ML20008G279

Final Design Review of Plant Shielding of Spaces for Post-Accident Operation, Revision 1

ML20008G279
Person / Time
Site:	Beaver Valley
Issue date:	04/25/1980
From:	Booth H, Larson A, Younger J QUADREX CORP.
To:
Shared Package
ML20008G276	List: ML20008G275 ML20008G279
References
RTR-NUREG-0578, RTR-NUREG-578 QUAD-1-80-040, QUAD-1-80-040-R01, QUAD-1-80-40, QUAD-1-80-40-R1, NUDOCS 8107070258
Download: ML20008G279 (101)

Text

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ATTACIDIENT 1 DUQUESNE LIGHT COMPANY Beaver Valley Power Station, Unit No. 1 1

Design Review of Plant Shielding of Spaces for Post-Accident Operation

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1 8107070258 010630 D, DR ADOCK 05000

QUAD-1-80-040

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DESIGN REVIEW OF PLANT SHIELDING OF f

SPACES FOR POST-ACCIDENT OPERATION NUREG-0578 SECTION 2,.1.6.b i

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DUCLEAR SERVICES CORPORATI0n t

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i WORDREX compomanom Campbell, California Prepared For DUQUESNE LIGHT COMPANY g

i Prepared by:

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Reviewed by:

. sluk #1umo i

j H. R. Booth M. Kasahara g-Q f d

J. J. Younger Y. Cheng q

A. Larsen S. A. Hobart Approved by:

RevisionNo.l Data NSC Job Humber l Issued By l

Data l2-19-f,0 (d9 L, t au 2-19-80 0

D00-0316 i

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QUAD-1-80-040 b

TABLE OF CONTENTS PAGE

1.0 INTRODUCTION

1-1 1.1 NUREG-0578 1 -1 2.0

SUMMARY

2-1 E

2.1 Chemical and Volume Control System 2-1 g

2.2 Containment Isolation 2-1 2.3 Safety Injection System 2-2 2.4 Containment Depressurization 2-3 2.5 Containment Vacuum and Leakage Monitoring System 2-4 2.6 Supplementary Leak Collection and Release System 2-4 2.7 Post DBA Hydrogen System 2-4 2.8 Post-Accident Sampling System 2-5 3.0 BASIS OF EVALUATION 3-1 3.1 NUREG-0578 3-1 3.2 Source Tenns 3-1 3.3 CIA, CIB Actuation 3-3 3.4 Radionuclide Flow Paths 3-4 3.5 Vital Areas 3-4 3.6 Occupancy and NRC Radiation Exposure Criteria 3-5 3.7 Shielding Calculation Description 3-6 4.0 SYSTEM EVALUATION AND RECOMMENDATIONS 4-1 4.1 Normal and Alternate Letdown Flow Paths Under Accident 4-2 Conditions 4.2 Safety Injection System 4-8 4.3 Containment Depressurization System 4-13 4.4 Containment Vacuum and Leakage Monitoring System 4-14

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4.5 Supplementary Leak Collection and Release System 4-16 l

4.6 Hydrogen Recombiner 4-18 l

4.7 Post-Accident Sampling System 4-21 l

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5.0 REFERENCES

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QUAD-1-80-040 LIST OF FIGURES FIGURE NUMBER NAM 2 l

3-1 Control Room 3-2 Post Accident Sampling Station O

33 n thway to Recombiner Room in Accidev'..andition a

1 4-1 Potential Radionuclide Flow Paths with Use of Nonnal CHS Letdown Line 4-2A Potential High Radiation Areas from Use of Normal CHS Letdown Line - Plan El. 722' -6" in Auxiliary Building 4-2B Potential High Radiation Areas from Use of Normal CHS Letdown Line - Plan El. 735' -6" in Auxiliary Building 4

4-2C Potential High Radiation Areas from Use of Nonnal CHS Letdown Line - Plan El. 752' -6" in Auxiliary Building 4-3 Alternate Letdown Flow Paths for Containment Isolation 4-4 Potential Radionuclide Flow Paths in Safety Injection System Recirculation Phase 4-5 Potential High Radiation Areas During Safety Injection System Recirculation Phase O

4-6 Potential Rsdionuclide Flow Paths Through Recirculation Subsystem with Outside Recirculation Spray Pumps 4-7 Radioactive Gas Flow Paths of Containment Pressure Sensing Subsystem L

4-8 Potential High Radiation Area of Containment Pressure W

i Sensing Subsystem l

4-9 Potential Radionuclide Flow Paths in Supplementary Leak Collection and Release System l

4-10 Potential High Radiation Areas of Supplementary Leak Collection and Release System 4-11 Simplified Flow Diagram of Hydrogen Recombiner System 4-12 Recombiner Room Layout 11 l

A QUAD-1-80-040 LIST OF FIGURES (Continued)

FIGURE NUMBER NAME 4-13 I

Existing Reactor Building Sampling System and Proposed Post-Accident Sampling System

,O 4-14 Location of Post-Accident Sampling System 4-15 Arrangement for Post-Accident Sampling System 4-16 Flow Diagram of Post-Accident Sampling System I

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QUAD-l'-80-040 LIST OF TABLES TABLE NUMBER NAME 2-1 Sumary of Dosage and Occupancy Requirements O

3-1 Radiation Source Tenn Under Post-LOCA Conditions 3-2 Reactor Parameters 3-3 Liquid Sa rce F

3-4 Gas Source 3-5.1 through Containment Penetration Checklist (Outside Containment) 3-5.8 3-6.1 through Containment Penetration Checklist (Inside Containment) 3-6.8 3-7 Identification of Radioactive Flow Path Systems 3-8 Total Gamma Ray Linear Attenuation Coefficients of Concrete (2.30g/cmz density) 4-1 Radiation Levels in CHS with Use of Letdown Line 4-2 Radiation Levels in Safety Injection System 1

4-3 Radiation Levels in Outside Recirculation Spray Pump System 4-4 Sumary, Radiation Level Outside Containment for a

Containment Vacuum and Leakage Monitoring System E

4-5 Radiation Levels in Supplementary Leak Collection and Release System 4-6 Hydrogen Recombiner Piping 4-7 Radiation Levels in Hydrogen Recombiner Control Area 4-8 Sampling Requirements from NUREG-0573 I

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QUAD-1-80-040

1.0 INTRODUCTION

1.1 NUREG-0578 g

The Office of Nuclear Reactor Regulation of the U.S. Nuclear Regulatory Comission established the TMI-2 Lessons Learned Task Force shortly after the TMI-2 accident in the spring of 1979. The purpose of the Task Force is to identify and evaluate those safety concerns raised by the TMI-2 accident that require generic licensing actions (beyond those already specified in I.E. E,ulletins and Comission Orders). The Task Force is charged to identify, analyze, and recomend changes to both specific licensing requirements and the general licensing process for nuclear power plants based on the lessons learned.

l In July 1979, the Task Force issued its first report, NUREG-0578, "TMI-2 Lessons Learned Task Force Status Report and Short-Term Recomendations".

This report contains a set of short-tem recomendations which are to be implemented in two stages over the 18 month period following publication.

The recomendations are to be implemented for operating plants, plants under construction, and plants with pending construction permit applications.

There are 23 specific recomendations in 12 areas where implementation

g is judged to provide substantial additional protection for the public 5

health and safety.

This study by Nuclear Services Corporatien specifically addressed im-5 plementation of Section 2.1.6b of NUREG-0578, " Design Review of Plant f

Shielding of Spaces for Post-Accident Operations" for Bdaver Valley Unit l

1. This particular section is concerned with post-accident control of l

radiation in systems outride containment. At TMI-2, systems outside the containment building contained radioactive material to the extent that l

personnel access was impaired due to high radiation levels. Several i

deficiencies were noted. For example, the licensee had little knowledge of the plant's operational leakage pathway characteristics, expected post-accident radiation levels, and area occupancy requirements.

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i l-QUAD-1-80-040 W

Shielding provision for personnel access were inadequate. Section 2.1.6b recommends:

(1) performing a shielding design review of systems processing primary coolant outside containment, (2) identifying any

O areas or equipment that are vital for post-accident occupancy or operations, and (3) taking steps to assure that access and perfonnance will not be unduly impaired due to radiation from these systems.

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l QUAD-1-80-040 2.0

SUMMARY

Dosaga and occupancy requirements of the systems which have post-accident radionuclide flow paths from inside containment to outside containment are sumarized in Table 2-1.

2.1 Chemical and Volume Control System Use of the Chemical and Volume Control System (CHS) must be avoided when the reactor coolent is highly radioactive due to an accident. A pathway for the radioactive reactor coolant is through the CHS letdown line. Use of the nonnal letdown line will most likely result in highly radioactive liquid and gas flow in the CHS, boron recovery system, and gaseous waste I

system.

Therefore, an alternate letdown flow path through the existing excess letdown heat exchanger, under containment isolation, was investigated.

In this scheme, the containment isolation is assumed so that the radioactive gases and liquid are contained in the containment. The charging pumps are assumed to take suction from the refueling water storage tank. The nonnal letdown line in CHS, however, is closed to preven *. radioactive reactor coolant from flowing out of the centtinment into the CHS and other systems.

It is y

further assumed that two reactor coolant pumps are running. Under these B

conditions, the excess letdown flow line through the excess letdown heat exchanger may be effectively used to maintain the water levels in the pressurizer and the reactor pressure vessel. The letdown flow is discharged to the containment sump through a relief valve in the bottom of the primary drain transfer tank (DG-TK-1). To ensure the letdown flow into the con-tainment sump, it is recomended that the presently existing lock-shut

, manual valve in the excess letdown bypass line to the containment sump be replaced with a remotely operated valve.

2,? Containment Isolation Containment Isolation Train A (CIA) signal will automatically close the letdown orifice isolation valves and stop the letdown flow from the reactor coolant system. The CIA signal is initiated by a containment high pressure signal or a safety injection. This project responds to TMI Lessons Learned, 2-1

QUAD-1-80-040 1

Section 2.1.6b, additional shielding required for the increased radiation levels. The response to this section does not address high containment pressure of CIA actuation in a high radiation situation without a pipe 8

break.

In the future, Duquesne may have to respond to a high radiation CIA actuation in response to radiation in the primary coolant loop.

l 2.3 Safety Injection System i

During the safety injection system recirculation phase, the contact dose rate on the floor above the pipe tunnel in the safeguards and containment contigous areas, when the low head safety injection pump discharge lines are recirculating the radioactive containment sump water, is estimated to be 5 R/hr one hour after the accident. Similarly, the contact dose rate on top of the pipe trenches in the Auxiliary Building at Elevation 722'-6",

wnere the charging pump lines are flowing the containment sump water in the recirculation phase, is estimated to be 22 R/hr at ene hour after the accident. Since these estimates are contact dose rates, the actual total body exposure rate to personnel wili be much less, and the personnel should be able to pass by the treas, as required, under administrative control.

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There is an area where recirculation lines are exposed in the west-wall E

pathway and over the pipe trench No.1 near the charging pump cubicles at elevation 722'-6" in the Auxiliary Building. A very high radiation background will exist in this area during the recirculation phase and e

personnel should be restricted from this area.

Since the boron injection tank (BIT) in the Auxiliary Building at

, Elevation 722'-6" has 900-gn11on capacity, the radiation level outside the BIT cubicle remains high with the radioactive containment sump water flowing through the BIT during the safety injection system recirculation phase. The l

calculsted contact dose rate at the outer surface of the 2-ft concrete shield wall is 810 R/hr at one hour after the accident. Although occupancy is not required in the vicinity of the BIT under accident conditions, strict administrative control is necessary to ensure personnel are restricted from the BIT area.

2-2

QUAD-1-80-040 Shortly after switch over to recirculation phase, the operator must operate the HHSI/ charging pump suction and discharge valves manually from the south wall of the pump cubicles to set up re Q ndant and independent flow paths from containment sump to cold legs.

Although the operator will be exposed to radiation from the pipe trench No. 2 and Boron Injection Tank, the operator will receive dosage which is appreciably less than the allowable limit of 5 Rem whole body.

The following cnange in operating procedure, however, is recommended to reduce radiation exposure:

o Switch from injecting through the Boron Injection Tank line to the HHSI discharge through cold legs prior to start of recirculation phase, or i

o Delay switching from injection through the Boron Injection Tank line to the HHSI discharge through cold legs during recirculation phase to allow sufficient time for radioactive decay in the Boron Injection Tank and recirculation lines.

Tne recombiner inlet manual valves and cross-connect instrument air valve (llA-90) to the containment instrument air are located in the pipe penetration area of safeguards where the LHSI lines with highly radioactive containment samp water 'are routed.

It is reconnended that these valves be made remotely operable.

2.4 Containment Depressurization Recirculation spray pumps of the Containment Depressurization System, which take suction from the containment sump, are used to tr3intain the containment at subatmospheric pressure following an accident.

Two of I

the four re~irculation spray pumps are located outside the containment.

c The outside recirculation spray pumps, components, and piping are in enclosed shielded a eas which are located in the west safeguards area.

There is no need for personnel access to the pump cubicle area.

Rev.-1 2-3

QUAD-1-80-040 2.5 Containment Vacuum and' Leakage Monitoring System I

The Containment Pressure Sensing Subsystem of the Containc.,ent Yacuum and Leakage Monitorinrl System will contain radioactive. gases after an accident.

Four 3/8" pipes are led from inside containment to the pressure sensing instruments which are enclosed by concrete shield walls outside the con-tainment.

Based on the estimated radiation levels and the occupancy re-quirements, it is concluded that personnel exposures can be maintained within the allowable limits without any shielding modifications.

2.6 Supplementary Leak Collection and Release System 4

The Supplementary Leak Collection and Release System ensures that radio-active leakage from the reactor containment is collected and filtered for iodine removal prior to discharge to the atmosphere at an elevated release point. The calculated radiation levels after an accident indicate that passage near the vent ducts and the shielded filter banks is permissible i

for personnel as the need arises. It is expected that post-accident occupancy requirements in the vicinity of system ducting and the filter bank will be minimal. Therefore, it is concluded that additional shielding

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for the system is not necessary. Fertonnel exposures can be controlled by implementing personnel access restrictions.

2.7 post-DBA Hydrogen System

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Two redundant hydrogen recombiner systems are installed in the safeguards

!O area to maintain hydrogen content in the containment below the lower flammability limit in the event of LOCA. There is a biological shield wall between the recombiners and the control panels. Calculations l

' indicate that an operator standing on grating in the centrol panel room L

will receive most of the radiation dosage from two low head safety injection lines which are routed approximately 14 feet below the 2-ft concrete floor which is adjacent to the grating. Ne appreciable shielding protection is afforded to the operator by the 2-ft concrete floor. A radiation level of 3000 R/hr at the operator's location one hour after l

2-4 1

QUAD-1-80-040 1

the accident has been determined. At one hour after the accident, the operator will have to start the H analyzer from the control panel. The 2

total amount of occupancy time required for this action should not be more than 30 minutes. At approximately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the acc' dent, the recombiner is started from the control panel. Approximately 30 minutes are required for starcup. Based on the dose rates and personnel occupancy requirements at this locativ, personnel exposures far in excess of permissible limits v?uld result. Therefore, the control panels must be relocated to a concrete floor area or the control panel room must be shielded from the LHSI lines. One additional problem area exists that requires resolution. Manual valves 101,102.103, and 104 on the recombiner inlet line have to be opened prior to start of the hydrogen analyrer and recombiner. These valves are located in the pipe penetration area of the safeguards area where the LHSI lines containing highly radioactive containment sump water are routed. Based on the high radiation level predicted for the LHSI lines and requirement for operation of these valves, it is expected that personnel overexposure would occur with the existing arrangement. Therefore, it is recomended that the manual valves be either changed to remotely operated valves or made to open from the floor above with shielding and using reach rods.

2.8 Post-Accident Sampling System It was determined that the existing reactor building sampling system was O

not designed w the post-accident sampling. The containment isolation valves (both inside as well as outside containment) of all sampling lines are operated from one solenoid valve and a limit switch inboard g

(SOV-SS-000DA), and one solenoid and limit switch outboard (SOV-SS-000B).

g The existing sampling panel was not intended for use with source term radiation levels. Therefore, it is recomended that a post-accident shielded,ampling system be built for sampling both, liquid hot leg and containment air samples.

