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Final Summary Report, Startup Retest Program Unit 1, Browns Ferry Nuclear Plant
	ML18283B545
Person / Time
Site:	Browns Ferry
Issue date:	07/11/1977
From:	; Tennessee Valley Authority
To:	; Office of Nuclear Reactor Regulation
References
	Download: ML18283B545 (62)
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{{#Wiki_filter:TENNESSEE VALLEY AUTHORITY Division of Power Production FINAL SIR2fARY REPORT STARTUP RETEST PROGRAM UNIT l BROWNS FERRY NUCLEAR PLANT Submitted by Approved by lani Superinterident Chief, N . lear Generation Branch

TABLE OF CONTENTS Page 1.0 In reduction I 1.1 Purpose ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 1.2 Plant Description ~ ~ ~ ~ ~ ~ 1.3 Startup Retest Program . 1.4 Startup Test Description ~ ~ P

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1.5 Startup Test Acceptance Criteria ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 2.0 5un~mar 2.1 Significant Dates of the Startup Test Program 2.2 Test Completion Dates for Startup Test 2.3 Power Flow Hap with Startup Test Conditions 3.0 Results 3.1 SRI-3$ Fuel Loading 3.2 SRI-4, Full Core Shutdown Margin . 3.3 SRI-5 Control Rod Drives 10 3.4 SRI>>6, SR% Performance and Control Rod Sequence 15 3.5 SRI-10% IRM Performance 17 3.6 SRI-11, LPIQi Calibration 3.7 SRX-12, APR'f Calibrat'on 20 3.8 SRX-14, ROTC RC C ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 0 3.9 SRX-15, HPCI 25 3.10 SRI-19, Core Performance . 28 3.11 SRX-22, Pressure Regulator 30

3. 12 SRI-23, Feedwater System . 34 3.13 SRI-25, Ifain Steam Isolation Valves 37 3.14 SRX-26, Relief Valves 38 3 15 SRI-30, Recirculation System .

     ~                                                                                         39 3.16 SRI-32 Recirculation Speed Control                                                     40 3.17 SRI-35, Recirculation Syst'm Flow Calibration                                          43 3.18 SRI-72, Drywell Atmosphere Cooling System 3.19 SRX>>73, Cool'ng Vater Systems                                                          47 3.20'TX-74, Off-Gas System                                                                  49

! FINAL'lkÃhRY REPORT STARTUP RETEST PROGRAM BROWNS FERRY NUCLEAR PLANT UNIT 1 Abstract The final report of the startup retest program performed at Browns Ferry Nuclear Plant unit 1 is presented in three parts: (1) Xptroduc'tion, (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. h'here deviations were noted, resolutions or correc'tive actions are also described. 1.0 Introduction 1.1 ~Pur ose The purpose of this report is to present a concise summary and pertinent detailed results obtained in the performance of startup retests at Browns Ferry Nuclear Plant Unit 1. The startup retest program embraced core physics, thermal-hydraulic, electromechanical and overall system dynamic performance. 1.2 Plant Descri tion Browns Ferry Nuclear Plant Unit 1 is a single-cycle, boiling water reactor designed by General Electric Company (GE) for the Tennessee Valley

'Authority (TVA) and is the third unit of a three-unit site to be placed in service following the March 22, 1975, cable tray fire. The plant is located on the Tennessee River in Northern Alabama.         The design gross electrical output is 1098 Mv'e,derived from a core thermal power of 3293 tMt.

1.3 Startu Retest Pro ram The objectives of the program are to assure that unit 1 has been restored to its pre-fire condition and still meets design, construction, con-tractual, and licensing requirements and that all tests are performed in a controlled and orderly manner. The startup retests are conducted at power levels of about 0 (fuel loading, initial criticality, and heatup), 10, 25, 50, '5, and 100 percent. 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 retest program began with fuel loading on July 4, 1976, and continued through completion of the 100K power testing.

STARTUP RETEST PROGI49't SU".MARY REPORT HFNP UNIT .1 1.4 Startu Test Descri tion'he startup retest instructions were prepared by the results supervisor. The Plant Operations Review. Committee (PORC) reviewed the instruc-tions and recommended 'approval. All.startup retests compi.led with the FSAR test description as modified by the startup retest scoping document. hhere tests had direct bearing on plant modifications, DED reviewed the speci.fic instructions on a selective basis. Final approval for use was made by the plant superintendent. The ~faster Startup Retest Instruction (HSRI) coordinated and docu"..:ented all test activities from initial fuel loading to the completion of all startup retests. This instruction provi.ded guidance for sequence of events, and control points for satisfactory test completion and review before power ascension. 1.5 Startu Test Acceptance Criteria The Startup.Retest 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 1 criteria include the values of process variables assigned in the design of the plant and equipment. If a level 1 criterion is not sati.,- fied, the plant is placed in a satisfactory hold condition until a resolution is made. Tests comparaole with this hold condition may be continued. Folio"ing resolution, applicable tests must be repeated to veri~ that the requirements'f the level 1 criterion are satisfied. The level 2 criteria are associated with expectations n regard to performance of the system. If a level 2 criterion is not satisfied, operatin" 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 retest results were reviewed and recommended for approval by PORC and the plant superintendent and are under-going a final review and evaluation by TVA DED. 2.0 Summar of Test Program 2.1 Chronolo,v of Test Propram

           Table 2.1 presents the dates for significant events in the unit                                   1 startup retest program.

2.2 Startu Rcte.,t Comiletion Dates Table 2.2 presents a summary of th'e dates of completion for all startup tests at each test condition.

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          'TARTUP RETEST PRCGIWf SU~MARY REPORT   - BFNP  UNIT 1 2.0        Summary   of Test Pro ram (Continued) 2.3         Power Flow Ma Figure 2.1 presents a power flow map for Browns Ferry unit 1, showing flow control lines          and the nominal positions of test conditions for the startup retest program.

Table 2-1 Major Events of Unit 1 Startup Retest Program Date Event July 4, 1976 First fuel assembly loaded. July 23, 1976 Core fully loaded to 764 fuel assemblies. September 3, 1976 STI-4, Shutdown Margin Demonstration completed.

   . September 14, 1976               Initial Criticality September 18,1976                Begin initial nuclear    heatup September 20, 1976               Reached rated temperature and pressure September      22,1976           Initial generator    synchronization September      24,1976           Completion  of Heatup Test Phase October 7, 1976                  Completion  of 15-40% testing November 1, 1976                 Completion  of 40-60% testing November 23, 1976                Completion   of 65-80% testing December 14, 1976               100% power first attained
     . May    14, 1977                 Completion of 100% testing

Table 2.2 UNIT 1 STARTUP R1".I'I:.S'1'Rrrr:Rh~i1 50% Plow 75% 1'low 100% Flow Control Line Control Line Control Line Retest Power T Open l5-li0 30-50 40-6 37-57 50-70 65-85 55 70-90 95-100 err. ber Flow % Vessel Heatup <47% ~70 ~104 ~48 ~70 ~102 ~48 ~70 "-100 Test Condition Cold 2D 2E 3C 31) 3E 4C 4D 4E Fuel Loading 7-23 Full 'Core SD'1 5 CRD 9-23 10-5 S?'1 Perf. 6 Control Rod Sequence -13 9-22 10-6 10 Ill.'1 Performance 9-14 9-18 10-1 ll a2

         . LP?..'l C Ai'R."

libration Cal'br tion 9-18 10-4 2-15 1-19 RC1C 9-24 9-30 15 11!'CI 9-24 Core Performance 9-27 10-31'0-10 2-14 Pres. Reg.:Setpt. Ch 9-30 11-14 5-14 22 11-14 Hacku Reg, '-14 23 F':.'vs:Fli Pump Trip Le;el Setpt. Chg. 9-30 10-29 11-15 2-16 25 1lS1V 9-23 Pelief Valve 9 )0 Peck!.c. 5 s. Perf. Bat; 16'- Recirc. Flow Control 10-2 2-16 33 Recirc. S s. Flow Cal. 1-20 D'. Coolinp 9-23 2-5 73 9-22 2-23..I 74 RBCC';.'ff-";as S stem 10-25 2-16

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10 20 0 30 40 50 60 70 80 90 100 CORE RECIRCULATION FLOW, (% of 102.5 x 10 1bs/hr) Test Condition No, 2D 2E 3C . 3) 3F. 4C 4D 4E ROD PATTERN a b

                                 %  Pu...p S eed                          'lI 41              ~68                                 E            ~41        ~68
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                                 /  Co e   Flow                           +4     >'           " 70< ~'u3.04';                              .
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onlltant ulllp eel T.ines P a Rod Pattern obtained at T.C. 2E A Natural Circulation b Rod Pattern obtained at T.C. 3E 8 20% Pump Speed c Rod Pattern obtained at T.C. 4E C Analytical lol"er limit of master flow control (~41% speed)

~ Asterisked values are set as initial D Contractual lower limit of flow control (~68% speed) test cond.; non-asterisked values are E Pump speed for rated flow at rated po::er eye J ~ ~

V V;rier Figure 2.1 Approximate Power Flow }!ap S}lowin8Unit 1 Startllp Retest Conditions

I STARTUP RETEST PROCsRAM SU~MARY REPORT BFNP UNIT 1 3.0 Results 3.1 SRI-3 Fuel Loadin 3.1.1 ~Pur. ese

1. To load fuel with the CRD's disarmed.

2. To reload the core to the same configuration that

                             ,existed prior to the 0farch 22, 1975, fire.

