ML19350B715

Forwards Updated Responses to NRC Re TMI Concerns.Zero Power Physics Testing Is Scheduled for Completion by May 1981

ML19350B715
Person / Time
Site:	McGuire, Mcguire
Issue date:	03/09/1981
From:	Parker W DUKE POWER CO.
To:	Harold Denton, Youngblood B Office of Nuclear Reactor Regulation
References
NUDOCS 8103230479
Download: ML19350B715 (62)

Text

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DUKE POWER COSIPANY g

Powra Uc!LDING N?

'Il 422 SocTu Cucucu STacET, CHARLOTTE, N. C. aea42 h;q w ww o aa a a r a. s a.

March 9, 1981 4:e3 s e c.

a.oove,1o-Mr. H. R. Centon, Director Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C.

20555 Attention:

Mr. B. J. Youngblood, Chief Licensing Projects Branch No. 1

Subject:

McGuire Nuclear Station Docket Nos. 50-369 and 50-370

Dear Mr. Denton:

Enclosed with this letter are forty copies of updated responses for the document " Duke Power Company, McGuire Nuclear Station, Response to TMI Concerns." This document was transmitted to the NRC via my letter of May 23, 1980 and updated via my letters of July 18, 1980, August 6, 1980, September 8, 1980, October 10, 1980, October 29, 1980, and February 6, 1981.

Zero Power Physics Testing at McGuire is currently scheduled to be com-pleted by mid to late May, 1981.

Please schedule your review of this document accordingly.

n V y truly yours /

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William O. Parker, J THH:scs Enclosures (40) 8103280 MS P

Mr. H. R. Denton, Director March 9, 1981 Page Two WILLIAM 0. PARKER, JR., being duly sworn, states that he is a Vice President of Duke Power Company; and he is authorized on the part of said Company to sign and file with the Nuclear Regulatory Commission this document, " Duke Power Company, McGuire Nuclear Station, Response to TMI Concerns, and that all statements and matters set forth therein are true and correct to the best of his nowledge.

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1. 1 Da 6

<.0 C7 William O. Parker, Jr.,

e President Subscribed and sworn to before me this 9th day of March, 1981.

Notary Public My Commission Expires:

September 20, 1984 l

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Mr. H. R. Denton, Director March 9, 1981 l

Page Three l

bec: With 1 Enclosure R. C. Futrell W. O. Parker, Jr.

D. L. Sweat ll S. K. Blackley S. T. Rose S. B. Hager J. W. Cox (CNS)

C. J. Wylie A. M. Segrest H. B. Tucker P. R. Herran K. S. Canady R. M. Koehler N. A. Rutherford M. L. Birch l

G. A..Copp H. T. Snead R. O. Sharpe W. A. Coley R. L. Gill P. H. Barton E. M. Geddie R. L. Weber R. E. Harris M. D. McIntosh (MNS) l J. M. McGarry G. W. Cage (MNS) j W. L. Porter W. M. Sample (MNS) l D. R. House R. S. Howard (W)

T. C. McMeekin Ozen Batum (SCS) l T. P. Harrall L. M. Mills (TVA)

D. L. Canup M. V. Williamson (PG&E)

D. B. Blackman F. A. Marian (PSE&G)

R. W. Quellette R. B. Clary (SCE&G)

J. H. Knuti (CNS)

E. S. Grecheck (VEPCO) i R. M. Glover Section File MC-801.01 l

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DUKE POWER COMPANY MCGUIRE NUCLEAR STATION Response to TMI Concerns Changes and Corrections Remove These Pages Insert These Pages Table of Contents pg. 11 02/06/81 Table of Contents pg. 11 03/09/81 I-16 02/06/81 1-16 03/09/81 II-2 07/18/80 II-2 03/09/81 II-3 02/06/81 II-3 03/09/81 II-3A 02/06/81 Carryover II-10 02/06/81 II-10 03/09/81 II-12 02/06/81 II-12 03/09/81 II-13 02/06/81 II-13 03/09/81 II-13A 02/06/81 II-13A 03/09/81 II-14 02/06/81 II-14 03/09/81 II-14A 03/09/81 II-14B 03/09/81 II-19A 02/06/81 II-19A 03/09/81 02/06/81 II-19B 03/09/81 O

II-198 II-19C 02/06/81 II-19C 03/09/81 II-19D 02/06/81 Carryover III-5 02/06/81 III-5 03/09/81 Appendix C WG System Leak Detection Weekly Surveillance Items l

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Appendix A McGuire Nuclear Station Procedures

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Station Directive 3.1.4, Conduct of Operations Station Directive 3.8.2, Station Emergency Organisation l

Station Directive 3.1.9, Relief at Duties of Plant Operation Periodic Test PT/1/A/4700/10, Shift Turnover Verification Station Directive 3.1.31, Duties, Responsibilities and Qualifications of the Shift Technical Advisor i

Charter of the Station Safety Review Group i

Appendix B Control Room Design August 15, 1980 letter from W. O. Parker to H. R. Denton Appendix C Leak Rate Procedures for Systems Containing Primary Coolant Outside Containment Periodic Test Leak Rate' Determination for NI Sy item Periodic Test Leak Rate Determination for ND System I

i Periodic Test Leak Rate Determination for NS System s_,/

Periodic Test Leak Rate Determination for NM System Periodic Test Leak Rate Determination for NB System l

Periodic Test Leak Rate Determination for NV System-Periodic Test Leak Rate Determination for FW System Periodic Test Liquid Waste System Leakage Check l

Periodic Test Waste Gas System Leak Detection Periodic Test Weekly Surveillance Items i

Appendix C NRC Requests for Additional Information June 4, 1980 letter from B. J. Youngblood to W. O. Parker June 30, 1980 letter from B. J. Youngblood to W.

O.' Parker July 2, 1980 letter from B. J. Youngblood to W. O. Parker July 23, 1980 letter from R. L. Tedesco to W. O. Parker August 25, 1980 letter from R. L. Tedesco to W. O. Parker September 17, 1980 letter from Ralph Birkel to W. O. Parker September 17, 1980 letter from B. J. Youngblood to W. O. Parker

. November 6, 1980 letter from R. L. Tedesco to W. O. Parker l

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PROCEDURES FOR VERIFYING CORRECT PERFORMANCE OF OPERATING ACTIVITIES

Reference:

Action Plan I.C.6 At McGuire Nuclear Station the designation " safety-related" is applied to all systems important to safety.

Operating and periodic test procedures that require valve movement in safety-related systems have been reviewed and revised as necessary to provide assurance that these valves are returned to their correct position.

These procedures require verification of the operability of a redundant system prior to the removal of any safety-related system from service, verification of the operability of all safety-related systems when they are returned to service, and notification of and action by the Shift Supervisor and reactor operators whenever any safety-l related system is removed from or returned to service. The removal from service of portions of safety-related systems (for example, pumps, filters, fans, etc.)

are treated in a like manner. Formal checklists are used to provide.tssurance that all valves in these safety-related systems are properly aligned.

These procedures will be further revised to require independent verification of proper valve alignuent. These revisions will be complete prior to Mode 4 operation of McGuire Unit 1.

A removal and restoration procedure governs the repositioning of valves in safety-related systems following maintenance activities or other non-normal activities which require valve movement. This procedure also governs the OS removal and restoration of portions of safety-related systems (for example, pumps, filters, fans, etc.).

A formal checklist provides assurance that all safety-related valves are properly aligned following these activities.

This procedure will be revised to require independent verification of proper valve alignment. This procedure revision will be complete prior to Mode 4 operation of McGuire Unit 1.

Notification of and action by the Shift Supervisor and reactor operators when-ever any safety-related system is removed from or returned to service is accomplished by the use of the operating and periodic test procedure checklists, red tags and the red tag logbook, white tags and the white tag logbook, out of j

service stickers, and the 1.47 oypass panel. Log entries denoting the removal and restoration are made in the Reactor Operator's Log. All of the above documents are reviewed during shift turnovers.

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RELIEF AND SAFETY VALVE TESTING

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References:

NUREG-0578 - 2.1.2 Action Plan - II.D.1 Test Program Duke Power Company is participating in and monitoring an EPRI testing program to qualify the McGuire pressurizer relief and safety valves under expected operating conditions for design basis transients and accidents.

By letter dated December 17, 1979, Mr. W. J. Cahill, Jr, then Chairman of the EPRI Safety and Analysis Task Force submitted to the NRC a " Program Plan for the Performance Verification of PWR Safety / Relief Valves and Systems." Revision 1 of the pro-gram plan was submitted to the NRC on July 8, 1980.

On December 15, 1980 Mr.

R. C. Youngdal, Chairman,EPRI Research Advisory Committee, provided the NRC with the PWR utilities response to the NUREG-0737 " clarifications" of the PWR pres-surizer relief and safety valve testing required by the NRC.

Duke Power Company concurs in this response.

The current schedule for completing this testing and providing the NRC with the test results is contained in Mr. Youngdal's letter.

The EPRI program calls for testing both the McGuire pressurizer relief and safety valves.

In addition a small number of pressurizer block valves including the McGuire block valves have been tested as part of the EPRI program. After com-pletion of the test program and evaluation of the test results Duke will submit

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test data to the NRC. This data will provide evidence that the McGuire pres-( )s surizer relief and safety valves will open and reclose under the expected McGuire flow conditions for the expected McGuire operar.ing and accident (non-ATb3) conditions. Test data submitted would include criteria for success and failure of valves tested and would permit evaluation of discharge piping and supports which are not tested directly.

The E/RI program also calls for the development of correlation codes for analyzing the effect of relief and safety valve discharge piping on valve operability.

Duke will use these codes to correlate the test loop piping with the McGuire discharge piping.

Duke Valve Testing Duke Power Company has established a full scale valve testing facility at Marshall Steam Station Unit No. 2.

This facility will be used to perform the EPRI steam flow testing of power operated relief valves (PORV). Valves identical to both the McGuire PORV and PORV block valve were tested at Marshall for steam flow at McGuire full temperature and pressure conditions.

This testing was part of a Duke program independent of the EPRI program.

Several iterations of valve modifications were performed and tested until both valves met all funcional and design requirements. These modifications were then performed on the PORV's and PORV block valves at McGuire.

