ML17266A517

Forwards Responses to NRC Requests for Addl Info Not Formally Submitted on Docket.Responses Will Be Incorporated Into Future FSAR Amend

ML17266A517
Person / Time
Site:	Saint Lucie
Issue date:	10/15/1981
From:	Robert E. Uhrig FLORIDA POWER & LIGHT CO.
To:	Eisenhut D Office of Nuclear Reactor Regulation
References
L-81-450, NUDOCS 8110280204
Download: ML17266A517 (118)

Text

REGULAITORY I>> ORMATION, DISTRIBUTION SYSl 'RIDS) g7 ACCES 8 I 0'V< NSR ' 1'1 0280 20 4 DOC ~ DA>>TEi 8 1/1 0/~ NOTAR'I ZED DOCKEITI FACILE:50-389 St. Lucie Plenty Unit 2g, F]orida Power L Liight Co~, 0'5090389 AUTHI,NAMEI AUTHOR AFFILIATtION UHRIGeR ~ Ee Floriidai Power E Light Co.

RECIP, VAREI RECXP.IENT AFFILtIATION F ISEVHUT'i D.G. Di visi on of Li censing

SUBJECT:

" Forwards responses to NRC requests for addi info noti formally suomitted on docket>>,Re'sponses will bei incorporated into future FSAR a@end.

aa,e DtISTRtIB>>JTION CODEI: B091$ COPIES RECEIVED:LITR l'NCL' 'IZE".:

TCTLEI:.. PSALM>>/FSAR AMDT6 and Related Corre'spondence NOTES'::

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FLORIDA POWER & LIGHT COMPANY October 4Q, 1981 L-81-450 Office of Nuclear Reactor Regulation Attention: Hr. Darrell G. Eisenhut, Director Division of Licensing U. S. Nuclear Regulatory Commission Washington, D. C. 20555 III'IIICLEARREGULA'ICti COhNI 55 QN

Dear Hr. Eisenhut:

Re: St. Lucie Unit 2 Docket No. 50-389 Final Safety Analysis Report Re uests For- Additional Information Attached are Florida Power 8 Light Company (FPL) responses to NRC staff requests for additional information which have not been formally submitted on the St. Lucie Unit 2 docket. These responses will be incorporated into the St. Lucie Unit 2 FSAR in a future amendment.

Very truly yours, Robert E. Uhrig Vice President Advanced Systems 8 Technology REU/TCG/ah Attachments cc: J. P. O'Rei lly, Director, Region II (w/o attachments)

Harold F. Reis, Esquire (w/o attachments) gooI ti 8ii0280204 PDR ADOCK 8ii0 05000389 A PDR PEOPLE... SERVING PEOPLE

0 l

Attachment to L-81-450 October 15, 1981 A. Minutes of a meeting held on October 15, 1981 on the subject of design values used for relative displacement between structures resulting from seismic event (attachment included).

B. Minutes of a meeting held on October 14, 1981 on the subject of compaction files.

C. Minutes of a meeting held on October 23, 1981 on the subject of Inadequate Core Cooling.

D. Revised response to SER open item: Adequacy of Station Voltages.

E. Additional information requested by NRC review Sal Salak:

l. Training for Mitigating Core Damage

2. System Difference/Requalification Training Outline 3~ Hot License Operator Training Program

4. Licensed Operator Requalification Program F. Minutes of a meeting held on October 16, 1981 concerning loss of all AC power (morning and afternoon'meeting) plus attachments.

G. Minutes of a meeting held on October 15, 1981 concerning Matrix Power Supply Test (plus attachments).

H. 1. 'ff/Normal 'procedures, Reactor Ttrip/Turbine Trip

2. Emergency procedure, Blockout Operation

MEETING MINUTES FOR MEETING HELD ON OCTOBER '15, 1981

Subject:

St. Lucie Unit 2 Design Values Used for Relative Displacement Between Structures Resulting From a Seismic Event Location: C-,E Conference Room, Triangle Towers, Bethesda, Maryland Attendees:

NRC NRC EBABCO V. Nerses E.W. Dotson J. Burket J. Ra)an W.F. Brannen P. Grossman R. Bosnak R.W. Gritz W. Fan J. Brammer p.p. Carier N.S. Whany H.E. Polk A. Bohen Proceedings:

Ebasco presented a comparison ,of the design values used for relative displacements between structures resulting from seismic event to values derived from an absolute summation of time history displace-ments. See attached handout for details of the presentation. It was demonstrated that the design values were comparable to the absolute values.

The NRC staff indicated that the values used for design, and not the method by which they were derived, are acceptable to the staff.

t I

COMPARISON OF MAXIMUM DIFFERENTIAL SE SMIC DISPLACEMENTS g DBE TIME HISTORY VS. DESIGN VALUES, ST LUCIE UNIT NO. 2*

On St Lucie Unit 2, the design values for displacement of structures not on a common mat were computed by a SRSS summation of maximum response spectra displacemen'ts. The maximum response spectra values are based on enveloping maximum response spectra displacements of each =building considering a range of various soil properties. In re-sponse to the NRC's request for justification of these design values, various scoping studies have been performed to determine appropriate values for maximum expected displacements at St Lucie Unit 2. These values are comp'ared to those actually used in the pipe stress analysis calculations.

Sco in Studies Summar

1) The most realistic indication of the true differential seismic displacements can be seen by comparing the time history values

-between two points on different structures and calculating the maximum difference. Inherent in this comparison are two as-sumptions for adjacent structures:

1. Motion of the two structures begin in-phase.

2. Differential ground motion between the two structures are negligible.

Table 1 indicates these maximum difference values for two cases:

for piping system going from18'-6" elevation 28'-0" in the reactor building (RB) to elevation in the reactor aux building (RAB) and for a- system from elevation 48'-0" in the RB to 42'-6" in the RAB. These two cases are typical for piping between these two buildings. Note that the time histpry values are in all cases substantially less than the St Lucie Unit 2 design values.

2) In order to account for any additional displacements resulting from inaccuracy due to the two assumptions stated above, the time history values hav'e also been combined by maximum summations of the values as a function of time. As would be expected, these absolute sum displacements are greater than those calculated previously; However, as shown in .Table 1, these values are in all cases less than the St Lucie Unit 2 design values.

3) An additional parameter that could be considered is variation in time interval. The time history earthquakes were generated using a time interval of .005 sec. This could be "spread" by 20% analagous to the peak spreading-employed for response spectra.

Thus time histories with intervals. from .004 to .006 sec would be considered. 'his spreading would account for uncertainties in design assumptions.

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Table 2 indicates that

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even assuming 20% time interval spread, the maximum differential displacements are in all cases less-

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than the St Lucie Unit 2 design values.

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• This supplements the response to MEB SER dated item 018 transmitted to the NRC (Uhriq to Eisenhut) via letter L-81-381 9/1/81.

) Finally, the displacement values from the "spread" time history have been combined by the maximum sum method. As shown in Table 2, the design values are in all cases within 15% of the values determined by this technique. The maximum difference is less than 1/8".

Discussion Seismic displacement is a secondary stress when applied to piping systems and is combined with other stresses to determine the total effe'ct. For supports and penetrations, seismic displacement is a primary load. However, this is combined with several other primary loads. Variations in the magnitude discussed above would have negligible impact on the total system and would therefore not require any further evaluation.

It should 'also be noted that exists only for the .006 sec the deviation in displacement values time interval. The .006 sec time interval reflects a 20% increa'se above the base case of .005 sec to cor-respond to the spectra broadening that was used on St Lucie Unit 2.

However, this interval could be reduced to 15% and be in compliance

.with the Standard Review Plan 3.7.2 (ll/24/75). Th'e effect would

,.be to reduce or eliminate the deviation between the time history ivalue and the design value.

onclusion ased on the'omparisons .presented in Tables 1 and 2 and the

discussions above, the specific seismic displacement design values

.developed for St Lucie Unit 2 are acceptable. This conclusion

-would apply to all the piping systems which penetrate the containment

,and Shield Building listed in FSAR Table 6.2-52 and located at elevation 37 feet and below.

Penetrations not included are:

Penetration ~Sstsm Containment Purge Exhaust ll 10 12, 13 HVAC HVAC Spares Containment Purge Supply 25 Fuel Transfer Tube 45,53 Fluid Instrument Line 55 Spare 56 HVAC Containment Mini Purge Inlet 57 HVAC Containment Mini Purge Exhaust 58 Spare 59,60 HVAC Shield Building Vent System 61 Spare

'2,65'.,66 HVAC Instrument Does not penetrate containment 63 Spare 67;68 HVAC Containment Vacuum Relief

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TABLE 2,

, ST LUCIE 02 EFFECT OF T VARIATION ON COMPARISON OF TIME HISTORY DISPLACEMENTS TO DESIGN VALUES ABS SUM'FT) MAX. DIFF. (FT) DESIGN VALUES N-S .0457 .0391 Displacem .05150515 hgpenRB . 005. .0484 . 0462 E118'-6'004 RAB .006 '. 0567 .0414

'515 VERT .004 .0357 .0309 Disp. bet- .0372'372 KIN-P s .005 .0371 .0286 r RAB E118'-6'-W

.006 .0431 ;0364

. 004 .0410 . 0392 Disp. Bet-ween RB E128'-0" & .005 .0462 .0426 RAB El 18'-6" .006 .0590 . 0394

.0372'0522'0522'0522'0649'0649'649s N-S .004 .0551 .0496, Disp. Bet-ween RB El.

48'-0" .'005 .0615 .0593 RAB E142'-6" .006 .0713 .0528 E-W . 004 .04806 .'04806 Disp. Bet-ween RB t

El. .0625'0625'0625' 48 Qll .005 .0591 .0531 RAB E142'-6" 006 .0728 .0502

MEETING MINUTES FOR MEETING HELD ON OCTOBER 14i 1981

Subject:

FPL St. Lucie Unit 2"- Compaction Piles Location: C-E Conference Room,. Triangle Towers,'Bethesda, Maryland Attendees:

NRC 'FPL EBASCO V. Nerses E.W. Dotson 'W.F. Mercurio G. Lear P.P. Carier G.M. Coscia L. Heller W.F. Brannen R. Pichumani Proceedings:

Ebasco opened the meeting by presenting a brief history of the compaction pile/slope stability issue. It was noted that:

1. The NRC's original concerns were related to the possi-bility of a sub-surface layer of in-situ soil liquefying during a seismic event resulting in a land slide that would block the flow. of cooling water through the intake canal.

2. The NRC was originally concerned about the switch yard area. However, it was determined that, due to the proxi-mity of the switch yard to the plant intakes, a slide of this area would not block cooling water flow.

3. Concerns related to the barrier wall area and the slopes north and south of the plant- intake structures were identified late in the discussions with the NRC.

4. FPL never recognized that the soils in question were potentially liquefiable. The instituted the compaction pile program only to satisfy the NRC's concerns.

5. Although the liquefaction issue was dismissed at the Unit 1 ACRS sub-committee meeting, no definite conclusion one way, or-the other, was'ever established.

Ebasco then reviewed the compaction pile program instituted at the slopes north and south of the plant intake, structures. The discussion was limited to three areas because the NRC had indicated that the com-paction piles placed under the barrier wall were acceptable. Ebasco attempted to demonstrate that the compaction piles placed in,the north and south slopes achieved the desired densification of the in-situ soils. It was noted that the pile program outgined in Ammendment No. 47 of the Unit 1 FSAR and supplement No. 2 of the Unit 1 SER did not agree. It was further noted that the piles were placed in accor-dance with the FSAR ammendement using pre-augering. Ebasco tried

0 to show that any pre-augering was not excessive and that the compaction piles were driven through the in-situ soils.

Ebasco also noted that the pile driving report dated January 1976

. contained several errors. One error was related to the procedure for pre-augering. Ebasco presented a memo dated October 22, 1981 which stated that only one pass with a auger was made for the full depth of a pile. Two additional passes were made only to the interface of the class II fill and the in-situ materials. Ebasco agreed to sub-mit an addenda to the report correcting this error and several other misleading statements.