3.1.2 Criteria None. 3.1.3 ~Anal s1s SRI-3 was conducted during open vessel testing as-defined on the power flow map in section 2.3. The core was reloaded in the reverse order of defuel-ing following the 5farch 22, 1975, fire. There .~ere no unexpect-ed reactivity anomalies and the core was fully loaded while remaining in a subcritical state. The final core inspection verified that all fuel assemblies were properly seated and in the correct core locations. F-el 'oading was completed in a

                 ,twenty day period, starting July 4 'nd finishing July 2~3 1976 fuel loading began, all control rods w re fu   ':'efore inserted, electrically disarmed and valved out such that the control rods could not be moved. Four gamma compensated fuel loading chambers were used with the normal SIQf electronics to monitor the core until the nor,.al SRN could detect >3 cps.

The FLC/S1Q( shorting links were removed, p'acing them in a non-coincidence scram mode. a An inverse count rate (1/H) ~lot was mainta'ined for each FLC/SIN to control the loading increment. Expected behavior was observed on all 1/H plots. The core verification for fue1 assembly seating was performed by lowering the empty refueling platform grapple to near the top of the fuel assembly handles and traversing across each fuel assembly. The core verification for fuel assembly location and orientation was performed by video-taping each fuel assembly. The core configuration was verified to be in the same configura-tion that existed before the Narch 22, 1975, fire.

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STARTUP RETEST PROG1UQf SU".f'fARY RFPORT - BFNP UNIT 1 3.0 Results (Continued) 3.2 SRI-. 4, Full Core Shutdown afar in 3.2.1 ~Pur ass The purpose of this test is to demonstrate that the reactor will be subcritical throughout the first fuel cycle with any single control rod fully withdrawn. 3.2.2 Criteria Level 1 The shutdown margin of the fully loaded core with the

analytically strongest rod withdrawn must be at least 0..38% hk/k on unit 1 at 68 F. Level 2 There are no level 2 'criteria. 3.2.3 ~Anal sls SRI 4 was conducted during open vessel testing as defined on the power flow map 'n section 2.3. The full core shutdown margin test'was perfor...ed in accordance with surveillance instruction SI 4,3.A.1. Nith the analytically strongest control rod, 26-07, fully withdra~m, an additional rod was withdrawn tc add '0.58% 'k/k at 106 F.. moderator temperature. This reactiv'ty insertion is equivalan" to 0 52% hk/k at 68 F, compared to the required 0.38% ~k/k.

Subcriticality at this point was sufficient to demonstrate the required shutdown margin. Rod worth data was obtained from General E'ctric as shown below. This data was generated using documented codes. s Rod Individual h'orth Cumulative Vorth

                            .22-03                 0.94%   hk/k                     0>>94% hk/k 0

Temperature coefficient at 68 F.

                                                                -5      o C ~  "1.72 x      10     hk/   F

STARTUP RETEST PROGRA"I SlMQRY REPORT - BFNP UNIT 1 3.0 Results (Continued) 3.2 SRI-4, Full Core .Shutdown Har,in (Continued)

3. 2. 3 Analvgis (Continued)

To demonstrate the required margin, rods 22-03 was withdrawn to notch 18. The reactivity insertion is thus

                 .(at- 106o F)

North 22-03 at 18 = 0.58% hk/k Total Insertion = 0,58% hk/k 0 To calculate the equivalent insertion at 68 F, the temperature coefficient is applied as follows: ht = 106o F '8 F 38o F Vorth withdrawn rods 8 106o F = . 0,58% hk/k c x R=(-1,72 x 10 OF

                                          ) (100%) (38 F)     = .   .06% hk/k reactivity equivalent at 68o F         =    0.52% ~k/k Subcriticality     was verified after rod 26-07, was  fully withdrawn,    and  22-03   at notch 18. The moderator temperature was   106o   F.

The demonstrated margin of 0.52% is clearly larger than the required 0,38% satisfying the test criterion.

~, ~ 5 R ~ STARTUP RFTRST PRIIPI(IQt SlltttiARY Rl'.Pd;llT BFNP UNTT 1 I 4

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4000 5000 6000 7ppp gppp 90pp 10 Average Core Exposure (N%/T) at 6S F. Figure SRI 4>>1 Sl>utdown ",jara'n Cffrve For Unit 1

STARTUP RFTEST PROCRAM SENARY REPOPT - HFNP UNIT 1'.0 Results (Continued) I 3.3 SRI-5 Control Rod Drive S stem 3.3.1 ~Pur osa The purposes of 'the control rod 'drive system test are:

1. To demonstrate that the CRD system operates properly at ambient and rated primary coolant temperature and pressure.

2. To verify that the operating char'acteristics of the CRD system meet the level 1 and level 2 criteria.

3.3.2 Criteria Level 1 Each CRD must have a normal withdraw speed less than-or equal to 3.6 inches per second (9.14 cm/sec.), indicated by a full 12-foot stroke greater than or equal to 40 seconds. The control rod scram insertion times must be wit1>- in the limiting conditions for operation specified in technical specification.

1. The average scram insertion time in the reactor power operation condition shall be no greater than:

Z Inserted From Avg. Scram inser-Fully Vithdrawn tion Times (sec.)

0. 375 20 0.90 50 2.0 90 5.0

2. The average of the scram insertion times for the three fastest operable control rods of all groups of four control rods in a two-bytwo array. shall be no greater than:

                            %  inserted   From         Avg. Scram       Inser-Full '4'ithdrawn           tion Times       (sec.)

5 0.398 20 0.954 50 2.120 90 5.300

s ~ 3 STARTUP RFTEST PROGRAM SttfltARY RFPORT BPNP UNIT 1. 3.0 Results (Continued) 3.3 SRI-5 Control Rod Drive S stem (Continued) 3.3.2 Criteria (Continued) 3s The maximum scram insertion time for 90/ insertion of any operable control rod shall not exceed 7.00 seconds. Level 2 Each CRD must have a normal insert or withdraw of 3.0 + 0.6 inches per second (7.62 + 1.52 cm/sec.), indicated by a full 12-foot stroke in 40 to 60 seconds. With respect to the control rod drive friction tests, if the differential pressure variation exceeds 15 psid (1 kg/cm2) for a continuous drive in, a settling test must be performed, in which case, the;di<Serential settling pressure".. ~ should not be less than 30 psid (2.1 kg/cm2) nor should it vary by more than 10 psid (0.7 kg/cm2) over a full stroke. Scram times with normal accumulator charge should fall within the time limits indicated in figure SRI 5-1. 3.3.3 ~Anal sls SRI 5 testing was conducted at open vessel, heatup, and test condition 1 as defined on the power flow map in section 2.3. The following Control Rod Drive (CRD) system tests "were performed after the completion of fuel loading:

1. A functional test of each CRD to check rod position indication, rod-to-drive coupling, and insert/with-draw speeds.

2. A continuous-in friction test of each CRD to check for drive line friction,

3. A scram time. test of each CRD to verify that scram times are within technical specifi.cation limits.

Rod Timin and Stall Flows The normal rod withdrawal and insert time, together with the stall flows were measured and verified to rive a total stroke time of 40-60 second . Some of the drives were adjusted

                     'o   that their times were within the above criteria.