II-2 03/09/81 O) i%-)

I AUXILIARY FEEDWATER INITIATION AND INDICATION g_,)

References:

NUREG-0578 - 2.1.7a and 2.1.7b Action Plan - II.E.L.2 Automatic Initiation Safety-grade automatic initiation and safety-grade emergency power for the Auxiliary Feedwater System (AFS) are features of the McGuire Nuclear Station design (reference FSAR Ch. 7 and 10).

The automatic initiation circuitry for the AFS meets the single failure criteria.

Additionally, for =ost failures which could prevent the automatic start of an individual auxiliary feedwater pump, manual initiation of the affected pu=p is available from the control room. However, should the auxiliary feedwater pump in one safety train not be available due to any single failure, the redundant safety train is available with no loss of system function.

In the final stages of plant shutdown, the main feedwater pumps must be tripped.

Thertaore, the automatic start of the motar-driven auxiliary feedwater pumps upon trip of both main feedwater pumps or steam generator low-low level must be bypassed. This bypass is accomplished manually by means of a bypass switch located in the control room. When the-bypass is instated a light is energized on the bypass control switch.

Additionally, status light indication of the by-pass is provided on the associated status light panel.

This bypass is adminis-tratively controlled by use of operating procedures.

When the plant is in the startup mode, station procedures require that the bypass of the above motor-driven auxiliary feedwater pump start signals be removed.

In addition to the station procedures, an automatic means to remove the bypass has been provided. This automatic bypass removal will be generated when the P-ll set point is reached. The P-11 set point is derived from Reactor Coolant System (RCS) pressure (% 1950 psig) and is the same signal used to unblock safety injection actuation. The P-il set point is considered the appropriate signal to automatically remove the bypass of the above motor-driven auxiliary feedwater pump start signals because the reactor is not brought critical until RCS operating te=perature and pressure conditions have been reached.

In the unlikely event that the following plant conditions exist: (1) the plant is in the startup mode and below % 1950 psig (P-ll set point), (2) the turbine -

driven auxiliary feedwater pump train is not available, (3) the bypass of the

=otor-driven auxiliary feedwater pump start signals has not been removed, and (4) either both main feedwater pumps trip on a low-low steam generator signal is generated due to a steam line rupture or feedwater line break; adequate time is available-for the reactor operator to manually initiate the auxiliary feedwater system. The low-low steam generator level alarm would alert the operator to the need to check the auxiliary feedwater pumps and assure they are running. The operator would have at least ten minutes upon receipt of this alarm to take this action. This is a conservative esti= ate of the time avail-able for operator action based upon a comparison of the stea=/ feed line break at full power with a similar accident under startup conditions; i.e. subcritical

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pressure less than 1950 psig, decay heat at =uch lower levels.

II-3 03/09/81 l

The turbine-driven auxiliary feedwater pump does not have a bypass feature.

Indication Safety grade indication of auxiliary feedwater flow to each steam generator has been provided in the McGuire control room.

Provisions for calibration and testing were incorporated into the design of this instrumentation.

Control grade flow instrumentation in the lines to each steam generator and in the suction piping to each auxiliary feedwater pump is also provided.

This control grade flow instrumentation is powered from the highly reliable battery-backed 120 VAC Auxiliary Control Power System (reference FSAR Ch. 8).

Provisions for calibration and testing are included in the design of this control grade flow instrumentation.

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Carryover

The system instrumentation permits vessel level measurement from the bottom to the top of the reactor vessel, utilizing taps off of an existing spare head penetration and a tap off of a thimble tube at the seal table. Two sets of differential pressure transmitters are provided which have differing measure-ment ranges to cover different flow behavior with and without pump operations.

The narrow range cells indicate water level when zero or one reactor coolant pump is operating. The wide range cells indicate the combined core and internals pressure drop for any combination of operating reactor coolant pumps.

The upper plenum measurement is taken by two differential pressure transmitters between the same spare head penetration, and taps off two hot legs.

To minimize containment post-accident environment effects in measurement accuracy, the system design is based upon locating the transmitters outside the containment.

Hydraulic isolators in the impulse lines provide the required double barrier protection between tite RCS and outside containment.

Reference leg temperature measurements, together with the existing RCS temperature and pressure, are utilized to automatically compensate for difference in coolant and reference leg temperature effects.

A more detailed description of this system was submitted to the NRC by Mr. W. O. Parker's letter of January 16, 1981.

Installation and functional testing is scheduled to be completed by January 1,1982.

Mr. W. O. Parker's letter of March 9, 1981 addressed the five open items con-tained on pages 22-52 of Supplement 4 to the McGuire Safety Evaluation Report and the documentation requirements for Item II.F.2 as delineated in NUREG-0737.

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ADDITIONAL ACCIDENT MONITORING INSTRUMENTATION 3

References:

NUREG-0578 - 2.1.8b Action Plan - II.F.1 Noble Gas Monitors Vent monitors for noble gases will be provided with a range adequate to cover both normal and postulated accident conditions. The presently installed noble gas monitors at McGuire cover the range of 10~7 uC1/cc to 10+3 uC1/cc. A gross gam =a detector will be added to these monitors to extend the range up to 10+5 uCi/cc.

Th'.s detector will be attached to the outside of the unit vent and shielded to minimize count rate contribution from other possible sources.

The detector will be sensitive to the 80 Kev energy range of noble gases and will have a minimum of one decade overlap with the existing noble gas monitor. If an event were to occur to cause the activity being released to be in the range of this additional detector, the noble gas monitor sample will be isolated.

This action will prevent the noble gas monitor from becoming contaminated and rendering erroneous indications when activity starts decreasing. The addi-tional detector will be installed by June 1, 1981.

Procedures for estimating noble gas release rates if the existing instrumenta-tion goes off scale have been written to cover the interim period between fuel loading and installation of the new detectors. These procedures require the

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use of portable high range survey instruments to measure the radiation levels k_ /

on lines going to the radiation monitors for the unit vents if the radiation levels are such that personnel exposure could exceed 3 rem /qtr whole body and 18 3/4 rem /qtr to the extremities in the collection of a sample. The contact dose rate (mR/hr) on the lines is used to estimate the concentration (uC1/cc) of gas in the line.

If the radiation levels do not exceed the above personnel exposure limits the procedures require collection of gas, particulate, and radioiodine samples. Silver zeolite cartridges are used for radiciodine samp-ling when noble gas interference is expected. Counting rooms in the Auxiliary Building and the Technical Training Center, located just outside the McGuire exclusion boundary, are available for sample analysis.

In addition, the whole body counting room in the Administrative Building can be made available for sample analysis.

The present radiation monitoring system provides detection of volatile and non-volatile radioactive contamination of the secondary. A condensor air ejector monitor continuously monitors gaseous activity released to the unit vent by the condensor air ejector exhaust. A steam generator sample monitor continuously monitors non-volatile activity in all steam generators.

An alarm on either of these monitors provides control room operators with an indication of steam generator tube failure. By cycling steam generator samples individually through the steam generator sample monitor control room operators can identify and if desired isolate the affected steam generator. The condensor air ejector monitor would quantify the level of radioactivity released to the environment prior to isolation of the affected steam generator.

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To quantify the level of radioactivity releassd in the event the affected steam generator is not isolated and the atmospheric steam dump valves open, Duke has a steam radiation monitoring system under design. This system will use four area radiation monitors (GA Model RD-1A) mounted in the doghouse, one near each of the four main steam lines.

Each monitor will be mounted upstream of its corresponding steam dump valves. Continuous display of the monitor readout will be provided in the control room. A strip chart recorder will also be provided. The radioactivity range covered will be 10-2 to 103 R/hr. These monitors will be calibrated every refueling outage using a 10 mci source.

In addition an ele.ctronic calibration will be performed biannually. This system is scheduled to be installed by January 1, 1982.

Open-closed indication of the atmospheric steam dump valves is provided in the i

control room. Procedures will be written to use the length of time each valve is open with the design steam flow per valve to estimate the total steam = ass released during a dump.

The containment hydrogen purge exhaust discharges through the unit vent and is monitored by L trir.ent radiation monitors.

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Containment High Range Radiation Monitors Two physically and electrically separated radiation monitors will be installed f

inside the McGuire containment.

These monitors will be supplied by General Atomics and will feature GA detector model number RD23. Each monitor will utilizeanionizatiogchambertomeasuregammaradiationandwillcoverthe 0 to 10 R/hr.

No overlapping of ranges is required. Monitor range from 10to 62 Kev is 9.8X10-12 Amps / Rad /hr and the sensivity to 52 Kev sensitivit[2 is 9.0X10-Amps / Rad /hr. Seismic qualification of the monitor is in accord-

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ance with IEEE344-1975 and environmental qualification is per IEEE323-1971.

s One monitor will be powered from the Train A vital instrument bus, and the other monitor will be powered from the Train B vital instrument bus. Analog meters (one per train) will continuously indicate monitor output in the control A continuous strip chart recorder (one train) will also be located in room.

the control room.

An electronic calibration of the monitors will be performed every refueling outage.

In addition a radiation source will be used to perform an in-situ calibration of the monitor range below 10 R/hr.

The monitors will be mounted on the primary shield wall at an elevation of at least 750+2 (10 feet or more above the maximum post-LOCA water level of 0 and 180o in the lower containment).

The following McGuire General Arrangement drawings show the plan and sectional views with the monitor locations drawn in.

The monitors will be installed by June 1, 1981.

Containment Pressure Continuous indication of containment pressure has been provided in the control room. Measurement and indication range extends from -5 psig to 60 psig.

Each of the redundant differential pressure transmitters is located in an electrical penetration room and is equipped with one-half inch tubing impulse lines.

Each impulse line has a fail-closed isolation valve located in the annulus. These

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valves are normally open and have position indication and manual control in the II-13 03/09/81

control room. Continuous indication from each transmitter is provided in the control room. In addition, one channel of containment pressure is recorded.

These instruments are completely independent of the existing containment pressure transmitters.

i Containment Water Level Two containment floor and equipment sumps are provided on the floor of the lower containment (El 725') to collect floor drains and equipment drains.

However, these sumps and their associated pumps and instrumentation serve no safety function.

The containment emergency recirculation su=p at McGuire encompasses the entire floor of the lower containment.

The two ECCS recirculation lines take suction just inside the Containment wall at elevation 725' and are oriented horizontally.

