ML20151X022

Forwards Request for Addl Info Re Instrumentation,Control & Power Sys to Support 670116 Applicatiion for CP & License

ML20151X022
Person / Time
Site:	Diablo Canyon, 05000000
Issue date:	05/18/1967
From:	Morris P US ATOMIC ENERGY COMMISSION (AEC)
To:	Peterson R PACIFIC GAS & ELECTRIC CO.
Shared Package
ML20151W779	List: ML20151W777 ML20151W807 ML20151W834 ML20151W945 ML20151W976 ML20151X022 ML20151X037 ML20151X057 ML20151X067 ML20151X076 ML20151X087 ML20151X097 ML20151X119 ML20151X125 ML20151X306 ML20151X322 ML20151X358 ML20151X381 ML20151X405 ML20151X436 ML20151X461 ML20151X466 ML20151X585 ML20151X612 ML20151X719 ML20151X755 ML20151X798 ML20151X821 ML20151X882
References
FOIA-88-156 NUDOCS 8808250198
Download: ML20151X022 (5)

Text

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H. Steele bcc: W. B. Cottrell, ORNL Pacific Gas and Electric Company 245 Market Street San Francisco, California 94106 Attention: Mr. Richard H. Peterson Senior Vice President &

General Counsel Gentlemen:

TF's refers to your application dated January 16, 1967, for a construc-tiva permit and facility license which would authorize construction and operation of a nuclear power reactor at the Diablo Canyon site located in San Luis Obispo County, California.

On March 21, 1967, and April 20-21, 1967, members of the regulatory staf f met with representatives of your company to discuss various aspects of the plant design. As a result of this meeting, we requested additional information pertaining to the s,dte, plant layout, and containment struc-tural design by letter dated May 5, 1967. We indicated in the referenced letter that questions related to other aspects of the design .would be forwarded in subsequent correspondence.

Accordingly, you are requested to provide the information listed in the enclosure pertaining to instrumentation, control, and power systems. ,

We are continuing our review and will develop further questions in the remaining areas of the plant design.

Your reply to these attached questions should be submitted as an amend-ment to your application. The staff, of course, will be available as may be required to discuss and amplify the meaning of the questions.

Sincerely yours, ORIGINAL SiONED BY Peter A. Morris Peter A. Morris, Director Division of Reactor Licensing

Enclosure:

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i RrgyEST FOR ADDITIONAL INFORMATION PACIFIC CAS AND ELECTRIC COMPAh*(

DIABLO CANYON REACTOR DOCKET NO. 50-275 IV. Instrumentation. Control, and Power Systema _

A. More specific design criteria should be provided for the protection systems which initiate reactor trip, containment isolation, emer-For gency core cooling, and other engineered safety features.

example, each of the items of section four of the IEEE proposed Standards for Reactor Protection Systems _ (Rev. 7 or 8) should be addressed, for those items o'f the proposed standard which are criteria for Disblo Canyon, discuss how the criteria are to be met.

If any are not to be met, discuss why they are not necessary for this design.

B. Please compare and identify any dif ferences between the reactor protection system and the instrumentation and controls for enoinected safety features for Df ablo Canyon and those for the H. B. Robinsen and Point Beach reactor plants.

C. Please supply clementary (schematic) diagrams of the logic circuits which actuate reactor trips, emergency core cooling, containment isolation,and other engineered safety features.

D. Identify and describe the control functions (in addition to actuation) required for successful operation of the engineered safety features.

E. Describe each of the reactor trip channels more fully, including:

1. T/pe of components (sensors , amplifiers, bistables , etc.).

2. Trip point arrangement fixed, chsnged by mode switch, changed by another variable, and means for bypassing.

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3. Where variable trip points are used, indica e whether the means for varying the trip point meet the single failure criterion.

4. With regard to testing, state the design criteria and indicate:

1 (a) What parts of the circuit can be tested at power (b) what parts of the circuit must be tested when shut down (c) how does the test check for the loss of redundancy

i 2

(d) where and how is the test signal injected (c) how is the result of the test detected.

F. Describe specifically the interlocking of reactor trip signals with measured nuclear and steam power (page 7-3).

G. Analyze the ' rod drop' protection in terna of the assurance the droppingthere is of any that 'out of core' neutron detector s will detect rod (consider the case for 1 of the 4 neutron Describe the protection circuits specifically channels including features toout of servic meet the single failure criterion and hor they will be tested.

H.

Does the 'lew rod insertion' limit alarm (page 7 4) n(et the single failure criterion? Describe the circuit.

safe:y I.

Describe the local s:ations where the operation of engineered features and plant shutdown equipeent can be controlled and monitored Where are in case the central control room becomes untnhabitable.

the stations located and how is the operator protected from excessive Include radiation? What indications of plant status are provided?

a discussion of the design features to be considered for trarsfer of operations from the central control room to the local stations.