'e 2-5 1

TAELE 2-1 SliyttY Or DUSAGE AND OCCUPAtlCY REQUIREMENTS RADIATION LEVEL VS. TIME (HRS)*

SYSTEM AREA EQUIPMENT S!ilEi nlNG TOi0 ONE 10 24 REQUIREMENTS RE50tuil0N CHS With Use Floor Above Chars ig and let-l' Thick 32 5.2 2.4 Unlikely Administrative of Letnown Penetration Roon down Lines Concrete Floor R/:'-

R/Hr R/Hr Control li"'

Auxiliary Bldg.

Pipe Chase 2' Thick 22 3.6 1.7 Possibly Administrative (Letdown or Concretc Shield R/Hr R/Hr R/Hr Control Charging Line)

Auxiliary Bldg.

Outside Volume 3.5' Thick 41 3.4 1.2 Possibly Ad:ninis t ra t ive Control Tank Concrete Wall R/Hr R/Hr R/Hr Control Enclosure Auxiliary Bldg.

Outside Charging 2' Thick 3.9 0.7 0.3 Possibly Administrative Pump Cubicle Concrete Floor R/Hr R/Hr R/Hr Control and Wall Safety West Safeguards LHSI Pump Cubicle -

2' Thick Floor 2.4 0.6 0.3 Unlikely Administrative injection' Directly Above R/Hr R/Hr R/Hr Control bYSl**

Safeguards and Pathvay Above Pipe 2' Thick Floor 4.5 1.0 0.4 Unlikely Administrative Rc culation Cont. Contiguous Tunnel - LHSI Pump R/Hr R/Hr R/Hr Control Areas Lines Auxiliary 2 da.

Pathway Above Pipe 2' Thick Floor 21.8 4.8 2.4 Possibly Administrative Vault Charging Lines R/Hr R/Hr R/Hr Control Auxille.y Bldg.

Charging Pump 2' Thick Floor 410 96 47 Unlikely Ad nini s tra ti ve Cubicle - Directly mR/Hr mR/Hr mR/Hr Control Above

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8 10 96 47 possi g i..

Administrative 1

Aux.11ary Bldg.

Charging Pump 2' Side Wall Cubicle - Standing mR/Hr mR/Hr mR/itr Control Aside Auxiliary Bldg.

Outside Baron 2' Thick Wall 810 180 07 Unlikely Administrative injection Tank R/Hr R/He R/tfr Control Containirent West Safeguards Outside Recirc.

2' Thick Floor 2.4 0.5 0.3 Unlikely Administrative Depressuri:a-Spray Pump Cubicle -

R/Hr R/Hr R/Hr Control tion Outside directly Above Recirc. Spray West Safeguards Recirc. Spray &

2' Thick Floor 4.8 1.2 0.6 Unlikely Administrative Pump System LHSI Pump lines -

R/He R/Hr R/Hr Control s Pipe Tunnel

• Hours elapsed following time zero of the accident or event. Radiation level is in contact with shieldiry.

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QUIRE S(Con ued)

RADIATION LEVEL VS. TIME (HRil*

SYSTEM AREA EQUIPME' SHIELDING ZERO ONE 10 24 REQUIREMENTS RESOLUTION Containment Contiguous Area 3/8* Pipe. Valves 2' Thick Wall 6.8 1.3 0.5 Infrequent Aeninistrative Vacuum and to Containment and Instrument mR/Hr mR/Hr mR/Hr Control Leakage (4 sets)

Monitoring

System, Containment Press. Sensing Subsystem Supplementary Safeguards and Hain Duct at 10' None 200 Infrequent Aaninistrative Leak Collection Containment Distance mR/Hr Control 9

5 Main Duct at 20' None 74 Infrequent Administrative t

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Distance mR/Hr Control Auxiliary Bldg.

Auxilfary Bldg.

Outside Filter 2' Thick Wall 290**

Infrequent Administrative Bank mR/Hr Control Recombiner Hydrogen LHSI Line Below Concrete Shield 3,000 1.400 1.000 One hr. Max.

Relocate control System Recombiner Control Floor and Control Between Control R/Hr R/Hr R/Hr residency per panels to concrete Panel Panel Room on Panel and person floor area or Grating Recombiner shield control panel room from LHSI lines Post-Accident Auxfilary Bldg.

(Radiation Level will be Low)

Infrequent Administrative Sampling Control System (Proposed)

Hours elapsed following time zero of the accident or event. Radiation level is in contact with shleiding.

• • Assume all lodines flitered for 30 minutes after time zero.

QUAD-1-80-040 3.0 BASIS OF EVALUATION 3.1 NUREG-0578 l

NUREG-0578, Section 2.1.6b is addressed in this study and described in the Introduction Section 1.0.

In analysis of the original source tems and sources for radioactive material transport outside containment, two general accident conditions are important in assessing the effects of system shielding.

In the first case, no breach in the Reactor Coolant System integrity is assumed and that water being handled originates in the reactor coolant system. The source terms used for this analysis are listed in Table 3-1

,y under Liquid Systems. Because of the assumption of no breach of the reactor coolant line, the fission gases are assumed to remain in the water where the dilution volume is only the contents of the reactor coolant system itself.

P In the second case, it is assumed that the systems are handling water which originates in the containment sump. Because the equilibrium value of gas removal from the water has not been reached at the time immediately following the accident, a maximum source tem value was assumed where g

all the noble gases and halogens remain the the water phase. The dilution volume is assumed to be all water in the refueling water storage tank, accumulators (3), chemical addition tank, and boron injection tank, or 3

65,500 ft including the reactor coolant system.

The containment atmosphere was assumed to follow the guidelines as outlined in Table 3-1 Containment Air. Although an intact coolant gaseous source term inside containment would be minor when compared to a

' LOCA condition, the maximum source tem was used for all situations (both LOCA and no breach of Reactor Coolant Systems conditions).

3.2 Source Terms The radiation source term used for the Beaver Valley Unit 1 post-LOCA radionuclide distribution and shielding study was based on the guidelines 3-1

QUAD-1-80-040 given in TID-14844 and TMI-2 Lessons Learned Short Tenn Reconrnendations, Section 2.1.6b as outlined in Table 3-1.

The digital computer code designated as "Origen" was used to obtain the initial radioactivity released and subsequent containment / reactor water O

radiation level calculations. Origen calculates detailed isotopic compositions for a range of reactor conditions including fuel irradiation, neutron activation, and radioactive decay. The particular parameters used in Origen for the Beaver Valley Unit 1 study were acquired from BVPS and are outlined in Table 3-2.

The final computer output is summarized in two printouts, one for the reactor coolant and one for containment atmosphere. Each printout is broken up into 12 groups of different mean energies in photons /second and MeV/ watt-sec units. The printouts are reproduced in Table 3-3 for Liquid and Table 3-4 for Containment Air.

The liquid source terms from Table 3-3 were used as follows:

o For " intact" coolant lines (no pipe break but high radioactivity),

the radioactivity was assumed to be unifonnly distributed in the 3

9400 ft reactor coolant volume. For example, dose rate calcula-O tions for the normal letdown flow path (Section 4.1.1) were based on this radioactive intact primary coolant flowing in the Chemical and Volume Control System.

o For a LOCA condition, it was assumed that all of the stored water in the refueling water storage tank, acc.umulators (3), chemical addition tank, and boron injection tank was injected into either the Reactor Coolant System or containment spray and ultimately appeared in the reactor containment sump. The total volume of that 3

water is approximately 65,500 ft, including the reactor coolant..

For the Safety Injection System and Containment Depressurization System which recirculate the reactor containment sump water, the radioactivity was assumed to be distributed in the 65,500 ft of water.

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QUAD-1-80-040 b

o The gaseous source tem from Table 3-4 was assumed to be unifomly 6

3

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distributed in the 1.86 x 10 ft containment free volum. The resulting concentration of radioactivity was used in the dose rate calculations of the Post-DBA Hydrogen System and containment pressure

!O sensing subsystem of the containment vacuum and Leakage Monitoring System.

3.3 CIA AND CIB Actuation The Containment Isolation phase A (CIA) signal isolates all non-essential process lines on receipt of a signal from the actuated Safety Injection System (SIS). The Containment Isolation phase B (CIB) signal isolates remaining process lines (which do not include safety injection lines) on receipt of two out of four high-high containment pressure signals. Each line penetrating the containment has redundant isolction valves.

Tables 3-5.1 through 3-5.8 give information on isolation valves installed outside containment. Tables 3-6.1 through 3-6.8 give similar information on isolation valves installed inside containment.

SIS and CIA are activated by any of the following.

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Low pressurizer pressure in coincidence with low pressurizer water level.

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High containment pressure.

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Pressure in one steam line lower than the pressure in both of the other lines by a predetemined and set amount.

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High steam flow in any of the three steam lines with low steem line f

pressure or low-low average tanperature in the Reactor Coolant System.

4 5.

Manual action.

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QUAD-1-80-040 l

CIB is actuated by high-high containment pressure.

3.4 Radionuclide Flow Paths Section 2.1.6b of NUREG-0578 is concerned with post-accident control of radiation emanating from systems outside containment. Table 3-7 summarizes I

the systems that were considered for this study and identifies those systems as to which could transport radioactivity to areas outside the containment.

I 3.5 Vital Areas Areas which mr.st be accessible for post-accident operations (vital areas) are discussed below and are taken into consideration in system evaluations presented in rection 4.0.

Accessibility to these areas is included in

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the discussion below.

3.5.1 Control Poom The main control room is located in the service building between the reactor and the turbine portions of the stations. The control room is in such an area as to allow easy access in an accident condition and remain relatively unaffected by the radiation levels.

The area and design are shown in Figure 3-1.

3.5.2 Shutdown Panel A shutdown panel is located two floors below the main control room.

This panel provides the capability for hot shutdown if the control room is uninhabitable.

Events making the control room uninhabitable will not render the shutdown panel area uninhabitable. The area is easily accessible and unaffected by accident conditions.

3.5.3 Emergency Power Supplies These supplies are located below the main control room and are an easy access area relatively unaffected by an accident condition.

3.5.4 Instrument Areas / Local Operating Panels There are no instruments or local operating panels in areas of high radiation deemed necessary for usage in an accident condition.

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QUAD-1 040 1

3.5.5 Radiochemis*.y Laboratory The hot laboratory is surrounded by the change area, and a clean 3

shop. These are areas of low contamination in an accident condition.

Although the laboratory would be accessible for chemical analyses, the background radiation may be too high to use the existing counting g

equipment for accident monitoring. This subject needs to be addressed P

further in response to NUREG-0578, Section 2.1.8.

a 3.5.6 Post-Accident Sampling Station The post-accident sampling station will be located on plan elevation 735'-6" on the back of the existing sample room wall as drawn in Figure 3-2.

3.5.7 Recombiner Room The Recombiners and hydrogen analyzers are vital components for post-accident operations. The entrance to the recombiner panel, under acc Hen? conditiont, is as drawn in Figure 3-3.

The control panel rown 1:, on grating and the operator will be exposed to high D.

radiation from the low head safety injection (LHSI) lines in the recirculation mode (maximum 3000 R/hr at one hour after accident).

The LHSI lines are routed below the control panel room floor.

Recommendation for telocating the control panel or shielding from B

the LHSI lines is provided in Section 4.6.

To turn the manual 7

valves (number 101,102,103, and 104), the operator must go to the 722' level. The predicted radiation level at this floor will be approximately 3000 R/hr at one hour cm.' the accident with the system in the recirculation mode. This radiation level is too high to permit personnel access to the aree.. Reconinendation for remote operation of these valves is provided in Section 4.6.

3.6 Occupancy and NRC Radiation Exposure Criteria Required occupancy of vital areas (specified above in Section 3.5) for post-accident conditions must be given considervle attertion in emeroency 3-5

QUAD-1-80-040 planning. These occupancy requirements must be carefully evaluated in concert with existing shielding design in vital areas and radioactive flow paths to ensure that plant personnel exposures are maintained O

within the limits specified by the NRC.

P; Based on the NRC definition of requirements for post-accident conditions specified in NUREG-0578, the basic dose rate (in vital areas) should be such that the guidelines of General Design Criteria 19 (GDC 19) of Appendix A of 10 CFR 50 should not be exceeded (during the course of the accident). GDC 19 limits the dose to an operator to 5 Rem whole body or its equivt. lent to any part of the body. These exposure limitations 4

are factored into system evaluations in Section 4.0 of this study. Areas which require higher periods of occupancy obviously require lower dose rates than areas where lesser occupancy periods are required.

Most of the post-accident dose rates presented in Section 4.0 of this study are near contact with existing shielding. Therefore, the total body dose rate for an individual passing near (or occupying the areal will be I

appreciably less than the dose rate in contact with the shield. This factor is taken into consideration in Section 4.0 in assessing the adequacy of existing shielding.

3.7 Shielding Calculation Description Thisprogram(SHD)isdesignedtocalculatedosagefromalinesource.

For a line source with shielding and assumed buildup factor of the form I

B (u r) = A e-** + (1-A) e-* T (Ref. 1)

(.3-1 )

3-6 1

(

QUAD-1-80-040 I

O J

N 4

/

g S

2 e

t z

9 I

The flux at point P is given by 4) fA F(e, (1 + a) pt + F(0, (1 + a) ut)

(3-2) 0 j

2

+(1.A F(e, (1 + 8) ut) + F(e ' II + 8) FT j

2 where (e

-Xsece F(e,X)

=

e de (3-3) 0 W

p = linear attenuation coefficient of shielding material B

3-7

QUAD-1-80-040 Values o' %, a, and a were Values of p were obtained from Table 3.8.

t obtained from Table 9.1.12-117 in Reference 1.

Subroutine INTERP then performed linear interpolation to calculate these values at the specific O

energy desired.

i Flux was calculated from equation 3-2 for each energy group. The flux, 2

phatot/cm -see was then onverted to equivalent dose rate in mR/hr. Dose conversionfactor,(MeV/cm -sec)/(mR/hr), was calculated with the following assumptions:

1.

DCF = 500 for photon energy between 0.3 and 1 Mey.

2.

DCF is a linear function between 1 and 8 MeV Men plotted on log-log paper.

(At 1 MeV, DCF = 500; at 8 MeV, DCF = 1000)

Therefore, for photon energy E between 1 and 8 MeV n

I log (DCF)= log (500)+ log (E)

(3-4) n 1g

-1g O

orlettingX= log (500)+ log (E)n og (

(3-5) l i

then DCF = 10*

(3-6) l l

l Then dose rate from group n is x E /DCF (3-7)

Dn"in n

n Dose rate from group n, m.ar where D

=

2 Flux of group n, photons /cm -sec.

4

=

Photon energy of group n, MeV/ photon E

=

n DCF Dose convenion facW for gmup n (Ref. 2)

=

n 2

(MeV/cm -sec)/(mR/hr) 3-8 g

1 CUAD-1-80-040 1

The total dose rate is then the sum of dose rates from all n groups.

Total Dose Rate =

Dn O

n (3-8)

I The dose rate for each energy group was first calculated and then sumed for a total dose for all 12 energy groups.

Program SHD was thoroughly checked against hand calculations. Six test I

runs were made on Program SHD, and the results agreed well with values given in Figure 12, Reference 3.

I

?

I 3

i 4

I a

i 3.g

1 i

i A

'.i -

Q" l! 21 9

\\l3 I I

l M

4-J

% 1 'r J

[.j

.g 7---

'a,,,c,,t,6 m

/

./.,,,.

ElE*U is. fit '

oc eest >=6 na a es gt,W

(

,unn a so-6stse n.. u.e i

O

..s....

.e6so.se g

v s a v is a 6_s a,c tio n f

x f

6 S

convee6samenmeane V

%g g

ie e

.e

. V

//

./

e

.....e6 t

lY ll'Yd5'"llW l

e nl+=== =4=

=

', i i I II r.e T.",e ll h

i i

_I elslslslelslaje 1.y E g

4

! ! / Bi!55-y s

g i 6 6 ? I I g

.. p.g

.....a T.t68T

[ 3.*., % C l vn.6 we case "m~

m""a5

-i Z

a-e -

l.

wo n.~

.a n,,a u,

" I ***

• "l,10.

A's, 3-.-

l Sam

.a m me.

g g

oc esse put-4 b

k g

N.

=

• W PLAN EL. 735* S*,

-esmatte mamationmonnowerte to tasa mesecast et e-e 1

.Bak..F..,

FIGURE 3-1 CONTROL ROOM

.1 4

1 CH-E O

-Y-@

- CH-I-1A N

CH-I-1B ELEVATOR

/

g O

CH-I-2 A

f.-

CH-I-3A CH-I-3B iA 1[g lol CH-FL-2

' ' e',

lol CH-R-3 CH-FL-4A S@Wbhk log g

L.