They are not located in the bottom of a recess er su=p in the floor. Redundant safety grade level instrumentation is provided to measure emergency recirculation sump level. The range of this instrumentation is 0-20 feet (El 725' to El 745')

which is equivalent to a lower containment volume of approximately 1.000,000 gallons. The accuracy of this instrumentation is 10% over the full range.

The redundant differential pressure transmitters utilized in this instrumentation have been relocated to the annulus where a filled capillary system connects its associated transmitter with bellous sensors located inside containment.

Continuous indication from each transmitter is provided in the control room.

In addition, one channel of containment water level is recorded.

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Containment Hydrogen Monitorinc Continuous indication of hydrogen concen,tration in the containment atmosphere will be provided in the control room. This hydrogen monitoring system will consist of two redundant Consip, Inc./Delphi Systems Division K-111 analyzer systems with a range of 0 to 30% hydrogen by volume.

These analyzers operate independent of the recombiner system and will be powered from redundant Class 1E power supplies. Each analyzer will have its cwn contain=ent sample and return lines, and will be able to monitor either of two identical contain=ent sampling headers or the calibration gases.

Each analyzer will have a local control panel indicator and alarm and a separate control room indicator and alarm.

In addition, one channel of containment hydrogen concentration will be recorded.

Each containment sample header will have five inlet samples available for monitoring.

1.

Top of containment 2.

Operating level 3.

Basement 4.

Radiation Monitor /Recombiner Inlet header 5.

Radiation Monitor /Recombiner Discharge header All sample selection and switching is accomplished manually by the operator from the local analyzer control panel.

C1 This instrumentation vill be installed by June 1. 1981.

II-13A 03/09/81

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POST-ACCIDENT SAMPLING j

References:

NUREG-0578 - 2.1.8a Action Plan - II.B.3 J

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A new sampling panel has been designed to allow analysis of reactor coolant j

samples under accident conditions. A new sampling line will be connected to l

the present sampling line and routed directly to the sampling panel.

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to minimize personnel access limitations, this routing will be in accordance with the findings of the plant shielding review. This system will allow 1

collection of reactor coolant samples under both pressurized and zero pressure conditions at any level of coolant activity. The design of this system signi-ficantly reduces radiation exposures during sample collection under accident conditions.

This liquid sample system consists of a sampler panel that houses the tubing, valving, instrumentation and other system components. The system is controlled i

and monitored remotely from the sampler control panel. Distance and shielding are utilized to reduce personnel dose rates. The sampler panel will be located in an area close to the containment that would normally have limited accessibility.

The sampler control panels will be located in an accessible area shielded from the sampler panel location by walls, system components, tanks, etc., up to 250 feet cabling distance from the sampler panel. Sampler lines are stainless stael to facilitate. flushing and cleaning and are rated to 2500 psig and 650 F.

A diagram of this system is provided on the following figure.

The control of valves and pumps outside the containment is from the sampler control panel. Sample isolation valves inside the containment and their associated con-tainment isolation valves are operated by the control room operator to permit l

sampling. The operator can override these isolation valves to permit sampling under accident conditions. Selection of sample lines is provided on the sampler control panel. The logic of the control system will operate the selected sample valves at the proper time in the sampling sequence. The logic system also operates the demineralized water valves associated with the selected sample valves to flush the sample lines at the appropriate time in the sampling sequence. The operator can i

complete the sampling sequence in approximately 20 minutes.

In addition to the reactor coolant sample line, a containment atmosphere sample line will be routed to a new accident level.sa2pling panel. The containment atmosphere sample will be obtained from the hydrogen analyzer sample lines.

- This new post-accident sampling system is scheduled to be installed before January 1, 1982. A revised sampling program for the interim p,eriod between fuel loading and completion of the new system has been developed.

Procedures for collection and transport of reactor coolant, sump water, and

-containment air samples under post-accident conditions have been revised to incorporate actions to be taken to minimizo rudiation exposures.

These proce-dures specify the preplenning to be performed as well as modifications and j

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Samples can be collected within 0% one hour in all instances where personnel exposure does not exceed 3 rem /qtr whole body and 18 3/4 rem /qtr to the extremities.

If the predicted personnel exposures exceed the above, samples can still be colle_ted and analyzed but not within one hour. The time required to install additional shielding and allow sample collection while minimizing personnel exposure will be dependent upon the nature of the event. The analytical procedures have been reviewed and determined to be adequate for the expected sample activity levels.

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The design of McGuire Nuclear Station does not feature a reactor trip on turbine

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trip. This trip was removed from the McGuire design to prevent unnecessary reactor trips, particularly during initial startup.

Unnecessary reactor trips should be avoided to minimize reactor coolant system thermal cycles and challenges to the reactor coolant system protective devices. The removal of this anticipatory trip was possible due to the full load rejectio. capability of McGuire.

The McGuire trip system keeps surveillance on process variables which are directly related to equipment mechanical limitations, such as pressure, pressurizer water level (to prevent water discharge through safety valu s) and also on variables which directly affect the heat transfer capability of *.he reactor (e.g., flow, reactor coolant temperatures). Still other paramete.rs utilized in the reactor trip system are calculated from various process variables.

In any event, whenever a direct process or calculated variable exceeds a setpoint, the reactor will be shut down in order to protect against either gross damage to fuel cladding or loss of system integrity which could lead to release of radioactive fission products into the Containment.

An analysis was conducted to determine the potential for pressurizer PORV challenges following a turbine trip f rom full power both with and without an immediate reactor trip on trubine trip. This analysis considered both normal plant response and cases assuming the failare of certain central systems that can influence challenges to the pressurizer PORV's.

Two types of control system failures were considered:

failure of all steam dump valves to open on demand (not including the steam generator PORV's); and complete failure of pressurizer spray to

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l function on demand. Partial failures (for example, failure of half of the steam V

dump valves) were not considered.

The analysis demonstrated that if all of'the steam du=p valves failed to open the pressurizer PORV's would be challenged regardless of the presence or absence of an immediate reactor trip on turbine trip at full power.

If there was no failure of the steam dump valves the absence of the subject trip would result in challenges to the pressurizer PORV's whereas the presence of such a trip would not challenge the FORV's.

Installation of a direct reactor trip on turbine trip would only protect against PORV challenges initiated by a narrow range of events, that is turbine trips not initiated by a reactor trip or a safety injection and occurring at or near full power.

Furthermore, valves identical to the McGuire PORV's and PORV block vr.lves have been subjected to extensive steam flow testing. This testing was conducted at Duke's Marshall Steam Station in conjunction with the EPRI valve testing program. The testing demonstrates that the McGuire PORV's and PORV block valves meet all functional and design requirements and provides added assurance of proper PORV and PORV block valve operation.

Duke will install a direct reactor trip on turbine trip to provide this additional protection against PORV challenges. This trip will be installed by June 1,1981.

The reactor trip on turbine trip will be generated by either of the following signals:

^

i.

Four-out-of-four turbine stop valves closed e

Two-out-of-three turbine auto-stop oil pressure low e

II-19A 03/09/81

The four turbine stop valve signals will be developed through the actuation of independent limit switches mounted on the stop valve assemblies.

Each of the four turbine stop valve signals can be tested individually trem the control room through the digital electrohydraulic (DEH) control panel. The turbine auto-stop oil system is the medium through which a turbine trip is initiated.

Turbine auto-stop oil pressure is =easured by three independent pressure switches which are mounted in a terminal box located adjacent to the turbina.

The pressure switches can be tested from the control room by applying power to three solenoid valves located in the auto-stop oil system. The test circuit is designed to allow one pressure switch test per test actuation.

The limit switches and pressure switches used in this application are similar to those used in other Class lE applications in the plant.

Although the main turbine-generator is not seismic Category I, these limit switches and pressure switches are seis=ically qualified and the associated cables will be installed in accordance with the McGuire separation criteria.

Each turbine stop valve limit switch and each turbine auto-stop oil pressure switch provide an input to both trains of the solid state protection system (SSPS).

If either logic function as described above is satisfied, a reactor trip signal will be generated provided reactor power is greater than approximately 48% (PS). A logic diagram for the reactor trip on turbine trip is provided on the following figure.

It should be noted that the reactor trip on turbine trip was originally part of the McGuire SSPS design with the exception that the PS interlock will be substituted for the P7 interlock.

It is therefore concluded that reinstituting this trip will not degrade the existing protection system since all separation, testing, and reliability considerations are in accordance with the original SSPS design.

C.3.17 The McGuire Nuclear Station Technical Specifications require that in the event the ECCS is actuated and injects water into the reactor coolant system, a Special Report shall be prepared and submitted to the Commission pursuant to Specifica-tion 6.9.2 within 90 days describing the circumstances of the actuation and the total accumulated actuation cycles to date.

The current value of the usage factor

'er each affected safety injection noz:le shall be provided in this Special Report

..c+.-er its value exceeds 0.70.

This report will satisfy the require =ents of recort.ng ECCS outages so as to establish if there is any need for modifications to the ECCS.

C.3.25 At McGuire Nuclear Station the reactor coolant pump seal water is supplied by Nuclear service the charging pumps and cooled by component cooling water.

water in turn cools the component cooling water.

In the event of a loss of offsite power at McGuire the component cooling water pumps, the nuclear service water pumps, and the charging pumps are all supplied with emergency power from the emergency diesel generators.

II-19B 03/09/81

. _ _ - - - _ _ ~.. ~ - -.. - _.

1 C.3.30 and C.3.31 j

Westinghouse feels very strongly that the small break LOCA analysis model currently approved by the NRC for use on Westinghouse designed NSSS is conservative and in conformance with Appendix K of 10CFR 550.46.

It is noted that the NRC has not indicated the contrary. However, Westinghouse believes that improvement in the realism of small break calculations is a worthwhile effort. Whenever possible and whenever allowed by Appendix K, more realistic modeling assumptions will be employed.

This will hopefully provide more con-sistency among analyses performed for licensing, training and procedure writing.

I A detailed outline of the scope and schedule for this Westinghouse effort was I

submitted to the NRC on September 26, 1980.

