ML17207A745

Forwards Response to NUREG-0578,Lessons Learned Task Force short-term Requirements,Per NRC .Engineering Designs Liable to Change in Response to Addl NRC Clarifications.Drawings Available in Central Files Only

ML17207A745
Person / Time
Site:	Saint Lucie
Issue date:	01/11/1980
From:	Robert E. Uhrig FLORIDA POWER & LIGHT CO.
To:	Eisenhut D Office of Nuclear Reactor Regulation
Shared Package
ML17207A746	List: ML17207A745
References
RTR-NUREG-0578, RTR-NUREG-578 L-80-17, NUDOCS 8001180249
Download: ML17207A745 (86)

Text

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REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIOS)

ACCESSION NBR;8001180249 OOC DATE'0/01/11 NOTARIZED NO P"ACYL".50 335 "St. Lucia Planti Unit ip F'lorida Power 4 Light Co.

AOTH BYNAME AU'fHOH AFFILIATION UHRiceR ~ ED Florida Power L'ight Co.

RECAP ~ NAME RECIPIENT AFFILIATION EISENAUT<D.G, f)3 vi sion of Oper ating Reactors DOCKET ¹ 0500tr33b

SUBJECT:

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FLORIDA POWER 4 LIGHTCOMPANY January ll, 1980 L-80-17 Office of Nucl ear Reactor Regul ati on Attention:

fir. Darrell G. Eisenhut, Acting Director Division of Operating Reactors U. S. Nuclear Regulatory Commission Hashington, D.

C.

20555

Dear Nr. Eisenhut:

Re:

St. Lucie Unit 1

Docket No. 50-335 NNEG-Q5 B St T'~Ri t

Pursuant to Nr. Denton's letter of October 30,

1979, as verbally amplified by the NRC Staff on December 3, 1979, Florida Pow r 5 Light Company is submitting herein addi tiorial details concerning the actions being taken to meet our coIr~nitIIIents to implement the Category A items contained in NUREG-0578.

The drawings associated with this submittal are being copied and will be forwarded to you within a few days.

FPL is coIIImi ted to meeting the intent of NUPEG-0578.

However, the actions and engineering designs described herein represent our best current thinking about how to implement'he NUREG recommendations.

The designs are subject to change in response to additional clarifications from the NRC Staff, changes in the state of'he art, and changes in commercia'lly.available hardware.

Very, truly yours,

/ g Robert E. Uhrig Vice President Advanced

Systems, 8 Technology REU/MAS/ah Attachm'ent cc:

J.

P. O'Reilly, Region II Harold F. Reis, Esquire 800>180 2A P PEOPLE... SERVING PEOPLE

f f

ST.

LUCIE UNIT 1 IMPLEMENTATION ACTIONS TO MEET NUREG-0578

~

CATEGORY A ITEMS

2. 1.1 EMERGENCY POWER SOURCES Our reviews have shown that St. Lucie Plant as originally designed meets the Category A requirements of NUREG-0578 for this item.

2.1.2 RELIEF AND SAFETY VALVE TESTING FPL will be participating in the EPRI program submitted to you by letter dated December 17,

1979, from Mr. William J. Cahill, Jr.,

Chairman of the EPRI Safety and Analysis Task Force, under the title "Program Plan for the Performance Verification of PWR Safety/Relief Valves and Systems",

December 13, 1979.

Reference FPL's letter L-79-362 from R.

E.

Uhr ig to H. Denton dated 12-21-79.

2.1.3.a DIRECT INDICATION OF VALVE POSITION The implementation schedule for this modification is being held in abeyance pending our review of the NRC's Show Cause Order for St. Lucie Unit 1 dated January 2, 1980.

However, equipment required for installation is on-site and installation of cable and components outside containment is in progress.

The modification itself is described below:

Desi n Bases Valve positions will be monitored acoustically and indicated in the Control Room.

Valve position monitors are powered from a vital instrument bus and are seismically qualified.

The sensors are designed to be qualified for the postulated post accident environment.'alve position indication is safety grade.

As backup to this single channel environmentally qualified system, limit switch position indication for the PORV pilot valves and discharge pipe temperature indication for the pressurizer safety relief valves are used.

The PORY's also are provided with an RTD in the piping downstream of the valves.

Function The acoustical valve flow monitoring system provides the operator with positive flow indication for each pressurizer safety valve and power operated relief valve (PORV).

System Descri tion The means of detecting pressurizer safety relief and power operated relief valve position is by continuously and automatically detecti ng acoustical signals (noise levels) generated by flow through the valve.

This is accomplished by utilizing accelerometers mounted on each valve's discharge piping.

The accelerometer converts flow-generated acoustical noise into an electrical charge which is then converted to a voltage by the charge converter.

This proportional voltage is then processed and a relative flow indication is obtained.

Five valve position monitors are provided, three for the pressurizer safety relief valves and two for the PORV.

A common.audio-visual alarm alerts the operator when flow through any of the five valves exceeds a pre-established set point.

These set points are front panel adjustable.

The system is powered from a 120Y AC 60 Hz Class 1E power supply.

An alarm is initiated upon loss of instrument power.

The indicator modules are located on

the Post Accident Panel in the Control Rocm.

The system is qualified in accordance with applicable IEEE codes.

The accelerometers and charge converters are located inside the containment and are subjected to the contai nment environment during and following a LOCA.

These ccmponents are designed and tested to withstand and remain operable following the postulated accident.

The indicator modules are located in the Control Room where the enviroanental conditions during and following an accident are the same as that prior to the postulated accident.

The components of the Accoustical Valve Flow Monitoring System have been designed as Seismic Category I.

The system components remain operable following seismic loads.

S stem Com onents The following equipment is furnished as part of the Accoustical Valve Flow Monitoring System:

5 Accoustical Sensors

-'ested and..qualified to IEEE* for the containment environment 5 Charge Converters tested and qualified to IEEE* for the containment environment

• The vendor is currently performing qualification test.

Cable furnished as Class 1E.

50 feet of low

noise, high temperature cable connects each valve sensor to its charge converter 5 Indicator Modules tested and qualified to IEEE* for the control rocm envirorment 1 Alarm Module tested and qualified to IEEE* for the control room environment Testin and Ins ection Vendors substantiate through test, calculations and/or operational data that the system components remain operable following seismic loads.

Vendors also qualify the equipment to the service environment-

• The vendor is currently performing qualification test.

2.1.3.b INSTRUMENTATION FOR INDE UATE CORE COOLING Installation of a Subcooled Margin Monitor is currently in progress.

Recent problems regarding the qualification of Class 1E isolators has caused a delay in the shipment of the isolators, but all other material is on-site.

We currently schedule complete installation of the monitor on or before February 15, 1980.

Both the need for and the functional requirements of a reactor vessel level instrument are under development by CE and the CE Owners Group.

When finalized, the information will be provided to the NRC for review.

Subcoolin Mar in Monitor Desi n Bases Continuous display of reactor coolant margin to saturation will be provided in the Control Room.

The inputs to the Subcooling Margin Monitor are from safety grade sensors qualified for the post accident environment.

Safety grade temperature signals from each RCS hot and cold leg will be utilized.

The monitor will be a highly reliable, non-redundant, envi roanental ly qualified system using the steam tables as a backup.

Subcooling margin monitor will be isolated from existing reactor protection or engineered safety features.