STARTUP RETEST PROGRAH SE&GKRY REPORT - BFNP UNIT 1'.0 Results (Continued)

3. 3 SRI-5 Control Rod Drive Svstem (Continued) 3.3.3 ~Anal sis,, {ContiAued)

Po"ition-Indicatinr..Check The rod position information system was extensively checked and was operating properly. Cou lin Check This check was performed after fuel'oading whenever a rod was fully withdrawn to.position 48. All rods were coupled to their drives. Friction Testing All of the CRD's were friction. tested by continu-ously inserting them from position 48 to position 0 and photo-graphing the insertion pressure throughout the insert process. The friction test data were acquired using a str<<.in gauge differential pressure cell and a storage oscilloscope. Polaroid photographs of the oscilloscope traces were taken to record the data. All control rods passed the continuous insertion Pmax, - 8'min. criteria.. Scram Testin During open vessel testing all control rods were individually scram tested. The average scram insertion times of all 185 CRD's are as follows: 5% insertion time ~ 0.246 seconds 20% insertion time = 0'.476 seconds 50% insertion time ~ 0.958,seconds 90% insertion time = 1.678 seconds The scram insertion times of the slowest rods in thc core are as follows: CRD 22-23 5% time = 0.302 seconds CRD 14-23 20% time 0.539 seconds

                              -CRD   30-19  20%   time.~ 0.539 seconds

t s STARTUP RET1'.ST PROGRAM SUtfKKRY RFPORT - BF?JP UNIT 1 3.0 Results (Continudd) 3.3 SRI-. 5 Control Rod Drive S stem (Continued)

3. 3. 3 ~Anal sls (Coo..lnuad)

CRD 22-23 50% time ~ 1. 088 seconds. CRD 22-23 90% time ~ l. 880 seconds CRD performance at the open vessel test all level 1 and level 2 criteria. condition'atisfied Pressurized Testine A scram test of each CRD was performed at rated pressure to verify that the scram times were within technical specifications limits. The average scram times for the 185 rods at rated pressure are as follows: 5% insertion time = 0. 298 seconds 20% insertion time = 0. 676 seconds 50% insertion time = 1. 440 seconds 90% insertion time = 2. 533 seconds One individual control rod was not in compliance with the level 2 scram time for 90% insertion. Rod Limit 06-27 90% insertion time = 3.101 sec.. > 3.000 sec. This rod will be retested and, if necessary, the drive will be examined and/or replaced during the first refueling outage. Subsequent examination of the 2 x 2 rocK arrays in which this rod exists indicated the average ~sertion times were within the level 1 criterion. The maximum 90% insertion tim(e was found to be seconds, meeting the 7.00 second requirement of the teclmical speci.f ications. All level 1 criteria were satisfied.

S; '.!Z'P 1.'EST 1 POGPA.'f SL+" fARY REPORT - BF~IP Il.r'TT 1

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STARTUP RETFST PROGRAM SlMfARY REPORT BFNP UNIT 1 3.0 Results (Continued) 3.4 SRI-6 SRM Performance and Control Rod Se uence 3.4.1 ~Put ose The purpose of this test is to demonstrate that the operational sources, SRM instrumentation, and rod withdrawal sequences provide adequate information to achieve criticality and to increase power in a safe and efficient manner. 3.4.2 Criteria Lovel 1 There must be a neutron signal-to-noise ratio of at least 2:1 on the required operable SRH's or fuel loading chambers prior to pulling rods. There must be a minimum count rate-of 3 cps on the required operable SR~i's or fuel loading chambers prior to pulling rods. The IR'f's must be on scale before the SRM's exceed the rod block setpoint. The RSCS shall be operable below 20% power. Level 2 Not applicable 3.4.3 Analvsis SRI-6 testing was conducted at open vessel and heatup test conditions as defined on the power flow map 'in section 2.3. Prior to performing the Shutdown Margin 'Test (SRI-4), the SRM's were verified to read at > 3 counts per second with a signal-to-noise ratio of 2:1. Table SRI 6-1 shows the SRM data. Table SRI.'6-1 SR.'f Count Rates SRM Channel SRM full out (counts/second) .2 .2 SQ! full in (counts/second) 10 28 25 21 Si nal-to-noise ratio 49 279 124 209

STARTUP RF.TF.ST PROGRAH SNQfARY REPORT <<HFNP UNIT 1

3. 0 Results (Continued) 3.4 SRI-6 SRH Performance and Control Rod Se uence (Continued) 3.4.3 ~Anal eie (Continued)

Eleven days later the reactor was taken critical in rod sequence A. Prior to rod withdrawal, the Rod Sequence Control System (RSCS) was proven operable by performing the startup functional surveillance test. The reactor became critical on the thirtieth rod withdrawal (42-23'otch 12; at a moderator temperature of 185 F. After SR~i/IR~f overlap was verified bv SRI-10, the SRH's and IR';f's were removed from the non-coincident serac: mode, and the SR~1 high level blocks set at their normal point of 1 x 105 cps. It was also shown that the SR~."s were capable of monitoring 7.5 x 105 cps without saturating. The RSCS was shown to be operable at 10, 20, and 25% power as evidenced by the inability to select out-of-sequence control rods. The RSCS stopped enforcing at 29 7Z thermal power. All test criteria were satisfied.

STAI<'I'ltl'ETEST PROGRAM SUl'fMARY REPORT - HFNP UNIT 1 3,0 RrsP3ll l ts (Continued) 3s 5 SRI-10 IRM Performance 3.5.1 ~Pan ese s The purpose of this test is to ad)ust the intermediate range monitor system to obtain an optimum overlap with the SP~f and APRM systems. 3.5.2 Criteria Level 1 Each IRM channel must be adjusted so that overlap with the SP"f's and APR.'f's is assured. The IR~f's must produce a scram at 120/125 (96%) of full scale. Level 2 Not applicable 3.5.3 ~Anal sls SRX-10 testing was conducted at open vessel, heatup, and tes t condition 1 as defined on the power flow map in section 2.3. The IL'f gains were initially set to maximum gain. The IRM scram setpoints were checked during preoperational testing and are maintained through plant surveillance test-ing at intervals of three months. The HRf's had been placed in a non-coincidence scram mode prior ta fuel loading. Rods were withdrawn in seauence "A" to bring the reactor critical. All the XR';f's were om scale before the normalized SRM readings reached the opmational limit of 2 x 107 cps. All the IRM's responded tc3 changes in the neutron flux. s After the IRM response and XR&f/SRM overlap were verified, the SRM's and IRM's were taken out of non-coin-cidence scram mode. Table SRX 10-1 comains the overlap data.

                                            -'TARTUP RFTFST PROGIUW SU MARY RFPORT                        - BFNP K!IT 1 3.0 Results     (Continued) 3.5   SRI-10     IfQf Performance       (Continued)

I

3. 5. 3 ~Anni sds (Cnntdnnnd)

Table SRI 10-1 IRM/SRM Overlap Data E s

                       'IR1 f'                                           SRN' Range         Reading                                  Reading 5

s A 2,53x10 cps B 1 5 B 1,87xlO" cps C 1 14 C 1,88xlO< cps D 2 35 D 1.13x105 cps 50 F 2 70 G 1 3.5 H 8 4 The IRM preamplifiers were adjusted for continuity between ranges six and seven during the initial heatup. Low power IRM/APRM overlap was verified as soon as the Apf~!'s came onscale. At 30% power,'he IP."f/APRM calibration was per-formed according to the normal surveillance test program. Proper SR~f overlap with the IRM's was reverified after this calibration. All criteria applicable to this test were net.

STARTUP RETFST PROGRAM SUANRY REPORT - BFNP UNIT 1 ~

~ '

3.0 Results

3. 6 (Continued)

SRI-ll LPRM Calibration 3.6.1 ~Pur ose The purpose of this test is to calibrate the local power range monitor system. 3.6.2 Criteria Level 1 The meter readings of each LPR3f chamber will be proportional to the neutron flux in the narrow-narrow water gap at the height of the chamber. e Level 2 There are no level 2 criteria." 3 ~w3 t'e t'.6.3

                      ~Anal  sis SRI-ll testing   was conducted at test conditions 1 and 4E as     defined on the power flow.map in section 2.3.

TIP sets were run at 19% and 98% thermal power to provide a base distribution in the process computer and the offline computer to allow accurate calculations or" core thermal limits above 25% power as required by technical specifications. The process computer calculations were verified by the offline computer program BUGLE and close agreement was seen for the core limits calculations for SRI-19 and the LPRM GAF's (Gain Adjustment Factor) for SRI-11. The calibration was performed according to the pl,.nt surveillance instruction,'I 4.1.B-3. This involves adjus"ing the meter readings of each LPRM chamber by the appropriate gain adjustment factor (GAF), thereby setting the LPRM to read proportional to the neutron flux in the narrow-narrow water gap at the height of the chamber. I In the cases where the LPKf GAF varied from leO after the calibration, the process computer corrects each LPRM reading using its corresponding CAF. Therefore, the cor~ calculations are still valid and the core monitoring is not affected. ~s All test criteria werc met.

s

                                                                                  ~ I STARTUP RETFST PROGRAM SECTARY REPORT  BFNP UNIT 1 "3.0  Results (Continued) 3.7  SRI-12  APRN  Calibration 3.7.1  Purpose s

The purpose of this test is to calibrate the average power range monitoring system., 3.7.2 Criteria Level 1 The APR~f channels must be calibrated to read equal to or greater than the actual core thermal power. Technical specification and fuel warranty limits on APEf scram and rod block shall not be exceeded. In the startup mode, all APR~f channels must produce a scram at less than or equal .to 15% of rated thermal power. Recalibration of the APR".f system will not be necessary from safety considerations if at least two APR.'1 channels per RPS trip circuit have readings greater than or equal to core power. Level 2 If the above criteria are satisfied, then the APRY. channels will be considered to be reading accurately if they agree with the heat balance to within 7% of rated power. 3.7.3 ~Anal sls SRI-12 testing was conducted during heatup and at test conditions 1 and 4E as defined on the power flow map section 2.3. The APEf's were calibrated using the low power heat balance based on the heatup rate. After the heatup rate had stabilized at 87.6 F per hour, the AP&f's were set to read 3% thermal power. This calibration was used until a more accurate heat balance could be performed at a higher power level. All applicable test criteria were satisfied. In the'startup mode, the APF~~t scram sctpoint was set at < 15% thermal power, and the rod block at < 12%.

