ML20056B715

Responds to NRC 700626 Request for Util Responses to Comments Cited in ACRS

ML20056B715
Person / Time
Site:	Monticello
Issue date:	07/30/1970
From:	Dienhart A NORTHERN STATES POWER CO.
To:	Morris P US ATOMIC ENERGY COMMISSION (AEC)
References
NUDOCS 9102080414
Download: ML20056B715 (5)

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M I N N E A PO W S, M I N N E S OTA 95401 July 30, 1970

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>fa Dr. Peter A. Morris, Director

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jj Division of Reactor Licensing 4

? 'i United States Atomic Energy Commission g', ' y; q. ~ y Washington, D.C.

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Dear Dr. Morris:

MONTICELLO NUCLEAR GENERATING PLANT E-5979 Comments on ACRS Letter of June 15, 1970 i

This is in reply to your letter of June 26 requesting NSP to sub: nit our responses to each of the comments cited in the ACRS letter of June 15, 1970.

The following paragraphs constitute our response to the ACRS letter items.

1. Independent Stress Analysis Check - Work is proceeding on an independent review of the thermal stresses in the as-built piping within the drywell including the recirculating system and associated portions of the residual heat removal systen.

1 Included will be a review of the piping layouts, design assumptions, computer input data, and calculational methods.

In addition, one or two stress analysis computations will be ver-ified by'perfonning a complete and independent stress computation.

This check review will be performed by groups in the design organizations different from the groups that conducted the original designs.

'I'he work is scheduled for completion in the fall of f

1970, and the results will be summarized in a report, copies of which will be furnished to you.

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2. Hot Displacement Measurements - The movement of piping systems j

within the drywell resulting from thermal expansion during system

- i heat-up will be evaluated by the following actions:

a)

Record the as built locations and ambient temperature settings j

for pipe hangers on pipe in the drywell over two inches in deameter.

l b)

Record accessible hanger positions at one intermediate l

temperature (3OOOF to 4000F) for' piping larger than two inches in diameter on the recirculating, feedwater, stean., core J

t spray, control rod drive return, and high pressure coolant I

injection systems in the drywell.

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c)

Inspection will be made at the inte'hmediate temperature to i

j visually verify that adequate piping clearance exist and

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1 freedom of movement through penetrations is possible.

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j d)

Extrapolation to operating temperature of the results and a

i observations made in Item b.

l e)

Adjustment of hangers, if necessary, to bring supports within specifications.

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f)

Where possible observe and record displacements at full i

operating temperature.

i g)

Recheck cold hanger settings on accessible hangers at the completion of the first total thermal cycle of the plant.

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3. Hich Points in Non-Flowino Parts of Primary System Pipino i'

A study has been initiated of non-flowing parts in the primary l

system piping aimed at disclosing pockets where gases could become trapped.

Systems involved in this study include the l

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core spray, recirculation, standby liquid control, control rod drive, head spray and vent, jet pump instrument, and other miscellaneous instrument lines Based on the study work completed to date, certain corrective measures have been taken in the form of added vents.

The study is continuing and completion is scheduled for the fall of 1970, after which a report will be produced and copies of this report will be furnished to you.

4. Bioloaical Shield Study - A study is now underway to determine the effects of a safe-end to nozzle weld failure on the biological shield surrounding the reactor vessel.

Completion of this study is scheduled for the fall of 1970, and the report will be produced and copies furnished to you.

Any plant modifications required as a result of the study will be made at the first scheduled 1

refueling outage.

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5. In-Service Inspection - Your staff has discussed with us a schedule of inspections for safe-ends and other plant components f

involved by our modification program related to sensitized j;

stainless steel.

The attached table outlines the additions tol our in-service inspection program appropriate to our modification fL i

progran.

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6. Leak Detection - The performance of the reactor coolant leakag k, Wib:;

detection system, including the speed and sensitivity for ik will be evaluated during the first eighteen months of plant W operation and the conclusions of this evaluation will be

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reported to the AEC.

Modifications to the leak detection v-

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if any are required, will be performed during the first scheduled refueling outage following the evaluation.

In addition, we will continue to study other techniques for detecting leaks and examine techniques for appropriate disposition in regard to the Monticello facility.

Yours very truly, o

Arthur V. Dienhart Vice President - Engineering Subscribed and sworn to before me This 30 day of July

, 1970 YV l

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Robert E. Hessian Notary Public, Hennepin County, Minnesota My Commission Expires May 15, 1976 (Notarial Seal) i i

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_________.___.__.__________._._______-____.___.__-_____m

SUPPLEMENTAL IN-SERVICE INSPECTION PROGRAM c

' Examination Area Examination Method Inspection Interval l

Field-clad safe-ends PT; UT One safe-end at or within first refueling outage Other safe-end at or within second refueling outage Welds of field-replaced safe-ends Pr; UT/RT Two safe-ends at or within first refueling outage Two safe-ends at or within the sece ~~',

refueling outage Severely weld-sensitized heat-PT; UT/RT 10% at or within first refueling outagd affected zones in wrought piping (Type 304, 316)

Field clad-repaired sensitized Visual Examination to the extent practicable components within reactor vessel at the first and second refueling whose failure could adversely outage, and whenever the reactor affect safety vessel head is removed for other purposes, within the first and second refueling i

PT - Liquid Penetrant Examination f

RT - Radiographic Examinetion UT - Ultrasonic Examination t

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