Function The subcooling margin monitor provides the Control Room operator with continuous indication of margin to saturated conditions in the reactor coolant system.

S stem Descri tion The Subcooled Margin Monitor is an on-line microccmputer based system which uses reactor coolant process signals to provide a continuous indication of the margin to saturation.

The system consists of a calculator module, a panel display module 'and cabling between the modules.

The calculator module receives temperature and pressure analog signals frcm the following primary loop parameters:

Pressurizer Pressure (2 Signals) b)

RCS Cold Leg Temperature Loop 1

Cold Leg Temperature Hot Leg Temperature

c)

RCS Cold Leg Temperature Loop 2 Cold Leg Temperature Hot Leg Temperature These signals are interfaced to a microcomputer.

The microcomputer contains steam tables and interpolation routines to calculate saturation temperatures and pressures.

By compari ng the-saturation temperature or pressure to the actual coolant temperature or pressure, a margin to saturation is calculated.

Either the temperature or pressure margin can be displayed on the digital panel meter.

The margin is ccmpared to a setpoint for a low margin alarm.

Input sensor ranges and setpoints are listed in Table 2. 1.3b-l attached.

See Figure 2. 1.3b-l for the instrument block diagram.

Software diagnostics are periodically executed to detect observable software malfunctions-A diagnostic failure is indicated by a flashing meter display.

The Subcooled Margin Monitor system is provided with a 120V AC 60 Hz safety grade power supply.

An alarm actuates in case of a power failure.

The system is qualified in accordance with applicable codes and will be upgraded to meet requirements of Regulatory Guide 1.97 prior.to January 1,

1981.

The display module is located on RTGB 103 in the Control Room.

The only portions of the system which will be subjected to the contairment environment associated with a LOCA are the input sensors.

These components are part of the Reactor Protection instrumentation and are designed to withstand the service environment during and following a LOCA as specified in Section 3. 11 of the FSAR.

The remaining portions of the system are located in

the Control Room where the envirormental conditions during-and following an accident are the same as that prior to the postulated accident.

S stem Com onents The following equipment is furnished as part of the Subcooled Margin Monitor System:

1 Calculator Module tested and qualified to IEEE* for the control room environment 1 Panel Display Module -

tested and.qualified to IEEE* for the control room environment Cable furnished as Class 1E, 15 feet inter-connecting cabl e 7 E/E Signal Isolators -

tested and qualified to IEEE* for the control room environment.

These i so 1 ato rs are requi red to ma inta i n the separation requirements between the different Reactor Protection channels and safeguard channels

• IEEE-323-1974 I E E E-344-19 7 5 10

Testin and Ins ection Vendors will substantiate through test, calculations and/or operational data that the system components will remain operable under seismic loads.

Vendors will also qualify the equipment to their service enviroment.

SUBCOOLED a~DKGIN a~fONITOR INPUT RA AGES AND'ALARM SETPOINT TABLE 2.1.3b-l Process In ut Process Signal Ran~e F

sia) 0 Electrical Signal Ran"e (V

mA PY-1103 0-1600 psia TY-1112CA 465-615 F

TY-1112CB 465-615 F

TY-1112HA 515-665 F

TY-1122EiA 515-665 F

~-1122CB 465-615 F

TY-1122CA 465-615 PY-1102A-1 1500-2500 psia 1-5 Vdc 1-5 Vdc 1-5 Vdc 1-5 Vdc 1-5 Vdc 1-5 Vdc 1-5 Vdc 1-5 Vdc Low Hargin Alarn Set F

30o Subcooling Reset F

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2.1.4 DIVERSE CONTAINMENT ISOLATION The original plant design provides for containment isolation on the receipt of either a high containment pressure signal or a high containment radiation signal.

Additionally, the design has been changed to add the Safety Injection Actuation Signal (SIAS) as an input to the containment isolation circuitry (CIAS).

Receipt of SIAS will now automatically initiate containment isolation.

Neither SIAS nor CIAS can be reset while an initiating signal is

present, and resetting of either SIAS or CIAS will not result in the reopening

'f any valves-This change resulted from a review of all lines penetrating contaiwent, and the following lines, not previously isolated on SIAS were found to be non-essential frcm a safety standpoint:

1.

Steam Generator Blowdown (2 lines) 2.

Primary Makeup Water 3.

Instrument Air Supply 4.

Containment Pure Air Exhaust 5.

Containment Purge Supply 6.

Nitrogen Supply to Safety Injection Tanks 7.

Reactor Coolant Sample 8.

Pressurizer Surge Line Sample 9.

Pressurizer Steam Space Sample 10.

Steam Generator Blowdown Sample (2 lines) 11.

Containment Vent Header

12. Reactor Orain Tank Pump Suction 12

13. Reactor Coolant Pump Controlled Bleedoff 14.

Containment Atmosphere Radiation Monitoring System (3 lines)

All other lines penetraing containment are either essential to core cooling capability, have locked closed manual valves, or isolate already on a safety injection actuation signal.

We conclude that St. Lucie Unit 1 meets requirement 2.1.4.

13

2.1.5. a DEDICATED PENETRATIONS FOR EXTERNAL HYDROGEN RECONBINERS St. Lucie Unit 1 is equipped with 2 hydrogen reccmbiners located within the containment.

This requirement is therefore not applicable.

2.1.5.b INERTING BMR CONTAINMENTS NOT APPLICABLE TO PWR's 15

2.1.5.c HYDROGEN RECOMBINER OPERATING PROCEDURES A detailed review of the existing bases and procedures for reccmbiner use at St. Lucie Unit 1 has been conducted in response to NUREG-0578 and other NRC documents.

Our review has concluded that no changes are necessary at this

time, and that St. Lucie Unit 1 meets the requirements of this item.

16

. 2.1.6.a SYSTEM INTEGRITY FOR HIGH RADIOACTIVITY In order to implement the recommendations of NUREG-0578, St. Lucie Plant has recently developed a program to identify and minimize leakage in systems outside containment that could contain highly radioactive fluids during a

serious transient or accident.

The procedure "RAG Fluid Systems Periodic Leak Test", OP-1300054, which has been written and approved, establishes and implements a leak test program to provide baseline data pertaini ng,to operational leakage characteristics and information from which leak reduction measures can be instituted.

The following systems are included within the scope of this program:

(1)

ECCS (HPSI, LPSI, and Containment Spray systems)

(2)

RCS Sample System (3)

Charging and Letdown System (CVCS)

(4)

Waste Gas System In order to perform the leak test, these systems are pr essurized for a minimum period of one hour duri ng which physical walk-downs and visual inspections of individual ccmponents are performed and all observed leakage is recorded.

Gaseous systems are tested using Volumetric Leakrate Monitors or the equivalent, bubble testing, and airborne activity sampling in conjunction with physical walk-downs and visual inspections.

This surveillance program is presently scheduled to be performed at 18 month intervals (maximum).

The following data was collected during initial testing of individual components and fluid systems:

17

(1)

ECCS Integrated Leak Rate <1.0 gpm (2)

RCS Sample Integrated. Leak Rate <1.0 gpm (3)

CVCS Integrated Leak Rate <1.0 gpm (4)

Waste Gas System Integrated Leak Rate <0.5 SCFM 18

2.1.6.b DESIGN REVIEW OF PLANT SHIELDING AND ENVIRONMENTAL QUALIFICATION OF E UIPMENT FOR SPACES/SYSTEMS WHICH MAY BE USED IN POST ACCIDENT OPERATIONS Desi n Bases Radiation fields are considered resulting from all equi pment and piping outside the containment that might contain primary coolant or radioactive gases during post-accident operations.

