ML19309C570

Suppl 3 to TMI-1 PSAR, Answers to AEC Questions 13.0-18.7

ML19309C570
Person / Time
Site:	Crane
Issue date:	05/01/1967
From:	JERSEY CENTRAL POWER & LIGHT CO., METROPOLITAN EDISON CO.
To:
References
NUDOCS 8004080805
Download: ML19309C570 (66)

Text

o METROPOLITAN EDISON COMPANY l

THREE MILE ISLAND NUCLEAR STATION UNIT 1

Preliminary Safety enalysis Repert 0

Volume 5

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0004 041

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TABLE OF CONTENTS Section Page 1

INTRODUCTION AND

SUMMARY

Volume 1.

. Tab 1 1-1

1.1 INTRODUCTION

1-1 1.2 DESIGN HIGHLIGHTS.

1-2 1.2.1 SITE CHARACTERISTICS 142 1.2.2 PCWER LEVEL 1-2 1.2.3 PEAK SPECIFIC PCWER LEVEL 1-2 1.2.4 REACTOR BUILDING SYSTEM 1-2 1.2 5 ENGINEERED SAFEGUARDS.

1-2 1.2.6 ELECTRICAL SYSTEMS AND ENERGENCY PCWER 1-3 1.2.7 ONCE-THRCUGH STEAM GENERATORS 1k 1.3 TABULAR CHARACTERISTICS.

1h 1.4 PRINCIPAL CESIGN CRITERIA 1-7 1.h.1 CRITERICN 1 1-7 1.h.2 CRITERION 2 1-9 1.k.3 CRITERION 3 1-9 1.k.h CRITERION k 1-1C 1.h.5 CRITERION 5 1-10 1.4.6 CRITERION 6 1-11 1.k.7 CRITERION 7 1-12 1.h.8 CRITERION 8 1-12 0

1.k.9 CRITERION 9 1-13 1.k.10 CRITERION 10 1-1h l

1.h.11 CRITERION 11 1-lh 1.h.12 CRITERION 12 1-15 l.h.13 CRIIERICN 13 1-16

1. h. lh CRITERION lh 1-17 1.h.15 CRITERION 15 1-17 1

1.h.16 CRITERION 16 1-18 1.h.17 CRITERICN 17 1-19 1.h.18 CRITERION 18 1-20 1.h.19 CRITERION 19 1-21 1.h.20 CRITERION 20 1-21 1.h.21 CRITERION 21 1-22 1.k.22 CRITERION 22 1-22 1.h.23 CRITERICN 23 1-23 1.h.24 CRITERION 2h 1-23 1.k.25 CRITERION 25 1-2h 1.4.26 CRITERION 26 1-2h 1.h.27 CRITERION 27 1-25 I

1.5 RESEARCH AND DEVELOPMENT REQUIREMENTS 1-25 1.5.1 ONCE-TERCUGH STEAM GENERATCR TEST.

1-25 1.5.2 CONTROL ROD DRIVE LINE TEST.

1-26 1.5 3 SELF-PCWERED DETECTOR TESTS.

1-26 l.5.4 THERMAL AND HYDRAULIC PROGRAMS.

1-26 1.6 IDENTIFICATION OF AGENTS AND CCNTRACTORS.

1-27 l

1.7 CONCLUSION

S 1-28 0004 042 i

i

' i Section P_ age 2

SITE AND ENVIRONMENT

• Tab 2 Volume 1

• 2-1 2.1 GENERE DF.SCRIPTION 2-1 2.2 LOCATION, DOPULATION,AND LAND USE 2-1 2.2.1 LOCATION 2-1 2.2.2 POPULATICN 2-2 2.2 3 LAND USE 2-2 s

23 METEOROLOGY 2-3 2.3.1

SUMMARY

2-3 2.3.2 SEVERE WEATHER 2h 233 AVERAGE ATMOSPHERIC DISPERSION.

2k 2 3.k ATMOSPHERIC DIFMSICN FOR ASSESSING ACCIDENTS.

2-7 2.h HYDROLOGY AND GROUNDWATER 2-8 2.k.1 CHARACTERISTICS OF STREAMS IN VICINITY 2-8 2.h.2 OTHER PCWER PROJECTS IN VICINITY 2-9 2.h.3 LCW FLOW STUDIES 2-10 2.h.h FLOOD FLCW STUDIES.

2-11 2.k.5 DESIGN OF PROPOSED DAMS AND SPILLWAYS 2-12 2.4.6 GROUNDWATER 2-14 2.5 GEOLOGY 2-lh 2.6 SEISMICITY 2-15 2.6.1 SEISMICITY 2-15 2.6.2 RESPONSE SPECTRA 2-15

2.7 REFERENCES

2-16 3

REACTOR Volume 1

. Tab 3 3-1 3.1 DESIGN BASES 3-1 3.1.1 PERFOR M CE OBJECTIVES 3-1 3.1.2 LIMITS 3-1 3.2 REACTOR DESIGN.

3-6 3.2.1 GENERE

SUMMARY

3-6 3.2.2 NUCLEAR DESIGN AND EVALUATICN 3-7 3.2.3 THERMAL AND HYDRAULIC DESIGN AND EVAWATION 3-26 3.2.4 MECFANICE DESIGN LAYOUT.

3-51 33 TESTS AND INSPECTIONS 3-82 331 NUCLEAR TESTS AND INSPECTION 3-82 3.3 2 THERMAL AND HYDRAULIC TESTS AND INSPECTION.

3-82 333 NEL ASSEMBLY, CONTROL ROD ASSEMBLY, AND CONTROL RCD DRIVE MECHANICE TESTS AND INSPECTION 3-8h 3 3.h INTERNALS TESTS AND INSPECTIONS 3-90

3.4 REFERENCES

3-91 4

REACTOR COOLANT SYSTEM Volume 1.

Tab h.

k-1 h.1 DESIGN BASES h-1 h.l.1 PERFORMANCE OBJECTIVES h-1 k.1.2 DESIGN CHARACTERISTICS 4-1 h.l.3 EXPECTED OPERATING CONDITIONS h-2 O

h.l.4 SERVICE LIFE h-3 h.1.5 CODES AND CLASSIFICATICNS 4-6 0C0t 043 ii

Section k Page O

h REACTCR COOLANT SYSTEM (CCNTINUED)

Volume 1.

Tab k k.2 SYSTEM DESCRIFIION AND OPERATICH.

<.6 k.2.1 GENERAL DESCRIPTION k-6 4.2.2 MAJOR COMPONENTS k-6 k.2 3 PRESSURE-RELIEVING DEVICES 4-12 k.2.h ENVIRCNMENTAL PROTECTION.

h-12 k.2.5 MATERIES OF CCNSTRUCSON 4-12 k.2.6 MAXIMUM HEATING AND COOLING RATES.

k-lk k.2.7 LEAK DETECTION k-lh 4.3 SYSTEM DESIGN EVALUATION k-16 k.3.1 SAFETY FACTORS k-16 k.3.2 RELIANCE ON INTERCONNECTED SYSTEMS k-23 k.3.3 SYSTEM INTEGRITY h-23 k.3.h PRESSURE RELIEF.

k-23 k.3.5 REDUNDANCY 4-2h k.3.6 SAFETY ANALYSIS h-2k k.3.7 OPERATIONAL LD4ITS.

k-2h k.h TESTS AND INSPECTICNS k-25 k.k.1 COMPONENT IN-SERVICE INSPF 2 ION k-25 k.k.2 REACTOR COOLANT SYSTEM TESTS AND INSPECTIONS k-25 k.k.3 MATERIE IRRADIATION SURVEILLANCE.

4-26 k.5 REFERENCES k-28 5

CCNTAIMMENT SYSTEM Volume 1.

Tab 5.

5-1 5.1 REACTOR BUILDING 5-1 5.1.1 DESIGN BASES 5-1 5.1.2 STRUCTURE DESIGN 5-2 52 ISOLATION SYSTEM 5-10 5 2.1 DESIGN BASES 5-10 5.2.2 SYSTEM DESIGN 5-10 5.3 VENTIIATION SYSTEM 5-12 5 3.1 DESIGN BASES 5-12 5 3.2 SYSTEM DESIGN 5-13 5.k LEAKAGE MONITORING SYSTEM 5-1h 55 SYSTEM DESIGN EVALUATION 5-16 5.6 TESTS AND INSPECTION.

5-16 5.6.1 PREOPERATICNAL TESTING AND INSPECTION 5-16 5.6.2 POSTOPERATIONAL LEAK MONITORING 5-17 6

ENGINEERED SAFEGUARDS Volume 1.

Tab 6.

6-1 6.1 EMERGENCY INJECTION 6-1 6.1.1 DESIGN BASES 6-1 6.

1.2 DESCRIPTION

6-2 6.1 3 DESIGN EVALUATION 6-3 6.1.k TESTS AND INSPECTICNS.

6-6 6.2 REACTOR BUILDING ATMOSPHERE COOLING AND WASHING 6-13 l

6.2.1 DESIGN BASES.

. '6-13 6.

2.2 DESCRIPTION

6-13 0004 044 iii i

Section P_ age 6

EIGINEERED SAFEGUARDS (CONTUTUED)

Volume 1.

Tab 6 6.2.3 DESIGN EVALUATION 6-14 6.2.h TESTS AND UiSPECTICNS.

6-19 6.3 ENGINEERED SAFEGUARDS LEAKAGE AND RADIATION CONSIDERATIONS 6-20 6.

3.1 INTRODUCTION

6-20 6.3 2

SUMMARY

OF POSTACCIDENT RECIRCULATION AND LEAKAGE CONSIDERATIONS 6-20 6.3.3 LEAKAGE ASSUMFTIONS 6-21 6.3.4 DESIGN BASIS LEAKAGE 6-22 6.3.5 LEAKAGE ANALYSIS CONCLUSIONS 6-22 7

INSTRUMENTATION AND CONTROL Volume.2.

. Tab 7 7-1 7.1 PROTECTION SYSTEMS 7-1 7.1.1 DESIGN BASES 7-1 7.1.2 SYSTEM DESIGN 7-4 7.1.3 SYSTEMS EVALUATION.

7-11 7.2

_ REGULATING SYSTEMS 7-15 7.2.1 DESIGN BASES.

7-15 7.2.2 SYSTEM DESIGN 7-17 7.2 3 SYSTEM EVALUATION 7-22 73 INSTRUMENTATION 7-25 O

7 3.1 NUCLEAR INSTRUMENIATION 7-25 732 NONNUCLEAR PROCESS INSTRUMENTATION 7-27 7.3.3 INCORE MONITORING SYSTEM.

7-29 7.h OPERATING CCNTROL STATIONS.

