ML20246A900

Startup Rept,Cycle 3

ML20246A900
Person / Time
Site:	Limerick
Issue date:	06/30/1989
From:	Leitch G, Mccormick M, Ruppert G PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 8908230179
Download: ML20246A900 (59)

Text

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Ts6911 3895247560 -

PH5LA6ELPHIA ELECTRIC COMPANY LIMERICK GENER ATING STATION P. O. DOX A S AN ATOG A. PENNSY LV ANI A 19464 (215) 3271100, EXT. 3000

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August 17, 1989 Docket No. 50-352 License No. NPF-39 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555

Subject:

Limerick Generating Station Unit 1 Startup Report - Cycle 3 Enclosed is the Limerick Generating Station Unit 1, Cycle 3, Startup Report. The report is being submitted in accordance with Technical Specification Reporting Requirements 6.9.1.1, and contains all pertinent information regarding the third cycle

.startup testing activities.

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

Very truly you s, b h.

G. M. A JJM/kk Attachment cc: William T. Russell, Administrator, Region I, USNRC T. J. Kenny, USNRC Senior Resident Inspector, IGS 1 8908230179 890630 PDR ADOCK 05000352 I P PDC l 9,

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June 1989 I

I PHILADELPHIA ELECTRIC COMPANY LIMERICK GENERATING STA"' ION UNIT NO. 1 STARTUP REPORT CYCLE 3 8

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I Preparation Directed by:

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M. J. McCormick, Jr, Manager I Limerick Generating Station f

Prepared by:

G. F. Ruppert, Reactor Engineering Limerick Generating Station Reviewed by:

M. L. Eyre, Reactor Engineer Limerick Generating Station iI

I l fE TABLE OF CONTENTS

.I INTRODUCTION /

SUMMARY

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,g 1.1 Report Abstract 1.2 Summary 1.3 Limerick Plant Description Table 1.3-1 Limerick 1 Plant Parameters

2. 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 Performan'e and Control Pod 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 l

2.16 STP-18, TIP Uncertainty f

i 2.17 STP-19, Core Performance I _

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TABLE OF CONTENTS 2.18 STP-20, Steam Production 2.19 STP-22, Pressure Regulator 2.20 STP-23, Feedwater System 2.21 STP-24, Turbine Valve Surveillance E 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 System 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 g 2.33 STP-36, Piping Dynamic Transients STP-37, Main Steam System and Turbine I 2.34 Performance and Plant Dynamic Response Verification 2.35 STP-70, Reactor Water Cleanup System 2.36 STP-71, Residual Heat Removal System lI I

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I INTRODUCTION /

SUMMARY

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7-i 1.1- . REPORT-ABSTRACT This Startup Report, written to comply with Technical j Specifications-paragraphs 6.9.1.1 thru 6.9.1.3, consists of

a. summary o'f.the~Startup and Power Escalation Testing performed at Unit 1 of the Limerick' Generating Statio.n.~ .

This report is required since fuel of a different. design was-installed during the second refueling outage of Unit 1.

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 1to obtain satisfactory operation.

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SUMMARY

Limerick Unit 1 was out-of-service from January 11, 1989 to The unit I Msy 19, 1989 to accommodate a refueling outage.

i returned to service on May 19, 1989 and rea,ched 80% power on May 25, 1989. Reactor Power is limited to 80% due to Due to the CRUD unsatisfactory levels of feedwater copper. failures found during fuel Induced Localized corrosion (CILC) inspections and reconstitution performed during the second refuel outage, the above restriction on power From versus past industry feedwater copper levels was developed.

experience it has been determined that the probability of CILC failures is greatly reduced when feedwater copper levels are kept below 0.2 ppb or fuel bundle power densities are l

limited to BWR 3 power densities, which corresponds to 80%

rated power at Limerick (BWR 4/5).

The successfully implemented startup program ensures that the l second refueling outage of Limerick Unit 1 has resulted in no conditions or system characteristics that diminishes the safe operation of the plant. The tests and data referenced in this report are on file at the Limerick Generating Station.

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I 1.3 LIMERICK PLANT DESCRIPTION I The Limerick. Generating Station is a two unit nuclear power plant. The two units share a common control room, refueling floor, turbine operating deck, radwaste system, and other I

auxiliary systems. ,,

The Limerick Generating Station is located on the east bank I of the Schuylkill River in Limerick Township of Montgomery County, Pennsylvania, approximately 4 river miles downriver from Pottstown, 35 river miles upriver from Philadelphia, and I 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 gross I electrical output of 1092 MWe. Approximately 37 MWe are u;ed for auxiliary power, resulting in a net electrical output of See Table 1.3-1 for Limerick Plant Parameters.

1055 MWe.

The containment for each unit is a pressure suppression type designated as Mark II. The drywell is a steel-lined concrete 5 cone located above t he steel-lined concrete cylindrical 5 pressure suppression chamber. The drywell and suppression chamber are separated by a concrete diaphragm slab which also serves to strengthen :he entire system.

I The Architect Engineer and Constructor was Bechtel Power Corporation.

The plant is owned and operated by the Philadelphia Electric Company.