The NRC stated that, based on the portions of the pile driving log they had reviewed, it did not appear that the compaction piles den-sified the in-situ soils. They noted several cases where the piles apparently droped through the in-situ soils under there own weight as a result of execessive pre-augering.

At this point itItwaswas suggested that agreed to end additional clarifying information the meeting and reconvene at a was required.

later date.

MEETING MINUTES ICC INSTRUMENTATION A meeting between FPL and the NRC was held on 10/23/81 for the purpose of closing open items associated with inadequate core cooling instrumentation, TMI Item II.F.2.,

The attendees of the meeting were:

Name Or anization Vic Nerses NRC Tai Haung NRC Vic Westoven C-E Remo Gritz FPL The following items were agreed to.be addressed in a revised FPL submittal:

FPL Revision will reference CENPD-181 and CENPD-185 for generic issues concerning the QSPDS.

2. Section 3.1 will be revised to reflect FPL commitment to input Core Exit Thermocouples into the subcooling margin monitor.

3. Section 3.2 will revised to better describe QSPDS and SAS.

4. Section 3.2.1 will be revised to include an alarm, set-point range. for the subcooling margin monitor.

5. Section 3.2.3 will be revised to include a description of the CET Processing Equipment.

6. Section 3.2.4 will be revised to provide trending information.

7. ---Section 5;0 will be "included in an expanded discussion on Equip-ment Qualification of out-of-vessel Class lE material (i.e.

cable, connectors, instrumentation in C.R.)

8. Section 6.2 will clarify FPL position on a Proto-type Test schedule.

9. Section 9.0 will be revised to discuss FPL commitment on imple-mentation schedule for ICC.

10. FPL will modify Table 1.9B 2:;:

Item 5: Revised information on the display will be submitted Item 8: Commitment on procedures before system operation be made. ' 'ill

11. FPL will provide discussions to resolve schedule issues in Item 9.

12. Table 1.9B-3 will be modified as follows:

a) Paragraph 2 will be expanded to address concerns from NUREG-0737 2a through f.

b) Paragraph 3 will be revised to incl'ude gSPDS c) Paragraph 7 will be revised to include discussion

. on qualification.

13. Table 1.9B-4 will be revised, where applicable to reflect changes made above, to include paragraphs 1, 4, 6, and 7.

Res onse to SER Open Item: Adequacy of Station Voltages Branch Technical Position PSB-1 requires that two levels of undervoltage protection be provided for the Class lE busses.

gn sooordenoe vith Pgg-1 the first 1evel og nndervoitsge proteotion is provided. to detect a loss of offsite power. One Type CV-2 inverse time voltage relay is provided for each Class 1E division and is set at time'ial 2 which will proivde undervoltage tripping in accordance with the re-tgr-ASH lay characteristic curv'e provided in Figure f. The tqp voltage vamp is set at 105VAC which produces a tripping characteristic of approximately 12 seconds at 79X v'oltage.

Each Class 1E division is provided with one'lass 1E relay as described above mounted in. the Class 1E 4.16KV switchgear." Upon detection of a loss of voltage condition this relay automatically initiates diesel gen-erator'tarting and disconnection of the'offsite source og a loss of offsite power.

Branch Technical Position PSB-1 requires that a second level of under-voltage protection be provided for the Class lE busses.'lordia Power and Light meets the requirements of the position for St. Lucie Unit 2 by providing for each Class lE division, a coincident logic protection scheme consisting of three definite time relays, ITE Type 27D or equiva-lent set at 92.5X of 4.16KV and provided with a 10 second time delay. The relay logic actuates control room annunication to alert the operator to a degraded voltage condition and align the~trip circuitry associated with

the undervooltage log j c such that subsequent occurance of a sa fety in) ection actuation signal s gna (SIAS) will separate the Class 1E system from the offsite power system automatically.

evaluate the acceptability of the relay setting an analysis of station electric system voltages was performed under steady state conditions with the full 'plant running leads and minimum design main generator voltage supp 1yingng the on site system through the Unit Auxilia Transformers.

The results of thi's analysis are shQn on Figure 2998-PSBl-A which demons-trate that voltages on Class 1E system at the 4.16KV level, the 480VAC level and the 120VAC level with the exception of PP247 remain above the design limits of the equipment.

The most limiting equipment .was considered to be the 460VAC motors rated at 90X 'of nameplate operating voltage. Inspection of Figure PSBl-A reveals that in this condition, on the worst case 480'AfCC the operating voltage remains above 90.8X of 480VAC which corresponds to 94.7X of motor nameplate voltage, safety above the 90X operating limit.

The voltage level on the 4.16KV busses remains above 94.5X of 4.16KV which insures that the. alarm and SIAS alignment relays described above are not picked up.during this steady state operating condition.

The minimum acceptable operating voltage" at the 120VAC level was established by equipment ratings to be 90X of 120VAC. Inspection of Figure PSB.-1A revelas that for all 120VAC panels 'except PP247 the operating voltage remain above the 90X limit. The unacceptably low voltage of PP27 will be corrected prior,to plant operation by load redistribution or other means such that under the defined operating conditions voltages on PP247 and ill 120VAC

ower panels remain above 90'f 120VAC.

f Analysis o stat ti t onn electric e ec system voltages was also performed under-steady state conditions with the full plant running load aud minimum .

design sid,ttch yarrd voltage vo supplyiug the ousite system through the Start u Transformers. The results of this analysis are shown on Figure 2998-PSB1-B which demonstrate that as shown in the analysis for the Unit A~ilary Transformer all voltages on the Class 1E system with the excep-tion of 120VAC panel PP247 remain above minimum acceptable design conditions.

Modification to panel PP247 will'be made as previously committed to insure all voltages remain above the acceptable minimum.

The worst case starting transient was also analyzed for the most limiting conditions which are on the 2A system since this is the most heavly loaded with all normal plant'running loads on the busses, when the startup trans-former is supplying the 2A system and ofgsite switchyard voltage is at the design minimum of 230KV. The results of this analysis is provided in Figure 2998-PSBl-C. The analysis indicates that following the starting transient voltages on all Class lE busses remain at values above the acceptable design limits and that the voltage on the 4.16KV busses returns to above the relay'setpoint of 92.5X within the timer setting of 10 seconds.

In accordance with PSB-1 relay actuation during the worst case motor starting transient does not, occur.

An additional analysis was performed on the onsite system to evaluate the impact of an SIAS and resultant fast dead bus transfer when the offiste s'ource is at the minimum design voltage conditions. The results of this

analysis are provided in Figure 2998-PSBl-D which demonstrate that the voltages on the 4.16KV level and 480VAC level and 120VAC level remain with acceptable design limits following the fast dead bus transfer.

The 10 second time delay is based on preventing the worst case motor starting transient, which is 'the 4.16KV condensate pump, which accelerates to full speed .upon the minimum voltage conditions expected o5 the main generator or switchyard in 6 seconds, from causing sperious.alarms in the control room.

The relays and all assocated equipment will be Cl'ass 1E and will be located in the Class 1E switchgear.

Capability for test and calibration of the relay scheme during power operation will be provident.

The above scheme meets the requirements of Branch Technical Position PSB-1 section B.1.5)1).

To meet the requirement of Branch Technical Position PSB-1 section B.l.b)2) a second set of ITE-27D or. equivalent definite time relays will be provided in a coincident logic arrangement for each Class 1E division. These relays will be set at 90X of 480VAC and located downstream of the 480VAC powercenter 2A5 and 2B5 reactors. The output of the relay logic will enable a CV-2 inverse time relay having a time voltage characteristic shwhn in Figure 2998-PSB1-E. The inverse time relay will separate the Class 1E system from the offsite source in accordance with the selected time dial setting should the operator fail to restore system voltages.

The setting of the definite time coincident logic relays at 90X of 480VAC insures positive operation within the defined region of the operating curve of the inverse time relay. The operating characteristics of the inverse thne relay assures that under the worst "case starting transient which is the 4.16KV condensate pump, when generator voltage is at its expected minimum or switchyard voltage is at its expected minimum inadvertent relay actuation will not occur..

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It can be seen from Pigure 2998-PSBl-E that with.a time dial setting in the range of 2 to 4 and with tap voltage selected such that a 90X tap voltage value corresponds to 90X of 480VAC~ 4e condensate pump starting transient which dips the 480VAC bus voltage to a minimum of 80X with a recovery to 91.7X within 6 seconds, with generator voltage are at their absolute expected minimums, will not produce inadvertant actuation of the protection feature.

Under steady state conditions with normal plant operating loads and with the switchyard or generator voltages at their expected minimum, 480VAC bus voltage remains above 90.6X of 480VAC which is above the setpoints of the definite time relays, preventing sperious actuation of the protection feature during steady state conditions.

Should 480VAC bus voltage decrease to 90X of 480VAC, voltage at the worst case motor control center will be at 90X of 460VAC or 86.2X of 480VAC.

The definite time logic relays will pick up emergizing the CV-2 relay which will transfer the Class 1E busses to the Diesel Generators in approxi-.

mately 20 seconds. Should the 480VAC bus voltage continue to decrease the inverse time function of the relay will shorten the time to trip in accordance

""with the time dial setting selected.

The relays and all associated equipment are Class lE and will be in the Class 1E switchgear. The capability for test and calibration during power operation will be provided.

Plordia Power and Light will provide test verification of the analysis peerformed,to establish adequate station electric system voltages prior to fuel loading optimum relay setpoints be determined and based on the test results.

"TRAINING FOR MITIGATING CORE DAMAGE" POWER PLANT THERMODYNAMICS (45 Hrs)

A. THERMODYNAMIC CONCEPTS

1. The Nature of Thermodynamics Introductio'n to the Laws of Thermodynamics

3. Thermodynamic System,.Property, State, and Process

,.4. Density, Specific Weight, Specific Volu'me, and Pressure .

5. Temperature and.the Zeroth.Law

6. Ideal-.Gas Equation of State

7. Hoyle's Law

8. Charle's Law

9. Combined 'Gas Law B. WORK AND ENERGY

1. The Concept of Work and the Adiabatic Process

2. Heat and Heat Units.

3.. P-V Work

4. The Nature of E, the Total Energy

5. Internal Energy (U)

6. The Conservation of Energy Principl'e for Simple Compressible Closed Systems

7. The First Law'f Thermodynamics 8.'he Gene'ral Energy Equation

.9. Entropy, The Second Law of Th'ermodynamics C. PROPERTIES OF WATER AND STEAM

l. -

Water and Steam Properties

2. Saturation Temperature and Pressure

3. Specific Volume

4. Quality

5. Wet Steam

6. Internal Energy

7. Enthalpy

8. Specific Heat

9. Entropy.

10. Superheated Steam Tables

. D. THERMODYNAMIC PROCESSES

1. T-S Diagram'.

Mollier Chart

3. Throttling Process for Steam 4; Flashing Water to Steam

. 5. Isentropic Processes on Mollier Chart

'(CONTINUED)

E. THERMODYNAMIC CYCLES 1.. Carnot Heat Engine

2. :Thermodynamics of 'a Basic Gycle

3. Pressure - Specific Volume Diagram 4;'ycle Diagrams

5. Rankine.'Cycle and Efficiency

6. Thermodynamic Cycles

7. Tuibine Work

8. Heat Rejected

9. Work

'0;

-ll.