7-31 7.h.1 GENERAL LAYOUT 7-31 7.k.2 INFORMATION DISPLAY AND CONTROL WNCTION 7-31 7.4.3

SUMMARY

OF ALARMS 7-32 7.k.h COMMUNICATION 7-32 7.k.5 OCCUPANCY 7-32 7.k.6 AUXILIARY CONTROL STATIONS 7-33 7.k.7 SAFETY FEATURES.

7-34 8

ELECTRICAL SYSTEMS Volume 2.

. Tab 8 8-1 8.1 DESIGN BASES 8-1 8.2 ELECTRICAL SYSTEM DESIGN 8-1 8.2.1 NET 40RK INTERCONNECTIONS.

8-1 8.2.2 STATION DISTRIBUTION SYSTEM.

8-2 8.2 3 EMERGENCY POWER.

8-5 8.3 TESTS AND INSPECTICNS 8-8 O

0004 045 LY

Section Pge 9

AUE LIARY AND EMERGENCY SYSTEMS Volume 2.

. Tab 9 9-1 91 MAKZUP AND PURIFICAMCN SYSTEM 9-2 9 1.1 DESIGN 3ASES 9-2 9 1.2 SYSTEM DESCRIPTICN AND EVAW ATION.

9-3 92 CHEMICAL ADDIHON AND SAMPLING SYSTEM 9-9 9 2.1 DESIGN BASES 9-9 9 2.2 SYSTEM DESCRIPTION AND EVALUAMON.

9-10 9.3 INTERMEDIATE COOLING SYSTEM 9-18 9.3.1 DESIGN BASES 9-18 9 3.2 SYSTEM DESCRIPHON AND EVEUATION.

9-18 9.h SPENT WEL COOLING SYSTEM 9-22 9.h.1 DESIGN BASES 9-22 9.h.2 SYSTEM DESCRIPU CN AND EVAMAS ON.

9-22 95 DECAY HEAT REMOVAL SYSTEM 9-25 9 5.1 DESIGN BASES 9-25 9 5.2 SYSTEM DESCRIPTION AND EVEUATICN.

9-25 9.6 COOLING WATER SYSTEMS 9-29 9.6.1 DESIGN EASES 9-29 I

9.6.2 SYSTEM DESCRIPHON AND EVAWATION.

9-30 97' WEL HANDLING SYSTEM.

9-5 9 7.1 DESIGN BASES 9-35 9 7.2 SYSTEM DESCRIPTION AND EVAWATION.

9-36 9.8 STATICN VENTILATICN SYSTEMS 9 41 Q

9.8.1 DESIGN BASES 9 kl 9.8.2 SYSTEM DESCRIPTION AND EVA WATION.

9 h1 10 STEAM AND PCWER CONVERSION SYSTEM Volume 2.

. Tab 10.

10-1 10.1 DESIGN BASES 10-1 10.1.1 OPERATING AND PERFORMANCE REQUIREMENTS 10-1 10.1.2 ELECTRICE SYSTEM CHARACTERISTICS.

10-1 10.1.3 WNCTIONAL LIMITATIONS 10-1 lO.l.h SECONDARY WNCTIONS 10-1 10.2 SYSTEM DESIGN AND OPERATICN 10-2 10.2.1 SCHEMAU C FLOW DIAGRAM 10-2 10.2.2 CODES AND STANDARDS 10-2 10.2 3 DESIGN FEATURES 10-3 10.2.k SHIELDING 10-3 10.2 5 CORRCSION PROTECTION 10-3 10.2.6 IMPURIMES CONTROL.

10-3 10.2.7 RADICACH VITY 10-3 10.3 SYSTEM ANALYSIS 10-3 10 3.1 TRIPS, AUTCMATIC CCNTROL ACTIONS,AND ALARMS 10-3 10 3.2

"'RANSIENT CONDITIONS 10 k 10.3 3 MEWNCHCNS 10 k 10.3.h OVERPRESSURE PROTECTICN 10-5 10.3.5 INTERAC"' IONS 10-5 10.3.6 OPERAn0N E uxITS.

10-3 10.k TESTS AND INSPECTIONS 10-5 i

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I 1

Section M

11 RADIOACTIVE WASTES AND RADIATION PROTECTION - _ _ -

Volume 2.

. Tab 11.

11-1 11.1 RADIOACTIVE WASTES 11-1 11.1.1 DESIGN BASES.

11-1 n.l.2 SYSTEM DESIGN 11-3 11.1 3 TESTS AND INSPECTIONS.

11-11 n.2 RADIATION SHIELDING 11-11 11.2.1 PRIMARY, SECONDARY, REACTOR BUILDING, AND AUXILIARY SEELDING ll-n 11.2.2 AREA RADIATION MONITORING SYSTEM 11-16 n.2 3 HEETH PHYSICS. '.

H-17 11 3 REFERENCES H-21 12 CONDUCT OF OPERATICNS Volume 2.

. Tab 12.

12-1 12.1 ORGANIZATION AND RESPONSIBILITY 12-1 12.1.1 FUNCTICNAL DESCRIPTION 12-1 12.1.2 QUEIFICATIONS 12-2 12.1.3 ORGANIZATION DIAGRAM 12-2 12.2 TRAINING 12-2 12.2.1 STATION STAFF 12-2 12.2.2 REPLACEMENT PERSONNEL.

12-5 12.2.3 CN-THE-JOB TRAINING 12-5 12.2.h ENERG2NCY DRILLS 12-6 12 3 WRITTEN PROCEDURES 12-6 12.4 rec 0RDS 12-6 12 5 ADMINISTRATIVE CONTROL 12-6 13 INITIAL TESTS AND OPERATION Volume.2.

. Tab 13.

13-1 13 1 TESTS 11IOR TO REACTOR MJELING 13-1 13 2 INITIAL CRITICEITY 13-1 13 3 POSTCRITICALITY TESTS 13-1 lk SAFETY ANALYSIS Volume 2.

. Tab 14 1k-1 1k.1 CORE AND COOLANT BOUNDARY PROTECTION ANALYSIS 14-1 1k.1.1 ABNORMALITIES 14-1 1k.1.2 WALYSIS OF EFFECTS AND CCNSEQUENCES.

1k-3 1k.2 STANDBY SAFEGUARDS ANALYSIS 14-19 14.2.1 SITUATIONS ANALYZED AND CAUSES.

14-19 14.2.2 ACCIDENT ANALYSES 14-20

14.3 REFERENCES

14-56 15 TECHNICAL SPECIFICATIONS Volume 2.

. Tab 15 15-1 0

0004 047 Y1

TABLE OF APPENDICES J

Accendix 1A TECENICAL QUALIFICATIONS.

. Vol=e 3... Tab 1A 2A ENGIN E NG GEOLOGY AND FOUNDATION CONSIDEPATIONS.

. Volu=e 3... Tab 2A 23 SEISMOLOGY AND METEOROLCGY..

. Volu=e 3... Tab 23 2C GROUND '4ATER HYDROLOGY.

. Volume 3... Tab 2C 2D GEOLOGY

. Volume 3... Tab 2D SA STRUCTURAL DESIGN 3ASES........ Volume 3... Tab SA 53 DESIGN PROGRAM FOR REAC'"0R SUILDING.. Volu=e 3... Tab 53 5C DESIGN CRITERIA FOR REACTOR SUILDING. Volu=e 3... Tab SC 5D QUALITY. CONTROL.

. Volume 3... Tab SD 5E LTTER PLATE SPECIFICATION

. Volu=e 3... Tab 5E 5F REACTOR SUILDING INSTRUMENTATION.

. Volume 3... Tab 5F Surnle=ent 1.

.. Volume k.. Supple =ent No. 1 2.

. Volume k.. Supple =ent No. 2 O

3.................... Volume 5.. S upple=e nt No. 3 4............

.. Volume 5.. Supple =ent No. h 5.................... Volume 5.. Supple =ent No. 5 t

0004 048 O

vii (Revised 6-23-63) 1

Dockst 50-289 Supplement No. 3 December 8, 1967 13.0 GFJERAL QUESTION Provide an outline of the emergency plans to be followed 13.1 in case of a major accident at the facility.

ANSWER I.

Introduction The Three Mile Island Nuclear Station vill have a Radiation Emergency Plan that vill be employed to minimize radiation exposure to inplant and outside personnel.

It vin describe action levels and specific duties required of station person-nel in the event of an accident or any unplanned incident producing high radiation levels.

Basis for the Radiation Emergency Plan vill be the recom-mendatiot.J of the Federal Radiation Council Reports #5 and #7 The ulti= ate responsibility for the coordination of duties for Radiation Emergency Plan vill be charged to the Radiation Protection Supervisor. Periodic drills win be held to insure that all station personnel have a working knowledge of action required of them by the Radiation Emergency Plan. Each individuel assigned to Three Mile Island vill have a written copy of his specific duty to be performed during the radia-tion emergency. Prior to receipt of an operating license an Emergency Planning Group vill be assigned to meet with state and local authorities to establish a clear line of action and responsibility during a radiation emergency.

Coordination vill also be established with the appropriate welfare groups to assure that temporary quarters etc. can be made available for an accident evacuation pericd if required.

The Health Physics Laboratory vill be located in the Control Tower and vill be utilized to manage the radiation accident.

This laboratory win be equipped with all the necessary instru-mentation and equipment needed to handle a radiation emergency.

In addition, emergency monitoring kits win be placed throughout the station at predetermined locations. These kits vill be inspected and used at periodic drills, which vin be conducted in order to =aintain the effectiveness of the plan.

Training programs vill be established and executed to insure that all selected personnel assigned to emergency monitoring squads have a working knowledge of Health Physics procedures and use of radiation instru=ents.

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II. Action Levels for Initiating the Radiation Emergency Plan

^

Initiation of any of tha following alarms vill alert the Control Room Operator to the possible existence of an accident in which the release of radioactive material could occur to the environ-meat.

1.

Unplanned, automatic initiation of the emergency injection safeguards system.

2.

Loss of primary coolant pressure.

3.

Reactor trip alarm.

4.

Alam of raactor building dome radiation detector.

5 Alam of reactor building air =enitor.

6.

Alarm of spent fuel building dose rate detector.

7.

Alam of spent fuel building air monitor.

8.

High readings on other radiation monitoring instru-mentation.

After receiving one or more of the above alarms, the operator l

vill evaluate the conditions producing the alarms, and if g

instrumentation indicates that radiation has resulted which presents a danger to plant personnel or the surrounding popu-i lation, he vill sound the Radiation Emergency Plan Alarm. This alarm vill have a visual and audible indication in the control The audible signal vill be unique and distinguishable room.

and vill sound throughout the station and surrounding site area.

l III. [tation High Radiation Evacuation Upon the sounding of the station radiation emergency plan alarm, all station employees and visitors except designated l

key personnel and shift workers vill be evacuated from the site.

l All evacuees vill be surveyed by an Emergency Radiation l

Monitor prior to leaving the plant boundary. All personnel remaining in the plant to perform emergency operating procedures vill be surveyed for possible contamination.