STARTUP RETEST PROCRAM SENARY REPORT - BFNP UNIT 3.0 Results (Continued) 3.7 SRI-12 APR"f Calibration (Continued) 3.7.3 ~Anal sis (Continued) At test conditions 3 (19%) and 4E (95%), the APRM's were calibrated to read equal to or greater than core thermal power. The core thermal power was obtained from the process computer heat balan'ce (OD-3), which had been verified to be accurate previously by a detailed manual heat balance. The APEf's were also recalibrated after each LPR.'f calibration. All test criteria were satisfied.

I) s a STARTUP RETEST PROGRAM SUM'lARY REPORT BFNP UNIT 1'.0

    . Results  (Continued) 3.8  SRI-14   Reactor Core    Isolation Coolinp      S       stem 3.8.1  ~Par ese Thee'purpose of this test is to verify the proper-,

operation of the reactor core isolation cooling system. I 3.8.2 Criteria NOTE: Applicable to both types of injections. Level 1 The time from actuating signal to required flow must be less than 30 seconds at any reactor pressure between 150 psig and rated (1020 psig). With pump discharge at any pressure b'etween 150 ps'ig " 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 startup. If either of the first two level 1 criteria is not: met, the reactor will only be 'allot"ed to operate up to a restricted power level defined by figure SRI 14-1. Level 2 The turbine gland seal condenser system shall be capable of preventing steam leakage to the at:mosphere. The AP swit:ch for t: he RCIC steam supply line high flow isolation trip shall be adjusted to actuate at 300K of'he maximum required steady state steam flow. 3.8.3 ~Anal sis SRI-14 test:ing was conducted at heatup and test condition 1 as defined on the power flow map in section 2.3. Flow test:s wcre conducted pumping to the test: line at 150 psig and 1000 psig. The results are summarized in table SRI 14-1. During this testing it was determinctl that the (1) 80 psi will bc u ed in lieu of 100 p" . i, a This measured line loss value of 60 psi plus 20 psi margin.

S TARTUP RETEST PROGRAM SlBÃARY REPORT BFNP UNIT 1 3.0 Results (Continued) 3.8 SRI-14 Reactor Core Isolation Coolin S stem {Continued) 3.8.3 ~Anal eie (Coneinued) AP switches for the RCIC steam supply line high flow isolation trip could not be adjusted to 300% of the maximum required steady state steam flow due to inadequate switch range. The switches are presently set to comply with the technical specifications. TVA Design is currently resolving this problem. Test Time to 600 m Pum Dischar e Pressure ( si ) Condition Measured Reauired Heasured Re uired 150 psip 9 sec. ~(0 sec 240 230 1000 psi 25.8 sec. ~0 sec 1080 M030 1000 si 26 sec. <30 sec 1090 M030 The flow test at 1000 psig was thought to have produced the required 600 gpm flow. However, due to an erroneous initial reading of 20 gpm on the flow meter, it was discovered that actual flow was 580 gpm. After a calibration check of the flow indicating circuit failed to reveal any problems, the test was repeated and proper flow verified by a reading of 630 gpm, thus compensating for the 20 gpm error. The error in the flow indication has since been corrected. The RCIC vessel injection test was performed at 33% reactor power, pressure 954 psig. The system reached rated flow in 27 seconds, delivering 640 gpm. The RCIC turbine did not trip off during any startup. There was no leakage of steam to the atmosphere during the tests at 150 psig or 3.000 psig. Except for the RCIC high steam flow isolation AF switch setpoint (Level II), all test criteria were satisfied.

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STARTUP RETEST PROGRAM

SUMMARY

REPORT - BFNP UNIT 1 3.0 Results (Continued) 3.9 SRI-15 Hi h Pressure Coolant In ection S stem 3.,9'. 1 Purpose The purpose of this test is to verify the proper operation of the high pressure coolant injection system over its required operati'ng pressure" range. 3.9.2 Criteria Level 1 The time for actuating signal to required flow must be less than 25 seconds at any reactor pressure between 150 psig and rated. With pump discharge at any pressure between 150'si'g and 1220 psig, the flow should be at least 5000 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. Level 2 The turbine gland seal condenser system shall be capable of preventing steam leakage to the atmosphere. The AP switch for the HPCI steam supply line high flow isolation trip shall be adjusted to actuate at 225% of the maximum required steady-state steam flow. 3.9.3 ~Anal sin SRI-15 testing was conducted at heatup and test condi-tion 2E as defined on the power flow map in section 2.3. The HPCI performed its required functions on the vessel injection test with no observed abnormalities. Maximum observed rpm indications were slightly high due to an erroneously high initial value. Transient traces did not substantiate an abnormal rpm. The indicator in question has since been repaired. Tran-sient traces of the test and observed parameters indicate smooth operation after the initial transient. Observation of the HPCI during and after the test run indicated that the gland seal condenser system was capable of removing gland leak-off, thereby preventing steam leakage to the (1) 60 psi will be used in lieu of 100 psi. This is a. measured line loss value of 40 psi plus 20 psi margin.

STARTUP RETEST PROGRAM SEMARY REPORT BFNP UNIT 1 e-3.0 Results (Continued) 3.9 SRI-15 Hi h Pressure'oolant In ection S stem ,(Continued) 3.9.3 ~Anal sis '(Cenninned) atmosphere. The HPCI turbine did isolate during the 1000 psig test run due to a high pressure between the rupture disks (RO 259/7615). 'his problem was corrected and the test successfully repeated. Maximum observed rpm indica-tions were well below the turbine trip. The steam supply line high flow isolation lines were within calibration and set in accordance with STI-15, which was performed during initial startup testing of the unit 1 HPCI. Based on the data taken and observation of the system, it can be concluded that the HPCI is fully opera-tional and capable of fulfilling its intended function. Flow tests were conducted pumping to the conden-sate storage test line at 150 psig and 1000 psig. A f3ow meter offset discovered after the 150 psig test caused a false indication of flow and thus caused the 150 psig test results to be inconclusive. The difficulty was subsequently corrected and the test satisfactorily repeated. The results are summarized in table. SRX 15-1. Table SRZ 15-1 HPCI Flow Test Test Time to 5000 vnm Pumo Discharre Pressure Condition Measured Required Measured Required

'<150  psig               14.2 sec.              25 sec.       360  psig         17  psig 150  psig                  21 sec.             25 sec.       300  psig        210  psi g 1000  psig                 '17 sec.             25 sec.      1050  psig      1020  psig
            *Actual flow    was 4100 gpm due      to meter error.

This data indicates the flow and pump discharge pres-sure criteria were satisfied. The HPCI vessel injection test was performed at 52% power, reactor prcssure, ~55 psig. The pertinent test data is listed in table SRI 15-2.

STARTUP RETEST PROGRAM Sb~ÃARY REPORT BFNP UNIT '-0 3.0 Results (Continued) 3s9 SRI-15 $ 1i h Pressure Coolant In ection S stem (Continued) i ~ 3.9.3 ~Anal sis (Continued) Table SRI 15-2 Vessel Injection Test Required Pump Measured Time to Pump 'Discharge Flow/ Time to Rated Flow Rated Discharge Pressure Required Flow Pressure Required Flow 24'.5 sec. 1015 5200 gpm/ 25 sec. 1020 psig psig 5000 gpm All test criteria were satisfied.

s3 I"I'AR'I'1l'l'.TEST PROGRAN SUIIRIARY RE1'ORT I31'NP UNIT 3.0 iiasults (Continued) 3,10 SRZ-lp Core. performance 3.10.1 ~i'ur >use The purpose of this test is to evaluate the core performance parameters of core flow rate, core thermal power level, the core minimum critical power ratio (HCPR), the maximum average planar linear heat generation rate (HAPL)IGR), and the maximum linear heat generation rate (LIIGR) of any

                        > od in. any, fuel assembly.

3. l0. " ('r I ter ia Level 1 The maximum linear heat generation rate (LI!GR) o:

any rod during steady-state conditions shall not exceed the limit specified by the Technical Specifications. s Steady-state reactor power shall be limited to 3293 hsHt and values on or below the design flow control line (defined as 3440 'iVt with core flow of at least 102.5 x 106 lb/hr.) exceed the The minimum critical limits specified poser ratio (".tCPR) shall not by the technical specifications.