I If the results of the new Westinghouse Model (and subsequent NRC review and approval) indicate that the present small-break LOCA analysis for McGuire are not in conformance with 10CFR 550.46, a new analysis utilizing the new and approved Westinghouse model will be submitted to the NRC.

i I

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II-19C 03/09/81

i i

l Peactor Trip on Turbine Trip I

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l TURBINE TURBINE STOP VALVE POSITION AUTO-STOP OIL PRESSURE E

m lPS l LS LS LS LS PS l PS i

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j 4/4 2/3 i'

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i REACTOR TRIP r

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NOTE: SHOWN 'FOR ONE TRAIN ONLY; other train is similar l

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I II-19D-02/06/81 j

Carryover 4

i 5

4 i

U PRIMARY COOLANT SOURCES OUTSIDE CONTAI.NMENT

References:

NUREG-0578 - 2.1.6a Action Plan - III.D.l.1 Periodic leak race test procedures have been written for systems carrying radioactive fluids outside of containment. Draft copies of these procedures are provided in Appendix C.

The following systems are included:

Safety Injection, Residual Heat Removal, Containment Spray, Nuclear Sampling, Boron Recycle, Chemical Volume and Control, Refueling Water, Liquid Waste, and Waste i

Gas. These tests, will be initially performed by June 1, 1981 and then during each refueling outage. The tests will be conducted by pressurizing a system or part of a system and checking non-welded pipe joints, penetrations, flanges, valve separations, packing, and pump packing for leakage. Where possible, pumps included in the leak test boundary will be run so that a more accurate determination of the leak rate may be made.

A separate periodic test procedure is being revised to assure that excessive leakage is de*ected on a timely basis. A draft copy of this procedure is provided in Appendix C.

This test will be run at least weekly and will require that systems carrying radioactive fluids outside of containment be visually inspected for excessive leakage. Appropriate corrective action will be taken if excessive leakage is detected.

I

/)

After the occurrence of an uncontrolled gaseous release at North Anna Unit 1 (ms in September 1979, and the NRC issuance of IE Circular 79-21, an evaluation of the potential release pathways for liquid and gaseous material was performed for McGuire Nuclear Station. This evaluation resulted in several minor station modifications to further reduce the possibility of an uncontrolled release at McGuire. These modifications included the installation of catch basins, curbs, shelters, and covers in various areas of the station.

'N III-5 03/09/81-

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DUKE POWER COMPANY N hE E M Y U'D McGUIRE NUCLEAR STATION WG SYSTEM LEAK DETECTION 1.0 Discussion 1.1 Purpose The purpose of this procedure is to dercribe the method of identi-fying leaks from the WG System to the Auxiliary Building atmosphere.

1.2 Principle Pressure gauges located throughout the WG System will allow us to monitor the WG System total gas inventory. Unaccountable drops in i

system pressure will indicate external system leaks to the Auxiliary Building. Detection of leak sources will be done both by isolating system boundary valves and by leak testing suspected valves, instru-mants, components and any associated tubing and piping. Both NUPRO

" Snoop" leak detector solution and a Matheson gas leak detector may be used for leak detection.

1.3 Limits and Precautions 1.3.1 The in-service U1 or U2 VCT cannot be isolated from the WG System more than hours at a time.

1.3.2 The NCDT cannot be allowed to pressurize above 5 psig.

1.3.3 If possible, the section to be leak tested should be purged of any radioactive gases before scaring any test-ing for leaks.

1.3.4 Anytime the "A" and "B" Recombiners are pressurized, the Control Valve Stem Leakoff header isolations (1WG203 and 1WG182 respectively) must be OPEN.

6 atm cc/see of Helium is the limit of 1.3.5 A leak rate of 10 detectability with the Model 8017 Matheson Gas Leak Detec-tor.

1.3.6 Contact Health Physics before entering any room or area for leak testing to determine the radiation levels for that area. Health Physics will designate what protective gear, if any, is needed before one can enter the desired area for leak testing.

h (a

2.0 Apparatus 2.1 One (1) Matheson Model 8017 Cas Leak Detector 2.2 One (1) two stage gas regulator designated for use with a Helium or Nitrogen gas cylinder.

2.3 Tygon tubing, 20 ft; 3/8" ID 2.4 Miscellaneous tubing and piping connections 3.0 Reagente 3.1 Helium cylinder 3.2 Nitrogen cylinder 3.3 Sulk Nitrogen 3.4 " Snoop" solution 4.0 Procedure The WG System has been divided into sections to be checked in order of highest priority of suspected leakage during system operation. The sections are as follows:

4.3 (Section 1)

WG System isolations to the Auxiliary Building 4.4 (Section 2)

WG Compressors 4.5 (Section 3)

Catalytic Hydrogen Recombiners (CHRs) 4.6 (Section 4)

WGDTs and S/D WGDTs 4.7 (Section 5)

Vent Header 4.8 (Section 6)

Miscellaneous system piping and components During initial and periodic scheduled leak testing, all of the above listed sections are to be completed. Otherwise when a decline in system pressure during normal system operation indicates a potential leak problem, it is necessary only to begin checking for leaks in Section 1 of the system and proceeding in a logical manner to the next Section which is suspect until the leak (s) is identified and corrected.

For valves When " snoop" testing, check any area where a leak may occur.

this would involve testing for leaks in the vicinity of both the valve stem and diaphragm while for instruments all associated tubing connec-tions as well as the instrument itself should be leak tested.

Results of all testing are to be recorded in the proper procedure Enclosure.

\\

r d 4.1 Initial Conditions 4.1.1 Pressure gauges have been installed at IWG314, IWG303 and 1WG304.

4.1.2 All pressure gauges throughout the system have current calibrations.

4.1.3 The section being tested for leaks is under the pressure i

it would be during normal system operation or at a higher test pressure.

If possible, the section being tested l

should be purged of radioactive gases and pressurized with either Nitrogen or Helium to at least normal operating pressure for that section of the system.

4.2 Identification of System leakage to the Auxiliary Building 4.2.1 Record the pressure on the following gauges and enter on.

OWGPS5100 (WGDT A,)

OWGPS5110 (WGDT B)

OWGPSS120 (WGDT C) l q

OWGPS5130 (WGDT D)

OWGPS5140 (WGDT E) i OWGPS5150 (WGDT F)

OWGPS5090 (S/D WGDT A)

OWGPS5090 (S/D WGDT B) f (IWG314)

(1WG303)

(1WG304)

(A COMP SEP)

(B COMP SEP)

OWGPG5500 (CHR "A" INLET)

OWGPG5750 (CHR "A" OUTLET)

OWGPG5770 (CHR "A" OUTLET TO WG DECAY TANKS)

OWGPG5540 (CHR "B" INLET)

OWGPG5740 (CHR "B" OUTLET)

)

OWGPG5760 (CHR "B" OITILET TO WG DECAY TANKS)

OWGPG5900 (DECAY TANK DRAIN PUMP DISCH)

V (DECAY TANK DRAIN PUMP SUCTION)

L

._ ()

4.2.2 After 1/2 hour, record the pressures shown on the gauges in Section 4.2.1 and enter on Enclosure 1.

4.2.3 If an unaccountable pressure drop occurs on any of the gauges, an external system leak is probably the cause, therefore proceed directly to the section of this pro-cedure which most nearly incorporates the gauge on which the pressure drop was noticed. Continue hunting for the leak using the format outlined in Sections 4.3 - 4.8 until the leak is found and corrected.

4.2.4 If a leak is detected from one section of the WG System to another, then refer to CP/0/B/

4.3 (Section 1) WG System Isolations to the Auxiliary Building 4.3.1 The following valves should be closed during normal system operation. Verify that they are closed.

VAI.VE OR INSTRUMENT VALVE PCSITION IWG321 Closed 1WG322 Closed

'A 1WG134 Closed w) 1WG136 Closed ANALYZER RACK ISOLATIONS Closed TO BUILDING EXHAUST Closed 1WG229 Closed 1WG202 Closed 1WG218 Closed 1WG294 Closed i

I IWG291 Closed i

1WG305 Closed 1WG217 Closed 1WG293 Closed 1WG300 Closed 1WG301 Closed l

IWG174 Closed 1WG75 Closed Any of 'these valves found partially open could be a system i

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leak source if the valve was not otherwise isolated from

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l the current system flow alignment.

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t 4.3.2 Check the following drain trap isolations by individually closing each valve and checking for leak stoppage.

1 1WG317 t

1WG296 IWG319 IWG297 1NB347 4.4 (Section 2) WG Compressors 4.4.1 Leak test the following potential leak points associated with WG Compressor "A" and record the results on Enclosure 2.

Align the valves as stated below before testing this section of the WG System for leaks.

VALVE OR INSTRUMENT VALVE POSITION IWG230 Open i

IWG26 Open 0WGFE6060 Closed IWG25

• 0WGPS5060

• 0WGPS5061 j

• 0WGPS5062

• 0WGPS5063 1

Closed 1WG23 closed 1WG30

• 0WGLT5040

• 0WGPS5040

• 0WGPS5041 Closed 1WG321 1WG31 Open 1WG32 Open 1WG233 open Closed 1WG236 1WG27

• Check all connecting points along the tubing that is associated with this instrumentation.

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h VALVE OR INSTRUMENT VALVE POSITION IWG234 Open 1WG48 Closed IWG19 Open 1WG29 Closed 1W 28 Closed OWGTH5600 "Y" STR1INER OWG7E6070 4

OWGTE6'080 OWG7E6090 WG COMP "A" top plug (used for lubrication purposes) 4.4.2 Leak test the following potential leak points associated with WG Compressor "B" and record the results on Enclosure 2.

Align the valves as stated below before testing this section of the WG System for leaks.

VALVE OR INSTRUMENT VALVE POSITION IWG232 Open

~

~

1WG40 Open OWGFE6050 IWG39 Closed

• 0WGPS5070

• 0WGPS5071

• 0WGPS5072

• 0WGPS5073 1WG37 Closed IWG44 Closed

• 0WGLT5050

• 0WGPS5050 j

I-

• 0WGPS5051 IWG45 Open 1WG46 Open 1WG233 Closed

• Cleck all connection points along the tubing that is associated with this instrumentation.

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! VALVE OR INSTRUMENT VALVE POSITION IWG236 Open 1WG237 Open 1WG48 Closed l

1WG322 Closed i

1WG19 Closed i

IWG42 Closed 1WG43 Closed OWGTH5610 "Y" STRAINER 1WG41 OWGFE6020 OWGFE6030 I

OWGFE6040 WG' COMP "B" top plug (used for lubication purposes) 4.5 (Section 3) Catalytic Hydrogen Recombiners (CHRs) 4.5.1 Leak test the following potential leak points associated O

with CHR "A" and record the results on Enclosure 3.

