ML20148H878

Forwards Request for Addl Info Re Outstanding Safety Review Issues at Subj Facil

ML20148H878
Person / Time
Site:	Allens Creek File:Houston Lighting and Power Company icon.png
Issue date:	11/07/1978
From:	Varga S Office of Nuclear Reactor Regulation
To:	Oprea G HOUSTON LIGHTING & POWER CO.
References
NUDOCS 7811150032
Download: ML20148H878 (13)

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o UNITED STATES

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NUCLEAR REGULATORY COMMISSION q

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c WASHINGTON, D. C 20555

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h NOV 7 1978 Docket No:

50-466 Mr. G. W. Oprea, Jr.

Executive Vice President Houston Lighting 6 Power Company P. O. Box 1700 Houston, Texas.77001

SUBJECT:

OUTSTANDING SAFETY REVIEW ISSUES - ALLENS CREEK NUCLEAR GENERATING STATION, UNIT 1 We forwarded requests for supplemental information on December 29 1977; February 17, 1978; March 8, 1978; March 24, 1978; and April 14, 1978.

On the basis of our review of your responses f

provided in the Preliminary Safety Analysis Report as amended by amendments through Amendment 45, we identified a number of outstanding safety review' issues in the enclosure to our letter of July 21, 1978.

On the basis of our i

review of your responses provided in the Preliminary Safety l

Analysis Report as amended through Amendment 47, we have I

concluded that the items described in the enclosure continue to be outstanding review issues.

Within 10 days after receipt of this letter, please advise us of the date by when you can provide information that we need for resolution of the issues in the enclosure.

j Sincerely, e 1

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Stev

. Varga, Chief i

Lig Water Reactors Branch No. 4 Div ion of Project Management

Enclosure:

Request for Additional Information cc:

See next page

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E Ho,iston Light'ing a Power Conipany ccs:

Mr. P. A. Horn Project Manager, ACNGS Houston Lighting & Power Coi.ipany P. O. Box'1700 Houston, Texas 77001 R. Gordon Gooch, Esq.

Baker & Botts 1701 Pennsylvania Avenue, N. W.

Washington, D. C.

20000

'd. Gregory Copeland, Esq.

Baker & Botts One Shell Plaza Houston, Texas 77002 Jack R. Newnan, Esq.

Lowenstein, tiennan, Rei s 5. Axelrad i

102b Connecticut Avenue,fl. W.

Washington, D. C.

20036 Mr. Ra'y Matzelle Project Manager', ACflGS Ebasco Services, Inc.

1() Rector Street New York, New York 10005 Hr. Hay Lebre Project Manager, ACi4GS General Electric 17S Kurtner Avenue i

San Jose, California 9b125 Mr. Carlos Syars Th'e Houston Chronicle 801 Texas Avenue Houston, Texas 77002 froy Webb, Esq.

Assistant Attorney beneral Environinental Protection Division P. O. Box 12548 Capitol Station Austin, Texas 78711

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i bb REQUEST FOR ADDITIONAL 'INFORARTION l

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019 Auxiliary Systems 010.5 In Section 3.1.2.4.17.1 of the PSAR, " Evaluation Against Criterion 46," you state "The essential Services Cooling Water System will be designed to permit testing of system operability with simulation of emergency reactor shutdown or LOCA conditions and transfer between normal and emergency power sources."

Provide clarification that this commitment includes consideration of the coincident loss of the cooling lake as included in the design basis events (Section 9.2.5.3.1.2 of the PSAR) in the evaluation against the requirements of Criterion 46 of the General Design Criteria.

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l 110.0 Mechanical Engineering I

110.2 For the PSAR to be in agreement with the staff position L

as discussed in Standard Review Plan 3.6.1 and 3.6.2 and i

Appendix F to the Safety Evaluation Report, the first sentence in Section 3.6.2.2.4(g)(3) must be removed.

That sentence now states, " Piping welds subject to 100 percent volumetric inspection will be those short sections of process pipe themselves which serve as a part of contain-

ment, i.e., no guard pipe exists to contain the rupture i

in this section."

110.3 The staff position remains as stated in tne enclosure to and 110.4 our letter of July 21, 1978.

110.6(3) In your response to Item 130.20, provided by Amendment 110.17 No. 47 to your PSAR, you stated, "The applicant commits 130.20 to apply the generic resolution of this issue to the design of Allens Creek.

Howe"er, for cases where the generic l

resolution cannot be practically implemented such as steel

. plate structures within the containment boundary the applicant will justify the acceptability of the design B

to the satisfaction of the NRC staff."

Provide c);.rification that for each such case construction or installation will not be crmpleted until NRC staff approval of the i

justification has been obtained.

110.7 You have not provided sufficient information in the PSAR to enable us to complete our review

,f the design criteria J

Mechanical Engineering Cont..

to be used for supports for ASME Class 2 and 3 components.

Specifically, design criteria and loading combinations have not been provided for standard and plate and shell type supports in the balance-of-plant scope of supply and for all ASME Class 2 and 3 supports in the nuclear steam supply system scope of supply.

For ASME Class 2 and 3 linear supports in the balance-of-plant scope of supply the stress limits and the methods used to combine responses ~ as described in the PSAR are not completely acceptable.

The information is not sufficient to enable us to complete our review of the design criteria proposed for these supports.

In particular and for the information in Table 3.9.8 of the PSAR (1) the factor of 1.2 under an upset condition does not exist in NF, (2) no faulted limits are given and (3) verification that the faulted' buckling limit complies with F.1370(c) should be

,provided.