Pump Reactor .Heat. Input "-

Thermal Efficiency

12. 'Feed Water Heating

13. Cycle Analysis

14. Turbine Work

15. Heat Rejected 16..Pump Work'7.

Feed Heater Pump Work

18. Reactor'Power

19. Thermal'fficiency with Feed Water Heating II. HEAT TRANSFER AND FLUID MECHANICS '(45 Hrs)

A. SPAT TRANSFER Conduction Principles and Application

a. ~ 'Specific Heat

b. Thermal Conductivity

c. Electrical Analog 2~ Heat Flux

'3. Convection 4, Radiative H eat Transfer B. HEAT EXCHANGERS

1. Counterflow

a. Principl'es .and 'Problem Solving

b. Log M or D Temperature Difference (LMTD)

2. Parallel Flow

3. Appli'cation to a PWR Steam Generator

a. Latent Heat of Vaporization

(CONTINUED)

C. CONDENSERS AND PEEDWATER HEATERS

1. Principles and Problem Solving D. 'VAPORATORS E. REGENERATIVE AND .NON-REGENERATIVE HEAT EXCHANGERS Application to Pg CVCS Systems

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Q)

P. HEAT LOSSES G. REACTOR POWER H. SECONDARY PLANT CYCLES

1. Heat Balances

2. Efficiencies I. BOILING REGIMES

1. Nucleate Boiling

2. Critical Heat Plux

a. DNB J. THERMAL,STRESS

1. Effects"of Heatup and Cooldown rates

.K. FLUIDS

1. Pressure

a. Hydro Static

b. Measurement Devices

1. Principles

2. Modes of Failure

2. Density', Spec. Volume, Spec. Gravity III ' MITIGATING CORE DAMAGE (9 CZARS A. DEGRADED CORE.

l. Introduction

a. Design Principles

b. Safety Function Concept

c. Defense in Depth

2. Typical Scenarios

a. LOCA

b. ~4'.

What is problem

2. Present state of the art

III. {CONTINUED)

B. DEGRADED CORE/SYSTEM RESPONSE,

1. Xntioduction

2. Inade q uate Core Cooling (I. C.C.)

a. Small b'reak LOCA Scenario

1. Sys'em Parameter. Responses

2. Variable with break s5.ze and operation of RCP's

b. Steam line break

c. ATWS

d. Indications of approach to I.C.C.

LOCA LOSS OF. HEATSINK

e. Recovery Large break LOCA C. TMI'CCIDENT SCENARIO

1. What Happened

a. Time Relationship

l. Operator Action 2.. Plant Status

b. Report Findings

2. What ifV D. ,DEGRADED CORE INDICATORS..

r

l. In-'Core D'etector Response

2. Ex-Core Detector Response

3. Radiochemistry Results

4. Radioactivity Levels

-. ~ =5.'. PSL Instrument Predicted Response

a. Validity of Instrument Response

b. Post LOCA Lifetime E. CORE MELTDOWN-

1. Sequence of Events

2. Melting Mechanis'ms 3; Vessel'ailure

4. Steam Explosions 5.. Interaction with Concrete and Soils F CONTAINMENT RESPONSE I. Design Criteria

a. Heat Sources

! b. Heat Sinks '

XXX: F. (CONTINUED)

2. LOCA Response Spray System

b. Fan Coolers

3. Hydrogen

a. Flamability and Detonation

b. St'earn, Air; Hydrogen Mixtures

4. Hydrogen Contermeasures

5. Containment Augmentation 6.'ummary.

IV.. ACCIDENT AND TRANSIENT ANALYSIS ..

A. TRANSIENT ANALYSIS (20 Hrs) ~

1. Detailed Disc'ussion of FSAR Analyzed Transients

a. 'Initiating Conditions

b. Plant and Systems Response

c. Off'ormal Procedures

2. In'tegrated Plant Response B. 'CCIDENT ANALYSIS (20 Hrs)

1. Detailed Discussion of FSAR Analyzed Accidents

a. Initiating Conditions'.

Analysis Assumptions

c. Plant and. Systems Response

d. Plant Emergency Procedures

2. Integrated Plant. Response C. ACCIDENT IDENTIFICATION

1. Available Indications

2. Determination of Incident using Available Indications.

FLORIDA POWER & LIGHT COMPANY ST. LUCIE UNITS 1/2 SYSTEM DIFFERENCE/REgUALIFICATION TRAINING OUTLINE Prepared by:

A. SPODIC Approved by: ', July 31; 1981 P. L. FINCHER

PRELIMINARY TRAINING SCHEDULE WEEKS 1 THRU 5 .Unit 1 and 2 System'omParisons and Differences Training.'EEKS 6 and 7 Reactor Theory Review .

Thermo dynamics Health Physics Others WEEKS 8 and 9 Procedures Transient and Accident Analysis WEEKS 10 THRU 12 Review System Tracing

WEEK 1

-- Rx Coolant System '(Including. Pressurizer and Quench Tank)

Rx.Vessel and Core Design CVCS and Boron Concentration Control Rx Protection System and Nuclear Instrumentation Reactor Regulating System Pressurizer Level and Pressure Control Control Element Drive Mechanism Control System RTGB Comparisons (RTGB 203; 204; 205)

E ergency Core Co ling Systems High Pre ssure Safety Infection b>) Low Pressuxe Safete y Inn)ection.an'd e Shutdown Cooling c) Safety Infection Tanks 4

Containment ra and.Xodi.ne Removal S pray S ys tern Containment Cooling System Component Cooling Mater System Intake Cooling Water and Lube Rater Diesel Generator and Auxiliaries Engineered Safe'guards s Actuation Systems Fuel Pool C ooling and Purification

- Ventilation Systems a) Shield Building Ventilation b> Control Room Ventilatiat on and Emergency Cleanup c) ECCS Area Ventilation.

d>) 'n Fuuel Pool and Fuel 'Handl ing Building Ventilation e) Intak ake Stxucture Ventilation RTGB 206 and HVAC Control Panel

WEEK 3 Main and Auxiliary Steam Extraction Steam Steam Bypass Control System Condenser and Circulating Water Feed and Condensate Heater Uents and Drains Turbine Cooling Water

~ Control Oi'1 Systems

-. Den and Reheater Controls Main Generator and Controls Xnstrument and Station Air

--Feedwater Regulating System Auxiliary Feedwater System

- RTGB Comparison (RTGB 202)

Steam Generator Blowdown System Pire and Domestic Water

,- Waste Management Systems Primary Sampling

- Secondary Sampling Condenser/Leakage Detection Systems Incore Instrumentation DDPS & Sequence of Events Recorder Radiation Monitoring Systems Miscellaneous Control Room Equipment

Electrical Distribution Miscellaneous Uentilation Systems Compressed Air Systems System Interties and Shared Systems Fuel Handling Miscellaneous Instrumentation Systems RTGB 201 Comparison

WEEKS 6 & 7

- Rx Theory Review

. Plant Operating Characteristics Thermodynamics Heat Transfer and Fluid Plow Electrical (Generator Theory)

Chemist'ry Health Physics

~ e ~

gEEKS 8 & 9

- Procedures a) Normal

~

b) Off-Normal c) Emergency d) Administrative e) E-Plan Implementing

- Transient Analysis

- Accident Analysis

WEEKS 10 12 Review and Self-Study System Tracing

EXAM SCHEDULE QUIZZES: 1. Week 83

2. Week 85

3. Week 88

4. Week i0'10 FINAL WRITTEN EXAM: Week 812 FINAL ORAL EXAM: Week 812

rage Revision 3 FLORIDA POtKR h LIGHT CO.'fPATA'T.

LUCIE PLANT iPilIT NO. 1 ADMINISTRATIVE PROCEDUPZ NO. 0005721

l. 0 Tin la:

'LICRiSE OPERATOR TRAINING PROGKQi

'. EHore use, HOT This d Umen

' t.

verify informabo vrith a controlled docurnen

3. 0 ~Approval:

Reviewed by 'the. P t Nuclear Safety Connittee ~

I':,I

'l +

19 Approved by 'V/ ~<w .Plant Manager. 19 .-

c l I Revision t Reviewed by Plant Nuclear Safety Committee I I Approved by - Plant ~anager 10 ~

Revision 3 ".eviewed by FRG lo8>

Approved ~y~ '

8 <<z .lant Manager Q~o

3.0 Scone

3.1 Purpose 1 3.2 Discussion

4.0 Precautions

None 4&

5. 0 Responsibilities:.

5.1 The training staff shall be responsible or:

5.1.1 Coordination and conducting of this orogram.

5.1.2 Lesson .Plan preparation and presentation.

5.1.3 Replication to audits conducted upon this program.

5.1.4 Evaluations and examinations as a part of this program.

5.1.5 Reports to Management.

5.1. 6 Coordination of screening examinations and all other vendor contracted training functions.

5.1.7 Provide required documentation for each Hot License Candidate.

5.2 The Training Supervisor shall be responsible for coordination of operator scheduling with the Operations Department.

5.3 Hot License Candidate students shall be responsible for:

5.3. 1 Attending lectures, examinations, d iscussion as scheduled.

5.3.2 Making changes to appropriate. documents as required.

0 0

Page 2 .of 5 Revision 3 ADMINISTRATIVE PROCEDURE NO. 0005721 HOT LICENSE OPERATOR TRAINING PROGRAif 6.0

References:

6.1 American National Standard Institute 18.1 6.2 F.S.A.R. 13, Part 13.2.3.

6.3 10 CFR 50 R3 6.4 10 CFR 55 R3 6.5 QI 2 PR/PSL-2 R3 6.6 Letter from H. R. Denton to all power reactor applicants and R3 licensees, dated March 28, 1980.

Subject:

'Qualifications of Reactor Operators.

7.,0 .Records and Notifications:

All records and notifications shall be in accordance with Indoctrination apd Tryining of Plant St. Lucie Power Resources Personnel, "Quality Instruction 2 PR/PSL-2".

rage a nr Revision 3 ADHINISTRATIVE PROCEDURE NO. 0005721 HOT LICENSE OPERATOR TRAINING PROGRAM

8.0 Instructions

8.1 This program should be accomplished in no less than 12 training months.

8.2 The St. Lucie Hot'icense Operator Training Program shall encompass the following outline:

1. Theory and Principles of Operation

2. General and Specific Plant Operating Characteristics

3. Plant Instrumentation and Control Systems

4. Plant Protection Systems

5. Engineered Safety Systems

6. Normal, Off-Normal, and Emergency Operating Procedures

7. Radiation Control and Safety,

8. Technical Specifications and Administrative Controls

9. Applicable Portions of Title 10, Chapter I, Code of Federal Regulations

.10. Thermodynamics, Heat Tr'ansfer, and Fluid Flow ll. Mitigation of Accidents Involving a Degraded Core 12., Fuel Handling and Core Parameters 8.3 Recommended Program:

8.3.1 Phase 1: Theorv 8.3.1.1 Nuclear Power Orientation 8 '.1.2 Basic Nuclear Concepts 8.3.1.3 Reactor Operation 8.3.1.4 Plant Performance: Thermodynamics, Heat Transfer and Fluid Flow 8.3.1.5 Radiation Protection 8.3.1.6 Waste .and Water Treatment 8.3.1.7 Instrumentation 6 Operational Analysis

Page 4 of 5 Revision 3 ADMINISTRATIVE PROCEDURE NO. 0005721 HOT LICENSE OPERATOR TRAINING PROGRAM 8.0 Instructions cont.):

8.3 (cont.)

8. 3.2'haae II: ~Sstems 8.3.2.1 Nuclear Steam Supply Systems Design and Operational Theory 8.3.2.2 Secondary Plant Systems Design and Operational Theory 8;3.2.3 Instrumentation Systems Design and Operational Theory f

8.3.2.4 Administrative Controlsp T~hnical Specifications, Procedures, Emergency Plans, Plant Transient Response.

8.3.3 Phase III: Practical 0 erations 8.3.3.3. Shift Operations:

8.3.3.1.1 Observation of Daily Operations 8.3.3.1.2 . Operational Practice 8.3.3.1.3 Systems Field Study 8.3.3.1.4 Systems Checkouts 8.3 .3.2 Simulator Training:

8.3.3.2.1 'eneral Operations Prac'tice 8.3.3.2.2 Plant & Reactor Startups Practice 8.3.3.2.3 Emergency/Casualty. Operations Practice 8.3.3.2.4 Operating Exam & Startup Certification 8.3.3.3 Cumulative Practical Operations Time' 3 months..

