ML20148A948

Responds to Re Evacuation Distances Around TMI-2. Forwards Table Developed on 790401 Giving Recommended Actions for Various Contingencies & Draft of long-term contingencies.Ten-mile Evacuation Sufficient

ML20148A948
Person / Time
Issue date:	12/05/1979
From:	Hendrie J NRC COMMISSION (OCM)
To:	Yellin J HARVARD UNIV., CAMBRIDGE, MA
References
NUDOCS 7912270107
Download: ML20148A948 (48)

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4 f((ean 'o g UNITED STATES gA g g NUCLEAR REGULATORY COMMISSION g(L  ;

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'%...+f December 5, 1979 1

CHAIRMAN j

Dr. Joel Yellin J Associate Professor, M.I.T. j John F. Kennedy School of Government 1 Harvard University 79 Boylston Street Cambridge, Massachusetts 02138

Dear Dr. Yellin:

This is in response to your letter of July 31, 1979.

The ad'vice given to Governor Thornburgh on March 30, 1979 regarding possible evacuation distances in the event of a major release of radioactivity at Three Mile Island was not explicitly derived from specific event sequences i in WASH-1400 or NUREG-0396. However, my advice took into account my general understanding of core melt. consequence modeling and analyses of the type relied on in both of these reports as well as some recollections of similar dose modeling in the WASH-740 study.

The Commission did consult subsequently with members of the staff of the Offices of Nuclear Reactor Regulation, Nuclear Regulatory Research, Inspec-tion and Enforcement and State Programs in attempting to understand the possible extent of evacuations that might be shown to be necessary if the accident were to become more serious or produce consequences beyond those already experienced by March 30, 1979. A table developed on April 1, 1979 giving recommended actions for various contingencies, which was developed under the direction of Corm 1issioner Gilinsky and given to State representatives on the same date, is enclosed together with a markup of an initial draf t of longer term contingencies.

As reflected in the March 31 transcript at the pages you cite (27-29), the ,

20 miles was meant as a rough estimate of a worst case situation, even '

though, as noted in NUREG-0396, a 10-mile evacuation would be sufficient for most of the severe core melt consequence scenarios. As you surmise, my response to Governor Thornburgh's question was made without " extensive consideration of the technical argum.ents." ,

.y  ;

iSincerely, l

1 k r[ &

'dosep6M.Hendrie i

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'- i HARVARD UNIVERSITY

) JOHN F. KENNEDY SCHOOL OF GOVERNMENT 79 BOYLsTON STREET CAMBRIDGE, MASSACHUSETTS 0213S (617)' 495-1328 i July 31,1979

.The Hon. Joseph Hendrie Chairman U.S. Nuclear Regulatory Commission Washington, D.C. 20555

Dear Dr..Hendrie:

In connection with our ongoing research on nuclear safety, we have a number of questions relating to the Three Mile Island accident.

. You advised. Governor Thronburgh, on March 30, 1979, that in the event .of a major release of radioactivity, evacuation out to 20 miles from the Three !sile Island generating station would be called for.

advice? What was the technical basis for that Did you consult with members of the Regulatory Staff, or with your colleagues on If so,thewith Commission whom andconcerning when? the " worst case" evacuation distance estirnates?

Were you influenced, in giving that advice, by information presented in WASH-1400, or by the material presented in NUREG-0396, the NRC-EPA emergency planning document? If so, what specific information did you use?

Did you receive, or have you received since, private memoranda from t'ne Staff or event fromofyour a majorcolleagues release?concerning the appropriate evacuation distance: in the If so, we would like copies of that material.

The' colloquy in the NRC transcript of March 31 (pp.28-29) suggests that your selection of a 20-mile distance was not the result of an extensive considerat of the technical arguments, but was in fact an extempore response.

fair evaluation? Is that a We look forward to hearing from you.

,'ncerely you ,

\

M '

oel Yelli.