To assure that key personnel can be contacted at all times, a

" Call List" will be established.

IV. Evaluation of On-Site Cenditions To eval th ha ation levels on-site, the Niloving procedure vill be used:

O 1.

Readings vill be taken from the station radiation monitoring system in the control room.

0004 0S0 13.1-2

2.

A member of the emergency monitoring team vill take air samples, beta-gn=ma swipe, and dose rate readings throughout the plent and immediate environment.

3 These survey results vill be recorded in a log book at the Health Physics Laboratory.

V.

Evaluation of Off-Site Conditions To evaluate the radiation levels off-site, the following pro-cedure vill be used:

1.

Members of the emergency monitoring team will take air samples, beta-gamma swipe, and dose rate readings at specific areas determined by the Radiation Protection Supervisor. The area surveyed depends on vind direction and.vind velocity as indicated by the Three Mile Island weather station.

\\, ' ' i Vi.' $ctective Action for Local Penulation If the emergency monitoring squads have reported high radiation levels on the off-site survey and these radiation levels are above the average projected dose limit recommended by the Federal O

Radiation Council, and radiation levels are increasing, the Radiation Protection Supervisor or his alternate vill call the appropriate authorities which will be designated by state and local governments, and the AEC New York Operations office.

These organizations vill be informed that a maj6r release of radioactive material occurred at Three Mile Island; evacuation of the low population zone vill then be performed by pre-arranged nthods by personnel notified in the preceding para-graph.

VII. Hoseitalization of Contaminated Casualties In the event a medical injury occurrs that necessitates hospitali-zation before decontamination, the folleving action vill be taken:

1.

A survey of the contaminated person vill be performed by a member of the emergency monitoring squad.

2.

The Radiatica Protection Supervisor or Shift Foreman vill alert the proper hospital authorities and advise them of the radiation levels of the contaminated person.

(Special arrangements vill be made with a designated local hospital on a radiation emergency plan.)

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0004 051 t

13.1-3

VIII. Emergency Reserve of Personnel If additional personnel are required in the event of an emergency, they vill be obtained from the folleving sources:

1.

Within the company and/or other operating nuclear power plants

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2.

GPU Nuclear Staff l

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AEC t

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Pennsylvania Public Health Department t

5 Outside Health Physics and Nuclear Decontamination vendors l

0004 052 O

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Dockst 50-289 Supplement No. 3 December 8, 1967

{

QUESTION Provide a parametric study of various = asses, energies 13.2 and i= pact areas which might result from a breakuu at overspeed of the last stage wheel of the turbine to show that the wheel section chosen is indeed the vorst missile.

ANS'ER The turbine-generator manufacturer has =ade a study of the different si:es of last stage wheel missiles ~that could be generated. The latest analysis indicates that the last stage wheel could fail at an overspeed of 169 percent of the rated speed. Properties and penetration into the Reactor Building for the last stage wheel missiles are shown in table 13.2.

The naximum penetration vill occur with the 120 degree segment as shown in Figure 13.2-1.

A comparison of the penetration depths of the 120 degree segment at a 69 percent overspeed, as stated above, and the 13h degree seg=ent at an 86 percent overspeed (as shown in section 5.1.2.T.2 of the PSAR) indicates that the difference in penetration depths is negligible and that the difference in percent of overspeed does not appreciably influence the penetration depth. For example, penetration in concrete is 12.8 in. at 69 percent overspeed as opposed to 13.2 in. at 86 percent overspeed.

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O TABLE 13.2-1 LAST STAGE WHEEL MISSILES 2

FRAGMENT WElGHT lMPACT AREA (FT )

FINAL FINAL DEPTH OF PENETRAT ANGLE (POUNDS)

SIDE CN END ON ENERGY VELOCITY SIDE ON END C 90*

4458 6.83 3.17 464.0 5.45" 11.8' FT,9 120*

5944 8.37 3.66 447.3 5.6 "

12.8' FT-p 180*

8916 9.66 4.83 351.0 5.04" 10.1' FT-#

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MISSILE PENETRATIONS FIGURE 13.21

Dockst 50-289 Supplement No. 3 December 8, 1967 QUESTION Provide a brief ' description on vendor qualification on 13.3 the vaste evaporators that vill assure a lok decentamina-tion factor.

ANSWER The analysis in Section 11 of the PSAR assumes a decontamin-ation factor of 10k inthevasteevaporgtors.

In practice, decontamination factors greater than 10 have been experien-ced and have been reported in the references given below.

The test results generally indicate that suitably designed evaporators with efficient demisting units vill achieve decontamination factors on the order of 106 Vendors who vill be invited to bid on this equipment include American Machine and Foundry, and Aquachem who have supplied most commercial vaste evaporator units to date.

These companies, and any others who are judged as qualified to bid, vill be required to submit evidence, as part of their bid, of test results an,d operating experience which vill demonstrate the required decontamination efficiency.

References (1) Saxton Op-ti' ; Experience, AEC Nuclear Safety Revies Spring, 196h (2) PM-1 Operating Experience AMF Report May, '.967 (3) Evaporation of Wastes at Brookhaven National Laboratory

" Chemical Engineering" March, 1955 (h) Radi: active Waste Concentration Griscem-Russell Co.

Report RT-19-56 October, 1956 0004 055 l

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Docket 50-289 Supplement No. 3 December 8. 1967 1h.0 SITE QUESTION Discuss the design of the dikes and intake structure for protection ik.1 against erosion during flood conditionc. Can " ice jams" occur on j

the Susquehanna, and if so what forces are considered in the design of the intake structure?

ANSWER The dikes are protected by a layer of dumped riprap of sufficient size and thickness, and with an adequate zone of sand and gravel embedmenc material, to withstand wave action of 2-1/k feet, and o velocity in excess of 12.0 feet per second, on a 2 on 1 slope.

The riprap was sized according to the standards of the Tennessee Valley Authority for riprap on earth dams and in powerhouse and spillway tailraces. The riprap will continue downward into natural ground for a minimum depth of two feet to provide a cut-off 1

against undermining.

j The intake structure vill have reinforced concrete ving valls, based on competent rock and with concrete cut-off valls to resist any undercutting. Each wall vill be designed so as to be fully stable under all conditions of loading, including that of any

" ice jams." The banks upstream and downstream vill be riprapped in accordance with the preceeding paragraph, to afford protection 1

against erosion during flood.

The Susquehanna River is susceptable to " ice jams;" however, these occur at floods of relative lov magnitude, since a flood flow in excess of 300,000 efs vill vash the ice downstream. The worst

" ice jam" in history occurred in 1904. The first jamming action occurred at river mile 60.8 in the vicinity of Sassafras Island.

As the river continued to rise the jam was swept dovestream where it jammed on the York Haven dam and headrace vall, and in the gorge between the York Haven headrace vall and the Red Hill tower.

i As the river continued to rise, this was swept downstream. No ja= ming of ice was experienced in the Three Mile Island reach. The flood peak of that year was 298,000 efs.

)

In recent years, the low-magnitude ficed " ice jams" have occurred between the upstream tip of Sand Beach Island and the east shore.

No " ice jams" have occurred at the location of the pr'oposed inteke structure. Without regard to jamming, the entire river surface can at times be covered with jumbled moving ice flows. The design leading of the intake vill be such as to take into account the following conditions:

0004 056 1.

Full thrust as developed by the maximum expected thickness (approximately 3 ft. ) of sheet ice, with the top surface elevation corresponding to the crest of the York Haven Dam.

2.

Hydrostatic forces caused by any possible down stream " ice jam."

considering the maximum vater rise of record at the site as the water level of the 1936 flood at the plant intake.

3.

Hydredynamic forces exerted by a =aximum river velocity, i= pinging against a major " ice jas" cccurring at the intake.

The loading conditions for the " Design Flood" vill be higher than any of the preceeding design items for icing conditions.

j 1h.1-1

Docket 50-289 Supplement No. 3 December 8, 1967

~,

O QUESTION The flow model of the river from which the flood protection level 14.2 vere established is not acceptable. The rationale for using such j

a program is that if a profile of a known flood can be matched by a computed profile over a significant reach of the river the para meters used in the calculation can be used in the computation of higher floods. Since the computed profile grossly conflicts with available data, the extension of these computations to a larger flood is not applicable. A recomputation in which a longer reach (perhaps 5 to 7 miles) is taken into consideration might produce acceptable results. Additional river cross sections might be required to justify the final results.

ANSWER The studies of the 1936, 1964, and design floods on the Susquehan River between York Haven and Middletown vere reviewed, using as a basis the 1936 flood. All known high water marks were utilized j

in the program, and a super-elevation of the water surface toward i

the east back of the river was taken into account for flow around f

the bend at Hill Island.

The 196k flood was recalculated based upon the same criteria and procedure as the 1936 flood. All known high-water marks were considered in the program, and a good correlation was achieved, v confirmed the rationale utilized for analyzing the 1936 flood.

The design flood was then recalculated on the same basis as the two preceeding floods. The top elevation of the protective dike at the tip of Three Mile Island has been raised to 310, providing a freeboard of approximately six feet above the design flood at this location. The dikes along both sides of the island descend uniformly from elevation 310 to elevation 305, extending over a distance of 2,h00 and 3.M0 feet on the east and vest sides, re-spectively. The freeboard at the intake structure is approximate three feet. Provision has been made for a cut-off at elevation 30k extending across the downstream end of the plant site.

The answer to Question number 2.3, Supplement No. 1, has been revised to incorporate the final results of the study.

The permanent access bridge extends from the east shore a distanc of 1600 feet to Three Mile Island, crossing the east channel and Sand Beach Island. The bridge is a combination rail-roadway stru l

vith concrete slab and steel girders supported on 16 concrete pie and end abutments. The elevations of the deck slab are 306.3 at east end of the bridge, 313.2 near the center, and 310 at the ves end. Throughout the main spans of the bridge the bottom of the supporting steel girders is approximately 9 feet below the deck.

The bridge has been designed to withstand the forces of the desig flood. The deck is high enough to ensure dry passage over the br O

during the design flood. The backwater effect of the bridge duri the design flood is esti=ated to be in the order of 0.h feet.

0004 057 1h.2-1 (Revised 12-22-67) l

Dockst 50-289 Supplement No. 3 Decamber 8, 1967 O

15.o srauctuait oss'ou QUESTION Clari.^/ the statement made in response to question 7.2 15.1 which indicates that horizontal and vertical " frequencies" vill be added.

Indicate that the appropriate stresses will be added directly and linearly in all cases.

ANSWER The respective vertical and horizontal seismic components at any point on the shell vill be added by summing the absolute values of the response (i.e., stress, shear, moment, or deflection) of each contributing frequency due to vertical motion and adding the resultants to the corresponding absolute values of the response of each contributing frequency due to horizontal motion.

