ML20202B209

Forwards Info Demonstrating Util Compliance w/10CFR50.62 Re Reduction of Risk from ATWS Events.Sys Design & Ordered Equipment Needs to Be Finalized in Fourth Quarter 1986. Prompt NRC Review Required to Meet Schedule

ML20202B209
Person / Time
Site:	Calvert Cliffs
Issue date:	06/27/1986
From:	Tiernan J BALTIMORE GAS & ELECTRIC CO.
To:	Thadani A Office of Nuclear Reactor Regulation
References
NUDOCS 8607100239
Download: ML20202B209 (6)

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B ALTIMORE GAS AND ELECTRIC t-CHARLES CENTER R O. BOX 1475 BALTIMORE, MARYLAND 21203 JOSEPH A.TIERN AN M

Vict PatesOENT NucLEAn ENEnOY June 27,1986 p.S. Nuclear Regulatory Commission Office of N.uclear Reactor Regulation Washington, D.C. 20555 ATTENTION:

Mr. A. C. Thadani, Project Director PWR Project Directorate #8 Division of PWR Licensing - B

SUBJECT:

Calvert Cliffs Nuclear Power Plant Unit Nos.1 & 2; Docket Nos.'50-3f 7 & 50-313

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Compliance with 10 CFR 50.62, Reduction of Risk from ATWS Events

REFERENCES:

(a) 1.etter' from L. E. Salyards, to A. C. Thadani, same subject, dated May 1,1986 Gentlemen:

e Attachrnent (a) provides information. to demonstrate we meet the requirements of Paragraphs 10 CFR 50.62(a)(1) and 10 CFR 50.62(a)(2) as required by 10 GFR 50.62(a)(6). Paragraphs 10 CFR 50.62(a)(3) through (5) apply only to boiling water reactors and therefore have not been addressed.

As stated in Reference (a), we need,to finafize the system design and order equipment in the fourth quarter of this year. As such, prompt NRC Staff review is needed to meet the agreed-upon implementation schedule in Reference (a).

Should you have additional questions regarding this matter, please do not hesitate to

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contact us.

l Very truly yours, l

g Sw se i

D. A. B' rune, Esquire l

cc:

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3. E. Silberg, Esquire i*

Mr. D. H. Jaffe, NRC Mr. T. Foley, NRC g

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Mr. A. C. Thadani 2-June 27,'1986 bcc: Messrs.

R. F. Ash /R. C. L. Olson C. H. Cruse /P. E. Katz R. E. Denton/3. A. Mihalcik

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R. M. Douglass/T. N. Pritchett M. Gavrilas/E.1. Bauerels

3. R. Lemons /R.'P. Heibel W. 3. Lippold/A. R. Thornton F. 3. Munno

- R. B. Pond, Jr./R..E. Cantrell L. B. Russell /3. T. Carroll V. F. Stricklin (3) '

W. P. McCaughey, Jr.

W. R. Horlacher, III B. E. Holian P. E. McGrane M. E. Bowman m

L. E. Salyards M. D. Patterson/3. D. Palank

, C. R. Mahon/R. B. Sydnor.

W. C. Holston/3. 3. Ihnacik

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y Mr. A. C. Th?dani Attachment A June 27,1986 A-1 Complia6ce with 10 CFR.50.62', Reductibn of Risk from ATWS Events

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10 CFR 50.62(c)(1)

Each pressurized water reactor must have equipment, from sensor output to final actuation device, that is diverse from the reactor trip system, to automatically initiate the auxiliary (or emergency) feedwater system -

and initiate a turbine trip under conditions indicative of an ATWS.' This equipment must be designed to perform

. its function in a. reliable manner and be independent

.(from sensor output to the final actuation device) from '

the existing' reactor. trip system.

Reference:

Letter from D. M., Crutchfield, to R. C. L. Olson, dated June 16,1936

Response

We participate in the. Combustion Engineering, Owners Group (CEOG) ATWS Subcommittee in order to respond to the ATWS Rule od generic basis wherever:

possible. In this instance, we endorse the Subcommittee's position that the e.xisting turbine trip and auxiliary feedwater actuation systems of CE-designed PWRs meet the requirements of the rule. The reference'noted above provides a status report of the NRC Staff review of the CEOG conceptual design. Please refer to this letter and to the information provided at the meetings identified in the letter for our position regarding this aspect of th'e rule.

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The following is a brief description of our Turbine Trip System and Auxiliary Feedwater Actuation Systems (AFAS). Also, see.our Final Safety Analysis Report (FSAR) Chapter 7 for detailed plant-specific information on our Turbine Trip System and AFAS.

Urider conditions indicative of an kTWS the existing Turbine Trip System automatically initiates a turbine trip diverse from the existing Reactor Protection System (RPS). Four undervoltage relays sense the loss of power to the Control Element Drive Mechanisms (CEDM) and provide input to two-out-of-four logic in the Engineered Safety Features Actuation System (ESFAS) which generate the turbine trip signal.

Under conditions indicative of an ATWS the AFAS automatically initiates independent of the existing RPS. Section 7.10 o.f the FSAR provides a" detailed description of the AFAS.