110.18 The staff in its generic review has not completed its review of recommendations 5,6 and 7 of the report ORNL/

SUB/2913.8.

Therefore, for Allens Creek, reliance or those recommendations is not acceptable at this time.

Revise your commitment to omit a dependency en those recommendations.

A commitment to the generic resolution which results,from the ongoing discussions between the NRC E.

Mechanical Engineering Con. _

and the BWR Mark II Owner's Group would also be accept.,le.

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130.0

'Struct0ral Engineering 130.6 On the basis of discussions with your representatives in a meeting in Bethesda, Maryland on October 30, 1978, we agreed to further review your proposed analysis methodology with modifications that you proposed to delineate in an amendment to your Preliminary Safety Analysis Report.

Generally, we understand that the methodology would be modified for frequencies less than 8 Hertz to utilize accelerations calculated using your methodology modified to accommodate a free-field Regulatory Guide 1.60 response spectra control motion at an elevation corresponding to the bottom of the reactor building mat.

In addition to a description of the modifications to the methodology and the bases as described in the meeting, provide the following additional information for our review:

(1)- For systems and components in the reactor building, auxiliary Building, and fuel handling building with l

natural frequencies less than 8 hz, we understand that you will increase the amplitude of the floor response

. spectra for design by a multiplier determined by the ratio of the Flush b/ Flush a response spectra within designated frequency ranges.

Provide the explicit criteria to be used to establish these multipliers.

The Flush b response spectra should be based on the use of the envelope of GAVE, GAVE *1.5, and G r the response spectra based on G AVE /1.5, AVE should be broadened by plus or minus 15%.

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Structural Engineering Cont..

(2)

In order to justify the use of the Flush a analysis,

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provide a comparison of the design shears and moments as computed by the Flush a analysis and the Spring an analysis for the (reactor building, including the shield building, steel containment, drywell, and

.t the RPV pedestal).

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y 211.0

_ Reactor Systems Branch 211.2 With regard to the postulated loss of a CRD pump (Item 211.2-07/27/78), our position remains:

(1)

You should provide the bases that will be used to determine that unacceptable impairment of control rod scram capability has occurred, (2)

You should provide automatic protection or demonstrate that 20 minutes is available for operator action, and (3)

Jou should describe initial and periodic test programs that will be used to demonstrate that this capability is maintained for plant lifetime.

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You should also describe how the leak performance of the h

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check valves will be continuously monitored via monitoring k

l of accumulator pressure, as stated in the response, i

211.3 Your response to Item 211.3 requires supplemental discussion.

t In particular, we note that this break size @.02 f' produces

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a peak cladding temperature in excess of the temperature 4

, produced by a large break DBA previously analyzed.

The following additional information should be provided.

(1)

Justify that the system provided for diversion of LPCI flow meets single failure criteria so that diversion before 10 minutes need not be considered.

(2)

Provide further justification that a diesel failure causing loss of the LPCS is more limiting than a loss of the LPCI for core cooling.

It is not apparent in your discussion that CCFL effects on e

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v Reactor Systems Branch Cont. l l

reflood times were included.

Discuss the relative effects of these low pressure systems upon the parameters in the LOCA calculations, e.g.,

reflood, core heat j

transfer.

(3)

Provide a sensitivity study showing peak clad temperature as a function of break size for small break LOCA's assuming diversion will be initiated at 10 minutes.

Perform this study for HPCS and recirculation line l

breaks.

For the most limiting break, provide the following figures:

(a)

Water level inside the shroud as a function of time during the LOCA (b)

Reactor vessel pressure vs. time (c)

Convective heat transfer coefficient vs. time (d)

Peak clad temperature vs. time (e)

ECCS flow rate vs. time.

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Justify that diversion at times greater than 10 minutes i

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will have less severe consequences than diversion at t

10 minutes (considering appropriate break sizes for t

later diversion).

(5)

Provide a discusssion which balances the need for 2

LPCI diversion for this break size (~. 0 2 f t )

with the need for abundant core cooling (GDC 35).

For example, this discussion could relate to Figure 6.2-3%

with regard to the likelihood of LPCI diversion for this size break.

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211.22 In your additional response in Amendment 47 to the PSAR fou stated, "The applicant will adopt the generic resolution of this issue for the ACNGS.

The applicant reserves the right to provide an acceptable alternative at a later date."

Provide a commitment that alternatives will be provided for staff review and will not be constructed

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or installed until staff approval in obtained.

211. 26 Additional information is needed relative to detection of leakage into the HPCS and the RCIC systems to either show conformance with the position of Regulatory Guide 1.45 or to demonstrate that such leakage does not need to be considered.

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363.6

.'Geosciences Branch 361.5' In Section 9.2.5.3.2 of the PSAR you state, "In the event that the rate of sediment accumulation is such that it appears that the allowable level of accumulation will be exceeded during the life if the plant, +he sediment will i

be removed before that allowable limit is reached."

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addition to level of sediment accumulation, limits on l

slope of the surface of the accumulated sediments should be considered to assure that unacceptable consequences will I

not result from sediment flow into pumps intakes during design basis events.

State the allowable configurations for 3

accumulated sediments within the cooling lake and provide a preliminary description of the technical specifications that will be used to assure maintenance l

of acceptable sediment configurations.

Include criteria, procedures, and technical specifications for maintaining sediment configurations.

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