' 2ro on the single tap AP transmitters. During test condiieros 2E and 3E, the indicated core flow was verified to be withini 2% of the calculated values. At these two test conditicas, the jet pump flow instrumentation provided an accurate indication of core flows such that adjustments were not necessary. Experience has shown that the accuracy of the core flow calibration increases with power level. Three sets of core flow data were taken at rated conditions. Based on this data, the gains of the jet pump loop and total core flow proportional amplifiers were adjusted to give the correct control room indications of total core flow and jet pump loops A and B flows. Comparison of the total core flow recorder and the process computer core flciT data point showed agreement within 0.08%. Subsequently, three additional data Sets were taken to confirm the recirculation flow nozzle transmitter spans. Based upon analysis of this data, the flow nozzle transmitters were subsequently spsmnad to 24.5 psid for Loop A, and 29.8 psid for Loop B. The X-ratios S .....
-127-FINAL SUMfARY RE?ORT - EF:'P UNIT 3 3.0 Results 3.31 STI-35, Recirculation System Flow Calibration (Continued) 3.31.3 Analysis (Continued) calculated via the computer program "JRPLV.P", were with-in the band of expected theoretical values. The gain adjustr-nt factors and as-left gains are as follows:
Instrument Gain As-Left Adjustment Factor Gains A .99 .495 1.01 .505 The APRM/RBM flow bias instrumentation was adjusted and found to perform satisfactorily. In addition, -1I jet pump riser plugging, nozzle plugging, and loop flow variation criteria were satisfied. t1 : .
-128-FINAL SU4-ARY 'REPORT - BPNP UNIT 3 3.0 Results 3.32 STI-70, Reactor Water Cleanup System 3.32.1 Purpose The purpose of STI-70 is to demonstrate specific aspects of the mechanical operability of the Reactor Water Cleanup System. (This test, per-formed at rated reactor pressure and temperature, is actually the completion of the preoperational testing that could not be done without nuclear heating.)
3.32.2 Criteria Level I Not applicable Level 2 The temperature at the tube side outlet of the non-regenerative heat exchangers shall not exceed 130 0 F Iz-any mode. The pump available NPSH will be 13 feet or greater during the hot standby mode defined in the process diuSr.=s. The cooling water supplied to the non-regenerarive heat exchangers shall be within the flow and outlet tempera-ture limits indicated in the process diagrams. (This is applicable to "normal" and "blowdown" modes.) 3.32.3 Analysis STI-70 testing was conducted during heatup as defined on the power flow map in section 2.3. The reactor water cleanup system was successfully tested at rated reactor pressure and temperature in the blowdoun, hot standby, and normal mode. It was demgnstrated that the service water could remove 24.70 X 10 Btu/hr from the non-regenerative heat exchangers when the cleanup system was in the blowdown mode. The regenerative 6 exchangers were found to have a capacity of 37.95 X i0 Btu/hr when the cleanup system was in the hot standby mode. The IPSE is strongly dependent on the temperature of the water on leaving the pressure vessel and entering the cleanup system. Because the actual value of the pump inlet temperature -as below the process diagram, the process A .' -. -:5n . .
-129-FINAL SUMIARY REPORT - BFNP UN'IT 3 3.0 Results (Continued) 3.32 STI-70, Reactor Water Cleanup System (Continued) 3.32.3 Analysis (Continued) diagram value of 5450 F was used for conservatism.
This temperature resulted in an available NPSH of 37.3 ft at 5450 F, considerably larger than the required 13 ft. Figure STI 70-1 summarizes the results of the reactor water cleanup system test in each mode of operation. All test criteria were satisfied.
.. ... ... N . .
-130-FINAL
SUMMARY
REPORT - BFNP UNIT 3 3.0 Results 3.33 STI-71, Residual Heat Removal System 3.33.1 Purpose The purpose of STI-71 is to demonstrate the ability of the Residual Heat Removal (RHR) system to remove residual and decay heat from the nuclear syste-so that refueling and nuclear system servicing can be performed. 3.33,2 Criteria Level 1 Not applicable Level 2 The heat removal capability of each RER heat exchanger in tte shutdown cooling node shall be at least 187 X 10 Btu/hr when the inlet flows and tempera-tures are as indicated on the process diagrams. (See section 8 of this test for sut.ary of flow rates.) 3.33.3 Analysis STI-71 testing was conducted at test conditions 1 as defined on the power flow map in section 2.3 and at hot shutdown. At test condition 1, the capacity of the RER heat exchangers from the shutdown cooling mode test could not be demonstrated due to insufficient decay heat. Also, the suppression pool cooling mode method was un-successful in determining the RER heat exchanger capacity because of an insufficient AT. Therefore, this test .-as repeated following the load rejection trip from test condition 4E. The calculated heat removal capacities tanged from 188.7 to 532 M1tu/lr. Additionally, the head spray capacity was verified by obtaining a rated flow of 1000 gpm. ... t . . .. .
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-131-MFLAL SUIAY REPORT - BF.NP UNIT 3 3.0 Results 3.35 STI-72, Dryvell Atmosphere CoolinS System 3.35.1 Purpose The purpose of this test Is to verify the ability of the drywell atmosphere cooling system to maintain design conditions in the dryuell during operating conditions a~i post-scram conditions.
3.35.2 Criteria Level 1 Not applicable Level 2 The heat removal capability of the drywell coolers shall be approximately 5.19 x 106 Btulhr. The drywell cooling system shall have a standby capability of > 25% of the design heat removal capability. he-drywell coolin; scytczn shall naintaia tar.p-eratures in the drywell below the following design valuas during normal operation. 1ormal reactor ooeration: During 1500 F average throughout drywell 50% relative humidity 1350 F maximu around the recirculating pump motors 2000 F maximum above the bulkhead 1800 F maximum for all other areas Tec hours after shutdown:
- Within 150 F of closed cooling water inlet temperature (average. throughout the drywel)
Cooling water suply-:
- 1000 F maximum
-132-FINAL S'JUARY REPORT - BFNP UMIT 3 3.0 Results (Continued) 3.35 STI-72, DRnyell Atmosphere Cooling System (Continued) 3.35.3 Analysis STI-72 testing was performed at heatup and test condition 4E levels as defined on the power flow map in section 2.3.