0004 058 O

oo 15.1-1

O Docket 50-289 V

Supplement No. 3 Dece=ber 8, 1967 QUESTION We understand that calculations of the response of the pri=ary sy 15.2 te= internals to si=ultaneous earthquake and accident leads are being perfor=ed in response to our previous Questien 3.3.

Outlin

'the scope of the calculations' and the technique to be used.

Pro-vide a schedule on which the preliminary and detailed calculation vill be performed and the date when the stress and defor=ation li its and leading ec=binatiens which you censider appropriate vill deter =ined.

Provide any results which you have obtained to date.

ANS'ER Secte of Calculations and Techniones to be E= cloyed All reactor internals and core ec=penents (including centrol rods vill be analyced separately for stresses and deflections resultin.

frc= accidents and earthquakes.

Static and/or dynamic analyses vill be e= ployed, as appropriate.

In general, dynamic analysis vill be used for the sub-cooled port.

of the LOCA, and earthquakes, and static analysis vill be used fo:

the relatively steady state portion of the LOCA.

Dyna =le analysis vill include the response of the entire syste: (1 applicable in each case) to the various excitations. For LOCA, t1 excitatica vill be applied at the appropriate no::le or internals cc=penent. Where appropriate, the response of the reactor vessel en its support skirt vill be used as input to the internals. The response of the internals vill then be used as input to the core.

Seismic excitation vill be handled in a similar manner, except the the ground motion vill be input at the junction of the support sk and the vessel foundation.

Lu= ped parameter simulatien vill be ut generally.

For LOCA, predicted pressure-ti=e histories vill be used as input For earthquake, actual earthquake records, nor=alized to the appr:

priate ground motion, vill be used as input. Output vill be in ti for of internal's motions (displacements, velocities and accelert tions), =otions of individual fuel asse=blies, impact leads betwee adjacent fuel assemblies and impact loads between peripheral fuel asse=blies and the core shroud.

Seis=ic analysis vill also be performed using the response spectra approach.

" p' -

3 The relative timing of the various aspects of a given LOCA vill be

'censidered only as indicated by the various local time histories associated with that particular accident, although sensitivity to the time duration of the pulses and other calculated input vill be investigated.

l u

Where simulta: acus occurrence of LOCA and the MHE is censidered, !

is intended that both excitations vill be input to the syste:

0004 059 23.2_1

simultaneously. Relative starting times vill be changed until maxi-h

=um structural =otions, indicating =aximum stresses, are obtained.

Alternately, the maxi =um stress from one ce=penent of the combina-tien vill be added to the square root of the sum of the squares of the other ec=ponents.

Schedule for Providinz Deformation Limits and Lead Cc=binations Defor=ation limits on internals and core vill be provided by mid-January, 1968. These " safety limits" vill be the maximum deforma-tions which can be permitted to occur without cc=prc=ising a safe and orderly shutdown. We vill describe the =cdes of deformation, the type of loading which causes the-deformation, the safety impli-cations, the limiting value and the permissible value.

For example, for the lover grid plate this is expected to ta.ke the form:

Cc=cenent:

Lover Grid Plate Mode:

First bending mode, devnvard (dished devn-vard)

Lead:

Devnvard pressure drop due to cold leg rup-ture Safety I=clication:

F.xcessive grid plate deflection =ay cause

)

unacceptable defo: ation of the 0.tel asse=bly guide tubes and/or disengage =ent of the fuel asse=blies frem the upper grid, resulting in unacceptable resista=ce to centrol red

=otion.

Limitin: Value:

at g-id center Allevable Value:

5 of limiting value A table of lead ce=binations vill also be provided in January,1968.

It is expected that a Wvdevn leads, seistic accelerations, te:-

perature gradients, and other needed input vill be available by July 1966, and stress analysis Of the inter.als vill be ec=pleted no la* =~ "a-July 1969 i

i l

0004 060 l

9 l

i hI k -

-:.=-=

. - =

Dockst 50-289 Supplement No. 3 December 8, 1967 O

QUISTION Clarify your response to question 3.2 to indi.cate the provisions 15.3 which vill be made to insure that cranes cannot be displaced from the track.

ANSWIR See revised answer to Question 3.2.h.

I

~

0004 061 4

J O

f s

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J l O i

i 15.3-1 l

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

Dockst 50-289 Supplement No. 3 December 8, 1967 QUESTION With regard to the location of vital structures on dissimilar 15.h foundation materials provide (1) a detailed description of the transient analysis to be perfor=ed to determine the stresses and deflections of critical piping between buildings and speciff the criteria for stresses and deflections, (2) discuss provisions to be made to prevent physical interaction between buildings due to different seismic response or settlement and tilting, and (3) the criterion for allevable settlement or tilting of the turbine building and the method used to calculate this phenomenon.

ANSWER Dynamie Piping Analysis The dynamic analysis of critical piping syste=s (i.e., Class I systems) vill be a modal analysis based upon either a distributed or lumped-mass solution depending upon the complexity of the system. The two approaches are performed as follows:

a.

Distributed-Mass Analysis The system is represented by a number of straight uniform beams with a distributed mass and stiffness. First, the transfer matrix for each of the straight beams is determined and the rotation transfer = atrix for each joint calculated.

Next, the equation of =otion is written in matrix form.

Previously determined transfer matrices are used. Considering the appropriate boundary conditions, the characteristic determinant is generated. When the natural frequencies are known, the corresponding mode shapes are determined. Then, by using the response spectrum for a single-degree-of-freedom system, the =ax1=um displacements are obtained as the root-mean-square sum of the modal maxis a.

Finally, after the maximum displacements are known, forces and moments are calculated at the structural joints.

b.

Lumped-Mass Analysis The system is represented by a series of concentrated masses.

First, the space coordinates are established for the system and coordinates of mass points are determined. Using a static analysis, flexibility =atrices corresponding to these = ass points are computed. Next, the equations of motion are writtet in matrix for=.

Force influence coefficients method is used.

Natural frequencies and mode shapes are obtained assuming harmonic motion of the system. Finally, using the same tech-nique as for the distributed-mass analysis, =axi=un internal forces and cements are calculated at the structural joints.

In addition to the earthquake response for the pipe system, the modals described above vill be used to deter =ine forces and coments with resulting stresses for any transient or per=anent displacements which vill be induced at the support points.

000.4 n

A 6 lHCH LINE PRIMARY COOLANT 8 N.G. OPEN CLO5ED CLOSED YE5 COLD 305 B lb lHCal LlHE LET DowH.MU & F 5YST. 9 N.G. P OPEN CLOSED CLOSED YE5 COLD O OZ SE AL RETURN WATER 10 H.G. P OPEN CLO5LD CLOSED YES COLD .g g 17 2-1-B B 4 INCH LINE MU & P $YST. II H.G. P OPEN CLOSED CLOSED YE5 COLD -4 Z OEm E ZC

• IQ R.C..' UMP I

ca gU 307 SEAL SUPPLY WATER 17.2-l A 12 5.w.G. P OPEH CLOSED CLOSED YES COL D A 4 8HCH LlHE F Cu MU & P SYST. m ~~ Eb O 4 as O N 310 17.21 A 13 5 w.G. E CLOSED CLOSED CLO5ED YES HOT A IC INCH LlHE w TE MW L E GE ND: r M -4 I O 30 VALVfk VALVE OPERATORh NORMAL TEMP.a l.W.G. a SPLli WE DGE CATE VALVE A = AIR OPERATED; OPEN & CLO5ED COL D a

• 254*F

-4 l l l l 2' PENETR ATION & ISOLATION TYPE YALVE PO58T M PO58T M POSIT M POSITION POilTW HORMAL --4 Z PENETR ASSOCI AT ED FIGURE VALVE OF OPERATOR DURING AFTER AFTER FOR INDICATION TEMP. REMARK 5 OE NO. SYSTEM NO. VALYE TYPF. NORMAL POWER LOCA LEAK RATE lH OF e=g OPERATION F AILUR E SIGNAL TEST CONT. ROOM CONTENT 5 so 4_. 3 19 R C 5 T' 47.2 A 14 5.t f.G. A.O. OPEN CLO5ED CLOSED YE5 COLD A 6 MCH LlHE N 320 87.21-A 15 5.w.G. P OPEN CLOSED CLOSED YES COLD A 6 NCH LlHE R 55 y r-u ? 3 ~O 328 I?.21-A le 5.w.G. A.O. OPEN CLOSED CLOSED YES COLD A 31NCH LlHE R 5TE YST mH-O43 + R.C. ORAN T ANK VENT 322 TO RAD. WA5TE SYST. 87.2 A 17 5.w.G. A.O. OPEN CLOSED CLO5ED YE5 COLD A 3 INCH LlHE 0 Ny SUPPLY TO R.C. la N A.O. OPEN CLOSED CLOSED YES COLD h 3 8 I NCH LME DRAIN T ANK 17.2-1-8 19 H OPEN CLOSED CLOSED YES COLD 0 326 whT UPP' Y 17.2 3-A 20 5.W.G. A.O. OPEN CLOSED CLOSED YE5 COLD A 6 INCH LINE N 327 IF.2 A 21 5.w.G. A.O. OPEN CLOSED CLOSED YES COLD A 6 8HCH L8HE WATE PLY COOLMG WATER SUPPLY 332 REACTOR COOLANT PUMPL 17.2 A 22 5.w.G. A.O. OPEN CLOSED CLOSED YES COLD A 6 INCH LINE f0R C ^ 333 17.21-A 23 5.w.G. A.O. OPEN CLOSED CLOSED YE5 COLD A 6 INCH LlHE P DRAM RL DRAM 24 H.C. A.O. OPEN CLOSED CLOSED YE5 COLD 337 17.2-1-8 8 21HCH LlHE TANK 25 H.G. A.O. OPEN CLOSED CLO5ED YES COLD' MAKE-UP WATER SUPPLY 26 N.G. A.O. OPEN CLOSED CLOSED YES COLD 33 R.C. DRAIN TANK II 2'I'0 27 H.G. A.O. OPEN CLOSED CLOSED YE5 COLD

1. I #^

2R H.G. E OPEN CLOSED CLOSED YE5 COLD 353 C.R.D'. SE ALS ~ 17.2 1-8 8 15 INCH LlHE 29 N.G. E OPEN CLOSED CLOSED YE5 COLD 437(A) PRE 55URIZER SAMPLE 17.2 1-8 30 d E OPEN CLOSED CLOSED YES HOT S % INCH LlHE 31 N E OPEN CLOSED CLOSED YES HOT RE ACTOR COOLANT 32 H E OPEN CLOSED CLOSED YES HOT d IE0I I* O B % M H l.mE SAMPLE 33 '4 E OPEN CLOSED CLOSED YES HOT STEAM GENERATOR A 17.2-l.8 B M lHC't LlHE O 5 AMPLE 35 N E OPEN CLOSED CLOSED YES HOT C STEAM GENERATOR 8 36 H E OPEN CLOSED CLOSED YES HOT C m e) 17.2+8 8 % lHCH LINE 3,,ptg 37 N E OPEN CL0$ED CLOSED YES HOT 4 LEGEND: Q VALVt5: VALVE OPER ATOR14 HodMAL TEMP.4 ( 5.W.G.