The maximum average planar linear heat generation rate (K'.PLI!GR) shall not exceed the limits oi the Plant Technical Specifications. Level 2 There are no level 2 criteria. 3.10. 3 Analvsis SRI-19 testing was conducted at test conditions 1 and 4E as defined on the power flow map in section 2.3. Core performance parameters were calculated by the offline and process computer. programs at 19Z and 96K tliermal po3:er. Proper operation of the process computer was verified by comparing the result with the offline computer. The data for. both test- conciitions are liste54 0.003 .54 Q jj 1.0 0.311 03109 0. 0001 .03 T. C. Core thea.wl nov~>> 3293 Mft sS 45. S4 '!'<t 3141. 07 Mi t 4.27 Nit. 0.14% 4 ki~/ft 16.24 k4/ft 16.17 kV/rt 0,07 kh/f t 0.44% h<( PRRAT 1.0 0. 970 0.967 0.003 0.31% y yP:q<7. 1.0 0.855 0.853 n,002 0,23%

*.":CPR, L'!GR, and iQPL!!GR are calculated by G!'. program BUCLE.
Core ther...al po"er calculated by TVA propan CORPR.
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STARTlJP RETEST PRQGfUQf SPi'MARY REPORT - HFNP WiIT 1 3.0 Resul ts (Continued) 3.11 SRX-22 Pressure Re ulator 3.1>.l ~Pur os@ The purposes of this test are:
1. To, determine the optimum settings for the pressure control loop by analysis of the transients induced in the reactor pressure c'ontrol system by means of the pressure. regulators.

2. To demonstrate the take-over capability of the back-up pressure regulator upon failure of the controlling pressure regulator and to set spacing between the set points at an appropriate value.
30 To demonstrate smooth pressure control transition between control valves and bypass valves when reactor steam generation exceeds steam used by the turbine. 3.11.2 Criteria Level 1 The decay ratio must be less than 1.0 for each process variable that exhibits 'oscillatory response to pres-sure regulator changes. Level 2 Xn all tests except the simulated failure of the operating pressure regulator, the decay ratio is expected to 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 Haster Flow Controller. Pressure control deadband, delay, etc., shall be small enough that steady-state limit cycles, if any, shall produce variations in steam flow to the turbine no larger than the values of rated flow specified in the following table. Percent of full ower Percent of rated flow 90-100 + 0.5 70-90 + 1.5 to + 0.5 70 and below + 1.5
,STARTUP RI'.TFST PROGRAH SUPIARY RI'.POINT HFNP U',lIT 1 3.0 Results (Centlnued) 3.11 SRI-22 Pressure Rer,ulator (Continued) 3.11.2 Criteria 'Continued)
Opt:inurn gain .values for the pressure cont:rol loop shall be determined in order to give the fastest retrrrn from the transient condition to the steady-stat;e condition within the limits of the above criteria. During the sir,".ulated failure of the prir..ary control-ling pressure re! ulator, if the set point of the bar!:-up:>."r .. re regulator is optimally set between 1-4 psi differential, the back-up regulator shall control the transi.ent suc!. e!rat the peak neut:ron flux and peak vessel pressure remain belo~ the scram settings by 7.5/ and 10 psi respectively. Following + 1-10 psi (0.7 kg/cm ) pressure set-
.point change, the time between. the setpoint change and the occurrence of the pressure peak shall be 10 seconds or less.
3.11-.3 Analysis SRZ-22 was conducted at test condit ons ', 2F., 3!:, and 4E as defined on the flow map in section 2.3. At: test: condition 1, step changes of + 10 p.,i were made on the prir.ary and backup pressure regulatorrr usinr; >he built-in test feature in the r.',I!C circu'try. r nsien.s wer'e recorded for the pressure step witit both the control a.;~'. by <<a ... valves functional and for the case with the bypass valves alone regulating the change. A smooth transition between the control valves and bypass valves was noted and neglig'ble oscilla"..i' of the recorded parameters was observed. Tire backup regu'ator takeover test failed durin~ t.lr simulated prir,:ary regulator failure te t at 5 psi dif fe: ent!,t3. This t;cot was successful for a 2 Psi rli feren"-s"
~
The gain values for tl>e pressure control loops were adjusted, and 'SRT-22 was 'repeated at T.C. 21'.. The + 10 Psi step changes were accomplished'uccessfully, and hackrrp takeover capability r as demon trated at a 1.6 psi. <Iifterent:i;rl. I.evel scram with a 1-4 psi spread between t: he regulators. Sirrilar results were obser ved at 'E.C. 3F,.
32>> STARTUP RETEST PROGRAM SIQfARY REPORT - BFNP UNXT 1 3.0 Results (Continued) 3.11 SRI-22 Pressure Re ulator (Continued) 3.11.3 ~Anal sis (Ceneinned) J The final system checkout was performed at test condition 4E. Positive and negative step changes of appro-ximately 5 psi were made on both the primary and secondary regulators. Again the tests were performed with the control-valves and bypass valves, and with the bypass valves alone negotiating the transient.. The decay ratio for the process variables were less than 1.0 as required by level 1 criteria. Steam rate flow variations, however, were not less than + 0.5%. The gain values for each pressure control loop were considered satisfactory. Pressure peaks after the step changes occurred before the acceptance criteria limit of 10 seconds. All observed process parameters had decay ratios less than .25. Table SRI 1 'summarizes the step change data The backup regulator takeover capability was demonstrated with a 1.5 psi differential. This setting provided the scram margin required by the level 2 criteria; yet prevented the backup regulator from "operating .during normal pressure transients. Except for steam flow variations, (Level IX), all applicable test criteria were satisfied. Additional testing. is planned to" resolve this problem.
Table SRI 22-1 Pressure Regulatory Response Summary (Recirculation in Ifastcr Manual Hode) Test Condition 2E ~ 3E Step Input (psi) -10 -10 +10 -10 -10 +10 -10 +10 -10 -10 +10 .-5
'
Rerulator A/B B BPV BPV BPV BPV BPV BPV BPV BFV SPV 'alves (CV/BPV) C.V. nc nt. C.V. 50Z C.V. 50Z C.V. Inc >nt. C.V. 50K C.V. 502 C.V. 502 Irr."nt <np First(Press; Initial Bone 964 962 950 960 980 973 980 974 968 960 967 960 972 .972 977 5 972 949 947 963 944 973 981 973 980 962 967 961 967 977 977 972.5 Final Done Press ~ Ti:e to 3.5 3.0 2.1 2.1 8.5 7.5 Fre<s Pea/ ) 2.5 3.1 3.1 4.0 5 Ri."best Decay Ratio <.25 <.25 < ~ 25 <.25 <.25 <.25 <.25 <.25 25 <.25 <+25 <.25 <.25 < ~ 25 <.25 Para-. eter 2 4 4 3 4 4 4 4 4 4 3 4 4 5 (5 6) (1) Level 2 criteria linit is 10 seconds. (2) Level 2 criteria is 0.25. (3) Later level. (4) APRN (5) Turbine inlet pressure. (6) Rx narra range done pressure. 1
STARTUP RETEST PROGRAM SlOL"1ARY REPORT - BFNP lJHEV 1 3~0 Resul ts (Cont] nued) I 3. 12.1 ~Pur a.",a 4 The purposes of this test are:
1. To adjust the feedwatcr control system for acceptable reactor water level control.

2. To demonstrate stable reactor response to sub-cooling changes.
a
3. To demonstrate the capability of the'utomatic core flow runback feature to prevent low water level scram following the trip of one feedwater pump.
3.12.2 Criteria Level 1 The decay ratio must be less than 1.0 for each process variable that exhibits oscillatory response to feedwater system changes. Level 2 The decay ratio is expected to be 1'ess than or equal to 0.25 for each process variable that exhibits oscillatory response to feedwater system changes when the plant is operating 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 time from set-point step change until the water level peak occurs shall be less than 35 seconds without excessive feedwater swings (changes 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 one of the operating feedwater pumps. Pith the condensate syst: em operating normally, the maximum turbine speed limit shall prevent pump damage due to cavitation. 3 ~ 12.3 ~Anal s1s The capability of thc feedwater system to atis-factorily control water level was demonstrated at test
STARTUP RFTEST PROGRAM SU. NARY REPORT HF'ip lf'NIT1 3.0
~ Results (Continued) 3:12 SRT-23 2 Feedvater System (Continued) 3.12.3 Analvsis (Continued) conditions 1, 2F., 3E, and 4E as defined on the flow map in section 2.3. Experience from the initial startup testinp provided preliminary controller setpoint, minimizing sub-sequent adjustments.
The proportional band was initially set at 390",'., resetting at 0.4 repeats per minute. A set of ~ 3-inch level setpoint ch,.nres vere performed at test condition 1 for a single feedwater pump in both single-element and three-element mode. The system responded smoothly and the water level peaked at about 30 seconds after the level setpoint change. The level setpoint changes ~ere also performed at test conditions 2E and 3E. Final sy tern checI;out was completed a" test condition 4E (97'; power). Table SRT 23-1 su;.=..arizes the system response time to + 3-inch setpoint changes. Table SRI 23-1. Feedwater System Response Controlling Control Setpoint Time to Peak Level Node Change LAater Level 3 A Flement -3'3 sec. 3 II sec Element 1 Flement sec.
-3'3I 1
I Element sec. Transient traces shoved no noticeable oscillations of the system, and tI!e decay ratio of the measured proce..s variables less than .25.
-3G- I s SThRTUP RETEST PROGKQ1 SU'INARY REPORT - HFNP UNIT 1
3. 0 Results

