ML20003E187

Forwards Revised Response to NRC 810225 Request for Info Re Major safety-related Civil Structures,Mechanical Sys & Components & Electrical Sys & Components.Response Supersedes Info Submitted 810325

ML20003E187
Person / Time
Site:	Mcguire, McGuire
Issue date:	03/31/1981
From:	Parker W DUKE POWER CO.
To:	Harold Denton, Youngblood B Office of Nuclear Reactor Regulation
References
NUDOCS 8104020454
Download: ML20003E187 (11)

Text

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s-DUKE POWER CO>iPANY l Powra Et:n.orwo 422 SocTn Catacu S tazzT. CruatoTTr. N. C. asa42 m u a o. .. u c a.s a. March 31, 1981

%CE Pets 60t=T TC.te-C=t.AeE4 704 Setano PeoDwet os. 3 ? 3- 4 53 Mr. Harold R. Denton, Director .

Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission ^

Washington, C. C. 20555 ,

~

Attention: Mr. B. J. Youngblood, Chief .

Licensing Projects Branch No. 1 -

u _$

Re: McGuire Nuclear Station 4* .s Docket Nos. 50-369, 50-370

Dear Mr. Denton:

The purpose of this letter is to transmit a revised response to Mr. Robert L. Tedesco's letter of February 25, 1981. This ;'esponse supersedes that transmitted in my March 25, 1981 letter. Attachment 1 te'lects the agree-ments reached in a conference call with representatives cf t* t Quality Assurance Branch and Duke Power Company on March 28, 1981.

If there are further questions regarding this matter, please advise.

Very truly yours.

0 hA(3 William t). Parker,Jr.>

h GAC:scs cc: T. J. Donat Senior Resident Inspector-McGuire Nuclear Station l

1 81040204-54

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ATTACHMENT 1 McGuire Nuclear Station Response to NRC Request for Information Transnitted by Mr. Robert L. Tedesco's Letter of February 25, 1981 Tables 3.2.1-1, 3.2.2-; and 3.2.3-1 are tabulations of the major safety-related civil structures, ruechanical systems and components and electrical systems and components which are part of the McGuire Nuclear Station. These tables are not intended to be a listing of every safety-related component in the station.

In many instances, various safety-related equipment are not listed in these tables. This is for one of several reasons; (1) The safety-related aspects of the equipment is discussed in other sections of the FSAR; (2) System flow diagrams depict the safety-related status of the stru. cure, system or component, (3) t he level of detail in the FSAR does not addresr. a particular component.

With regard to the Quality Assurance Branch, " Request for Additional Information" 260.1, numerous items contained in the request are not identified as safety-related in Tables 3.2.1-1, 3.2.2-2 and 3.2.3-1 either because of one of the three reasons stated above or because they are not appropriate for inclusion in the tables, i.e. , not a safety-related structure, system or component.

Part a. Response:

The following items are considered safety-related and are included in the FSAR.

A new table will be added to the FSAR or existing tables will be modified at a future date to indicate that these items are subject to the provisions of the Duke Power Company Quality Assurance program as outlined in the Quality Assurance Topical Report - Duke 1-A. (Numbers correspond to those in the

• request for information):

12) Fuel asse 'olies.

13) Cont.ol rod assemblies.

14) Core support structure.

15) Reactor vessel internals other than items 12, 13, and 14 above.

16) Control rod drive mechanisms.

17) Pressurizer spray nozzles.

18) Steam generator steam flow restrictors.

20) Sampling system lines connected to safety system components up to and including the isolation valve or block valve.

21) Spent fuel storage racks.

22) On-site Power Systems (Class IE)

a. Diesel generator package including auxiliaries (e.g. , governors, voltage regulator, excitation system).

b. Protective relays and control panels.

c. Containment electrical penetration assemblies.

d. Other cable penetrations (fire stops).

23) DC Power Systems (Class IE)

a. Battery racks.

b. Protective relays and control panels.

25) Tornado missile protection for the refueling water storage tank.

28) Essential heating and ventilation. (Control Room and Diesel Generator Ventilation)

29) Missile barriers protecting safety-related equipment, including missile barriers around air intakes, vent stacks, and other outside structures as applicable.

33) Non-safety systems that penetrate containment and are an extension of the containment boundary up to and including the containment isolation valves.

34) Residual heat removal system piping, spray headers, spray nozzles, and related supports.