• SPLIT WEDGE GATE VALVE A a AIR OPERATED 6 OPEN & CLOSED COLD a (250*F

[ N.G. m SOLIO WEDGE GATE V ALVE AO a AtR OPER ATED; AIR TO OPEN, SPRlNG TO CLOSE NOT a > 250*F N GL.

• GLOSE VALVE AC a A4R OPERATED 4 AIR TO CLO5E, SPRING TO OPEN REMARK 5e N

a NEEDLE VALVE P a Pl5 TON OPERATED l A a SINGt.E 450LATION VALVE FLut0 SLOCK R a AtliTFRF6 Y vat VF F = F9 Frf. OPFR ATFD fi laalTOROUF TYPF OPFR ATORt t 9 O 9 O WATER 3 ) COMPRESSEC SUPPLY 4 4 GASSUPPLY 3Z XI i V 1 l l 1 1 ~ I ~ l l l SEAL WATER L dh SUPPLY TANK G J AS50 PLY WATER SUPPLY -5EAL WATER SUPPLY HEADER TEST N.C. N.C. TANK TEST FEEDER 2 SPLIT WEDGE GATE VALVE J CONNECTIONS % LINE N.C. (/ FOR FLOWMETER C O y en a w, REMOVABLE v DUTCHMAN s g AI R '. rp TO BE USED RB U., DURING TEST SUPPLY g ONLY INTERLOCK C 0 FEEDER LINE N.C. ES AIR SUPPLYbC TYPE A FLUID BLOCK ISOLATION 4V INTERLOCK gg p..Tg ES t i ES NOTE: SYSTEM IN STANDBY C V C O PLANT IN NORMAL OPERATION b REGULAR GATE OR GLOBE RB VALVES IN SERIES 0004 083 TYPE B FLUID BLOCK ISOLATION TYPICAL FLUID sLOOK ISOLA FIGURE 17.21 Dockst 50-289 Supplement No. 3 December 8, 1967 QUESTION We believe that the ability to take samples frem the recirculated 17.3 vater after a loss-of-coolant accident should be maintained to allow determination of the boren content and chemical composition of the liquid. This is par'.icularly important when a chemical iodine removal agent is relied on. Indicate (1) the method by which sampl!ng vill be acccmplished, (2) how soon after an accident the srmple could be taken, (3) whether adequate equipment and qualified rersor.nel vill be available on-site or if the sample vould nave to be transported, and (h) precautions to be taken in sampling to avoid major radioactivity releases. ANSWER 1. Following a LOCA, a sample of recirculated water can be ob-tained locally in the Auxiliary Building throur,h the use of existing facilities. The sample is obtained by draining a quantity of water from a vent / sample valve from the decay heat removal pump discharge. The sample vould be taken in an evacuated sample bowl. Samples of a highly radioactive nature such as the recirculated water vould be transported in a shielded pig. 2. A sa=ple can be obtained in approximately 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> following an accident. 3. Plant personnel vill be trained in the taking of a sample in the manner described above and en-site radio-chem lab (} equipment will be used to analyze the sample. The plant radio-chem lab will be equipped with instrumentation that vill measure boron content, isotope identification and sodium thiosulphate concentration. It vill not be necessary to transport the sample elsewhere for analysis. h. Plant personnel vill be instructed in the sa=pling of radioactive materials and procedures established for their handling and analysis. 0004 084 O 17.3-1 \\ Docket 50-289 Supplement No. 3 December 8,1967 QUESTION Provide a complete,, detailed list of experiments which must be 17.h performed to substantiate the use of a che=ical spray system which is relied on to meet Part 100 site guidelines. You may indicate those experiments which you believe vill be carried out by other programs but you should indicate clearly your respon-sibility for obtaining the necessar/ data in case programs other than your own do not produce conclusive results. The folleving list of areas constitutes a necessary, but not necessarily suf-ficient, list of items which must be resolved by the research and development program effort. (1) Verification that Griffith's model is applicable to an air film around a reactive spray drop over the temperature and pressure range of interest. (2) Determination of the effect of simultaneous water condensation on the iodine removal rate. (3) Verification that Griffith's model is valid over a vide range of initial iodine concentrations and forms. (k) Stability of the solution in the pressure-te=perature-r:.diatio: environment over long time periods. (5) verification of scaling assumptions. (6) Projection of efforts if (1) Griffith's model and/or (2) the sodium thiosulfate reagent must be abandoned. (7) Simulation of injection system. ANSWER The design of the Chemical Spray System for Iodine Removal is based on sound engineering principles and a large margin of conservatism is deliberately built into this system. Information generated by R&D programs related to this system indicates that the system vill function as an effective engineered safeguard for iodine removal following an ?EA. As indicated in answer to Question 5.13, Supplement No. 1, there is valid reason to assume that the solution does function as an effective system for iodine removal. However, the ansvers also indicated that some R&D is desirable to provide more detail in order to improve the system and to verify that the system function: effectively under post-accident conditions. The specific objective and scope of these R&D progra=s are described in Section A below. The above sub-questions are answered in detail in Section B. O 0004 085 mn 17.h-1 ( h A. R&D Programs A.1. Objectives The objectives of these programs as indicated in ceply to Question 5.13, Supplement No. 1, are to verify that the solution: 1. Does not exhibit chemical or physical changes which are detri-mental to the solution's effectiveness or which produce exces-sive amounts of undesirable decomposition products, when it is exposed to temperature, pressures and irradiation doses associ. ated with the post-accident conditions. 2. Does not risult in significant chemical attack or corrosion of the primary construction materials (Zirealoy, Inconel, stainle-steel, carbon steel, paint and concrete) during the post-acci-dent period. 3. Rapidly reduces iodine concentration to tolerable levels and retains it under post-accident humidity, temperature and pres-sure conditions. h. Is compatible with soluble boren ecmpounds. A.2 Scope i B&W, ORNL, and others have R&D programs ubich vill provide the necessary data to satisfy the above objectives. B&W will have this data available in time to be incorporated into the system design and to meet requirements for an operating license. R&D data from the experiments discussed belov vill be adequate to confirm that (1) the solution would not react significantly with any materials proposed for surfaces of the containment (reactor building), the reactor or the recirculation system (5.13.1); (2) the solution is compatible with boric acid and satisfies the sta-bility requirements over long periods in the accident environment (513.2); (3) the system effectively removes radiciodine from the reactor building environment (5.13.3); and (h) an adequate A, is obtained in the post-accident environment (5 13.h). A.2.1 Radiation Stability These experi=ents are designed to provide the data which vill de=enstrate the effects of radiation on the solution in the post-accident environment. Sealed a=poules containing a solution in a liquid-air vol-ume ratio which approxi=ates that in the reactor building vill be irradiated at temperatures which correspond to g l 3 q,' 9.g'j' ;', those experienced by the bulk solution under pset-accident w l i.' conditions. A calibrated spent fuel element vill be used to irradiate the solutien to integrated doses up to 0004 086 l 17.h-2 1 N O 8 =aes, which.1s the maximu= teta1 eose received b 3 x 10 7 the bulk solution from all sources over the post-accident period. The dose rate of the experiments approximates the average dose rate to the bulk solution during the accident, but some data vill be collected at a higher dose rate to establish the effect of the dose rate. The experiments are scoped to provide data on decomposition rate of solution, the chemical nature of the major decomposi tien products, the iodine removal capacity and iodine reten-tion of the irradiated solution. A.2.2 Thermal Stability These experi=ents are designed to provide the data which vill demonstrate the ther-1 effects resulting from contact between the solution and heat transfer surfaces under the vorst expected post-accident thermal conditions. They are adequate to confirm the thermal stability of the solution, since they vould reveal changes which are detrimental to the solution's effectiveness or which produce excessive amounts of undesirable decomposition products. i The solution vill be exposed to simulated fuel pins which approximate the hot spot heat flux in the core as a function of time. The bulk liquid temperature is controlled by thm { cooling of the apparatus at a rate which reproduces a con-servatively simulated post-accident pressure vs. time rela-tionship. It vill also =easure the evolution of gaseous decomposition products and changes in the iodine absorption capacity. A.2.3 Corrosion and chemical Attack These experiments vill determine the ec=patibility of the ] solution with the prhry construction materials. Specimens of stainless steel, Zirealoy, Inconel, carbon steel liner, concrete and specimens of the liner paint and the seal coating applied to the inside of the raactor build-ing vill be checked for chemical attack and corrosion by the solution. The tests vill be conducted under temperature vs. time conditions which correspond to those of the post-accident period. A.2.h Iodine Pe= oval Th w e teste experimentally measure the rate at which chemi-cal sprays remove iodine frem a steam-air atmosphere under o post-accident te=peratures and pressures. They are pres-C ently being perfoemed in the US?? facility at CRNL. '.dden-du= 1 is an outline of these tests which vill verify he spray system's iodine removal capability under pest-accident m- .u<.s, 0004 087 m 17.h-3 I a. conditions and the applicability of Grif fith's mode. Pre-liminary results are available from these tests. (2 l A.2 5 Boron Comratibility 1 All solutions used in any of the a' cove experiments contain 6 l boric acid; therefore its compatibility with the iodine removal reactant is demonstrated in each experiment. A.3 Program Evaluation B&W will perform the R&D outlined above except for the spray tests on iodine removal which are an integral part of ORNL's Spray Technology Program. The ORNL Spray Technology Program has already produced data from spray tests of the type listed above. This program also includes some experiments similar to those included in.B&W's R&D Program, such as irradiation tests at ele-vated temperatures and pressures. The complementary nature of these programs should improve the statistical confidence of the experiments or results and lead to engineering i=provement of the Chemical Spray System. l l B&W, ORNL and others are presently and have been generating very significant quantities of data which aid in the evaluation of the Chemical Spray System. Results from these experiments indicate that the Chemicel Spray System vill perform satisfactorily and ) that radiation does not produce changes in alkaline thiosulfate solution which are htrimental to the solution's effectiveness or which otherwise make it unsuitable. It is expected that the experiments listed above vill confirm this indication. We believe the experiments outlined above are adequate to verify that the Chemical Spray System functions effectively under post-accident conditions. With the exception of the Iodine Removal tests indicated in A.2.k, the program is being carried out by B&W. In the event that ORNL does not complete the iodine removal tests, sufficient tests vill be performed by B&W to verify the effectiveness of iodine removal for this plant. B. Specific Answers to Sub-Questions Bl.1 Applicab'ility of Griffith's Model i l The basic concept of the chemical spray system has been experimen-l tally demonstrated. The system effectively removes radioicdine fr l the at=osphere under post accident conditions (2). Experimental i data is in good agree =ent with Griffith's model which is based on I accepted theory for =oment'.:=, heat and mass transfer from a contin phase to a boundary system, and heat-mass momentum trnasfer analog provides a basis for adapting it to the spray system. The gas pha l transfer of iodine into the reactive spray has been experimentally de=onstrated man-/ ti=es. (3) (h) Thus, Griffith's medel is based l i a good mass transfer correlatien and a sound ens;ineering principle Alg% gh experimental data to substantiate its adaptability to the hou 1, +" 0004 088 x 17.h h ) Chemical Spray System under post-accident conditions is limited, all present data shows good agreement with calculated values base on Griffith's model. For example, an experiment ORNL in the Containment Mockup Facility (CMF) (1)perfor=ed at gave excellent agreement with the calculated value. Recent chemical spray tests in NSPP (2) conducted in steam-air environ =ent at elevated tem-peratures and pressures are in agreement with calculated values. This series of spray tests is yielding considerable experimental data which verify calculated values. These tests being carried out by ORNL also provide sufficient data to modify Griffith's model to account for effects such as pressure, a steam-air environment, spray drop size distribution, and variation in drop velocity. B.2. Condensation Effects Condensation film formation on the outside of the chemical spray droplet does not appear to be significant. This has been demon-strated by recent tests at NSPP in steam-ai. mixtures at elevated temperatures and pressures. Run 28 was performed with a hot spras solution (2) so that condensation on the spray drop is negligible. Run 2h was performed with a cold spray so that there was conden-sation on the spray drop. Iodine removal in these experiments was similar thus indicating that there is no significant mass transfer barrier formed by a condensation on the spray drop. Additional experiments under conditions which approximate post-accident conditions vill be performed in NSPP (Addenda 1). B.3. Experimental Range of Ccucentrations and Forms The program outlined in Addendum 1 is being run with concentratior and forms of iodine similar to those which might exist as a result of the assumed MF.A in the plant. Thus, the spray tests are adeque for verification of the chemical spray system capability to ruduce radiciodine to a tolerable level and for verification of the anal-ytical model for the conditions of importance in the plant. B.h. Solution Stability As indicated above in A.1 and A.2, B&W's R&D is designed to dem-enstrate the ther=al and radiation stability of the solution under conditions applicable to accident. {T b{ii.!. B.S. Scale-Up i All R&D related to this question is performed under as realistic conditions as experimentally practical. The radiation, thermal and pressure conditions for stability and corrosion tests are O conservatively simulated and exposure times under these conditions are longer than those predicted for the post-accident period. 0004 089 I 17.h-5 l Application of spray test data to conditions in the reactor building can be made with confidence since NSPP is a conservative representation of a full scale building. The average drop veloci* in NSPP is higher than in the reactor building, a smaller fractier of the volume is vasher by sprays in NSPF, and there is little if any wall run off in NSPr. Each of these factors vould lead to a lover relative iodine removal effectiveness in the NSPP than in the reactor building. B.6. Alternate Plans As indicated pre'tiously, the basic concept of Chemical Spray Syste has been experimentally de=enstrated to remove radiciodine under post-accident conditions and calculational models based on Griffit work are appli, cable. The' alkaline sodium thicaulfate solution performs satisfactorily over the range thnt it has been studied. If it becomes desirable to do so alternate solutions may be sub-stituted. Present R&D programs include parallel tests on the alternate chemical solutions and such a change could easily be made. B.T. Simulation of Injection System The ability of '.h2 iodine removal system to inject chemicals as designed vill be demonstrated during the pre start-up test progra: References (1) G. W. Parker, et.al., " Reaction of Molecular Iodine and Methyl Iodine with Sodium Thiosulfate Sprays" ORNL h071 p.18k-191. Dec. 31, 1966. (2) L. Parsley, Jr., "Results of Runs 21 - 28 in NSPP." (Unpublished) (3) L. F. Parsley, Jr., "Bemoval of Elemental Iodine frem Steam-Air-Atmos-phere by Reactive Sprays" CRUL-TM-1911. October 1967 (h) R. F. Taylor " Absorption of Iodine Vapor by Aquecus Solutions" Chemical Engr. Science, 1959, vol. 10 pp 68-79. 0004 090 t l O l f, b b8 ) -I l?.k-6 k O O O ADDENDUM 1 - NSPP EXPERIMENTS Exp. Ter$ative Spray Drop No. Date Source Spray Size,p Conditions Purpose 20 6-20-67 12 I w/o Na2 2 3, 3000 ppm B, 100-200 1 ata Baseline S0