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L TABLE 1.3-1 Limerick 1 Plant Parameters Parameter Value Rated Power (MWt) 3293 f

i Rated Core. Flow (Mlb/hr) 100 (1)

Reactor Dome Pressure (psia) 1020 Rated Feedwater Temperature (Deg. F) 420 (4)

Total Steam Flow (Mlb/hr) 14.159 Vessel Diameter (in) 251 i

l Total Number of Jet Pumps 20 Core Operating Strategy Modified Control Cell Core l Using Al/A2 Sequences Number.of Control Rods 185 f

Number of Fuel Bundles 764 i Fuel Type 8 x 8 (Barrier)

Core Active Fuel Length (in) 150 Cladding Thickness (in) 0.032 Channel Thickness (in) 0.100 for initial core fuel

) 0.080 for GE8B fuel (2)

MCPR Operating Limit 1.26 (3) f f

Maximum LHGR (KW/ft) 13.4 for BP/P8X8R fuel 14.4 for GE8X8EB fuel 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

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

(2) Except for LYG 644 which required a replacement 100 mil channel (3) With EOC-RPT Breakers operable, increased core flow and/or a final feedwater temp reduction of 60 degrees F for BOC to EOC

-2000 mwd / ton.

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

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2.1 STP-1, CHEMICAL AND RADIOCHEMICAL OBJECTIVES-The principal objectives of this test are a) to secure information on the chemistry and radiochemistry of the reactor coolant, and b) to determine that the sampling I equipment, procedures and analytical techniques are adequate to supply the data required to demonstrate that the chemistry of all parts of the entire reactor system I meet specifications and process requirements.

ACCEPTANCE CRITERIA 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 1 reactor, following its second refueling outage, reactor I coolant chemistry parameters as well as radioactive gaseous waste releases and radioactive liquid waste releases were maintained within the limits set forth in the LimerickThe I Generating Station Unit 1 Technical Specifications.

following is a list of Chemistry related surveillance tests satisfactorily performed in support of unit startup I activities:

ST-5-041-800-1, ST-5-041-875-1, ST-5-041-876-1, I ST-5-041-877-1, ST-5-041-878-1, ST-5-041-879-1, ST-5-041-885-1, ST-5-076-810-1, ST-5-076-815-0, ST-5-076-820-0, ST-5-076-815-1, ST-5-076-820-1, ST-5-070-885-1 I

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l I In addition to the surveillance tests, routine tests and normal analyses were pegformed. Maximum Dose Equivalent I Iodine was 4.11 x 10- uCi/g (Tech Spec 0.2 uCi/g).

Fuel Warranty Appendix I - Water Quality Requirements were met during startup.- From 5/15/89 through 5/26/89 with reactor-power greater than 0%, reactor water conductivity I averaged 0.148 umho/cm (Fuel Warranty limit 1.0) chlorides ranged from less than 2.0 to 9.2 ppb (Fuel Warranty limit I 100 ppb), and pH ranged from 6.55 to 6.90 (Fuel Warranty I I Range 5.6 - 8.6). Above 50% power, feedwater copper concentration was a maximum of 0.292 ppb, iron was 0.612 ppb and total metals were 1.357 ppb (Fuel Warranty limit 2 i

-E 5 ppb, 10 ppb, and 15 ppb respectively). The highest ]

condensate demineralized effluent conductivity above 50% l j

power was 0.59 umho/cm. 1 I Condensate and reactor water cleanup demineralized performance was monitored closely during the startup.

Demineralizers were regenerated as necessary to maintain I

l reactor water conductivity less than 0.3 umho/cm. J l

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,.I 2.2 STP-2, RADIATION MEASUREMENTS OBJECTIVES

' I The objectives of this test are to a) detr -i.ne the background radiation levels in the plant irons prior to operation for base data to assess future . tivity buildup I-.

and b) monitor radiation at selected power levels to assure the protection of personnel during plant operation.

ACCEPTANCE CRITERIA Level 1 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 I protection against radiation as outlined in 10CFR20

" Standards for Protection Against Radiation".

Level 2 None RESULTS Health Physics procedure HP-203, "HP Startup Surveillance I Procedure" was implemented during reactor startup. This procedure directs Health Physics surveillance throughout the plant to help ensure plant posting and RWP's are I updated as reactor power increases. It should be noted that Radiation Areas have been identified in general access and hallway areas which are designated as Radiation Zone II, therefore, should be less than 2.5 mr/hr. The Health I Physics department has identified the various sources causing the increased radiation levels, and will prioritize the affected areas so that plant staff can take action to I attempt to return the plant to pre-outage condition. Table 2.2-1 provides a list of these radiation sources.

Radiological postings in the plant are in place so that I personnel exposures can be controlled in accordance with 10CFR20.

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TABLE 2.2-1.

Radiation Sources Causing' Dose Rates Greater-than:2.5 mR/hr In Normal' Access Paths.or Areas Designated as Zone II HRx'313' Dose Rate (mrem /hr) System

' Source HV-1-23 300 RWCU Demin Fuel Pool HCC-108-North & South 50-600 HCC-106 South 6-20 Fuel Pool "HCC-133 North 10-40 Fuel Pool AllLFuel Pool Piping (North) 2-10 -Fuel Pool Rx 283'

-10 RHR

. GBB 105 Rx:253' Fdwtr Main Stream Entrance'to Steam' Chase 4-16

Floor / Equip. Drains 16-70 R/W Scram Discharge Tanks 40. CRD Chem Sample-Sink 2-6 RWCU 1

GBB.118 2-15 RER

.GBB,119. 40 RHR HBB'119 10-40 RHR HBC 103 100 Fuel Pool'

HCC 101 2 Fuel Pool HBC 133 2 Fuel Pool Rx 217'-

_ GBli 119 5 (Field) RHR GBB 107 2-6 RER

'GBB'108 2-6 RER HBC 103 .

