W3P86-0048, Addendum 1 to Waterford Steam Electric Station Unit 3 Startup Rept

Addendum 1 to Waterford Steam Electric Station Unit 3 Startup Rept

ML20140C441
Person / Time
Site:	Waterford
Issue date:	03/14/1986
From:	Barkhurst R, Cook K, Kenning R LOUISIANA POWER & LIGHT CO.
To:	Martin R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
References
W3P86-0048, W3P86-48, NUDOCS 8603250358
Download: ML20140C441 (16)

Text

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2a Approval for Issue Addendum No. I to the Waterford Steam Electric Station Unit 3 Startup Report has been approved for issue by the members of the Plant Operations Review Committee (PORC) and the Plant Manager-Nnclear.

PORC MEMBER ,

SI ATURE DATE Maintenance Superintendent (,, bb d/.2g/g4 Operations Superintendent , dM[

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Radiation Protection Superintendent '

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Plant Quality Manager <f//[mi- g/A 7/P(.

Technical Support Superintendent /w/Ooaxpu-r

'2 7 -Vltg Assistant Pl~ ant Manager '2 / '7[W PORC Chairman bm 2.-7_7-8[p PLANT MANAGER-NUCLEAR Q l >AJA -

Pages affected by this addendum: 2a, 7, 75, 241, 348, 361, 375, 385, 392, 414-420 m,A J

8603250358 860314 PDR P

ADOCK 05000382.

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PAGE 6.4 PHYSICS TESTING 289 6.4.1 Core Performance Record 289 6.4.2 Variable Tavg 310 6.5 REACTOR COOLANT SYSTEM TESTING 318 6.5.1 RCS Calorimetric Flow Measurement 318 6.5.2 Natural Circulation Demonstration Testing 329 6.6 TRANSIENT TESTING 337 6.6.1 Remote Reactor Trip With Subsequent Remote Cooldown 337 6.6.2 Load Changes 348 6.6.3 Loss of Offsite Power Trip 350 6.6.4 80% Total Loss of Flow Test / Natural Circulation 356 6.6.5 100% Turbine Trip 373 6.7 PLANT TESTING 381 6.7.1 NSSS Plant Data Record 381 6.7.2 Transient Data Record 383 6.7.3 Biological Shield Effectiveness Survey 385 6.7.4 Power Ascension Testing Ventilation Capability -

388 6.7.5 Atmospheric Steam Dump and Turbine Bypass Valve Capacity Checks 392 6.7.6 Initial Turbine Startup 401 6.7.7 BOP Data Record 405 6.7.8 Level 2 Piping Vibration Testing 407 6.7.9 Pipe Whip Restraint Monitoring 409 l 1

6.7.10 Inspection of Mechanical Snubbers and Spring Supports 411 1 6.7.11 Chemistry 414 I 7.0 APPENDIX A - List of Acronyms. 416 l

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i 75' I

3.1 INSTRUMENTATION TESTING / CALIBRATION i

3.1.1 Intercomparison of PPS. CPC. and PMC Inouts (SIT-TP-501) l-PURPOSE:

.1 1

, The purpose of this test was to Jemonstrate that the inputs a- and appropriate outputs of the. Plant Protection System (PPS),

, the Core Protection Calculators (CPC's), and the Plant Monitoring Computer (PMC) were in satisfactory agreement with one another. Permanent plant instruments (meters and q recorders) were also intercompared.

j This test satisfied the commitments of FSAR section 14.2.12.3.4.

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METHOD:

+

1 Plant conditions were stabilized at each' of the three test plateaus -- 120 F, 345 F, and 545'F -- during the heatup.

" following initial tuel load. Data from each of the four

, sources (PPS, CPCs, PMC and' meters) were simultaneously

gathered for each.of the following parameters

1

1. RCS cold leg temperature l 2. RCS hot leg temperature

3. RCP differential pressure i 4. RCP speed ,

4

i. 5. RCS pressure ,

s. Pressurizer level #C

7.. Steam generator level

8. Steam generator pressure

9. Steam generator primary side differential' pressure

10. Reactor vessel differential pressure

, . . __, - . - , . . . ,_ . - - . , ,. , . . _ _ , ._u..__ -

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241 6.3.3 CPC/COLSS VERIFICATION (SIT-TP-717)

PURPOSE:

The CPC/COLSS Verification test was performed to verify the Core Protection Calculator (CPC) and Core Operating Limit Supervisory System (COLSS) calculations of departure from nucleate boiling ratio (DNBR) and local power density (LPD).