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.b MEMDRARDUM FDR: linger S. Boyd. Director. Division of Pmject Management. NRR FROM: Norman C. Moseley. Director. Division of Reactor Operations Inspectice. IE SUEJECT: PUSLIC SERVICE ELECTRIC MD GAS COMPANY SALEM UNIT 2 We have been informed by our Pacion ! Sffice, based on their inspection findings, that construction and preoperational testing of the subject l facility have been completed in substantial agreement with docketed l commitments and regulatory requiremects, with the exceptions listed in l the enclosures. The Office of Inspection and Enforcement has no further items which would preclude isssance of so operattag License to permit facility operatioe up to its full design rating. It is reconnended that the operating license be conditioned with the information contained is i

the enclosurws.

j We have reviewed the licensee's preparations for implementation of the '

quality Assurance Program for Operations and have found that they meet the requirent,nts of 10 CFR So, Appendix 3, as specified in the licensee's Quality Assurance Program (Appendiz D of the F5A2) which was reviewed by the Office of Muclear Reactor Regulation.

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. Norman C. Moseley Director Division of Reactor Operations Ingection Office of laspection i

and Enforcament Enclosuees:

A. Items to be Completed Sefore Fuel Loading B. Items to be Completed Before Mode 2 [o 4/l cc: See neat page '

J. I. Riesland 79060200 & 90024144 C0!iTACT:

(x28019) l FC:ROI:IE ADFC:ROI:IE DIDR;I;IE D:DSI:IE D:DFF&MSI:IE D:DROI:IE JIRiesland:LD SEBryan HDThornburg EMHoward JHSniezek NCMoseley

, afufva si fio si rio 3/ f70 3/ /79 3/ /79

Roser S. Soyd 2-l cc: D. B. Yassallo, nRR i

0. D. Parr, HRR l A. W. Dromerick, NRR B. H. Grier, RI E. M. Howard. IE H. D. Thornburg, It i J. !!. Sniezek, IE E. J. Brunner, RI R. R. Kaleig, RI l L. k rrholm, RI J. I. Riesland, IE w r, r ow n an J cs 1 ; lu o bi _ ( _ m 90024145 l

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- ENCLOSURE A Items to Complete Before Loadina Fuel j .The following items shall be completed prior to loading fuel in the Salem Unit 2 reactor core. All items will be verified complete through inspection by NRC: Region I inspectors.

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1. Complete the review of all Phase II testing.

2. Complete the'following Phase II tests

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a. SUP S1 Integrated Test of Engineered Safeguards and Emergency .

Power System i b. SUP 20.2 Reactor Pro,tection Operational Check j

c. SUP 10.4 Boric Acid Blender Performance Test 4

d. SUP 21 Radiation Monitoring: Complete for the following channels:

f l Local indication and alarm (8).

Local and remote indication and alarm (S). 1 E. SUP'20.3 Safeguards System Operational Test j

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f. SUP 48 Preservice Test of Pumps a1d Valves 9

SUP 49 Emergency Lighting r

h. SUP 24 Nuclear Instrumentation j i. SUP 17.1 Manipulator Crane Indexing

j. SUP 17.4 Fuel Handling Tools and Fixtures

k. SUP 17.5 Fuel Transfer System

1. SUP 18.1 Containment Spray System

m. SUP 19.1 Auxiliary Building Ventilation System

n. SUP 19.2 Containment Iodine Removal

o. SUP 19.6 Control Room Ventilation

p. SUP 19.7 Chilled Water System l

q. SUP 19.8 Fuel Handling Suilding Ventilation

r. SUP 20.1 Reactor Protection / Emergency Safeguarcs ilme Response Measurement 90024146 J

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Enc 1osure A 2

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. s. SUP 20.5 Reactor Plant Systems Setpoint Verification  !

t. SUP 23 Communications System

u. SUP 25B Heat Tracing System-Vital

v. SUP 26 Computer Input and Data Printout Verification (NSSS

only) t

w. SUP 28 Energizing Electrical Buses - General

x. SUP 32 Service Water System

y. SUP 38 115 VAC System

3. Complete data review of Containment Integrated Leak Rate Test.

1 4. Complete Pre Operational Test and Turnover (POTT) of the following systems:  !

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a. Control Air

b. Auxiliary Building-General

, c. Auxiliaries Controlled Systems

d. Chemical and Volume Control System

e. Communications

f. Containment 9 Containment Spray

h. Demineralized Water

i. Fire Protection - Water

j. Fire Protection - CO2

k. Incore Instrumentation

1. Lighting

m. Nuclear Instrumentation

n. Radiation Monitoring

o. Rod Control

p. Security Systems  ;

q. Control and Relay Room HVAC

5. Prepare and issue Core Loading Procedure and Startup and Low Power Ascension Test Procedures for tests identified in the FSAR.