Data recorded at each plateau of heatup indicated a uniform temperature increase as was expected. All ter-n eratures were within design limits for this level of testirg. (See table STI 72-1) The estimated beat removal rate of thn drywell coolers was 4.4 x 106 Btu/hr. Drywell humidity could not be evaluated due to the inoperability of instrsent MR-80-36. This item was carried as an exception. It should be noted that the cooling water inlet temperature was 9S!o V. Extrapolation of data to a design maximum of 1000 F inlet temperature indicates that all temperatures will be within design limits. Data recorded at test condition 4E indicated the; all normal operational temperature limits were within design limits. (See table STI 72-1) Extrapolation of data d6rn-.
.hcatup testing to a design maximmn inlet water Lemperacure ox 1000 IF.indicates that all temperatures will be within design limits. The estimated heat removal rate of the drywell coolers was 5.13 x 106 Btu/hr. This meets level 2 criteria, that the cooler heat removal rate be approximately 5.19 x 106 Btu/hr.
Instrument HR-80-36 was repaired prior to reaching test condition 4E. Channels A and B indicated 36% and 53X relative humidity. This cleared the exception to STI-72 daring beatup testing. Level 2 criteria required drywell humidity to be below 50X. Drywell humidity was therefore carried as an exception to STI-72. Following inerting of the unit 3 dry-well HR-60-36 indicated 29% and 33% relative humidity on channels A and B. respectively. This cleared the associated exception. During test condition 4E testing, drywell cooler fans A2 and 12 were inoperative. This prevented testing following a full power.scram to determine if level 2 criteria, requiring that the average drywell temperature be within 15° F of the closed cooling water inlet temperature 10 hours after shutdown, can be met. This item is carried as an exception tc STX-72. Drywell cooling fans A2 and Z2 have been repaired. This test will be performed as soon as plant conditions permit.
-133-FINAL SU,*A.RY REPORT - BFNP UNIT 3 3.0 Results (Continued) 3.35 STI-72, Drywell Atmosphere Cooling System (Continued) 3.35.3 Analysis (Continued) a Table STI 72-1 Parameter Design Limit Heatup T.C. 4E Avg. DWJ Temp. 1500 F 1260 F 130.60 F Recirc. Pumap Temp. 1350 F 109 0o F 1080 F Above Bulkhead Temp. 2000 F 1530 F 1570 F Miax. Temp. Other Areas 1800 F 1500 F 1560 F 6
-134-FINAL
SUMMARY
REPORT - BFWP UNiIT 3 3.0 Results 3.36 STI-73, Cooling 'dater Systems 3.36.1 Purpose The purpose of this test is to verify that the performance of the Reactor Building Closed Cooling Water (RBCCW) system is adequate with the reactor at rated conditions. 3.36.2 Criteria Level 1 Not applicable Level 2 Verification that the system performance meets the cooling requirements constitutes satisfactory completion of this test. The RBCCW was designed to transfer a maximum.heat load to 31.3 x 106 Btu/hr. in order to limit equipment inlez water tezpezature of 1000 F assuming a service (raw cooiing) water inlet temperature of go9 F. 3.36.3 Analysis STI-73 testing was performed at heatup and test condition 4E levels as defined on the power flow map in section 2.3. At hot standby the calculated heat load was 18.98 x 10 Btu/hr on the RBCCW side of the heat exchangers and 21.0 x 106 Btu/hr on the RCW side. At test condition 4e the heat load was 24.86 x 106 Btu/hr on the RBCCW side and 21.86 on the R67 side. It should be noted that the Rai flcw was extremely low at test condition 4E due to cold river water. Therefore, the RCW side heat balance cannot be con-sidered reliable due to inaccuracies in the flow measurement system at low flow rates. Data indicates that the RBCCW system component flow and heat exchangers are properly balanced. Significant para-meters are summarized In table STI 73-1. i *.v
-135-FINAL SU'!flRY REPORT - BFXP UNIT 3 3.0 Results (Continued) 3.36 STI-73, Cooling Water Systems (Continued) 3.36.3 Analysis (Continued)
Due to low RCW flow and temperature it is not possible to extrapolate the data to design rated condit'ons. Therefore, it cannot be determined if design criteria will be met at rated system beat load and temperatures. All criteria were met for conditions at which testing was con-ducted. The RBCCW system is adequate for handling system heat loads until the fuel pool heat exchangers approach design heat load. The Division of Engineering Design is evaluating system performance at rated system heat load and temperature. When RCW teiperatures approach design values, additional testing will be performed to clear this exception. Table STI 73-1 RBCCW Operation at T.C. 4E Max. or Design Measured
-Farameter .Value ';alue Total RBCCW Flow 3369.5 gpm 3648.5 RECCW Inlet Temp.
at. x A 118.50 F 96.2 Et. x B 118.50 F 96.2 RBCCW Outlet Temp. Et. x A 1000 F 84.5 Et. x B 1000 F 80.5 RCW Flow Et. x A 2550 gpm ' 331 gpm Et. r B 2550 gpm "689 gpm RCW Inlet Temp. o Et. z A gO° F 44.4 F at. XB 0 F gO0 44.50 F RCW Outlet Temp. Et. x A 102.3° F 88.80 F Et. x B 102.3 0 F 87.0 0 F Heat Removal. Rate RBCCW Side' 24.86 x 106 Btu/hr. RCW Side . 21.9 x 106 Btu/hr.
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-136-FINAL SMUMARY REPORT - BFNP UNIT 3 3.0 Results (Continued) 3.37 STI-74, Modified Off-Gas System 3.37.1 Purpose The purposes of this test are:
- 1. To verify the proper operation of the off-gas system over its expected operating parameters.
- 2. To determine the performance of the activated carbon idsorbers.