Align the valves as stated below before testing this section of the WG System for leaks.

VALVE OR INSTRUMENT VALVE POSITION OWGTE5370 IWG186 Closed OWGPG5550 1WG187 Open 1WG286 1WG69 closed OWGFE5360 IWG202 Closed OWGTE5650

^

OWGTE5380 i

IWG188 IWG286

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_ VALVE OR INSTRUMENT VALVE POSITION IWGl91 Closed 1WG192 1WG193 Open 1WG229 Closed 1WG200 OWGTE5390 1WG196 Closed OWGTE5520 OWGTH5680 OWGTE5530 OWGLT5710 IWG287 Closed 4

ILG199 open 1WG201 Closed 1WG203 Open 1WG197 Closed 1WG71 Closed 1WG198 Open OWGPG5750 OWGPG5770 All sample gas tubing from CHR "A" to the "A" analyzer rack The point at which the "A" analyzer rack compressor suction vent line connects with CHR "A" packing stem leak-off line 4.5.2 Leak test the following potential leak points associated with CHR "B" and record-the results on Enclosure 3.

Align the valves as stated below before testing this section of the WG System for leaks.

VALVE OR INSTRUMENT VALVE POSITION OWGTE5200 Closed IWG55 1WG54 Open OWGPG5540 O

Closed i

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1WG64

-.. VALVE OR INSTRUMENT VALVE POSITION OWGFE5620 Closed 1WG75 OWGTE5640 OWGTE5210 1WG168 IWG284 1WG171 Closed IWG172 1WG173 Open 1WG174 Closed OWGTE5220 i

LWG176 Closed OWGTE5350 i

OWGTH5690 OWGTE5360 OWGLT5700 1WG285 Closed IWG179 Open 1WG180 1WG181 Closed 1WG177 Closed OWGPG5740 4

IWG57 Closed I

IWG178 Open OWGPG5760 All sample gas tubing from CHR "B" to the "B" analyzer I

Rack.

l The point at which the "B" Analyzer Rack compressor suction vent line connects with CHR "B" packing stem leak-off line.

4.6 (Section 4) WGDTs and S/D WGDTs 4.6.1 Leak test the following potential leak points associated l

with WGDT "A" and record the results on Enclosure 4.

Align the valves as stated below before testing this section of the WG System for leaks.

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VALVE OR INSTRUMENT VALVE POSITION j

IWG264 Closed 1WG124 (Relief valve)

OWGPT5100 1WG267 Closed 1WG127 Closed 1WG2o6 Closed WGDT "A" MANWAY WELD WGDT "A" PLUG 4.6.2 Leak test the following potential leak points associated with WGDT "B" and record the results on Enclosure 4.

Align the valves as stated below before testing this section of the WG System for leaks.

VALVE OR INSTRUMENT VALVE POSITION IWG260 Closed 1WG117 (Relief Va'lve)

OWGPT5110 IWG262 Closed 1WG118 Closed IWG261 Closed WGDT "B" MANWAY WELD WGDT "B" PLUG 4.6.3 Leak test the following potential leak points-associated with WGDT "C" and record the results on Enclosure 4.

Align the "- es as stated below before testing this section of W System for leaks.

VALVE OR INS q

VALVE POSITION IWG257 Closed 1WG112 (Relief Valve)

OWGPT5120 Closed 1WG259 IWG113 Closed LWG258 Closed WGDT "C" MANWAY WELD WGDT "C" PLUG N-4.6.4 Leak test the following potential leak points associated with WGDT "D" and record the results on Enclosure 4.

Align the valves as stated below before testing this section of the WG System for leaks.

VALVE OR INSTRUMENT VALVE POSITION IWG253 Closed 1WG104 (Relief valve)

OWGPT5130 IWG256 Closed IWG107 Closed 1WG255 Closed WGDT "D" MANWAY WELD WGDT "D" PLUG 4.6.5 Leak test the following potential leak points associated with WGDT "E" and record the results on Enclosure 4.

Align the valves 'as stated below before testing this section of the WG System for leaks.

VALVE OR INSTRUMENT VALVE POSITION IWG249 Closed 1WG97 (Relief valve)

OWGPT5140 1WG251 Closed 1WG98 Closed IWG250 Closed WGDT "E" MANWAY WELD WGDT "E" PLUG l

4.6.6 Leak test the following potential leak points associated f

with WGDT "F" and record the results on Enclosure 4.

l Align the valves as stated below before testing this section of the WG System for leaks.

VALVE OR INSTRUMENT VALVE POSITION

-1WG246 Closed l

1WG92 (Relief valve)

Cg I

_ VALVE OR INSTRUMENT VALVE POSITION OWGPT5150 IWG248 Closed IWG93 Closed IWG247 Closed WGDT "F" MAWAY WELD WGDT "F" PLUG 4.6.7 Leak test the following potential leak points associated with S/D WGDT "A" and record the results on Enclosure 4.

1 Align the valves as stated below before testing this section of the WG System for leaks.

VALVE OR INSTRUMENT VALVE POSITION IWG271 Closed 1WG146 (Relief valve)

OWGPT5080 Closed

?WG151 IWG274 Closed 1WG273 Closed S/D WGDT "A" MANWAY WELD

-S/D WGDT "A" PLUG 4.6.8 Leak test the following potential leak points associated with S/D WGDT "B" and record the results on Enclosure 4.

Align the valves as stated below before testing this section of the WG System for leaks.

VALVE OR INSTRUMENT VALVE POSITION IWG275 Closed 1WG133 (Relief valve)

OWGPT5090 IWG156 Closed 1WG277 Closed 1WG276 Closed S/D WGDT "B" MANWAY WELD S/D WGDT "B" PLUG i

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13-4.7 (Section 5) Vent Header 4.7.1 Shutdown the WG System by taking both compressors ans i.

recombiners out of service.

4.7.2 Isolate the Vent Header by closing the following valves:

l VALVE OR INSTRUMENT VALVE POSITION t

i IWG3 IWG317 l

1' 1WG169 IWG2 1NB160 IWG9 J

l IWG8 IWG11 IWG13 IWG206 4

f IWG14 IWG319 l

1WG215 1FG296 i

IWG308 IWG315 i

1WG316 4

IWG19 i

i 1WG314 IWG230 1WG232 t

l_

1WG297 IWG245 IWG268 1WG135 IWG136 L

l 1WG302 I

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VALVE OR INSTRUMENT VALVE POSITION 1WG289 IWG290 IWG203 IWG182 IWG235 IWG231 IWG20 "A" ANALYZTT. RACK EXHAUST TO COMPRESSOR SUCTION VALVES i

"B" ANALYZER RACK EXHAUST TO COMPRESSOR SUCTION VALVES 4.7.3 Open the following valves:

IWG4 IWG5 IWG7

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IWG15 1WG16 IWG18 1WG314 i

IWG19 4.7.4 Connect a helium cylinder at OWGPG5000 and OWGPG5010.

4.7.5 Pressurize the Vent Header to 5 psig at OWGPG (1WG314 location). Record this pressure, wait 15 minutes and f-record any pressure drop on Enclosure 5.

NOTE: Maintain at least 3 psig in the vent header as f

i read at OWGPG throughout the vent header-leak test.

4.7.6 Leak test the following Vent Header items and record the

(

results on Enclosure 5.

Align the valves as stated below before testing this section of the WG System for leaks.

VALVE OR INSTRUMENT VALVE POSITION 1WG3 Closed l

IWG4 Open 1

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VALVE OR INSTRUMENT VALVE POSITION OWGPG5000 1WG317 Closed 0WGFT5020 (UNIT I VCT TURBINE FLOW METER) 1 IWG5 Open i

IWG6 f

IWG7 Open INV169 Closed P

l 1WG2 Closed INB160 Closed

-r IWG9 Closed IWG8 Closed IWGli Closed 1WG13 Closed 4

1WG206 Closed 1WG18 Open i

IWG17 4

IWC16 Open 5

OWG7T5030 (UNIT II VCT TURBINE l

FLOW METER) f 1WG15 Open IWG14 Closed

-OWGPG5010 l

IWG319 Closed 1WL215 Closed 1WG296 Closed OWGPT5170 1

Closed-f' IWG297 l

1WG135 Closed l

IWG136 Closed 1WG139 1WG268 Closed 1WG245 Closed 1WG289 Closed 1WG290 Closed 1WG203 Closed f

g

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v VALVE OR INSTRUMENT VALVE POSITION "A" ANALYZER RACK EXHAUST TO Closed COMPRESSOR SUCTION VALVES (and associated tubing)

"B" ANALYZER RACK EXHAUST TO Closed COMPRESSOR SUCTION VALVES (and associated tubing)

IWG308 Closed 1WG314 Open 1WG316 Closed 1WG315 Closed 1WG19 Open 1WG232 Closed 1WG230 Closed 1WG20 Closed 1WG231

~

Closed 1WG235 Closed

\\_/

1WG302 Closed 1WG182 Closed 4.7.7 Record the pressure at OWGPG on Enclosure 5 (should be between 3.0 psig and 5.0 psig; if not, pressurize the vent header to approximatelv 5 psig then record that pressure.)

4.7.8 Record on Enclosure 5 the pressure reading at OWGPG then open 1WG296. After 15 minutes, record the' pressure t

i at OWGPG____ on Enclosure 5 then close 1WG296. During the 15 minute period, leak test LWG306 and 0WGLS6120. Record I

the results on Enclosure 5.

l l

4.7.9 Record on Enclosure 5 the pressure reading at OWGPG then open 1WG317. Af ter 15 minutes record the pressure at i

OWGPG on Enclosure 15 then close IWG317. During 15 minute period, leak test 1WG318 and OWGLS6150. Record these results on Enclosure 5.

t 4.7.10 Record on Enclosure 5 the pressure reading at OWGPG then open 1WG319. After 15 minutes record the pressure N-at OWGPG on Enclosure 5 then close IWG319. During the 15 minute period, leak test IWG320 and OWGLS6160.