CH-FL-4B y

,<f l

^

STAIRWAY POST-ACCIDENT CCEMA l SAMPLING STATION I

F

[

[

~

~

CC-E-1B l

0 l_ _... __ _ CC-E-1C

,,_ j s

f f o','

e's

~P

i DEGASIFIER s

c,.,

L,' '

>--SHIELDED AREA

,s

_\\

,e

/

'9,d,

'5 6

i PLAN EL. 735'-6" IN AUXILIARY BUILDING NOTE: Shaded areas with dash lines indicate potentially radioactive areas.

FIGURE 3-2 POST-ACCIDENT SAMPLING STATION 1

w 1

b 1

II J

10 RECOMe1NER ROOM i

l O

CONTAINMENT RECIRC. SPRAY O

L W HEAD SAFETY INJECTION PUMP STAIRS LEVEL dt 747'-0" i, p

5 i

~

I FIGURE 3-3 PATHWAY T0 RECOMBINER ROOM IN ACCIDENT CONDITION i

1 ie ii

TABLE 3-1 RADIATION S0tlRCE TERM UNDER POST-LOCA C0 DfTI'ONS CONTAINKNT AIR O

Noble gases 100% of core inventory Halogens 25% of core inventory LIQUID SYSTEMS Noble gases 100% of core inventory i

Halogens 50% of core inventory Others 1% of core inventory TABLE 3-2 REACTOR PARAMETERS Power 2766lW Burnup 1797888. N D Time 650 days O

I3 2

Flux 4.80 x 10 N/Cm - Sec Decay Periods: initial,1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />,10 hours, 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />,120 hours, and 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> i

J ki 1

.y

. ~

~

7

'z.'

TABLE 3-3 LI D SOURCE

)

,o

• • • • • • .*** canna source **********

• O PRINCIPAL PHOTON $00RCES IN CROUP.je_M. EV/WAf f-5EC

,0 MEAN ENERGf =

.300PEV INIIIAL

1. H

2. H S. H

10. H

24. H 820. H T20. H

'P 70181 4 22E*08 4.4SE*08 4.64E*08 4.97E*08 S.CSE

• 08 4.2SE*08 2.16E*08

1. 7 3E *0 F l#Et/WalI-5ECB 101AL 3 89E*IS 4.llE+IS 4.28E*18 4 19E*IO 4.66E*18 3.92E*18 1.99 Eel 8_,)_.59(fjf

,P (PHOTON!/5tCI

~

PR7NCTPIE 7H51UN 15uitEl~1N-'tlRtiup-E. MEV/ watt-itC

'e

~~

MEAN ENERGY =

.630PEV INillAL

). H

2. H

5. h_,IO. H _

24. H, 120. H _,,F20. H TOIAL 4.CTE*01 3.S2E*09 2.87E*09 1 98E*01 8.71E*09 1.40E*09 S.0lE*08 2.59E*0F (PEv/ Wall =$fC),

i TOIAL 1.79E*19 1.54t*19 1.26E*19

8. ),CE + 18 7.53 Eel 8,,. _,,.6 13E*18,,,,P.20E*18._ 1.14E *1 f EPHot0N5/5ECB l

l

_ PRINCI PAL, PHCTON, $CURCES,8 N,, CROUP..,3e, MEW /W A TT-SEC _

~

l MEAN ENERGY = 1.100MEV

' ~ liittIAL

1. H

2. H S. H

10. H

24. H 120. It 720. H infat 4.FFE*09 1.56E+C9 8.34t*09 9.40E*08 6.42E*08 3.86E*08 9.56E*0F 4.6tE*05 (PEV/hAII-5EC8 10 At 4.4SE*l8 3.92E*S

3. ire *IS 2 36E*ls I.6 E*18,_7.93E*17 2.40 Eel F_J. lfELS (PrCIDst/sECI i

PKirictME piettbr ScuRCis-IN CROUW 4* MEV/wArr-stC~ -

FEA*I E NE RGY = 1.SSOPEV

(

INIIIAL

1. H

2. H

5. H

10. H

24. H 120. H 720. H TOTAL i.11t*08 6.8M *64 S.S8E

• b8-~~ M 4E*08 7.68E+08 'l.73E*08 7.52E*0f' 6.SIE*06

~

IPEV/ haft-SECS 10rAL 1.4sE*ie 1.23Eil6 % 4tilf-L50E*M 4.r9E+1r 3 08Eear 1.34E+1r "I'.1tE*l4 tr>0 toss /sECl PRINCIPAL PHCf0N SOURCES IN CROUP

$* MEV/ WAIT-SEC dihM 'i NE R 4Y' 'e~i.490PEV

' 7 EITIIL

1. H

2. H S. H

10. H '

24. H 120." H'~~~720. H 101AL 4.87E*09 4.10E*08 3.48E*08 2.2 8 t +0 8 1.22E+08 4.71E*07 1.38E*0F 2.78E*05 (PEv/WAII-5ECS T01A1 6.7FE+1T S.70E*17 4.83E*l7 3.C 8E

• I F

'"1.69E*17~ ~ '6.SSE*16 1 92E*16 IPHCION$/5ECl

~~~

~~~3.7FE*14 PRIIeCl FAC #iat t ch ICuR C E S ' ! N ' CROUP ~4E"MEV/ WAIT-SEC -

s REAN ENERGY 2 380PEV

=

I

• s '

INiilAL

1. H

2. H

5. H

10. H

24. H 120. H 720. H TOTAL 6.0TE*08 4.Tbt*BI%3E*68 1.7tE *05 ~S.03E* 0 F " 2.9)E *06 1.20E 46 TJ.15E*05 (met /wAtI-SECl v

10lAt

f. CS E

• 11 ~BR

• l)~C22 E

• 17 ' !.99E

• 17 5.85E*lt 3.40E*l5 1.39E*lS "3.66t+14

~~

s I PHCID A S /SE C )

()

~ ~ - ~ ~

PRINCIPAL PHCION SOURCES IN CROUP 7* MEV/WAff-5EC

~~

~

~ 6 hEAN INERCV o' 2.7SOMEV

TABLE 3-3 LIQUID OURCE(Continu:d)

~

i

_a m __ _..

m

! Mo

- __. o !

l g

INiflAL

1. H

2. H S. h

10. H

24. H 320. H F20. H

.{

101AL 4 24E*08_ (s48[*0g_ l.44E*06 2.79E*07,J.84E*06 2.80E*03 S.96E-04.,0.

- O t=Ev/W4TI-5EC)

,, e

• ;/

total ___4,{FE+1T 2.50E*l7 1.45E*lF 2.8tE*14 1.05E*tS 2.!!E*t2 E.00E*01 0.

V S PHClu A5/5E C 8 o

~

PRlW,lPAL PHO EN7 6 ItN 5~lii 6ROUP~ h~HEV/ WAIT-SEC'

'e N

Mf.A. N_,,1 AE R GY = 3 450MEV 4

i 40

,O INITIAL

1. H

2. H

9. M

10. H

24. H tJ0. H 720. H 10l&L 4.60E*07 F.30E*06 2 2AE*06 4.60E*05 4.00E*04 8.566*01 1 22E-!!

0.

(MEv/hAtf-5ECl

'O I

10tAL 3.92E+16 6 21E *[$

1 90Eiil F.'92E*14' ~4700 Eel 3' F.29E*le 1 04E-00 ' ~ 0.

I i

(PHOffN5/5EC) f

,O 1

PRINCIPAL PHCTON SOURCES IN CROUP

9. MEv/W4TI-5EC i

3.800MEV gb MEAN ENERGY

=

O

.s. H

10. H

24. H.....120. H F20. H

. _ _.liAL

1. H

2. h

..O INi I

TDTAL 3 2TEt0F 9.70g*06 2 6tE*06 S.16E*04 7.46E*01 8 34E-0T C.

O.

O eMevisatt-SEC O

10 At 2 4*E*le 7.25E*ts 1.95E*is 3 05E *ti s.stE*to 4.23E*02 c.

O.

tPoCIDAS/SEC4 lO PiTEllPAL PHOION~500ECEs'IN' CROUP'l07 HEV/ haft-5EC"'~~~

O g

la ME74 ENER0Y = 4.220PEV.

3 0 C

INIflAL

1. H

2. H S. H

10. H

24. H 120. H 720. H IOTAL 1,15E*03 0.36E*00 4.9 4 E- 06 1.02t-24 0.

O.

O.

O.

O t*Evihart-5EC l

10fAL 1.5 3 E ill $.43E*09 3.24E*03

4. I t tGi f 67"~ ~ ~~'0 7 '

'C.'

O.

O 1

I P HO T ON 5 / 5'.C l PRINCIPAL PHolt1N SCURCES IN GADUP lie MEV/WAff-5EC i

ME AN E AERGY = 4.r00HEV i

O O

' ~tWif th I. H

2. H S.-~ i4~ 1DT H -

24. H 12 0. I' f 26. N I

total 9.!FE*03 4.F4E-01 2.45E-03 3.40E-10~~1.26E-21~ 0. '

O.

~ ~ ~ ' '.

~ O

}

O I4E vtisaf f-5ECf

~

~

O.

10141 5.40E*to 2 79E*08 1.44E*06 2.00E-Ot 7.4tE-13 0.

O.

