ML19345A254

Forwards Request for Addl Info Consisting of Reactor Sys Branch & QA Branch Questions from Review of Fsar.Fsar Should Be Amended to Reflect Responses to Requests

ML19345A254
Person / Time
Site:	05000147, Grand Gulf
Issue date:	10/02/1980
From:	Tedesco R Office of Nuclear Reactor Regulation
To:	Mcgaughy J MISSISSIPPI POWER & LIGHT CO.
References
NUDOCS 8010210320
Download: ML19345A254 (4)

Text

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UNITED STATES

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NUCLEAR REGULATORY COMMISSION 3

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OCT 2 1980

%..v Docket Nos.: 50-416/417 Mr. J. P. McGaughy, Jr.

Assistant Vice President - Nuclear Production tiississippi Power and Light Company P. O. Box 1640 Jackson, Mississippi 39205

SUBJECT:

REQUEST FOR ADDITIONAL INFORi.4 TION - GRAND GULF NUCLEAR STATION, UNITS 1 AND 2 As a result of our review of the infomation contained in the Final Safety Analysis Report for the Grand Gulf Nuclear Station, Units 1 and 2, we have Included are developed the enclosed request for additional information.

questions from the Reactor Systems Branch and the Quality Assurana Branch.

We request that you amend your Final Safety Analysis Report to reflect your responses to the enclosed requests as soon as possible and to infonn the Project Manager, Joseph A. Martore, of the date by which you intend to respond.

Sincerely,

%W Robert L. Tedesco Assistant Director for Licensing Division of Licensing

Enclosure:

As stated cc:

See next page 8olosleSW

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4 cc: RobertEf.'McGehee,'Esq.

Wise, Carter, Child Steen & Caraway-7 P.- 0. Box.651.

Jackson. Mississippi 39205

~ Troy B. Conner, Jr., Esq.

. Conner, Moore &'Corber 1747 PennsylvaniaiAvenue, N. W.

Washington, D. C.

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i Mr. Adrian Zaccaria, Project. Engineer Grand Culf Nuclear Station i

Bechtel Power Corporation Caithersburg,f4aryland -20760-4 L

Mr. Alan G. Wagner,' Resident Inspector P. O. Box 469

- Port Gibson, Mississippi 39150 Mr. John Richardson P. O. Box-1640 Jackson, Mississippi 39205 f

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Fire Protection - Quality Assurance Request for Additional Information s

Section C of Table 9A-1 of Amendment 41 of the FSAR for the Grand Gulf Nuclear Station, pl. 3 Unit Nos.1 & 2 indicates that formalized quality assurance requirements have not been Ep9A) applied to the fire protection system during design and construction, but that the fire protection system will be covered by applicable policies in MP&L's operational The fire protection sys-QA program following. turn over of the system to Plant Staff.

tem is not subject to the quality assurance requirements as defined in Appendix 9 to 10 CFR Part 50, since it does not directly prevent or mitigate the consequences of It is nonetheless true that proper functioning of the fire protection sys-accidents.

tem must minimize the adverse effect of fires upon structures, systems, and components important to safety, as made clear in Criterion 3 of Appendix A to 10 CFR Part 50. ~ Con-sequently, according to Criterion 1 of Appendix A to 10 CFR Part 50, quality assurance requirements should be applied to the fire protection program to an appropriate extent.

The extent of a suitable program is described in Section C of Appendix A to Branch Tech-This position provides that the quality assurance organi-nical Position APCSB 9.5-1.

zation manage the QA activities for fire protection and that 10 specific quality assur-ance criteria be applied to fire protection activities.

Although the design and construction of the fire protection system for Unit 1 is

18. 1980 conference call with MP&L) we believe that the QA

. completed, (Ref. September activities for fire protection should be appropriately applied to any ongoing or remaining portions of design and construction dealing with the fire protection system Accordingly, we request ifor Unit 2 as well as to the operational phase of both units.

you modify your, response in Section C of Table 9A-1 of the FSAR to address this position.

In' addition, describe the measures which assure that the QA activities that apply to fire protection for Grand Gulf Unit Nos.1 & 2 are under the management control of the Please identify the specific organization (s) that exercise this con-

-QA organization.

The description of the management control measures should address the activities trol. (a) formulating and/or verifying that the fire protection QA program elements in-for:

corporate suitable rer,uirements and are acceptable to the management responsible for '

7 fire protection; and (b) verifying the effectiveness of the QA activities for fire protection through review, surveillance, and audits.

(Performance of other QA program functions for meeting the fire protection program requirements may be performed bi qualified personnel outside of the QA organization.)

Also, describe the measures which assure the quality of purchased equipment and material when the supplier is not qualified under MP&L QA program (Ref. first sentence of note Describe the involvement-of the QA organization when material / equipment is 3-a).

identified as being non-safety related (Ref. second sentence of note 3-b).

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REACTOR SYSTEMS BRANCH 0211. 202

'Rcrerence and provide the lavout studies dnne tn assure that nn interferecce 4

(4.6.2.3.1.2) exists that will restrict the _ passage of control rods.

Also reference the pre-operational, tests that are used to show acceptable performance.

0211. 203 For the majority of events analyzed in Section 15, the (15.0) recirculation flow contrcl mode (automatic or mantal) asstmed in

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the analysis is not specified.

Our concern is that the mode selected mcy not result in the most severe margins on MCPR and peak vessel presstre.

a) Spe:ify the recirculation flow control mode asstmed for each event analyzo1 in Sc: tion 15.

b) Spe:ify the change in MCPR and peak vessel pressure for each event if the opposite re:iretdation flow control mode had been asstmed in the analysis.

Q211. 204 The resconse to item 221.3 indicates that 8x8 fuel bundles (15.0) with two water rods will be used at Grand Gulf instead of the 8 x 8 fuel btndles with one water rod.

a) Have the transients and a:cidents in Section 15 been evaltated with 8 x 8 fuel bundles using one or two water rods?

b)

If the transients and a::Idents in Se: tion 15 were analyzed with the one water rod fuel bundles, would any significant changes in MCFR, ceak vessel pressure, percent of rods experiencing boiling transition, and radiological consequences be expe:ted if.the two water rods design was used in the analyses? Dis:uss any changes to the above event parameters in quantitative terms.

C211 205 Explain why a s:rr. does not o::tr for the analysis of the " fast (15.3.2 3 3)' closure of one recirculation valve" transient in the Grand culf FSARg Q211. 206 (5.2.2)

Provide (6.3) assurance that your relief valve design is qualified (including testing after being subjected to an environment representative of an extended time period at normal operating conditions) to support yote asstaption that seven of the eight ADS valves will operate.

A. quantitative history of safety / relief valve operation, incitding i

similar valves in other plants, should be included in this evaluation.

0211. 207 If the air-supply line upstream of the ball check valve for all (5.2.2) non-ADS safety / relief valve air acetaulators were to break upstream of the b.all check valve, would there be an indication in f

the control rocm'of this break and an irdication of the acctmtdator riatus? If indications are given, what operator action would.. t required? If no indication is given,' justify why none is needed.

0211. 208 Soecify whether the fast scram system has been acenunted for in the (5.2.2) overpresstrization analysis?

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