6. Update Emergency Plan and implementing Procedures and conduct emergency drill from Unit 2 Control Room

7. Submit satisfactory response to IE Sulletin 77-04 (Containment sump pH control).

E. Complete stem-mounted limit switch modifications per ECN's 25195 and 35294 (Containment isolation valves).

9. Verify anchor bc1t embedment acceptability (Containment structures).

90024147

Enclosure A 3

- i 10. Revise Performance Department Manual to address:

-f a. Qualification of Performance Supervisor-HP and Chemistry

! b. Designation of Corporate Health Physicist c.

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Qualifications of Technicians-Nuclear (ANSI 18.1-1971) i 11. Prepare and issue procedures to address the following health

physics areas

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a. MPC-Hour Accountability

b. Extremity Monitoring

c. Beta Dose Rate Determination l d. Overexposure Investigation

e. Lost, Damaged, or Off-Scale Dosimeter or TLD Evaluation

f. Bioassa ry

g. Nasal Smears i

h. Respiratory Protection Procedures

i. Decontamination of Personnel

12. Prepare and issue revisions to health physics procedures to address

+he following:

a. 15.3.009 - Form NRC-5 applicability

. b. 15. 4. 008 and 15. 4. 009 - Relate air samples to exposure

c. 15.3.009 - Address current requirements of 49 CFR and 10 CFR 71

13. Revise, issue and implement procedures or changes as necessary, to <

address the following in Station Administrative Procedures:

a. AP Define organization as specified by ANSI N18.7-1976

b. Prepare summary document required by ANSI N18.7-1976 Section 5.1

c. Define the six specific ope.ator responsiblities listed in ANSI N18.7-1976 in station procedures.

d. AP document verification of jumper placement.

e. Maintenance procedure A make consistent with AP-13 regarding temporary jumpers.

f. AP-1 Include method of assuring that department manuals are consistent with Station Administrative Procedures.

g. Issue and implement draf t Station Adminstrative Procedures which reflect requirements of ANSI N18.7-1976.

h. Establish preventive maintenance program via Inspection Order system for Unit 2.

i. Maintenance Procedure A Include ANSI N18.1-1971 minimum qualification requirements for Level I maintenance personnel,

j. AP Clarify authority to use Night Order Book to oromulgate temocrary procedures.

K. Ciarify control of temporary changes to system valve lineups.

1. Include procedure adherence requirements of A"SI N18.7-1976 in station procedures.

90024148

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r.nclosure X 4

m. Establish a system for documented periodic review of Adminstrative Procedures.

1 n. Establish system for periodic review of Station Plant i

Manual procedures.

o. Establish consistent format requirements for Emergency and Operating Instructions.

p. Provide approval status documentation for refueling l instructions in Station Plant Manaul.

I q. Issue and implement Unit 2 Reactor Engineering Manual.

14. Issue and implement operating procedures or changes to existing procedures to address the following areas listed in Regulatory i Guide 1.33: l l a. Operation at Hot Standby l

. b. Pressurizer Pressure and Spray Control System i j c. Reactor Control and Protection System

! d. Loss of Condenser Vacuum

e. Loss of Service Water (present procedure covers only partial loss of service water)

.! f. Loss of Protective System Channel (present procedure covers only loss of nuclear instrumentation)

g. Malfunction of Pressure Control System  ;

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15. Issue and' implement Station procedures for Unit 2 covering the following areas:

a. Alarm procedures of the Station Plant Manual

b. Mode 3 and Mode 4 Surveillance Checkoff Lists of 01 1-3.2 (including 500KV and 13KV systems)

c. Maintenance Department Manual revisions required for consistency with station procedures, j d. Fire Fighting and Organization Manual and associated

, Surveillance Procedures.

. 16. Issue and implement Surveillance Procedures covering the following surveillance requirements in the proposed facility Technical Specifications:

a. 4.1.1.5.b j. 4. 8.1.1. 2. a

b. 4.1.3.5 k. 4.8.2.3.2.b

c. 4.4.9.3.2 1. 4.8.2.5.2.b

d. 4.5.3.2.

e. 4. 7.1. 2

f. 4.7.10.1.1,2,3

g. 4.7.10.3

h. 4. 7.10. a

i. 4.7.11 ,

i 90024149 1

Enclosure A 5

' 17. Issue and implement the following station Plant Manual Operating Instructions (01's):

a. II - i.3.5, RC Leak Detection;

b. II - 4.3.1, Safety Injection System - Normal Operation;

c. II - 5.3.1, Fill and Vent the CS System; i

d. II - 9. 3.1, Manipulation Crane;

e. II - 9.3. 2, Fuel Transfer System; I

f. II - 9.3.3, Control Rod Shaft Unlatching Tool; i g. II - 9.3.5, Thimble Plug Handling Tool; i

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h. II - 9.3.10, RCC Changing Fixture;

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i. II . 9.3.11, Burnable Poison Rod Assembly Handling Tool;

j. II - 9.3.12, New Fuel Assembly Handling Fixture; 3 k. ' II - 9.4.1, Fuel Transfer System; s.