O l

IPHoloAS/5tC8 1

O

~

i PKITe(IML'FitCl6PI 51iURCEI IN ' GROUP 12e HEV/ Weit-5EC ~ ~~~

g i

at M E_AN_E N_E R GY =

5._25.CPEV.-

IklilAL

1. H

2. H S. H

10. H

24. H 120. H 720. H

~~~10TAL 2.0tE*06 4 4FE*00

8. 6 F E-O i 1.80E-2 5 O.

O.

O.

O.

f l

1 IPEV/dall-5ECS O '

105AL 1.06E*1 M '72D M 4.57 66 N.4tfM17

'0.

~~O.'~~~'"~'O..

O.

j I PHCf 0N5/5EC. S y

i i

g l

t

- - - ~ -

- ~ ~ ~ ~ ---

}

~

~

TABLE 3-4 N SOURCE 0 - -.. -..-

eO..

• O O

...*..... 4APRA,0tRm.......+..

i O

PU NCIPAl,_Ptf fCN SOURCES Ih CRCUP I d Evf,hAff-5tC

,0 f

HEAN kNERGV e

.3J0Pif V

• O I Ni ll At

". H

2. H

5. H

0. H 7 4. H If0. H 730. H

'e TCrat 3.21r*J3 3.?2E*03 3.70t*00

%. C. E

• 01 e.14E*0s 3.let*Ce I.PE*05 1.C3F*0' O

PglNCIPAt P,ioren suuRCCS t h ceCUP 2e MEW /hATI-set

• O

\\

.Nin-iKthai.

. 30*tEv bO tr.1] l Au

1. H

2. Is

2. H 3C.H v.H " 12a: W '-' 'hT. H

• C T tat 1.8?f*0?

1.73 E *J' l.3JE*07 f.'r E*01 P.ltt*C3 S.t (*08

2. 30F 6 08 1.C?t*0S 9O _. _,,

PRif4C,lPAL PHulura 5'JJRCE S t h CRCUP 3 d LV/t.All-$f(

,p

[

NE A:4 Lt.lRhy. tilJO SEW t

e IO ~ ~ ~ ~

~lilITat

1. H

2. Il

7. H IC. H 24.*H'**'120. ti ~ ~72d. H

(

f*lat C.11F*CT 7.t'E*03 t.?ft*J*

.t3E*0R 3.16E*08 1.'.*E*C3 8.

7tE*C7 2.28t

• 0 5

~" ~ ~ ~ -

Ptif4CIPAL Ptfull1N $UURCES IN GROUP 4* MED/ hall-SEC ~~

C

• O UP.i U.[i;y -

1. 10 It v ~ ~ ~ ~ ~ '

' ')

~

~ ll.'II I AL

1. H

2. H

3. H 10~ R 2 '. H'~~ ~ 12 0.~ H ~~ 72 0.~ H C

l* fat 3.7)E*C1 3.2tE*00 2.*Cl*00 1.CtE*0?

1.88Ee09

7. SL*C7 0.5'E*d7

.!.7"E*05 1

xt&TUhr. 176 bytv

~ ~ - ~ ~ ' ~ ~ ~ ~

C

?

PRlr4CIPAL PHCluta $UURCES IN GROUP fe MED/kAll-SEC O' ' ~ ' -

~~1 Ell l A*.

a. H

2. H

. H

10. H 2 r. it -- g 2 0.172 0. H O

I J i&'.

3. ( 'd

• C 3 3.0l E

• 05 2..CE*03

. 4" E *0 3-

'.01E*d7

?.?lt*07 6.49E*06 9.!4E*04 i

I 5

- - - - ~ ~ - - -

ar.WEKERM - n3snty-- - '--- - -

O

. Q P4lNCIPAL PHOIUta 50URtt$ IN GktrJP to ME v/W A II-SL C I

-~~~ Kill ~At

- I. e-

2. H f.' H '-~~ In." li'*

2 4.. H 120. H~~~720. Il O

1

• )

10 fat

!.7' E *C3 4.46t*0C

?. 4 Jt

• 00. 6tF401 4.9'E*05

?.SJkout 7. l ' E-0*

0.

Pee l*JCI 5 AL Ptt'l.73 $'.'JEC ES (f. CR"UP 7, '4E t /h A ll-SE C O -l O.,

~ ~ ~ ~

n Irn u ti M 1 2 :s u ts e

i j

O" "

~[Tj[JI AL

1. H

• .4?t*03

2. 7 4t

• 0

l.73F*06 8.30f*02 c.

O.

2. Te Wii--'-

lo.~H 2 ; H ~126."h ~'~~72 0. H O,

4 IP.f AL A.*1F*C3

• 45t*0C O,

Pfl:4CI F At FITT *.f3 $'sJP.tts I's CR30P 8e '9E b/ hall-5EC O

at44 (6t RGW-5'$0st v '---- ~~- * -- ~ -

Oll 4

i O'

' il l' IIIL

3. H

~~l.~ ll

.H

30. H 23. 'er -'.120."'M ~79 0.~~ H O

O.

tefat O.

C.

C.

O.

O.

O.

]

g O*

PR l**CI F AL PH t'.84 soORCLS l'4 C%Mir 9 Mtv/ watt-stC e

• ~ ~ ~

t.

~~

MEAV tt.tRGV = GiddMC D -

C' IellIA.

1;'t.--

2rH

s. H

• 0. H "4.

D.

• td. H 7*0. H ~~~~~~~

~ ~ ~ ~ ~ ~ -

~ 9 it.f At 1.748:*3' 4.7-L*J.

l.Jil

• 06

2. *

• E

• C.

3.*iE*01

• .llt-C' d.

O.

PRINCIPAL. PittilJ. StupCES lb.GRCUP Ice PEV/hATI-5EC O

4O'

- - - - - ~ ~ -

MrAps thLtit,F

s. 22JPLy -

" ~ ' ' " " *

• ~~~~~ ~ ~~"~~

l f gs

~

q

.C, G.

.G. fe.

2) o.

e o

e a

e e

a m

-n n

a a

=

s s.

e O

4 be I

e E

K E

e e

e O

O O

p%

g em sw Y

p.

+=

e e

o O

O u

3E E

E ed o

e C

O O

Q M

se4 L>

w e

w e

'*s*

O esa O

ees O

en

+

0 0

E a.

3

=.

Z

,e e-s e

e e

se

=

e N

• w N

M w

et, e

O a

O, a

- O m

z i

CD e

e le i

g O

O a

O

)

e c>

e e

.e e

O o

O o

O LaJ d

.d.

l" M

1 Z

e=

2 lZe' M

e e

e e

H em twD n

wD M s

Y$

Y$h.

= a w a O

e"*

.)3 O

w"..

O r

wesw e

e E

7 B%

E c

h e

o e m

g e

=

e E n L >

J O

e O

.0K O

mot O

4ans er us u 2 Z

ed I

mW e"

0 g

E

)

u u.-.

.z-4 ee e.

~

e

~

e

.f

.e a

e a

se a s a m i

EM E E e

o e

a O

O O

j I

[

I d.

W O

es em

.e l

M E

i e

• . e me e me e

• i l

O p

O i

I

=0

=O

.O 0

1 4

ei M

4 l

C[

a=

8 e,

CJ

• si g

e=

so W

?

4 L

t 4

y g.

O O

M 4

9 0

tl 6

4 6

6 4

1 0

== e a

TABLE 3-5.1 CONTAINHENT PENETRATION CllECKLIST l

(Outside Containment) l PENET.

ISOLATION VALVES ISOLATION VALVE ISOLATION VALVE l

HO.

SERVICE PROVIDED IDENTIFICATION NUMBER POSITIONS INIT.

i AREA SilUT ESF NORHAL FAIL.

g ACTION 1-D CCR to RilS II/X 1A & PJIS Hanual ICCR-247 Shut Open AS-IS Hone l

Pump 1A Seal Cooler j

2-D CCR From RIIS II/X IB & RllS Hanual 1CCR-252,

Shut Open AS-IS None l

Pump IB Seal Cooler 3-Spare l

4-D CCR From RilS II/X 1A & RilS Hanual 1CCR-251 Shut Open AS-IS None Pump 1A Seal Cooler l

5-D CCR to RllS II/X IB & RHS Hanual ICCR-248 Shut Open AS-IS None i

Pump 1B Seal Cooler l

62B Spare l

7-A liigh llead S.I. To llot Rem-Han HOV-lSI-869A Shut Shut AS-IS Hone Legs l

8-C CCR From RCP B & C Auto-Trip TV-lCC-107D2 Open Open Shut S-CIB Thermal Barriers 9-B CCR From Shroud Coolers.

Auto-Trip TV-1CC-lllD2 Open Open Shut S-CIB

}

10-B Spare l

11-B Air Recirc. Cooling Auto-Trip TV-1CC-110F2 Open Open Shut S-CIB Water Out Auto-Trip TV-lCC-110F1 Shut Shut Shut S-CIB l

12-A Spare l

13-D Spare l

14-D Air Rectre Cooling Auto-Trip TV-1CC-110E2 Open Open Shut S-CIB Water IN 15-A Coolant System Charging Auto-Trip HOV-1CH-289 Open Shut AS-IS S-SIS I

i 16-B CCR to Shroud Coolers Auto-Trip TV-lCC-lllAl Open Shut Shut S-CIB l

17-A CCR to RCP IB Auto-Trip TV-lCC-103B Open Open Shut S-CIB l

18-A CCR to RCP IC Auto-Prip TV-lCC-103C Open Open Shut S-CIB 19-A RCP's Seal Water Return Auto-12ip HOV-lCll-381 Open Open/ AS-IS S-CIB Shut 20-C S.I. Accum. Hakeup Hanual ISI-41

, Shut Shut AS-IS Hone 4

d I

a

b TABLE 3-5.2 CONTAINHENT PENETRATION CllECKLIST (Outside Containment) 1 1

PENET.

ISOLATION VALVES ISOLATION VALVE ISOLATION VALVE NO.

SERVICE PROVIDED IDENTIFICATION NUMBER POSITIONS INIT.

SHUT l

' AREA FAIL.

DOWN ACTION.

t 21-B Spare l22-B Spare 4

l

'23-B Spare 1Ril-15 Shut open/ AS-IS None

?la-hgD RilS to RWST Manual Shut l

25-B CCR From RCP 1B & lc Auto-Trip TV-lCC-105D2 Open Open Shut -

S-CIB Motors j

26-C CCR From RCP 1A Thermal Auto-Trip TV-lCC-107E2 Open Open/ Shut S-CIB Shut l

Barriers i

27-C CCR From RCP 1A Hotor Auto-Trip TV-lCC-105E2 Open Open Shut S-CIB l

28*-A RCS Letdown Auto-Trip TV-1CH-204 Open/

Shut Shut S-CIA Shut 29-A Pri. Drains Trans. Pump Auto-Trip TV-lDC-1085 Open/

Shut Shut S-CIA 4

Shut l

No. 1 Discharge 1

30-8 Spare l

31-D Spare l

32-C Spare Rem-Man HOV-ISI-8698 Shut Shut AS-IS None

{

33-C liigh Ilead S.I. to llot l

Legs 34-A Spare I

35-A Seal Inj. Water RCP 1A Rem-Man HOV-1Cil-308A Open Open AS-IS S-CIB HOV 1Cil-3088 Open Open AS-IS S-CIB l

36-A -

Seal Inj. Water RCP IB Rem-Man i

37-A Seal Inj. Water RCP 1C Rem-Man HOV-1Cll-308C Open Open AS.IS S-CIB

' CIA 38-A Cnmt Sump Pump Discharge Auto-Trip TV-lDA-100B Open Shut Shut l

39*-C Stm Cen lA Blowdown Auto-Trip TV-1BD-100A Open Shut Shut S-CIA, 40*-A Stm Cen IB Blowdown Auto-Trip TV-1BD-1005 Open Shut Shut S-CIA 41*-B Stm Cen 1C Blo'wdown Auto-Trip TV-1BD-100C Open Shut Shut 3-CIA 42-C Compressed Air to Fuel Hanual 1SA-14,

Shut Open AS-IS None llandling Equipment 43-D Air Activity Honitor -

Auto-Trip TV-lCV-102-1 Open Shut Shut S-CIA, Out

j TABLE 3-5.3 I

i CONTAINHENT PENETRATION CllECKLIST (Outside Containment) i PENET.

ISOLATION VALVES ISOLATION VALVE ISOLATION VALVE NO.

SERVICE

~

PROVIDED IDENTIFICATION NUMBER POSITIONS INIT.

AREA i

SilUT ESF NOM y

DOWN ACTION j

44-B Air Activity Monitor - IN Series TV-1CV-10IA Open Open Shut S-CIA Auto-Trips TV-1CV-101B Open Open Shut S-CIA

]

45-B Pri. Grade Water to Par.

Auto-Trip TV-1RC-519 Open Open Shut S-CIA

{

Relief Tank 1

46-A

Charging. Fill llender.

Rem-Man FCV-1CH-160 Shut Shut Shut None i

47-B Instrument Air Manual IIA-90 Shut Shut AS-IS None

~

l 48-B Primary Vent lleader Auto-Trip TV-101 109Al Open Shut Shut S-CIA

]

49-C Nitrogen Supply to Pzr.

Auto-Trip TV-lRC-101 Open Shut Snut S-CIA Relief Tank a

50-C Spare 51*-C Spare 4

1 52*-C Spare l

53-C Nitrogen Supply to S.I..

Auto-Trip TV-1SI-101-1 Shut Shut Shut S-CIA Accumulators 54-B Spare l

55*-1-A S.I. Acuum. Sample Auto-Trip TV-:1SS-109A2 Open Open Shut S-CIA 55*-2-A CNHT Leakage Honitoring Series TV-11M-100A1 bpen.

Open Shut S-CIA Open Taps 55*-3-A Spare 55*-4-A Pzr. Re1lef Tank Gas Auto-Trip TV-1SS-111A2 Open Open Shut S-CIA Sample 56*-1-A Par. Liquid Sample Auto-Trip TV-ISS-100A2 Open Open Shut S-CIA 56*-2-A RCS Cold Leg Samples Auto-Trip TV-1SS-102A2 Open Open Shut S-CIA 56*-3-A RCS Ilot Leg Samples Auto-Trip TV-1SS-105A2 Open Open Shut S-CIA TV-1SS-117A Open Open Shut S-CIA 56*-4-A Stm Cen IA Blowdown Auto-Trip Sample 57*-1A Cnmt Leakage Monitoring Series TV-Illi-100A1 Op e.-

Open Shut S-CIA Open Open Shut S-CIA Open Taps Auto-Trip TV-1LM-100A2 57*-2-A CNHT Leakage Honitoring Open Tapa

,s

~~M MM M"M M

TABLE 3-5.4 CONTAINHENT PENETRATION C11ECKLIST_

(Outside Containment)

ISOLATION VALVE ISOLATION VALVES ISOLATION VALVE POSITIGUS INIT.

IDENTIFICATION NUMBER F

PENET.

PROVIDED 8"U SERVICE NORMAL DOWN_

ACTION _

FAIL.

NO.

AREA' Shut Shut AS-IS None ICV "J5 Hanual Shut Shut.

None 57*-3-A 57**4-A CNHT Leakage Honitoring' ICV-36 TV-1CC-103A Open Open/

Shut S-CIB Hanual System - Pressurized Auto-Tri,p Shut 55-d CCR to RCP 1A HOV-1SI-890A Shut Shut AS-IS None 59-C Spare Rem-Man Low Ilead S.I. to Hot HOV-1SI-890C Open Shut AS-IS None 60-SgD Lens Rem-Han Low Head,S.I. to Cold Sliut Shut AS-IS Hone 64-SgD HOV-1SI-890B Legs Rem-Han Low JIead S.I. to Hot HOV-1QS-1018 Shut Shut AS-IS 0-CIB 62-SgD Legs Parallel 63-SgD Quench Spray Pump Disch HOV-1QS-101A Shut Shut AS-IS 0-CIB 360* Ileader Parallel Quench Spray Pump Disch Shut Open AS-IS None 64-SgD 360* I!cader LATER Fuel Transfer Tube HOV-1RS-155A Open -

Open AS-IS 0-CI,5 Manual 65 Rem-Han 66-SgD Outside Recirc Spray Pump 2A Suct. From CNMT HOV-1RS-155B open Open AS-IS 0-CIB Rem-Man Outside Recire Spary Pump 2B Suct. From CNHT HOV-ISI-860A Shut Shut AS-IS None 67-SgD Rem-Man Low Head S.I. Pump 1A Suction From CNHT Sump HOV-1SI-8608 Shut Shut AS-IS Hone 68-SgD Rem-Han Low itead S.I. Pump IB Suction From CNHT Sump HOV-1RS-156B open open AS-IS 0-CIB 69-SgD Auto-Trip 70-SgD Outside Recire Spray LATER LATER LATER LATER Pump 2B Discharge.

LATER LATER 72-SgD Argon Supply (RP-14A) l g

.1

,r--------------

M i

m W

\\

4 TABLE 3-5.5 CONTAINHENT PENETRATION CHECKLIST i

(Dutside Containment) j l

PENET.

ISOLATION VALVES ISOLATION VALVE ISOLATION VALVE j

NO.

SERVICE

• PROVIDED IDENTIFICATION NUMBER POSITIONS INIT.

U SF AREA j

NORHAL FAIL.

DOWN ACTION 1

i 73*-SgD' Hain Steam Loop 1A Auto-Trip TV-1HS-101A Open Shut Shut S-SLI Hain Steam Line Drain Auto-Trip TV-lHS-111A.

Open

.open

. Shut S-SLI Main Steam Atmos. Dump PCV PCV-1HS-101A Shut Open Shut None

, Main Steam Safety Valves Safety Safety.

Shut Shut Shut None i

Valves Valves j

I Main Steam to Aux.

Rem-Han HOV-1HS-105 Open Shut AS-IS None i

2 Feed Pump 74*-SgD, Main Steam Loop 1B Auto-Trip TV-1HS-1018 Open Shut Shut S-SLI 4

Open Shut S-SLI Main Steam Line Drain Auto-Trip TV-1HS-IllB Open S!ut Open Shut None Hain Steam Atmos. Dump PCV PCV-1HS-101B l

Main Stm Safety Valves ;

Safety Safety Shut Shut Shut-None i

Valves Valves i

Hain Steam to Aux.

Rem-Han HOV-1HS-105 Open Open AS-IS None i

Feed Pump

)

75*-SgD Hain Steam Loop IC Auto-Trip.

TV-lHS-101C Open Shut Shut S-SLI i

Hain Steam Line Drain Auto-Trip TV-1HS-111B Open Open' Shut S-SLI Hain Steam Atmos. Dump PCV PCV-1HS-101B Shut Open Shut None Main Stm Safety Valves Safety Safety Shut

' Shut Shut None l

Valves Valves Hain Steam to Aux.

Rem-Han 110V-1HS-105 Open Open AS-IS None Feed Pump 76*-SgD Feedwater Loop 1A Non-Return HOV-1FW-156A Open Shut Shut S-FW1 Aux Feedwater Loop 1A Rem-Han HOV-lFW-158A I

Open Open AS-IS None 77*-SgD Feedwater Loop 1B Non-Return HOV-1FW-156B Open Shut Shut S-FWI Aux Feedwater.Lo'op 1B Rem-Han HOV-1FW-158B Ope's Open AS IS None 78*-SgD Feedwater Loop IC Non-Return HOV-1FW2156C Open Shut Shut S-FWI Aux Feedwater Loop 1C Rem-Han

'HOV-1FW-158C Open Open AS-IS None 79-SgD RW to 1A Recire.

Rem-Man HOV-1RW-104A Open Open AS-IS None Spray Heat Exch.

80-SgD RW to IC Recire.

Rem-Han HOV-1RW-104C Open Open AS-IS None I

Soray Heat Exch.

W~~'~~~'

6 1

h W V M

TABLE 3-5.6 CONTAINHENT PENETRATION CllECKLIST (Outside Containment)

PENET.

ISOLATION VALVES ISOLATION VALVE ISOLATION VALVE NO.

SERVICE PROVIDED IDENTIFICATION NUMBER POSITIONS.

INIT.

AREA NORMAL FAIL.

DOW:1 ACTION 81-SgD RW to IB Recire.

Rem-Han H0V-1RW-104B Open Open AS-IS None Spray lient Exch.

82-SgD RW to ID Recire.

Rem-Han HOV-1RW-104D, Open Open AS-IS None Spray lleat Exch.

H0,-1RW-105A Open-Open AS-IS None V

83-SgD RU from 1A Recire.

Rem-Han Spray IIcat Exch.

HOV-1RW-105C Open Open AS-IS None 84-SgD RW from IC Recire.

Rem-Han Spray llcat Exch.

85-SgD RW from IB Recire.

Rem-Han HOV-1RW-105B Open Open AS-IS None Spray llent Exch.

86-SgD RW from ID Recire.

Rem-Man H0V-1RW-105D Open Open AS-IS None Spray lleat Exch.

87-Sgp Post DBA Ilydrogen Manual illy-110 Shut Shut AS-IS None Control 88-SgD Discharge to CNHT Manual Illy-111 Shut Shut AS-IS None 89-SgD Hain Condenser Auto-Trip TV-ISV-100A Shut Shut Shut S-CIB Ejector Vent 90-SgD CNHT Purge Exhaust Auto-Trip "VS-D-5-3A Shut Open AS-IS S-RH 91-SgD CilHT Purge Supply Auto-Trip VS-D-5-5A Shur Open AS-IS S-RH

~~

TV-ICV-ISOC Open Open Shut S-CIA 92-A CNtfr Vacuum Pump IB &

Series 112 Recomb. Suction Auto-Trips TV-1CV-150D Oren Shut Shut S-CIA Hanual 111Y-102 Sliut Shut AS-IS None 93-B CHHT Vac'uum Pump 1A &

Series TV-1CV-150A Open Ope..

Shut S-CIA lig Recomb. Suction Auto-Trips TV-ICV-150B Open Shut Shut S-CIA Hanual 1HY-101 Sliut Shut AS-IS None 94-C CNHT Vacuum Hanual llCV-1CV-151-1 Shut Shut AS-IS None Ejector Suction 95-C Spare 96-B liigh llead S.I. to Rem-Han HOV-1SI-836 Shut Shut AS-IS None Cold Lega

f%

7 g

g~

U-g

-' g ~

TABLE 3-5.7 CONTAINMENT PENETRATION CilECKLIST (Outside Containment)

PENET.

ISOLATION VALVES ISOLATION VALVE ISOLATION VALVE NO.

SERVICE PROVIDED IDENTIFICATION NUMBER POSITIONS THIT.

SilUT ESF AREA NORME pg DOWN ACTION 97*-1-A RIIS Inlet Sample

. Auto-Trip TV-1SS-104A2..

Open Open Shut S-CIA 97*-2-A RilS Outlet Sample Auto-Trip TV-1SS-103A2 Open Open Shut S-CIA 97*-3-A CNMT Leakage Series TV-1LM-100A1 Open Open Shut S-CIA Honitoring Open Taps Au'to-Trips TV-11R-100A2 '

Open Open Shut S-CIA 97*-4-A Stm Cen,1C Blowdown

• Auto-Trip TV-ISS-117C Open Open Shut S-CIA Sample Capped.

N/A N/A N/A N/A 98-1-C 0xygen Capped 98-2-C Argon Capped Capped N/A N/A N/A N/A 98-3-C Acetylene Capped Capped N/A N/A N/A N/A 89-4-C Spare 99-C Spare 100-B Spare 101-B Spare 102-B Spare Shut Open AS-IS None 103-A Refueling Cavity Purif Manual IPC-37 Outlet 104-A Refueling Cavity Purif Hanual IPC-10 Shut Open AS-IS None Inlet 105*-1-B Stm Con IB Blowdown Auto-Trip TV-ISS-117B Open Open Shut S-CIA, Sample 105*-2-B Pzr Vapor Sample Auto-Trip TV-1SS-112A2 Open Open Shut S-CIA 105*-3-B Spare I

105*-4-0 Spare 8

Shut Shut' Shut.

S-CIA l

106-SgDl S.I. Accum. Test Line Auto-Trip' TV-1SI-889 Spare

'~C B !

Spare I

109-C Spare 110-1-C' Press Dead Weight Series 1RC-277 Shut Shut AS-IS-None Calibrator PT-RC-455A Manual IRC-278 Shut Shut AS-IS None 110-2-C Spare i

110-3-Ci Spare l

T I

N I

NO SS e

HHII e I

T n

RRSS n o

- - - - o C

N O000N A

E S

SSSSS V

L I

IIIII LS I

AN A

S SSSSS VO F

A AAAAA I

NT OI N

n t t t t t W

e uuuuu

"" O I S TO p

hhhhh AP D

O SSSSS LO S

L I

AM t

t t t tt R

u uuuuu O

h hhhhh N

S SSSSS RE B

EM VU T

LN CD S

A I

VN 77 L

O 66, K

NI A

AB88 C

OT 6

11 - -

t E

)

IA II 9_9SS1 t

TC 4

I I

- 119 C

n AI 8

e LF D

DD - - -

N m

OI

- - VV1 5 O n

ST S

SSOOS I

i IN V

VVHH1 3 T a

E A

t D

ne e

E R n

I e

L T

o B

E A

7]

T E

e.

S P

d E

i V

T s

L N

t AD s

E u

VE H

O D

p p

lili N

(

NI erer I

OV A

I O l

lTlTl a

l l - a T

TR u

aoaou N

AP O

L n

rt rt n a

auaua C

O M

PAPAM S

g I

eg d

e l

L oc g

t d

d s

l o

l u

o t

B a

C E

h s

C l

x o

s I

e E

t a

V u

p R

F g

k y

E d

n B

S dtl a

esB T

T. s r r ee et u rual gaa I.

alhe I

epp pi xu SDEF B

BLSS C

A A

AA 4C C

1 2

34 TE

.A NOE 0l 2