• 10 mg/m3 pli = 5,1/14 gpm 70 F 2]

7-18~67 12 Same as No. 20 100-200 1 c.m Baseline Replica-3 ~ 10 mg/m 7oo F tion 22 8-7-67 I 1 w/o Na2 2 3, 3000 ppm B, 100-200, I atm Spray Solution 80 2 10 mg/m3 0,15 M Na0li, pH =~ 9.3,1/4 70 F Variation gpm 23 9-lh-67 Water, 9.5 gpm 700 his psig Ileat Removal Check 265" F L' 23 ~ 9-20-67 I2 0.15 M NaOH, 3000 ppm B, 700 !414 psig Spray Solution 4 k-100 mg/m3 pil = 9. k. 9. 5 gPm 266 F Variation 14 25 10 le-67 I2 Same as No. 22 @ 9.5 gpm 700 ble psig Drop Size and Tem-3 100 mg/m 266 F perature Effect 26 10-17-67 12 Same as No. 25 700 14 psig Replication 4 3 100 mg/m 266" F 27 10-2h-67 Io Water at 9.5 gpm 700 414 psig Spray Solution o IDO mt/m3 266 F Variation C O 28 10-31-67 I2 Same as No. 21 with spray 700 Ih psig Condensation Film 4 t 100 mg/m3 solution heated to 21a8 F 266 F Effect h before spraying 29 11-13-67 I2 Same as No._2h at 15 gpm 700 1144 psig Spray Flow Rate 100 mg/m3 266 F Effect 30 11-20-67 12 Same as No. 22 at 15 gpm 700 1 atm Drop Size Effect 100 mg/m3 0 70 y 9 ADDt.fwuM 1 - NSPP EXPElt1ME!rls Exp. Tentative Spray Drop flo. Date Source Spray Size,p Conditions, Purpose 31 11-28-67 CII I Feedback from No. 20-30 700 lfe psig Cll 1 Removal 3 i 3 5 mg/m3 266 F 32 12-5-67 Cil31 Same as No. 31 700 1:16 psig Replication 5 mg/m3 266 F 'j_ 33 12-12-6*( Cli I Feedback from No. 20-32 700 14 1: psig Ch3I2 Interac-3 5 mg/m3 266 F tion I2 100 mg/m3 31 12-19-67 CII 1 Feedback from No. 20-33 700 lle psig Spray Solution 4 3 i f2 plus single drop and solu-266 F Var,iation M 5 mg/m3 tion search k ln 12-25-67 to 1-8 Analysis of first half work, detailed planning of follovup runs as a result of other parts of program. O O k Om N 9 9 9 =5 O QUESTION In view of the large reduction factor which must 17.5 be supplied by the proposed chemical spray in order to meet part 100 and in view of the uncert: in the fraction of iodine in the containment whi. is in the organic or particulate forms, we belie-that all practical measures should be taken in other areas to reduce the potential of f-site dos. after an accident. This should include reducing the leak rate to the lowest practical value. ANSWER We believe that a leak rate of 0.2 per cent per day is adequately low to provide protection with the guide lines of 10CFR part 100, even in the event of a fission product release to the contai ment such as is assumed for a maximum hypothetic. accident. Releases of this magnitude would be prevented by the core cooling systems. If the premises underlying the above belief, in-cluding those concerning the effectiveness of th iodine spray removal cystem, are shown by the research and development program (outlined in 17 above) or other information to be unjustified, a adjustment in allowable leak rate can be made. O addition, the containment isolation system, cons ing of fluid block and pressurized penetration systems is being provided and is being designed to reduce the leak rate substantially below the allowable rate. To illustrate the effect of variations in the premises underlying the belief stated above (and underlying the dose estimates in Figure 14-56, Volume 2 of the pSAR), Figures 17.5-1 A & B are attached to show the variation of dose rate with assumed amount of unremovable iodine (i.e., methyl iodide), iodine removal coef ficient ( 24 ) and containment leak rate. t 0004 093 0 t Figure 17.5 n ,) !.,_i. a i ,-~ o i , t.

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• Notes Concerning Figures 17. 5-1A and B Figure 17.5-1A shows maximum estimated thyroid dose at the minimum exclusion radius (2,000 ft.) for the two hour period following a maximum hypothetical accident as a function of:

(1) containment leak rate (% of total volume / day) (2) DC, the constant fraction of initial iodine inventory in the containment which is. assumed not to be removable with iodine sprays. (3) 7\\ 2t) iodine spray removal coefficient (hr Figure 17.5-1A indicates that there is margin for un-certainty in the two parameters (C( and3h ) at a leak rate of 0.2% per day. For example, even if 04 is () assumed to be 0.1 (twice the referegce value=* of 0.05) and 2\\ is assumed to be 4 hr (about 1/6 of the reference value of 25.3), the estimated 2-hour dose is about 250 rem as is shown by point 1 on Figure

17. 5-1A f or a 0.2% per day leak rate.