3.37.2 Criteria Level 1 The release of radioactive gaseous particulate effluents must not exceed the linits specified in BrNP technical specifications 3.8.B. There shall be no loss of flow for dilution steam.-. to the noncondensing stages when the steam jet air ejectorz are pumping. Level 2 The system flmw, pressure, temperature, and relative humidity shall comply with the design specifications shown in form 74.6-1. The catalytic recombiner, the hydrogen analyzer, the activated carbon beds, and the filters shall be working as designed. 3.37.3 Analysis STI-74 testing was performed at test conditions 1, 2E, 3E, and 4E as defined on the power flow map in section 2.3. Airborne Releases - Airborne releases during testing were aocumenred-t n surveillance tests SI 4.8.B.l-a and SI 4.8.B.2-6. There were no violations of the BFNP Tech. Specs.' 3.8.B limits at any test condition. Therefore, level 1 criteria were fully satisfied.
FINAL SIMIARY REPORT - B3NP UNIT 3
-137- I 3.0 Results (Continued) 3.37 STI-74,. Modified Off-Gas System (Continued) 3.37.3 Analysis (Continued)
Dilution Steam Flow - There were no losses of dilution steam flow to the noncondensing stages of the punping SJAE during any testing. The total dilution steam flows are recorded in table STI 74-1. Level I criteria were fully satisfied.
.System Parameters - Table STI 74-1 uimrizes system operating parameters during startup.
The system temperatures, pressures, flow, and relative humidity complied with design specifications, except for the following:
- 1) A malfunctioning gauge prevented SJAE outlet pressure from being obtained during test condition
.1. However, the gauge was repaired before subse-quent test conditions where the pressures vere.-
maintained within the normal operating range. TThis was a level 2 crlterion exception.
- 2) Adsorber bed F temperature-anomaly was reported at all test conditions and is believed to be due
; to a cooling effect of moisture being removed from the bed. In addition, the thermocouple that provides
; this temperature as recorded on TRS-66-115 seems to be responding properly, but, as outage time permits, will be examined at the adsorber bed inside the vault.
This was a level 2 criterion exception.
- 3) Hydrogen analyzer malfunctions are discussed below.
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-138-FINAL SPURY REPORT - BFNP UNIT 3 Table STT 74-1 X Pow'er .15-35 40-60 65-85 ,
.Date
_ . 10/4/76 10/11/76 11/3/76 1liu_ System Parameters _____ _ 820 1937 2531
.Operating TC1 TC2E TC3E :C , _
_______________________ Pange DIL Steam low (Total) 9100ihr 9350 -3 9700 1 SJAE OUtlet Pressure 5-10 psir op.(. 5 S OG Preheater T Outlet 275°-360° F 350 350 350 ._6 Bottom 2_50-8750 F 425 555 605 c)3 Middle I_ 2750-87503 F 420 543 605 _ _ TOP 27504-750 F 405 535 585 1 s-: StandbY Recomb. TemP. _ Bottom 275u-360 FI 320 325 320 325 Middle 2750-363 305 315 315 335 Top 275° 360° F 295 315 315 OG Cond. Coolant Out 120 F 110 103 109 , OG Cond. Outlet 'Temp. .Y 140 F 123 119 117 .: ctration 0-11 .05 .05 0 (1? OG Flow 20-40 SCfm 35 35 30 GIXcol Pup P 20-40 P-SiR 19 32 31 38 2 G T330-380 F 34 36 36 Moist. Sep. T Out -S5° F SO 49 55 __ Reheater Dey it poS 42 - 4
-42 3 43_ _____ 4 .
49 74 74 _ _ ,---_ Reheater T Out ________7 Prefilter D.P. 02c aer .05 .2 _ 0 0 3-2.6 ps 2.2 *8 .75
- Adsorber D.P.
Bypass D.P. 0_2" water O O OO 0 Adsorber Vessel T - - _ _ 680-790 F 70.0 72 69(2) _ - 9 ! Bed A Pt. 1 Bed A Ft. 2 680-79" F 71.0 71 68(2) -_ _' _ Bd
- t.3. 68O-790 F 70.0 67.5 68(2) 60 68°-79° F 68.0 70 69(2) C__.
Bed B Pt. 4 Bed C Pt. 5 680-79O r 68.5 69.5 68(2) 69.5 Bed D Pt. 7 680-79O F 70.0 75.5 68.5(2) 69.5 t 680-79° F 62.0 58 64(2) 52.5 i Vault T 730-81-0 F 75.0 73.5 75.5 76.5_I Adeorb After Filter D.P. 0-2" water .35 _ .5 0 .4 401 1 32 32 34 32 Z Rel. Eum. (C)Dats not obtained or was out of operating range and carried as an STI exception . (2)These readings were taken on 11/4/76 at 2490 fWt and the same test condition. I..
-139-FINAL SIMMARY REPORT - ZFNP UNIT 3 3.0 Results (Continued) 3.37 STI-74, Modified Off-Gas System (Continued) 3.37.3 Analsss (Continued)
Tables STI-74-2A and -2B summarize hydrogen analyzer performance data taken during startup. Table STI 74-2A
- Power 15-35 40-60 65-85 00569-)
Date 9/27/76 10/11/76 11/3/76 11/22/76 HYDROGEN AYALYZER PERPOLWAYCE Hut 1038 1937 2531 3274 Normal K2 Analyzer Operating Range T.C. 1 T.C. 2E T.C. 3E T.C. AE Process Reading 2 H2 0-1 .08 .05 0 VfOP Satple Flow scfh 3-4 4 4.0 c2 Demin. Water flow gph .1-2 2 1.5 1.5 Vacuum regulator water 10-25 20 17 10-40 I
'Calibration Standard scfh 3-4 3.5 4.0 2' _
Calibration Standard H2. - 1.0 1.0 1.0 1.0 1 Calibration Gas Results X H2 1.0 1.0 1.1 1.0 92 Free Standard scfh 3-4 3.5 4.0 c 2 U2 Free-Standard Z E2 0. 0 0 0 22 Free Results Z52 .. 0 0 0 0
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. .. .I .