35) Containment spray systee. piping, spray headers, and related supports.

36) Annulus ventilation system ducting and supports.

37) Ice basket supports.

38) Supports for containment isolation system valves.

39) Ducting and supports for the containment air return and hydrogen skimmer system.

40) Ducting, rupports, dampers, and operators for the ES' air handling units of the auxiliary building ventilation system.

42) Containment pressure indication systems (with input to ESF).

43) Containment sump lesel indication systems.

The following three items are considered safety-related with some qualifying explanation. These items will be included in the aforementioned table subject to the qualifying remarks below.

2) Masonry walls per IE Bulletin No. 80-11.

Masonry walls or " block" walls are treated as safety-related if, a) they provide support to any safety-related components or piping, or b) failure of the wall could potentially threaten other safety-related structures, systems or components.

24) Expendable and consumable items necessary for the functional performance of critical structures, systems, and components (i.e., weld rod, fuel oil, boric acid, snubber oil, etc.).

These items are treated as safety-related although they are not spe-cifically identified in the FSAR since the FSAR does not go to this level of detail. These items are included in the more detailed ,

" Nuclear Safety Related - Structures, Systems and Components" list which is the working document used at the station.

, 32) Biological shielding within contai ment and auxiliary buildings and other radiation shielding.

Installed biological shielding consists of reinforced concrete walls and masonry walls. These walls are considered safety-related to the extent that their structural integrity is required under seismic conditions.

Most of the equipment associated with the remaining items is not considered safety related; however, selected activities involving these items are cur-rently subject to or will be made subject to pertinent requirements of the Appendix B Quality Assurance Program. The extent to which these activities are or will be controlled by the Quality Assurance Program is outlined briefly below for each item. These items will also be added to the FSAR table mentioned in the first part of this response.

1) Measuring and test equipment used for safety-related structures, systems, ,

and components.

9) Instrument storage, calibration and maintenance.

. As required by Appendix B, Duke Power Company has in placa measures to assure that measuring and test equipment used in activities affecting quality are controlled, calibrated and adjusted to maintain accuracy within specified limits.

2 3) Radiation monitoring (fixed and portable).

4) Radioactivity monitoring (fixed and portable).

5) Radioactivity sampling (air, surfaces, liquids).

7) Personnel monitoring internal (e.g., whole boev counter) and external (e.g., TLD system).

41) Hydrogen analyzer system.

Calibration of the equipment associated with items 3, 4, %. 7 and 41 which is utilized under-post accident conditions will be performed. The perti-nent requirements of the Appendix B Quality Assurance Program will be applied to this equipment to assure that these calibrations are acceptable; e.g.,' procedure co..trols, audits, etc.

6) Radioactive contamination measurement and analysis.

9) Decontamination (facilities, personnel, and equipment).

10) Respiratory protection, including testing.

11) Contamination control.

Items 6, 9,10 and 11 involve programmatic activities rather than measuring or monitoring equipment. To assure these activities are performed in a quality manner under post-accident conditions, safety-related procedures will be utilized. These procedures will be subject to pertinent require-ments of the Appendix B Quality Assurance Program.

19) Sampling system delay coils up to and including tne containment isolation valve.

Delay coils are no longer installed at McGuire in the sampling systems.

26) Meteorological data collection programs.

Meteorological instrumentation is calibrated periodically to assure the validity of the meteorological data as required by Technical Specifications.

Calibration of this instrusentation will be subject to the pertinent requirements of the Appendix B Quality Assurance Program. Procedures for use of the instrumentation under post-accident conditions (i.e. , those condicions covered by the Emergency Plan) will be controlled by the per-tinen. portions of the Appendix B Quality Assurance Program.

27) Post accident monitoring instrumentation.

Regulatory Guide 1.97 specifies criteria for various categories of post-accident instrumentation. These criteria include requirements for some instruments to be fully qualified and safety grade while other instru-ments are only required to be "high quality commercial grade." The current instrumentation at McGuire covers this entire range. Those instruments that were designed and installed ac safety-related are so designated and maintained accordingly.

The calibration of all this instrumentation will be subject to the per-tinent requirements of the Appendix B Quality Assurance Program. Proce-dures governing the use of this instrumentation are designated safety-related and are also subject to pertinent requirements of the Appendix B Quality Assurance Program.

30) The Leak Detection System discussed in FSAR Section 5.2.7 should be explicitly identified, or all of its constituent parts should be included as sub systems or components of other entries on the Q-list.