60 Fuel Pool 8-26 Floor / Equip. Drains R/W l'

Rx 201' Fuel Pool-HBC 103 120 HCC-101 4 Fuel Pool HBC 133 12 Fuel Pool Floor / Equip. Drains 4-20 R/W Rx 177' Core / Spray A Core / Spray <2-38 C Core / Spray <2-36 Core / Spray B Core / Spray <2-8 Core / Spray D. Core / Spray- <2-12 Core / Spray Floor / Equip.. Drains 4-38 R/W HPCI Pump Room 6-8 R/W f=

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

ACCEPTANCE CRITERIA Leveel 1 The partially loaded core must be subcritical by at least 0.38% delta k/k with the analytically determined strongest rod fully withdrawn.

Level 2 None RESULTS

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The beginning of cyc;1e shutdown margin calculated in the Core y Cycle Management Repart 23A5802 was 1.89t delta K/K.

.cload was conducted in accordance with Technical Specifications. Equipment required to be operable to I- ensure that the shutdown margin is maintained was verified operable by various performances of ST-6-107-630-1 and ST-6-107-591-1 between March 17, 1989 and March 28, 1989.

Post alteration core verification was completed on March

-I 29, 1989 after all refueling operations were completed by the performance of ST-3-097-355-1. All fuel bundles were verified to be in their proper core locations and properly I' oriented in the control cell. The bundle seating pass identified two fuel bundles improperly seated (35-50, LYM-384 AND 33-28, LYN-132). The bundles were properly I reseated, and the location and orientation was reverified after reseating.

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2.4 STP-4, SHUTDOWN MARGIN DEMONSTRATION j 1

OBJECTIVES The purpose of this test is to demonstrate that the reactor will_be sufficiently subcritical throughout the cycle with i any single control rod fully withdrawn. .

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

Level 2 Criticality should occur within +1.0% delta K/K of the

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RESULTS A shutdown margin of 1.497% 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 May 15, 1989.

L Using the data obtained during the shutdown margin demonstration, the difference between criticality and predicted critical was -0.32% delta F/K. This was within the Level 2 acceptance criteria and documented in f ST-3-107-800-1 completed on May 17, 1989.

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2.5 STP-5, CONTROL ROD DRIVE SYSTEM 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 I characteristics of the CRD system.

ACCEPTANCE CRITERIA 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.

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 Prom Fully Withdrawn Scram Time (Seconds) 45 0.43 I 39 0.86 25 1.93 05 3.49 The mean scram time of the three fastest CRD's in a two by two array must not exceed the following times (Scram time I is measured from the time the pilot scram valve solenoids are de-energized):

Position Inserted to I From Fully Withdrawn 45 Scram Time (Seconds) 0.45 39 0.92 2.05 8 25 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 I stroke in 40 to 60 seconds.

RESULTS 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 I thermal limits such as critical power ratio are not exceeded.

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I Level'1 and Level 2 stroke time acceptance criteria were E fully satisfied by the performance of ST-6-047-760-1 on May 8, 1989 prior to and during the operation hydrostatic test.

Level 1 scram time acceptance criteria were fully satisfied by the performance of ST-3-107-790-1 on May 1 & 2, 1989 during the operational hydrostatic test. ,

Six control rods were retested per ST-3-107-790-1 on May 8, 1989 for the reasons listed below and found to satisfy level 1 scram time acceptance criteria.

CONTROL ROD REASON FOR RETEST 30-15 REPLACE CRD DUE TO UNCOUPLING PIN MISALIGNMENT.

22-35 REPLACE CRD DUE TO UNCOUFLING PIN MISALIGNMENT.

22-19 TIGHTENED PACKING ON SCRAM INLET /0UTLET VALVES.

38-19 REPAIR OF CRD O-RINGS.

38-27 REPAIR OF CRD O-RINGS.

18-35 TIGHTENED PACKING nN SCRAM INLET / OUTLET VALVES.

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I 2.6 STP-6, SRM PERFORMANCE AND CONTROL ROD SEQUENCE OBJECTIVES The objective of this test is to demonstrate that the operational neutron sources, SRM instrumentation, and rod withdrawal sequences provide adequate information to l .'

achieve criticality and increase power in a safe and efficient manner.

g ACCEPTANCE CRITERIA

!3 Level 1

' There must be a neutron signal to noise count ratio of at least 2:1 on the required operable SRMs.

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

Level 2 None RESULTS Minimum SRM count rate was determined to be greater than 3 CPS by the performance of ST-6-107-591-1 prior to the I withdrawal of control rods on May 15, 1989 in OPCON 4 and performance of ST-6 in OPCON 1 and 2 on May 15,

-107-590-1The signal-to-noise ratio 1989 through May 25, 1989.