Radiation fields are considered as resulting fran gaseous nuclides released from the core into the containment atmosphere as the result of a postulated LOCA.

Source terms for liquid systems assume 100$ of the core inventory of noble

gases, 50% of the core inventory of halogens, and 1$ of the core inventory of other nuclides are contained in the primary coolant.

Approprite dilution and decay factors are applied.

Source terms for containment atmosphere follow the guidance of Regulatory Guide 1.4 and assume that 100$ of the core inventory of noble gases and 255 of the core inventory of halogens are released into the containment atmosphere.

Appropriate decay factors are applied.

Source terms for gaseous systems are derived from the appropriate activity in either the primary coolant or containment atmosphere.

Dilution and decay factors shall be applied.

19

Dose rates 'at all vital areas are reviewed to ensure that personnel access to perform required tasks will be possible.

A vital area is defined as any area which will or may require occupancy to permit an operator to aid in the mitigation of, or recovery from, an accident.

t Radiation exposure to personnel is deemed unacceptable if exposures exceed the limits of 10CFR20, and GDC 19, 60 and 64 of Appendix A to 10CFR50.

Radiation exposure to the control room operators is limited to 5

Rem whole body in 30 days following the accident.

Exposure to other operating personnel performing necessary functions in vital areas is limited to 3

Rem whole body and 18 3/4 Rem to the extremities.

Acceptance criteria for dose rates in vital areas consider the appropriate dose limit, required occupancy time, round-trip access route, required frequency of access, and a safety margin.

Integrated doses are estimated for safety equipment required for post-accident operation to ensure that GDC 4 requirements are met.

Acceptance criteria for equipment doses are determined either by equipment specification, actual equipment qualification, or generic material damage data.

Doses are found by integrating dose rates over the time the equipment is required to.function.

The radiation fields present do not impair the function of counting rocm instruments or other Health Physics equipment required for sample analyses.

20

8 Function The design review evaluates the radiological envi rorment of the plant following an accident in which significant core damage has occurred.

The objective of this review is to:

a)

Insure that required access to vital areas and equipment needed for post-accident operations will not be compromised by high radiation fields, and b)

Insure that the operation of safety equipment needed for post-accident operations will not be impaired by high radiation fields.

Shieldin Descri tion Shielding is presently provided around the majority of radioactive and potentially radioactive equipment and piping.

The original design criteria and shielding descriptions are found in Section 12.1 of the FSAR.

Shielding protection is generally provided by concrete walls which are shown on Reactor Auxiliary Building General Arrangement sketches 2.1.6b-12 thru 2. 1.6b-17.

Desi n Evaluation The following systems or portions of systems have been identified as containing reactor coolant or radioactive gases:

21

a)

Containment Spray System b)

Safety Injection System

- Low Pressure Safety Injection

- High Pressure Safety Injection c)

Shutdown Cooling System d)

Sampling System

- Liquid

- Contairment H2 Analyses These systems or portions of systems that have the potential for containing highly radioactive fluids are indicated on Flow Diagram sketches

2. 1.6b-l thru 2.1.6b-11.

Source Terms The total core inventory at the time of the accident is shown in Table 2. 1.6b-1 and was taken from FSAR Table 15.4. 1-1 for Noble Gas and Halogens.

Core inventory of other isotopes were taken from Table 4.3-1 of Combustion Engineering document SYS80-PE-RG, Rev. 4, Radiation Desi n Guide, with a factor of.643 applied to account for power difference.

22

Multigroup source strengths were calculated at various times following the accident for Reactor Coolant (undiluted),

Sump Water (diluted reactor coolant),

and Co'ntairment Atmosphere.

These source strengths are shown in Tables

2. 1.6b-2,

2. 1.6b-3, and 2. 1.6b-5.

Source strengths were calculated J

from the original isotopic inventory data by separating the specific activities of the radionuclides in the various fluids into appropriate energy groupings.

The GROUP> code performed the grouping in some instances.

The specific activity of the reactor coolant was used as the source term for the liquid sample system at one hour.

The specific activity of the sump water was used as the source term for all other liquid systems, and for the sample systems at times greater than one hour following the accident.

The specific activity of the containment atmosphere was used as the source term for the Hydrogen Analyzer and gas sample.

Dose Rate Calculations Dose rate calculations were performed in areas identified as vital areas requiring occupancy and in areas containi ng safety equipment.

Sources were determined as stated in Section 4.2 and appropriate geometry factors were applied to the piping and equipment of the systems identified in Section 4. 1 as containing radioactive sources.

The shielding effect of the equipment, the fluid, and the present shield wall arrangement was considered.

The effect of

rebar, embedded

plates, and structural steel was neglected, thereby giving high (conservative) dose rates.

Dose rates were calculated primarily by the ISOSHLD2 point-kernel integration code for "simple" geometry cases, such as 23

single pipes and tanks.

The SPAN-IV3 point-kernel integration code was used to find dose rates for selected "complex" geometry cases involving a number of

pipes, tanks and shields.

Vital Areas Re uiri n Occu anc Vital areas requiring occupancy include:

-Control Rocm

-continuous occupancy

-Sample Room

-1 minute, 1 time occupancy starting at T=l hr.

-Counting Room

-Single 1 time occupancy to move equipment

-Hydrogen Analyzer 3 minutes occasional occupancy starting at T=l hr.

-Charging Pump 1A cubicle

-requires 2 men, each for 5 min.

-Charging Pump 1B cubicle

-requires 1 man for 5 min.

-Charging Pump 1C cubicle

-requires 1 man for 5 min.

-Letdown Heat Exchanger Area (Hot Leg Injection)

-requires 4 men, each for 5 min.

24

-LPSI Pump lA cubicle

-requires 3 men, each for 5-10 min.

-LPSI Pump 1B cubicle

-requires 3 men, each for 5-10 min.

-LPSI Pump 1B cubicle (Shut Down mode)

-requires 1 man for 10 min.

-Electric Penetration Room (El. 19 '6)

-requires 2 men, each for 5 min.

-Shutdown Heat Exchangers 1A, 1B (RAB, EL-5, Corridor)

-requires 4 men, each for 5 min.

Potential Vital Areas

-Power to IC-HPSI pump and IC - charging pump - frcm switchgear "C"

located on RAB, elev. 19.5',

near the CEDM MG sets (adjacent to the Sample Rocm)

-Valve station outside Equipment Drain Tank

-Valve station outside Chemical Drain Tank

-Valve station - Waste gas decay tank area 25

E ui ment Re uirin Protection Radiation damage was addressed on a generic basis.

The majority of the dose following an accident is experienced in the first few days, and 90$ of the dose occurs in the first 30 days.

Thirty day integrated doses were determined for selected

areas, and one-year post accident doses estimated from these.

Doses frcm di rect shine only are considered; airborne activity could increase the levels.

Integrated doses to safety equipment located outside the contairment were estimated.