10 CFR 50.62(c)(2)

Each pressurized water reactor. manufactured by Combustion Edgineering-or by Babcock and Wilcox must have'a diverse scram system from the sensor output to.

interruption of power to the control rods. This scram system must be designed to perform its function in a reliable manner and be independent from the existing

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reactor trip system (from sensor output to interruption

- of power to the control rods),

Response

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We again endorse the CEOG ATWS SubcommitteeTposition with respect to this aspect of the rule'and will install a new Diverse Scram System;(DSS). System specifications A

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Mr. A. C, Thadani Attachment A 3 tine 27,1986 A-2

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2 presented below should provide enough information to demonstrate the adequach,of this' new system in meeting the requirements of the rule.

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Safety Related (IEEE-279)

Our design of the DDS will be Safety Rsiaie2.'(SR)'frpm -

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sensor output to actuation relay coils? The acttfalion relay will proyide SR/NSR separation. The actustion -

relay contact, the cable run to the motor generator contactors, and the contactors will be NSR. Therefore, the design of the DSS exceeds the"NRC guidance!

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Redunciancy NRC guidance does not require redundancy.

Diversityfrom the Existin'g The DSS will utilize ESFAS logic and actuation.

Trip System,

equipmeo,t supplied by Vitro versus the RPS equipmeitt' supplied by Cembustion Engineering. Th,e: DSS will'use the CEDM motorgenerator output contactors to interrupt power to the OEDMs versus the reactor --

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protection breakers used in the RPS. Existing RPS sensors will be used to provide signals to the D.SS,l' however adequate means of isolation will be provided.

. The DSS design will consist of four sensor channels utilizing a two-out-of-four logic to generate a.re'a~ctor trip signal. Existing spare ESFAS logic modules, isolators and relays will be utilized. The output' contactors of the CEDM ' otor generators will open on. -

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receipt of this' reactor trip signal.

Electrical Independence The DSS will be electrically independent, separate, and' y from'the Existing. Reactor and diversh from the existing RPS from the se,nsor -

Trip System output ~ current loop.to and including the actuation circuitry.

Physical Separation from Not included as part of NRC guidan' e. Ho' wever, the c

the Existing Reactor Trip CCNPP design from the sensor output to the actuation System relay coils will be physically separate from the existing RPS and therefore will not violate separation criteria applied to the' existing RPS.

l Environment,al Qualification All DSS equipment except for the contactors are.

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qualified for accidents. The motor generator'csutput contactors are part of the existing Control Element Drive System' and will not be upgraded. "This exceeds NRC guidance.

s Seismic Qualification NRC guidance does not require seismi qualification.

Quality Assurance for Test, The Quality Assurance level for all DSS equipment is' Maintenance and Surveillance in compliance with 10 CFR 50 Appendix B requirements, which exceeci the requirements in,

l Generic Letter 85-06," Quality Assurance Guidance /for

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ATWS Equipment that is Not Safety Related".

i Safety-Relateif (IE)

The DSS will use four redundant IE Safety Related, Power Supply inverter fed, AC power sources shared by the existing 1

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Mr. A. C. Thadant Attac-f*nant A

. June 27,166

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w RP'S. Each inverter is po'wered from a IE DC bus and mch DC bus is backed up by a IE 125V DC station i -

battery. Tperefore, the DSS is fully capable of

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performing its safety function with loss o fisite power. Both the RPS and the DSS power s plies

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pre /will be isolated from safety busses. Th DSS will

- consist of four sensor channels and t,wo actuation

~ hannels. DSS sensor char)nel and actuation channel.

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- power supplies will be diverse from existing RPS s

W cfiannel pwer supplies.

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v Testability, at Power The DSS. Trip function will be bypasidd to prevent e

inadvertent actuation during testing at' power. The bypass condition will be automatically and. continuously indicat'ed in the main co.ntrol room.

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Inndvertent Actuation ~

The use of two-out-of-four logicg' trips and challenges to ill minimize the frequency.of inadvertent reactor

.other safeiy systems.

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Maintenance Bypasses i'he system'dgsign will permit bypass of the diverse trip function for all maintenance, repair, test or calibration of one channel during, operation to avoid inadvertent agtuation of prote.ctive action at the system level. The m

' use of maintenance bypasses will be restricted as governed by technical specification out of,servic6 times

'and limiting conditions for operation. _The bypass

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condition will be automatically and contfnuously iridicated in the*. main control room. The use of maintenance bypasses will be restricted as governed by administrative controls commensurate'witi; the safety significance of the system. As previously stated, 10 CFR 50 Appendix B requirements shall apply.

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Operating Bypasses There is no heed for bperat,ing bypasses. None pro'vided.

Indication of Bypasses

'If the protective action of some part of the' DSS has been by. passed or deliberately' rendered inoperative for any purpose, this fact will be continuously and'i '

automatically indicated in the control, room.

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Means of Bypassing A pefmanently installed bypass switch will be used for.

this purpose.

Completion of Protective Jhe design will be such that, once initiated, the Action Once it is Initiated protective action at the system level will go to e

completion. An operator initiated reset sighal will be provided for return to operation.

Manual Initiation Manual initiation of the DSS will be provided M several locations outsid,e the control room. Manualinitiation of

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the DSS is not' required in the control room because there already exists a means of manually, tripping the reactor from inside the control room that is diverse e

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e Mr. A. C. Thadani Attachment A

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June 27,' 1986 A te

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from the RPS inanual trip; specifically, pianually

' tripping the MG set input breakers.

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u Information Readout The* diverse reactor toip syst.em will be designed to provisi.e the operator with accurate,.ceanplete, and timely information pertinent to its own status.

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