J. Wha t ins trumentation is required for post accideat recovery cperat tens?

State your criteria for indicating to the operator the reactivity s tat il ef the reactor and the pressure, temperature, water levels following an, hydrogen, and boron concentration in the containment accident.

and electrical K. Evaluate the ability of the reactor protecticn equiprent equipment for the engineered safety features te withstand the environ-ment in which it must operate. The evaluation should include but not be lirited to the following:

1. Identification of the equipment, including cables, inside the containment which State themust functionand environment in an accidentlength expected environment. Describe of time for which this equipment must function.

the tests or test dat'. which will demonstrate the ability of the coeponent to function in the accident envirennent.

2. An evaluat'.on of the performance of the control room equipment i ndi e abnormal conditions, such as heating or air condittoatag failure.

Is forced cooling required for any of the sensors?

What are the ef fects of loss of forced cooling? Is 3

the coeling ronitoredt

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4 An evaluation of the ability of the electrical equipment such as pump motors, esbles, and instrumentation in the emergency cere cooling systems to withstand the radiation to which it will be exposed during recircula-tion af ter a major accident.

5. A description of the containment sump level instrumentation and a discussion of their capabilities in terms of environ-mental conditions and maximum water levels.

L. Are the ECCS accumulator level sensors connected to commen sensing lines?

M. What prevents improper cperation of the safety injection block switch?

N. Discuss the use of the coincidence of pressurizer level and pressure for the actuatten of energency core cooling. Is there ro ccndition in which emergency core cooling might be required when either Icw Could the reference leg pressure or low level eight not be sensed? Are of the level sensors be voided as the result of the blewdown?

spurious trips from the pressure or level instruments a serious enough threat to justify a system which requires tripping both for ECCS actuaticn?

O. Do the containment high activity and high pressure signals which initiate purge valve closure each meet the single fatture criterion? Is each channel of these systems continuously recorded?

Describe the type and location of local indications.

P. Evaluate the ability of the rod position indication to provide the operator with continuous information on the reactivity s tatus of the cere. For example, what information is indicated in the control room To what by means of readouts, indicator lights, print outs, etc.?

extent do the pulse counter digital readouts back What up theerrors readouts night an fron the differential transformer transmitters?

oe led to make as the result of a single failure in the rod operator position indication 1 Q. The maximum number of RCC assemblies which can be moved If the nuclear and their power speed will be determined by detail plant design.

level trip is unable to protect against the simultaneous withdrawal of all RCC assenblies, will interlocks be provided to prevent sicula taneous withdrawalt Will the interlocks meet the single failure criterion?

R. Evaluate the ability to supply electric power to engineered safety features under accident conditions from the incoming pcwer lines.

The evaluation should include but not be limited to the effect of abrupt loss of the Diablo Canyon plant, fault on the iccoming lines,.

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4 and faults and equipment failures in the plant or substatten.

Specifically address the problems associated with the use cf a singic startup transformer. Are the buses containing the enginetred safety f eatures (buses number 2 and 3) normally supplied from the star tup transformer or must they be transferred from the main unit to the s tartup trans former under accident conditions? Describe the motor operated breaker in the main bus cad describe the signal for operation and tiee-to open characteristics.

S. Evaluate the ability of the electric system to supply powet for emergency core cooling and other engineered safety features under conditions of a najor reactor accident with concurrent loss of of f-site power and a single f ailure in the on-site electric system. The evaluation should include but not be limited to the fc11 ewing:

1. A specific description of the sequence af ter a major a:cident with a 1 css of external power . State which circuit breakers are cpen and which closed, the sequence of operations , the time ir.tervals involved, which buses are isclated, which buses are connected together, which generators supply en:h bus, and the instrumentation and relaying which initiates the various operations.

2. Saeo as (1) except with a single failure in the en site electric system. Include, but do not limit the discussion to failure of any diesel engine to start, failure of a supply or load circuit breaker, sequencer failure, bus fault, and failure in the d-c control circuit. Specifically, address the question of whether reliability is lost by connecting the engineered safety features to buses which also supply other loads.

3. Ratings of the diesel generators and the station batteries.

4. List each component of the engineered safety features and its load requirenent.

T. Evaluate the ability to provide power to engineered safety features from off-site and on site sources with any single failure in the d c systee which supplies control power for circuit breakers, valves. etc.

Discuss more specifically how the d c buses. are arranged to supply alternate power sources for systems where redundancy is employed.

U. Describe the switching sequence to energize the boric acid pumps and the mator driven emergency feedwater pumps in the event of the loss of of f site power, including design provisions to meet the singic failute criterion.

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