-140-FINAL SMlU4ARY REPORT - B'FNP IMIT 3 3.0 Results (Continued) 3.37 STI-74, Modified Off-Gas System (Continued) 3.37.3 Analysis (Continued)
Table STI 74-2B _ Power 15-35 40-60 65-85 95-109 1 Date 9/27/76 10111176 11/3/76 - 11/7f HYDROGEN AMTALYZER PERFORXANCE 11t 1038 1937 2531 327L Operating 87 Analyzer B Rante T.C. 1 T.C. 2E T.C. 3E T.C. 4 Process Reading 2 H2 Sample Flow scfh Demin. Water flow gph 0-1 3-4 1-2
.05 4
2
.1
.2-4 1.5 IIOP I i .r.-,
i Vacuum regulator water 10-25 17 15 Calibration Standard scfh 3-4 3.8 4.0 Calibration Standard Z H2 1.0 1 1.0 __i Calibration Gas Results Z 82 1.0 1 1.5 E2 Free Standard 6cfh 3-4 3.8 4.0 E2 Free Standard 2 H2 0 0 0 Hz Free Results X H2 0 0 0 a
-141-FINAL S MOARY REPORT - BFNP UNIT 3 3.0 Results (Continued) 3.37 STI-74, Modified Off-Gas System (Continued) 3.37.3 Analysis (Continued)
The hydrogen analyzers were not reliable for continuous process use. This was attributed to moisture which, when condensed, caused erratic sample flow and improper sensor response. Engineering Change Notice, ECK 1825 will Install the required modifications to the hydrogen analyzers to resolve this problem. both bydrogen analyzers failed to perform satis-factorily at test conditions 2E, 3E, and 4E, and, therefore, do not fulfill level 2 criteria. Grab samples taken and analyzed by the radiochenical laboratory insured that the hydrogen concentration was less than 4%. Catalytic Recombiner - Table STI 74-3 summarizes catalytic recombiner performance during startup. Table STI 74-3 Power % 15-35 40-60 65-85 95-100 1 Date 9/27/76 10/11/76 11/3/76 11/22/76 Wt 1038 1937 2531 3274 REC:B: n:R PERFOL:NCE TC T.C. 1 T.C. 2E T.C. 3E T.C. 4} Radiolytic Gas Production Rate, CFNl)Nt .03 .04 .038 .035 Active Recombiner-Temp, 'FO 425 555 605 605 OG Preheater Temp Outlet, OF 350 350 350 340 T Actual, OF 75 250 255 265 AT Expected, °F 87 225 288 261 The catalytic recombiners performed satisfactorily during startup. Level 2 criteria was satisfied.
- . I
-142-FINAL S!M4ARY REPORT - BFMMP UIT 3 3.0 Results (Continued) 3.37 STI-74, Modified Off-Gas Svste:a (Continued) 3.37.3 Analysis (Continued)
Adsorber Beds - Table STI 74-4 su~narizes the calculated residence times for four radionuclides and the Xe/Kr ratios across the six charcoal adsorber beds operated in series.. Table STI 74-4 r.
. Power 15-35 I40-60 65-85 95-100 Charcoal Adsorber Date 9/27/76 0l/9/76' 11/5/76 11/22/76 Performance }$Mt 1038 1890 2555 3274 (Residence Time) T.C. T.C. 1 T.C. 2E T.C. 3E T.C. 4E Kr88 (Actual), Hr. 33 7.6 10.4 15 Kr8Sm (Actual), Hr. 43 7.3 10.1 13 Yr (Expeated), Er. 1R.2 11.5 0,7 15 Xel35 (Actual), Day 7.3 7.8 10.1 10 Xel33 (Actual),-lay 89.7 68 23.8 16 Xe (Expected), Day l1.5 8.8 7.3 12 Ratio Xe/Kr (Actual) 5/1(1) 25/1(l) 23/1(1) 22:1 Ratio Xe/Kr (Expected) 18/1 15/1 18/1 19:1 (1) Xel33 was not averaged into ratio because it was not in equalibrium.
This Was the result of the unilt 1 offgas flow, heavily laden with Xe133, being routed through unit 3 adsorber beds during unit 1 maintenanCe. A.large.X433 inventory remained to slowly be eluted from the unit 3 adsorber beds. Calculated and expected radionuclide delay times
- through the adsorber beds showed good agreement at all test conditions. In particular, fuel power testing performed after several days.of steady reactor operation represented the expected adsorption of the Xe and Kr radionuclides.
Level 2 criteria has been satisfied. - .. ... .:.. . i - ,
. . , 7. . - .
-143-FINiAL SWUM REPORT - BFnP UNIT 3 3.0 Results (Coatinued) 3.37 STI-74, Modified Off-Gas Syste=_ (Continued) 3.37.3 Anaiysis (Continued)
System PEPA Filters - Table STI 74-5 summarizes the results of radiochemical testing of the offgas system prefilter and after filters. Table STI .74-5 I.
. -Footnotes on next page w,
. .s .
-144-FINAL SUMbARY REPORT - BFNP UNIT 3 3.0 Results (Continued) 3.37 STI-74, Modified Off-Gas System (Continued) 3.37.3 'Analyis (Continued)
Table STI 74-5 (Continued) (1) Activity levels of Bal4O before and after both the prefilters and afterfilter were too low to detect statistically. Therefore, the calculated effice+/-ciz^_ were meaningless and were omitted from this test. (2). ">)t means that the actual efficiency is some Statue larger than this value, but because a concentration (or both) used to calculate the efficiency was +/-tscrL less than the detectable concentration, the actual value could not be determined. (3) When the afterfilter outlet concentration was dcc 7-corrected to sample time, this effluent appeared ta; have more activity than the inlet. (The efficienrcies were negative.) Actually, both the inlet and out.l.s-t had activity levels too low to detect statistica._:;. This was remedied at test conditions 3E and 4E bv using a partial prefilter bypass. Efficiencies of the prefilters were measured and found to be satisfactory. Laboratory analyses of the afterfilters indicated that they were operating properly. Level 2 criteria were satisfied. All required startup testing for the modified oFf-c-s system has been satisfactorily completed with those exceptions listed.