Record these results on Enclosure 5.

l

~

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-. _ _ 1 4.7.11 Record on Enclosure 5 the pressure reading at OWGPG then open 1WG297. After 15 minutes record the pressure ac OWGPG on Enclosure 5 then close IWG297. During the 15 minute period, leak test 1WG307 and OWGLS6130. Record these results on Enclosure 5.

4.8 (Section 6) Miscellaneous System Piping and Components 4.8.1 Leak test the following WG Compressor Discharge Header valves and record the results on Enclosure 6.

Align the valves as stated below before testing this section of the WG System for leaks:

VALVE OR INSTRUMENT VALVE POSITION IWG238 open 1WG239 Closed 1WG244 Closed 1WG263 Open IWG12' 1WG252 Open l

1WG102

,L IWG270 Open 1WG145 1WG110 Closed 1WG90 Closed 4.8.2 Leak test the following Compressor and WGDT Discharge Header valves and record the results on Enclosure 6.

I Align the valves as stated below before testing this section of the WG System for lesks.

VALVE OR INSTRUMENT VALVE POSITION 1WG126 l

IWG265 IWG137 l

IWG254 1WG106 l

IWG288 IWG269

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.- VALVE OR INSTRLMENT VALVE POSITION W

1WG240 IWG241 IWG53 IWG67 4.8.3 Leak test the following S/D WGDT Discharge Header valves and record the results on Enclosure 6.

Align the valves as stated below before testing this section of the.WG System for leaks.

VALVE OR INSTRUMENT VALVE POSITION i

1WG225 IWG220 1WG88 1WG149 IWG272 4.8.4 Leak test the following CHR Discharge Header valves and record the results on Enclosure 6.

()

Align the valves as stated below before testing this section of the WG System for leaks.

VALVE OR INSTRUMENT VALVE POSITION IWG243 1WG62 IWG242 i

1WG59 IWG77 1WG244 IWG245 4.8.5 Leak test the following WGDT and S/D WGDT Sample valves and record the results on Enclosure 6.

Align the valves as stated below before testing this section of the WG System for leaks.

s

1 f VALVE OR INSTRUMENT VALVE POSITION g

IWG134 4

IWG135 1WG127 1WG118 f

1WG113 t

IWG107 IWG98 LWG93 1WG151 IWG156 4.8.6 Leak test the following potential leak points associated with the N and H headers by aligning the valves as 2

2 stated below and then pressurizing both purge gas supply lines to the CHR's with Bulk Nitrogen.

Record the results on Enclosure 6.

7 VALVE OR INSTRUMENT VALVE POSITION a

I IWG82 1WG80 1WG84 1WG86 IWG83 IWG85 IWG81 OWGPG5160 (TUBING CONNECTIONS) 1WG204 IWG183 IWG185 IWG186 IWG312 IWG74 j

IWG58 IWG55 IWG313 4

IWG90 1

IWG110 IWG88 4,

.i m.,

..._,7_, _,

-.--..- 4.8.7 Leak test the following WG Release Header items and record the results on Enclosure 6.

Align the valves as stated below before testing this section of the WG System for leaks.

VALVE OR INSTRUMENT VALVE POSITION IWG158 IWG159 IWG160 1WG161 1WG303 "A" ANALYZER RACK BUILDING EXHAUST VENT ISOLATIONS "B" ANALYZEE RACK BUILDING EXHAUST VENT ISOLATIONS 1WG140 OWGFE6140 OWGFT6140 (TUBING CONNECTIONS)

WG DISCHARGE MONITOR 1 EMF 50 1WG300 1WG142 IWG301

-4.8.8 Leak test the following Degas Header valves and record the results on Enclosure 6.

Align the valves as stated below before testing this a

section of the WG System for leaks.

VALVE OR INSTRUMENT VALVE POSITION IWG162 IWG163 1WG164 IWG165 IWG166 IWG167

\\

U 7

• we

--,y-'--

ye-w-

ygr=

,y

-y eg, -t y

-r4 e

yvep e "

g-

-Ww,-

g y

gy---q-yw--

w

• • ~f--

e

...m_._._.

I i

l I 1 l

4.8.9 Leak test the following WG System Compressor and CHR's l

drain line items. Record the results on Enclosure 6.

i Align the valves as stated below before testing this 1

i section of the WG System for leaks.

4 VALVE OR INSTRUMENT VALVE POSITION 1

]

T-01 T-02 T-03 4

1 T-04 T-05 T-06 i

IWG283 1

IWG213 1

)

IWG279 IWG280 4

l IWC217 1WG215 IWG311 IWG219 j-1WG226 i

IWG218 IWG310 IWG308 i

i e

!~

l l

l l

+

i i

i k'

4

..,.-._.-4,,,..m.-,-,.4,.

,,,-.,~,.x,-,,m.,,-,_,m.-...,.-...--

~

PAGE 1 of 1 ENCLOSURE 1 s

(

)

OP/0/3/8600/10 u_/

WG SYSTEM INVENTORY GAUGE DATE/ TIME PRESSURE DATE/ TIME PRESSURE OWGPSS100 (WGDT A) l

/

l i!

/

l OWGPSS110 (WGDT B) l

/

l ll

/

l OWGPS5120 (WGDT C) l

/

0

/

I OWGPS5130 (WGDT D) l

/

l l

/

l OWGPS5140 (WGDT E)

{

/

/

OWGPSS150 (WGDT F)

/

l

/

l OWGPS5080 (S/D WGDT A)

/

/

I OWGPS5090 (S/D WGDT B)

/

l II

/

l (IWG314) l

/

ll

/

l l

/

I II

/

I (1WG303

/

/

(1WG304 f

("A" COMP SEP) l

/-

l

!l

/

l

("B" COMP SEP)

/

l l

/

^ 1WGPG5500 (CHR "A" INLET) l

/

l d

/

l

' 0WGPG5750 (CHR "A")

l

/

l

/

I OWGPG5770 (CHR "A" OUTLET)

/

I II

/

OWGPG5540 (CHR "B" INLET)

/

l

!l

/

I

/

ll

/

I OWGPG5740 (CHR "B")

OWGPG5760 (CHR "B" OUTLET)

/

/

OWGPG5900(WGDTDRAINPUMPDISCHi!

/

l h

/

i

PAGE 1 of 2

('.

ENCLOSURE 2 (j

OP/0/B/8600/10 SECTION 2: WG COMPRESSORS WG COMPRESSOR "A" DATE/ INITIALS ITEM

• RESULTS REMARKS

/

l IWG230

/

IWG26

/

OWGFE6060 l

l l

/

l 1WG25 l

/

OWGPS5060

/

OWGPS5061 l

l

/

f OWGPS5062 f

f f

/

OWGPS5063

/

1WG23

/

IWG30

/

OWGLT5040 l

l

/

l OWGPS5040 l

l

/

l OWGPS5041 l

/

IWG31

/

IWG32 l

l f

l

/

l IWG233

/

IWG234 l

l l

/

1WG48 f

f

/

IWG19 l

l l

l

/

l 1WG28 l

/

l 1WG29

/

l OWGTH5600 l

l l

/

l Y STRAINER

/

f 1WG27 l

/

l OWGFE6080

/

f OWGFE6090

/

OWGFE6070

/

WG COMP "A"

/

top plug O

• / - no leaks found x - leak found and corrected o - leak found but could not be corrected by Radwaste Chemistry Personnel L

PAGE 2 of 2 ENCLOSURE 2 OP/0/B/8600/10

(~~'s SECTION 2: WG COMPRESSORE

\\

WG COMPRESSOR "B"

w. '

DATE/ INITIALS ITEM

• RESULTS REMAKRS

/

1WG232

/

IWC40

/

OWGFE6050

/

1WG39

/

OWGPS5070

/

OWGPS5071 l

/

OWGPS5072 l

/

l OWGPS5073

/

1WC37

/

IWG44 0WGLT5050

/

/

l OWGPS5050

/

l OWGPS5051 l

IWC45

/

/

1WG46 l

/

l 1WG236 f1WG48

/

/

l IWG322

/

l 1WG19 l

l

/

l IWG237 l

/

l 1WG42 l

/

1WG43 l

/

OWCTH5610

/

l Y STRAINER

/

l 1WG41

/

l OWGFE6020 l

l l

l OWGFE6030

/

/

OWGFE6040

/

WG COMP "B" top pl.ug

• / - no leaks found

(',

x - leak found and corrected o - leak found but could not be corrected by Radwaste Chemistry Personnel

( )

_ _. = - - -. - _

= -

I PAGE 1 of 4 i

ENCLOSURE 3 I

OP/0/B/8600/10 i

O SECTION 3: CHRs I

U ca,,,,,,

DATE/ INITIALS ITEM

• RESULTS REMARKS l

/

OWGTE5370

/

IWG186

/

OWGPG5550 i

/

1WG187

/

IWG68

/

1WG69 l

/

OWGFE5630 l

f

/

1WG202

/

OWGTE5650 l

I

/

OWGTE5380

)'

/

IWG188

/

IWG286 i.

/

IWG191 l

/

1WG192

/

1WG193

/

1WG229

/

1WG203

/

OWGTE5390

\\

/

IWG196

/

OWGTE5520

/

OWGTH5680

/

OWGTE5530 1

j,

/

OWGLT5710

/

IWC287

/

IWG199 1

/

IWG200 l

/-

IWC201 l

/

OWGPG5770

/

l IWG197-

/

1WG71'

/

1WG198

• / - no leaks found x - leak found and corrected o - leak found but could not be-corrected by Radwaste Chemistry Personnel l

1 i

~...,....... _ _ - _, - _

. _ _ - ~ _.. _.., -.,,,. - -

-m..c--

-...-m,.

PAGE 2 of 4 l

1 ENCLOSURE 3 OP/0/B/8600/10 SECTION 3:

CHRs CHR "A" j

DATE/ INITIALS ITEM

• RESULTS REARKS

/

OWGPG5750 Sample gas tub-ing from CHR "A" I

to "A" Analyzer

/

Rack l

Connection point of "A" Rack Comp Suct Vent Line &

CHR "A" Packing Sten Leak-off I

/

Line

• / - no leaks found x - leak found and corrected o - leak found but could not be corrected by Radwaste Chemistry Personnel i

1 6

-,we-,,,..

.,-,,,,,n,-

_-, ~,.