One-year post accident doses to equipment located outside the highly radioactive cubicles on elevations - 0.5'nd 19.5're estimated as less than 105 rads in all cases, and doses of 10" rads or less are typical.

For elevation 43', the maximum dose in most areas is estimated as 103 rads.

However, doses of 10 rads may be experienced near the containment structure.

Doses on elevation 62're limited to a few r ads.

Safety equipment housed inside the highly radioactive cubicles,in the Reactor Auxiliary Building or inside the contairment could experience a post-accident dose higher than 105'ads.

Outside.the containment, this would mainly apply to instruments,

valves, pumps, and motors located directly on the lines handling the primary coolant fluids.

These systems are identified in section

3. 1.

26

Sam lin S stem A redesign of the sample system is under evaluation and is discussed in the response to requirement 2.1.8.a.

Miseel 1aneous Access may be required, for flexibility of operations, to "C" switchgear located on RAB Elevation 19.5'ear the CEDM motor generator sets.

The primary source of radiation is equipment and tubing in the nearby sample

room, which could contribute a dose rate of approximately 290 rem/hr 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after the accident, 130 rem/hr 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> later, and 40 rem/hr 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> later.

This would normally preclude access to the switchgear; However, the existing sample room will be used only once following an accident.

An initial, small

(

1 ml) sample will be drawn and the sample line then flushed with demineralized water.

The flush will eliminate the sample lines as the primary source of radiation, and the primary source of radiation then becomes the ECCS piping and equipment on the floor below, elevation - 0.5'.

This source far exceeds the remaining sources (excluding the sample room) on the 19.50'evel; i.e.,

the hydrogen analyzer and the letdown heat exchanger.

The dose rates and doses (assuming five minute exposure) in the switchgear area as a function of time are:

Time hr.

Dose Rate rem/hr.

Dose mr 920 10 330 27

100 80 These doses are sufficiently low to permit brief access to the "C" switchgear.

Estimates of Ex osure Dose rates were calculated for numerous areas in the RAB and are shown in Table 2. 1.6b-6.

Several time periods from 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to 1000 hours0.0116 days <br />0.278 hours <br />0.00165 weeks <br />3.805e-4 months <br /> following the accident were considered to allow for decay until access is required.

Personnel access and exposure are summarized below.

Primary sources of radiation on elevation -0.5'f the Reactor Auxiliary Building are the Shutdown Cooling Heat Exchangers, and the ECCS pumps.

Dose rates inside cubicles with equipment containing primary coolant range from several hundred to several thousand rem per hour, and these cubicles could be expected to be inaccessible for a period of time following an accident.

Dose rates decline to.the neighborhood of 1 to 10 rem/hr after 30 days, and access to these cubicles would be limited to a few minutes.

Dose rates in the main access corridor at elevation -0.5'ange from 1 to 10 rem/hr after one hour, decaying to less than 1 rem/hr after three days.

Access to equipment in this corridor is possible, but also must be limited to short periods of time.

Equipment requiring frequent access will be relocated, redesigned or provided with additional shielding.

Primary radiation sources on the 19.5'levation of the RAB include the sample

station, letdown heat exchanger, and the Yolume Control Tank and associated 28

piping.

In addition, a significant contribution to the radiation level originates from equipment and piping on the floor below.

Cubicles housing equipment containing primary coolant have dose rates similar to the cubicles on'levation -0.5'nd are considered inaccessible.

As

, discussed earlier, the dose rate in the sample room could preclude the use of the sample room with the presently designed sample panel.

Redesign of the sample panel is discussed in section 2. 1.8a.

Dose rates in the main access corridor of elevation 19.5's calculated at a

maximum of 14 rem/hr at one hour near Staircase RA3 decaying to less than 1

rem/hr after 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

Dose rates in the corridor are generally less than 1

rem/hr after 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />.

Access to the corridor is possible although limited in certain areas.

Dose rates in the vicinity of the contaireent hydrogen analyzer exceed 10 rem/hr and access would be limited.

Use of this analyzer for grab samples will require modifications and this is discussed in section 2. 1.8a.

Although readout is in the Control

Room, access may be required to this panel for calibration purposes.

Dose rates in the radiochemistry lab and the counting rocm allow access, but may negate the room's main function.

Therefore, a

procedure has been written and approved which provides, in case of accident, for the relocation of needed equipment frcm the counting rocm to one of four specified locations on-site.

Dose rates on elevation 43'f the RAB are generally less than 100 mrem/hr and access is generally possible except in the vicinity of the containment 29

ventilation filters and contairment building wall.

A dose rate of approximately 10 rem/hr would be experienced within 10 feet of the containment wal 1.

Exposure in the Control Room at elevation 62'f the RAB will not exceed a

total integrated dose of 5 rem for the duration of the accident in accordance with 10CFR50 Appendix A GDC 19.

The above dose rates for all elevations assume the Holdup Tanks do not contain an appreciable volume of primary coolant.

The use of these tanks to store liquid wastes with specific activity of the sump water will significantly decrease the accessibility of elevation 19.5'nd 43'.

Many locations requiring access to operate manual valves are located in areas expected to be in extremely high radiation fields following an accident, such as the ECCS room and the pipe tunnel..

Since these valves are surrounded by radioactive equipment= and piping, and are themselves located in radioactive pipe, shielding does not appear to be feasible.

Therefore, if our ongoing evaluation determines that it is necessary to ope'rate these

valves, then remote actuation will be required.

Other valves and locations, such as in the waste management

system, may be accessible following an accident with the possible addition of local shielding providing sufficient time has elapsed to allow for radioactive decay.

Further classification will be required to establish exactly when a particular valve or area must be accessed.

30

The accessibility of specific valves and locations are discussed below.

Access routes and vital areas are indicated on figures 2. 1.6b-18 and 2. 1.6b-19.

SDCS Val ve Lineup:

Prior to operation of the shutdown cooling system, several 10" valves must be manually operated:

I-V-07-008 (bypass);

V3452 and Y3453 (Shutdown Cooling Heat Exchanger inlet isolation valves);

and V3456 and V3457 (heat exchanger outlet isolation valves).

The bypass valve is located in the LPSI cubicle, and will, therefore, be inaccessible.

The four shutdown cooling heat exchanger valves are located in the heat exchanger cubicles, but are manually operated with reach rods from'utside the South wall of the cubicles.

Thus, the operator would be shielded from both the shutdown cooling and ECCS sources.

It has been estimated that one person can operate each valve in from 5-10 minutes.

The dose rates (frcm all sources) and exposures for all valve operations are as follows:

Time hr.

Dose Rate rem/hr.

Ex osure rem 10 100 17 6.3 1.7 5.7 - 11 2.1 -

4.2

.57-1.1 In addition, contributions to the total dose while travelling to and frcm the valves must be considered.

Table 2.1.6b-6 lists dose rates at various 31

locations in:the Reactor Auxiliary Building (RAB).

For example, the dose rate in the vicnity of'taircase RA-7 (near these valves) is 7.8 rem/hr at 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, 2.3 rem/hr at 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />, and 0.43 rem/hr at 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />.

Fran the foregoing exposure values, it appears possible for two people to operate these valves 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> after an accident.

Local shielding near the valve operators is being evaluated.

References

- Section 2.1. 6b 1.

E.

Ochoa, 0. Vories, "GROUP - An Isotopic Source Generation Program",

(1977).

2.