3 3

33 ENR 1l 1

1 1

11 P

A 1i 1

1 1

11

-r- +

a.

o

.g

,a, s

o

.,,,, s TABLE 3-6.1 CONTAINHENT PENETRATION CilECKLIST (Inside Containment)

ISOLATION VALVES ISOLATION VALVE ISOLATION VALVE PENET.

NO.

SERVICE PROVIDED IDENTIFICATION NUMBER POSITIONS INIT.

^

NORMAL FAIL.

DOWN ACTION 1-D CCR to RHS 18/X 1A &

Rem-Man HOV-1CC-112A2 Shut open AS-IS Nc,ne RilS Pump 1A Seal Cooler 2-D CCR From.111S H/X IB &

Rem-Han HOV-1CC-11283 Shut Open AS-IS Hone RilS Pump 18 Seal Cooler 3-Spare 4-D CCR Froni RHS H/X 1A &

Rem-Man HOV-1CC-112A2 Shut Open AS-IS Hone RilS Pump 1A Seal Cooler 5-D CCR,to RilS H/X 1B &

Rem-Han HOV-1CC-11252 Shut Open AS-IS None RllS Pump 1B Seal Cooler 62B Spare 7-A litgh llead S.I. to Hot Check 1S1-83 Shut Shut AS -IS None Legs 8-C CCR From RCP B & C Auto-Trip TV-1CC-107D1 Open Open/ Shut S-CIB Shut Thermal Barriers i

9-B CCR From Shroud Coolers Auto-Trip TV-1CC-111D1 Open

Oper, Shut S-CIB 10-B Spare

)

11-B Air Recirc Cooling Auto-Trip TV-1CC-110D Open Open Shut S-CIB Water Out l

12-A Spare 13-D Spare 14-D Air Rectre Cooling Auto-Trip TV-1CC-110E3 Open Open

' Shut S-CIB Il Water IN 15-A Coolant System Charging check ICII-31 i

16-B CCR to Shroud Coolers Auto-Trip TV-1CC-111A2 Open Shut Shut S-CIB 17-A CCR to RCP IB Auto-Trip TV-1CC-10381 Open Open/ Shut S-CIB Shut 18-A CCR to RCP IC Auto-Trip TV-1CC-103C1 Open Open/ Shut S-CIB Shut 19-A RCP's Seal Water Return Auto-Trip HOV-1CH-378 Open Open/ A';-IS S-CIB i

Shut

,. s TCCHF 1

h&

S h

~

TABLE 3-6.2' CONTAINHENT PENETRATION CilECKLIST (Inside Contnihment) l PENET.

ISOLATION VALVES ISOLATION VALVE ISOLATION VALVE NO.

SERVICE

• PROVIDED*

IDENTIFICATION NUMBER POSITIONS INIT.

SilUT ESF AREA FE DOWN AC; ION 20-C S.I. Accum. Hakeup Check ISI-42 Shut Shut AS-IS None 21-B Spare 22-B Spare 23-B Spare Shut Open/ AS-IS None 24-SgD Ells to RWST Hanual 1Ril-14 Shut 25-B CCR From RCP IB & IC Auto-Trip TV-1CC-105D1 Open Open/ Shut S-CIB Shut Hotors 26-C CCR From RCP 1A Thermal Auto-Trip TV-1CC-107El Span Open/ Shut S-CIB Shut Barriers 27-C CCR From RCP 1A Hotor Auto-Trip TV-1CC-105El Open Open Shut S-cab 28*-A RCS Letdown

' Auto-Trip TV-1CH-200A Open/

Shut Shut S-CIA Shut Auto-Trip TV-1Cll-2005 Shut Shut S-CIA Auto-Trip

- TV-1Cll-200C Shut Shut S-CII.

Rem-Han HOV-1CH-142 Shut Open AS-IS None 29-A Prl. Drains Trans Pump Auto-Trip TV-1DG-108A Open/

Shut Shut S-CIA Shut No. 1 Discharge 30-B Spare 31-D Spare 32-D Spare 33-C IIIgh IIcad S.I. to Check 151-84 Shut Shut AS-IS Cone llot Legs '

34-A Spare 35-A Seal Inj. Water RCP 1A Check 1C11-181 Open Open AS-IS S-CIB 36-A Seal Inj. W.ater RCP IB Check 1C11-182 '

Open Open AS-IS S-CIB 37-A Seal Inj. V.ater RCP IC Check 3 C11-183 Open Open AS-IS S-CIB 38-A Coat Sump Pump

, Auto-Tr1~p TV-1DA-100A Open/

Shut Shut S-CIA 4

TABLE 3-6.3 CONTAINMENT PENETRATION CllECKLIST (Inside Containment) i PENET.

ISOLATION VALVES ISOLATION VALVE ISOLATION VALVE NO.

SERVICE PROVIDED IDENTIFICATION NUMBER POSITIONS INIT.

AREA SIIUT ESF pg DOWN ACTION 39*-C Stm Ccn IA Blowdown System System 40*-A Stm Cen IB Blowdown System System i

41*-B Sta Cen 1C Blowdown System System j

42-C Compressed Air to Check ISA-15 Shut Open AS-IS None Fuel llandling Equipment, d

43-B Air Activity Honitor -

Auto-Trip TV'1CV-102 Open Shut Shut S-CIA Out 44-B Air Activity Honitor -

None None IN 45-B Pri. Grade Water to Check 1RC-72 Open Open Shut S-CIA Pzr. Relief Tank 46-A Charging Fill llender Check 1C11-170 Shut Shut Shut None j

47-B Instrument Air Check and 11A-91 and Sliut Shut AS-IS None Manual IIA-91-1 Open Open AS-IS Hone j

48-B Primary Vent llender Auto-Trip TV-1DC-109A2 Open Shut Shut S-CIA j

49-C Nitrogen Supply to Chr.ck 1RC-68 Open __

Shut Shut S-CIA

}

Pzr. Relief Tank 50-C Spare 51*-C

, Spare i

52*-C iSpare i

53-C gNitrogen Supply to S.I.

I Auto-Trip TV-1SI-101-2

, Shut Shut Shut

'S-CIA l

• Accumuletors j

,54-B

' Spare l55*-1-A S.I. Accum. Sample Auto-Trip TV-ISS-109Al

0 pen Open Shut

>S-CIA 155*-2-A CNMT Leakage Monitoring

.Open Taps None

~

None s

~55*-3-A Spare g

556-4-A

Pzr. Relief Tank Cas Auto-Trip TV-ISS-111A1 Open Open Shut S-CIA Sample I

' 5 6*-1--A iPzr.LiquidSample.

Auto-T. rip TV-ISS-100A1 Open Qpen phut S-CIA

TABLE 3-6.4 CONTAINHENT PENETRATION CllECKLIST I

l (Inside Containment)

P l

l PENET.

ISOLATION VALVES ISOLATION VALVE ISOLATION VALVE l

NO.

SERVICE -

PROVIDED IDENTIFICATION HlAIBER POSITIONS INIT.

~ SilUT ESF AREA g

FE DOWN ACTION 1

56*-2-A RCS Cold Leg Samples Auto-Trip TV-1SS-102A1 ~

Open Open Shut S-CIA I

56*-3-A RCS Ilot Leg Samples Auto-Trip TV-1SS-105A1 Open Open Shut S-CIA 56*-4-A STM Cen 1A Blowdown System System l

Sample 57*-1A CNHT Leakage Monitoring, Open Taps None Noite l

57*-2-A' CNHT Leakage Honitoring 8

j Open Taps None None l

57*-3-A 57*-4-As CNHT Leakage Monitoring Auto-Trip TV-11R-101A Open l Open Shut S-CIA System - Pressurized TV-ILH-101B 1

i 1

I TV-1CC-103Al l

Open j Open[ Shut S-CIB' 58-B 3

CCR to RCP 1A Auto-Trip 1

l l

Shutg i

i 1

59-C Spare

}

g i

Check 151-13 l

Shut Shut AS-IS None j

60-SgDj Low llead S.I. to llot j

I Legs j

I j

61-SgD j Low Ilead S.I. to Cold checks 151-10,11,12

'Open Shut' AS-IS Hone

• i

~

l i

Le gs.

i

{

l 62-SgD Low ilead S.I. to llot Check I

151-14 i

I4ut j Shut; AS-IS None j

i Legs l

Shut!

AS-IS 0-CIB

.ihut j

63-SgD i Quench Spray Pump Disch Check IQS-4

~

l 360* Ileader 6

l l

l i

IQS-3 Saut Shut' AS-IS!

O-CIB.

I 64-SgDj Quench Spray Pump Disch' Check l

360* llender Flange Shut Open/ AS-IS None !

65 Fuel Transfer Tube Flange e

l Shut None

. Nonit l

l l

66-SgDl Outside REcire. Spray I

J s

Pump 2A Suct From CNHT

)

67-SgD i Outside Recire. Spray None t

.None l

I I

l l

s' e

Pump 2B Suct from Cl#1T.

)

l

~

I l

j e

i 1

I 4

l l

TABLE 3.6-5 l

CONTAINHENT PENETRATION Ci!ECKLIST i

(Inside Containment) i l

PENET.

ISOLATION VALVES ISOLATION VALVE ISOLATION VALVE PROVIDED IDENTIFICATION NUMBER POSITIONS INIT.

I NO.

SERVICE f

^

NORMAL FAIL.

j DOWN ACTION l

68-SgD Low Ilead S.I. Pump 1A Hone None i

Suction from CNHT Sump 69-SgD Low Head S.I. Pump IB None None Suction from CNHT Sump 70-SgD Outside REcire Spray Check 1RS-101 Open Open AS-IS 0-CIB Pump 2B Discharge 72-SgD Argon Supply (RP-11eA)

LATER IATER MTER IATER LATER LATER 73*-SgD Hain Stenta Loop 1A System System i

Hain Steam Line Drain System System-Main Steam Atmos. Dump System System Hain Stm Safety Valves System System

)

Hain Steam to Aux.

System System i

Feed Pump 74*-SgD Hain Steam Loop IB System System j

Hain Steam Line Drain System System i

Main Steam Atmos. Dump System System j

Hain Stm Safety Valves System System

}

Hain Steam to Aux.

System System Feed Pump 75*-SgD Hain Stearn Loop IC System System Hain Secam Line Drain System System j

Hain Steam Atmos. Dump System

• System j

Hain Stm Safety Valves System System

)

Hain Steam to Aux.

System System Feed Pump 76*-SgD Feedwater Loop 1A System System

~

j Aux Feedwater loop 1A System System 77*-SgD Feedwater Loop 1B System System i

Aux Feedwater Loop 1B System System l

i

g' g

g g

f v

d g

O 4

TABLE 3-6.6 CONTAINNENT PENETRATION CilECKLIST 4

(Iniside Containment) l PENET.

ISOLATION VALVES ISOLATION VALVE ISOLATION VALVE j

NO.

SERVICE PROVIDED IDENTIFICATION NUMBER POSITIONS INIT.

AREA SilUT ESF g

FAIL DOWN ACTION 78*-SgD Feedwater Loop IC System System i

Aux Feedwater Loop IC System System 79-SgD RW to 1A Recirc System System Spray IIcat Exch.

80-SgD RW to IC Recire System System Spray llent Exch.

81-SgD RW to IB Recirc System System Spray llent Exch.

4 I

82-SgD RW to ID Recirc System System i

Spray liest Exch.

l 83-SgD RW from 1A Recirc System System l

Spray llent Exch.

84-SgD RW from IC Recirc System System i

Spray llent Excii.

85-SgD RW from IB Recirc System System j

Spray llect Exch.

86-SgD RW from ID Recire.

System System Spray llent Exch.

87-SgD Post DBA Ilydrogen LATER

- LATER l

Control l

88-SgD Discharge to CNMT

. LATER LATER 89-SgD Hain Condenser Check 1AS-278 l

Ejector Vent j

90-SgD CNMT Purge Exhaust Auto-Trip VS-D-5-38 l

91-SgD CNMT Purge Supply Auto-Trip VS-D-5-5B l

92-A CNMT Vacuum Pump 1B &

None None 11 Recomb. Suction 2

9 3'-B CHHT Vacuum Pump 1A &

None 112 Recomb. Suction l

I l

I

TARLE 346.7

_CONTAINNENT PENETRATION CliECKLIST (Inside Containment)

PENET.

ISOLATION VALVES ISOLATION VALVE ISOLATION VALVE NO.

SERVICE PROVIDED IDENTIFICATION NUMBER POSITIONS INIT.

j AREA SilUT ESF i

gg

pggL, DOWN ACTION 94-C CNHT Vacuum Hanual IIVC-1CV-151 -

Shut Shut AS-IS None s

j Ejector Suction j

95-C Spare

}

96-B liigh IIcat! S.I. to Check 1S1-95 Shut Shut AS-IS None Cold Legs l

97*-1-A RilS' Inlet Sample Auto-Trip TV'1SS-104A1 Open Open Shut S-CIA j

97*-2-A RilS Outlet Saa:ple Auto-Trip TV-ISS-103Al Open Open Shut S-CIA 97*-3-A CNMT Leakage Hone

.None i

Honitoring Open Taps 9)*-4-A Stm Cen 1C Blowdown System

{

Sample 98-1-C 0xygen Capped Capped N/A N/A N/A N/A 98-2-C Argon Capped Capped N/A N/A N/A N/A

}

98-3-C Acetylene Capped Capped N/A N/A N/A.

N/A l

98-4-C Spare

{

99-C Spare 100-B Spare 101-B Spare

}

102-B Spare 103-A Refuleing Cavity Purif Nanual IPC-38 Shut Open AS-IS None l

Outlet 104-A Refueling Cavity Purif Manual 1PC-9 Shut Open AS-IS None i

Inlet 105*-1-B Stm Lan IB Blowdown System System Sample 105*-2-B Pzr Vapor Sample Auto-trip TV-1SS-112A1 Open Open Shut S-CIA I

105*-3-B Spare 105*-4-B Spare i

106-SgD S.I. Accum. Test Line Auto-Tri'p HOV-IS1-842 Shut Shut AS-IS S-CIA 107-C Spare-108-B, Spare i

i

TABLE 3-6.8

_CONTAINHENT PENETRATION CllEt,.? IST (Inside containment)

}

PENET.

ISOLATION VALVES ISOLATION VALVE ISOLATION VALVE NO.

SERVICE

• PROVIDED IDENTIFICATION NUMBER POSITIONS INIT.

AREA SilUT ESF i

NORMAL FAIL.

g ACTION s

109-C Spare 110-1-C Press Dead weight Hone None i

Calibrator PT-RC-455A 110-2-C Spare 110-3-C Spare 110-4-C Spare 111-C Diluted Fuel Bldg Hanual VS-D--4-6B Shut Open AS-IS None j

Exhaust 11,2-C Fuel Bldg Exhaust Hanual VS-D-9-2 Shut Shut AS-IS 0-RH 113-1-A B.I. Tank to cold lege Check

~ 151-94 Shut Shut AS-IS 0-SIS 113-2-A B.I.T Bypass to cold Check 1S1-94 Shut Shut AS-IS None j

legs

~

l 113-3-A Spare 113-4-A Spare i

l e

e e

I l

4 TABLE 3-7 IDENUFICATION OF RADIONUCLIDE FLOW PATHS 1

RADIONUCLIDE FLOW SYSTEM PATH C0!HENTS Reactor Plant Vents & Drain None Containment sump. primary drain transfer tank and gaseous releases are isolated by CIA l

Supplementary Leak Collection Yes Any leak from containment

,p and Release System into the safeguards area and areas contiguous to contain-ment is vented through the offgas system Boron Recovery System (BR)

None Provided that Chemical and l

Volume Control System (CHS) l is not used Radiation Monitoring System (RM)

None Containment Structure and Mone Personnel air lock and equip-l Air Lock ment hatch are closed and not L

available for personnel c

access

~

Post LOCA Sampling System Yes Not 1n existence but required by U.S. NRC as defined in NUREG-0578 for post-accident conditions

[

l Liquid Waste System (LW)

None Connected to vent and drain i

systems which have no flow path from containment SolidWaste/ Decontamination (SW)

None Connected to CHS, BR, LW, and h

fuel pool G::seous Waste (GW)

None Connected to BR. As long as CHS remains isolated GW should l

not be used.

" Area Ventilation None Separate air supply provided in control room if accident l

occurs Chemical and Volume Control None Except charging pumps in SIS (CHS) recirculation phase., Also pump seal injection water would be used if any RC pumps are operable.

1

1 table W

TABLE 3-7 (Continued)

RADIONUCLIDE FLOW SYSTEM PATH COMMENTS p

Reactor Coolant System (RCS)

None Inside containment SafetyInjectionSystem(SIS)

Yes Recirculate containment sump water in recirculation phase Containment Depressurization, Yes Recirculate containment sump Recirculation Spray water through spray (2 pumps outside containment)

Containment Vacuum Leakage Yes Pressure taps inside contain-Monitoring System, Containment ment and.tample line extending Pressure Sensing Subsystem outside containment Residual Heat Removal System None Inside containment l

(RHR)

Post DBA Hydrogen Yes Draw containment air for hydrogen monitoring and I

recombination of H and 0 2

2 f

l

'I l

I

I

b 1

~

TABLE 3-8 I

\\

TOTAL GAMMA RAY LINEAR ATTENUATION COEFFICIENTS 3

OF CONCRETE (2.30 g/cm densitvl cm"I GAMMA ENERGY, MeV

+2,

0.5 0.2017 1.0 0.1473 l

2.0 0.1034 t

3.0 0.0841 4.0 0.0732 5.0 0.0663 l

6.0 0.0615 i

8.0 0.0554 i

10.0 0.052 i

i

!)

h 1

}

QUAD-1-80-040 4.0 SYSTEM EVALUATION AND RECOM4ENDATIONS This section describes the systems that are necessary for mitigating an accident but which have the potential of carrying radioactivity outside the reactor containment. Existing shielding, resultant radiation levels, I

occupancy requirements, and proposed resolution of problem areas are discussed. - Drawings, diagrams, and tables are included for these systems to supplement the discussions. Specific flow paths, locations, and values of radiation levels are included.

J 9

i O

J

~

I k

4-1 1

1 5

QUAD-1-80-040 4.1 Normal and letdown Flow-Paths Under Accident Condition If the reactor coolant is highly radioactive due to an accident, one pathway out of the reactor containment would be through the Chemical and VolumeControlSystem(CHS). During the TMI-2 accident, the letdown line in the CHS was reopened after isolation which resulted in flow of radionuclides into the CHS and subsequent release to systems not designed for this high radioactivity. The CHS should, therefore, be isolated from the reactor coolant system (RCS) if the reactor coolant is highly radioactive. For a minor accident, where lesser amounts of radioactivity are involved, it may be possible to use the normal letdown line in the CHS without releasing radionuclides into the CHS and other systems to the extent that a serious radiation hazard is created. This section considers the uses of the nonnal and alternate letdown flow paths.

4.1.1 Nonnal Letdown Flow Path

}

Use of the nonnal letdown line is not a safe practice, and therefore should not be used when the radiation level is high in the RCS.

Radiation hazard with highly radioactive reactor coolant flowing thr6 ugh the nonnal letdown line is delineated in this section.

)

The charging and letdown functions of the CHS are employed to maintain a progransned water level in the reactor coolant system

)

pressurizer during all phases of plant operation. This is achieved by means of continuous feed (charging) and bleed (letdown) process.

p During normal operation, the charging rate is controlled by the charging line flow control valve. The~ reactor coolant system I

pressurizer water level error signal resetting the flow controller setpoint in the proper direction satisfies operational requirements resulting from the selection of one or combinations of three flow l

restriction orifices in the letdown line.

l 4-2

QUAD-1-BG-040 The letdown coolant flow path leads to the purification system k

wtere the coolant nonna11y flows through one of the two mixed bed dem m eralizers, the reactor coolant filters, and into the volume control tank (VCT). The chargi.ng pumps take suction from the VCT O

and return the cooled, purified reactor coolant to the RCS via the charging line where flow rate is controlled by the flow control valve.

A minimum flow for charging pump protection is provided by continuously diverting a portion of the charging pump discharge back to the VCT through the charging pump orifice and the seal water heat exchanger.

Another portion of the charging flow is diverted to the reactor coolant pump seals via a seal water injection filter.

The potential radionuclide flow paths resulting from use of the CHS letdown line are shown in a simplified flow diagram, Figure 4-1.