I verification is only required to be performed in accordance 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-6-107-875-1 on May 15, 1989 until criticality was achieved.

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lI 2.7 STP-9, WATER LEVEL REFERENCE LEG TEMPERATURE OBJECTIVES The objectives of this test are to measure the level instrumentation reference leg temperature, recalibrates the water level instruments if the mea,sured temperature is significantly different from'the value assumed during the I. initial end points calibration, and to obtain baseline data on the Narrow Range and Wide Range water level instrumentation.

ACCEPTANCE CRITERIA Level 1 None Level 2 The difference between the actual reference leg I temperature (s) and the value(s) assumed during initial calibration shall be less than that amount which will result in a scale end point error of 1% of the instrument  ;

span for each range.

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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I 2.8 STP-10, IRM PERFORMANCE OBJECTIVES The objectives of this test are to adjust the Intermediate Range Monitoring (IRM) System to obtain an optimum overlap with the SRM and APRM s,ystems.

ACCEPTANCE CRITERIA Level 1 Each IRM channel must be on scale before the SRM's exceed their red block setpoint.

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

Level 2 Each IRM channel must be adjusted so that one-half decade I overlap with the SRM's is assured.

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

RESULTS, i

Technical Specification SRM/IRM overlap was satisfiedThisby the performance of ST-6-107-884-1 on May 15, 1989. I test demonstrated at least a half decade SRM/IRM overlap.

During startup, all required APRM's were verified to be on scale before any IRM exceeded their scram setpoint of 120%

of scale. This was documented on GP-2, Normal Plant I Startup, on May 17, 1989. One-half decade IRM/APRM is verified in accordance with Technical Specifications overlap {

during each controlled shutdown by the performance of I ST-6-107-886-1. i i

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I 2.9 STP-11, LPRM CALIBRATION OBJECTIVES

'The objectives of this test are to calibrate the Local Power Range Monitoring (LPRM) System and to verify LPRM Flux 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 80% power and June 13, 1989 per ST-3-074-505-1 on May 22, 1989the On June 13, 1989 LPRM's were calibrated respectively.

to within 4% of their calculated value.

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

ACC$PTANCE CRITERIA Level 1 The APRM channels must be calibrated to 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 startup mode, all APRM channels must produce a scram ,

at less than or equal to 15% of rated thermal power.

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

RESULTS By various performances of ST-6-107-885-1 from May 22, 1989 I to May 25, 1989, Level 1 acceptance criteria was met by verifying APRM Cannels were indicating greater than or equal to actual core thrrmal power and below the scram and

- rod block setpoints wbr:n thermal power was greater than 25%. Level 2 acceptance criteria was also met in this surveillr..cen tat by adjusting indicated APRM reading to

+2, -0% (not to exceed 100%) of the greater of I

within fraction of rated power or maximum fraction limiting power density.

The Level 1 acceptance criteria of APRM scram setpoint of 15% was met by performance of channel functional 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 I. performed on May 5 through May 17, 1989.

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2.11 STP-13, PROCESS COMPUTER OBJECTIVES I -The objective'of this test is to verify the performance of the Process Computer under plant operating conditions.

ACCEPTANCE CRITERIA I-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 l

- 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 within 2%.

.The maximum LHGR calculated by BUCLE and the Process Computer either:

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

- in the case in which the maximum LHGR calculated by the Process Computer is in a different assembly than that calculated by BUCLE, of each assembly, the m/." mum LHGR I and the LHGR calculated by the two methods sc.'.1 agree within 2%.

The MAPLHGR calculated by BUCLE ar.d the Process Computer either:

are in the same fuel assembly and do not differ in value I -

by more than 2%, or in the case in which the MAPLEGR calculated by the I

Process Computer is in a different assembly than that calculated by BUCLE, of each assembly, the MAPLEGR 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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I RESULTS l

On June 13, 1989 at 80% core thermal power, the accuracy of the thermal limits and LPRM gain adjustment factor I calculated by the Process Computer 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 I all cases.

data.

Table 2.11-1 Eummarizes the thermal limits Also, all LPRM gain adjustment factors calculated by BUCLE and the Process Computer for operable LPRM's were

-determined to be within 2%.

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i-TABLE 2.11-1 LGS 1 BOC 3 80% Power P1 to BUCLE Comparison  ;

lI I I Value P1 Data 6-12-87, 1200 BUCLE Data 6-12-87, 1200 ,

CMWT 2636 2636 MFLPD Location _ Value P1 BUCL.; 1 i

11-18-4 0.723 0.722 0.723 l I 31-20-5 49-18-4 17-34-4 0.723 0.723 0.718 0.722 0.717 0.795 0.795 I

25-34-4 43-28-4 0.718 0.717 11-44-4 0.723 0.722 i 31-42-4 0.723 0.723 -

49-44-4 0.723 0.722 MFLCPR Location Value BUCLE P1-19-22 0 791 0.792 25-20 0.813 0.813 41-22 0.791 0.792 34 0.804 0.804 I 25-34 43-28 19-40 0.867 0.804 0.792 0.868 0.804 0.792 25 0.813 0.813 3

g 41-40 0.792 0.792 MAPRAT Location Value P1 BUCLE 0.722 I 11-18-4 31-20-11 49-18-4 6 722 0.723 0.722 0.724 0.722 17-34-4 0.717 0.716 I 25-34-4 43-21-4 0.795 0.717 0.722 0.794 0.716 0.722 11-44-4 I 31-42-11 48-44-4 0.723 0.722 0.724 0.722 I

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I 2.12 STP-14, RCIC SYSTEM OBJECTIVES The objectives of this test are to verify the proper I operation of the Reactor Core Isolation Cooling (RCIC)

System over its expected operating pressure and flow ranges, and to demonstrate reliability in automatic I starting from c.old standby when the reactor is at power conditions.