The instrumentatica used for detection of Reactor Coolant System leakage is identified in the FSAR. It will be further identified in the FSAR in the new table mentioned earlier in this response. This table will indicate that the calibration of this instrumentation will be subject to the pertinent requirement of the Appendix B Quality Assurance Program.

31) Pressurizer relief piping from pressurizer to the pressurizer drain tank.

Piping between the pressurizer and the pressurizer safety valves and power operated relief valves is Class A (Class I) and as such is safety-related. The piping between the safety / relief valve discharge and the pressurizer relief tank is Class E (nonsafety) and, therefore, is not included within the purview of the QA program (See Table 3.2.2-3 in the FSAR for a list of piping classification and criteria).

Part b. Response:

1) Clarify that the Q-list includes the pressurizer PORVs and associated block valves (including their actuators).

The pressurizer PORV's, associated block valves and their actuators are safety-related components which are subject to appropriate QA requirements.

2) Clarify that main steam isolation valves are included on the Q-list. It is not clear from the designation "D" in Table 3.2.2-2 whether matn steam isolation valves are "Q required."

The main steam isolation valves are safety-related components which are subject to appropriate QA requirement. (The "D" in Table 3.2.2-2 is a typographical error, it should be "X").

3) Clarify that main steam piping (S.G. to MSIV) is included on the Q-list.

Main steam piping between each steam generator and its associated isola-tion valve (MSIV) is Duke Class B piping and is subject to appropriate QA eequirements.

4) Clarify that the containment sump, sump screen, and vortex suppression devices are included on the Q-list.

The containment emergency sump components were designed and installed as safety-related components and are maintained as such.

5) Clarify what is meant by " selected" in Table 3.2.3-1. When applied to valve motors or solenoids, exempted valves should be justified specifically by their function (s) or lack thereof.

" Selected"' valve motor operators or solenoids in Table 3.2.3-1 refers to those operators attached to valves that serve a . safety function and are con-nected to a Class IE power source. Valve operators in this category are classified as safety-related and are subject to appropriate QA requirements.

6) Identify the safety-related instrumentation and control systems to the same scope and level of detail as provided in Chapter 7 of the FSAR.

Table 3.2.3-1 and the various sections in Chapter 7 describe'the instru-mentation and control systems in sufficient detail. McGuire station drawings, procedures and equipment listings provide the deta 1 necessary for determining the safety-related status of individual switchgear, area termination cabinets, switches, relays, etc. A footnote will be added to Table 3.2.3-1 to indicate that the instrumentation described in FEAR Sections 7.2 through 7.6 is. subject.to pertinent requirements of the

. Appendix B Quality Assurance Program.

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7) Clarify that " Containment" (Table 3.2.1-1) includes (a) Personnel access hatches and associated seals, valves, piping, and tanks.

(b) Equipment access hatch and seals.

(c) Divider barrier seal.

The criteria for the personnel airlocks (including associated seals, valves, piping and tanks), equipment access hatch and seals and the containment divider barrier seal are discussed in Section 3.8.2 of the FSAR. All of these components are or will be considered safety-related and treated accordingly.

8) Clarify that operators of salves which require safety-related quality assurance also require safeuy*related quality assurance.

See response to item b.5. above.

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I l Part c. Response:

NUREG 0737 outlines requirements for operating plants and NTOL plants which must be implemented according to specified schedules. NUREG-0737 does not i address whether an item is safety-related and does not imply that the items should be on a "Q list".

Several items in NUREG-0737 involved reviews or upgrading of procedures, plans, programs, etc. For these items, it is inappropriate to require that they be safety-related. For some items, hardware changes were ende or new equipment was added. For those items identified as safety-related the Appendix j B Quality Assurance Program applies. For other items pertinent requirements of j the' Quality Assurance Program will be applied as discussed below.

J Several items in' NUREG-0737 have not been implemented for various reasons.

Appropriate quality requirements will be imposed when these items are implemented, i

i i

Responses to each of the identified items are given below:

1) Plant Safety Parameter Display Console I.D.2 i

i This item is not installed. Quality assurance criteria appropriate to j the method of implementation will be utilized.

2) Reactor Coolant System Vente II.B.1 This system was installed as a safety grade system. It is described in the document "McGuire Nuclear Station - Response to TMI Concerns."

This information will be incorporated into the FSAR at a future date.