-145-9 a . -
FINAL SUM}ARY REPORT - BFNP UNIT 3 3.0 Results 3.37 STI-75, Reactor Scram From Outside Main Control Room 3.37.1 Purpose The purposes of STI-75 are:
- 1. To demonstrate that the plant design permits safe reactor shutdown from outside the main control room.
- 2. To demonstrate that the reactor can be maintained in a safe condition after shutdown from outside the main control room.
- 3. To demonstrate that the minimum number of personnel required by the tech specs is adequate to perfcr.-
steps 1.1 and 1.1.1 without affecting the safe continuous operation of the other units.
- 4. To demonstrate that EOI-34, Control Room Abandon-ment, is adequate to perform steps 1.1, 1.1.1, and 1.1.2 without affecting unit safety.
3.37.2 Criteria Level 1 Not applicable. Level 2 Initiation of reactor scram must occur from out-side the main control room. Reactor water level must be maintained greater than 490" above vessel zero level and less than the high level turbine trip point. The RER and RERSW pumps and control valves shall be operable from the backup controls to initiate suppression pool cooling. The metnmum number of personnel as required by the tech specs can conduct this test. 3.37,3, An sis STI-75 was conducted at a power level of 11.5% with the .f. &. . .
6-FINA1L SUMXOARY REPORT - BPNP UNIT 3 3.0 Resilts (Continued) 3.37 STI-75 -Reactor Scram From Outside Main Control Room (Coutinu-) 3.,37.3 Ar.alysis (Continued) turbine/generator off-line. Control was transferred fro-. the varin control room to the remote panel 25-32 prior to initiating a reactor scram by closure of the MSIV's. Rcacto-.water level on a Yarway initially started at +L5" and decreased to +10" after the scram. The reactor core isolation cooling system initiated to maintain level 2t
+10". The minimum water level observed was 538 inches above vessel zero (+10 inches on Yarvay A). The maxi-.u water level observed was 566 inches above vessel zero, well below the high level turbine trip setpoint at 532 inches.
There were no unexpected events during the performance of this test and all test criteria were satisfied. Prior to terminating the test (at = 17 minutes), the following plar.t conditions were ob-servedt RIp - 155& ig iMR, A _ Z:C:.; P1t. A 3.50 crz PRHR1DR B - 6 0 psig EECW Pcmp 3 -O REIHDR C - 40 psig EECW Pump C.- 0 RER 11DR D - 70 psig EECW Pump D - 3500 gpm RCIC Flow - 520 g8! Drywall Temp - 85 F Suppression Chamber Temp - 1l0 F Suppression Chamber Level - 2 inch Reactor Pressure - 660 psig
F ENCLOSURE 6 TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT (BFN) UNITS 2 AND 3 EXTENDED POWER UPRATE RS-001 REVISED TEMPLATE SAFETY EVALUATION The attached pages have been revised. On the affected pages, the revised portions have been highlighted. A line has been drawn through the deleted text and a double underline for new or revised text.
2.8.4.4 Residual Heat Removal System Regulatory Evaluation The RHR system is used to cool down the RCS following shutdown. The RHR system is typically a low pressure system which takes over the shutdown cooling function when the RCS temperature is reduced. The NRC staff's review covered the effect of the proposed EPU on the functional capability of the RHR system to cool the RCS following shutdown and provide decay heat removal. The NRC's acceptance criteria are based on (1) draft GDC-40 and 42, insofar as they require that ESFs be protected against dynamic effects; and (2) draft GDC-4, insofar as it requires that reactor facilities shall not share systems or components unless it is shown safety is not impaired by the sharing; and (3) draft GDC-6. insofar as it requires that decay heat removal systems shall be provided for all expected conditions of normal operation. Specific review criteria are contained in SRP Section 5.4.7 and other guidance provided in Matrix 8 of RS-001. Technical Evaluation [Insert technical evaluation. The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.] Conclusion The NRC staff has reviewed the licensee's analyses related to the effects of the proposed EPU on the RHR system. The NRC staff concludes that the licensee has adequately accounted for the effects of the proposed EPU on the system and demonstrated that the RHR system will maintain its ability to cool the RCS following shutdown and provide decay heat removal. Based on this, the NRC staff concludes that the RHR system will continue to meet the requirements of draft GDC-4, ft40 6 and 42 following implementation of the proposed EPU. Therefore, the NRC staff finds the proposed EPU acceptable with respect to the RHR system. INSERT S FOR SECTION 3.2 - BWR TEMPLATE SAFETY EVALUATION DECEMBER 2003
2.8.5 Accident and Transient Analyses 2.8.5.1 Decrease in Feedwater Temperature, Increase in Feedwater Flow, Increase in Steam Flow, and Inadvertent Opening of a Main Steam Relief or Safety Valve Regulatory Evaluation Excessive heat removal causes a decrease in moderator temperature which increases core reactivity and can lead to a power level increase and a decrease in shutdown margin. Any unplanned power level increase may result in fuel damage or excessive reactor system pressure. Reactor protection and safety systems are actuated to mitigate the transient. The NRC staff's review covered (1) postulated initial core and reactor conditions, (2) methods of thermal and hydraulic analyses, (3) the sequence of events, (4) assumed reactions of reactor system components, (5) functional and operational characteristics of the reactor protection system, (6) operator actions, and (7) the results of the transient analyses. The NRC's acceptance criteria are based on (1) draft GDC-6, insofar as it requires that the reactor core be designed to function throughout its design lifetime without exceeding acceptable fuel damage limits; (2) draft GDC-9. insofar as it requires that the reactor coolant pressure boundary shall be desined and constructed so as to have an exceedingly low Probability of gross rupture or significant leakage throughout its design lifetime: (g3) draft GDC-14 and 15, insofar as they require that the core protection system be designed to act automatically to prevent or suppress conditions that could result in exceeding acceptable fuel damage limits and that protection systems be provided for sensing accident situations and initiating the operation of necessary ESFs; and (+34) draft GDC-27 and 28, insofar as they require that at least two reactivity control systems be provided and be capable of making and holding the core subcritical from any hot standby or hot operating condition sufficiently fast to prevent exceeding acceptable fuel damage limits. Specific review criteria are contained in SRP Section 15.1.1-4 and other guidance provided in Matrix 8 of RS-001. Technical Evaluation [Insert technical evaluation. The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a INSERT 8 FOR SECTION 3.2 - BWR TEMPLATE SAFETY EVALUATION DECEMBER 2003
clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.] Conclusion The NRC staff has reviewed the licensee's analyses of the excess heat removal events described above and concludes that the licensee's analyses have adequately accounted for operation of the plant at the proposed power level and were performed using acceptable analytical models. The NRC staff further concludes that the licensee has demonstrated that the reactor protection and safety systems will continue to ensure that the AFDLs and the RCPB pressure limits will not be exceeded as a result of these events. Based on this, the NRC staff concludes that the plant will continue to meet the requirements of draft GDC-6, K 14, 15, 27, and 28 following implementation of the proposed EPU. Therefore, the NRC staff finds the proposed EPU acceptable with respect to the events stated. INSERT 8 FOR SECTION 3.2 - BWR TEMPLATE SAFETY EVALUATION DECEMBER 2003
2.8.5.2 Decrease in Heat Removal by the Secondary System 2.8.5.2.1 Loss of External Load; Turbine Trip; Loss of Condenser Vacuum; Closure of Main Steam Isolation Valve; and Steam Pressure Regulator Failure (Closed) Regulatory Evaluation A number of initiating events may result in unplanned decreases in heat removal by the secondary system. These events result in a sudden reduction in steam flow and, consequently, result in pressurization events. Reactor protection and safety systems are actuated to mitigate the transient. The NRC staff's review covered the sequence of events, the analytical models used for analyses, the values of parameters used in the analytical models, and the results of the transient analyses. The NRC's acceptance criteria are based on (1) draft GDC-6, insofar as it requires that the reactor core be designed to function throughout its design lifetime without exceeding acceptable fuel damage limits; (2) draft GDC-9, insofar as it requires that the reactor coolant pressure boundary shall be designed and constructed so as to have an exceedingly low probability of gross rupture or significant leakage throughout its desicn lifetime; and (go) draft GDC-27 and 28, insofar as they require that at least two reactivity control systems be provided and be capable of making and holding the core subcritical from any hot standby or hot operating condition sufficiently fast to prevent exceeding acceptable fuel damage limits. Specific review criteria are contained in SRP Section 15.2.1-5 and other guidance provided in Matrix 8 of RS-OO1. Technical Evaluation (Insert technical evaluation. The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.] Conclusion The NRC staff has reviewed the licensee's analyses of the decrease in heat removal events described above and concludes that the licensee's analyses have adequately accounted for operation of the plant at the proposed power level and were INSERT 8 FOR SECTION 3.2 - BWR TEMPLATE SAFETY EVALUATION DECEMBER 2003
performed using acceptable analytical models. The NRC staff further concludes that the licensee has demonstrated that the reactor protection and safety systems will continue to ensure that the AFDLs and the RCPB pressure limits will not be exceeded as a result of these events. Based on this, the NRC staff concludes that the plant will continue to meet the requirements of draft GDC-6, 9. 27, and 28 following implementation of the proposed EPU. Therefore, the NRC staff finds the proposed EPU acceptable with respect to the events stated. INSERT SFOR SECTION 3.2 - BWR TEMPLATE SAFETY EVALUATION DECEMBER 2003
2.8.5.2.2 Loss of Nonemergency AC Power to the Station Auxiliaries Regulatory Evaluation The loss of nonemergency ac power is assumed to result in the loss of all power to the station auxiliaries and the simultaneous tripping of all reactor coolant circulation pumps. This causes a flow coastdown as well as a decrease in heat removal by the secondary system, a turbine trip, an increase in pressure and temperature of the coolant, and a reactor trip. Reactor protection and safety systems are actuated to mitigate the transient. The NRC staff's review covered (1) the sequence of events, (2) the analytical model used for analyses, (3) the values of parameters used in the analytical model, and (4) the results of the transient analyses. The NRC's acceptance criteria are based on (1) draft GDC-6, insofar as it requires that the reactor core be designed to function throughout its design lifetime without exceeding acceptable fuel damage limits; (2) draft GDC-9. insofar as it requires that the reactor coolant pressure boundary shall be designed and constructed so as to have an exceedingly low probability of gross rupture or significant leakaae throughout its design lifetime; and (2_3) draft GDC-27 and 28, insofar as they require that at least two reactivity control systems be provided and be capable of making and holding the core subcritical from any hot standby or hot operating condition sufficiently fast to prevent exceeding acceptable fuel damage limits. Specific review criteria are contained in SRP Section 15.2.6 and other guidance provided in Matrix 8 of RS-OO1. Technical Evaluation [Insert technical evaluation. The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.] Conclusion The NRC staff has reviewed the licensee's analyses of the loss of nonemergency ac power to station auxiliaries event and concludes that the licensee's analyses have adequately accounted for operation of the plant at the proposed power level and were INSERT 8 FOR SECTION 3.2 - BWR TEMPLATE SAFETY EVALUATION DECEMBER 2003
performed using acceptable analytical models. The NRC staff further concludes that the licensee has demonstrated that the reactor protection and safety systems will continue to ensure that the AFDLs and the RCPB pressure limits will not be exceeded as a result of this event. Based on this, the NRC staff concludes that the plant will continue to meet the requirements of draft GDC-6, L 27, and 28 following implementation of the proposed EPU. Therefore, the NRC staff finds the proposed EPU acceptable with respect to the loss of nonemergency ac power to station auxiliaries event. INSERT 8 FOR SECTION 3.2 - BWR TEMPLATE SAFETY EVALUATION DECEMBER 2003
