ML20115A546

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Rev 0 to Limerick Generating Station,Unit 1 Startup Rept, Cycle 5
ML20115A546
Person / Time
Site: Limerick Constellation icon.png
Issue date: 09/30/1992
From: Doering J, Helwig D, Mowry C
PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To:
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9210140366
Download: ML20115A546 (58)
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Text

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_9250 3 .. 70 333 T.S.6.9.1'.1

- T.S.6.9.1.2 T*S*6'9'1'3 PHILADELPHIA ELECTRIC COMPANY LIMERICK GENFRATING STATION P. O. BvX 2300 SANATOGA, PA 19464-2300 (215) 3271200. EXT. 3dC DAVIC R. HELWlO vtCE FRES! CENT vurqu orveauwo surou October 5, 1992 Docket No. 50-352 License No. NFF-39 U. S. Nuclear Regulatory Corr:nissicn Attn: Document Centrol Desk Washington, DC 20555

Subject:

Limerick Generat_ng Station, Unit 1 Startup Report - Cycla 5 Enclosed is the Limerick Generating Station Unit 1, Cycle 5, Startup Report. The report is being submitted in accordance with Technical Specificatiens (TS) Reporting Requiremente 6.9.1.1, 6.9.1.2, and 6.9.1.3. The Report contains all pertinent information regarding the fifth cycle startup testing activities.

If you have any questions, or require additional information, please do not hesitate to contact us.

Very truly yours, T \

KWM/dtc Enclosure ac: T. T. Martin, Ad.inistrator, Region I, USNRC (w/ enclosure)

T. J. Kenny, U3"RC Senior Resident Inspector (w/ enclosure) 13003G l 1

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TABLE OF CONTENTS

'l . INTROD'UCTION/

SUMMARY

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1.1 Introduction 1.2 Summary ,

1.3 Limerick Plant Description Table 1.3-1 Limerick 1 Plant-Parameters ,

3.- RESULTS 2.1 STP-1, Chemical and Radiochemical 2.2 STP-2, Radiation Measurements 2.3 STP-3, Fuel Loading 2.4 STP-4, Shutdown-Margin Demonstration.

2.5 STP-5, Control Rod Drive System-2.6 .STP-6, SRM Performance and Control Rod Sequence 2.7 STP-9, ' Water Level Reference Leg-Temperature 2.8 STP-10, IRM Performance 2.9 STP-ll, LPRM Calibration 2.10 STP-12, APRM Calibration 2.11 STP-13, Process Computer 2.12 STP-14, Reactor Core. Isolation Cooling System 2.13 STP-15, High Pressure Coolant Injection

. System-2.14 STP-16, Selected-Process Temperatures 2.15. .STP-17, System Expansion 2.16 STP-18, TIP Uncertainty 2.17 STP-19, Core Performance

E TABLE OF CONTENTS 2.18 STP-20, Steam Production-2.19 STP-12, Pressure Regulator

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2.20 STP-23, Feedwater System ,,

2.21 STP-24, Turbine Valve Surveillance 2.22 STP-25, Main Steam Isolation Valves 2.23 STP-26, Relief Valves 2.24 STP-27, Main Turbine Trip 2.25 STP-28, Shutdown From Outside the Control Room 2.26 STP-29, Recirculation Flow Control Systh.a 2.27 STP-30, Recirculation System-2.28 STP-31, Loss of Turbine Generator and Offsite Power 2.29 - STP-32, Essential HVAC System Operation and Containment Hot Penetration

  • Temperature Verification 2.30 STP-33,-Piping Steady-State vibration 2.31 STP-34, Offgas Performance-Verification 2.32 STP-35, Recirculation System Flow Calibration-2.33 STP-36, Piping. Dynamic Transients 2.34 STP-37, Main: Steam System and Turbine Performance and Plant '

Dynamic-Response Verification 2.35 STP-70, Reactor Water Cleanup System -

2.3C STP-71, Residual Heat Removal System-e ec=-- w - y

Rrv. 0 Sei ember 1992 PHILADELPHIA ELECTRIC COMPANY LIMERICK GENERATING STATION 3 UNIT NO. 1 STARTUP REPORT CYCLE 5 Preparation Directed by:

J. Doering, Plant Manager Limerick Generating Station

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Prepared by:

C. M. Mowry, Reactor Engineer Limerick Generating Station 4

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1.'l >' REPORT ABSTRACT ThisLStartup Report, written to comply with. Technical Specifications paragraphs 6.9.1.1 thru 6.9.1.3,. consists of a summary:of the Startup and Power Escalation-Testing performed at Unit 1 of the Limerick Generating Stati"on.

This report is required since fuel of a different design-was installed during the fourth refueling outage of. Unit _l.

.- During this refueling outage, 240 bundles of Gell fuel,,were  ;

loaded into the core.

The report addresses each of the Startup Tests identified in chapter 14 of the FSAR and includes a description of the measured values of the operating conditions or characteristics obtained during-the test ~ program with a comparison of these values to the Acceptance Criteria.

Also included is a description of any corrective actions required to obtain satisfactory operation.

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SECTION 1 .

1 INTRODUCTION /

SUMMARY

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il.'2 Y SUMIMRY [

Limerick Unit'1 was out-of-service from March 20, 1992 to .

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' accommodate a refueling o'tage. . The unit returned to service-on July 4, 1992 and reached' full power operation =

-July 15, 1992. -

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The succesarully implemented startup program ensures that. y t' the fourth refueling outage of Limerick Ut.it:1 has resulted in no conditions or system characteristics that diminishes-

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the safe operation of the plant. The tests and data referenced in this report are on file at the Limerick Gererating' Station.

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-1, 3 + LIMERICK PLANT-DESCRIPTION e

The Limerick ~Generoting Station is a two unit nuclear. power.

plant.. - The-two units share a common control room,

i. refueling floor, turbine operating deck, radwaste system, and other auxiliary systems. -

The Limerick Generating Station is located on_the east bank

.- of the Schuylkill River in Limerick Township-of Montgomery County,_ Pennsylvania, approximately 4 river miles downriver from Pottstown, 35 river miles upriver from Philadelphia, e

and'49 river miles above the confluence of the Schuylkill with the-Delaware River. The-site contains 595 acres - 423 acres in Montgomery County and 172 in Chester County.

Each of the LGS units employs a General Electric Company boiling water reactor (BWR) designed to_ operate at a rated

. core thermal power of 3293 MWt with a corresponding groLa electrical output of'1092 MWe. Approximately 37 MWe are used for auxiliary-power,=resulting in a net electrical

-output of 1055 MWe. See Table 1.3-1 for Limerick Plant

, Parameters.

The containment for'each unit is a pressure suppression type designated as Mark II. -The drywell is a steel-lined concrete cone. located above the steel-lined concrete-cylindrical pressure suppression chamber. The drywell and suppression chamber are separated by a concrete diaphragm p slab which also serves tc strengthen the entire system.

The Architect Engineer and Constructor was Bechtel Power

-Corporation.

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The-plant is~ owned.and operated by-the Philade'.phia Electric Company.