The test also collected input recordings used to evaluate the

. effects of process noise on the CPC system.

This test satisfied the commitments of FSAR section 14.2.12.3.27.

METHOD:

This test was performed at all test plateaus including hot zero power (HZP, 20%, 50%, 80% and 100%). The CPC DNBR and LPD calculations were verified using CEDIPS, a CE-Windsor FORTRAN simulation of the CPC software. Input to CEDIPS consists of the maximum and minimum values (observed over a l period of up to 30 seconds) of the following:

1) RCP Speed

2) Cold leg-temperature

3) Hot leg temperature

4) Pressurizer pressure

5) Upper excore signal

' 348 6.6.2 Load Changes (SIT-TP-721) -

PURPOSE:

The purpose of this test was to demonstrate that the integrat-ed plant control systems operate satisfactorily in automatic to maintain plant parameters within specific limits.

If during performance of this test plant parameters were not maintained within or restored to specific operating bands, new-setpoints were to be determined for the affected control system (s) and the test repeated to verify proper system operation.

This test satisfied in part the commitments of FSAR section 14.2.12.3.39 (see also section 6.2.5). l METHOD:

The satisfactory completion of the individual system and integrated system automatic steady state operation checkouts (see section 6.2.5) at 100% power was a prerequisite for the performance of this test.

Plant conditions were stabilized at approximately 95% power with the steam bypass control system (SBCS) the feedwater control system (FWCS), the reactor regulating system (RRS),

the pressurizer level control system (PLCS), the pressurizer pressure control system (PPCS) and the digital-electro-hydraulic (DEH) system in automatic, and the control element drive mechanism control system (CEDMCS) in manual sequential (MS) with the CEAs fully withdrawn. Tavg was maintained within 10.5(f of Tref.

361 TABLE 6.6.4.1 CESEC SINGLE VALUE ACCEPTANCE CRITERIA PARAMETERS DURING TIE FIRST 60 SECONDS FOLLOWING LOSS OF FLOW.

Max-(or Min) Acceptance Parameter Value Criteria (SVAC)

Pressurizer Pressure 2247 psia 5 2339 psia Pressurizer Level 15.1% > 10.5%

RCS Hot Leg 1 Temp. 600 F $ 608 F RCS Hot Leg 2 Temp. 600 F $ 608*F Steam Generator 1 Pressure 964 psia 5 964 psia Steam Generator 2 Pressure 964 psia 5 964 psia l

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TABLE 6.6.5.1 TURBINE TRIP SINGLE VALUE ACCEPTANCE CRITERIA Original Updated SVAC SVAC Limiting Parameter Value Value. Test Value Pressurizer Pressure (psia) 52316 52318 2271 (max.)

Pressurizer Water Level

(% indicated level) >14.4 >18.4 19.0 (min.)

RCS Hot Leg 1 Temperature 5613 5604 600 (max.)

RCS Hot Leg 2 Temperature. 5613 5604 602 (max.)

Steam Generator 1 Pressure (psia) 51044 51088 1036 (max.)

Steam Generator 2 Pressure (psia) 51044 51086 1035 (max.)

385 6.7.3 Biological Shield Effectiveness Survey (SIT-TP-715)

PURPOSE:

The purpose of this test was tc obtain baseline radiation levels in order to trend radiation level buildup with operation; to measure and document radiation levels in locations outside of the biological shield while at power; to establish the adequacy of the biological shield and to identify high-radiation zones. This test satisfied the commitments of FSAR section '.4.2.12.3.15.

METHOD:

Portable neutron and gamma survey equipment was used in performing all phases of the biological shield survey.

RESULTS:

Overall, the radiation levels in the RCB were lower than design basis estimates in the FSAR by factors ranging from 2 t o l'2. ' Maximum neutron dose rates (extrapolated to 100%

power) of 10 rem /hr were observed at the south side of the refueling cavity, with a gamma dose rate of 1.25 rem /hr.