2.8.5.2.3 Loss of Normal Feedwater Flow Regulatory Evaluation A loss of normal feedwater flow could occur from pump failures, valve malfunctions, or a LOOP. Loss of feedwater flow results in an increase in reactor coolant temperature and pressure which eventually requires a reactor trip to prevent fuel damage. Decay heat must be transferred from fuel following a loss of normal feedwater flow. Reactor protection and safety systems are actuated to provide this function and mitigate other aspects of the transient. The NRC staff's review covered (1) the sequence of events, (2) the analytical model used for analyses, (3) the values of parameters used in the analytical model, and (4) the results of the transient analyses. The NRC's acceptance criteria are based on (1) draft GDC-6, insofar as it requires that the reactor core be designed to function throughout its design lifetime without exceeding acceptable fuel damage limits; (2) draft GDC-9, insofar as it requires that the reactor coolant pressure boundary shall be designed and constructed so as to have an exceedingly low probability of gross rupture or significant leakage throughout its design lifetime; and (23) draft GDC-27 and 28, insofar as they require that at least two reactivity control systems be provided and be capable of making and holding the core subcritical from any hot standby or hot operating condition sufficiently fast to prevent exceeding acceptable fuel damage limits. Specific review criteria are contained in SRP Section 15.2.7 and other guidance provided in Matrix 8 of RS-001. Technical Evaluation [Insert technical evaluation. The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.] Conclusion The NRC staff has reviewed the licensee's analyses of the loss of normal feedwater flow event and concludes that the licensee's analyses have adequately accounted for operation of the plant at the proposed power level and were performed using acceptable analytical models. The NRC staff further concludes that the INSERT 8 FOR SECTION 3.2 - BWR TEMPLATE SAFETY EVALUATION DECEMBER 2003
licensee has demonstrated that the reactor protection and safety systems will continue to ensure that the AFDLs and the RCPB pressure limits will not be exceeded as a result of the loss of normal feedwater flow. Based on this, the NRC staff concludes that the plant will continue to meet the requirements of draft GDC-6, L 27, and 28 following implementation of the proposed EPU. Therefore, the NRC staff finds the proposed EPU acceptable with respect to the loss of normal feedwater flow event. INSERT 8 FOR SECTION 3.2 - BWR TEMPLATE SAFETY EVALUATION DECEMBER 2003
N 2.8.5.3 Decrease in Reactor Coolant System Flow 2.8.5.3.1 Loss of Forced Reactor Coolant Flow Regulatory Evaluation A decrease in reactor coolant flow occurring while the plant is at power could result in a degradation of core heat transfer. An increase in fuel temperature and accompanying fuel damage could then result if AFDLs are exceeded during the transient. Reactor protection and safety systems are actuated to mitigate the transient. The NRC staff's review covered (1) the postulated initial core and reactor conditions, (2) the methods of thermal and hydraulic analyses, (3) the sequence of events, (4) assumed reactions of reactor systems components, (5) the functional and operational characteristics of the reactor protection system, (6) operator actions, and (7) the results of the transient analyses. The NRC's acceptance criteria are based on (1) draft GDC-6, insofar as it requires that the reactor core be designed to function throughout its design lifetime without exceeding acceptable fuel damage limits; (2) draft GDC-9, insofar as it recuires that the reactor coolant pressure boundary shall be designed and constructed so as to have an exceedingly low probability of gross rupture or significant leakage throughout its design lifetime: and (23) draft GDC-27 and 28, insofar as they require that at least two reactivity control systems be provided and be capable of making and holding the core subcritical from any hot standby or hot operating condition sufficiently fast to prevent exceeding acceptable fuel damage limits. Specific review criteria are contained in SRP Section 15.3.1-2 and other guidance provided in Matrix 8 of RS-O01. Technical Evaluation [Insert technical evaluation. The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.] Conclusion The NRC staff has reviewed the licensee's analyses of the decrease in reactor coolant flow event and concludes that the INSERT 8 FOR SECTION 3.2 - BWR TEMPLATE SAFETY EVALUATION DECEMBER 2003
licensee's analyses have adequately accounted for operation of the plant at the proposed power level and were performed using acceptable analytical models. The NRC staff further concludes that the licensee has demonstrated that the reactor protection and safety systems will continue to ensure that the AFDLs and the RCPB pressure limits will not be exceeded as a result of this event. Based on this, the NRC staff concludes that the plant will continue to meet the requirements of draft GDC-6, 9, 27, and 28 following implementation of the proposed EPU. Therefore, the NRC staff finds the proposed EPU acceptable with respect to the decrease in reactor coolant flow event. INSERT 8 FOR SECTION 3.2 - BWR TEMPLATE SAFETY EVALUATION DECEMBER2003
1 2.8.5.4.3 Startup of a Recirculation Loop at an Incorrect Temperature and Flow Controller Malfunction Causing an Increase in Core Flow Rate Regulatory Evaluation A startup of an inactive loop transient may result in either an increased core flow or the introduction of cooler water into the core. This event causes an increase in core reactivity due to decreased moderator temperature and core void fraction. The NRC staff's review covered (1) the sequence of events, (2) the analytical model, (3) the values of parameters used in the analytical model, and (4) the results of the transient analyses. The NRC's acceptance criteria are based on (1) draft GDC-6, insofar as it requires that the reactor core be designed to function throughout its design lifetime without exceeding acceptable fuel damage limits; (2) draft GDC-9. insofar as it reauires that the reactor coolant pressure boundary shall be designed and constructed so as to have an exceedingly low probability of gross rupture or significant leakage throughout its design lifetime: (23) draft GDC-14 and 15, insofar as they require that the core protection systems be designed to act automatically to prevent or suppress conditions that could result in exceeding acceptable fuel damage limits and that protection systems be provided for sensing accident situations and initiating the operation of necessary ESFs; (4i) draft GDC-32, insofar as it requires that limits, which include considerable margin, be placed on the maximum reactivity worth of control rods or elements and on rates at which reactivity can be increased to ensure that the potential effects of a sudden or large change of reactivity cannot (a) rupture the reactor coolant pressure boundary or (b) disrupt the core, its support structures, or other vessel internals sufficiently to impair the effectiveness of emergency core cooling; and (45) draft GDC-27 and 28, insofar as they require that at least two reactivity control systems be provided and be capable of making and holding the core subcritical from any hot standby or hot operating condition sufficiently fast to prevent exceeding acceptable fuel damage limits. Specific review criteria are contained in SRP Section 15.4.4-5 and other guidance provided in matrix 8 of RS-001. Technical Evaluation INSERT 8 FOR SECTION 3.2 - BWR TEMPLATE SAFETY EVALUATION DECEMBER 2003