..,.na,-

.<...-,~-~,~.,..,,,n.-

,w,-

-.. _. -. = _..

j PAGE 3 of 4

)

i ENCLOSURE 3 J

OP/0/B/8600/10 SECTION 3: CHRs CHR "B"

)

DATE/ INITIAL ITEM

• RESLT.TS REMARKS

/

OWGTE5200

/

IWG55 i

/

1WG54 i

/

OWGPG5540

/

IWG64

/

1WG65

/

OWGFE5620

[

/

IWG75 1

I

/

OWCTE5640

/

OWGTE5210

/

IWG168

/

IWG284

/

1WG171 l

/

IWG172 i

/

b IWG173 l

/

l IWC174 l

l

/

I OWGTES220 IWG176 i

/

OWGTE5350 I

/

OWGTH5690

/

OWGTE5360 l

/

OWCLT5700 t

-/

1WG285

/

IWG179 l

/

1WG180 l

/

OWGPG5760

/

IWG181-

/

IWG177

/

OWGPG5740-

/

1WG57 f

f

/

1WG178

• / - no leaks found x - leak found and' corrected o - leak found but could not be corrected by Radwaste Chemistry Personnel l

L.. -.

. =.

_ = _

i Page 4 of 4 B

ENCLOSURE 3 l

OP/0/B/8600/10 l

SECTION 3: CHRs CHR "B" DATE/ INITIAL ITEM

• RESULTS REMARKS Sample gat tub-i ing from CHR "B" to "B" Analyzer

/

Rack I

Connection point of "B" Rack Comp Suct Vent Line and CHR "B" Pack-ing Stem Leak-off

/

Line

l

• / - no leaks found l

x - leak found and corrected o - leak found but could not be corrected by Radwaste Chemistry Personnel i

O

PAGE 1 of 2 ENC 2,0SURE 4 OP/0/B/8600/10 SECTION 4: WGDTs and S/D WGDTs DATE/ INITIAL ITEM

• RESULTS REMARKS WGDT A

/

OWGPT5100 I

/

IWG264 l

l

/

IWC124 l

/

I IWG267 l

/

1WG127 l

/

1WG266 I

/

l MASWAY l

/

l PLUG l

WGDT B

/

1WG260

/

1WG117 l

/

OWGPT5110

/

IWG262 l

l

/

l IWG118 l

/

l 1WG261 l

/

! MANWAY I

l

/

l PLUG l

l WGDT C

/

1WG257

/

l IWG112 l

/

I OWGPT5120 l

/

l 1WG113 l

/

l IWG258

/

IVG259 l

l

/

MANWAY

/

l PLUG l

WGDT D

/

IWG253

/

I IWG104

/

l OWGPT5130

/

l 1WG256

/

1WG107 l

/

l IWG255 l

/

l MANWAY A

/

PLUG

• / - no leaks found x - leak found and corrected o - leak found but could not be corrected by Radwaste Chemistry Personnel

PAGE 2 of 2 ENCLOSURE 4 OP/0/B/8600/10

,m, SECTION 4: WGDTs and S/D WGDTs t

i

(_/ DATE/ INITIAL ITEM

• RESULTS REMARKS WGDT E

/

1WG249

/

1WG97 l

/

l OWGPT5140 l

l

/

1WG251 l

/

1WG98 l

l

/

1WG250

/

l MANWAY l

l

/

l PLUG WGDT F

/

1WG246

/

IWG92 l

l

/

OWGPT5150 l

/

1WG248 l

/

l IWG93 l

l

/

l 1WG247 l

l

/

l MANWAY l

l

(

l

/

l PLUG l

l l

3 S/D WGDT A l

/

1WG271 1

l

/

l 1WG146

/

l OWGPT5080 l

/

l 1WG151 l

l

/

l IWG274 l

l

/

l 1WG273 l

f

/

l MANWAY l

l

/

PLUG l

l S/D WGDT B l

l

/

1WG275 i

/

l IWG153 l

l

/

l OWGPT5090 l

l

/

l 1WG156 l

l

/

l 1WG276 l

l l

l

/

l 1WG277 l

l

/

l MWAY I

j

/

l PLUG

• / - no leaks found x - leak found and corrected o - leak found but could not be corrected by Radwaste Chemistry Personnel

PAGE 1 of 2 ENCLOSURE 5 OP/0/B/8600/10

/

8

)

SECTION 5: VENT liEADER DATE/ INITIALS ITEM

• RESULTS REE\\RKS

/

lIWG3

/

l IWG4 l

l

/

OWGPG5000 l

l

/

lIWG317 l

/

fOWGFT5020 l

l

/

!IWG5 l

/

l1WG6 l

l

/

l IWG7 f

f lINV169 l

f

/

/

I IWG2 l

l

/

l INB160 l

l

/

l IWC9 l

l

/

IIWG8 l

/

l IWG11 f

l

/

l IWC13 I

I

/

l IWG206 l

l

/

lIWG18 l

l

/

I IWG17 I

/

l IWG16 l

}

/

l OWCFT5030 l

l

/

l IWGIS l

l

/

I IWG14 l

/

I GWGPG5010 l

l

/

l IWG319 l

}

l l

/

l 1WL215 I

/

l IWG296

/

l OWGPT5170 l

l

/

l IWG297 l

l

• / - no leaka found x - leak found and corrected o - leaks found but could not be corrected by Radwa.2te Chemistry Personnel

J A

PAGE 2 of 2 ENCLOSURE 5 OP/0/3/8600/10

/3 SECTION 5: VENT HEADER DATE/ INITIALS ITEM

• RESULTS REMARKS l

/

IWG135 l

/

IWG136 f

/

l 1WG139

/

1WC268

/

1WG245 l

/

l IWG289 l

l

/

l 1WG290

/

f IWG203 l

l "A"

RACK COMP

/

SUCT ISOL VALVES f

l

/

l 1WG182 "B" RACK COMP

/

SUCT ISOL VALVES l

l

/

IWG308

/

l IWG314 l

l

/

l 1WG316 l

l l

)

/

1WG315 l

l

/*

\\

v

/

IWG19 l

l

/

1WG230 l

l l

l

/

IWC232 l

l

/

IWG20

/

l 1WG231 I

I

/

l IWG235 l

l l

l

/

l IWG302

• / - no leaks found x - leak found and corrected o - leak found but could not be corrected by Radwaste Chemistry Personnel DATA PROCEDURE STEP 4.7.5 Pressure at OWGPG after initial pressurization to N 5 psig 4.7.6 Pressure at OWGPG after 15 minutes 4.7.7 Pressure at OWGPG 4.7.8.1 Pressure at OWGPG before opening 1WG296 Pressure at OWGPG after 1WG296 has been open for 15 min.

l 4.7.8.2 Pressure at OWGPG before opening IWG317 Pressure at OWGPG after 1WG317 has been open for 15 min.

/N 4.7.8.3 Pressure at OWGPG before opening IWG319

(#l Pressure.st OWGPG after 1WG319 has been open for 15 min.

4.7.8.4 Pressure at OWGPG before opening IWG297 Pressure at OWGPG after 1WG297 has been open for 15 min.

PAGE 1 of 4 ENCLOSURE 6 OP/0/B/8600/10

[v SECTION 6: MISC SYSTEM PIPING AND CGMPONENTS DATE/ INITIALS ITEM

• RESULTS REMARKS COMP DISCH HEALER

/

IWG238

/

I IWG239

/

l 1WG244 l

l l

/

l 1WG263

/

f IWG122 l

l

/

1WG252 l

l

/

IWG102 l

/

IWG270 l

l

/

1WG145 l

/

l 1WG110 l

/

I IWG90 l

WGDT DISCH HEADER

/

IWG126

/

l 1WG265 l

l m

(Q

/

l 1WG137 l

l l

/

l 1WG254

/

l 1WG106 l

l

/

l 1WG288 l

l

/

l 1WG269 l

l l

/

1WG240

/

l 1WG241 l

l I

l

/

1WG53 l

/

l 1WG67 S/D WGDT DISCH EEADER.

/

1WG255 l

/

1'4G220

/

1WG88 l

l

/

l IWG149 l

l

/

IWG272

• / - no leaks found x - leak found and corrected

\\ )

o - leak found but could not be corrected by Radwaste Chemistry Personnel

PAGE 2 of 4 ENCLOSURE 6 OP/0/3/8600/10 SECTION 6: MISC SYSTEM PIPING AND COMPONENTS DATE/ INITIALS ITEM

• RESULTS REMARKS DECAY TAh"stS SAMPLE HDR

/

IWG134

/

l 1WG127 l

l l

l

/

l 1WG118

/

l IWG113 l

/

l 1WG107

/

l 1WG98 l

/

l 1WG93

/

l IWG151

/

l 1WG156 N3 and H2 HEADER

/

1WG82

/

l 1WG80 l

/

l IWG84

's

/

1WG86 l

V

/

IWG83 l

l l

/

IWG85 l

l

/

1WG81

/

OWGPG5160 l

l l

l

/

l 1WG204 l

l

/

l 1WG183 l

/

1WG185 l

IWG312

/

l l

/

l 1UG74

/

l IWG313 l

l l

l

/

IWG58

/

l 1WG55 l

l

• / - no leaks found x - leak found and corrected o - leak found but could not be corrected by Radwaste Che:sistry Personnel CN fV

PAGE 3 of 4 ENCLOSURE 6 OP/0/B/8600/10 7-t.

)

SECTION 6: MISC SYSTEM PIPING AND CCMPONENTS

'v

~

DATE/ INITIALS ITEMS

• RESULTS REMARKS WG RELEASE HEADER

/

IWG158

/

l IWG159 f _.

/

l IWG160 l

/

IWG161 l

l

/

IWG303 l

l "A" RACK BLDG j

/

VENT ISOLS "B" RACK BLDG l

/

VEhi ISOLS 4

/

IWG140 l

l

/

OWGFE6140 f

l

/

OWGFT6140 l

l l

WG DISCH l

l

/

MONITOR 6

/

1WG300 l

l

/

IWG142 l

l 9

/

1WG301 DEGAS HEADER I

/

IWG162 I

/

l IWG163 I

l

/

l 1WG164 l

l

/

l IWG165 I

l

/

l 1WG166 i

/

l IWG167 l

l DRAIN & FILL j

HEADER, l

/

T-01 i

/

l T-02 I

l l

l

/

l T-03 l

T-04 l

/

/

I T-05 l

}

/

I T-06 l

l

• / - no leaks fotnd x - leak found and corrected o - leak found but could not be corrected by Radwaste Chemistry Personnel

,o N_j

PAGE 4 of 4 ENCLOSURE 6 OP/0/B/8600/10 SECTION 6: MISC SYSTEM PIPING AND COMPONENTS DATE/ INITIALS

_ ITEM

• RESULTS REMARKS i DRAIN AND FILL HEADER (CONT.)