R. L. Engel, J.

Greenberg, M. M. Hendrickson, "ISOSHLO - A Computer Code for General Purpose Isotope Shielding Analysis",

BNWL-236 (1966).

3.

0. J. Wallace, "SPAN-IV:

A Point-Kernel Computer Program for Shielding",

WAPO-TM-809 (L) (1972).

32

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Core Inventory release fraction to containment air Noble Gas -

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4

@ J:>A. )M ..wR i f3 ~ Aloe keffe.fHS <<g>~7pxff /~ftflVJf>gf ....t~e.'l<4p Gr -'-c~./- 18 gag 4lt-g C. ~ JV~ c~H 4ii-3 ft ~/u~ ~7 4'Q. 1 ~ T ~ ~'/< / YP~o 4-703 O'. Z-( '3.77'2. /,7y > L <7' /,.a~ ~ (.4+ ( I 7.P / '/ '/-I. 5'7< 0'.z..'o I v-3 ~ ~ ~ I ~ ~ Cco tM 4'0 %,J hl11lff < P7> J df Qy J 4<< C e>.s ~ f=uC. S'~ Jfe,i J O XMC. HI Ji>CJ jlJx (Jv r (J,i(r) ro RLC ~ 11 ACY 1 1 1 2.1. 7a AUTO INITIATION OF THE AUXILIARY FEEOWATER AFW SYSTEM As discussed in our earlier responses, as well as in numerous NRC information requests and telephone conferences, there is a significant concern regarding the applicability of current FSAR safety analyses if the NRC requirements in this area are implemented. To resolve this issue, Florida Power 8 Light Company, in conjunction with other owners of the CE NSSS, has contracted CE to perform analyses which are expected to yield acceptable results which will resolve the outstanding questions and allow full implementation of an automated auxiliary feedwater system. Pending the completion of these analyses and subsequent NRC approval, we do not intend to implement the fully automatic system required by NUREG-0578 at this time. However, installation of the semi-automatic system described below is in progress. The isolator qualification problems discussed in response to requirement 2.1.3.b also apply to this system, but implementation is expected on or before February 15, 1980. Oesi n Bases The design pr ovides automatic start of the AFW pumps. A failure in the automatic circuit does not result in loss of manual capability. The automatic pump starting circuit is designed so that single failure does not result in loss of system function. Manual capability of pump start from the Control Rocm is retained. Single failure in the manual circuits does not result in loss of system function. 33 The AC powered pumps and valves are included in the diesel generator sequence. The system is designed as control grade. S stem Function The Auxiliary Feedwater System (AFWS) pumps are automatically started to satisfy the requirements of AFWS automatic start set forth in GDC 2 Appendix A to 10CFR50. S stem Oescri tion The AFW pumps are started once any of the steam generators reach a preset low level. The automatic AFW initiation signal starts the two motor driven AFW pumps and opens the two steam supply valves to the turbine driven pump. In case of loss of offsite power the AFW pumps are automatically loaded on the diesel generators. The system is designed as a control grade'system. Separation between safety grade components and the new control grade equipment is maintained with safety grade isolation devices. S stem Com onents The following majot equi pment is furnished as part of the control grade system: ( 8 ) Level Bistables - ccmmercial grade ( 8 ) Signal Isolators - tested and qualified to IEEE for the control rocm envi rorment ( 9 ) Relays - tested and qualified to IEEE for the control rocm envirorment Testin and Ins ection The equipment and components are purchased as specified in "System Components" above for the safety grade equipment. Yendors substantiate through test, calculations and/or operational data that the ccmponents remain operable under the seismic loads. 35 2.1.7b AUXILIARY FEED FLOW INDICATION The St. Lucie Unit 1 design incorporates safety-grade auxiliary feedwater flow indication in the Control Room which meets the intent of this requirement. 36 2.1.8a POST - ACCIDENT SAMPLING CAPABILITY A plant procedure (C-100) has been written to provide interim guidelines and instructions for sampling and analysis of reactor coolant, containment, gas decay tanks, liquid waste, plant vent and fuel building after a loss of coolant or other similar accident. Desi n Bases Plant operators shall be able to obtain a sample under accident conditions without incurri ng a radiation exposure in excess of 3 Rems whole body and 18 3/4 Rems to extremities. Plant operators shall be able to obtain samples within one hour after start of accident. Release of fission products shall correspond to assumptions made in Regulatory Guide 1.4 (Bases for Source Terms). Effects of direct radiation. from piping and canponents in the Auxiliary Building shall be considered. Effects of direct radiation from airborne effluents shall be considered. Samples shall indicate the degree of cladding failure, high fuel temperatures and degree of fue1 mel t i ng. 37 The sampling system shall have the capability to analyze RCS samples for pH, activity, Boron, Hydrogen and Oxygen. The sampling system shall have the capability to analyze the contairment air under positive and negative pressure. Function Provide information related to the extent of core damage that has occurred or may be occurring during an accident. Determine the types and quantities of fission products released to the containment in the liquid and gas phase. Provide information on coolant chemistry and contairment hydrogen. S stem Descri tion The criteria used in the design of the modified sampling system to have a post accident sampling capability shall be based on the calculated post accident radi ation dose in the area where all the sample taking, handling and analysis activities are conducted duri ng normal operation. All radiochemical and chemical analyses will be done in the laboratory facilities away from high radiation zones. 38 Sample collection, and handling, as well as in line monitoring of reactor coolant is accomplished with equipment and instrumentation that are solely dedicated to function during accident and post accident conditions. This facility is designated as the Post Accident Sampling System. The Hot Leg Loop 8 Shutdown Cooling Suction line are sampled. Figure 2. 1.8.a-1 thru 3 shows the routing. Table 2.1.8.a-2 lists the chemical and radiochemical analyses. The guideline used in the sampling and monitoring of reactor coolant is to limit the quantity of the reactor coolant in the post accident sampling facility to a minimum at all times. Sample collection and in-line monitoring is done in sequence with an intervening step consisti ng of a demineralized water purge. These procedures are graphically shown in Figures

2. 1.8a-4 through 2.1.8.a-7.

Reduction of sample collected to less than 1 ml for radiological spectrum analysis and Boron concentration. Intermittent inline monitoring of pH and dissolved gases (oxygen and hydrogen). Provisions for ccmplete purging of the sampling lines and associated equipment with demineralized water. Purging is done by remote manual operation. 39 Provisions for use of special ly designed shielded sample container that can be safely transported to a remote laboratory for analysis. The hotleg sample line (3/8" stainless steel) ties in downstream of the .primary containment isolation valve (V-5208). The high pressure and high temperature reactor coolant sample is cooled to 120'F and depressurized to approximately 25 psig. The sample line then branches out to supply a grab sample station, inline analyzers and monitors. The pressure reduction station and inline probe holders will be shielded. The shutdown cooling suction line is routed in the same manner as the hotl eg sample. The pressure reducing station and in-line probe holders will be ccmmon to both sample lines. Sample coolers will be moved to the penetration room. Remote purging of the system with demineralized water is provided. The sample lines emerging from the sample station shall be discharged back into the flash tank. Dose rate data for the optimum canbination consistent with the NUREG-0578 requirements will be provided. This combination shall be adopted as the final design of the post accident sampling system. S stem Com onents The following system ccmponents are designed to quality group D and nonseismic category I. 40 Piping, valves and associated ccmponents - constructed of stainless steel. Sample Coolers - Shell and tube type heat exchange sample stream flows through the tube while the cooling water passes through the shellside. Cooling water is supplied frcm the Component Cooling Water System. Dissolved

Oxygen, Hydrogen and pH Analyers - analyzers are in-line type with remote indicator transmitter, for readout, and/or recording.