8;3.4 Phase IV: Review and Examinations 8.3.4.1 Course Review (40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> min.)

8.3.4.2 Pre-Licensing Written Examination similar in length and content to NRC Licensing Examinations 8.3.4.3 Pre-Licensing Oral Examination similar in length

--.and content to NRC 3 icensing Examinations

ADMINISTRATIVE PROCEDURE NO. 0005721 HOT LICENSE OPERATOR TRAINING PROGRAif 8.0 Instructions (cont.):

8.3 (cont.)

8.3.4 (cont.)

8.3.4.4 Periodic written and oral examinations shall be given throughout the program to gage the candidate's progress and overall performance.

8.4'hanges to the specific content and schedule of training program may be made at the discretion of the Training Supervisor.

8.5 For License Candidates with previous 'nuclear experience, applicable portions of the training program may be deleted based upon evaluation of the individual's .background.

~ ~

FLORIDA POWER. & LIGHT CO."fPANY ST. LUCIE UNIT 51 ADHINISTRATIVE PROCEDURE NO'005720, REVTSION 8 vv ~v

1.0 Tlrle

Licensed Operator Requalification Program COPY 2.0 ~Aroval:

Reviewed by PNSC '

kl g >P "

J 3 19 7~

Approved by '.k Af H a <,'S . Plant ?ianager +bi w yP. ~d 19 '7W Rev. 8 Reviewed b Facility Review Group er k/ 19 PO Approved by / ~~~ ~ant tanager 19 0 3.0 ~Sco e:

3.1 ~Pur oee:

This procedure provides instructions for conducting a requalification program which will assure that licensed operators and senior reactor operators maintain their proficiency.

3.2 .Discussion:

Indiv'iduals actively engaged as licensed operators or senior operators are required by federal regulations to participate in a USNRC approved requalif ication program.

-Non-shift 'personnel who hold licenses and desire to keep their licenses effective shall participate'ully in the program except for off-shift sessions devoted'o subjects where their specialized knowledge precludes the need to. attend.

As individual licenses near the expiration date, participation in the program will be reviewed, evaluations of performance will be

'analyzed, deficiencies will be rectified and a request for license renewal will be submitted.to'the USNRC.

The program will repeat on a continuing one year cycle, and in brief, will consist -of:

3.2.1, Off Shift Training which will consist of a series'of lectures and examinations, and.may also include discussion sessions and simulator training ~

/R8

Page 2 of 1 ADttINISTRATXVE PROCEDURE NO. 0005720, REVISION 8 LICENSED OPERATOR R UALIFICATION PROGRAtl 3.0 ~Sce e: (ccncinned) 3.2 (continued) 3.2.2 On Shift Training wherein plant changes/modifications, oper'ating procedure revisions and reportable occurrences are discussed,'nd performance under both simulated and actual off-normal and emergency situations is evaluated. The required control manipulations may be performed either at a simulator or as a part of this on shift training.

~*

10 CFR 50.54 (i-l), 10 CFR 55.31 (e) and Appendix A to 10 CFR 55

3.4 Definitions

3.4. 1 Licensed 0 erator or Licensee An individual who holds an effective USNRC operator or senior operator license for St.

Lucie Plant and who intends to maintain his license current.

3.4.2 Lecture Se ent t&y consist of several lecture sessions covering a particular topic or category.

3-4-3 Off-Site Licensee A licensee who is:

1 Not permanently assigned to the St. Lucie Plant, or

2. Assigned to the St. Lucie Plant but who has been absent from the site for greater than 90 days at a time.

3'.4.4 Trainin De artment Staff The training organization, consisting of the Training Supervisor, and Licensed Instructors who are permanently assigned to the department.

4.0 Precautions

None 5.0 Res ns ibilities:

5.1 The Trainin Su ervisor:

5.1.1 From the results of prior examinations, evaluations and suggestions by licensed operators, determine topics to be included in regularly scheduled lectures. Prepare the material and give lectures on'these specific topics or appoint a competent individual to prepare and give lectures on specific topics.

/R8

Page 3 of 16 ADHINISTRATIVE PROCEDURE NO. 0005720, REVISION 8 LICENSED OPERATOR R UALIFICATION PROGRA?f 5.0 Res onsibilities: (continued) 5.1 .The Trainin Su ervisoz': (continued) 5.1 ~ 2 Prepare examinations on lecture topics and an annual operator and senior operator NRC type examination.

5.1 3 Schedule off.-shift lecture and/or 'discussion sessions on

." topics suggested by licensed operators or of. current interest, such as plant changes and modifications, plant operating problems, reportable occurrences and plant operating procedures. Mhere deemed advisable, appoint an individual fully conversant with the topic to lead the dis'cussion.

5.1.4 ArrAnge for c'ompetent examiners'o perform 'oral evaluations as necessary.

5 1.5 Review the results of written exams'and evaluations and, where required, provide intensive 'individual training.

1 5.1.6 Haintain the records required to document the training given

( 5.2 The 0 each licensee.

erations Su ervisor Nuclear:

'Shall coordinate with th'e Training Supervisor to prepare a schedule for off-shift.'raining sessions.

5.3 The Nuclear Plant Su ezvisor in char e of each shift:

5~3~1 Shall be familiar wi'th the performance of operators on his shift.= He shall, on the form provided, evaluate the performance of members of his crew during =actual off-normal/emergency situations and discuss the evaluations with the individuals 5.3 2 Cogies of all documents distiibuted pursuant to Section 7.1 will be, posted in a'inder provided zor their receipt. Each Nuclear Plant Supervisor shall ensure that each member of his crew reviews these documents and'nitials the acknowledgment section of the cover letter associated with these, changes.

( /R8

Page 4 of 16 ADMINISTRATIVE PROCEDURE NO. 0005720, REVISION 8 LICENSED OPERATOR RE UALIFICATION PROGRAH 5.0 Res onsibilities: (continued) 5.3 (continued) 5.3.3 An individual. log of'control manipulations, for licensed crew

. members shall be kept. The licensee is to make entries which shall. be validated by the Nuclear Plant Supervisor, Operations Supervisor, Operations Superintendent, or Training Department Staff 5fember. A't intervals, the supervisor is'o review the logs of his shift members to assure that the required manipulations will be performed within the

. appropriate cycle.

5.4 Each Licensee:

'.4;1 Attend required off-shift lectures, discussion sessions, and any secheduled .simulator training, and take schedu1ed written examinations and evaluations.

5.4.2 1faintain his control manipulation log.

5.4.3 Ensure .that he understands documents distributed pursuant to Section 7.1 and signify his understanding by initialing the acknowledgment letter'ssociated with these changes.

5.4.4 Participate in special training programs where writt'en

'- examinations and evaluations indicate the rice.

6.0

References:

.6.1 10 CFR 50.54 (i;1), Conditions of Licenses 6.2 10 CFR 55.31 (e), Conditions of Licenses 6.3 10 CFR 55, Appendix A, Requalification Programs for Licensed Operators of Production and Utilization Facilities 6.4'etter fiom Harold R. Denton to all power reactor applicants and licensees, dated Harch 28, 1980.

Subject:

Qualifications of Reactor Operators..

7.0 Records and Notifications:

7.1 Trainin Re ort

. A copy of each plant change/modification (or a summary of the plant

'hange/Hodification),'eportable occurrence, FSAR" supplement, faci1ity license amendment, and applicable procedure change shall

/R8

Page 5 of 16 ADlENISTRATIVE PROCEDURE NO. 0005720, PZVISION 8 LICENSED OPERATOR R UALIFICAT'ION PROGRAH 7.0 Records and Notifications: (continued) 7.1 Trainin Re ort (continued) be sent to the Nuclear Control Room, (for shift operators) and to the Training Department (for non-shift operators). These locations should maintain these documents in loose leaf binders, designated for'he receipt. of Training Reports. Each licensee shall review thepe documents and shall so .indicate by initialing the acknowledgment section of the Training Report Cover Letter associated with each posting. The canpleted training report cover letters shall be filed in Document Control. (See Figure 1).

7.2 Control Hani ulation Lo Each licensee shall maintain his own log; 'The, logs for shift operators will be maintained in the Nuclear Plant 6upervisor's office. The logs for'on-shift operators'will be maintained in the Training Department.. At the completion of each cycle, the log will be placed in the licensee's Requalification File. When the control manipulations take place at a simulator, the word "simulator" may be entered in the signature column, and the documentation provided by tQe simulator facility attached to the form. (See Figure 2 and 3).

7.3 Lo of On-Watch Dut Off-Site Licensed Personnel:

This log will be displayed in a conspicuous location. Its purpose is to enable both the licensee and the Training Supervisor to assure that shift operation requirements are net. The logs need not be retained. Refer to Figure 4.

7.4 Records Relati to Off-Shift Traini 7.4.1 Posted Schedule:

To be displayed'n conspicuous locations and shall show:

Shifts on Training Schedule Dates of lecturess and the subject matter Dates of procedure review and prcedures to be discussed Dates of discussion sessions and the topics to be discussed Dates of scheduled examinations Dates of any simulator. training

Page 6 of 16 ADMINISTRATIVE PROCEDURE NO. 0005720, REVISION 8 LICENSED OPERATOR R UALIFICATION PROGRAH 7.0 . Records and Notifications.: (continued) 7.4 (continued) 7~4~ 1 (continued)

Multiple sessions will be required and the schedule shall be posted. sufficiently in advance,to make the schedule-arrangements necessary to allow licensees. to attend a session.

Completed schedules shall be kept in the requalification program' ile.

~

7.4.2 Attendance Records:

The lecturer or discussion leader shall maintain a record of attendance. lfultiple schedu1ing of sessions should assist in assuring that all licensed operators can attend. Where circumstances cause an absence, the necessity of make up shall be determined by the Training Supervisor. The Attendance Record is to be kept in the Requalification Program File.

7.4.3 Written Examina'tion:

The questions, together with the answers for examination on lecture topics and the annual OP, SOP examination together with the grades made by each licensee shall be kept in the Requalification Program File.

7.5 A licable NRC Bulletins Circulars and Information Notices Copies of the applicable correspondence shall be routed to the Training Department, posted in the binder provided, and logged on the cover letter.'ach member of the Training Department Staff shall review these documents and shall so indicate by initialing the cover letter for the applicable month. Completed cover letters shall be filed in Document Control. (See Figure 5).

7.6 Lo Revi'ew Sheet Each member of the Training Department Staff shall maintain his own sheet. The sheets will 'be maintained in the Training Department.

Control Room Logs (NPS and RCO) shall be reviewed by the Training Department Staff at regular intervaLs not exceeding one month, and review documented by initialing this sheet.. When the sheet is 'his completed, it shall be placed in the individual's Requalification File. (See Figure 6).

/RO

Page 7 of 1t ADMINISTRATIVE PROCEDURE HO. 0005720, REVISION 8 LICENSED OPERATOR R UALIFICATION PROGRAM 7.0 Records and Notifications: (continued) 7 7 Re uest for License Renewal As the date for license renewal approaches, the Training Supervisor shall review program participation to insure eligibility and prepare and submit applications for license renewal to the USNRC.

8.0 Instructions

8.1 Off-Shift Trainin 8.1.1 A minimum of 160 hours0.00185 days <br />0.0444 hours <br />2.645503e-4 weeks <br />6.088e-5 months <br /> of off-shift training shall be scheduled for each operating shift. 'This training may include discussion sessions and simulator training, but as a minimum shalk consist of lectures covering .,topics which fall under general'eadings as listed.