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/ TABLE 1.3-1

Limerick-1-Plant-Parameters

. Parameter Value Rated = Power-(MWt). 3293

' Rated Core _ Flow (Mlb/hr) 100 (1)

Reactor Dome: Pressure (psia) 1020

< Rated Feedwater Temperature (Deg. F) 420 (4)

Total Steam Flow (M1b/hr) 14.159 Vessel Diameter (in) 251 Tota) Number of Jet Pumps 20 Core Operating Strategy Control Cell Core / Spectral Shift Number of Control Rods 185 Number of Fuel Bundles 764 Fuel' Type 8x8 (Barrier) and-9 x 9.(Barrier)

- Core active Fuel Length (in) 150

-Cladding Thickness (in) 0.032 for GE7 GE8B, and GE9B 0.0281for Gell Channel Thickness (in) 0.080 for GE7B, GE8B, and GE9B.(2) 0.100 for Gell MCPR Operating _ Limit 1.32'(3);

Maximum ~LHGR (KW/ft) 13.4 for GE7B fuel _ ,

14.4 for GE89, GE9B and-Gell fuels i

Turbine Control Valve Mode Partial Arc Turbine Bypass Valve Capacity (% NBR) 25 Relief Valve Capacity (% NBR) 87.4

Number of Relief Valves 14 Recirculation Flow Control Mode Variable Speed M/G Sets-4 s

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s NOTES FOR TABLE 1.3-1 (1) Unit 1 is analyzed for increased core flow to 105% -

(2) Except for-LGY 644 which required a replacement 100 mil channel

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(3) See Core Operating Limits Report for LGS Unit 1 Reload 4, Cycle 5 for spt ifins.

(4) Unit 1 is analyzed for a 60 degrees F final feedwater temperature reduction h

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9 SECTION 2 -

RESULTS ,

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Y.1 STP-1,iCHEMICAL'AND RADIOCHEMICAL -

" OBJECTIVES?

The principallobjectives of thisitest-are'a) to secure information on the chemistry and radiochemistry of_the reactor coolent, and b) to determine'that the sampling equipmenti procedures and analytical techniques,are adequate to supply-the data required to demonstrate that the-chemistry of;all parts of the entire reactor rystem o meet specifications and_ process requirements.. ,

ACCEPTANCE' CRITERIA '

9 Level 1 -

Chemical factors defined in the Technical Specifications and Fuel Warranty must be maintained'within the_ limits specified.

The activity of gaseous and liquid effluents must conform to license limitations.

Water quality must be known at all times and must-remain within the guidelines of the Water Quality. Specifications.

Level 2 None -

RESULTS During Startup-of. Limerick Generating Station Unit 11 reactor, following.its fourth refucling outage, reactor. ,

coolant chemistry parameters as'well as radioactive gaseous waste releases _and radioactive _ liquid waste releases-were.

maintained within the-limits set forth in the Limerick Generating Station-Unit 1 Technical Specifications. The following is a list of Chemistry related surveillance tests:

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satisfactorily performed in support of unit startup activities:

ST-5-041-800-1, ST-5-041-875-1, ST-5-041-876-1,

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normal;ana' lyse'swerefpqformed.

(Iodinelwas:1.71 Ex -10 Luci/gt-(Tech Speci0.2JuCi/g). _.

Ji Fuel 1 Warranty--Appendix I" "WaterLQualitr Requirements.lwerel 9

met during startup.: ,From17-'4-92tthrough-7_-17-92'with --

reactor' power greatersthan40%,zreactor water conductivity. .

averaged 0.178 umho/cm,f(Fuel ~Warrantyflimit/liO)achlorides}

, remained:lessEthnn-2?ppba(Fuel Warranty limit 1100 ppb),x::andi , @

pH ranged =from 6.323to07.01'-(Fue10WarrantyFRangej5.6f j 1 8.6). 'Above 50%'powerfifeedwater coppericoncentration_ . ,3 reached-a' maximum of-1.22: ppb, ironFreachedi;0.247 ppb.'and .

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> total metals' reached-1.55; ppb (Fuel Warrantyj! limit 12ippby fi

'10-ppb, and-15; ppb respectively). _TheLhighestEcondensate demineralizer effluent conduct 4vity above150%-powerLwas .

O.055 umno/cm.;- ,

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During the; Refuel.Ou.i.ageifa' Condensate Deep-Bed. .

Demineralizer System was installed?to improveJthe chemistry _

-of f eedwater :. injected _:.into.i thef reactor. -: _of jpr imary. concern- .

was improving control'of the feedwater~ copper 5 concentration. =The' modification has not impacted the safe 3 operation of the plant, i condensate'and reactor water.cleanupJdemineralizer: . .;

performance was monitored.closelyfduring'the..startup. l

'Demineralizers were: regenerated as.necessary to maintain'

reactor water._ conductivity lesslthan 0.3.umho/cm.
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I 2.2 STP-2, RADIATION MEASUREMENTS

' OBJECTIVES The objectives of this test are to a) determine the background radiation levels in the plant environs prior to operation for base data to assess future activity, buildup and b) monitor radiation at selrs:ted power levels- to assure the protection of personnel during plant operation.

ACCEPTANCE CRITERIA ,

Level i 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 10CFR20

" Standards-for Protection Against Radiation".

Level 2 None RESULTS Health Physics procedure HP-203, "HP Startup Surveillance

, Procedure" was iraplemented during rea. tor startup. This-procedu e directs Health Physics surveillance throughout

  • the plant to help ensure plant posting and RWP's are updated as reactor power increases.

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2.3 STP-3, FUEL LOADING OBJECTIVE The objective of this test is to load fuel safely and efficiently tv the full core size.

ACCSPTANCE CRITERIA ,

Level 1 The partially loaded core must be subcritical by at least 0.38t delta k/k with the analytically determined strongest-rod fully withdrawn.

Level 2 None RESULTS The beginning of cycle shutdown margin calcalated in the Cycle Management Report Limerick 1 Cycle 5 was 1.17% delta K/K. Core reload was conducted in Lccordance with Technical Specifications. Equipment required to be operable to ensure that the shutdown margin is maintained was verified operable by various performances of ST i-107-630-1 and ST-6-107-591-1 between March 20, 1992

. and July 4, 1992 Post alteration core verification was

-completed on May 8, 1992 after all refueling operations were completed by the performance of ST-3-097-355-1. All fu:l bundles were verified to be in their proper core locations and properly oriented in the control cell. The bundle seating pass identified three fuel bundles improperly seated (51-10, 49-08, 31-20). The bundles were o properly reseated, and the location and orientation was reverified after reseating.

2.4 STP-4, SHUTDOWN MARGIN DEMO!1STRATION OBJECTIVES The purpose of this test is to demonstrate that the reactor will be sufficiently suberitical throughout the cycle with any single control rod fully withdrawn.

ACCEPTANCE CRITERIA -

Level 1 The shutdown margin (SDM) of the fully loaded, cold (68 degrees P), xenon-free core occuring at the most reactive time during the cycle must be at least 0.38% delta K/K wit.

the analytically strongest rod (or it's reactivity equivalent) withdrawn. If the SDM is measured at sometime during the cycle other than the most reactive time, compliance with the above criteria is shown by demonstrating that the SDM is 0.38% delta K/K plus an exposure dependent correction factor which corrects the SDM at that time to the minimum SD:1.

Level 2

. Criticality should occur within +1.0% delta K/K of the predicted critical.

RESULTS An "In Sequence" shutdown margin of at least 1.73% delta K/K was obtained during the reactor startup. This satisfies the Level 1 acceptance criteria. Test data is documented in ST-6-107-875-1 completed on July 4, 1992.

'Jsing the 'ata obtained during the shutdown margin Lemonstration, the difference between criticality and predicted critical was -0.56% delta K/K. This was within t1e Level 2 acceptance criteria.

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-2 5 STP-5, CO!3 TROL ROD DRIVE SYSTEF OBJECTIVES The objectives of this test are to demonstrate that the Control Rod Drive (CRD) System operates properly over the full range of primary coolant operating temperatures and pressures, and to determine the initial operating' characteristics of the CRD system. .

ACCEPTANCE CRITERIA l .