To minimize the radionuclide flow paths within the CHS, E11 the charging flow can be taken from the VCT by diverting all of the letdown flow into the VCT. In this operating scheme, the letdown flow is circulated back to the RCS from the charging line through the VCT. Since the main objective is to maintain a sufficient water level in the reactor vessel and not the purification of the reactor coolant water, the demineralizers and reactor coolant filter may be bypassed.

Radiation levels and occupancy requirements in the CHS are sunnarized in Table 4-1.

If the level in the VCT increases, the letdown flow may have to be J

.4 diverted to the degasifier of the boron recovery system.

If the

~

pressure builds up in the VCT, the gaseous content of the VCT is released to the degasifier through the pressure relief line. The degasifier and associated equipment in the feed, overhead, and discharge lines (exchangers, heater, condenser, chiller, and cooler) 4-3

QUAD-1-80-040 are in a shielded enclosure. The personnel access to the degasifier l

shielded enclosure should be controlled. Degassed water from the degasifier, which is relatively clean and low in radioactivity, can be either pumped back to the VCT or stored in the coolant recovery tank.

Radioactive gases from the degasifier are directed ty the system pressure gradient to the gaseous charcoal delay subsystem upstream of the overhead gas compressor. The overhead compressor directs the radioactive gas stream to the gas surge tank. Most of the gas flow is reduced in pressure and returned to the VCT. A quantity of gas is periodically discharged from the surge tank to one of the three decay tanks for eventual release to the atmosphere via the l

process vent on top of the cooling tower.

The release of the letdown flow into the degasifier should be held to a minimum to prevent spread of the radioactive gases to the gsseous waste system. Radioactive gases are difficult to contain i

l even in closed systems and often result in airborne contamination.

Use Jf the normal letdown line will most likely result in radioactive h

liquid and gas flow in the CHS, boron recovery system, and gaseous waste system.

Figures 4-2A, 28, and 2C indicate potentially radioactive areas in the Auxiliary Building due to radionuclide transport through the CHS.

4.1.2 Alternate Letdown Flow Path Under Containment Isolation In this scheme, the containment is assumed to be isolated so that the radioactive gases and liquid will be contained in the reactor l

containment. The RCS is charged through the normal charging line; the charging pumps take suction from the refueling water storage l

tank. The normal letdown line,'cwever, is closed to prevent radioactive re~ctor coolant from flowing Out of the containment into the Chemical l

4-4

QUAD-1-80-040 j

and Volume Control System (CHS) and other systems. Figure 4-3 presents a simplified flow diagram of the alternate letdown flow paths for containment isolatioa.

I The Containment Isolation Train A (CIA) signal will automatically close the letdown orifice isolation valves and.itop the letdown flow from the RCS. The CIA signal is initiated by a containment high pressure signal or a safety injection signal (SIS). A con-tainment high' pressure signal simultaneously actuates safety injection. This project for Duquesne responds to TMI Lessons Learned, Section 2.1.6.b, which addresses additional shielding required for post-LOCA radiation levels. This report does not, therefore, address whether high radiation without a pipe break will I

result in a CIA. In the future Duquesne may have to respond to

+

the question of whether high primary coolant loop radiation will i

activate a CIA. This study assumes that a CIA signal does occur and in the event of a fuel related accident, therefore, containment lE isolation is actuated.

The safety injection signal is activated simultaneously with CIA.

l The safety injection signal will start the high head safety injection /

charging pumps. The suction of the high head safety injection / charging pumps is diverted from the volume control tank to the refueling water storage tank.

Isolation valves on the reactor coolant pump seal water return line are also automatically closed. Other sources of water such as emergency boration could add to the accumulated water volume.

The automatic closing of the letdown line, volume control tank discharge line, and reactor coolant pump seal water return line by SIS and CIA will isolate the CHS from the RCS. The CHS is isolated at the containment boundary except for the charging pumps and the piping in the safety injection flow path.

4-5

QUAD-1-80-040 b

With the CHS isolated, overflow from the RCS resulting from the safety injection and other water input sources will be confined within the reactor containment. If there is no pipe break, the excess water in the RCS will overflow from the pressurizer into the pressurizer relief tank through the pressure control valves. The pressurizer relief tank content can be transferred to the primary drain transfer tank by manually opening the motor operated valve (MOV-RC-523) in the discharge line. As the primary drain transfer tank fills up, the relief valve will open to relieve pressure in the tank. The liquid which is forced out will drain into.the reactor containment sump.

All gases and liquids from the pressurizer overflow will be confined i

within the reactor containment. The pressurizer relief tank vent line contains two containment isolation trip valves which will

(

close on a CIA signal. The primary drain transfer tank vent, tied into the pressurizer relief tank vent line, is also isolated. The primary drain transfer tank discharge is isolated automatically on a CIA signal by two in-line containment isolation trip valves.

i Similarly, the reactor containment sump discharge is also isolated h

automatically on a CIA signal by two in-line isolation trip valves.

Isolation of the vent and drain lines assures that the pressurizer l

overflow remains in the reactor containment.

l Rupture discs are installed on the pressurizer relief tank to l

protect the tank from overpressurization. High-pressure gas from the pressurizer could rupture the discs. Installation of a relief valve should be considered to prevent rupturing the discs and thus educing the amount of activity release to the containment atmos-r phere through water and steam released from ruptured discs.

The alternate route to letdown from the reactor coolant system utilizes the excess letdown heat exchanger (CH-E-4). The letdown flow can be discharged from the reactor coolant loop (or loops) 4-6

MJAD-1-80-040 into the primary drain transfer tank. This can be achieved by opening one or more of the motor operated valves (MOV-RC-557A, 5578, and 557C) on the reactor coolant legs a.

lining up the excess letdown divert valve (HCV-CH-389) to the drain system (see Figure 4-3).

_O With the hypotheses that there is either no break or only a minor pipe break, the CIA is activated and two of the three reactor coolant pumps are running, the excess letdown flow-line may be effectively used to maintain the water levels in the pressurizer and the reactor pressure vessel.

It is assumed that the charging pump (or pumps) are running, the minimum flow and test line in the pump discharge is opened, and the normal charging line through the regenerative heat exchanger (CH-E-3) to the RC-P-1B cold leg is opened. Under these conditions, letdown outflow has to balance not only with inflow through the chargino line, but also with seal water leakage into the RCS fru.4 the reactor coolant pumps. Since letdown through CH-E-4 is limited by line flow resistance, the

=

charging flow rate must be reduced with flow control valve FCV-CH-122 in the charging line to balance the letdown outflow.

!O To reduce the charging flow rate, the following operator actions may be necessary:

(1) leave only one charging pump running, and (2) open the needle valve which is in parallel with the orifice in the minimum flow and test line in the pump discharge.

It is assumed that cooling water is available to CH-E-4.

The letdown flow is discharged to the containment sump through a relief valve in the bottom of the primary drain transfer tank (DG-TK-1).

Since the capacity of DG-TK-i (700 gallons) is limited, the letdown flow depends on proper operation of the relief valve. To ensure flow of the letdown stream into the containment sump, it is recom-mended that the present locked-shut manual valve in the excess letdown bypass line to the containment sump be replaced with a l

4-7 l

QUAD-1-80 040 remotely operated valve. The remotely operated valve wculd be E

operated from the Control Room and norna11y key-locked.

1 4.2 Safety Injection System O

The Safety Injection System operates in two phasss; the injection phase and the recirculation phase. The injection phase involves any combination of the following operations: 1) injection of borated water into the reactor vessel from the passive accumulators, 2) initial injection of borated water into the reactor vessel from the boron injection tank by the high head safety injection (HHSI)/ charging pumps taking suction from the refueling water storage tank (RWST), and 3) continued injection with the high head safety injection / charging pumps and with the low head I

safety injection (LHSI) pumps both drawing borated water from the refueling water storage tank.

The transfer from safety injection phase to recirculation phase will automatically take place on a low level signal from the RWST. The recirculation phase involves recirculation of coolant and injection water released to reactor containment back to the reactor vessel from l.

the reactor containment sump, by using the LHSI pumps and HHSI/ charging O

pumps if the reactor pressure is still high.

i In the recirculation phase, if the reactor containment sump water is highly radioactive, the radioactive water is recirculated through the Safety Injection System. Figure 4-4 presents a schematic drawing which shows the flow paths of the radioactive water in the recirculation phase.

'F.edundant motor operated valves (MOV's) in the minimum flow and test line from the HHSI/ charging pump discharge are closed en safety injection signal, so the potential radionuclide flow path through the minimum flow line into the volume control tank of the Chemical and Volume Control l

System is effectively cut off. Similarly, redundant MOV's in the reactor coolant pump seal water return line, witich is taken from the HHSI/ charging 4-8

QUAD-1-80-040 pump discharge, are closed on containment isolation signal A (CIA).

g During the automatic switchover from the injection to the recirculation E

phase, redundant M0V's in the minimum flow line from the LHSI pump discharge are automatically closed and the HHSI/ charging pump suction is realigned from the RWST to the discharge of the LHSI pumps. Therefore, J

the potential radionuclide flow path from the LHSI pump discharge to the RWST is blocked off. A pressure relief valve is installed in each of the LHSI pump discharge lines plus the coninon LHSI lines to the RCS cold leg loops.

If the LHSI pump discharge line is blocked while the pump is running, the reactor containment sump water may be drained into the safeguards area sump through the pressure relief lines. LHSI pump discharge high pressure alarm may be needed to warn the operator of blocked pump discharge. However, blockage of the LHSI pump discharge line is not likely to occur, because the MOV's in the connecting lines between the LHSI pump discherge and HHSI/ charging pump suction are open upon receipt of the switchover signal during the transfer of injection I

mode to recirculation mode. The design effectively prevents any radio-active water from leaking into other systems from the Safety Injection System.

1 Table 4-2 summarizes predicted post-accident radiation levels and occupancy j

requirements during the recirculation phase. The LHSI pumps (two pumps) are housed in separate pump cubicles in the west safeg'uards area. Each r

pump cubicle is totally enclosed by shielded concrete walls and concrete floor above the pump cubicle. The only access into the pump cubicle is i

by removing the shielded plug on the floor above and descending to the 8

pump floor by use of a ladder or through the open doorway on the 747' 1

level. Streaming should not be a problem out the open doorway; it should be administrative 1y controlled such as by roping it off. The

.sucti n line t each LHSI pump from the reactor containment sump is lE W

routed to the pump cubicle through a dedicated penetration in the reactor

{

containment wall. Similarly, the discharge lines (cold and hot legs) from the LHSI pumps back to the reactor vessel are routed through dedicated l

penetrations in the reactor containment wall. The LHSI pumps are redundant pumps, each with 100 percent capacity. There is no need for personnel 4-9

QUAD-1-80-040 b

access into the shielded pump cubicles during an accident condition.

Based on the indicated radiation levels and no expected personnel access requirements, additional shielding for the pump cubicles is not necessary.

p potential high radiation areas during the recirculation phase is indicated in Figure 4-5.

M0V's in the LHSI pump suction lines are installed in the valve pit which is on the 687'-11" level in the west safeguards area. The valve pit is well shielded from floor level 747', where access to the valve pit is possible by use of a ladder. These M0V's can be manually opened or closed with a reach rod from the shield floor above, if necessary.

4

)

The LHSI pump discharge line to the suction of HHSI/ charging pumps (recirculation lines) runs under the concrete floor which is used as i

personnel pathway in the safeguards area and areas contiguous to contain-ment and runs in the shielded concrete pipe vaults in the Auxiliary Building. Radiation levels in the pathways in the safeguards area and Auxiliary Building indicate that these areas are accessible for passing throughunderadministrativecontrol(Figure 4-5).

Exception is the area where the recirculation lines are exposed in the west-wall pathway l

Dg and over the pipe trench No. 1 near the charging pump cubicles at 722'-6"

's elevation in the Auxiliary Building. This area will be highly radio-active during the recirculation phase and should be administrative 1y controlled to prevent the personnel from passing by the exposed recircu-F lation lines. Radiation levels in these areas decrease rapidly with time (Table 4 2).

. Shortly M ter switchover to recirculation phase, the operator must operate the HHSI/ charging pump suction and discharge valves manually from the south wall of the pump cubicles to set up redundant and indepen-j dent flow paths from containment sump through HHSI discharge to cold j

legs. The operator will be exposed to radiation from the pipe trench No. 2 and also from the Boron Injection Tank which is about 30 feet l

away. Assuming that it takes abaut 5 minutes for the valve switching 4-10

QUAD-1-80-040 u

operation, the operator will receive dosage which is appreciably less than the allowable limit of 5 Rem whole body. The following change in operating procedure, however, is recomended to reduce radiation exposure:

O o

Switch from injection through the Boron Injection Tank line to the HHSI discharge through cold legs prior to start of the recirculation phase, or o

Delay switching from injection through the Boron Injection Tank line to the HHSI discharge through cold legs during recirculation phase to allow sufficient time for radioactive decay in the Boron Injection Tank and recirculation lines.

There is no need for personnel to be on the floor level below the pathway where the LHSI pump discharge lines are routed in the safeguards area except in the pipe penetration area where access into this area is needed to line up the hydrogen recombiners (inlet manual valves 101 and 102 or 103 and 104) and cross-connect instrument air to the containment l

instrument air (Valve No. ISA-90).