ACCEPTANCE CRITERIA Leve] 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 I manual start tests.

Level 2 In order to provide an overspeed and isolation trip avoidance margin, the transient start first and subsequent speed peaks shall not exceed 5% above the rated RCIC I turbine speed.

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

The turbine gland seal condenser system siell be capable of I preventing steam leakaw to the atmosphere.

The delta P switches of the RCIC steam supply line high flow isolation trip shall be calibrated to actuate at the 5 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 i.he acceptance criterin of this test.

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I l 3 2.13 STP-15, HFCI SYSTEM l g OBJECTIVES I The objectives of this test are to verify the proper operation of the High Pressure Coolant Injection (HPCI)

System over its expected operating pre tsure and flow 5 ranges, and to demonstrate reliability in automatic 3 starting..from cold standby when the reactor is at rated pressure conditions. l ACCEPTANCE 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 1 I psig and rated. ,

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

Level 2 3

l In order to provide an overspeed isolation trip margin, the transient first peak shall not come closer than 15% (of  ;

i g rated speed) to the overspeed trip, and subsequent speed g peaks shall not be greater than 5% above the rated turbine speed. ]

l I The speed and flow control loops shall be adjusted so that j the decay ratio of any HPCI system related variable is not i greater than 0.25. l The turbine gland seal 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 tett.

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g 2.14 STP-16, SELECTED PROCESS 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 modes that cause I temperature stratification, (3) to determine the proper setting of the low flow control limiter of the 3 recirculation pumps to avoid coolant temperature stratification in the reactor pressure vessel bottom head 5 region.

ACCEPTANCE CRITERIA Level 1 The reactor recirculation purps shall not be started, flow increased, nor power increased unless the coolant temperatures between the steam dome and bottom head drain I 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 I increased unless the idle loop suction temperature is *re within 50 degrees F of the active loop suction tempt and the active loop flow rate is less than or equa) 30%

I of rated loop flow. If two pumps are idle, the l' suction temperature must be within 50 degrees F o.

steam dome temperature before pump startup.

'I Level 2 During two pump operation at rated core flow, the bottom i head temperature, 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 I systems needed to satisfy the acceptance criteria of this test.

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).. -4 2.15 STP-17, SYSTEM EXPANSION OBJECTIVES This-test verifies that safety related piping systems and other piping systems as identified in the FSAR 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 spring hangers remain within operating travel ranges at various piping temperatures.

I. Piping is free to expand without interferences.

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 I shall not exceed the expected values.

Snubbers and spring hangers do not become extended or compressed beyond allowable travel limits (working range)

I and snubbers retain swing clearance.

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

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

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

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I; 2.16 STP-18, TIP UNCERTAINTY OBJECTIVES The objective of this test is to determine the I-< reproducibility of the Traversing Incore Probe system readings.

- ACCEPTANCE CRITERIA Level 1 None Level 2 The total TIP uncertainty (including random noise and geometrical uncertainties) obtained by averaging the I uncertainties of all data sets shall be less than 8.7%.

RESULTS Total TIP uncertainty was determined by the performance of RT-3-074-850-0 on June 20, 1989. Level 2 acceptance criteria was met by all data sets with a maximum uncertainty of 4.04%.

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lL 2.17 STP-19, CORE PERFORMANCE OBJECTIVES The objectives of this test are to:

a) Eva _te the core thermal power and core flow rate; and b) Evaluate whether the following core performance parameters are within limits:

- ' Maximum Linear Heat Generation Rate (MLEGR),

- Minimum Critical Power Ratio (MCPR), (

- Maximum Average Planar Linear Heat Generation Rate (MAPLHGR).

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

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

The Maximum Average Linear Heat Generation Rate (MAPLEGR) shall not exceed the limits specified by the r. ant Technical Specifications.

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

Core flow shall not exceed its licensed value (105 Mlb/hr).

Level 2 4 8 None l

RESULTS 4

With thermal power limited to 3293 MWth and core flow i limited to 105 Mlb/hr per GP-5 Power Operation, Level 1 j I acceptance criteria were met. Level 1 acceptance criteria for thermal limits and steady-state reactor power were met j

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j and documented throughout the startup by various I performances of ST-6-107-885-1 from May 22, 1989 to May 25, 1989.

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

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

The NSSS shall be capable of supplying 14,159,000 pounds I 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 I feedwater temperature of 420 degrees F 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 tnis test.-

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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 transients induced in the reactor pressure control system by means of the pressure regulators set point changes.

To demonstrate the take-over capability of the back-up

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pressure regulator upon failure of the controlling pressure regulator, and to set spacing between the setpoints at an appropriate value.