Level 1 Each CRD must have a normal withdraw speed less than or equal to 3.6 inches per second, indicated by a full 12 foot stroke in greater than or equal:to 40 seconds. i The mean scram time of all operable CRD's must not exceed the following times (Scram time is measured from the time '

the pilot scram valve solenoids are de-energized):  :

Position Inserted to From Fully Withdrawn Scram Time (Seconds)

, 45 0.43 39 0.86 25 1.93  ;

05 3.49

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The mean scram time of the three fastest CRD's in a two by -

two array must not exceed the following times (Scram time is measured from the. time the pilot scram valve solenoids are de-energized):

Position Inserted to Prom Fully Withdrawn Scram Time (Seconds)-

45 0.45 '

39 0.92 25 2.05 05 3.70

  • Level 2 Each CRD must have normal insert and withdrawn speeds of 3.0 + 0.6 inches per second, indicated by a full 12 foot -1 stroEe in 40 to-60 seconds.

!' RESULTS l

L Although the performance of the Control-Rod Drive-System was not affected by the installation.of the new fuel

. design, the scram time limits are required to assure-thermal limits such.as critical power-ratio _are not exceeded.

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. 6 Lovol I cnd Lovel 2 stroke tima ecccptanco critoric woro '

. fully satisfied by the-performance of ST-6-047-760-1 on June _1, 1992 during_.the operation hydrostatic test.

Level I scram time acceptance criteria were fully satisfied by the performance of ST-3-107-790-1 on June 2, 1992 during  !

. the operational hydrostatic test.

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i 206 STP-6, SRM PERFORMANCE AND CONTROL ROD SEQUENCE l OBJECTIVES The objective of this test is to demonstrate that the operational neutron sources, SRM instrumentation, and rod 6 withdrawal sequences provide adequate information to achieve criticality and increase power in a safe and ef ficient n.anner .-

f ACCEPTANCE' CRITERIA ,

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Level 1 There must be a neutron signal to ncise count ratio of a  :

least 2:1 on the required operable SRMs.

Th'ere must be a minimum count rate of 3 counts /second on the required operable SRMs.

Level 2 None RESULTS Minimum SP1 count rate was determined to be greater than 3 '

CPS-by the performance of ST-6-107-591-1 prior to the withdrawal of control rods on July 4, 1992. The .

signal-to-noise ratio verification is only required to be performed in accorcance with Tech Specs if the SRM count rate is less than 3.0 CPS. ,

-Since at no time during the startup was the count rate less-than 3.0 CPS,-this verification was not performed.- SRM- >

response was verified by the performance-_of ST-6-lu7-875-1 on July _4, 1992 until criticality was achieved.

2.7 STP-9, WATER LEVEL REFERENCE LEG TEMPERATURE L

-OBJECTIVES The ob4ectives of this test are.to measure the level instrumentation reference leg temperature, recalibrate the water level instruments if the measured temperature is-significantly different from the value assumed during the i initial end points calibration, and to obtain baseline data j on the' Narrow Range and Wide RLnge water level instrumentation.

ACCEPTANCE CRITERIA Level 1 I

None i I

Level 2

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The difference between the actual reference leg ,

temperature (s) and the value(s) assumed durino initial calibration shall be less than that amount which will result in a scale end point error of 1% of the instrument 1 span for each range. l RESULTS The new fuel design did not affect the performance of ,

systems needed-to satisfy the acceptance criteria of this test.

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2.8 STP-10, IRM PERFORMANCE j

. OBJECTIVES The objectives of this test are to adjust the Intermediate .i l

Range Monitoring (IRM) System to obtain an optimum overlap with the SRM and APRM systems. ,

ACCEPTANCE CRITERIA .

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Each IRM channel must be on scale before the SRM's exceed  :

their rod block setpoint.

i Each APRM must be on scale before the IRM's exceed their rod block setpoint.

Level 2 t Each IRM channel must be adjusted so that one-half decade '

overlap with tne SRM's is assured.

Each IRM channel must be adjusted so that one decade overlap with the APkM's is assured.

RESULTS

- Technical Specification SRM/IRh overlap was satisfied by the performance of ST-6-?.07-884-1 on July 4~, 1992. This test demonstrated at least a half decade SRM/IRM overlap.

During startup, all roquired APRM's were verified to be on_

ncale before any IRM-exceeded their scram setpoint of 120% ,

of scale. This was documented on GP-2, Normal Plant Startup, on July 8, 1992. One-half decaue IRM/APRM overlap is verified in accordance with.Tectrica) Specifications during each controlled shutdown by the performance of' ST-6-107-886-1.

2.9 STP-ll, LPRM CALTBRATION

, OBJECTIVES ,

The objectives of this test are to calibrate the Local Pcwer Range Monitoring (LPRM) System and to verify LPRM Plux Response.

ACCEPTANCE CRITERIA-Level 1 None ,

Level 2 -

Each LPRM reading will be within 10% of it's calculated value.

RESULTS LPRM calibrations were performed at 25% power and 100%

power per ST-3-074-505-1 on July 12, 1992 and July 22, 1992 respectively. On July 22, 1992 the LPRM's were calibrated to within 4% of their calculated value.

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2.10 STP-12,-APRM CALIBRATION OBJECTIVES The objective of this test is to calibrate the Aterage Power Range Monitor (APRM) Systen..

ACCEPTANCE CRITERIA ,

Level 1 The APRM channels must be. calibrated tu read equal to or greater than the actual core thermal power.

Technical specification and fuel warranty limits on APRM scram and Rod Block shall not be exceeded.

In the s'artup mode, all APRM channels must produce a scram at less than or equal-to 1"4 of rated thermal power.

Level 2 If the above criteria'are satisfied, then the APRM channels w131 be considered to be reading accurately if they agree with the heat balance or the minimum value required based on peaking factor, MLHGR, and fraction of rated power to within (+7,-0)% of rated power.

RESULTS By various performances of ST-6-107-885-1 from July 12, 1992 to July 15, 1992, Level 1 acceptance criteria was met by verifying APRM channels were indicating greater than or equal to actual-core thermal power.and below the scram and rod block setpoints when thermal power-was greater than 25% . Level 2 acceptance criteria was also met in this surveillance test by adjusting indicated APRM reading to within +2, -0%-(not to exceed 100%) of the greater of

! fraction of rated power or maximum fraction limiting _ power density.

The Level 1 acceptance criteria of APRM scram setpoint of 15% was met by per*ormance of channel functionai tests ST-2-074-412-1, ST-2-074-413-1, ST-2-074-414-1, ST-2-074-415-1, ST-2-074-416-1, and ST-2-074-417-1 L performed on June 29 through July 2,.1992.

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i 2.11 STP-13, PROCESS COMPUTER

_OBJECTIVEj The objective of this test is to verify the performance of ,

the Process Computer under plant operating conditions.

ACCEPTANCE CRITERI A ,

Level -1

/ None ,

Level 2 The MCPR calculated by BUCLE and the Process Computer either

- are in the same fuel assembly and do not differ in value by more than 2% or

- in the case in which the MCPR calculated by the Process Computer is in a different assembly than'that calculated by BUCLE, of each assembly, the MCPR and the CPR calculated by the two methods shall agree w4 thin 2%.

The maximum LHGR calculated by BUCLE and the Process Computer either are in the same fuel assembly and do not differ.in value by more than 2%, or

- in the case in which the maximum LUGR calculated by.the Process Computer is in a different assembly'than that calculated by BUCLE, of each assembly, the maximum LHGR and the LHGR calculated by the two. methods shall agree within 2%.

The MAPLHGR calculated by BUCLE and the Process Computer either:

are in the same fuel assembly and do not differ in value by more than 2%, or

- in the case in which the MAPLHGR calculated by the Process Computer is in a different assembly than that calculated by_BUCLE, of each assembly,-the MAPLHGR and-APLHGR-calculated by the two methods shall agree-within 2%.