3. 12 (Continued)
SRI-23 s Feedwater S stem (Continued) a s 3.12.3 ~Anal sis (Cnnnlnnad) 0 The automati'c recirc-flow runback feature prevented a low-level scram after a feedwater pump was tripped. The transient recording indicated that the vessel water level reached a minimum at 21 inches, well above the scram setpoint of 11 inches. Pith the condensate system operating normally, the feedwater pumps were able to provide full power flow with I no evidence of cavitation, I In summary, the feedwater system responded smoothly and predictably to induced transients. All test criteria were fully satisfied. 4
STARTUP RFTl'.ST PROGRAH SPFGKRY RI'.PORT 11FNP UNIT < 3 0 Re""ules (ConeSnued) 3.13 SRI-?5 Hain Steam I .oJation Valves 3.13.1 ~Pur ose The purpose. of this test are:
1. To functionally check the Hain Steam I.ine Isolation Valves (HSIV's) for proper operation at rated temp-erature and pressure."'.
To determine isolation valve closure ti...e.
3. 13. 2 Criteria Level 1 Closure time must be greater than 3 and less than 5 seconds.
Level 2
,There are no level 2 criteria.
3. 13. 3 Analysis SRI-25 was conducted during heatup as defined on the flow map in section 2.3.
The main steam isolation valves were checked "or proper operation at rated temperature and pressure. The closure time was measured in accordance with the normal surveillance instruction, and the results are she'-n 'n table SRI 25-1. The routine 3.0 closure surveillance was also conducted. 41o unusual response was noted, and all test criteria were satisfied. Table SRI 25-1 HSIV Valve Closure Tine" Valve T lme* FCV 1-14 3.6 FCV 1-26 4 ..7 FCV 1-37 3.8 FCV 1-51 3.3 FCV 1-15 3.5 I'CV 1-27 3.5 FCV 1-38 3.2 *Fcq ui r ed FCV 1-52 3.1 3 se c < t l me < 5 s ec .
-38<<
~ ~ STARTUP RETEST PROC14N SUKQRY 'REI ORT '-'"'BW4Ps-UN'IT 3.0 Results (Continued) 3.14 SRX-26 Relief Valves
~ ~
3.14.1 ~Pun ese The purposes of this test axe:
1. To verify the proper operation of the primary system relief valves.

2. To vexify the proper seating of the relief valves following operation.
3.14.2 Criteria Level 1 Hach valve will open and allow steam passage when manually actuated. Level 2 h Relief valve leakage must be low enough so that the temperature measured by the thermocouples in0 tIie discharge side of the valves returns to within 10 F of the temperature recorded before the valve was opened. 3.14.3 ~Anal sis SRI-26 was conducted during heatup as defined by the flow map in section 2.3. All the main steam relief .valves were manually actuated from the control room, and each ADS valve was operated from the backup control panel. Reactor pressure was 256 psig. Hach relief valve gave a positive indication of steam passage as evidenced by an abrupt increase in the relief valve tailpipe temperature. Six of the eleven relief valve tailpipe temperatures did not return to within 10 F of the original value, as required by the level 2 criteria, These temperatures were monitored as the reactor was further pressurized. The tailpipe temperatures showed no abrupt incxeases and the final equilibrium values at rated pressure were typical of normal operating range, indicating that the relief valves reseated properly,
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STARTUP RETEST PROC}AH SEQ'IARY. Rl'.PORT - DFNP UN1T 1 3.0 Results (Continued) 3-.16'ISI-32 Recirculation M-G Set S eed Control 3.16.1 ~Pur ose The purposes of this test. are:
1. To determine correct gain for op timum perfor-mance .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. 16. 2 Criteria Level 1
,The decay ratio must be less than 1.0 for each process variable that exhibits oscillatory response to flow control changes. 'I 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 over the entire range from 20% to 100% speed.
Steady state limit cycles, if any exist, must. not cause turbine steam .flow to vary in excess of + .0.5% rated flow as measured by the gross generator electrical power output. 3.16.3 Anal~sis SRI-32 was performed at test conditions 2D, 2E, 3E, and 4E as defined on the flow map in section 2.3. The test was performed to determine the corr'ect gains for stable performance of the recirculation loops. A series of + 10% speed changes in the recirculation H-G set speeds was performed. at specified core flows along the 50% load line (test conditions 2D and 2E) and a gain curve was generated; H-G set response was adequate, however, the gain curves showed a displacement from the desired curve. This condition did not affect the stability of the sy. tern, and the -test program was continued pending maintenance at the first opportunity on the positioner cams.
STARTUP RETEST PROGRAM
SUMMARY
REPORT BFNP UNIT 1 3.0 Results (Continued) 3.16 SRI-32 Recirculation M-G Set S eed Control (Continued) 3.16.3 Anelvel., (Continued). SRI-32 was repeated at test condition 3E. The controller's were initially set at 750% with a reset of 10 repeats-per minute. These settings provided a slow steady pump response,on both pumps with no noticeable oscillations. The final system checkout was performed at test condition 4E, nominal full load, following replacement of the Bailey positioner cams. The loop controllex settings were adjusted to 500% with a reset of.10 repeats per minute. The transient response to + 10% speed changes was then recorded. There was no .noticeable oscillation of the observed process variables." Similarly, during ste dy state oper tion, the steam flow to the turbine did not vary more than + 0.5%, as required by the level 2 cx'iteria.
"
The gain curve was plotted for several flows alone the test condition 4 load line. The gain curves are very nearly linear as shown on figure SRI-32-1. The slight offset between the t~~o curves is within acceptable limits. The mechanical and elect ical stops were set at 10';,"" flow on the 100% load line. All test critexia were satisf'iud.
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STARTUP RFTEST PROC'IbQ! SU.1."lARY REPORT HFNP UNIT 1 3.17 SRI-35 Recirculation S stem Flow Calibration 3.17.1 ~Par osn The purpose of this, test is to perforn a complete calibration of the installed recirculation system flow instru-mentation. 3.17.2 Criteria Level J There are no level 1 criteria. Level 2 Jet punp flow instrumentation shall be adjusted such that the jet puno total flow recorder will provide a correct core flow indication at rated conditions. The APR';l/PB~l flow-bias instrumentation shall be adjusted to function properly at rated conditions. 3.17.3, Analvsis SRI-35 was performed at test condit on '~E as defined on the flow nap n section 2.3. three sets of core flow data were t: en at =.",'. power. Based on thi.s data the gains of the jet pump loop and total core flow proportional amplifiers were adjusted to give correct control room indications of total core flc>>r and jet pump loop A and loop B flows. Comparison of the total core flow recorder and tne process computer core flew data point showed agreement within 1.0%, well within the accuracy of the plant instrumentation. These data" sets were also used to confirm the span range of the recirculation flow nozzle transmitter.'ased on analyses of this data, the flow nozzle transmitters should be respanned to 25.5 psig.for the A loop. A set-point change request has been completed and approved. 'i'hese changes were implemented and the APR~!/RBil flow-bias inst ru-mentation adjusted to function as designed at rated condi.- t" ons. lt
>>44-STARTlJP RETEST PROGRAH SUiLMRY REPORT - BFNP WiTT 1 3.0 Results (Continued)
3. 18 SRI-/2 Drvwell Atmos here Coolin System 3.'18. 1 Purpose The purpose of this test is to verify the ability of the drywell atmosphere cooling system to I,".aintain design conditions in the drywell during operating conditions.
3.18. 2~ Criteria Level 1 Not applicable. Level 2 (A) The heat removal capability of the drywell coolers shall be approximately 5.19 x 10~ Btu/hr, (B) The drywell cooling system shall. have a standby capability of > 25% of the heat removal capability shown in (A). This criteria will be satisfied if it can be shown that criteria (A) and (C) can be satisfied with eight fans and coils in operation. (C) The drywell cooling system shall maintain temp-eratures in, the drywell below the listed design values during normal operation. Design temperatures are:
1. During normal reactor operation:
0 150 - F maximum average throughout drywell 50 percent maximum relative humidity 0 135 F maximum around the recirculating pump motors 0 180 F maximum for all other areas 0 200 F maximum above the bulkhead
2. Uniform circumferential temperature at which the refueling bellows/bulkhead assembly must be maintained:
'>H.thin 25 0 F maximum point-to-point variation
STARTUP RETEST PROGRAM SURPIARY REPORT HFNP UNIT 1 3.0 Results (Continued) 4 3.18, SRT-72 Dr ~el% Aenosr3hc re Coolin,"..S "..tern (Continued) a 3.18. 3 ~Anal sin a s Temperature data was collected during the heatup eo rated pressure, .and at full load (TC 4E) eo determine if the drywell aemospf>eric cooling system was adequately perforr,".ing its intended function. Table SRI 72-1 summarizes the test
'esults at 95%,power.
Table SPT 72-1 Drvwell Ileat Load Tes t Paraneter Results Reauired Drywell cooler heat load 6 u (eight coolers) 6.15 x 10 Btu/hr + 5.19 x 10 6 ".~~x. avg. drywell temp. '138. 7 F < 150 F P~x. relative humidity 11% 50% Max. te. p. around recirc. pump noeors 124 F < 135 F .lax. temp. in all other areas 177 . F < 180 F Max. temp. above the bulkhead 216 F < 200 F K~x. cooling water temp. 75.9 < 100 F Hax. point-to-point ter..p. variation on circunference of bellows/bull;head assenblt =18 s F 25 F
,, ~
T.evel 2 criteri.a (A) was not met si,nce the drywell coolers were operating with a heat load 18% lriglrer than design capabilitys
STARTUP RETEST PROGRAM
SUMMARY
REPORT - BFNP UNIT 1 3.0 Results (Continued) 3.18 SRI-72 Dr well Atmosphere Coolin S stem (Continued) 3.l8.3 ~Anal ebs (Continued) Criteria (C-1) was met with two exceptions. The only temperature higher than criteria value was from TE 80-27 above the bulkhead. However, TE 80-14-, vessel to shield air, showed marginal values which would certainly exceed the limit with. warmer RCW temperatures. It should noted that these two exceptions appeared at both test conditions. In addition, only one channel of the humidity sensor 5R-80 was operable and it was experiencing intermit-tent failures, thus casting doubt'on the validity of the reading.'lans have been made to.replace the cables to these sensors.
=
Level 2 criteria (B) states the 25% standby capa-bility must be shown by meeting both criteria (A) and (C) using only'ight of the ten fans and coils available. Level 2 criteria (A) could not be met using only 8 fans and coils due to high drywell heat load, and since the temperature requirements of criteria (C)'ere not totally satisfied, cannot yet be said that criteria (B) has been satisfied. it The deviations from the predicted design conditions discovered in the test results were tentatively approved. The Division of Engineering Design is presently evaluating the test data in detail.
STARTUP RFTEST PROGR3Q1 SUK~tARY RF1'ORT BFVP UNIT 1 3.0 Results (Continued) 3.19 SRI-73 Coolin 4'ater S stems 3.19. 1 Purpose The purpose of this test is to verify that t1re performances of" the Reactor Building Closed Cooling 4'ater (RBCCN) and P~t Cooling 't~'ater (RCh) syster.;s are adequate with the reactor at xated conditions. 3.19.2 Criteria Level 1 Not applicable. Level 2 Verif cation that the system pexformance meets the cooling requirements constitutes satisfactory completion of this test. The RBCP was designed to transfer a maximum heat "load of'31.3 x 10 Btu/hr. in order to limit equipment inlet wa"er temperature to 100 .F, 'assuming a service (raw coolin~) water inlet temperature of 90 F. 3.19.3 Analvsis SRI-73 was conducted at hot standby nd at test condition 4E, as defined on the power flow map in section 2.3. Since the heat load is greatest at ful1 load, the test results at condition 4E are discussed below. Table SRI 73-1 Test Tlatc Parameter >fax. /Desir,n Value HX. A RBCQC flow 3369.5 gpm 2632 Cpm RBCCld inlet temp. 118.5 F 89.7 F S9.7 RBCQC outlet temp. 100 - F 78 F 76 F RCW flow 2550 gpm RQ4 inlet temp. 90 F 39 F 39 F RQ4 outlet temp. 102. 3 F 84 84.4 F 6 Heat removal rate 31.3 x 106 Btu/1rr 16.62 x 10 Bt u/hr
~Flow was t:oo low to bu me:rrrrrred:recur.at:ely.
s s
' STARTUP RETEST PROGRAll SUHllARY REPORT - EPEE UNIT 1 3.0 Results (Continued) 3.19 SRI-73 Cooiin listen S stems (Continued) 3.19.3 ~Anal sis (Continued)
The calculated heat on the RBCCW side of the heat exchangers was 16.62 x 10 broad Btu/hr. Of this total,
-
approximately ?.05 x 106 Btu/hr was heat load from the dry-well coolers and recirculation pumps (as calculated from SRI-?2 data which was taken simultaneously with the data for this test). The fuel pool heat exchangers, which were not in service at the time of this test, are rated for a maximum heat load of 8.8 x .10 Btu/hr. Adding this to the previously determined 16.62 x 10 6 Btu/hr yields a total heat load of 25.42 x '1%) Btu/hr. This compares to a design maximum of 31.3 x 10 Btu/hr with 90 F RCW inlet'water. This heat load is well within design limits. However, the drywell portion of the heat load is rather high. Although the test criteria have been met (e~;cept the drywell heat load difficulties) for the conditions at the time of testing, it is difficult to compare these results to a criteria based on much warmer RCW inlet water to the heat exchangers. Also, the low RCW flows create dif-ficulty in analyzing the RCW and RBCCW flow balance. A rough evaluation of system performance assuming maximum flows and maximum temperatures for the RCW inlet and RBCCW outlet shows the system performance would be marginal. Thorough testing will be repeated when RCM flows and temperatures are higher in order to better evaluate performance at the most limiting conditions.
~ ~ STARTUP RETEST PROGRAM SU&aM<Y RFPORT - BFNP lPilIT 3.0 Results (Continued) 3.20 STI-74 Modified Off-Gas S stem 3.20.1. ~Pur osc The purposes of this test are:
\ f
1. To verify the proper operation of the Off-Gas system over its expected operating parameters.