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

To demonstrate the stability of the reactivity-void

'I feedback loop to pressure perturbations in conjunction with STP-21, Core Power Void-Mode Response.

ACCEPTANCE CRITERIA .

Level 1 I 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 I oscillatory modes of response. In these cases, the decay ratio of eacn controlled mode of response must be less than (This criterion does not apply to tests or equal to 0.25.

involving simulated failure of one regulator with the backup regulator taking over.)

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

-.10 seconds.

Pressure 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 steam flow.

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

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The variation in incremental regulation (ratio of the E maximum to the minimum value of the quantity, " incremental

.3 change in pressure control signal / incremental change in steam flow", of each flow range) uhall meet the following: l

I  % of Steam Flow Obtained With Valves Wide Open Variation .

O to 85% $4:1 85% to 97% $2:1 97% to 99% $5:1 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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2.20 STP-23, FEEDWATER SYSTEM OBJECTIVES The objectives of this test are:

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

To demonstrate 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 the maximum feedwater runout capability is compatible 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 I <100 deg. F. The resultant Tuel thermal safety limit.

MCPR must be greater than the I The increase in simulated heat flux cannot The predicted Level 2 value by more than 2t.

exceed the predicted value will be based on the actual test values of feedwater temperature changes and initial power level.

Maximum speed attained shall not exceed the speeds which will give the following flows with the normal complement of pumps operating.

a. 135% NBR at 1075 psia

b. 146% NBR at 1020 psia .

Level 2 Level 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 I or equal to 0.25.

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

a. Maximum time to 10% of a step disturbance $1.1 see

b. Maximum time of 10% to 90% of a '

step disturbance 11.9 sec

c. Peak overshoot (% of step disturbance) 115%

d. Settling time, 100% +51, 514 see The average rate of response of the feedwater actuator to large (>20% of pump flow) step disturbances shall be I 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 I the 10 percent and 90 percent response points.

As steady-state generation of the 3/1 element 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 +1 inch.

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

.I The reactor shall avoid low water level scram by three inches 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 - 115% NBR at 1075 psia

b. One feedwater pump tripped conditions - 68% NBR at 1025 psia.

i 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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2.21 STP-24, TURBINE VALVE SURVEILLANCE OBJECTIVES l The~ objectives of this test are to demonstrate acceptable

.W 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 I 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.

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p 2.23 STP-25, MAIN STEAM ISOLATION VALVES I OBJECTIVES The objectives'of this test are to functionally check the I Main Steam Isolation Valves (MSIV's) of proper operation at selected power levels, to determine the MSIV closure times, and to determine the maximum power level at which full I

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closure of a single MSIV can be performed without causing a ..

reactor scram.

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

ACCEPTANCE CRITERIA Level J MSIV MSIV stroke time shall be no faster than 3.0 seconds.

closure time shall be no slower than 5.0 seconds.

The positive change in vessel dome pressure occurring within 30 seconds after closure of all MSIV's must The not exceed the Level 2 criteria by more than 25 psi.

I positive change in simulated heat flux shall not exceed the Level 2 criteria by more than 2% of rated value.

I Feedwater control system settings must prevent flooding of the steam lines.

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

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

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

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on the discharge i The temperature measured by thermocouple l side of the safety / relief valves must return to within 10 l degree F of the temperature recorded before the valve was I opened.

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I 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 referenced to actual test conditions of initial power level and dome pressure and I will use beginning of life nuclear data. ,

If water level' reaches the reactor vessel low water level I (Level 2) setpoint, RCIC and HPCI shall automatically initiate and reach rated system flow.

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

I 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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. 2.23 STP-26, RELIEF VALVES

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OBJECTIVES l 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 reseat properly ,

after actuation, c) to verify that there are no major to  ;

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blockages in the Relief Valve discharge piping, and d) demonstrate system stability to Relief Valve operation.

ACCEPTANCE CRITERIA 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 FSAR accident analysis at normal operating Reactor pressure.

Level 2 Pressure control system-related variables may contain I 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.

I on the The temperature measured by the thermocouple discharge side of the valves shall retarn to within 10 DEG F of the temperature recorded before the valve was opened.

During the low pressure functional test, the steam flow through each Relief Valve, as measured by Bypass Valve I 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 measured by Generator Gross MWe, shall not be lower than the average valve response by more than 0.5% of rated MWe.

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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I 2.24 STP-27, MAIN TURBINE TRIP OBJECTIVES _

The objectives of this test are to demonstrate the response I- of the Reactor and its control systems to protective trips of the Main Turbine and to evaluate the response of the

• bypass and safety / relief valves.

ACCEPTANCE CRITERIA Level 1 For Turbine and Generator Trips at power levels greater than 50% Nuclear Boiler Rated, there should be a delay of I less than 0.1 seconds following the beginning of Control or Stop Valve closure before the beginning of Bypass Valve opening. The Bypass Valves should be opened to a point I corresponIiog to greater than or equal to 80% of their capacity within 0.3 seconds from the beginning of Control or Stop Valve closure motion.

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

The positive change in vessel dome pressure occurring 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 Flux shall not exceed the Level 2 criteria by more than 2% of Rated Value.

The recirculation 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 the start of the Turbine Stop Valve or Control Valve motion to the complete suppression of the electrical arc between the fully open contacts of I the RPT circuit breakers shall be less than or equal to 175 milliseconds.