The LPRM gain adjustment factors ' calculated by BUCLE and the' Process Computer agree to within 2%.

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s HESULTS v

.On July 15, 1992 at 99.8% core thernal power, the accuracy

'of the thermal limits and LPRM gain adjustment factor calculated by-the Process Comp' iter were compared to the values calculated by an offline computer program call Backup Core Limits Evaluation (BUCLE).- The acceptance criteria for thermal limits determination was satisfied in all cases. Table 2.11-1 summarizes the-thermal 11'mits ,

data. <Also, all LPRM gain adjustment factors calculated by BUCLE and the Process Computer for operable LPRM's were e determined to be within 2%. ,

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t TABLE 2.11-1 LCS 1 BOC 5 100% Power P1 to 3UCLE Comparison Value P1 Data BUCLE Data 7-15-92, 1644 7-1A-92, 1644 CMWT 3288 3288 ,

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e Location Value ,

P1 BUCLE ,

19-16-4 07829 0.829 23-12-4 0.874 0.873 41-16-4 0.827 0.829 ,

11-24-4 0.869 0.869 31-24-12 0.808 0.806 49-24-4 0.869 0.869 19-46-4 0.829 0.829 23-50-4 0.874 0.873 41-46-4 0.829 0.829 MPLCPR Location Value P1 BUCLE 19-18 57833 0.832 35-14 0.880 0.879 41-18 0.833 0.832 13-36 0.870 0.869 27-36 0.829 0.829 47-26 0.870 0.869 19-44 0.833 0.832 25-/.8 0.88.1 0.879 41-41 0.833 0.832 ,

MAPRAT Location Value P1 BUCLE-16-4 0.854 0.853 +

37-12-4 0.905 0.905-41-16-4 0.854 0.853 11-38-4 0.903 0.902 <

29-38-12 0.825 0.823 49-38-4 0.903 0.902 ,

19-46-4 0.854 0.853 3*,-50-4 0.905 0.905 41-46-4 0.854 0.853 u _. _ _ .- _

t

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'b W & -- , e , , -, - -

2c12 STP-14, RCIC SYSTEM OBJECTIVES The objectives of this test are to verify the proper operation of the Reactor Core Isolation Cooling (RCIC)

System over its expected operating pressure and flow ranges, and to demonstrate reliability in automatic starting from cold standby when the reactor is at' power  :

conditions.

' ACCEPTANCE CRITERIA

  • evel 1 The average pump discharge flow must be equal to or greater than 100% rated value after 30 seconds have elapsed from automatic initiation at any reactor pressure between 150 psig and rated.

The RCIC turbine shall not trip or isolate during auto or manual start tests.

Level 2 .

In order to provide an overspeed and isolation trip avoidance margin, the transient start first altd subsequent speed peaks shall not exceed 5% above the rated RCIC turbine speed.

The speed and flow con:rol loops shall be adjusted so that the decay ratio of any RCIC system related variable is not greater than 0.25.

The turbine gland seal condenser system shall be capable of preventing steam leakage to the-atmosphere.

The delta P switches of the RCIC steam supply line high flow. isolation trip shall be calibrated to actuate at the value specified in the plant technical specifications (about 300%).

The RCIC system must have the capability to deliver specified flow against normal rated reactor pressure without the normal AC site power supply.

RESULTS The new fuel design did not affect the performance of systems needed to satisfy the acceptance criteria-of this test.

_ _ _ - ~ _ - ,

2013 STP-15, HPCI SYSTEM 0,BJ ECTITES_

The objectives of this test are to verify the proper operation of the High Press'.re Coolant Injection (HPCI)

System over its expected operating pressure and flow ranges, and to demonstrate reliability in automatic starting from cold standby when the reactor-is at' rated j pressure-conditions.

ACCEPTA:1CE CRITERIA ,

Level 1 The average pump discharge flow must be equal to or greater than 100% rated value after 30 seconds.have elapsed from automatic initiation at any reactor-pressure between 200 psig and rated. ,

The HPCI turbine shall not trip or isolate during auto or manual start tests.

Level 2 ,

In order to provide an overspeed isolation trip margin, the transient first peak shall not come closer than 15%.(of rated speed) to the overspeed trip, and subsequent speed '

peaks shall not be greater than 5% above the rated turbine cpeed.  ;

The speed and flow control loops shall be adjusted so that-the decay ratio of any HPCI system related variable is not greater than 0.25.

The turbine gland acal condenser system shall be capable of preventing-steam leakage to the atmosphere. ,

The delta P switches of the HPCI steam supply.line high flow isolation trip shall be calibrated to actuate at the value specified in plant technical specifications (about 300%).

RESULTS The new fuel design did not affect the: performance of ,

systems needed to satisfy the acceptance criteria of this test.-

l L

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i.

i,'

t-2014 STP-16, SELECTED PROCE!!S TEMPERATURES OBJECTIVES ,

The objectives of this test are (1) to assure that the measured bottom head drain temperature corresponds to bottom head coolant temperature during normal operations, ,

(2) to identify any reactor operating modeu that cause ' '

temperature stratification, (3) to determine tt.e. proper ,

setting of the low flow control limiter of the recirculation pumps to avoid coolant temperature

  1. stratification in the reactor pressure vessel bottom, head region.

ACCEPTANCE CRITERIA Level 1 The reactor recirculation pumps shall not be started, flow increased, nor power increased unless the coolant temperatures between the steam dome and bottom head drain are within 145 degrees F.

The_ recirculation pump in an idle loop must not be started, active loop flow must not be raised and power must not be increased unless the idle loop suction temperature is within 50 degrees F of the active loop suction temperature and the active-loop flow rate is less than or equal to 50%

. of rated loop flow. If two pumps are idle, the loop suction temperature must be within 50 degrees P of the steam dome temperature before pump startup.

Level 2 During two pump operation at rated core flow, the bottom head torperature, as measured by the bottom head drain line thermocouple, should be within 30 degrees F of the recirculation loop temperatures.

RESULTS The new. fuel design did not affect the-performance of-systems needed to satisfy the acceptance criteria of this test.

l 1

2.15 STP-17, SYSTEM EXPANSION OBJEC' '/ES This test verffles that safety related piping systems and other piping systems as identified in the PSAR expand in an acceptable manner during plant heatup and power escalation.

Specific objectives are to verify that: ,

Piping thermal expansion is as predicted by design calculations.

Snubbers and cpring hangcre remain within operating travel ranges at various piping temperatures. ,

Piping is free to expand without interferences.

1 ACCEPTANCE CRITERIA Level 1 There shall be no obstructions which will interfere with the thermal expansion of the Main Steam (inside drywell) and Reactor Recirculation piping systems.

The displacements at the established transducer locations shall not exceed the allowable values.

Level 2 The displacements at the established transducer locations shall not exceed the expected values.

Snubbers and spring hangers do r.ot become extended or compressed beyond allcwable travel limits (working range) and snubbers retain swing clearance.

Measured displacements compared with the calculated displacements are within the specified range.

Residual displacements measured following system return to ambient temperature do not exceed the greater of + 1 or + 25%of the maximum displacements measured durIng/16 in.

system initial heatup.

RESULTS The new fuel design did not affect the performance of systems needed to satisfy the acc'ptance criteria of this test.

-- " ~~~ _ , - , - _a ,

2.16 STP-18, TIP UNCERTAINTY OBJECTIVES The objective of this test i s to determine the reproducibility of the Traversing Incore Probe system readings.