It must be assumed that any accident in containment will result in loss of instrument air in containment.

O The only way to restore instrument air into containment is by cross-connecting it from the station instrument air. The radiation level is estimated at 3,000 R/hr at one hour after accident in the pipe tunnel where the LHSI I

lines are routed. It is recommended that the recombiner inlet manual valves and the instrument air valve (ISA-90) be made remotely operable.

The HHSI/ charging pumps (three pumps) are housed in separate pump cubicles n the Auxiliary Building at fl r level 722' 6".

Each cubicle is E

e enclosed with the shielded walls and the floor above the cubicle (elevation 735'-6"). Personnel access into the pump cubicles is not needed during the recirculation phase when highly radioactive sump water may be pumped.

Access to other areas of the Auxiliary Building is readily available on the floor levels 722'-6" and 735'-6" by passing around the pump cubicles.

If* containment isolation signal B (CIB) is activated, the HHSI/ charging l

4-11

)

QUAD-1-80-040 b

1 pump cubicles are automatically vented to the Supplementary Leak Collection and Release System. It may be a good safety precaution to rope off the floor area directly above the pump cubicles and make certain tnat the 3

shielded plugs (for personnel access to the pump cubicles) are firmly in place.

The HHSI/ charging pump discharge line is routed from the Auxiliary Building through the area contiguous to containment to the reacter vessel through a dedicated containment wall penetration. In the Auxiliary Building the pump discharge line is routed through the recessed pipe tunnel at floor level 722'-6" and the shielded concrete pipe vault.

In the area contiguous to containment, the pump discharge line runs in the l

pipe tunnel below floor level 735'-6".

There is no equipment on the I

floor below elevation 735'-6" in the area contiguous to containment and personnel access is not needed during operation of the Safety Injection System recirculation phase.

IE i

The high head safety injection line (MOV-SI-836 to cold leg loops) is

{

opened during the recirculation phase. The injection line through the boron injection tank (BIT), which is used during the injection phase, is left open during the recirculation phase. Since the BIT has a 900-gallon capacity, the radiation level outside of the BIT cubicle remains high with the radioactive sump water flowing through the BIT during the re-circulation phase. The calculated contact dose rates at the outer surface of the 2-ft. concrete shield wall are as follows: 810 R/Hr at one hour after shutdown,180 R/hr at 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> after shutdown, and 87 R/Hr at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after shutdown (Table 4-2). The BIT is located in the

' Auxiliary Building at elevation 722'-6".

Although occupancy is not required in the vicinity of the BIT under an accident condition, strict administrative control is necessary to keep personnel away from the BIT.

This will ensure that personnel exposures are maintained within the limits specified in Section 3.6.

There is a pressure relief valve on the boron injection recirculation return line. The pressure relief line is routed to the floor drain via 4-12

QUAD-1-80-040 1

boron injection surge tank. However, the pressurs relief valve setting is higher than the shutoff head of the HHSI/ Charging pump.

0 y

Based on: (1) review of the pipe equipment layouts, (2) the flow paths of the Safety Injection System in the recirculation phase, (3) predicted radiation levels, ano (4) personnel occupancy requirements, it is con-cluaed that the existing shielding is sufficient to protect the personnel L

from over-radiation - osure during an accident condition with strict administrative contec 4.

It is believed that administrative controls I

can be appropriately implemented to maintain personnel exposure well within limits specified in Section 3.6.

To make certain that equipment and instruments in the Safety Injection System are in good operating condition at all times, the plant preventive maintenance procedures should be reviewed and revised as necessary and surveillance and other tests should be perfonned as required. Roping off or barricading g

affected areas and administrative 1y controlling access to these areas E

are considered effective means for controlling personnel exposure.

Consideration should be given to developing methods for draining and decontaminating any pump and the pump cubicle if maintenance becomes necessary due to pump malfunction.

g 4.3 Containment Depressurization System The Containment Depressurization System is designed to cool and depres-g surize the containment to subatmospheric pressure in less than 60 minutes following a design basis accident (DBA). The Recirculation Spray Subsystem of the Containment Depressurization System is capable of maintaining containment at subatmospheric pressure for several months following a DBA. There are four recirculation spray pumps which take suction from the containment sump. Two of the recirculation spray pumps are located outside the containment.

If the sump water is highly radioactive, the radioactive water is recirculated through the recirculation spray pumps.

The outside recirculation spray pumps, components and piping are contained within the annular safeguards structure adjacent to tile west side of 4-13

QUAD-1-80-040 f

containment. Potential radionuclide flow paths through the recirculation spray subsystem, using the outside recirculation spray pumps, are shown in Figure 4-6.

O Table 4-3 sunrnarizes radiation levels in the outside recirculation spray pump system. The outside recirculation spray pumps (two pumps) are l

housed in separate pump cubicles in the west safeguards area (see Figure 4-5).

Each pump cubicle is totally enclosed by shielded concrete walls and concrete floor above the pump cubicle. Access into the pump cubicle is by removing the shielded plug on the floor above and descending to the pump floor by use of a ladder or by entering through an open doorway into the cubicle on the 747' level. Streaming should not be a problem, but the open doorway should be administrative 1y controlled, such as by i

roping it off. The suction line to each outside recirculation spray pump of the containment sump is routed to the pump cubicle through a dedicated penetration in the containment wall. The outside recirculation spray pumps are redundant pumps. The pumps and valves are all remotely operable and located in the west safeguards area which is not normally occupied.

There is no need for personnel access into the shielded pump cubicles during an accident condition. Additional shielding for the pump cubicles is therefore not necessary.

M0V's in the outside recirculation spray pump suction lines are installed j

in the valve pit which is on the 687'-11" level. The valve pit is well shielded from floor level 747' where access to the valve pit requires 4

the use of a ladder. These MOV's can be manually opened or closed with a reach rod from the shielded concrete floor above, if necessary.

E

. It is'n t anticipated that the outside recirculation spray pumps will t

G have to function for more than a few days at most; all need for them

• should end within a few months.

i 4.4 Containment Vacuum and Leakage Monitoring System The Containment Vacuum and Leakage Monitoring System consists of five (5) subsystems, as follows:

4 14

QUAD-1-80-040 1.

Containment Pressure Sensing Subsystem (Safety-Related)'

j 2.

Reference Volume Subsystem (Non-Safety Related)

O 3.

Containment Atmosphere Monitoring Subsystem (Safety Related) 4.

ContainmentVacuumSubsystem(SeeNote1,below)

J 5.

ContainmentBlowdownSubsystem(Non-SafetyRelated)

NOTE 1: Containment Vacuum Subsystem must be actuated, per Technical Specification, to reduce the containment atmosphere pressure to a minimum of 8.9 psia prior to becoming critical and during all power reactor operation. The subsystem must maintain the containment atmosphere within the specified subatmospheric range during reactor operation or the reactor must be shut down. This system should not function after either a LOCA or TMI-2 type event.

t l

The safety-related functions of the Containment Pressure Sensing Subsystem are:

a)

Monitor the containment atmosphere pressure for post-LOCA or TMI-2 type event, and g

b)

Actuate the Containment Isolation System when a pre-set pressure is achieved inside containment.

The safety-related function of the Containment Atmosphere Monitoring

' subsystem is to provide periodic evaluation of the concentration of radioactive particulates and noble gases in the containment atmosphere.

The existing equipment is probably inadequate for the post-event situation and isolates on CIA and should remain isolated.

4-15

QUAD-1-80-040 1

The Containment Pressure Sensing Subsystem is the only subsystem of those indicated above that will contain radioactive gases after a LOCA or THI-2 type event. Referring to Figure 4-7, the four 3/8" pipes D

between the inside of containment to the pressure sensing instruments outside of containment pass through containment at an azimuth of approx-imately 20' on the far side of the shield wall to the right of pumps CV-P-1A and CV-P-1B. The system contains flow rate restricting orifices and particulate filters inside containment, with manual valves and l

instruments outside containment.

?

The source term is the volume of containment atmosphere contained within 4-3/8" diameter pipes, each 30 feet long. This radiation source is l

conservatively approximated by a single 1" diameter pipe located 3'-0" above the 722'-6" floor elevation and 3' horizontally from the " Receptor" outside the 2' thick concrete shield wall. Figure 4-8 shows the potentially radioactive area of the Containment Pressure Sensing Subsystem.

Table 4-4 provides pertinent data for evaluating radiation hazards for i

post-accident release of gases to the Containment Vacuum and Leakage Monitoring System. Based on the radiation levels (contact with shielding) and the occupancy requirements, it is concluded that personnel exposures can be maintained within the allowable limits specified without any shielding modifications.

D I

4.5 Supplementary Leak Collection and Release System The function of the Supplementary Leak Collection and Release System is

. to ensure that radioactive leakage from the reactor containment following a Design Basis Accident (DBA), radioactive release due to a fuel handling l

accident, or radioactive material released in the waste gas storage area j

or areas contiguous to containment is collected and filtered for iodine i

removal prior to discharge to the atmosphere at an elevated release point.

l 4-16

QUAD-1-80-040 1

The Supplementary Leak Collection and Release System consists of two 100 percent capacity leak collection exhaust fans. Air is 'exhaurited from the fuel building, waste gas storage area, blowdown tank room, personnel 3

access hatch area, east and west cable vaults, pipe tunnel, north and west safeguards, MCC-1 area and the Hydrogen Recombiner-1 room.

During normal operation, the exhaust to the fan does not go through filters. On a containment isolation phase I, signal (CIA) or a high-high radiation signal from the monitor serving areas contiguous to the con-tainment, from the fuel building monitor, the waste gas storage area monitor, or the leak collection system exhaust monitor, flow is diverted so that it first flows through one of the two parallel main filter banks before flowing to the leak collection exhaust fans.

During normal power operation, the three charging pump cubicles are ex-g hausted by the Auxiliary Building Ventilation System A.

In the event of E

a containment isolation phase B (CIB) signal, the isolation dampers to the Auxiliary A Exhaust System are closed and the parallel dampers which connect the charging pump cubicles to the Supplementary Leak Collection and Release System are opened. This assures that any leakage or release from the charging pump cubicles is filtered via the main filter bank prior to release at the elevated release point.

Figure 4-9 presents a simplified flow diagram of the Supplementary Leak Collection and Release system. The primary potential for radionuclide transport is through leaks through containment penetrations. Other potential sources of leakage are the external recirculating loops of the Safety Injection System and Recirculation Spray Subsystem. However, leakage from the external recirculation loops (pump seals, valves, and flanges) is negligibly small compared to the containment leak rate, unless there is failure of a pump seal or excess leakage from a valve or flange (BVPS FSAR, Tables 6.3-9 and 6.4-4).

Pump seal failure and excess leakages from valves and flanges are not considered in the dose rate calculation. The containment leakage rate was assumed to be 0.10 percent by weight of the containment air per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (limit of Technical Specications, Appendix A).

4-17 1

1 QUAD-1-80-040 1

Table 4-5 summarizes radiation levels frora the Supplementary Leak Collection and Release System. The radiation levels from the main ducts were calculated at the distances of 10 ft and 20 ft. Dose rates are the i

estimated maximum values. As the reactor containment is depressurized with spray cooling, the radiation levels will decrease rapidly. The radiation level of the filter bank with two-foot thick shielded wall was based on iodine removal for 30 minutes at the assumed maximum containment leak rate. The 30 minutes duration was used with the assumption that the containment can be depressurized in approximately 30 minutes after a LOCA. Figure 4-10 illustrates radiation levels from the Supplementary Leak Collection and Release System.

The radiation levels indicate that passage near the vent ducts and around the shielded filtir bar.ks is pemissible as the need arises. It is expected that post-acciden. Mcupancy requirements in the vicinity of g

system ducting and filter bank will be minimal. Although the exhaust E

fans are in an open area at elevation 768'-7" i.' the Auxiliary Building, the radiation level is much lower than in the main ducts upstream of the filter banks.

iherefore, it is concluded that addition of shielding is not necessary. Personnel exposures can be controlled by implementing personnel access restrictions.

4.6 Hydrogen Recombiner 3

Two redundant Hydrogen Recombiner Systems are installed to maintain hydrogen content in the containment below the lower flammability limit in the event of a LOCA. A simplified flow diagram of the Hydrogen Recombiner System is shown in Figure 4-11. The Hydrogen Recombiner room

' consists of two hydrogen recombiner units, two hydrogen analyzers and one control and power panel for each of these systems.

It is assumed that the hydrogen analyzer is operated M thin one hour after the accident and the hydrogen recombiner at appre,N aly 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the accident.

There is a biological shield wd A tu the recombiners and control panels. The shield wall is mat; ot a e are with a thickness of 1.5 feet and extends 12 ft. from the containment wati, Two 6" Low Head Safety 4-18 1

QUAD-1-80-040 Injection (LHSI) discharge pipes run 167" below the control panel area, but the control panel area is on grating. Operator in the control panel room will be directly exposed to the LHSI lines. ine radiation dosage 0

received by an operator working inside the control panel area is sunined over contributions from the H recombiner units, the LHSI pipes, and 2

several other pipes that go in and out of the recombiner room.

The source terms f'or radiation and shielding calculations were conservatively obtained as follows:

1)

The total amount of radiation for each of 12 mean energy groups was obtained from the Origen Code for 2766 MW and 650 days of operation, 6