Figure 17.5-1B shows estimates of the variation of maximum cumulative thyroid dose at the low population distance (two miles) for the 30 day period following a maximum hypothetical accident as a function of the same parameters ( C4, A and leak rate) as are treated in Figure 17.5-1A for the two hour dose. Figure 17.5-1B indicates that if C4 is assumed to be 0.1 (twice the reference value) and Jh is assumed to be ~t 3 hr (less than 1/8 of the reference value) a dose of about 180 rem is estimated for a leak rate of 0.2% per day. Even if o4 is assumed to bg 0.2 (four times the reference value) and A is 8 hr-(1/3 of the reference value) the maximum estimated 30 day low population distance dose would not exceed 300 rem. Figures 17. 5-1A and B were made using the methods described starting on page 14-47 (Revision dated 7/21/67) of the PSAR. O Reference values are those assumed in making iodine == dose calculations in Chapter 14 of the pSAR. 1 0004 096 O QUESTION In view of the large reduction factor which must 17.5 be supplied by the proposed chemical spray in order to meet Part 100 and in view of the uncert in the fraction of iodine in the containment whi is in the organic or particulate forms, we belie that all practical measures should be taken in other areas to reduce the potential of f-site dos after an accident. This should include reducing the leak rate to the lowest practice.1 value. ANSWER We believe that a leak rate of 0.2 per cent per day is adequately low to provide protection with the guide lines of 10CFR part 100, even in the event of a fission product release to the contai ment such as is assumed for a maximum hypothetic accident. Releases of this magnitude would be prevented by the core cooling systems. If the premises underlying the above belief, in-cluding those concerning the effectiveness of th iodine spray removal system, are shown by the research and development program (outlined in 17 above) or other information to be unjustified, a adjustment in allowable leak rate can be made. () addition, the containment isolation system, cons ing of fluid block and pressurized penetration systems is being provided and is being designed to reduce the leak rate substantially below the allowable rate. To illustrate the effect of variations in the premises underlying the belief stated above (and underlying the dose estimates in Figure 14-56, Volume 2 of the PSAR), Figures 17. 5-1 A & B are attached to show the variation of dose rate with assumed amount of unremovable iodine (i.e., methyl iodide), iodine removal coef ficient (A) and containment leak rate. 0004 097 (2) l l l Dockst 50-289 Supplement No. 3 December 8, 1967 (I QUESTION Discuss the consequences of opening a valve from,the containment 17.6 sump during injection after a loss-of-coolant accident if the sump is empty and the corresponding case for the borated water storage tank. ANSWER The emergency procedures vill be well established and rehearsed. Therefore, it is not considered reasonable that the operator voul inadvertently open the valve before it is prudent. If a valve is opened after h to 6 minutes have elapsed, sufficient coolant inve tory will be present in the reactor building sump to maintain a flooded suction for the decay heat removal pumps. Under these ci cumstances cooling for the subsystem with the open sump valve vil be provided in the recirculation mode. In the event that a valve in a sunp line is opened before the reactor building sump is flooded, coolant flow to the suction of the decay heat pump in that subsystem could be temporarily prevented because the reactor building pressure in the sump line would close the check valve from the borated water storage tank and because the reactor building sump vauld not contain sufficien coolant to cover the connected to the sump line. Coolant would continue to be supplied by the other emergency core cooling sub-system. Within 6 minutes, sufficient coolint vill have been pumped into the reactor building by the ot aer emergency core cooling subsystem and the reactor building spray pump to cover () the connection to the sump line with the open valve, thereby establishing cooling in the recirculation mode in this subsystem. There are no consequences to leaving the boratti vater storage tank valve open because the valve is of a stop-check design, and the check portion of the valve vill prevent reverse flow. l l 0004 098 l (12) l l l 17.6-1 (Rs rised 12-22-67) l Figure 17.5-s e i s l-t t l a r i. t i i..' V .,_.m 4 6 .m,,. n

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• Notes Concerning Figures 17.5-1A and B Figure 17.6-LA shows maximum estimated thyroid dose at the minimum exclusion radius (2,000 ft.) for the two hour period following a maximum hypothetical accident as a function of:

(1) containment leak rate (% of total volume / day) (2) D(, the constant fraction of initial iodine inventory in the containment which is assumed not to be removable with iodine sprays. (3) 2\\ 2t) iodine spray removal coefficient (hr Figure 17.5-1A indicates that there is margin for un-certainty in the two parameters (C( andJh ) at a leak rate of 0.2% per day. For example, even if 04 is (~~/) assumed to be 0.1 (twice the referegce value== of 2\\ is assumed to be 4 hr-(about 1/6 of 0.05) and the reference value of 25.3), the estimated 2-hour dose is about 250 rem as is shown by point 1 on Figure

17. 5-1A for a 0.2?. per day leak rate.

Figure 17.5-1B shows estimates of the variation of maximum cumulative thyroid dose at the low population distance (two miles) for the 30 day period following a maximum hypothetical accident as a function of the same parameters ( oL, A and leak rate) as are treated in Figure 17.5-1A for the two hour dose. Figure 17.5-1B indicates that if C4 is assumed to be 0.1 (twice the reference value) and 5% is assumed to be t 3 hr (less than 1/8 of the reference value) a dose of about 180 rem is estimated for a leak rate of 0.2% per day. Even if o( is assumed to by 0.2 (four times the reference value) and A is 8 hr-(1/3 of the reference value) the maximum estimated 30 day low population distance dose would not exceed 300 rem. Figures 17. 5-1A and B were made using the methods described starting on page 14-47 (Revision dated 7/21/67) of the 3SAR. Reference values are those assumed in making iodine == dose calculations in Chapter 14 of the pSAR. 0004 101 J l l Dockst 50-289 Supplement No. 3 December 8, 1967 QUESTION Discuss the consequences to containment integrity and doses to the 17 7 public if :he design basis less-of-coolant accident vere to occur after the plant had been operating with steam generator tube leaka (at least 30 gym) and secondary safety valve leakage. Discuss the implicatiors of the calculation with respect to technical specific tion limits on operation of the plant with generator tube leakage i secondary yafety valve leakage. ANSWER The consetuences to the public of simultaneous safety valve leakas of 10,000 lbs/hr, steam generator tube leakage of 10 gpm, and loss of-coolant accident have been analyzed. This series of events cou result in leakage from the reactor building, through the primary s-tem, through the leaking steam generator tube, and then through th safety valve to the atmosphere. As was demonstrated in the answer to Question 5 3 of Supplement No.1, leakage from the reactor buil ing to the atmosphere via the leaking safety valve cannot occur un the pressure within the steam generator decreases to reactor build pressure. For a steam gene ator which contains 20,000 lbs of wate the minimum water inventors, blowdown to reactor building pressure requires 3.6 hours. Thus, the 2-hour doses at the exclusion dista are unchanged from the values reported in Section ik.2.2.3 of the After blevdown of the steem generator to reactor building pressure j the pre-- L differential no longer prevents leakage. The leak ra i fu this path would result in an increase in the 30-day dose. It O 1 h e thro = h th r tv v 1 1 a to 1====

• t tar the entire 30-day period, the thyroid dose at the 2-mile zone is i:

creased to 3.3 rem. This dose, although higher than the 1.9 rem r-j ported in the PSAR, is well below the allowable limits of 10 CFR 1-The consequences of simultaneous safety valve leakage of 10,000 lb: steam generator tube leakage of 10 gpm, and the maximum hypothetic. accident have also been analyzed. As with the loss-of-coolant acc dent, the 2 hour doses are unaffected as steam generator blevdown requires more than 2 hours. The long term thyroid doses at the 2 mile zone are increased to Th rem for the first 2k-hours and 117 r i for the 30 day period instead of the comparable doses reported in the PSAR or k9 rem /24 hours and 72 rem /30 days. l When the technical specifications are prepared, due regard vill be given to plant operating limits imposed by potential steam generati tube and secondary safety valve leakage. When operational limits i steam generator tube leakage are prepared, the effects of leakage. secondary system vater chemistry, primary system makeup rate, and boiler blowdown rate must also be considered. Similarly, when sec-ondary system safety valve leakage limits are prepared, secondary system makeup and heat loss must be considered. l C 0004 102 17.7-1 (Revised 12-22-67) / Deckat 50-289 Supplement No. 3 18.0 CONTROL mfd INSRUMERATION December 8, 1557 () QUESTION In response to a previous staff questien (reference Supplement 18.1 No.1, Question No. 9.2) relating to the performance of equipment located within containment, you stated that the fan and valve motors vill have a system of insulation and enclosure which has demonstrated capability to perform under the post-accident environ ment. Please confirm that this capability vill be deternined by prototype environmental tests conducted under Metropolitan Edison' supervision that vill illustrate the ability of the equip =ent, including insulation and lubricating systems, to function in the accident environ =ent. Indicate the conditions and length of time for which the prototype vill be tested. Sections 3 7 and h.h of the IEEE Jtandard, Nuclear Power Plant Protection Systems (Revisio:

9) should be addressed.

ANSWER The fan motors to be furnished with the Reactor Building ven-tilators will be designed so that vindings and bearing surfaces are protected against the accident ambient. Motor housings vill withstand 60 psi and vill be provided with an air-to-vater heat exchanger to be supplied frem the sa=e source as its acccmpanying ventilating cooling coil. The heat exchanger vill be selected to maintain a lov humidity internal ambient with safe winding temperatures. The winding insulation type vill have been demonstrated to withstand an accumulated dose greater than the expected lifeti=e and LOCA dosages. /~ \\ Bearings vill be of a seal type which vill withstand the LOCA pressure pulse and vill be eccled along with the motor internal air. We do not believe that an actual environmental test of the conservative motor design described vill be necessary. If, after the =anufacturer's preli=inary design is finali zed and a presen-tation =ade, it is determined that environmental test confirmation of bearing design is required, we would propose that a prototype driving end bearing and bearing seal be tested under the folleving conditions: 300 F at 60 psi, 100 percent humidity for 2 hours, followed by straight line decay to 150 F and 15 psi in 3 hours, folleved by 30 days subsequent operation in that ambient. Since the question concerns an industry problem, ve do not believe that the test should necessarily be under Metropolitan Edison supervision but Metropolitan Edisen would witness tests and approve the test results. With respect to section 3.7 and h.h of the IEEE Standard for Nuclear Pever Plant Protection Syste=s, the motor vill be separately tested to withstand a centinuous ecmbined frequency ,_,() and voltage variation of plus er =inus 10 percent and starting voltage of 75 percent of ner=al. 0004 103 18.1-1 \\ / Dcekst 50-289 Supplement No. 3 December 8, 1967 CUESTION Assuming that it becc=es necessary to abandon the control room 18.2 during full power operation, please discuss the precedures that vould have to be perfor=ed external to the control room which would ensure a safe shutdevn of the plant for an indefinite time. Include, in your discussion (1) the instrumentation that would be required to monitor vital plant para =eters, (2) components that would require manual actuation, and (3) the available times to take the actions necessary. Coincident loss of off-site power should be considered. How much of the above capability ~ vill be incorporate 4 in your design? ANSWER The question assur2es that it becomes necessary to abanden the control room, hevever, such a circumstance is not considered credible for tv.e following reasons: 1. Adequate shielding is provided to =aintain tolerable radiation levels in the Control Room during an MHA. 2. Accessibility is assured by having at least four points of entry from outside the " control tcver." 3 Nenflammable construction h. Cables and switchboard viring pass flame test per IPCEA publication S-61-402 and NE4A WC5-1961. O 5 Metal furniture is used throughout. 6. Smoke protection and detection equipment will be provided. While control room abandonment is not considered credible, the following infor=ation is supplied in answer to the assumed con-ditions postulated by this question. In order to maintain the reactor in a safe shutdown condition without access to the control room, it vill be necessary for the operator to have the availabilit of the information and the ability to take the assorted actions listed below. The information as listed indicates the presently contemplated design arrangemeat. Information Recuired Availability 1. Pressurizer Level Available 2. Reactor Outlet Temperature Available 3. Feedvater Pressure Available i O 4'# O .? l 18.2-1 ] L 4 Information Recuired Availability k. Steam Pressure Available 5. Steam Generator Water Level Available 6. Station Paging System Available Action Recuired Availability 1. Operation of high pressure Available injection pumps to maintain pressurizer level. 2. Operation of the emergency Controls available feedvater pump at the pump 3 Operation of =ake-up valves Manual by use of hand jacks at valv location 4 Operation of borated water Manual by use of storage tank supply valves hand jacks at valv to high pressure injection location pumps. 5 Ability to remove steam from Automatic by actio: the steam generator of the safety valvi 6. Ability to supply emergency Emergency diesels power (in the event of loss start automatica11 of outside power) In developing the action necessary outside of the control room, it has been assumed that after the operator makes the decision to evacuate the control room, he immediately takes the following actic prior to departing: 1. Actuates the scram button which will cause the following to occur: a. Insertion of the control rods into the core b. The turbine generator to trip 2. Starts emergency feedvater pumps 3. Trips =ain feedvater pumps h. Trips reactor coolant pumps 5. Trips high pressure injection ptmrps 6. Closes the letdown valve 7. Closes the seal return valve 8., Closes the reactor coolant pump seal injection water valve