Level 2 There shall be no MSIV closure during the first three I minutes of the transient and operator action shall not be required during that period to avoid the MSIV closure.

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g The positive change in vessel dome pressure occurring g within the first 30 seconds after the initiation of either Generator or Turbine Trip must not exceed predicted values.

The positive change in simulated Heat Flux occurring within the first 30 seconds after the initiation of either Generator or Turbine Trip must not exceed predicted values.

Feedwater level control shall avoid loss of feedwater flow due to a high (LB) water level trip during the event.

Low (L2) water level recirculation pump trip, HPCI and RCIC shall not be initiated.

on the discharge I The temperature measured by thermocouple side of the Relief Valves must return to within 10 Degree F of the temperature recorded before the valve was opened.

For the Turbine Trip within the Bypass Valves 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 perfc<rmance of I systems needed to satisfy the acceptance criteria of this test.

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2.25 STP-28, SHUTDOWN FROM OUTSIDE THE CONTROL ROOM

" OBJECTIVES The objectives of this test are to demonstrate that the

'I 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 safsly cooled from I hot to cold shutdown from outside the Control Room. In addition, it will provide an opportunity to demonstrate that the procedures of Remote Shutdown are clear and I comprehensive and that operational personnel are familiar 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 under control, and Reactor vessel pressure and water I level are controlled using equipment and controls located outside the Control Room.

The Reactor can be safely shutdown to a Hot Standby condition from outside the Control Room using the minimum shift erew 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 I permit operation of the Shutdown Cooling Mode of the Residual Heat Removal System.

The Shutdown Cooling Mode of the Residual Heat Removal I System can be initiated from outside the Control Room with a heat transfer path established 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 I Limit.

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

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I STP-29, RECIRCULATION FLOW CONTROL SYSTEM 2.26 I. OBJECTIVES The objectives of this test are to demonstrate the flow E control capability of the plant over the entire pump speed 5

range, in both Individual Local Manual and Combined Master

' Manual operation modes and to determine that the I controllers are set of the desired system performance and stability.

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

Level 2 A scram shall not occur due to Recirculation flow control The APRM neutron flux trip avoidance margin I maneuvers.

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.

I 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 mater within 6% of rated generator speed.

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

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"I STP-30,-RECIRCULATION SYSTEM 2.27 OBJECTIVES

'The-objectives of this test are to:

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.Obtain recirculation system performance data during steady-state conditions, pump' trip,-flow coastdown, and I pump restart, Verify that the feedwater control system can satisfactorily control water level on a single recirculation pump trip I without a resulting turbine trip and associated scram.

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

Verify the adegaacy of the recirculation runback to avoid a scram upon simulated loss of one 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 flow coastdown transient should be <2.5 seconds I from 1/4 to 2 seconds after the pumps are tripped and 33.0 seconds 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 be 17.5% during the pump trip recovery.

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

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

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

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The drive flow shortfall shall not exceed 5% at rated I power.

The measured recirculation pump efficiency shall not be >8%

points below the vendor tested efficiency.

The nozzle and riser pluggi,ng criteria shall not be exceeded.

I The recirculation pumps ohall runback upon a trip of the runback circuit.

I Runback logic shall have settings adequate to prevent recirculation pump operation in areas of potential cavitation.

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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2.28 STP-31, LOSS OF TURBINE GENERATOR AND OFFSITE POWER

. I.: OBJECTIVES This test determines electrical equipment and reactor I- system transient performance during a loss of main turbine-generator coincident with loss of all sources of l.

offsite power.

ACCEPTANCE CRITERIA Level 1 All safety systems, such as the Reactor Protection system, I the diesel-generators, and HPCI must function properly without manual assistance, and HPCI and/or RCIC system action, if necessary, shall keep the reactor water level above the initiating level of Low Pressure Core Spray, I LPCI, Automatic Depressurization System, and MSIV Closure.

Diesel generators shall start automatically.

Level 2 Proper instrumentation display to the reactor operator shall be demonstrated, including power monitors, pressure, I water level, control rod position, suppression pool temperatures, and reactor cooling system status. Displays shall not be dependent on specially installed I instrumentation.

Reactor pressure shall not exceed 1250 psig.

If safety / relief valves open, the temperature measured by thermocouple on the discharge side of the safety / relief valves must return to within 10 degrees F of the I temperature recorded before the valve was opened.

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

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 STP-32, ESSENTIAL HVAC SYSTEM OPERATION l

2.29 AND CONTAINMENT HOT PENETRATION I. TEMPERATURE VERIFICATION OBJECTIVES The objectives of this test are to demonstrate, under actual / norma'l operating conditions, that the various HVAC

'I systems will be capable of maintaining specified ambient, temperatures and relative humidity within the following areas:

a) Primary Containment (drywell and suppression chamber) 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 between 65 degrees F and 150 degrees F.

The reactor pressure vessel (RPV) support skirt surrounding I 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 control enclosure are maintained between 65 degrees F and 104 degrees F except the battery rooms, which are maintained at 88 I 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 I

1 55% R.H. t I- ,.

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'The. Control Room is maintained at a temperature b3twoon 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 daintained between 65 degrees F and 110 degrees F: rooms 502, 503, 504, and 505.