ACCEPTANCE CRITERIA -

Level 1 None Level 2 The total TIP uncertainty (including random noise and geometrical uncertaintien) obtained by averaging the uncertainties of all data cets shall be less than 7.1%.

RESULTS Total TIP uncertaint) was determined by the performance of RT-3-074-850-0 on September 22, 1992. Level 2 acceptance criteria was met by all data sets with a maximum uncertainty of 1.46%.

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- - - - - - - --- ----_..------------------------,-_---,-----u_-----a----- - - - _ - _ _ _

2.17 STP-19, CORE PERPORMANCE

. OBJECTIVES The objectives of this test are tot a) Evaluate the core therual power and core flow rater and b') Evaluate whether the following core performance parameters are within limits Maxir.um Linear Heat Generation Rate (MLHGR),

Minimum Critical Power Ratio (MCPR),

Maximum Average Planar Linear Heat Generation Rate (MAILHGR).

ACCF9TANCE CRITERIA Level 1 The Maximum Linear Heat Generation Rate (MLHGR) of any rod during steady-state conditions shall not exceed the limit specified by the Plant Technical Specifications.

The steady-state Minimum Critical Power Ratio (MCPR) shall exceed the minimum limit specified by the Flant Technical Specifications.

T.1e Maximum Average Linear Heat Generation Rate (MAPLHGR) t shall not exceed the limits specified by the Plant 2 Technical Specifications.

Steady-utate reactor power shall be limited to the rated core thermal power '3293 MWt).

Coce flow shall not exceed its rated value (105 Mlb/hr).

Level. 2 None RESULTS With thermal power limited to 3293 MWth and core flow limited to 105 Mlb/hr, Level 1 acceptance criteria of thermal limits were met and documenteo throughout the startup by various performancea of ST-6-107-885-1 from July 12, 1992 through July 17, 1992. .

I ___ _ _.__ ___- _._ _ . ___m- _ _ _ _ . .__.__..m._______.-_________-_________-.______u._._____._______._m____.m_____-_-_a_----_-.__u_.---___-__-__---_ - - - - _ - _ M

2.18 STP-20, STEAM PRODUCTION OBJECTIVES The objectives of this test are to demonstrate that the Nuclear Steam Supply System (NSSS) can provide steam sufficient to satisfy al? appropriate warranties as defined in the NSSS contract.

ACCEPTANCE CRITERIA Level 1 The NSES parameters as determined by using normal operating procedures shall be within the appropriate . license restrictions.

The NSSS shall be capable of supplying 14,159,000 pounds per hour of steam of not less than 99.7% quality at a pressure of 985 psia at the discharge of the second main steam isolation valve, as based upon a final reactor feedwater temperature of 420 degrees P and a control rod drive feed flow of 32,000 pounds per hour at 80 degrees F.

The reactor feedwater flow must equal the steam flow less the control rod drive feed flow.

Level 2 None RESULTS The new fuel design did not affect the performance of systems needed to satisfy the acceptance criteria of this test.

2.19 STP-22, PRESSURE REGULATOR OBJECTIVES The objectives of this test are as follows:

To demonstrate optimized controller settings of the pressure control loop by analysis of the transientr. induced in tne reactor pressure control system by means of the pressure regulators set point changes.

To demonstrate the take-over capcbility of the back-up pressure regulator upon failure of the controlling pressure regulator, and to set spacing between the setpoints at an appropriate value.

To demonstrate smooth pressure aontrol transition between the turbine control valves and the bypass valves when reactor steam generation exceeds the steam flow used by the curbine.

To demonstrate the stability of the reactivity-void feedback loop to pressure perturbations in conjunction with STP-21, Core Power Void-Mode Response.

ACCEPTANCE CRITERIA Level 1 The transient response of any pressure control system related variable to any test input must not diverge.

Level 2 Pressure control system related variables may contain oscillatory modes of response. In these cases, the decay ratio of each controlled mode of response must be less than or equal to 0.25. (This criterion does not apply to tests involving simulated failure of one regulator with the

( backup regulator taking over.)

The pressure response time from initiation of pressure setpoint change to the turbine inlet pressure peak shall be

<10 seconds, s Pressere control system deadband, delay, etc.,-shall be small enough that steady state limit cycles (if any) shall produce steam flow variations no larger than +0.5 percent of rated ateam flow.

The peak neutron flux and/or peak vessel pressure shall remain below the scram settings by 7.5 percent and 10 psi respectively of all pressure regulator transients performed at Test Condition 6.

+

. _ - _- - . - - _ _ . - - . _ - . . - - _ _ _ _ _ . - . - - . _ _ . - - - - - _ - - - _ - - - _ _ _ _ _ _ , _ _ - - - - - = . - - - -

Tha variction in incrcmantal regulation (ratio of tho '

maximum to the minimum value of the quantity, " incremental change in pressure control signal / incremental change in

' steam flow",_of each flow range) shall meet the following:

% of Steam Flow Obtained With Valves Wide Open Variation 0 to 85% $4 : l' 85% to 97% 12:1 97% to 99% $5:1 RESULTS

-The new fuel design did not affect the performance ef .

systems needed to satisfy the acceptance critet!* of this-test.

I

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2.20 STP-23, FEEDWATER SYSTEM OBJECTIVES i

The objectives of this test are:

To demonstrate that the feedwater system has been adjusted to provide acceptable reactor water level control.

To demonstrcte an adequate response to a feedwater temperature reduction.

To demonstrate the capability of the automatic core ' flow runback feature to prevent low water level scram following the trip of one feedwater pump at high power operation.

To demonstrate that che maximum feedwater runout capability is compatibic with the licensing assumptions.

ACCEPTANCE CRITERIA Level 1 The transient response of any level control system-related variable to any test input must not diverge.

For the feedwater heater loss test, the maximum feedwater temperature decrease due to a single failure case must be

<100 deg. P. The resultaitt MCPR must be greater than the Tuel thermal safety limit.

The increase in simulated heat flux cannot exceed the predicted Level 2 value by more than 2%. The predicted value will be based on the actual test valites of feedwater temperature changes and initial power level.

Maximum speed attained shalf ?t exceed the speeds which will give the fe lowing flowa sith the normal complement of pumps operating.

a. 144% NBR at 1075 paia
b. 144% NBR at 1020 psia Level 2 Level control system-related variables may contain oscillatory modes of responso. In these cases, the decay ratio of each controlled mode of response must be less than or equal to 0.25 l

__. . _ . _ _ _ _ _ _ _ _ . _ _ . _ _ _ _ _ - _ _ __.__.m.___.______m. _ _ _ _ _ _ _ _ . _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ , ,__ ..

The open loop dynamic flow response of each feedwater actuator (turbine) to small (<10%) step disturbances shall be:

a. Maximum time to 10% of a step disturbance $1.1 sec
b. Maximum time of 10% to 90% of a step disturbance , $1.9 sec
c. Peak overshoot (% of step disturbance) 115%
d. Settling time, 100% 15% $14 sea The average rate of respanse of the feedwater actuator to large (>20% of pump flow) step disturbances shall be between 10 percent and 25 percent rated feedwater flow /second. This average response rate will be assessed by determining the time required to pass linearly through the 10 percent and 90 percent respunse points.

As steady-state generation of the 3/1 elecent systems, the input scaling to the mismatch gain should be adjusted such that the level error due to biased mismatch gain output should be within il inch.

The increase in simulated heat flux cannot exceed the predicted value referenced to the actual feedwater temperature change and initial power level.

The reactor shall avoid low water level scram by three inchcs margin from an initial water level halfway between the high and low level alarm setpoints.

The maximum speed must be greater than the calculated speeds required to supply:

a. With rated complement of pumps - 1151 NBR aL 1075 psia L. One feedwater pump cripped conditions - 68% NBR at 1025 psia.