2. To determine the performance of the activ..ted carbon adsorbers.
3.20.2 Criteria Level l The release of radioactive gaseous particulate effluents must not exceed the limits specified in BFNP technical specifications 3.8.3. There shall be no loss of flow oi dilution ste m to the noncondensing stages when the steam jet air ejectors are pumping. Level 2 The system flow, pressure, temperature, and relati:e humidity shall comply with the design specifications shown in table 74-6. The catalytic recombiner, the hydrogen analyzer the activated carbon beds, and the filters shall be working as designed. 3.20.3 Analvsis
-STI-74 testing was performed at test conditions l, 2E; 3E, and 4E as defined on the power flow map in section 2.3. This test was not done during the original startup test program. Extensive off-gas treatment equipment has been added since the initial plant startup.
Airborne Effluent Releases Airborne releases "during testing were documented in surveillance te: ts SI 4.8.B.l-a and,SI 4.8.8.2-6. There were no violations of the HFNP technical specifications 3.8.I( limits at any test condition; therefore, level l criteria were fully sati"fied.
STARTUP RETEST PROGRAM SU"QMRY REPORT BFNP UNIT 1 3.0 Results (Continued) ~ o 3.20.3 ~Anal sis (Continued) Dilution Steam Flow There were no losses of dilution steam flow to the noncondensing stages of the pumping SJAE during any testing. The total dilution steam flows are recorded in table STI 74-1. Level 1 criteria were satisfied. 1 S stem Parameters Table STX 74-1 summarizes system operating parameters.
-5 1-Table STI 74-1 g Power 15-40 40-60 65-85 95-100 Data 10/4/76. 10/25/76 11/12/76 12/16/76 System, >>Mt 1127 1969 2542 3127 Parameters orma Operating TC1 TC2E TC3E TC4E !
Ran,e'9100 DIL Steam Flow (Total) ~r 9500 .9975 10200 10250 SJAE Outlet Pressure 5-10 psig 5.5 7.0 OG Preheatcr T Outlet 275-360 F 350 340 340 340 Active Recomb. Temp. Bottom 275-875 F 460 610'05 Hiddle 275-875 F 457 555 640 Top 275-875 F 460 610 640 Standby Recomb. Temp. Bottom 275-360 F 372 440 480 2/0
-"-Middle . -27.5-360 F 370 445 485 285 Top 275-360 F 360 445 (1) 490 (1) 285
) OG Cond. Coolant Out <120 F 108 (g2
,
OG Cord. Outlet Temp. <140 F 138 130 131 123 1 "2 H Concentration 0-1% 0.1 .03 0 (2) OG Flow 20-40 sc Em 50(3) 40 40 25 20-40 psig 32 38 32 I Glycol Tank T 33-38 F 36 38 3!> I titoist Sep. T Out <55 F 50 55 55 55 ! Reheater Dewpoirt Upscale(4) 'pscale( ) > ~eb(."ter T Out 72-/6 F 74 74 . 74 ) Piefilter D.P. 0-2" water .2 .2 0
',sorbcr 1).P. ~ 5 2-6 ps1 .5 .45 (5)
Lypass D~P ~ 0 2 water 0 0 0 0 Adsorber Vesstl T Bed A Pt. 1 68-79 F '72 73 Ped A Pt. 2 68-79 73 73
!><:d A Pt. 3 1'8-79 F 72 71 72 /3 lied !3 Pt. 68-79 F 72 .72 72 iiv<1 C Pt. 5 68-79 F 72 $ 72 t li<<'I D 1' . 7 6H-79 F 70 72 / ~<
1't,. 68-7'9 74'3 1!e<! 6 F 70 "7'J 72 /7~> A<is<) be< < Paul t T 73-S1 75 ' 7 c) A << r~ 1'I 1 t(r D.P. 1'-2" watel .2 .4 0. '> ('>) t'call Q 18.5
!l<'1 . 11<<m. lips<:a] e 32
STARTVP RETEST PROGRAM SU%VARY REPORT BFNP VNXT 1 3.0 Results (Continued) 3.20 STX-74 Modified 'Off-Gas S stem (Continued) 3.20.3 he~el sos (Cone2nned) Table STI-74-1 Footnotes
1) Leaking standby recombiner outlet. valve allowed
~
some recombination to occur.
2) Hydrogen analyzers were inoperable.
l
3) Excessive inleakage caused high o'ff-gas flow

4) Dewpoint sensor ME-66-110 was inoperable.
' stem Parameters Temperatures, pressures, flow and relative humidity values complied with 'the design specifications except:
1) Hydrogen analyzers were inoperable at test condition 4E.