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b U t,i t, 0004 105 18.2-2 . ~. _ _ _. _.. _. _ _... _ _ _ _.. _. 1 i In the event that off-site power is lost, items 2, 3, h, and 5 above will occur automatically, t It is anticipated that the operations in the centrol room can be executed within one or two minutes. There would be several hours available to place the make-up system into operation. i 0004 106 1 i 1 i O i .i l l I l o 18.2-3 boc'";t 50-289 Supplement No. 3 December 8, 1967 QUESTION Analy:e the consequence of an accidental phase reversal at an 18.3 emergency bus under accident conditions. The single failure criterion should not be used as a basis for the analysis in those cases where machinery rotating in reverse has an adverse effect on its redundant counterparts. ANS'4ER The Engineered Safeguards Systems, Makeup and Purification (High Pressure Injection), Decay Heat Re= oval System and Reactor Building Spray System were analyzed for the consequences of an accidental phase reversal at an emergency bus under accident condit For this analysis it was assumed that the consequences as a result of a phase reversal vould be the reverse rotation of a pump or the opposite movement of a motor operated valve from its intended direction. The analysis presented in Table 18.3-1 is based on the assumption that a major less-of-coolant accident had occurred and that a phase reversal at en emergency bus exists. All ar,tive components, as listed in Table 18.3-1 of the engineered safeguards vill be tested periodically to demonstrate system readiness and we do not believe that a total three phase reversal could occur. Table 18.3-1 Phase Reversal Failure Analysis Compenent Failure Comments & Consequences A. Makeup and Purification (High Pressure Injection) System Makeup Pu=p Reverse Rotation Standby pump is available of Pump for operation B. Decay Heat Removal System Decay Heat Pump Reverse Rotation Two remaining pumps vill of Pump deliver required flov Electric Motor Valve Remains Two lines & valves Operated Permitting Closed are provided suction from Reactor Building Su=p C. Reactor Building Spray System Reactor Bldg. Reverse Rotation Flov & cooling capacity Spray Pump of Pump reduced to 50 percent of design. In ecmbination >:,i, with emergency coolers, M 150 percent of total design O requirements is still provided. 0004 107 u.2-1 l Component Failure Ccmments & Consecueness Electric Motor Valve Remains Second header delivers SC Operated Valve Closed percent flow. See Ccmmer. in spray header for C-1 above. Each of the following systems was analyzed from a single failure phase reversal plus single accident consideration. While both phase reversal of any one valve and of a ecmplete bus were considered, in all cases reversal of the complete bus presented the more difficult handling. Main Steam Comtenent Failure Accident Remarks Main Steam Line Valve Open LOCA or Steam Turbine Stop Isolation Valves Line Rupture Valves Close DELBc. Condensate-Feedvater Under any ecmbination of single bas or single valve reversal plus single lh accident,- condensate-feedvater flov to a steam generator via one of the emergency pumps from one of the three storage areas is possible by realigning the flow pattern. The following presents a restriction which is tolerable: Comtenent Failure Accident Remarks Isolation valve Valve Closed Rupture of Isolation va to emergenev opposite steam to normal fe feed no::le on generator or its nozzle on st steam generator steam piping generator op Reversed val vould be man opened. Circulating & River Water No combination of single valve or single bus reversal plus single accident vill prohibit flow thru the nuclear services coolers, 0004 108 O c:

tuo, 1c.3-2 (Revis ed 12-22-67)

Nuclear Services Cooling Water Cemeenent Failure Accident Remarks Energency Valve closes LOCA One of three Cooling Unit cooling units Isolated Valve is lost. Rene Building spray gystem is redu dant and avail able. Phase reversal of the 3 chase power to the control red drive motor power supply will have no effect on direction of rod movement. Rod drive motors are fed from an electronic power supply which generates two-phase power whose phase secuence is determined by a culse generator that is indepen-dent of source feeder phase rotation. General Motor driven emergency safeguard pumps all have redundancy. If one pump runs in reverse, it vill pump ahead at reduced htad against its closed check valve, and the resultant ammeter reading vill be lover than the nornal pump. Phase reversal after initial check out is in itself a most remote possibility but should it ever occur, the above analysis proves that the remaining systems () and auxiliaries can nerform all the required safeguards functions. 0004 109 OV 13.3-3 (Revised 1-S-6a) O Docket 50-289 Supplement No. I Dece=ber 8, 1961 QUESTION Discuss, in detail, your criteria relating to the physical separatic 18.h

• of the installed instru=ent and logic channels which initiate protec i

tive and emergency safety feature action. This discussion should ir clude: (a) separation between redundant instru=ents, (b) separaticn between redundant relays and breakers, (c) routing of redundant viring, 7,.(d),,persanently installed test equipment which say be ec= men to re-i; auiundant instrument channels. ANSWER Within the protective systems, physical isolation vill be provided to minimize the likelihcod of a single event within or outside of a system cabinet or equipment housing frem i= pairing the operation of more than one protective channel and its associated instruments. Physical isolation is achieved by using construction elements withir and between cabinets, metal electrical conduit, and physical separa-tion of primary elements. () (a) Physical isolation of redundant instruments is achieved througr physical separation. For example, pri=ary reactor coolant prer sure sensors are located on individual pressure taps and divide between the two primary coolant loops. This principle of sepa-ration applies to all protective system sensors. (b) Redundant protective system relays are packaged in relay module and these in turn are divided between the protective systems cabinets. Physical isclation is therefore two-fold, being fire provided in the relay module package or box, and second by the division of =cdules between individual protective syste= cabi-nets based upon the relay's function and relationship to the system. Reactor protective system breakers are physically separated anc are in separate cabinets separated by metal barriers. (c) The cabling, both into and away frem the protection systems cal inets, vill be carried in clcsed =etal trays er soft iron con-duit. The cables will be assigned to separate conduits based upon separating individual channel signals and electrical func. tiens. Typical of the separation of electrical function vould be physically isolating between neutron signal cables for the pcVer range instrumentation which will be carried in separate soft iron conduits for each channel. Cable carrying trays and () conduits vill be color coded to identify the protective system usage. 18.u-1 0004 110 (d) The protective systems design criteria require that the built-in test equipment preserva the systems isolation require =ents. This requires that any permanently installed test equipment ec=- =cn to redundant channels be independent of the channels; there-fore, no two channels =ay be simultaneously connected to the test equipment without causing a system trip. Co==0n test equip-ment vill be physically isolated by being packaged as an inte-gral, independent ec=penent. The equip =ent would be located in a cabinet housing a single protective channel. 0004 111 O. O ~ ,,e t't.. a s 13.k-2 Dockst 50-289 Supplement No. 3 December 8, 1967 () QUESTION List the emergency equipment which is povered by the engineered 18.5 safeguards busses. List the expected leads and loading sequence and discuss your philosophy of Icading the diesel above its nameplate rating. Under what conditions vill the engineered safeguards busses be tied together? Will the tie be automatic or manual? ANSWER (a) The diesel generator units have been sized at 2850 kv, with the folleving automatically and manually applied loading in the event of a LOCA: Loading Expected Secuence Number Description Lead (Total) Block 1 1 Makeup Pump 700 hp 1 Decay Heat Pu=p h00 hp Miscellaneous Leads (Emergency lighting, 150 hp valves etc.) Block 2 2 Reactor Bldg. Ventilators h00 hp 1 Reactor 31dg. Vent. Energ. Pump 450 hp Block 3 1 Reactor 31dg. Spray Pump 250 hp 1 Nuclear Services Closed Cycle Pu=p 150 hp 1 Nuclear Services River Water Pump 75 hp ( Block h 1 Decay Heat Closed Cycle Pump 125 hp 1 Decay Heat. River Water Pu=p 125 hp Miscellaneous Loads 150 hp Manually Control Bldg. Chiller 150 hp Applied Centrol Bldg. Vent. Fan 75 hp Loads Spent Fuel Pump h0 hp 3240 hp The above represents the heaviest leading on one diesel-generator in the event that the other failed to start and would result in a total load of: 2630 kilevatts. This includes a minimum of 100% of the required capacity for all engineered safeguards functions. l The total =axt=cn applied lead vill therefore be less than the diesel-generator na=eplate rating. (b) Engineered safeguards busses vill be tied tcgether manually only. Short of a requirement to feed a condensate pu=p (See question h.9) there is no e=ergency require =ent for tying the safeguards busses together when fed frc= the diesels. O" 0004 112 15.5-1 '?evised 12-22-67} l l l 1 Dockat 50-289 SupplemInt No. 3 December 8, 1967 QUESTION On page 9 kl of the PSAR you state that there is manual 18.6 provision for switching to full recirculation for post-accident control recm ventilation. Please discuss your , justification for the absence of automatic switching in response to a signal indicative of an accident condition. ANSWER We are designing for automatic switching with a =anual override. 0004 l13 t O O 18.6-1 l Dockst 50-289 Supplemsnt No. 3 December S, 1967 QUESTION What are the expected ranges of the atmospheric and liquid =onitor O 18.7 systems? Indicate the relationship between your design basis accident analysis and the ranges, sensitivities and detector j 1 locations of the radiation monitoring system. ANSWER The expected ranges of the atmospheric and liquid monitoring zystems are covered in the revised ansvers to questions 10.h and 10 5 The relationship between the design basis accident and the ranges, sensitivities, and detector locations is covered in the first four paragraphs of the revised answer to question 10 5 0004 114 l O l 1 \\) i i 13.7-1 .}}