RESULTS The new tuel design did not affect the performance of systems needed to satisfy the acceptance criteria of this test.

2.21 STP-24, TURBINE VALVE SURVEILLANCE i OBJECTIVES

-The objectives of this test are to demonstrate occeptable procedures and maximum power levels of periodic surveillance testing of the main turbine control, stop and bypass valves without producing a reactor scram. ,

ACCEPTANCE CRITERIA

, Level 1 None Level 2 Peak neutron flux must be at least 7.5% below the scram trip setting.

Peak vessel pressure must remain at least 10 psi below the high pressure scram setting.

Peak steam flow in each line must remain 10% below the high flow isolation trip setting.

RESULTS The new fuel design did not affect the performance of systems needed to satisfy the acceptance criteria of this test.

9

2.23 STP-25, MAIN STEAM ISOLATION VALVES OBJECTIVES The objectives of this test are to functionally check the Main Steam Isolation Valves (MSIV's) of proper operation at selected powet levels, to determine the MSIV closure times, and to determine the maximum power level at which . full closure of a single MSIV can be performed without.~ causing a reactor scram.

The full isolation is performed to determine the reactor transient behavior that results from the simultaneous full closure of all MSIV's at a high power level.

ACCEP*SNCE CRITERIA Level 1 MSIV stroke time shall be no faster than 3.0 seconds. MSIV closure time shall be no slower than 5.0 seconds.

The positive chenge in vessel dome pressure occ;rring within 30 seconds after closure of all MSIV's must not exceed the Level 2 criteria by more than 25 psi. The positive change in simulated heat flux shall not exceed the Level 2 criteria by more than 2% of rated value.

Feedwater control rystem settings must prevent flooding of the steam lines.

Reactor must scram to limit the severity of the neutron flux and simulated heat flux transients.

Level 2 The reactor shall not scram The peak neutron flux must be at least 7.5 percent below the trip setting. The peak vessel pressure must remain at least 10 psi below the high pressure scram setting.

The reactor shall not isolate. The peak steam flow on each line must remain 10 percent below the high steam flow isolation trip retting.

The temperature measured by thermocouples on the discharge side of the safety / relief valves must return to within 10 degree P of the temperature recorded before the valve was opened.

e The positive change in vessel dome pressure and simulated heat flux occurring within the first 30 seconds after the closure of all MSIV valves must not exceed the predicted values. Predicted values will be referenaed to actual test conditions of initial power level and doras pressure and will use beginning of life nuclear dmta.

If water level reaches the reactor vessel low water level (Level 2) setpoint, RCIC and !!PCI shall automatically initiate and reach rated syste; flow.

Recirculation pump trip shall be initiated if water Level 2 is reached.

RESULTS The new fuel design did not affect the performance of systems needed to satisfy the acceptance criteria of this test.

2.23 STP-26, RELIEF VALVES

. OBJECTIVES The objectives of this test are a) to verify that the Relief Valves function properly (can be manually opened and closed, b) to verify that the Relief Valves rescat properly after actuation, c) to verify that there are no major blockagts in the Relief Valve discharge piping, and d) to dernonstrate system stability to Relief Valve operation.

ACCEPTANCE CRITERIA f

Level 1 There should be a positive indication of steam discharge during the manual actuation of each Relief Valve.

The flow through each Relief Valve shall compare favorably with value assumed in the PSAR accident analysis at normal operating Reactor pressure.

Level 2 Pressure control system-related variables may contain ,

oscillatory modes of response. In these cases, the decay ratio of each centro 11ed mode of response must be less than or equal to 0.25.

The temperature measured by the thermocouples on the

discharge side of the valves shall return to within 10 DEG F of the temperature recorded before the valve was openeu.

During the low pressure functional test, the-steam flow-through each Relief Valve, as measured by Bypass Valve position, shall not be less than 90% of the average Relief Valve steam flow.

During the rated pressure functional test, the steam flow through each Relief Valve, as .seasured by-Aenerator Gross MWe, shall not be lower than the average valve response by more than 0.5% of rated MWe.-

RESULTS L

, The new fuel design did not affect the performance of j systems needed to satisfy the acceptance criteria of this test.

l' 1

i 2.24 STP-27, MAIN TURBINE TRIP OBJECTIVES ,

The objectives of this test are to de'nonstrste the response of the Reactor and its control systens to protective trips of the Main Turbine and to evaluate the response of the bypass and safety / relief valves.

4 ACCEPTANCE CRITERIA Level 1 .

4 For Turbine and Generator Trips at. power levels greater than 50% Nuclesr Boiler Rated, there should be a delay of less thar. 0.1 seconds following the beginning of Control or 1 Stop Valve closure before the beginning of Bypacs Valve opening._ The Bypass Valves shou'id be opened to a point corresponding to greater than or equal to 80% of their '

capacity within 0.3 seconds fron the beginning of Control ot Stop Valve closure motion.

Feedwater System settings must prevent flooding of the steam lines following these transients.

The positive change in vessel dome pressure occut..ing "m within 30 seconds after either Generator or Turbine Trip must not exceed the Level 2 criteria by more than 25 psi.

The positive change in simulated Heat Flex shall not exceed the Level 2 criteria by more than 2% of Rated Value, The recirculate ar pump and motor time constants of the two pump drive flow coastdown transient should be <2.5 seconds from 1/4 to 2 seconds after the pumps are tripped.

The total time delay from tue start of the Turbine Stop.

Valve or Control Valve motion to the complete suppression of the electrical arc cetween the fully open contacts of the RPT circuit breakers shall be less than or equal to 175 tr il liseconds .

Level 2

( There shall be no'MSIV closure during the first three minutes of the traasient and operator action shall not be required during that period to avoid the MSIV closure.

1

, y.

The positive change in vessel dome pressure occurring within the first 30 seconds after the initiation of either l

Ger.erator or Turbine Trip must not exceed predicted values.

The positive change in simulated Heat Flux occurring within the first 30 seconds after the init'.ation of either Generator or T"rbine Trip must not exceed predicted values.

Feedwater level control shall avoid loss if feedwa'ter flow due to a high (L8) water level trip during the event.

Low (L2) water level recirculat'on po' ' rip, HPCI and RCIC-shall not be initiated.

The temperature mea pre *' by therrscouples on the discharge side of the Relief Valves must return to within 10 Degree F of the temperature recorded before the valve was opened.

For the Turhine Trip within the Bypass Va3ves capacity, the Reactor shall not scram.

The measured Bypass Valve capability shall be equal to or greater than that used in the FSAR analysis (25% of Nuclear ,

Boiler Rated Steam Flow).

RESULTS The new fuel design did not affect the performance of systems nee 6ed to satisfy the acceptance criteria of this test.

2.25 STP-28, SHUTTOWN PROM OUTSIDE THE CONTROL ROOM OBJECTIVES The objectives of this test are to demonstrate that the Reactor a) can be safely shutdown from outside the Control Room, b) can be maintained in a Hot Standby condition from

outside the Control Room and c) can be safely cooled from i

hot to cold shutdown from outside the Control Room'. I n-addition, it will provide an opportunity to demonstrate that the procedures of Remote Shutdown are clear and comprehensive and that operational personnel are familiar e with their applications.

ACCEPTANCE CRITERIA Level 1 ,

None Level 2

  • During a simulated Control Room evacuation, the Reactor must be brought to the point where cooldown is initiated and onder control, and Reactor vessel pressure and water level are controlled u' ng equipment and controls located outside the Control Rc ..

The Reactor can be safely shutdown to a Hot Standby condition from cutside the Control Room using the minimum shift crew complement.