2) A faulty dewpoint sensor hindered the determination of relative humidity at test conditions 3. and 2E. The replacement sensor, installed prior to test. condition 3E, restored operability.
e
'
dro en Anal zers Table STI 74-2 summarizes hydrogen analyzer per-formance data taken during startup.
~
5 3>>
~ J STARTUP RETFST PROGF4QI SENARY REPORT - BFNP UNIT 1 3.0 Results (Continued) 3.20 STI-74; 'Hodified Off'-"Gas S stem (Continued) 3.20.3 ~Ansi sis (Continued)
H dro en Anal zers ul Table STI 74-2 Hydrogen Analyzers:, g Power 15-40 40-60 65-85 95-100 Date 10/4/76 10/25/76 ll/12/76 12/16/76 Hydrogen Analyzer Performance MVt 1127 1969 .2542 3127 Normal Operating Tcl Ran,e B A Process Reading % H2 0-1 01; 01 .05 03 0 '0 Ino .Ino Sample Flow scfh 4.0 3.8 2.5 3.5 2-3 2-4 e"in. Vater flow gph 1.6. '1.5 1.5 1.5 2.0
'.acuum regulator water 10-25 20 17 15 15 15 20
alibration Standard scfh 3-4' 4 2,5 c2.0 C libration Standard % H2 1.0 1,0 1 1.0 '.0
! Calibration Gas Results % H 1.0 1,0 1.0 .03 ,8 H~ Free Standard scfh 4 c2 e2 0 3.5 3.4 I H Free Standard % H 0. 0 0 '"0 0 0 0 i H~ Free Results % H2 0 .05 .05 .05 .03 0 The process hydrogen analyzers were not reliable for ha a continuous service. This was attributed to condensed moisture 3 .
causing erratic sample flow and improper sensor response. "Ia a t e Both hydrogen analyzers failed to perform satis-factorily. Therefore, Leveg II criteria was not satisfied. The hydrogen analyzers have since been modified and are now operating.
STARTUP RETEST PROGRAH SUl'DSRY RFPORT BFNP UNIT 1 a 3.0 Results (Continued)
3. 20 STI-74, Modified Off-Gas S stem (Continued)
~~ 3.20.3 ~Anal "da (Cantlnued) Catal tic Recombiner catalytic recombiner l TaMe STI 74-3 summarizes performance during startup. Table STI 74-3 Recombiner Performance
% Power 15-40 (1) 40-60 65-85 ; 95-100 Recombiner '10/4/76 Performance Date . 10/25/76 13,/12/76 . 12/16/76 1127 '
1969 2542 3127 lÃt TC-1 TC-2E TC-3E TC-4E ludiolytic Gas Production Rate, cfm/Mdt 0. 036 0.036 0. 0. 038 040'10 0 460 560 640 Recombiner Tem
'ctive F
0 350 340 340 340 OQ Preheater Tem Outlet F hT Actual, F 110 220 270 300 hT Ex ected, F 120 210 286 286 (1) 1fethod used for this calculation was approved by revision dated ll/2/76. The catalytic recombiners performed satisfactorily during startup. Level II criteria was satisfied.
e* STARTUP RETEST PROGRAM SUsQMRY REPORT BFNP UNIT 1 3.0 Results (Continued) 3.20 STE-74, Hodified Off-Gas S stem (Continued)
3. 20.3 ~Anal sts (Continued)
Adsorber Beds Table STI 74-'4 summarizes the calculated residence times for four radionuclides and the XE/KR ratios across the six charcoal adsorber beds operated in series. Actual and expected delay time ratios (XE/KR) were in close agreement at all test conditions.
~ 'able ST1 74-4 Charcoal Adsorber Performance Charcoal Adsorber % Power 15-40 40-60 65-85 95-100 Performance Date 9/29/76 10/25/76 11/12/76 12/16/76 (Residence Time) li27- 1969 2542 3127 Mft TC-1 TC-2E" TC-3E TC-4E Kr88 (actual), Hr 37 15.8 16.9 Kr85m (actual) Hr 80 15.4 22.9 21 Kr (expected), Hr 9.7 9.7 9.7 19 Xe 135 (actual),day 13.9 14.3 15 Xe 133 (actual),day 2256 48.6 36.2 32 Xe (expected),day 7.3 7.3 7.3 15 Ratio Xe/Kr (actual) 16:1 21:1 17.3:1 17:1 Ratio Xe/Kr (expected) 18:1 18:1 18:1 18:1
~ ~
STARTUP Rl.'TEST PROGRA'll SUÃfARY REPORT BptlP UMTT l 3.0 Results (Continued)
3. 20 STI-74 Hodified Off-Gas S seen (Continued)

3. 20.3 ~Anal sis (Continued)
System HEPA Filters Table STI 74-5 summarizes the results of radio-chemical testing of the off-gas system prefilter and afterfilters. (See next page.) Efficiencies of the prefilters were satisfactory. However, some difficulty measured'nd found to be was encountered in measuring concentrations that were used to evaluate afterfilter performance. The activity levels at the afterfilters were too low to detect with statistics. At TC-3E and 4E, a temporary bypass reliable'ounting of the in-service prefilter allowed small amounts of particulates to reach the afterfilter inlet. A more mean-ingful comparison could then be made between inlet "and outlet concentrations and, consequently, efficiency. In all cases, the activity levels of particulates leaving the filter was near or below the level'f detection. These tests and DOP test performed prior to initial operation of filter provide adequate justification that the filters 'ach were operating properly. Level II criteria have been satisfied. All required startup testing for the unit 3. modified off-gas system have been satisfactorily completed with those exceptions listed.
STARTUP RETEST PROGRAM SU>~is tARY REPORT BFHP UNIT ll 2 ls s Q P IJ 3.0 Results (Continued) u K
\
3. 20 STI-74 Modified Off-Gas S stem (Continued) Pl, 3.20.3 ~desi sis (Continued)
S stem ttEPA Filters (Continued) Table STI 74-5 11EPA Filter Performance
% .Power '1'5-40 40-60 65-85 95-100 HEPA Filter (1) Date 10/4/76 10/25/76 11/12/76 12/16/76 Deficiencies 1127 1969 2542 3127 I
TC-1 TC-2E TC-3E TC-4E Prefilter A Csl38 99. 9 99.91 99 4 99. 9
.Rb88 99.5 99.93 99.76 99.8 Prefilter B
99. 99 99. 92 >99.49(2) 99 9 Csl38 Rb88 99. 97 99.87 99. 76 99.9 Afterfilter A Cs138 .97.75 >42.05 >96 8 Rb88 -867 (3) 98.16 >59.6 >99.4 ',Y$
Afterfilter B 98.13 ) 97 (2) Cs138 Rb88 -41. 9 99. 05 >93.8 99.4 (1) Activity levels of Ba-140 entering and leaving the prefilters and after filters were too low to detect st:atistically. Therefore, the calculated efficiencies were meaningless and were omitted from this table. (2) ">"means that the actual 22 efficiency is some value larger than this value,but because a concentration ( or both) used to calculate the efficiency was itself less than the detectable concentration, the actual value cannot be determined. (3) When the afterfilter outlet concentration was decay-corrected to sample time, the effluent appeared to have more activity than the inlet:. (The efficiencies were negative.) Actually, both the inlet and outlet had activity levels too low to detect statistically. This was remedied at TC-3E and 4E when a partiial bypass of the pre-filter allowed particulates to reach t: he afterfil,ter. Thus, a comparison of inlet and outlot concent:rations could be made.
nEcEivcn oocuvrm PAOCESSl~G Ut<lT STD JUL 18 AM t0 29}}

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