/

/

1WC213 l

/

! IWG279

/

j 17G280

/

lIWG217

/

IWC215

/

l IWG311

/

IWC219

/

1WG226

/

IWG218

/

IWG310

/

l IWG308 l

l f'i

• / - no leaks found V

x - leak found and carrected o - leak found but could not be corrected by Radwaste Chemistry Personnel A

N

PT/1/A/4600/03C PAGE 1 0F 2 r

,/ :'-

}

DUKE POWER COMPANY j S-l] I, F,O"' ' AT!C : '

McGUIRE NUCLEAR STATION g ' j g ;rg e:,Ly WEEKLY SURVEILLANCE n aiS 1.0 Purpose To verify the proper operation and/or condition of various instruments and/or systems.

2.0 References 2.1 McGuire Technical Specifications 3.0 Time Required One operator for 30 minutes at least once per 7 days.

4.0 Prercquisite Tests None 5.0 Test Equi;=ent One pyrometer or digital thermometer capable of reading ambient temperature.

NOTE Insure calibration'date is current.

6.0 Limits and Precautions 6.1 If an acceptance criteria is not ret, a report to the Nuclear Ec;..... cry Commission may be necessary, the Shif t Supervisor

_.. Control Room Operator shccid be notified i= mediately.

a v.

n.e.,scating Engineer and the Teianical Services Engineer may need to be notified if deemed appropriate by the Shif t Supervisor.

6.2 If...j instrument or systen is dct.;mir. ' to be inoperable, censult the appropriate Technical Specification Action State =ent to ac6 ermine any necessary immediate and/or corrective action.

6.3 If the Unit status or System condiL *cn prevents the performance of a surveillance item, the item should be noted.

7.0 Required Unit Status See Enclesure 13.1.

8.0 Prerequisite System conditions See Eacicsere 13.1.

9.0 Test Method Verificet'en that each acceptance crit- '- '

stisfied by a visual examinctica of the appropriate syster....

.atation.

~.

- ~... _.

-_=_.-.

l Pu is.i/~w6w,uSC I

PAGE 2 0F 2 I

l i

10.0 Data Required See Enclocure 13.1.

I 11.0 Acceptance Criteria See Enclosure 13.1.

j I

12.0 Procedure

)

12.1 Obtain the Boron Concentration of the Boric Acid Tank and the Refueling Water Storage Tank from the Chemistry Data Logbook.

If the Unit is in Mode 1, 2, or 3, also obtain the Boron Concentration of the Boron Injection Tank.

12.2 Check gross leakages on the ND and NI pumps; accessible NV, ND, and NI relief valves; accessible NV, ND, and NI equipment drains; and accessible NV, ND and NI valves for package leakages and record on Enclosure 13.1 i

Page 3 of 4.

12.3 Record any data noted on Enclosure 13.1.

12.4 Verify each applicable surveillance item on Enclosure 13.1.

12.5 If a surve.111ance item was.not completed due to Unit status or System ennditions, identify that surveillance item by writing j

"N/R" in the appropriate initial blank.

i j

12.6 If the Reactor Coolant Pumps are in operation, record the 1

vibration data for each pump.

13.0 Enclosures 13.1 Weekly Surveillance Items Checklist i

f-j s

Work Performed lly s'1/1/A/4ouutux. 3.1 g,re t,. _.-...+ w.

PdSe 1 of 4 1:. de ; o f og l on _ _.,_

cc et i I.

a nc ~

-Cal. au( sa t i.-

_ _ _. _... _ -W 1. C_I.Y _:.Ill_:_ _ V_d._.1.: _1._A_N_CE_ ITD15 CilECKLI ST 1 ilRVI.i t.LAt CE ITEM AG hPTAt!Ci: CI:l TERI A 123456 REOUIRED NI fl AI.

Axial Flux Difference

• AFD shall be maintained within the

/

following target band (flux differ-I ence units) about the target flux diffurence:

a. +5% for core aver-age accumulated burnup of <3000 MWD /MTU.

b.

+3%-12% for core average accumulated burnup of

>3000 MWD /MTU.

Quadrant Power Tilt Ra t io **

<l.02

/

FWST Solution Boron Concen-2000 < C < 2100 ppm boron y

/ j j

/ /

ppm tration 472 < L < 498 in.

/

/ / /

in W.C.

FWST Solution level Boric Acid Storage Tank

>37%

y y y y Solution level d

Containment Purge Time Accu-Less than 90 hours0.00104 days <br />0.025 hours <br />1.488095e-4 weeks <br />3.4245e-5 months <br /> for the pre-mulation f re:n Computer vious 365 days.

/

/ / /

hrs.

1. 33 in, FWST Solution level y

/

In W.C.

Boric Acid Storage Tank

>4. 5%

j j

Solution level Boric Acid Storage Tank Sol-7000 < C < 7 700pimi boron ution 15cron Concentration

/

/ / / /

/

ppm Boric Acid Storage Tank So-

>65'F j

j j j j j

lution Temperature Spent Fuel Pool Level ***

>(-2.8 ft.)

I/

//

/ / I ft.

i BIT Solution Boron Concen-20,000 < C <22,500ppai boron y

jj tration I

l BIT Solution level Positive Pressure ludication on BIT Pressure meter

/

/ y N/R

, Containment Purge Time Accu-Leno than 90 hours0.00104 days <br />0.025 hours <br />1.488095e-4 weeks <br />3.4245e-5 months <br /> for the pre-

)

lmulation from Computer vious 365 days l/

v; d /

Ilrs.

Work Performed liy PT/1/A/4600/03C Date

/~N 1:nclosure 13.1

'S

, Mode of L,,,,' tion Page 2 of 4 a

Time WEEKLY SURVEILLANCE ITEMS CilECKLIST APPLICABLE HODE

• LATA SURVEILLANCE ITEM ACCEPTANCE CRITERIA 123456 REQUIRED INITIAL Boron Injection flowpath (BAT

>65*F y

y j j j j

,F to NC System) * ***

Ice Condenser Inlet Door Alarm not illuminated.

/

/

/' /

N/A Positioning Monitoring Sys.

Groundwater Lovel Alarms No alarms illuminated

/

/

/

/

/

/

N/A 1WM-46 Liquid Waste

, Discharge Valve *****

Verify Closed

/

/

/

/

/ /

N/A IWC-160 WC Decay Tank Outlet to Unit Vent Control Verify Cloacd

/

/

/ / / /

N/A Ik'G-161 WG Decay Tank Outlet to Unit Vent' Control Verify Lochol Clonc.1

~

/

/

/ / '/ /'

~

N/A Inninrinn *****

• Mode 1 above 15% rated power.

• • Mode 1 above 50% of Rated Power

• • • Irradiated Fuel assemblies seated in

• • • • Use calibrated pyrometer

• • • • • Valves are to be verified in the storage racks.

position unless a release is Check One:

in progress.

No discrepancies Discrepancy sheet attached e

4

I L

x A

I I

T I

N I

D E

A A

^

A A

A A

A A

A A

AR

/

/

/

/

/

/

/

/

/

/

TI N

N N

N N

N N

N N

N AU s

Q

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1 3

4 "1

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f "r

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0

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/

/

/

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/

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/

h 3

l

" s f

c 0

E. 5

/

/

/

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/

/

/

/

/

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a

'o t

' l

' c 8 4

/

/

/

/

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/

/

/

/

/

/

/

t l

i n A f

A

'E P

AC 3

/

/

/

/

/

/

/

/

/

/

/

/

/

T I.

t St 2

/

/

/

/

/

/

/

/

/

/

/

/

/

ee I

/

h I

/

/

/

/

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/

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/

_/

1 L

P S

KA C

y E

c l

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l r

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I i

)/

A D

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x N

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(

A.

T g

g I

I a

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R k

k I

C a

a E

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V E

l l

R C

U l

g g

l S

A n

n T

i i

Y P

k k

L E

c c

K C

a e

a a

e e

e e

e e

e e

E C

j g

p p

g g

B n

r g

g g

g A

a a

a a

a a

a a

a E

i k

m m

k k

k k

k k

k k

k ka a

u u

a a

a a

i a

a a

a e

e m

m e

e e

e e.

e e

e e

d l

l i

i l

l l

l l

l l

l e

t a

n N

N H

M N

N N

N to o

o o

o u

o o

i i

o o

o o

N N

N N

J ac i

s s

s t

f f

f a

f f

f a

o o

o p

k k

k e

M a

a a

r E

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e s

s s

c,.

T l

l l

n n

n I

i i

i i

s g

s g

s g

a a

a D

E e

n e

n e

n r

r r

C v

i v

i v

i d

d d

o N

l k

l k

l k

J.

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c a

c a

c t

t t

L.

v a

V a

v a

n n

n A

B A

B p

p p

e e

e l

1 1

1 1

f f

f m

m m

e l

y n

E e

e e

e e

e p

p p

n p

p p

p i

v i

v i

v i

i i

O l

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f i

R m

m m

m l

l l

l t

l_

u u

u d

t U

u u

u u

e a

e a

e a

q q

q k

e S

P P

P P

R v

R v

d v

E E

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m i

r D

D 1

I V

V I

i e

)

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N N

N N

N N

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h.

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1 l

N N

N C

I f

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P o

k a2 o3d r

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l p!

l s

I'e'rferred 3y PT/1/ A/ 'ie00/03C 6,3 g, 3.1 Page 4 of 4 REACTOR COOLANT PUMP VIBRATION DATA REACTOR COOLANT PUMP CHANNEL NUMBER

• VIBRATION MILS (Peak to Peak) 1A 1

2 3

4 1B 1

2 3

4 1C 1

2 3

4 l

l l

p 1D 1

V 2

i 3

4 i

' Channel 1 - Pump Shaft Vibration l

f Channel 2 - Pump Shaft Vibration Channel 3 - Motor Vibration Monitor Channel 4 - Motor Vibration Monitor O

r