The Reactor coolant temperature and pressure can be lowered sufficiently (at a rate that does not exceed the Technical Specification Limit) from outside the Control Room to permit operation of the Shutdown Cooling Mode of the Residual Heat Removal System. -

The Shutdown-Cooling Mode of the Residual Heat Removal System can be initiated fru' outside the Control Room with a heat transfer path establisnci to the Ultimate Heat Sink.

The Shutdown Cooling Mode of the Residual Heat Removal System can be used to reduce Reactor coolant temperature at a rate which does not exceed the Technical Specification Limit.

RESULTS The new fuel design did not affect the performance of systems needed to satisfy the acceptance criteria of this test.

2c26 STP-29, RECTRCULATIOff FLOW CONTROL SYSTEM OBJECTIVES The objectives of this test are to demonst. ate the flow control capability of the plant over the ent-ire pump speed range, in both II.dividual Local Manual and Combined Master Manual operation modes and to determine that the controllers are set of the desired system performa'nce and stability.

ACCEPTANCE CRITERIA .

Level 1 The transient response of any recirculation system-related variable to a'ny test input must not diverge.

Level 2 A scram shall not occut due to Recirculation flow control maneuvers. The APRM neutron flux trip avoidanca margin shall be >7.5% when the power maneuver effects are extrapolated to those that would occur along the 100% rated rod line.

The decay ratio of any oscillatory controlled variable must be <0.25.

Steady-state. limit cycles (if any) shall not produce turbine steam flow variations greater than +0.5% of rated steam flow.

The speed demand meter must agree with the speed meter within 6% of rated generator speed.

~

RESULTS The new fuel design did not affect the performance of systems needed to satisfy the acceptance criteria of this test.

>. 2 , _ . . _. _ _ .

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!2.27 -STP-30,JRECIRCULATION SYSTEM:

OBJECTIVES The objectives of this test are to

Obtain recirculation system performance data during steady-state conditions, pump trip, flow coastdown, and pump restart.

Verify that the feedwater control' system can satisfactorily control water level on a single recirculation pump trip without a resulting turbine trip and associated scram.

Record and verify. acceptable performance of the circuit off a two-recirculation pump trip.-

- Verify the adequacy'of the recirculation runback to avoid a scram upon simule.ted loss of or.e feedwater pump.

Verify that no recirculation system cavitation will occur in the operable region of the power-flow map.

ACCEPTANCE CRITERIA Level 1 The reactor shall not scram during the one pump trip recovery.

The recirculation pump and motor time-constant of the two-pump drive fivw coastdown transient should-be <2.5 seconds-from 1/4 to 2 seconds after the pump's are tripped and 13.0 seconde from 1/4 to 3 seconds after the pumps are tripped.

Level 2 The reactor water level margin to avoid a high level trip-shall be 13.0 inches during the one pump trip.

The APRM margin to avoid a scram shall tx! >7.5% during the pump. trip recovery.

The core flow shortfall shall not exceed 5% at rated-power.

The measured care-delta P shall not be >0.6 PSI above prediction.

The calculated jet pume M ratio shall not be less than, 0.2 points below prediction.

, - ,., , . - ,- 4

A _. JA

.h g-Tho; drive" flow" shortfall chhll.not'excoed 50 at rated-

. power.--

t

'The measured recirt,ulation-pump' efficiency shall'not be >8%

. points-below the vendor tested efficiency.-

The~ nozzle and riser plugging criteria shall not be exceeded. ,

The recirculation pumps shall runback upon a trip of the runback circuit. ,

' Runback logic shall'have settings ad' equate to prevent recirculation pump operation in areas of potential' cavitation.

RESULTS

. ss

-The new fue.1' design did not affect the performance-of systems needed to satisfy the acceptance criteria of this test. .

1 t

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!2.28 STP-31,fLOSSLOF TURBZNE GENERATORLAND'OFFSITE-POWER' OBJECTIVES This; test-determines electrical equipment and reactor system transient performance during a loss of main turbine-generator coincident with loss of all sources-of

, offsite power. .

ACCEPTANCE CRITERIA a Level 1 All safety systems, such as the Recctor Protection system, the diesel generators, and HPCI must funct. ion properly-without_ manual assistance, and HPCI and/or RCIC system-action, if necessary, shal: keep the reactor water level above the-initiating' level of Low Pressure Core Spray, LPCI, Automatic Depressurization System,-and MSIV Closure.

Diesel Generators shall start automatically.

Level 2 Proper inst"1 mentation display to the reactor operator shall be demonstrated, including power monitors, pressure, water level, control rod position, suppression pool temperatures, and reactor. cooling system status. Displays.

shall not be dependent on specially installed instrumentation.

Reactor pressure shall not exceed 1250 psig, if safety / relief valves open, the temperature measured-by thermocouples on the discharge-side of the safety / relief valves must return to within 10 degrees F of the temperature recorded before the valve was opened.

Normal cooling systems shall be capable of maintaining adequate drywell cooling and adequate suppression pool water temperature.

RESULTS The new fuel design did not affect the performance of systems needed to satisfy the acceptance criteria of this.

test.

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2.29 STP-33, ESSENTIAL HVAC SYSTEM OPERATION AND CONTAINMENT HOT PENETRATION

  • EMPERATURE VERIFICATION OBJECTIVES The objectives of this test are to demonstrate, under actual / normal operating conditions, that the various HVAC systems will be capable of maintaining specified ambient temperatures and relative humidity within the following areas:

a) Primary Cortainment (drywell and suppression ch' amber) b) Reactor Enclosure and Main Steam Tunnel c) Control Room d) Control Enclosure e) Radwaste Enclosure In addition, this test shall verify that the concrete temperature surrounding Main Steam and Feedwater containment penetrations remains within specified limits.

ACCEPTANCE CRITERIA Level 1 The drywell area volumetric average air temperature is not to exceed 135 degrees F.

Level 2 The drywell area and suppression chamber are maintained ,

betw(: . 65 degrees F and 150 degrees F. -

The reactor pressure vessel (RPV) support skirt surrounding air temperature is maintained above a minimum of 70 degrees F.

The concrete temperatures surrounding primary containment Main Steam line and Feedwater line penetrations are maintained at less than or equal to 200 degrees F.

All areas listed in Subtest 32.3 of the centrol enclosure are maintained between 65 degrees F and 104,Jegrees j F except the battery rcoms, which are mainta.ined at 88 degrees maximum (at float charge rate) and^the auxiliary equipment room, which is maintained between 74 degrees F and 78 degrees F and relative humidity between 45% R.H. and 55% R.H.

l

Tha Control Room 10 maintaincdLot.a tcmparcture b0tw:On 74-degrees F and 78 degrees F and relative humidity between 45% R.H. and 55% R.H.

The following areas of the Reactor Enclosure are maintained between 65 degrees F and 104 degrees F: rooms 111, 118, 200, 207, 210, 304, 402, 406, 500, 506A, 506B, 506C, 506D, 507, 508, 509, 511, 519, 601, 602,-605, 612, and 618.

The following areas of the Reactor Enclosure are maintained-between 65 degrees F and 110 degrees F: rooms 502, 503, 504, and 505. ,

The following areas of the Reactor Enclosure are maintained between 65 degrees F and 115 degrees F: rooms 102, 103, 203, 204, 108, 109, 110, 113, 114, 117, 288, 289, 501, 510, 512, 523, and 599.

The following areas of the Reactor Enclosure are maintained between 65 degrees F and 120 degrees F: rooms 209, 306, 307, 309, 407, and 518.

The following areas of the Radwaste Enclosure are maintained between 65 degrees F and 76 degrees F: rooms s 410, 411, 412, 415, 417 and 418.