Instrumentation - Pressure, temperature, and flow instrumentation will be provided to monitor system performpnce. TABLE 2.f.8.a'-l, CALCULATED DOSE RATES AT ACCIDENT CONDITIONS EXISTING FACILITIES AREA DOSE RATE 1. Primary Sample Room (General Dose) a)'rom a single sample cooler b) Prom single 3/8", sample line 2. Chemistry Counting Room (Hot Lab) 3. Radiochemistry Laboratory 1,000 R/HR 14,000 R/HR one ft dist. 350 R/HR 12 R/HR 12 R/HR REFERENCE 1. Marked up drawing No. 8770-G-070. General Arrangement Drawing RAB elevation 19.5. Figure 2.1.8.a-l TABLE 2.1.8.a-2 POST ACCiDENT SAtG'LiNG REACTOR COOLANT CONSTiTUENT ITS 1 METHOD OF SAtPLi~ 1G SPBPLE VOUM CR FLOWRATE RENDU HZ~T TiwiNG RESULTS l. Radiological Analysis iodine Cesium Non Volatile isotopes 2. Chemical Analysis a) Diss O~Jgen b) Diss hydrogen c) pH d) Boron GRAB in line Monitoring $ I Grab <1 ml 200 mlpm <1 ml oA 0 'rl MC ~ri O g O Cl c5 Q ca w Q C4 0A 0 C 50 r< ~R C4 g Vl 0 d 4 N 211 ~ 5m~ o< i,122,5 SAC"lPL6 GOOLEP TATioh PlP E tuvH<L PR p P0CrO >ac+ SEP 1&AVG'5 VP >4C V. 2998-5-o2'9 ~V52o5 '~~ V5202 "-p ( i5i~SI 1ll a z 111 1 Ci l2 PRCsoay RC OVC12'9 ST22 T2OS C. 22n 222 I P) I ~l) Vl Q A < X ~n Y ~ ~.2 oy ra p P 12 l5 X D + vl )W t.lVF r Roe.ee R C < E R C 2'4 C b~ E -13112 -'Alp -12'p CC52.07 S,XHVl.i><C 2'-VAEH Pl] X W/CRAH V>l1T HO, 2 ~ WOlCmt Wa Ti~ P 'llc+1P'cc / / ro cuss TANK P()gT P,QC,)DF NT SAMPLING SVSTF g ROv T'1HG SCt.lC. HA,7lc I!olI.CG F20)-I <A'APLE COCLE P ~~TAT)ohl .,iivI Poyid Cnni.iv. 5<<.'ii ii< Li>>G FVoa .',s Hr L c.DOLE't

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g.x,. MATEO TO fL V~~ t'.I'H& Q>, Pi I.GGEH9 'CHOI'E Oar:W>XTi:. 9 VS LVE YWLVc OPCAATG D PV R E/cfl PoD F Zn ~ oi I " IOi= P IIiGLD=-S IN LING PROSE Q Vlcg Dig<:ola'!'ClS REACTOR COOI.Al4T SAHPI.C r>.X. VATES g-0 lg CO.C>>l.: C Vio<<"i'. Il'roV.O I'i"..I g I/ 4/ Y-D IO g I ~5g~ 7 AINE GTc P 'go f (ip, @j, (<3~,,ii )~(-D \\ 2~ ~5 gAYipLE COLI.E c TIOH FLo~ PA,T4l yp Lgp C LprE: G 7 IO yr LYE PoLP; 8 GAHr L I:O'ECTIOH PP gC," DO E FI6VRC.: 2.!8. a R LIOfLES FaOW '5/iHPI.E CDOLE 0 STAT L OW ':.LL<rfPo eriL cnoLLLL-14>l L LLUE 'PL 0+ V>>l'~'Tao A LEGEHP DI. %PTER TO F LvC LILNS 0 REZOrF nPERr TED VAC/E YALVE OPER ATfi D .REPCtl PoG FRoM BLD E p S I<<EI.OES r.'.ZGA IN ilyE Pf-Oe, E Qlllcg DL< coL<Hf.-cl'~ B.o TE: P FTGR TIIIg .'TEl THE S>HPLE C!> BICLE CAR gP ~~LCCouggCT EO P.AD TAME H To THe. -~ 9 CONTE CT>o rC I nL Ilvoao TEsT P~ Q) gz Xu3 2q gc To Fv sp 7ANK REACTOR Cool AFOOT SANPL E. D.I MATEQ STEP go P PEt.OTE Poealw,- of ~rL ~ALE LINE AFTL=R S~HpLE, lg IDOL~TED vALUE ct.ocE: ) VALV~ OPE~ 5 l(oTLES Fao~ 5/ii 1P' COOLFP. C,fA7 ['9)J 1H LllgE YlOH', l OR1NC Pt4OC E. DVRE. 5gupooWH copula S" [i~ii l LtuE FFoH LAMPLE C,oPLE" Aflo l4 LEGEND l P.X. VIA,TEA aO FLVutONS Qi Q SIOC A ~~h(t;LD-"-> lH LlgE Pr OP.,E Q QlCg Dl<CG14$ 'GCrS hPCr'i REHOTF-OPE ~Afc' VA'C Yaiirz OVr=P~TC'C REPCtl Pc O r oH 0'~l REACTOR COOL4.V.T SAMPl E.. D, 1. V/ATER go g! +ij'")='o Fc.Asp Q4'TG P go FLOVJ'ATH. I-Ol-t>>'9 l"lOl.llf0l~lllG VALVE. CLOG"-: 6 I 9 9.o S 9,i 2.3 g/gLyr. 0 pe y; P~) Qq +[g g~~) Qe5 IIOfLEC f:ROI-I 5WI<PLa COOLE~ STAT IOhl 1H LiHE HOHlrOai<<t-P~OCEt usa F IC VR0: 2,.t.a. CL y gllvTDO+0 cooLlv-li~ ia4 LINC I R'iH Q+MPLg ( oOLGR Gras>ou 9.X. WATER TO FLV<IIIHC 2 p~ -00 gi-fMI4. 0~i LEGEHO l 1 I.EHOff," Or Ct.rrz.n WI.Vr-YALVE OpfRA14 5 QY R Cr C ll Po D rre Ig OOl SIOe A lqlI:-LOS> ~PC A IN LIHt: PROP.C. .jI.- Rotc@ Dt~r.OltttfctS RBACToR COOLANT SANpl. E. n.X. wwTEZ l9 ggggCCyiol4 fot'. }lYoao TEST P~. (~)i gm x< y'< t, (I),,,i R3 2q o> FLaSu PAN)(, &YEP go QFLOW PP.7(.l FOR Rt= HO7E P URGlHS VALVC. C LOC"-: '""E "=":O'SOSSS' 2.1.8b INCREASED RANGE OF RADIATION MONITORS A plant procedure (C-101) has been written to provide interim capabilities to measure high level noble gases through the use of standard instrumentation by providing sample dilution with preci sion flowmeters. Under accident conditions, radiation fran the exhaust air is measured through the Plant Vent and the Fuel Building Effluent Monitors utilizing the Condenser Air Ejector and the Drumming Station Monitors, respectively. In both cases continuous Control Room readout is provided and portable NMC monitors will be used for backup. Appropriate graphs will be used to determine the activity release from the monitor readouts. Permanent high range monitors are scheduled for installation by January 1, 1981. 2.1.8.c IMPROYED IN-PLANT IODINE INSTRUMENTATION A plant procedure has been written which provides that, if required during the interim period (1/1/80 to 1/1/81), air samples will be drawn through silver zeolite filters, purged of noble gases by bypassing air over the silver

zeolite, and analyzed for Iodine on a GeLi detector.