3 I

then divided by containment free volume,1.86 x 10 ft, to give the concentration of radioactivity level per unit volume of the gas l

inside containment.

2)

For each recombiner unit, effects from two different sources are combined: one source is the tank containing the reaction chamber with heating and coolig pipes around it; the other is the vessel containing the blower and motor.

3)

The gas volume in the reaction chamber and pipes surrounding the chember is estimated as follows: the reaction chamber is assumed to be a right cylinder with 12" I.D. and 4'-0" height; the pipes O

surrounding it are estimated to have a total length of 140'; thus 3

the total gas volume in the tank is about 6 ft. This amount of radioactive gas is then assumed to be a uniformly distributed line l

source of 3.3' located at the edge of tank closer to the control l

panel.

4)

The blower and motor are mounted in a sealed stainless steel vessel of 20" 1.D., and 47-1/2" in length. The gas volume is conservatively estimated as 8.7 ft.3 Again, this amount of gas is assumed as a line source 4' long located at the edge of the vessel closer to control panel.

4-19 1

QUAD-1-80-040 5)

There are many pipe lines inside the recombiner room. Eight 3/8" pipes and two.l' pipes are assumed to contain radioact.iye gas. (.For a list of these pipes refer to Table 4-6).

For conservatism, each pipe is assumed to be 30' long and instead of going in the direction perpendicular to. shield wall, it is assumed all of them are parallel to the shield wall and 12' away from the operator.

Because the biological shield wall extends only 12' from the containment wall, back-scattering of photons from the side wall should be considered.

However, based on the calculation, most of the dosage that an operator would receive would be from two LHSI lines; e.g., at I hour after the accident he would receive a maximum of 3,000 R/hr from the LHSI lines versus 25 mR/hr from the recombiner room due to operation of the hydrogen analyzers. Thus, t: yen if the dose rate from back-scattering is on the order of 25 mR/hr, it will be negligible compared with that from the LHSI lines.

At one hour after the accident, an operator will have to start the hydrogen analyzer anc check the readings on the control panel. The occupancy i

time required for this activity should not exceed a total of 30 minutes.

At approximately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the accident the recombiner will be started.

I Approximately 30 minutes is required for this startup. Table 4-7 presents the radiation level and occupancy requirement in hydrogen recombiner control area. Figure 4-12 presents the recombiner room layout and radiation level at the control panel.

The estimated dose rate is maximum 3,000 R/hr at one hour after accident in the control panel area where operator is standing on the grating and

~

directly exposed to the LHSI lines. The operator will be exposed to a l

maximum dose of 1,500 Rem during the occupancy time of 30 minutes. The l

radiation exposure far exceeds the limit of General Design Criteria 19 I

(i.e., 5 Rem total body). Therefore, the control panels must be relocated to a concrete floor area or the centrol panel room must be shielded from the LHSI lines.

l 4-20 l

1

E QUAD-1-80-040 Manual valves (101 & 102 or 103 & 104) on recombiner line have to be opened first to start the hydrogen analyzer and recombiner. These valves are located in the pipe penetration area of safeguards where the D

LHSI lines with radioactive containment sump water are routed. The raf : ion level is estimated at 3,000 R/hr at one hour after accident in the pipe tunnel where the LHSI lines are routed. Therefore, the operator cannot manually open the recombiner inlet manual valves.

It is recom-mended that the manual valves be either changed to remotely operated valves or open from the floor above with shielding and using reach rods.

4.7 Post-Accident Sampling System The existing reactor building sampling system draws liquid samples from the primary coolant cold leg, hot leg, pressurizer relief tank liquid and gas, residual heat removal and safety injection accumulator water as l

shown in Figure 4-13. Upon CIA actuation, the containment isolation valves close inside and outside containment. The inside and outside l

cor.tainment isolation valves are operated from one solenoid valve and a 1M: switch as, shown on Figure 4-13. Therefore, if only one sample is required, all eight containment isolation valves must be opened. The liquid sample panel now in existence does not have facilities for handling such a highly radioactive sample.

l There are no provisions for taking a sample of post-accident containment air in existence at this time. During an accident conditior, the line

)

which contains a radiation monitor in the Containment Yacuum and Leakage Monitoring System is isolated, thereby leaving no method to monitor i

containmenc activity. Although the hydrogen analyzers are turned on

' within one hour after the accident, there is no method of extracting a sample of the containment from this system either. Table 4-8 describes basic sampling requirements as outlined in NUREG 0578.

It is therefore recommended that a post-accident sampling system be built for sampling both liquid hot leg and containment air samples.

lW Design of such a sampling system is not within the scope of this project.

t 4-21 1

QUAD-1-80-040 However, a conceptual design is being proposed and a copy of the simplified flow diagram, plan, and evaluation views, location drawings, and arrangement drawings are included in Figures 4-14, 4-15, and 4-16.

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FIGURE 4-1 POTENTIAL RADIONUCLIDE FLOVATils WITil USE OF NORMAL CHS LETDOWN LINE

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FIGURE 4-2A POTENTIAL HIGH RADIATION AREAS FR0ft USE OF NORMAL CHS LETDOWN LINE

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FIGURE 4-28 POTENTIAL HIGH RADIATION AREAS FROM USE OF NORitAL CHS LETDOWN LINE

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FIGURE 4-2C POTENTIAL HIGH RADIATION AREAS FROM USE OF NORf1AL CHS LETDOWN LINE

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i FIGURE 4-3 ALTERNATE LETDOWN FLOW PATH FOR CONTAINMENT ISOLATION i

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FIGURE 4-4 POTENTIAL RADIONUCLIDE FLOW PATHS IN SAFETY INJECTION SYSTEM RECIRCULATION PilASE

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(1)

Shaded areas with dashed lines indic ate potential radiation areas.

(2)

Radiation levels are estimated at or.e hour after accident and in contact with shielding FIGURE 4-5.POTENTI L HIGH RADIATION AREAS DURiilG SAFETY INJECTI N SYSTEM RECIRCULATION PHASE

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1 Note: (1) Shaded area with dashed lines indicates potential D

radiation area.

(2) Radiation level is estimated at one hour after I

accident and in contac+. with shielding.

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I FIGURE 4-9 POTENTIAL RADIONUCLIDE FLOW PATHS IN SUPPLEMENTARY LEAK COLLECTION AND RELEASE SYSTEM i

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TYFICAL VENT DUCT (54" x 45") IN SAFEGUARDS AREA.

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1 FIGURE 4-10 POTENTIAL HIGH RADIATION AREAS OF SUPPLEMENTARY LEAK COLLECTION AND RELEASE SYSTEM

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NOTE: Radiation level shown is at control panel area and one hour after accident.

I FIGURE 4-12 RECOMBINER ROOM LAYOUT 1

INSIDE OUTSIDs CONTAINENT CONTAINENT CIA CIA db

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N COOLANT COLD LEG i

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FIGURE 4-13 EXISTING REACTOR BUILDING SAMLING SYSTEM AND PROPOSED 4

POST-ACCIDENT SAMPLING SYSTEM

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FIGURE 4-14 LOCATION OF POST-ACCIDENT SAMPLING STATION I

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An automatic titrator is one option available l

FIGURE 4-16 FLOW DIAGRAM OF POST-ACCIDENT SAMPLING SYSTEM 1

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b bW TABLE 4-1 RADIATION LEVELS IN CHS WITH USE OF LETDOWN LINE DESCRIPTION EXISTING SilIELDING RADIATIONLEVE(

OCCUPANCY AREA 0F fu1D THICKNESS VS. TIME (HRS)

REQUIREMENTS RESOLUTION EQUIPHENT Zero 10 hrs 24 hrs Floor above Charging and Let-2 ft. thick concrete 32 5.2 2.4 Unlikely Administrative Penetration Room down Lines floor R/Hr R/Hr R/Hr Control (Area Contiguous to Containment)

Auxiliary Bldg.

3-inch Pipe Section 2 ft. thick concrete 22 3.6 1.7 Possibly Administrative in Pipe Chase shield R/Hr R/Hr R/Hr Control (Letdown or Charging Line)

Auxiliary Bldg.

Outside of Volume 3.5 ft. thick 41 3.4 1.2 Possibly Administrative Control Tank concrete shield wall R/Hr R/Hr R/Hr Control (CH-TK-2) Cubicle Auxiliary Bldg.

Outside of 2 ft. thick concrete 3.9 0.67 0.33 Possibly Administrative Charging Pump floor and wall R/Hr R/Hr R/Hr Control (CH-P-1A, IB, & IC)

Cubicle NOTE:

1.

Hours elapsed following tine 0 of the accident or event.

Radiation levels in contact with shielding.

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g-y TABLE 4-2 RADIATION LEVELS IN SAFETY INJECTION SYSTEM DESCRIPTION EXISTING SHIELDING RADIATION LEVE OCCUPANCY VS. TIME (HRS)( REQUIREMENTS RESOLUTION AREA 0F AND THICKNESS EQUIPMENT 1 hr 10 hrs 24 hrs West Safeguards LHSI Pump Cubicle-2 f t. thick floor 2.4 0.6 0.3 Unlikely Administrative directly above R/Hr R/Hr R/Hr Control LHSI Pump Cubicle 2 ft. thick floor 7.7 0.9 0.3 Unlikely Administrative standing aside and side wall mR/Hr mR/Hr mR/Hr Control Recir. Spray & LHSI 2 f t. thick floor 4.8 1.2 0.6 Unlikely Administrative Pump Discharge Lines-R/Hr R/Hr R/Hr Control Pipe Tunnel Floor Safeguards and Pathway above pipe 2 ft. thick floor 4.6 1.0 0.4 Unlikely Administrative Contiguous Areas tunnels R/Hr R/Hr R/Hr Control to Containment Auxiliary Bldg.

Pathway above pipe 2 f t. thick floor 21.8 4.8 2.4 Possibly Administrative vaul t R/Hr R/Hr R/Hr Control Auxiliary Bldg.

Charging Pump Cubicle 2 ft. thick floor 410 96 47 Unlikely Administrative directly above mR/Hr mR/Hr mR/Hr Control Charging Pump Cubicle 2 ft. thick floor 1.2 0.1 0.04 Possibly Administrative standing aside and side wall mR/Hr mR/Hr mR/Hr Control Outside Boron 2 ft. thick shield 810 180 87 Unlikely Administrative Injection Tank wall R/Hr R/Hr R/Hr Control 30 Feet from Boron 2 ft. thick Boron 15.5 3.7 1.8 Likely Administrative Injection Tank at Injection Tank R/Hr R/Hr R/Hr Control South Wall of Shield wall Charging Pump Cubicle NOTE:

1.

Hours elapsed following time 0 of the accident or event.

Hadiation levels in contac.t with shielding

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TABLE 4-3 RADIATION LEVELS IN OUTSIDE RECIRCULATION SPRAY PUMP SYSTEM l

DESCRIPTION OF HISTIE SHIRDING MDIATION LEVR OCCUPANCY AREA

. RESOLUTION EQUIPMENT AND THICKNESS VS TIME (HRS)1 REQUIREENTS l

1 hr 10 hrs 24 hrs l

West Safeguards Outside Recirc 2 feet thick 2.4 0.6 0.3 Unlikely Administrative Spray Pump concrete floor R/hr R/hr R/hr Control Cubicle Directly Above Outside Recirc 2 feet thick 7.7 0.9 0.3 Unlikely Administrative Spray Pump concrete floor and mR/hr mR/hr mR/hr Control Cubicle Standing' side wall Aside l

Recirc Spray &

2 feet thick 4.8 1.2 0.6 Unlikely Administrative LHSI Pump concrete floor R/hr R/hr R/hr Control l

Discharge Lines -

i Pipe Tunnel Floor l

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NOTE:

1.

Hours elapsed following time 0 of the accident or event.

- Radiation levels in contact with shielding.

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TABLE 4-4:

SVfEARY, RADIATION LEVEL OUTSIDE CONTAINMENT FOR CONTAINMENT VACUUM AND LEAKAGE tiONITORING SYSTEM, DESCRIPTION OF EXISTING SHIELDING RADIATION LEVEL OCCUPANCY

. RESOLUTION AREA EQUIPMENT AND THICKNESS VS TIME (HRS)1 REQUIREMENTS 1 hr 10 hrs-24 hrs Containment 3/8" pipe, valves 2 feet thick 6.8 1.3 0.45 Infrequent Administrative Atmosphere and instruments concrete wall mR/HR mR/HR mR/HR Control Monitoring 4 sets System Outside 3 ft. from pipe Containment Ref. Volume Not safety N/A Subsystem related Containment Existing System.

Investigate Atmosphere may not be range Moni toring adequate capability System Cont. Depress.

Not safety N/A subsystem related Cont. Blowdown subsystem Not Mafety N/A related I

i l

'l NOTE: 1 Itours elapsed following time 0 of the accident or event and all radiation levels in R/hr.

- Radiation levels in contact with shielding i

y TABLE 4-5:

RADIATION LEVELS IN SUPPLEMENTARY LEAK COLLECTION AND RELEASE SYSTEH 1

i DESCRIPTION OF EXISTING SHIELDING RADIATION LEVEL OCCUPANCY AREA

• RESOLUTION j

EQUIPMENT AND THICKNESS

_VS TIME (HRS)l__

REQUIREMENTS l

Zero l

Containment Main Duct None 200 Unlikely Administrative Contiguous (at 10 ft mR/HR Control Areas and Safe-distance) guards area Main Duct None 74 Unlikely Administrative (at 20 ft mR/HR Control j

distance)

Auxiliary Bldg.

Main Duc' None 200 Infrequent Administrative (at 10 i.

mR/HR Control distance) i Hain Duct None 74 Infrequent Administrative l

(at 20 ft mR/HR Control j

distance) i Auxiliary Bldg.

Filter Bank 2 ft thick concrete 290*

Infrequent Entrance of wall mR/HR the shielded i

enclosure i

should be

{,

controlled.

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NOTE: 1 Hours elapsed following time 0 of the accident or event.

• Assume all iodines filtered out for 30 minutes after time zero.

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TABLE 4-6 HYDROGEN RECOMBINEP, PIPING APPROXIMATE LENGTH IN PIPE SIZE PIPE NUMBER RECC'3INER ROOM 1

0 3/8" HY-14-N8 284" 3/8" HY-13-N8 368" i

3/8" HY-15-N8 368" l

3/8" HY-17-N8 311" 3/8" HY-5-N8 181" 3/8" HY-3-N8 202" 3/8" HY-4-N8 181" 3.8" HY-7-N8 74" 2"

HY-35-151-Q2 372" 2"

HY-36-151-Q2-240" i

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M*M W

O TABLE 4-7 RADIATION LEVELS IN HYDROGEN REC 0lEINER CONTROL AREA DESCRIPTION OF EXISTING SHIELDING RADIATION LEVEL OCCUPANCY AREA 0WTION EQUIPMENT AND THICKNESS VS TIME (HRS)1 REQUIREMENTS 1 h-10 hrs 24 hrs Hydrogen Control panel 1.5 ft biological 3000

1400*

1000 1 hr. max Rei>cate Recombiner room on grating concrete shield R/HR R/HR R/HR residency per cor.imi panels Control Panel and directly between control person.

to ctnerete exposed to LHSI panel and recombiner floor area or lines shield control panel room from LHSI lines NOTE: 1 llours elapsed following time 0.of the accident or event - Radiation levels near contact with shielding

• At this time 11 rec mbiner is not operating, no contribution from recombiner units is considered.

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Press./Unpress reactor coolant sarnples within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of incident J

l e,s s -

Containment atmosphere sample within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of incident i

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5.0 REFERENCES

5.1 " Engineering Compendium on Radiation Shielding," Volume II O

Shielding Material, edited by R. G. Jaeger and others, Springer-Verlag, New York,1975.

5.2 " Nuclear Reactor Engineering," by S. Glasstone and A. Sesouske.

D. Van Nostrand Company, June 1963.

5.3 " Nuclear Engineering Handbook," edited by Harold Etherington, McGraw Hill Book Company, 1958.

4 5.4 Beaver Valley Unit 1 Final Safety Analysis Report.

5.5 Ducuesne Letter on " Plant Shielding Review and Post-Accident Sampling," dated December 19, 1979.

5.6 NUREG-0578 (TMI-2 Lessons Le rned Task Force Report).

I 5.7 Response to NUREG-0578 Item 2.1.6 a Part II, M.0. Sanford, December 1979, METN-2005 Westinghouse Owners Group.

5.8 Operating Manual Figure Number 6-1 through 46-1.

5.9 Operating Manuals 6 through 46 5.10 10 CFR 50, Appendix A, Criterion 19 5.11 TID 14844, Calculation of Distance Factors for Power and Test Reactor Sites, March 1962.

5-1

.