1. II - 8.3.3, Filling the Reactor Refueling Cavity;

m. II - 8.3.4, Draining the Reactor Refueling Cavity;

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n. II - 10.3.2, Gas Analyzer Operation;

' o. II - 11.3.1, Waste Holdup Tanks - Normal Operation;

p. II - 14.3.1, Containment Drains System Operat on;

q. II - 15. 3.1, Hydrogen Recombiner - Normal Operation;

r. II - 15. 3.2, Personnel Locks and Containment Entry; 90024150

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Enclosure A 6 l .

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s. II - 17.3.1, Auxiliary Building Ventilation Operation;

t. III - 13.3.1, Draining the Steam Generator;

u. IV - 1.3.1A and B, 500 KV - Normal Operation; i

v. IV - 3.3.1 13 XV - Normal Operation;

w. IV - 4.3.3 A and B, 4 KV Group Busses - Normal Operation;

x. IV - 5.3.1, Battery Ground Detection;

y. IV - 5.3.2, Battery Charger Operation; j z. IV - 6.3.2, Operation of the Axial Flux Deviation System; aa. IV - 7.3.1, Flux Mapping System - Normal Operation; j ab. IV - 11.3.1, Area Radiation Monitors - Normal Operation; ac. IV .11.3.2, Process Filters Radiation Monitors-Normal Operation;  ;

ad. IV - 11.3.3, Process Radiation Monitor-Normal Operation; I

ae. IV - 11.3.4, Operation of the Control / Plant Ventilation Sampler; a f. V - 1.3.1, Service Water - Normal Operation;

__ ag. V - 3.3.1, Fire Protection System Operation; ah. v'- 3.3.2, Refilling Liquid Foam Storage Tank; ai. V - 3.3.4, Notification of Impairment to the Fire Protection System; aj. V - 9.3.k, Chilled Water System - Normal Operation.

18. Define the functions and responsibilities of the off-site QA engineer.

19. Complete all construction and testing open items identified in SMII-6 (Salem Unit 2 - Startup Manual Implementing Instruction - 6 Outstanding Item List) Currently tracking items required to be completed prior to core load. Justification for listed items deferred to a milestone beyond core load will be subject to review by NRC Region I inspectors.

90024151 r= "

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Enclosure A 7

20. Complete installation and construction testing of the Radiation Monitoring System. Inspection of this item by NRC Region I will complete the required construction inspection program.

21. Modify emergency diesel generator controls to bypass additional trips on emergency (Safeguard equipment control) start in accordance with the application as amended.

f 90024152

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Enclosure B

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Items to be Completed Before Achievina Mode 2 4

.1 . Complete the following Phase II tests:

a. SUP 6C Compressed Air System - Automatic Start

b. SUP 16.2 Liquid Waste Processing - Waste Evaporator

c. SUP 16.4 Resin Removal System

d. SUP 16.6 Gaseous Waste Processing

e. SUP 10.3 Boron Recycle Process

f. SUP 16.1 Liquid Waste' Receipt and Storage

g. SUP 20.4 Control- System Test for Turbine Runback i

i h. SUP 21 Radiation Monitorina

i. SUP 7 Con. trol Air System f

j. SUP 42 Hydrogen Recombiner

! 2. Complete Emergency Diesel Generator reliability testing.

3. Demonstrate abi.ity to provide alternate source of Auxiliary Feedwater within 30 minutes.

4. ~ Complete all construction and testing open items identified in SMII-6 Open Items List for completion by Mode 2 or initial criticality.

All listed items deferred to a milestone beyond criticality will be subject to-inspection and review by NRC Region I inspectors.

5. Issue chlorination and chemical inventory procedures.

6. Install control board information placards similar to those in place at Salem 1.

90024153 I

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