Permanent, portable monitors are being evaluated for installation by January 1, 1981. 43 2.1.9 REACTOR COOLANT SYSTEM VENTING The following is a description of the proposed design package. Desi n Bases The vent flow rate capability shall be based upon the following considerations: The vent rate shall be sufficient to vent one-half of the RCS volume in one hour. The vent mass rate shall not result in heat loss from the RCS in excess of the normal pressurizer heater capacity. The vent system shall be designed Seismic Category I and safety related. The vent system shall be designed for a single active failure. The system shall be designed to limit RCS mass loss to below the definition of a LOCA in 10CFR50, Appendix A. The vent system shall be operable following design basis events and loss of offsite or onsite AC power. The vent system shall be capable of venting directly to contaianent or to the quench tank. The vent system shall be designed to vent superheated

steam, steam water mixtures, water, fission gases, helium, nitrogen and hydrogen.

The system design parameters are as follows: Flow Design Temperature Design Pressure Line Size 100 scfm 700'F 2500 psia Control Rocm position indication is provided for the power operated valves. The system is designed not to interfere with refueling maintenance operations. Function The system allows for ranote venting of the RCS via the reactor vessel head vent or pressuri zer steam space vent during post accident situations when large quantities of non-condensible gases may collect in these high points. The system shall also be used in normal maintenance procedures. S stem Descri tion The RCS Vent System design as shown in Figure 2.1.9-1 is consistent with the design philosophy used in all engineered safety feature systems. The system is redundant and functions with allowance for a single active failure. The system is not designed to be used during plant normal operation. The power operated valves are provided with Class 1E power supplies. The system valves are disconnected from their power sources and administratively locked to eliminate the possibility of inadvertent system actuation during normal operation. In the event of an accident in which large quantities of non-condensible gases are generated within the RCS, operating procedures allow the RCS vent system valves to be connected to their power sources. The valves are also capable of being connected to an emergency power source in the event of loss of offsite power. Pressure reduction of the gas is acccmplished through the use of orifice plates in the piping system. Although the primary function of the orifice plate is to reduce vented gas pr essure, they are sized such that in the event of system failure or rupture, the RCS.mass loss through the orifice is less than a LOCA as defined in 10CFR50, Appendix A. The RCS Vent System will also be used in conjunction with plant maintenance procedures. Venting of the pressurizer and reactor vessel venting can be accomplished by directing vent flow to the quench tank for this operation. A single failure analysis verifies that the failure of any single power operated valve to open or close is not detrimental to system operation. This is accomplished through the redundancy in system valves. Consistent with NRC requirements, the system shall be designed to limit mass loss to less than a LOCA as defined in 10CFR50, Appendix A, and thus a separate analysis of inadvertent system operation or pipe breakage is not required. A Failure Mode and Effects Analysis (FMEA) is provided by the HSSS Supplier. The ccmponents, piping and supports of the RCS vent system are specified as Seismic Category I. P Piping and valves are constructed of austenitic stainless steels. The system shall be designed to Seismic Category I and ASME Code Class 1 or 2 as indicated on Figure 2. 1.9-1. Power operated valves are solenoid operated type designed to fail,closed. Redundancy in valve configuration and power supplies is designed to meet single failure criteria. Instruments,

Controls, Alarms and Protective Devices-The system shall be designed to be controlled remotely frcm the Control t

Room. Instrumentation shall be powered frcm emergency power sources and alternate sources are used as necessary to meet single failure criteria. Position indication (open/shut) shall be provided for the remotely operated valves and displayed in the Control Room. Pressur e instrumentation shall be also provided to monitor system performance and valve leakage. Pressure indication and a high pressur e alarm are also located in the Control Roan. 47 X Figure 2.1.9-1 REACTOR COOLANT GAS VENT SYSTEM RCS XS-GL-A-1-1 Reactor Vessel Vent 3~4x 1 XS-GL-A- /A-!J 3 XS-GL-A-1-2 RV-101 LC RV 102 XS-GL-A-1-2~ FC XS-88-A-!-8~ Pressurizer Vent 3)x 1 XS-HB-A-1-2 l RV-103 XS-HB-A-3/q-1 XS-HB-A-1-1 Vent To Containment KB-HB-A-1-4 RV-108 FC LC 9 R'l-10a S l=C LC 4xl; 2 XS-'.":8-A-1-2 S R'l-1C5 FC LC L{~g Qt XS-HB-A-1-2 ~ ~ ~ A -- PI R!l i ' R'll ]07 XS-HB-A->~-2 Quench Tan!< Grain XS-HB-A-1-2 S RV 106 RCS KB-HB-A-1-4 RV-109 ~. FC. LC Figure 2.1.9-1 S mbols and Abbreviations Line Oesignation: XX - XX - X X X + Safety Class (1,2,3,4) Size (Nominal Pipe) Haterial: A-Stainless Steel Design Temp ( F) Design Pressure (psig) Pressure/Temperature

Code, ls-. Letter 2nd Letter 00 YA 50 85 Class Break:

Class , Class Fail Close: FC Lock ('<ei, =I=-c-'i-,c=i } Close: LC Reactor C;o'i =-."-;.=ress "re Bou'ndary: Drain.. Solano; ' ~l Reducer/>> xpardar: Orifice Flanged Connection: Hl-Valve Designation: RV-XXX Pemote P:=ssure instrument Peadou and Higk Alarm: 2.1.9.a ANALYSIS OF DESIGN AND OFF-NORMAL TRANSIENTS AND ACCIDENTS Transient and Accident Anal sis Analyses of small break loss-of-coolant accidents, symptoms of inadequate core cooling and required actions to restore core cooling have been performed on a generic basis by the Combustion Engineering Owners Group of which Florida Power 8 Light Company is a member-The small break analyses have been ccmpleted and are reported in CEN-114, which was submitted to the Bulletins and Orders Task Force by the Owners Group. The guidelines of these analyses have been incorporated into plant procedures and initial training of operators on the new procedures has been completed. The inadequate core cooling analyses have been ccmpleted and are reported in CEN-117, which was submitted to the Bulletins and Orders Task Force by the Owners Group. The guidelines of these analyses have been incorporated into the appropriate plant procedures and training of operators on the new procedures has been ccmpleted. 2.2.1a SHIFT SUPERVISOR RESPONSIBILITIES The existing plant procedure, Duties and Responsibilities of Operators on Shift (OP-0010120) has been revised to comply with the requirements of NUREG-0578. Additionally, a corporate directive has been issued which delineates Shift Supervisor duties and responsibilities and the role of the new Shift Technical Advisor position. 2.2.1b PROVIDE ON-SHIFT TECHNICAL ADVISOR Personnel have been selected to fill the Shift Technical Advisor positions and have been given initial training to the extent possible. The Shift Technical Advisor position was filled on a continuous, round the clock basis beginning January 1, 1980. 50 2.2.1c SHIFT TURNOVER PROCEDURE Appropriate plant procedures have been revised to institute a system of shift turnover checklists which have been developed for each operating station to be ccmpleted by the off-going watch stander and reviewed by the oncaning operator. 51 2.2.2a CONTROL ROOM ACCESS Appropriate plant procedures have been revised to include means for limiting Control Room access to meet the intent of this requirement. 52 2.2.2b ON-SITE TECHNICAL SUPPORT CENTER TSC St. Lucie Plant has designated the training classrocm area (formerly the Unit 2 Control Rocm located adjacent to the Unit 1 control room) as the On-Site Technical Support Center (TSC). A new St. Lucie Plant procedure, which has been drafted and reviewed by the Facility Review Group, delineates the activation, manning, and use of the TSC during emergencies. The following is a description of the proposed design package. Desi n Bases Acccmodation for twenty-five occupants shall be provided. Telephone communications shall be provided to reach: the offsite emergency response