The following areas of the Reactor Enclosure are maintained rooms 102, 103, between 65 degrees-F and 115 degrees F:

203, 204, 108, 109, 110, 113, 114, 117, 288, 289, 501, 510, 522, 523, and 599.

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

307, 309, 407, and 518.

The following areas of the Radwaste Enclosure are maintained between 65 degrees F and 76 degrees F: rooms 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.

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I 2.30 STP-33, PIPING STEADY STATE VIBRATION OBJECTIVE The objective of this test is to verify that the steady I- state vibration of Main Steam, Reactor Recirculation and selected BOP piping systems is within acceptable limits.

ACCEPTANCE CRITERIA ,,

Level 1

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Operating Vibration: The measured amplitude (peak to peak) of each remotely monitored point shall not exceed the

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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.

I 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.

I 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 I 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 satisfy the acceptance criteria of this test.

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g 2.31 STP-34, OFFGAS PERFORMANCE VERIFICATION OBJECTIVES The objectives of this test are to verify that the Offgas I Recombination and Ambient Charcoal System operates within the technical specification limits and expected operating 3

conditions.

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ACCEPTANCE CRITERIA Level 1 The allowable dose and dose rates from releases of radioactive gaseous and particulate effluents to areas at

E and beyond the SITE BOUNDARY shall not be exceeded.

5 Allowable limits on the radioactivity release rates of the I 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 I steam jet air ejectors are in operation shall exceed 9555 lbs/hr.

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

The preheater, catalytic recombiner, after condenser, Hydrogen Analyzers, cooler condenser, activated charcoal I beds and the HEPA filter shall be performing their required functions adequately. The automatic drain systems function adequately.

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

systems needed to satisfy the acceptance criteria of this test.

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l 2.32 STP-35, RECIRCULATION SYSTEM FLOW CALIBRATION og OBJECTIVES

..e The objectives of this test are to perform a complete  :

I. calibration of the recirculation system flow instrumentation, incidi .; rpecific signals to the plant I

1 process computer and te adjust the recirculation flow control system to limit.. maximum core flow to 109% of rated

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core flow.

ACCEPTANCE CRITERIA Level 1 None Level 2 I 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 bias instrumentation shall be adjusted to ;

function properly at rated conditions.

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

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

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I 2.33 STP-36, PIPING DYNAMIC TRANSIENTS OBJECTIVES

'E The objectives of this test are to verify that the 5 following pipe systems are adequately designed and restrained to withstand the following respective transient loading conditions:

I. Main Steam 2 Main Turbine Stop Valve / Control Valve closures at approximately 20-25%, 60-80%, and 95-100% of rated thermal power.

Main Steam and Relief Valve Discharge - Main Steam Relief Valve actuation.

Recirculation - Recirculation Pump trips and restarts.

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

Feedwater - Reactor feed pump trips /coastdowns.

ACCEPTANCE CRITERIA Level 1 l

Operating Transients: The measured amplitude (peak to I peak) of each remotely monitored point shall not exceed the allowable value of that point.

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

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

In the judgment of the qualified test engineers, no signs I of excessive piping response (such as damaged insulation; markings on piping, structural or hanger steel, or walls; damaged pipe supports; etc.) are found during a I post-transient walkdown and visual inspection of the piping 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 I test.

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JI 2.34 STP 37, MAIN STEAM SYSTEM AND TURBINE PERFORMANCE AP9 PLANT DYNAMIC RESPONSE VERIFICATION OBJf.CTIVES The test objectives are to demonstrate (1) the satisfactory performance of the main steam system and the main turbine;

.m and (2) that the dynamic response of the plant to the desi.9n oad Swin9s, inc udin9 Step and ramp changes, is in B 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 I. ramp load changes are induced at each power level to verify plant dynamic response.

ACCEPTANCE CRITIERIA The main steam system operates properly at the specified I power levels. The main turbine operatesThe limits throughout the full power range.

within specified dynamic response of the plant to design load swings is within I 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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t 2.35 STP-70, REACTOR WATER CLEANUP SYSTEM l .

OBJECTIVES

[~ The objective of this test is to demonstrate specific l aspects of the mechanical operability of the Reactor Water Cleanup (RWCU) System.

..-ACCEPTANCE' CRITERIA Level 1 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 mode and shall not exceed 120 Deg. F in the

.I normal mode.

g The pump available NPSH shall be 13 feet or greater during 3 the Hot Shutdown mode as defined in the process diagram.

The cooling water supplied to the non-regenerative heat I exchangers shall be less than 6% above the flow corresponding to the heat exchanger capacity (as determined from the process diagram) and the existing temperature I . differential across the heat exchangers.

temperature shall not exceed 180 Deg. F.

The outlet Pump vibration shall be less than or equal to 2 mils

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

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2.36 STP-71, RESIDUAL HEAT REMOVAL SYSTEM OBJECTIVZS The objectives of this test are to demonstrate the ability of the Residual Heat Removal (RHR) System to remove

.I residual and decay heat from the nuclear system so that refueling and nuclear servicing can be performed.

Additionally, this test will demonstrate the ability of the

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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 I capacity specified.

The RER System shall be capable of operating in the I Shutdown Cooling Mode at the heat exchanger capacity specified.

Level 2 None 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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