Neutron dose rates ranged from 100 to 2100 mrem /hr on +46',

with corresponding gamma doses of 14 to 300 mrem /hr. On

+35', dose rates ranged from 20-300 mrem /hr neutron and 2-5 l mrem /hr gamma. Neutron doses of 5-540 mrem /hr were found on l

+21', with gamma doses of 2-90 mrem /hr. On -4', general area e -

l dose rates ranged from 8-100 mrem /hr for neutrons an 4-100 l

mrem /hr for gammas. Dose rates at the four blowout areas '

adj acent to the secondary shicld wall at each RCP bay were significantly higher than the rest of -4'. Neutron doses of l 350 -450 mrem /hr and gamma doses of 180-200 merm/hr were found- l at these locations at 100% power.

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. , 392 6.7.5 Atmopheric Steam Dump and Turbine Bypass Valve Capacity Checks (SIT-TP-707) 1 PURPOSE

The purpose of this test was to verify that the steam flow capacities of the two atmosphere dump valves (ADVs) and of the six turbine bypass valve (TBVs) are in accordance with design requirements and safety analysis assumptions within the WSES-3 FSAR, as described below:

1. Verify that the maximum capacity of each TBV is less than

, that assumed in the analysis of the most severe excess heat removal accident, as described in FSAR section 15.1.1.3.

2. Verify that the maximum capacity of each ADV is less than that assumed in the analysis of the inadvertent opening of an A7V accident, as described in FSAR section 15.1.1.4.

3. Verify that the minimum capacity of each ADV is greater than that assumed in the post-LOCA long-term decay heat removal analysis, as described-in FSAR section 6.3.3.4.

4. Verify that the total capacity of all six turbine bypass valves is greater than or equal to the design capacity I

specified in FSAR.section 10.3.3.

This test satisfied the requirements of FSAR section 14.2.12.3.29. l l

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414 6.7.11 Chemistry PURPOSE:

The objective of the chemistry testing was to verity proper operation of chemistry sampling and analysis equipment in addition to establishing the adequacy of procedural controls utilized to monitor and maintain chemistry and radiochemistry conditions within the Waterford-3 primary and secondary systems.

METHOD:

Chemistry and radiochemistry specifications were procedurally established from system vendor recommendations and previous industry experience. Monitoring, sampling and analysis methods were also procedurally implemented and then verified by comparison of laboratory standards to unknown test variables. Sample sets were obtained and analyzed during individual system hydro and pre-operational testing to assure that proper chemistry controls were implemented prior to system operation. Throughout the Phase III test program, normal operational chemistry procedures were utilized to establish the adequacy of the chemistry program. This included calibration and maintenance of instrumentation, implementation of sampling and analysis frequency and tech-niques and establishment of baseline chemistry / radiochemistry conditions in the primary and secondary systems.

415 -l RESULTS:

All data documenting successful completion of the chemistry monitoring and control program were logged and filed in plant records as procedurally required. These data demonstrated that:

1. Chemistry of the RCS and steam generators can be maintained within established specifications.

2. Procedurally established sampling frequencies are appropriate for monitoring and controlling system chemistry.

3. Baseline RCS and steam generator chemistry is established.

4. Procedures for sample collection and analysis are acceptable.

5. Baseline activities for the RCS are established.

6. Laboratory analysis agrees satisfactorily with process radiation monitor data.

CONCLUSIONS:

The chemistry monitoring and control program implemented in

'daterford-3 procedures is adequate to meet the required specification and commitments delineated in FSAR section 14.2.12.3.16.4.

a. t 416 l SECTION 7.0 APPENDIX A LIST OF ACRONDIS I

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. APPENDIX A Part 1 of 4 LIST OF ACRONYMS A00 anticipated operational occurrence ARI all rods inserted ARO all rods out ASI axial shape index ASME American Society of Mechanical Engineers BAMT boric acid makeup tanks BOC beginning of cycle BOL beginning of life BOP balance of plant BPPCC boundary point power correlation constant CCWS component cooling water system CE Combustion Engineering CEA control element assembly CEAC control element assembly calculator CEDM control element drive mechanism CEDMCS control element drive mechanism control system CESEC Combustion Engineering System Excursion Code CET core exit thermocouple CIAS containment isolation actuation signal CIS containment isolation system CIWA condition identification and work authorization COLSS core operating limits supervisory system CPC core protection calculator CPS counts per second CSAS containment spray actuation signal CSB core support barrel CSP condensate storage pool

. CST central standard time 1

. .. . 418-l APPENDIX A (continued)