center, the NRC, the control roan, the operational support center, Combustion Engineering and the FPL General Office build'i ng.

The following documents shall be available to the TSC: System Descriptions FSAR 8 Emergency Plan Emergency Procedures P&ID's and Electrical Schematics General Arrangement Drawings 53 Mater shall be provided for the twenty-five occupants for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> following a LOCA 8 DBE (75 gal. 9 I gal/man/day). Food shall be provided for the twenty-five occupants for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> following a LOCA 8( DBE. Two sanitation kit III's shall be provided. (One kit provides for 25 persons for two weeks). First aid supplies shall be provided in a 36 unit industrial first aid kit. Portable fire protection shall be provided. Several systems are being evaluated for the display of plant data in the TSC. The HVAC system shall be upgraded to maintain a "Control Room Level" envirorrnent during POST/LOCA envirormental conditions. Portable monitors will be provided in the TSC to monitor both direct radiation and airborne.radioactive.contaminants. Function The Technical Support Center shall be an area outside of and adjacent to the Control Rocm for the use of technical and management personnel in support of the operating staff during Post Accident conditions. Figure 2.2.2.b-l shows short and long term general arrangement of the TSC- System Descri tion While several systems. for the display of plant data in the Technical Support Center are under consideration, only one is sufficiently developed and reviewed at this time to the point that a detailed system description can be provided. This is the closed-circuit television system described in the following paragraphs. Other systems undergoing evaluation include a ccmputer-based system designed by our NSSS vendor and other designs which duplicate, in the TSC, selected Control Rocm meters, recorders and associated equi pment. The television system under consideration for the Technical Support Center includes six high resolution black and white cameras and one color camera. Each camera will be complete with monitor and camera control, for the purpose of monitoring meters,

gages, recorders,

controls, switches and status lights mounted on control and protection panels in the main Control Room.

The television cameras, located in the Control Roan, are ceiling mounted on separate pan - tilt mounts. Each camera, including the color camera, is equipped with a 10:1 zocm lens and with the necessary amplification for the purpose of close observation of panel meters and control status (1/8 of an inch minimum char aeter size). The remote controls for the pan - tilt and zocm functions are located in the Technical Support Center. The color camera is provided to ascertain the status of certain color recorders on the Post Accident Monitoring Panel. 55 A 17 inch high-resolution monitor is provided for each black and white camera

output, as well as a

17 inch color monitor for the color camera output. These monitors are ceiling mounted in the Technical Support Center. Corresponding remote pan - tilt and zoom controls are located beneath each monitor. This mode of operation provides easy access for the viewi ng of any one of the critical control panels in close-up or expanded view. A five foot, high resolution, screen television projection system is provided in the Technical Support Center, and a 14 x 1 video switcher is provided for the purpose of viewing any one of the high resolution camera outputs. The remote camera electronics are located in a separate cabinet in the Technical Support Center, and will be used to align the cameras and maintain adequate picture quality. Since the Technical Support Center is designed for two unit operation (St. Lucie Unit 1 and the St. Lucie Unit 2, now under construction), a 4 x 1 switcher is provided for each monitor and control location to select between the two Control Rooms. The switcher for the large screen display includes the expanded facility for two Control Rooms. These additions assume the duplication of the~cameras and their respective controls. The CCTV system shall have the capability to historically record all monitored parameters via CCTV video tape playback. Initiation of the historical record shall be from the Control Room. 56 ,r ft V'o ~ - ~ o ~... ~. ~ ~ ~ o ~ e ~ ~ t~ ~ 0 ~0 ~ ~ a~ 4 t.oo Q. ~.0V AI %1 ~ ~ to oo rt'L 1 m ~ 01 1 >>gO oo 1 O0 0 00 0 OOWP00OOI o ~ ~ ~ W ~ ~ W ~ ~ ~ ) ~ ~ ~ ~ ~ ~ g. ~ ~ ~ "t. t ~ ~ 1 ~ ~ ~ ~ W', 100% 0 W SOI0 V Por ~ ~ Po' ~ ~ ~ 1 ~ ~ C ~ ~ 1<<11 ~ ~ ~ ~ ~ ~ '\\ ~ ~ ~ ~ i ~ ~ ~.i ~ ~ ~ 0 ~ ~ ~ ~ 0 4 ~ ~ ~ ~ ~ I ~ ota ~ ~ ~ 0 ~ ~ f ~ ~ to1 ~ ~ ~ ~ o ~ I ~ ~ ~ '<<Ptr<<<<<<10 PV>>V0 t04OOVO I ~ I 4LGB.. 0 I,~ ~ ~ ~ aEVTt<c ~Co ~ecPmk ~ 1 ~ .4 ~ ~ ~,I ~ Io,g t ~ ~ ~', ~\\ ~ wo ~ ~ ~ ~ ~ ~ I f ~ ~ I ~'o I1 ~ ~ ~I, I ~ ~ ~ ~ ~ ~I t ~ * ~ ~ 0 ~ ~,t ~ ~ ~ ~ 0 1 I I 0 0 V Pt , ~ ' ~ IIP ~ ~ I ~ ~ 1 ~ qI~, ~ ' ~ 1% 0 <<1 No,I ~ ~ ~ 10 wo' ~ ~ ~ ~ 01 ~ ~ ~ ~ ~ I P~POW04 ~ ~ I gt ~ ~ ~ W ~ 1' ~ I<<v4 110 ~ ~ ~ 0 ~ ~ 1 W I ~ t' ~ ~ ~ ~ p &d'. ~ ~ 0 ~ 0 ~ 10 ~ ~ 11 ~ ~ t ~ I ~ Q ~ 'I " ".USQKY. ~ ~ ~ I 0' ~ ~ ~~ ~ ~ ~ 0 'V ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ g ~ gating~<< cVc4ha= -~>~~'% ~ ~ ~ 1 ~ I ~.Q;Q Ciy.Q ~ f' ~ ~ .14 C 0>> l A~ cp'=kC< t"0 'QlS Q g t+IA Qf;( Hgj'3 5'/;I Ca{. bt'd ag ~

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