RESULTS The new fuel design did not affect the performance of systems needed to satisfy the acceptance criteria of this test.

l 3.30 STP-33, PIPING STEADY STATE VIBRATZON

OBJECTIVE The objective of this test is to verify that the steady state vibration of Main Steam, Reactor Recirculation and selected BOP piping systems is within acceptable limits.

ACCEPTANCE CRITERIA Level 1 Operating Vibration: The measured amplitude (peak to peak) of each remotely monitored point shall not exceed the allowable value of that point.

Level 2 Operating Vibration: The measured amplitude (peak to peak) of each remotely monitored point shall not exceed the expected value of that point.

The steady state vibrations of visually examined balance of plant piping are acceptable if the vibration levels are judged by a qualified test engineer to be negligible.

Vibration levels judged to be potentially significant are evaluated as determined necessary by BPC Project Engineering.

The vibration measured by a remote accelerometer is acceptable if the acceleration frequency spectrum falls in the negligible region of the acceptance chart of that accelerometer. If the acceleration frequency spectrum crosses the negligible region boundary, the test results shall be evaluated by BPC Project Engineering.

RESULTS The new fuel design did not affect the performance of systems to aatisfy the acceptance criteria of this test.

l 2.31 STP-34~, OFFGAS PERFORMANCE VERIFICATION OBJECTIVES 1 The objectives of this test are to-verify that the Offgas Recombination and' Ambient Charcoal System operates within the technical specification limits and expected operating conditions. ,

ACCEPTANCE CRITERIA e Level 1 The allowable dose and dose rates frcm releases of radioactive' gaseous and particulate effluents to areas at and beyond the SITE BOUNDARY.shall not be exceeded.

Allowable limits on the radioactivity release-rates c2 the H six noble gases measured at the after condenser discharge shall not be exceeded.

The hydrogen content of the offgas effluent downsteam of the recombiner shall be equal to or less than 4% by volume.

The total flow rate of dilution steam plus offgas when the steam jet air ejectors are in operation shall exceed 9555 lbs/hr.

Level 2 System flows, pressures, temperatures and dewpoint shall be within e.xpected performance values.

The Preheater, catalytic Recombiner, After condenser, Hydrogen Analyzers, Cooler Condenser,' Activated Charcoal Beds and the HEPA filter snall be performing their required functions adequately. The automatic drain systems function i adequately.

TEST RESULTS The new fuel design did not affect the performance of systems needed to satisfy the acceptance criteria of this i test.

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I 2.32 STP-35, RECIRCULATION SYSTEM FLOW CALIBRATION

.. i-OBJECTIVES The objectives'ofzthis test are'to_petform a complete

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calibration of the r circulation-system flow instrumentation, including specific signals-to the-plant-process' computer and to adjust the recirculation flow control system to limit maximum core flow to-1090'of: rated.

,c core flow.

.- ACCEPTANCE CRITERIA Level 1 None Level 2 Jet pump flow instrumentation shall be adjusted such that the? jet pump total-flow recorder will provide correct core flow indication at rated conditions.

The APRM/RBM flow blac. instrumentation shall be adjusted.to function properly at rated conditions.

The flow control system shall be adjusted to limit maximum core flow'to 109% of rated.

RESULTS

.The new fuel design did not affect the performance of systems needed to satisfy the acceptance criteria of~.this test.

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2.33: STP-36,1 PIPING DYNAMIC TRANSIENTS  ;

OBJECTIVES The objectives of this test are to verify that the following pipe systems are adequately designed and-restrained to withstand the following respective transient loading conditions: ,

Main Steam - Main Turbine Stop Valve / Control Valve closures at approximately 20-25%, 60-80%, and 95-100% of rated o thermal power. ,

Main Steam a"d 9911ef Valve Discharge - Main Steam Relief Valve actuat3 '

Recirculation - au-irculation Pump trips and resta.its.

High Pressure Coolant Injection steam supply - High Pressure _ Coolant Injection turbine trips.

Feodwater - Reactor-feed pump trips /coastdowns.

a ACCEPTANCE CRITERIA Level 1 Operating Transients: The measured amplitude (peak to' peak) of each remotely monitored point shall not exceed the allowable value of that paint.

Level 2 Operating. Transients: The measured amplitude (peak to peak) of each remotely monitored point shall not exceed!the expected value of that point.

The maximum measured loads, displacements, and/or velocities are less than or equal to the acceptance limits specified.

In'the judgment of the qualified test engineers, no signs of excessive piping response (such as damaged insulation; mackings on. piping, structural or hanger-steel, or. walls; damaged pipe supports; etc.) are found during a 4

post-transient walkdown and visual inspection of'the piping- 'F' tested and associated branch lines.

RESULTS The new fuel design did not affect the performance of ,

systems needed to satisfy the-acceptance criteria of-this

test.

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2a34 STP-37,_ MAIN STEAM-SYSTEM AND-TURBINE PERFORMANCE.

AND PLANT DYNAMIC RESPONSE VERIFICATION-

' OBJECTIVES TheLtest objectives are to demonstrate (1) the satisfactory performance of the main steam _ system and the main turbine, and (2) that the dynamic response of.the, plant'to_the design load swings, i.' c ul ding step and-ramp change's, is in accordance with design.

- TEST METHOD ,

Reactor power is brought-to 25, 50, 75, and 100% percent to verify operability and design performance requirements of the main steam system and main turbine. Design step and ramp load changes are induced at each power level to verify plant dynami response.

ACCEPTANCE CRITIERIA-The main steam system operates properly at the specified power levels. -The main turbine operates within specified limits throughout the full power range. The dynamic.

response of the plant to design load swings is.within specified limits.

RESULTS The new fuel design did not affect the performance of systems needed to satisfy the acceptance criteria of this test.

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2.35- STPi"0',: REACTOR WATER CLEANUP SYSTCM h

OBJECTIVES The objective of this t.est is tc demonstrate-specific

aspects of the mechanical operability of the Reactor Water

_ Cleanup _(RWCU)-System.

ACCEPTANCE CRITERIA Level 1 s ,

None Level 2 The temperature at the tube side outlet of the non-regenerative. heat exchangers shall not exceed 130 Deg F in the blowdown-node-and shall not exceed 120'Deg. F in the normal mode.

The pump available NPSH shall be 13 feet or greater during the Hot Shutdown mode as defined in tha process diagram.

The co 1.ing water supplied to the non-regenerative heat.

exchanqies shall be less than 6% above the flow corresponding tc the heat exchanger capacity-(as' determined from the process diagram) and the existing temperature differential across the heat exchangers. The outlet-temperature shall not exceed 180 Deg. F.

Pump vibration shall be-less-than or-equal to 2 mils peak-to-peak (in any direction) as measured on.the bearing '

housing,;and 2' mils peak-to peak shaft vibration as measured on the coupling end.

RESULTS The new fuel design _did-not affect-the-performance of-systems needed to satisfy the acceptance criteria-of this test.

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2,36. -STP-71,-RESIDUAL HEAT REMOVAL SYSTEM LOBJECTIVES The' objectives of_this test are to demonstrate the ability; of_the Residual Heat Removal (RHR) System to remove residual and decay. heat from the nuclear. system so that' refueling ard. nuclear servicing can be performed. ,

-Additionally, this test-will demonstrate the ability of the RHR System to remove heat from the suppression pool.

Level 1 -

The RHR System shall be capable of operating in the Suppression Pool Cooling Mode at the heat exchanger capacity specified.

The RHR System shall be capable'of operating in the Shutdown Cooling Mode at the heat exchanger capacity s specified.

Level 2 None RESULTS The new fuel design did not affect the pecformance of

.- systems needed to satisfy the acceptance criteria of this test.

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