Part 2 of 4 LIST OF ACRONYMS CVCS chemical and volume control system DEH digital electro-hydraulic DNB departure from nucleate boiling DNBR departure from nucleate boiling ratio DRC digital reactivity computer EARO essentially all rods out ECBC estimated critical boron concentration EFAS emergency feedwater actuation signal EFFD effective full power days EFWCS emergency feedwater control system EFWS emergency feedwater system EOC end of cycle ESF engineered safety feature ESEAS engineered safety feature actuation system FHB fuel handling building FLCEA full length control element assembly FSAR final safety analysis report FTC fuel temperature coefficient FWCS feedwater control system HPSI high pressure safety. injection HVAC heating, ventilation, air-conditioning HZP hot zero power ICI in-core instrumentation ILRT integrated leak rate test ITC isothermal temperature coefficient LCO limiting condition for operation LOOP loss-of-offsite-power LPD local power density LPSI low pressure safety injection LTOP low temperature overpressure protection M&TE measuring and test equipment

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419l APPENDIX A (continued)

Pa rt 3 o f 4 LIST OF ACRONYMS MG manual group MI manual individual MICDS moveable in-core detector system MS manual sequential MSIS main steam isolation signal MSIV main steam isolation valve MTC moderator temperature coefficient MWE megawatt electric MWTh megawatt thermal NRC nuclear regulatory commission NSSS nuclear steam supply system OL operati2g license 00S out of sequence PAT power ascension testing PCHFT post-core hot functional testing PDIL power dependent insertion limit PEIR project evaluation /information request PLCEA part length control element assembly PMC plant monitoring computer PMU primary make-up PORC plant operations review committee PORV power operated relief valve PPDIL pre power dependent insertion limit PPS plant protection system PWR pressurized water reactor PZR pressurizer QSPDS qualified safety parameter display system RAB reactor auxiliary building RCP reactor coolant pump RCS reactor coolant system RM refueling machine

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. . . - 420 l APPENDIX A (continued)

Part 4 of 4 LIST OF ACRONYMS RMS root mean square RPCS reactor power cutback system RPF radial peaking factor RRS reactor regulating system RSPT reed switch position transmitter RTD resistance temperature detector RV reactor vessel RWSP refueling water storage pool SAM shape annealing matrix SBCS steam bypass control system SER safety evaluation report SFHM spent fuel handling machine SFP spent fuel pool SG steam generator ,

SIAS safety injection actuation system SIS safety injection system SIT safety injection tank SONGS San Onofore Nuclear Generating Station TLOF total loss of flow UGS upper guide structure VCT volume control tank VLPMS vibration and loose parts monitoring system WSES Waterford Steam Electric Station l

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LOUISIANA 34a mAnoNoe S1m1 POWE R & LIG HT P o Box 6008

• NEW ORLEANS. LOUISIANA 70174 * (504) 366 2345 sus $dvsM March 14, 1986 W3P86-0048 A4.05 QA Robert D. Martin _=

Regional Administrator, Region IV U.S. Nuclear Regulatory Commission g %h ""j

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611 Ryan Plaza Drive, Suite 1000 ) Ijj Arlington, TX 76011

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Subject:

Waterford 3 SES _ _ _

Docket No. 50-382 License No. NPF-38

_d Startup Report Addendum No. 1

Reference:

W3P85-3218 dated October 10, 1985

Dear Mr. Martin:

Enclosed are two copies of Addendum No. I to the Waterford 3 Startup Report which was provided under the referenced letter. ihe original report was submitted pursuant to Section 6.9.1.1 in the Waterford 3 Technical Specifications (NUREG-lll7) of Appendix A to Facility Operating License No.

NPF-38. Addendum No. 1 is being submitted to correct typographical errors and to include a chemistry section in the report. These changes have been discussed with your Mr. H. Bundy, w

The enclosed copies are replacement pages which should be placed in the original Startup Report.

Very truly yours,

/(/

K.W. Cook Nuclear Support & Licensing Manager KWC:BGM:ssf Enclosure cc (w/ enclosure): NRC Director, Office of I&E (Document Control Desk) (6)

NRC Resident inspectors Office cc (w/o enclosure): G.W. Knighton, NRC-NRR J.H. Wilson, NRC-NRR D.W. Churchill W.M. Stevenson

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