[Insert technical evaluation. The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.] Conclusion The NRC staff has reviewed the licensee's analyses of the increase in core flow event and concludes that the licensee's analyses have adequately accounted for operation of the plant at the proposed power level and were performed using acceptable analytical models. The NRC staff further concludes that the licensee has demonstrated that the reactor protection and safety systems will continue to ensure that the AFDLs and the RCPB pressure limits will not be exceeded as a result of this event. Based on this, the NRC staff concludes that the plant will continue to meet the requirements of draft GDC-6, 9 14, 15, 27, 28, and 32 following implementation of the proposed EPU. Therefore, the NRC staff finds the proposed EPU acceptable with respect to the increase in core flow event. INSERT 8 FOR SECTION 3.2 - BWR TEMPLATE SAFETY EVALUATION DECEMBER 2003
2.8.5.5 Inadvertent Operation of ECCS or Malfunction that Increases Reactor Coolant Inventory Regulatory Evaluation Equipment malfunctions, operator errors, and abnormal occurrences could cause unplanned increases in reactor coolant inventory. Depending on the temperature of the injected water and the response of the automatic control systems, a power level increase may result and, without adequate controls, could lead to fuel damage or overpressurization of the RCS. Alternatively, a power level decrease and depressurization may result. Reactor protection and safety systems are actuated to mitigate these events. The NRC staff's review covered (1) the sequence of events, (2) the analytical model used for analyses, (3) the values of parameters used in the analytical model, and (4) the results of the transient analyses. The NRC's acceptance criteria are based on (1) draft GDC-6, insofar as it requires that the reactor core be designed to function throughout its design lifetime without exceeding acceptable fuel damage limits; (2) draft GDC-9, insofar as it requires that the reactor coolant Pressure boundary shall be designed and constructed so as to have an exceedingly low Probability of gross rupture or significant leakage throughout its design lifetime: and (23) draft GDC-27 and 28, insofar as they require that at least two reactivity control systems be provided and be capable of making and holding the core subcritical from any hot standby or hot operating condition sufficiently fast to prevent exceeding acceptable fuel damage limits. Specific review criteria are contained in SRP Section 15.5.1-2 and other guidance provided in Matrix 8 of RS-001. Technical Evaluation [Insert technical evaluation. The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.] Conclusion The NRC staff has reviewed the licensee's analyses of the inadvertent operation of ECCS or malfunction that increases reactor coolant inventory and concludes that the licensee's INSERT 8 FOR SECTION 3.2 - BWR TEMPLATE SAFETY EVALUATION DECEMBER2003
analyses have adequately accounted for operation of the plant at the proposed power level and were performed using acceptable analytical models. The NRC staff further concludes that the licensee has demonstrated that the reactor protection and safety systems will continue to ensure that the AFDLs and the RCPB pressure limits will not be exceeded as a result of this event. Based on this, the NRC staff concludes that the plant will continue to meet the requirements of draft GDC-6, 9, 27, and 28 following implementation of the proposed EPU. Therefore, the NRC staff finds the proposed EPU acceptable with respect to the inadvertent operation of ECCS or malfunction that increases reactor coolant inventory. INSERT 8 FOR SECTION3.2 - BWR TEMPLATE SAFETY EVALUATION DECEMBER 2003
2.8.5.6 Decrease in Reactor Coolant Inventory 2.8.5.6.1 Inadvertent Opening of a Pressure Relief Valve Regulatory Evaluation The inadvertent opening of a pressure relief valve results in a reactor coolant inventory decrease and a decrease in RCS pressure. The pressure relief valve discharges into the suppression pool. Normally there is no reactor trip. The pressure regulator senses the RCS pressure decrease and partially closes the turbine control valves (TCVs) to stabilize the reactor at a lower pressure. The reactor power settles out at nearly the initial power level. The coolant inventory is maintained by the feedwater control system using water from the condensate storage tank via the condenser hotwell. The NRC staff's review covered (1) the sequence of events, (2) the analytical model used for analyses, (3) the values of parameters used in the analytical model, and (4) the results of the transient analyses. The NRC's acceptance criteria are based on (1) draft GDC-6, insofar as it requires that the reactor core be designed to function throughout its design lifetime without exceeding acceptable fuel damage limits; (2) draft GDC-9. insofar as it requires that the reactor coolant pressure boundary shall be designed and constructed so as to have an exceedingly low probability of gross rupture or significant leakage throughout its design lifetime: and (a_) draft GDC-27 and 28, insofar as they require that at least two reactivity control systems be provided and be capable of making and holding the core subcritical from any hot standby or hot operating condition sufficiently fast to prevent exceeding acceptable fuel damage limits. Specific review criteria are contained in SRP Section 15.6.1 and other guidance provided in Matrix 8 of RS-001. Technical Evaluation CInsert technical evaluation. The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.] Conclusion INSERTS FOR SECTION 3.2 - BWR TEMPLATE SAFETY EVALUATION DECEMBER 2003
The NRC staff has reviewed the licensee's analyses of the inadvertent opening of a pressure relief valve event and concludes that the licensee's analyses have adequately accounted for operation of the plant at the proposed power level and were performed using acceptable analytical models. The NRC staff further concludes that the licensee has demonstrated that the reactor protection and safety systems will continue to ensure that the AFDLs and the RCPB pressure limits will not be exceeded as a result of this event. Based on this, the NRC staff concludes that the plant will continue to meet the requirements of draft GDC-6, L 27, and 28 following implementation of the proposed EPU. Therefore, the NRC staff finds the proposed EPU acceptable with respect to the inadvertent opening of a pressure relief valve event. INSERT 8 FOR SECTION 3.2 - BWR TEMPLATE SAFETY EVALUATION DECEMBER 2003 I}}