1 Theory and Principles of Operation

, 2. General and Specific Plant Operating Characteristics

3. Plant Instrumentation and Control Systems Plant Protection Systems

'5 ~ Engineered Safety Systems

6. Normal, Off-Normal, and Emergency Operating Procedures

7. Radiation Control and Safety

8. Technical Specif ications

9. Applicable portions of Title 10, Chapter I, Code of Federal Regulations

10. Thermodynamics, Heat Transfer, and Fluid Flow ll. Mitigation of Accidents. involving a Degraded Core 8.1.2 When discussion sessions are scheduled to fulfillany off-shift training requirement, a member of the Training Department Staff, the Nuclear'Plant Supervisor, or an individual-'ell versed in the topic shall be designated as-leader of the session.

page 8 of 16 ADHINISTRATIVE PROCEDURE NO. 0005720, REVISION 8 LICENSED OPERATOR R UALIFICATION PROGRAM

8.0 'nstructions

(continued) a 8~ 1 (continued) 8.1 3 Hhen a simulator is used, for off-shift training, one hour of simulator t9me shall count as two hours of training time for the purpose o fulfilling the min9mum time requirement of Step 8. 1. 1 8 '.4 The results of the annual examinations shall be analyzed to determine subjects to be emphasized. Additionally, 'any plant.

or ind us txyxy problems which develop and are deemed as useful

.information for the opePetor shall be discussed; underlying theoxy, nuclear, mechanical, or thermodynamics shall be taught to the depth required.

8.1.5 Each shift shall be rotated from the operating shift schedule to the training shift schedule in order to pxovide off-shift training.

.8.1.6 The'majority of the off-. shift tra'ining will be conducted by

'he'raining Department Staff. /here the specialized k 1 d f in individual outside the training organization is available or desirable, this individual may conduc t thee

".session.

8.1.7 ~

The Training Department Staff. shall participate in any simulator training which is conducted as part of this progxam.

8.2 On-Shift Trainin 8.2.1 Control Hani ulations All licensees, within the cycle. indicated, shall, where

• possible, perform or direct. the activities listed on .the Control Manipulations Logs (Figures 2 and 3). ese manipulations may be performed either on the plant or a imul t %hen'ctual performance of certain manipulations i's not .possible, oral examinations shall be administere ~ n accoxdance with step 8.3.3. Those manipulations indicated by an asterisk (*) on 'the log shall be performed either on the

-p 1 antt or r a ssimulator.

a o Oral examination is not sufficient to

@eet the requirement for these 'manipulations- See Figu res 2 and 3 for details.

/R8

Page 9 of 16 ADMINISTRATIVE PROCEDURE NO. 0005720, REVISION 8 LICENSED OPERATOR RE UALIFICATION PROGRAM I

8 0 Instructions: (continued) 8.2 On-Shift Traini (continued) 8.2.2 Trainin Re orts Training Report Cover Letters shall accompany 'documents distribut'ed pursuant to Section 7.1. Each shift operator should review and initial the Cover Letter acknowledgment section within t9me specified. See Figure 1'or details.

8.2.3 Of f-Site Licensee

1. Each off-site 3icensee shall participate fully in the .

Licensed Operator Requalification Program. In addition, each off-site license'e shall stand periodic control room

, watches in order to maintain his proficiency as a watch stande'r The minimum requirement is 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of watch standing within any three I

month period.

e 2 Periods exceeding 4 months without actively performing the functions of a licensed operator shall require' demo'nstration of competency prior t.o resuming licensed activities by passing a written and'ral examination. To the extent practicable, the licensee is to stand the watch when major evolutions are taking place. Refer to Figure 4.

8 2 4 Trainin De artment Staff

1. The Training, Department Staff shall participate fully in the On-Shift Training portion of this program.

2. In addition, the Trainigg'Department Staff shall review the following::

(1) Applicable NRC Bulletins, Circulars, and Information Notices. This review shall be documented by initialing the cover letter accompanying these documents. See Figure 5.

(2) Control Room Operating Logs (NPS and RCO). This review shall be documented by initialing the Log Review Sheet at one month intervals. See Figure 6.'

3 Examinations 8.3.1 Periodic written quizzes shall'e administered following lecture segments.

/R8

page 10 of 16 ADHIHISTRATIVE PROCEDUPX NO. 0005720, REVISION 8 LICENSED OPERATOR RE UALIFICATION PROGRAM

8.0 Instructions

(continued) 8 ' 'xaminations 'continued)

.3.2 Qritten examinations similar in scope andd de dep th to NRC written examinations shall be administered on an annual basis to eac h licensed c individual. The Reactor Operator E i ation shall be completed within 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br />, an d thee Senior Reactor Operator Examination shall be completed w t n ithin 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />' 8.3.3 Oral examinations shall he conducted on control manipulations

~

not performed on the plant or a simulator.

.3.4 The Training Department Staff shall not be required. to take the Examinations since the staff administers these i

examinations.'y Th shall s 'owever, demonstrate and'maintain

~ their proficiency 'by. preparing, conducting, and evaluating training programs designed to maintain." or increase roficiencies of licensed operators and license carididates, and in addit ion, by fulfilling the requirements of Step 8.2.

8.4 Evaluation Critex'ia 8.4 1 A scoxe of .less, than 80Z on periodic quizzes administered in accordance with Step 8.3 1 shall requixe further study and completion of 'an additional quiz within 60 days of the pxevious quiz date.

8.4.2 The following criteria shall apply to written examinations administered in accordance with Step 8.3.2.

1 A score of less than 80Z on 'any section of the i i exam nat on s hall a r require mandatory attendance sessions covering material falling under thatt sect at lecture section on inn the following requalification pxogram year

2. A score of less than 80% overall, or less than 70Z on any section of the annual examination shall require that t e individual be placed in an intensive training program until the responsible'upervisors are satisfied that" the individual is again 'prof icient. Prof iciency shall be demonstrated by successfully passing another written examination. The licensee 'shall be relieved of duties requiring a license until deemed proficient.

r 8.4.3 Unsatisfactoxy per oxmance on control manipulations perfoxmed in accordance with Step 8.2.1 shall require further study and completion of an additional evaluation withih 60 days of the previous evaluation date.

/R8

~ ~ ~ IJUI'll i I FIGURE 1 LICENSED OPERATOPS TRAINING REPORT COVEP. LETTER OUARTER: ITEN IOENT'IF ICATION NUiMBER:

1st. ( JAN-s 1AR) 1 - Plant Change/iiodirication 2nd. (APR-JUN) 2 - Rep'ortable Occurrence.

3i d. (JUL-SEP) 3 FSAR Supp'fement 4th. -(OCT-OEC 4 - Operating License Arrendrent (Tech Spec Chang ) .

t)EEK NO. (1- 13) . 5 -. Operating'Procedure Change DATE DISTRISUTEO 6 - 0'ther O. IOENTIF ICATION REV. NO./OATE ITEit NO. IOENTIFICATION PEV. NO. /OAT"

'~I'ave reviewed and.discussed the data f'urnished by this cover letter, and hav so

'<icated by initialing below.

'. INITIALS/ DATE:

'NITrALS/

Nm~f -

. r,WE OATE

/

/

/

/

/

/

/

/

/

/

/

/

/

/

/

/

" TE TO LICENSEE AS= REV IEH ANO ACK iO'ILEDGE THE CHANGES ID N IF:"0 ON THIS: COVER LETTER '1ITHI AFTER THIS TRAINING PEPORT IS POSTED. 'AYS

• r< be filled by'rai.ning Dept.

~ ~

~ I

~ ~ ~ ~

~ ~

~ ~

~ ~ 1 I ~

~ ~ ~ ~ ~ 0 ~ ~ ~ V~ ~ I ~ le ltt ~ ~ N ~

FIGURE 2

~ t

~ ~

CONTROL HAI'JIPULATION LOG ~

a ~

AIIi'IUAL ~ I

~ 'rom To.

Hara: ~ ~

'icense SAT/UNSAT tto-OATE SIGNATURE

.actor Startu to Nuclear Heat "t)anval 5 G Level..Control Ourin S/U.or S/0

• po<;~r Chanae ) 105 in i'tariual Rod Control Loss of Reactor Coolant Flc>g Loss of All FeedMater Normal and Em rcenc Exc. RCS Leakage Incl. Rate Oeteraination Steam Cenerator Tube Ru ture Loss of Reactor Coolant:

Inside Containment Outside Containm nt Laree Break Small Brea'c Satur< ted RCS Res onse

.,')ust be perfor7>ed on plant..ot simulator.'roc'edure pialk through not sufficient.

RBNRKS: (Unsat evaluation requires explanation) .

~

J

FIGURE 3

~ ~ )

~ ~

' i CONTROL tlAilIPULATIOi'( LOG ~~ ~4 T'rlQ YGLR CYCLE Fl om To

~ ~

~ ~

i'lame: License Ho.

DATE SAT/Ui'lSAT S I Gi'!ATURE "Fl ntnSnutdnbn

~At Pov,er Bor ation or .Oi 1uti on > l00 cm Loss of Instrument Air Loss of Offsite Porver Loss of Condenser Vacuum Loss of Intake Cool in k'ater Flo<v Loss of Shutdown Cool i n RHR Loss of'omoonent Coolin Flow Loss of i<orral Feedwater Loss of Protective S stem Channel

~

Hisvositioned or Oro ed Control Rod

<<bilitv to Drive Cortrol Rods

>>'.eraency Sorati on Excessive RCS Activit Peactor lri Turbi>>e or Generator Tri Boron Concentration. Control Halfunction halfunction of Pressurizer Pressure Control f

i'!a unct on of Pres suri zer Level Contrdl 1

Hain Steam Line Break Source Ranae illS Failure rr,edizte Ranae itlIS Failure

'nt Po:.er Ranae flIS Failure

(:t be performed on p1ant or simulator. Procedure s'alk through not sufficient-RE) tARXS: '(Unsat eva1uation requires explanation)

URE. 4 OFF-.SITE LICENSED PERSONNEL - LOG Ol OH MATCI{ OUTY IN'ONTROL ROOM INSTRUCTIONS:

I. 0 llours within any 3 month period is recommended, Over 4 months without watch time - suspension until competency i demonstrated.

2. Oocument date, time start - end, evo'lutions directed/performed, sign.

Hake entry in shi ft personnel record book kept in control room to document on watch duty.

H ~

YEAR 19

~ r HAI lE JAN or. JULY FEB or AUG "MAR or SEPT APR or OCT MAY. or'HOV JUNE or OEC

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HUl1Ll'<1> I NHI 1 Vt YKUt t,UUHE REV 8 LICENSED OPERATOR REOUALIFICATION PROGRAM FIGURE 5 0 TRAINING DEPARTMENT STAFF NRC CORRESPONDENCE COVER LETTER Honth 19 Each applicable NRC Bu1let)n, Circu)ar, or Infornation Notice shall be placed in the binder following this cover 1etter. and logged in the appropriate'pa'ce below. Each staff member shall review these docUments and indicate by .initialing below >Iithin the erst 14 days of the following tiionth.

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FORMATION

<OTIC ES:

INITIALS/DATE

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~ t ~ t INITIALS/DATE January February

" i'farch April Hay June July. ~ ~

.August September October i'lo renb,er

'December

FLORDIA P{WER 5 LIGHT Cob%ANY St. Lucre Unit 2 NRC - FPL - Ebasco - CE Beetles on Loss of A'll AC Paar October 15, 1881, 9:30 e m.:

Those present at this meeting, were:

NRC CE E. Dotson V. Nerses D. Ruggiera S. Ritterbusch R. Gritz Kennedy G. Attarian g. 9ehr J. Franklin , J. RMgely 8 Gardener 8, Pagnozzi N. Iqlubin P. Caner At this meeting N. Rubin of the NRC asked that the follower)ng information be proviM to help the %C complete theB review:

Provide a list of insolvent systems required for this event.

Ust all loads on the station battery and show battery can operate for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

30 What, plant actions are requiredV How $ s plant monitoredT Rhere $ s action there sufficient 1)ghting taken's and cennunicationsV

@hat $s the impact of loss of ventklationt

5. Is diesel generator starting air a description of the air start exhausted'rovide systea.

Mhat procedures have been developed end when will they be l

avai ab) H A second meeting was scheduled for 2 pm to fur@er discuss these items.