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3) Plant Shielding II.B.2 I

i This item requirer a review of the accessibility of various station areas under-post-accident conditions. This review is not considered i

safety _related (See item 32 in Part a.).

~4) Post-Accident Sampling II.B.3 .

l' Two new sample panels have been designed to meet the requirements of

,this ites. These panels meet the criteria of NUREG 0578. ' Procedures for calibration and use of this equipment will be subject to pertinent requirements of the Appendix B Quality' Assurance Program.

i 5). Valve Position Indication II.D.3 l

This' item is discussed in the' document "McGuire Nuclear Station -

i Response'to TMI Concerns" and is identified as safety-related.

_ _ . _ _ _ _ _ _ _ _ _ _ .___ _ __ _ _ . _= . . . . . ._ _ _ _ . _ . . _ _ . - _ _ . _ _ _ . _ ___

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6) Auxilia ry Feedwater System II.E.1.1 The Auxiliary Feedwater System (AFS) is a safety-related system.

Item II.E.1.1 required that 1) a reliability analysis of the AFS be ,

, performed, 2) a review of the AFS against the criteria in SRP 10.4.5, be performed and 3) the AFS flow rate design bases be re-evaluated.

These reviews and evaluations were performed.

7) Auxiliary Feedwater System Initiation and II.E.1.2 Flow (Indication)

The automatic initiation circuitry for the AFS is safety-related. The flow indication which was added as a result of Ites II.E.1.2 is safety grade and will be maintained accordingly.

8) Emergency Power for Pressurizer Heaters II.E.3.1 Two banks of pressurizer heaters are provided with pcwer from the on-site essential power system. This power system is safety-related.

9) Dedicated Hydrogen Penetrations II.E.4.1

.This requirement is not applicable to McGuire.

10) Containment Isolation Dependability II.E.4.2 t

The McGuire containment isolation valves and their associated controls are safety-related. ,

11) Accident Monitoring Instrumentation II.F.1 See response to Item a.27.

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12) Instrumentation for Inadequate Core Cooling II.F.2 See response to Item a.27.

13) Power Supplies for Pressurizer Relief Valves, II.G.1

-Block Valves and Level Indicators The power supplies for this equipment consist of station emergency and ,

vital busses which are identified and treated as safety-related.

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14) Automatic PORV Isolation II.K.3.(1)

15) Automatic. Trip of Reactor Coolant Pumps II.K.3.(S)  ;

Neither of these functions is provided as part of the McGuire design and no decision has been made to insta11'them.

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16) PID Controller II.K.3.(9) i .

l The requirement contained in this ites was to prevent the derivative action of the PID controller from opening the pressurizer power operated relief valve. This was done for McGuire.

17) Anticipatory Reactor Trip on Turbine Trip II.K.3.(12)

This trip, like all reactor trips, is processed through solid state i protection system which is a safety-related piece of equipment.

18) Power on Pump Seals II.K.3.(25) i The component cooling water pumps provide cooling water to the reactor coolant pump seals. -This system is provided with onsite emergency power so that in the event of a loss of offsite power cooling water to

the pump seals would be provided. Both the Component Cooling Water System and the onsite emergency power systep are safety-related.

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19) Emergency Plans III.A.1.1/III.A.2 Emergency plans involve administrative controls to assure that various actions are taken fa the event plant conditions warrant these actions.

The McGuire Emergency Plan is a controlled document and the implementing procedures for the Emergency Plan are designated as safety-related pro-cedures. These procedures and the Emergency Plan are subject to the per-tinent requirements of the Appendix B Quality Assurance Program.

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20) . Emergency Support Facilities III.A.I.2 NRC Generic Letter 81-10 dated February 18, 1981 outlines the NRC requirements for each of the emergency support facilities. This letter does.not require or imply that any of these facilities must be safety-related. None of the Duke emergency support facilities are considered safety-related.

21) Inplant Iodine Radiation Monitoring ,

III.D.3.3 This monitoring would be performed under post-accident conditions using portable survey instruments and silver zeolite cartridges as discussed in the document "McGuire Nuclear Station - Response to TMI Concerns." .

Calibration and use of these devices under post-accident conditions will be controlled by procedures subject to the pertinent requirements of the Appendix B Quality Assurance Program.

22) Control Room Habstability III.D.3.4 The Control Room Ventilation System at McGuire is considered safety-related and is treated as such.

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