FLOROIA POLfBt 8 LMN COI NfY t'ucie-Unit 2 sac FPL Ebasco CE Vheting on Loss of Nl AC Peer October 16. l98l, 2:00 p.m.

This aeetfng is a continuation.ot the amming session, Those present et this afternoon meeting mre:

Ebasco

8. Franklin V. Nerses D. Ruggiero R. Grit@~ N. Rubfn 6, Attarian M. Mindake? M. Kennedy This meeting fo71aed the items as requested at the a@ming session. (minutes)

A list of'nstrument systems required (attached}. ms given to the NRC.

2. A preliminary load list for the betterI and a pre3iadrery profile ms given to the NRC (attached).

3. The p1ant All be monitored through the instr'tation system located in the control room. The control room is provided v$ th DC

.li9ht<ng. Action All a3sa be provided in the switchgear roam Dere selected DC load shedding sfll take ploce. Battery pack

'lighting e$ 'll be provided 5n this arei as <el) os sound premed cemunicat$ on.

The 5RC requested that additional description be provided f'r

, control rema habitability.

The diesel generator start system ~ described: Basicelly the diesel generotor air start motors draw air for a 5-6 second period end then ere cut out. This is on the basis that Cf the diesel generator has not started by that time they probably vill not.

Sufficient eir for at least one more start is. left in the receivers.

Kn addi Non a diesel driven hardcrank cowpressor is provided-.that v$ 71 fu)ly recharge the air receivers 4n 30 minutes. A fatal ll description e$ be provided.

Precedures All be developed and suhaftted to the NRC on October 28th XQSl.

The above items nfl'l be formally submitted to the HRC (less item 63 by October 20-21, 1M'.

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Florida Power 6 Light St. Lucie Unit 2 NRC-FPL-Ebasco-CE Meeting on Matrix Power Supply Test October 15, 1981 A meeting was held at the CE offices 'in Bethesda, Maryland which was attended by the people listed in Attachment 2. The items in Attachment 1 were discussed at that time.

Attachment 2 Heeting FPL/NRC/CE October 15, 1981 St. Lucie 2 Hatrix Power Supply Test ORGANIZATION Patrick Carier DPP/FPL/Licensing Stan Ritterbusch C-E'RC/NRR/DL/LB83 V. Nerses R. Stevens ,

NRC/NRR/ I CSB J. Joyce NRC/NRR/ I CSB B. Pagnozzi EPP/FPL/EKC E. Dotson EPP/FPL/Project J. Franklin EPP/FPL/Elect.

Remo Gritz EPP/FPL/Project

ATTACHMENT 1 e 2.

Discuss Document reliability of inverter outputs capability of inverter to limit surge.

qualification of inverter to and voltage suppression supress 4000 volt peak. Show references and/or test reports.

3. Equipment adjacent channel must operate before, during and after surge (add tolerance valise on surge) (make sure perturbation is within acceptance).

4. Basis for fault testing 600 VAC and 400 VDC as maximum credible (from inveCter) .

5. Surge test model'after fault testing procedure showing surge appl,i-cation points and monitoring points.

6. Recheck dwgs. to show that they agree with writeups for fault test (and surge test).

7. Surge test max valves show via inverter limiting criteria how it cannot propagate.

Apply surge (120 YAC side) to one channel, trip the bistables, cycle relays to show that the surge has no effect on that channel.

0'onitor adjacent matrix (120 VAC source) to show that surge does not cause perturbation on the AC side such that acceptance criteria is within 120 YAC + 105.

10. Surge will be limited to 300 VAC or less by showing surge suppression via qualification test on inverter (including documentation) or use of CE System 80 TOPAZ transformers showing 3000 volt surge limitation to 1 volt output. For the inverter the actual surge and suppressed surge output values are to be specified.

Give history of 300 VAC surge valve and engineering basis.

Surge test description should include test points and similar detail as for the faul t test.

12. Revise and resubmit total Matrix test writeup for both fault and surge test.

~

13. Basis for 600 YAC a 400 VDC 600 VAC is 480 VAC ~ 10%%d plus (323-1974) + 155. = 600 VAC 400 VDC is assumed that a 300 VDC bus + lOX I 14.

~

FPL will provide basis to 480 VAC and 325 VDC by show max fault on 120 VAC inverter bus is deta'iled description of cable routings worst case single failure.

and power sources together with

Psge 1 of 11 Eraergency Procedure 2<<0030140 Rev 0 BO FLORIDA POWER & LIGHT COMPANY ST+ LUCIE UNIT 2 EMERGENCY PROCEDURE NUMBER 2-0030140 REVISION 0 October 22, 1981 BLACRDUT OPERATION APPROVAL PLT MNGR TOTAL NUMBER OF PAGES: ll

Page 2 of 11 BO FLORIDA POWER 6 LIGHT COMPANY ST LUCIE UNIT 2 EMERGENCY PROCEDURE NUMBER 2-0030140 REVISION 0 1 0 SCOPE:

This procedure provides instructions to be followed in the event of a total loss of AC power for up to four (4) hours.

2.0 SYMPTOMS

2 1 Loss of Control Room AC 2 1 DC Lighting energized Lighting 2.2 Loss of AC Power 2 ' Indications

.RPSWH10, ADS WM-881, DMW881; W-REC-881 AM-881, VM918, VM916, VM919, VM917 2 2 Alarms A-23, B-23, B-39, B-48 2.3 Loss of Load ' Indications 3

RPSWH10, W-REC-881, DEH-System, PI-22, PI-24, PI-25 3 Alarms L-21, D-8 2.4 Loss of Feedwater 2 4 Indications AM-615, AM-620 PI-12-19, PIS-09-5, FI-09-1A, FI-09-1B, FIC-9011, FIC-9021, FR-09-1A, FR-09<<IB

2. 4 Alarms C-2, C-S, C-IO, C-ll, C-13, C-19, C-26, C-27 2.5 Loss of Forced RCS Flow 2 5" Indications PDI-1101A, PDI-1101B, PDI-1101CP PDI-1101D RPSWH6, AM-101, AN-109, AM-105, AM-113 2;5 Alarms L-10, L-18

Page 3 of 13 BO FLORIDA POWER 6 LIGHT COMPANY ST. LUCIE UNIT 2 EMERGENCY PROCEDURE NUMBER 2-0030140 REVISION 0 3.0 AUTOMATIC ACTIONS:

3.1 Reactor Trip 3 1 ,RCS Low Flow (93X) 3.2 AFW Auto Actuation 3 ' 5X S/G Narrow Span 3 3 PORV's open 3o3 2400 PSIA 3.4 MSSV's open 3 4 995 PSIG

'.5 Turbine Trip 3 5 Reactor trips turbine 4 0 IMMEDIATE OPERATOR ACTIONS:

4 ' Ensure immediate operator actions for a reactor trip have been accomplished 4.1 ~ 1 Any automatic functions that were required have operated properly 4.2 Open Atmospheric Dump Valves 4.2 This closes MSSV'.s and to maintain 900 PSIG maintains Hot Leg subcooling 4 ' Ensure auto AFW restoring 4 3 With steam driven AFW pump S/G level 4.4 Open DC breakers on buses 4.4 To minimize battery drain A, B, and AB as indicated 4 4 Appendix A lists loads by white breaker numbers REEUIRED for loss of AC power

Page 4 of 11.

BO FLORIDA POMBR & LIGHT COMPANY ST. LUCIE UNIT 2 EMERGENCY PROCEDURE NUMBER 2-0030140 REVISION 0 5+0 SUBSE VENT ACTIONS:

CHECK F 1 Ensure adequate natural circulation flow 5.1.1 Loop T.is less than full power T .(<44'F) 5,1.2 T is constant or decreasing 5 1~3 Th is stable or decreasing 5 1-4 No abno~l differ'ences between,TC -RTD 8 and exit thermocouples. 'oze 5.2 If natural .circulation is not assured 5 2.1 Check RCS Temperature/Pressure to ensure subcooling.

5.2.2 Ensure Aux Feed flow to the S/G 5 '.3 Ensure Steam dump to atmosphere is in service 5 ' Take manual control of APW Pumps and maintain S/G levels stable 5.4 Monitor status of off-site power and'iesels so that AC power can be restoied as soon as available' 5.5 When primary and secondary systems'have stablized, begin RCS depressurization and cooldown by means of the atmospheric dump'alves. Control'the dump valves to maintain at least 10 F subcooling in the hot leg.

CAIJTIOH:

As cooldown progresses,'verify shutdown margin per OP 2-0110056, Surveillance Requirements for Shutdown Margin in Modes 3 and 4 NOTE Since RCS pressure decreases at a higher. rate than RCS temperature, pressure approaches saturation. Saturation occurs in the vessel .

head. Continued pressure decrease without corresponding temperature decrease will result in a saturated condition in the hot leg.

5.6 Ensure SIT's dump their contents to the core as RCS pressure decreases below SIT pressure.

Page 5 of 11 BO FLORIDA POWER & LIGHT COMPANY ST. LUCIE UNIT 2 EMERGENCY PROCEDURE NUMBER 2-0030140 REVISION 0 5.0 SUBSE UENT ACTIONS: (Cont.)

ROTE: Prior to reaching on shutdown cooling temperature and pressure, AC power should be restored either by the diesels or off-site power.

CHECK 5 7 When AC power becomes available, close DC breakers on buses A, B and AB. which were opened in Appendix A.

5.8 Re-energize AC power supplies as availability and capacity of diesels (if applicable) permit.

5.9 Ensure that automatic sequencing of necessary loads (Table 8 3 ~ 2) takes place.

5 10 Check the ESFAS status board to ensure equipment availabili ty 5.11 Locally CLOSE (HSR Hale Steam Block Valves)

MV-08-4 MV-08-6 MV-08-8

, . HV-09-10 5.12 Ensure MSR Warm-up valves closed MV-08-5 MV-08-7 MV~08-9 HV-00-10 5.13 Verify one (1) set of cavity and support cooling fans operating, or start.

5.14 Lock out automatic equipment that is not in service 5.15 Manually open all breakers on any non-vital bus or motor control center that is to be energized.

5.16 Reset lockout relays for each required bus to allow closing of feeder breakers.

5.17 Enerize 4160V buses 2A2, 2B2 as follows:

5.17.1 Strip non-vital 4.16 KV bus breakers

~

(All should be opened automatically) 5 17 2 Synchronize and:

Close 2-20109 Close 2-20309 Hold acd Close 2-20209 Close 2-20411

Page 6 of 11 BO PLORIDA POWER 6 LIGHT COMPANY STe LUCIE UNIT 2 EMERGENCY PROCEDURE NUMBER 2-0030140 REVISION 0 s

5.0 SUBSE VENT ACTIONS: (Conte)

CBSCK 5.18 Energize non-vital load centers 2A1, 2Bl as follows:

S.lg.l ~gtri load centers 2A1 ~ 2BI 5.19.2 Close 2-20110 Close 2-20310 5.19 Enerize 480V-MCC s 2A1, 2BI, 2A4, 2B4, 2C as follows:

5.19.1 ~gtri MCC's, 2A1, 2B1, 2A4, 2B4, 20 5.19.2 Close 2-40115 Close 2-40410 Close 2-(later)

Close 2-(later)

Close 2%0119 or 40409 5 ~ 20 At MCC-2C Close breakers for:

Turbine gear 42510 Bearing oil pump - 42506 Air side seal oil pump - '42507 Hydrogen side seal oil pump 42504 5 '1 Place TCW pp in operation 5.22 Ensure TCW system to the instrument air compressor in noxmal alignment.

5.23 Place turbine drain valve control to'he ~o en position 5.24 Before turbine bearing oil pressure drops to 12 PSIG; Start bearing oQ. pump.. IP oil pressure drops to 10 PSIG; Start emergency D.C. oil pump. Do not run both pumps simultaneously.

5.25 Remove the following components from'service.

I a ) Steam get air egectors b.) Priming effector, ..

c ) Aux Priming effector d ) Aux Steam to RAB e ) Gland Seal

Pager,7 of II BO PLORIDA POWER 6 LIGHT COMPANY ST LUCIE UNIT 2 EMERGENCY PROCEDURE NUMBER 2-0030140 REVISION 0 5.0 SUBSE VENT ACTIONS: (Cont )

5 '5 (Cont ~ )

CJ!DTIOH:

Consider equipment starting requirements. Alternate operation of equipment may be required to avoid overloading the diesel generators (3685 KW max cont rating)

CHECK 5.26 Start CEDM cooling fans A or B 5 '7 Start reactor support cooling fans A and B 5.28 Close breakers for pressurizer heater buses 2-20204 2-20403 5.29 Ensure Bearing oil lift pump is running 5.30 Start turbine lube'oil vapor extractor and generator oil vapor extractor NOTE If Turbine'as been at reset longer than 15 minutes, rotate I/4 turns until bow is gone.

5 '1 At approx 0 RPM turbine speed, 'verify turning gear operation or initiate manually 5.32 Reduce turbine oil temperature to 95-100 P 5.33 Isolate TCW to the hydrogen coolers 5 34 If additional CST water is available and sufficient power is av'aQ.able, place the water treatment plant, in service 5.35 Place the spent fuel cooling system in operation 'as necessary NOTE: ~

With 3 I/3 fuel cores stored, before reaching the boiling point it will take 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> without cooling

Page 8 BO of ll FLORIDA POWER 6 LIGHT COMPANY STe LUCIE UNIT 2 EMERGENCY PROCEDURE NUMBER 2-0030140 REVISION 0 5.0 SUBSE VENT ACTIONS: (Cont.)

5.36 Sample and analyze the RCS to determine-if. fuel element failure has occuxred.

s CAUTIOH:

Insure that one BAlK remains in service as a source of borated water in Mode 5.

5-37 Determine expected duration of power outage. If unable to do so or the outage is to be extensive, boxate the RCS to cold shutdown concentration.

t 5 38 If the outage will exceed 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and the RWT is available, proceed to cold shutdown conditions utilizing thermal circulation atmospheric steam dump, and feedwater addition. Place SD cooling in service when conditions permit. Proceed to Step 5 ~ 40 If the outage willexceed 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and the RWT is not

'.39 available, the SIT s should be used for makeup to the RCS. Make the following preparations 5 39.1 Verify operation of the instrument air systems.

5 39 2 OPEN 480V AC breakers for:

MV-2504 MV-2501 5 39.3 OPEN and lock SIT test"line return to RWT V-07009 SIT test line return to RWT V-3463 SIT test line tie to VCT V-03920 5 39 4 Select a SIT to use as a makeup source to the VCT Operate the appropriate and drain valve, 2A1, -AOV-3621, 2A2, fill

-'OV-3611, 2Bl, -AOV-3631, 2B2, -AOV-3641 CAUTION:

USE ONE SIT AT A TXME. Insure RCS is >1750 PPM

Page 9 of 11 BO FLORIDA POWER 6 LIGHT COMPANY ST LUCIE UNIT 2 EMERGENCY PROCEDURE NUMBER 2-0030140 REVISION 0 5.0 SUBSE UENT ACTIONS: (Cont.)

CHECK 5.40 If przr cooldown cannot be accomplished satisfactorily by auxiliary spray," proceed with the alternate positive means of depressurization as follows:

5 '0.1 Place power operated relief valve V1402

.and V1404 switches in override 5 40 2 Initiate a high przr pr'essure trip signal on two RPS channel trip units 5 '0.3 Place either power operated relief valve (V1402 or V1404) switch in normal range position.

NOTE This will vent the pressurizer to the quench tank.

Eo close valve, place svitch dc override 5 40.4 Control rate of cooldown/depressurization by selective operation of power operate relief valves in this mode, untQ. cooldown vie the aux spray valves can be initiated 5.41 When normal AC power is avaQ.able 5 41 1 Restore bus sections to normal supply 5.41 ' Place diesel generator system in standby per 2-2200020 5.41 ' Restore'all systems to normal r

Page 10 of ll BO FLORIDA POWER 6 LIGHT COMPANY ST LUCIE UNIT 2 EMERGENCY PROCEDURE NUMBER 2-0030140 REVISION 0 6 0 PURPOSE AND DISCUSSION:

This procedure provides the action to be taken in the event of a complete loss of AC electrical power concurrent with a turbine trip. OP (2-0030130) will be used concurrently with this pxocedure.

Discussion A loss of power to the 4160 V buses, results in a loss of power to all 480 V load centers and motor control centers .and to all instrumentation not fed directly or indirectly from the A reactor trip will occur from a low reactor coolant flow station'attery.

signal due to the loss of powex'o .the 6900 V buses supplying the reactor coolant pumps and will be accompanied by a turbine trip, and generator lockout.

Steam dump to atmosphere must be used to remove reactor decay heat. Initially, steam generator safety valves may actuate to augment the s'team flow and to help control steam genexator pressure immediately after the trip.

A rapid reduction in steam generator water levels will occur due to the reduction .of the steam generator void fraction on the secondary side and also .because steam flov will continue after normal feedvater flov stops. Auxiliary feedwater flow vill automatically initiate.

Core"decay'eat removal is accomplished by natural circulation in the reactor coolant loops.

Core damage is not expected as a result of loss of power condition as the steam generators are maintained as a heat sink and no loss of watex occurs fram the pressurizer.

If operating under blackout conditions and an engineered safety features actuation signal occurs, any non-emergency loads that are running will be automatically tripped and the required emergency loads will be automatically started.

7aO

REFERENCES:

7 ' FSAR, Section 15 7 2 FSAR, Section 8 7 ' 'Opeiating Procedure 00030130, Shutdown Resulting From Reactor/Turbine Trip 7.4 Operating 'Procedure 00210020, Charging and Letdown (Unit tl)

Page BO ll of ll FLORIDA POWER & LIGHT COMPANY ST. LUCIE UNIT 2 EMERGENCY PROCEDURE NUMBER 2-0030140 REVISION 0 7 0 REFERENCES'Cont )

7,5 Operating Procedure 00330020, Turbine Cooling Water Operation (Unit Pl) 7.6 Operating Procedure 40250031, Boron Concentration Control, Off-Normal (Unit fl) 7.7 Operating Procedure f1010040, Loss of Instrument Air (Unit fl) 7.8 Operating Procedure 8'1540020, Water Plant Startup and Shutdown (Unit tl) 7 ' Operating Procedure 02200020, Emergency D'iesels Standby Lineup (Unit Pl) 7 ~ 10 Operating Procedure 5'0700022, Aux Feedwater System Opereation (Unit fl) 7.11 Operating Procedure 80030142, RCS Cooldown During Blackout (Unit dl) 8 ' RECORDS/NOTIFICATION:

Normal Log Entries.

Notify Duty Call Supervisor.

9 0 APPROVAL:'eviewed by Plant Nuclear Safety Committee. 19 Approved by Plant Manager 19 Rey Reviewed by RRG 19 Approv'ed by .Plant Manager 19 "LhST PhGE" Emergency Procedure 2-0030140 Rev 0 Total Number of Pages ll

/~ of Page kS 13 BO FLORIDA POWER & LIGHT COMPANY ST. LUCIE UNIT 2 EMERGENCY PROCEDURE NUMBER 2-0030140 REVISION 0 The following DC Instrumentation and Equipment is required for operation during loss of all AC power. All- other DC loads may be disconnected during the period of blackout. When either the diesel(s) or offsite power becomes available, the remaining DC loads should be returned to service.

REQUIRED FOR LOSS OF AC POWER DC Bus 2A Circuit Control Room Lighting 60111

'Safety Infection Tanks 60119 Vital AC Instrument Bus 2A MA" 60120 V-1475 (PORV) 60124 Vital AC Instrument Bus 2MC 60129 DC Vital Bus 2MA (Safety) 60133 DC Vital Bus 2MC (Safety) 60134 MV-08-18A (AFW) 60138 MV-08-18B (AFW) . 60141 DC Bus 2B Control Room Lighting 60202 Vital AC Instrument Bus 2MB 60220 Vital AC Instrument Bux 2MD 60222 Safety Infection Tanks'C 60231 Vital Bus 2MB'(Safety) 60233 DC Vital Bus 2MD (Safety) 60234 V-1474 (PORV) 60236 MV-08-19A (AFW) 60239 Mv-08-19B (AFW) 60241 DC Bus 2AB AFW Pp 2C,= Valve Indication and AFW Actuation Signal 60320 MV-09-11 AFW Isol Valve 60327 MV-08-12 Stm Supply to Turbine 60328 MV-09-12 AFW Isol Valves 60329 MV-08-13 Stm Supply to Turbine 60330

Page of 13 BO PLORIDA POWER & LIGHT COMPANY ST LUCIE UNIT 2 EMERGENCY PROCEDURE NUMBER 2-0030140 REVISION 0 APPENDIX A

,The follavlng instrumentation is located in either DC Vital Bus Cabinets 2MA, 2MB, 2MC, or 2MD or AC. Instrument AC Bus 2MA, 2MB, 2MC, or 2MD:

DC Vital Buses 2MC ESC-SA (ESFAS) CKT 6&8 6&8 ESC-SB (ESPAS) CKT 6&8 6&8 RPS CKT Instrument Inverter Buses Przr Press PIA-1102A, PIA-1102B, PIA-1 102C, P IA-1 102D CKT 4 Press Level later later later later S/G Press PIA-8013A

~

PIA>>8013B, PIA-8013C,

~ PIA-8013D CKT 'later .

later later later S/G Press PI 8023A, PI 8023B, PI 8023C PI 8023D CKT'/G

'ate'r later later later Level LIC 9013A, LIC 9013B, LIC 9013C, LIC 9013D CKT S/G Level'LIC 9023A, LIC 9023B, LIC 9023C, LIC 9023D CKT 3 CEDM's CKT Containment Radiation Monitors later later later later Containment Pressure PIS 072A, PIS 072B,,

PIS 072C PIS 072D later -latex later later

OFF-N0191AL PROCEDURE 2-0030130 REV.

0'x/n FLORIDA 'POWER AND LIGHT COMPANY ST LUCIE PLANT UNIT 2 OFF-NORMAL OPERATING PROCEDURE NUiiBER 2-0030130 REVISION 0 September 30, 1981 RECTOR TRIP/TURBINE TRIP REVo FRG.

APPROVAL PLT. MGR TOTAL NO OF PAGES 14

L ksva'La STs LUCIE PLANT UNIT 2 OPP-NORMAL PROCEDURE NUMBER 2-0030130 REVISION 0 1 0 300PZ This procedure provides guidelines for an operator following a reactor trip. The circumstances which contributed to the trip may require the use of another procedure concurrent with this procedure 2 0 SYiilPTOMS 2.1 Indications and annunciator alarms associated with any of the following trips:

SYMPTOM INDICATION/ALARM 2.1.1 Hi power 2 1~1 Indication J1002A/004A, B, C, D, JR009/010 RPS Ch. 1 TR1111/1121 2.1 1 Alarms L-9, L-17 2.1.2 Hi rate of change/power 2.1 ~ 2 Indication JK1-001A, B, C, D, RPS << Ch. 2 2.1.2 Alarms L-2, L-33 2~1 3 Lo .RC -Plow- 2s 1 3 Xndications PD1 1101A$ BP CP DP RPS << Ch 6 2 ~ 1.3 Alarms L-10i L-18 2 1 4 Lo S/0'level 2.1.4 Indications LIC-9013 A, B, C, D, LIC-9023 A, B, C, D, LR-9011/9021 RPS Chm 7 2.1.4 Alarms L-3, L-11

4 oj'j-num REVISION 0 2 0 82MPTOM8 (Cont )

2. 1 (Cont.)

2.1.5 Lo S/G Press 2.1 ~ 5 Indications PI-08-1A, 1B, PR-08-1, 2, RPS - Ch ~ 8 PI-8013A, B, C, D,, (RTGB 206)

PI-8023A, B, C, D, {RTGB 206) 2.1.5 Alarms L-19, L-27 2.1 6 Local Power Density 2.1.6 Indications JR-012, JO-012-'1 JI-005A, B, C, D.

JI-006A, B, C, D.

RPS - Ch 3 2.1 ~ 6 Alarms L-22, L-30 2.l. 7 Thermal margin/low pressure 2.1.7 Indications PIA-1102A, B, C, D, RPS Ch 4 2 '.7 Alarms L>>36, L-41 2.1.8 Hi Przr Press ,2 1 ' Indications PI-11026., B, C, D, FR-1100 PIC-llOOX/Y RPS Ch'. 5 2.1.8 Alarms L-20, L-'28 2.1 ~ 9 Hi Containment Press 2.1 9 Indications PIS<<07-2A, 2B, 2C, 2D ~

PI-07-4A/5A

'PR-07-4B/5B 2.1.9 Alarms L-5, 1-13

i OFF-NORMAL OPERATING PROCEDURE NUMBER 2-0030130 REVISION 0 2.0 8DMFLOM8 (Cont.)

2. 1 (Cont. )

2. 1. 10 Turbine trip l.

2. 10 Indications WM-881 DM',7-88 1 W-REC-881 RPS Ch 10 DEH SYSTEM 2 ~ '1.10 Alarms L-21, D-8 2 1 11 I.oss'f CCW 2.1.11 Indications FIA-1158, 1168, 1178, 1188 RPS - Ch. 11 FIS-14-1A, 1B 2.1.11 Alarms L-6, L-14 2 1 12 Turbine Overspeed 2.1 ~ 12 Indications RPS - Ch 10 DEH SYSTEM MW-881 2.1.12 Alarms D-5, L-21 2 1 13 Condenser low vacuum 2.1.13 Indications RPS Ch 10 PI-10-7A, 7B 2.1.13 Alarms D-3, L-21, D-13 2.1.14 Thrust bearing failure , 2.1.14 Indications

~

RPS Ch 10 TR-22-1 2.1.14 Alarms D-6, D>>16

OFL NOR.

REVISION 0 2.0 SYMPTOMS (Conn.)

2. 1 (Cont.)

2.1.15 Generator Lockout 2.1 ~ 15 Indications RPS Ch. 10 AN '881 AN 872 VH 872 2.1.15 Alarms D-7 2.1.16 Exhaust Hood Hi Temp 2 ' ~ 16 Indications RPS Ch 10 TR>>22-6 2.1s16 Alarms D-4, D-14 2 1.17 Turbine Bearing Oil Press 2ils17 Indications RPS<<Ch 10 PI-22-25 2, 1. 17 Alarms D-2, D-12

.2 1 18 Auto-Stop Oil Pressure 2 1.18 Indications RPS Ch 10 PI-22-25, 26 2 1 ~ 18 Alarms D-7, D-17 2.1.19 Hi S/G Level 2.1 ~ 19 Indications LIC-9013 A, B, C, D LIC-9023 A, B, C, D LR-9011/9021 RPS Ch. 12 (FUTURE) 2;1.19 Alarms L-'77 L-7 7 (FUTURE)

G-17 G-9

OFF-NORMAL OPERATING PROCEDURE NUMBER 2-0030130 REVISION 0 2 ' SYMPTOMS (Cont.)

2~1 (Cont,)

2 ' '0 Manual Trip 2.1 '0 Indications All CEA's fully inserted ADS, DEH- SYSTEM W-REC<<871 DX<-871 Core Mimic Digital Pos. Readout 2 1 ~ 20 Alarms L-1, D-10 EMOTE: In every case where a reactor trip occurs, the following alarms should aiso energize, K-l, -2,

-3, -4, -5, -9, -10, -12,

-13, (RTB's) 2.2 Reactor Trip Breakers Open 2.2 Indications RPS 2 2 Alarms K-l, -2, -3, -4, -5, -9, -10,

-12, -13 2.3 CEA's Insert 2.3 Indications ADS, Core Hormic, Digi"al Pos.

Readout, and Backup 2.3 Alarms 2 4 Generator MW Output Reduces to Zero 2. 4 Indications DEH, W-REC-871 DMW-871, RPS Ch. 10 2.4 Alarms D-21, D-8 2.5 After trip, the following Parameters:

Reactor Power - Decrease Pressurizer Pressure Decrease

~ .

Pressurizer Level .Decrease RCS Temperature Decrease Steam Generator Pressure Increase Steam Generatox L'evel- =- Decrease

OFF-NO REVISION 0 3 0 AUTOMATIC ACTIONS 3.1 The turbine generator will trip wf.th EITHER A Turbine Trip Signal OR A Reactor Trip Signal 3 2 'he reactor will trip with EITHER A Reactor Tx'ip Signal

-'OR' Turbine Trip Signal Power is > 15X 3~3 Atmospheric Dumps and/or SBCS actuate following a trip or runback.

3.4 Plant electrical auxiliaries *

, transfer from auxiliary to SU transformer.

3 5 urbine trip on Hi S/G lev 1 will also close 100X feedwater bypass valves.

3 6 S/G level < 34X initiates

'uxiliary FW auto start.

If the conditions for a reactor trip are present and the reactor has not tripped, proceed IMMEDIATELYto EP 2-0030132, Anticipated Transient Without a Scram (ATLAS) .

OFF-NORMAL OPERATING PROCEDURE NU BER REVISION 0 4.0 IMMEDIATE OPERATOR ACTIONS 4.1 Verify all applicable automatic actions ~

4.2 Manually depress reactor PB on RTGB 201 and 204.

'rip 4.3 Manually depress turbine trip PB'n RTGB 201.

4 4 Verify reactor trip bkrs are open.

4.5 Verify all full length CEA s If more than one CEA are fully inserted; is NOT fully inserted, initiate emergency 4.6 Verify all valves are closed'.5 turbine control boration.

4.7 Verify gen bkrs're open (240W40349) (240W03520) 4.8 Close reheater block valves MV-08%, 6,,8, 10 4 9 Verify that aux pwr has transferred to the S/U transformer.

OR DG 2A and/or 2B running In the event of a COMPLETE loss of offsite power, refer immediately to Blackout Operation, 2-0030140 breakers 20211 and/or 20401 closed 4.10 Verify main feedwater feed 5X through 15X values OR Initiate Auxiliary FW r

OFF-NOR'%PERATING PROCEDURE NUMBER 2-0030130 REVISION 0 4 0 IMMEDIATE OPERATOR ACTIONS 4.11 Verify that T avg is being maintained at 532 F by SBCS or atmospheric dumps.

4 12 -

Consult Break Diagnostic Chart (Fig. 1) and determine if anothei emergency procedure is required. If so, refer EaKDIA.TEM to, that procedure. If not, refer.

to subsequent acti'ons for this procedure.

BREAK D}OGNOSTIC CHART IF PRESSURIZER LEYEL'HANGING AND PRESSURIZER. PRESSURE RAPIDLY DECREASING AND

&TEAM GENERATOR PRESSURE

~ABNORMALLY LOW IN ONE OR NORMAL OR RISING%

0TH STEAM GENERATORS.

THEN STEAM LINE BREAK LOCA AND IF-- AND.IF CONTAINMENT PRESSURE CONTAINMENT PRESSURE UP .NORMA> NORMAL 7

THEN THEN THEN T EN MS/FW BREAK LOCA,.

MS/FN BREAK OUTSIDE OF S/6 - TUBE IN CTMT RUPTURE CTM T

%MAY D.ECR EASE SLIGHTLY."AFTER 'REA CTQR TRIP

Page 11 of 14 OPP-NORMAL OPERATING PROCEDURE NOSER 2-0030130 REVISION 0 5 0 SUBSE VENT ACTIONS CAUTION If trip was caused by High Pressurizer Pres'sure, ensure the PORV's are closed or isolate manually when pressure <2300 psig CHECK 5.1 Place the feedwater bypass valves in the "auto" position to maintain no load levels.

CAUTION 3o not overfeed the S/G s; This could cause T avg to go below 532oP and tnermal shock the S/G s.

5. 2 If the emergency .diesels are running, s tar t both motor driven aux fd pumps and maintain.S/G level 5.3 If both the S/G feed pumps and motor driven auxiliary feed pumps are inoperable, then start steam driven auxiliary feed pp CAUTION Peed should be diverted from any S/G showing abnormally low steam pressure or high feed flow.

5.4 Verify przr level is being returned to setpoint by auto control of charging pumps and letdown control valves 5 ' Verify przr pressure is being controlled automatically by the heaters and spray valves

age o OFF-NORMAL OPERATING PROCEDURE NUMBER 2-0030130 REVISION 0 5 0 SUBSE UENT ACTIONS (Cont )

CHECK 5.6 If steam pressure cannot be maintained above 800 psig, close HSIV's or atmospheric dump valves, as appropriate, to, avoid excessive cooldown and depressurization of the RCS.

5.7 Verify shutdown margin.. Xf it is below tech spec, emergency borate 5.8 As the turbine slows down, verify the bearing oil pumps start as follows:

5.8.1 'C bearing oil pump and seal-oil backup pumps start together 9 11-12 psig 5.8 ~ 2 Emergency D.C. oil pump 8 1G-11 psig 5.8.3 Bearing oil lift pump starts when turbine reaches 600 RPM NOTE If any of the above pumps fail to start, manually start.

5 ' Verify.turning gear motor is operable prior to reaching 0 RPM 5.10 Verify 8 0 RPM the turning gear engages to engage it manually 5,11 Place turbine drain selector to OPEN

Page 13 of 14 OFF-NORMAL OPERATING PROCEDURE NUMBER 2-0030130 REVISION 0 5~0 SUBSE UENT ACTIONS (Cont )

CAUTION Do not isolate steam to gland seals as long as condenser vacuum is maintained.

CHECK 5.12 Reduce bearing oil temperature to 90 F 100 F In the event the condenser is not available, evaluate condensate storage tank inventory 5 ~ 13 it If is unsafe to maintain the plant in hot standby condition; bring the unit to cold "shutdown per OP 82-0030127 5.14 If immediate recovery and return to power is planned, commence CEA b'lock circuit check, if.required (see OP 82-0030122 Para. 8.6)

CAUTION If RCP's have tripped, they may be restarted if not it has been confirmed that a lock has services occured, pump CAN be restored and RCS pressure-temperature conditions permit restart.

OFF-NORMAL OPERA'ZING PROCEDURE NUMBER 2-0030130 0 'EVISION

6. 0 PURPOSE/DISCUSSION This procedure provides the entry to a sequence of events that will lead to the safe termination of any of the emergency events considered for our plant. The first few immediate actions verify that all has performed as it should. The operator verifies CEA's in, or if not, is referred to the ATLAS procedure. He then verifies off-site po~er and if it is not available, he is referred to the Blackout Procedure. He then scans a Diagnostic Chart and, if necessary, is directed to the appropriate emergency procedur'es. Each of these emergency procedures will provide specific instructions for the particular circumstances and verify adequate core cooling, or if not, refer to the .operator to the inadequate core .cooling procedure.

If no emergency conditions exist, the operator continues to the subsequent actions, which guide him to a safe shutdown and preparation for return to power, if conditions. warrant.

7 0 REFERENCES 7.1 St. Lucie Unit gl Off-Normal Procedures 7 ' St. Lucie Unit 82 FSAR, Sect. 7

~

7 ' CEN 128, C.E. NSSS transients 6 accidents 8.1 Normal log entries and trip details 8.2 Startup/Shutdown log entry 8 ' Reactor trip log entry 8.4 Applicable log chart recorders 9 0 APPROVAL Reviewed -by the Facility Review Group 19 Approved by Plant Manager 19 Rev. reviewed by Facility Review Group 19 Approved by Plant Manager 19 "LAST PAGE"

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