ML19332A877

Forwards Request for Addl Info to Complete Evaluation of FSAR

ML19332A877
Person / Time
Site:	LaSalle
Issue date:	09/10/1980
From:	Tedesco R Office of Nuclear Reactor Regulation
To:	Abel J COMMONWEALTH EDISON CO.
References
NUDOCS 8009180515
Download: ML19332A877 (7)

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SEP 101980 Docket Nos.:

50-373 and 50-374 Mr. J. S. Abel Director of Nuclear Licensing Coninonwealth Edison Company Post Office Box 767 Chicago, Illinois 60690

SUBJECT:

REQUEST FOR ADDITIONAL INFORMATION FOR THE REVIEW 0F THE LASALLE COUNTY STATION, UNITS 1 AND 2, FINAL SAFETY ANALYSIS REPORT (FSAR)

As a result of our continuing review of the LaSalle FSAR, we find that we need additional information to complete our evaluation. The specific information required is listed in the enclosure.

Please inform us after receipt of this letter of the date you can supply the requested information.

Please contact us if you desire any discussion or clarification of the enclosed request.

Sincerely, Robert L. Tedesco, Assistant Director for Licensing Division of Licensing

Enclosure:

Request for Additional Information cc w/ encl.:

See next page c

80091806/8

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ENCLOSURE RE0 VEST FOR ADDITIONAL INFORMATION

'LASALLE COUNTY STATION, UNITS 1 & 2 DOCKET NOS. 50-373 AND 50-374 121.0 Materials Engineering Branch 121.17 High values of Charpy V Notch energy can be achieved at the nil ductility temperature for reactor vessel forging material.

You have stated that the RT f r y ur reactor vessel forging material NDT is assumed to be the nil ductility temperature as determined by the dropweight test. You did not supply any corroborative data or analyses to verify that this assumption would lead to conservative val' es of RT essel forging material. Consequently, before we u

NDT can complete our evaluation of compliance of the forging material, RT to the requirements of Paragraph IV.A.2.a of Appendix G, provide NDT additional data, information available in the literature, and/or analyses to verify the conservativeness of using nil ductility temperature alone to define the values of RT f r the vessel forging material.

NDT 121.18 Weld materials from weld seams 4-308, A, B, C and 3-308-A, B, C were not tested at temperatures that resulted in upper shelf energies greater than 75 ft-1b as required by Paragraph IV.B of Appendix G.

To demonstrate compliance with Paragraph IV.B of Appendix G, show that the six affected weld seams would have attained the minimum required upper-shelf energy if this weld material had been tested at higher temperatures.

To provide an acceptable basis for an exemption from the requirements of Paragraph IV.B, provide additional data, information from the literature, and/or analyses to demonstrate that a margin of safety equivalent to that of the requirement has been attained. The additional data may be from tests of similar welds; that is, welds of similar base material made by F

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. using the same weld wire, flux, welding process and heat treatments

'as'those in weld seams 4-308 and 3-308. Since one group of affected welds (seams 3-308) is represented in the reactor vessel materials surveillance program, you should test the baseline specimens at temperatures sufficient to define the upper-shelf energy, or to at least demonstrate that the 75 ft-lb requirement is met.

121.19 Identify the fabrication codes (edition and addenda) and the specific paragraphs in these codes that specify the fracture toughness require-ments and acceptance criteria for weldments and base metals. Codes and code paragraphs should be identified for all materials which constitute J

part of the containment pressure boundary (e.g. drywell head, piping penetrations, personnel airlockr. equipment hatch).

121.20 Specify the materials test data that certify that the fracture toughness acceptance standards have been met for each of'the identified materials in the containment pressure boundary.

330.0 RADIOLOGICAL ASSESSMENT BRANCH i

331.32 Provide a summary of the shielding design review results required by our letter dated November 9,1979, implementing the Lessons Learned

' item 2.1.6.b of NUREG-0578, and provide a description of the results of this review.

Include in your description:

a.

source terms used in the evaluation (NUREG-0578 specified that -

source terms in Regulatory Guide 1.3,1.4 and 1.7 be us,ed).

l b.

a listing of systems assumed in your analysis to contain high levels of radioactivity. in a post-accident situation including, but not limited to, containment, residual heat removal system, safety injection systems, CVCS, containment spray recirculation system, sample lines and gaseous radwaste systems, (or equivalent i

of these systems).

If any of these systems or others that could contain

.f high level radioactivity were excluded, explain why such systems F

were excluded from review. Verify that direct radiation from i

field-run piping and scattered radiation (such as shine over shield walls) were included in the analysis.

c.

Your April 1930 response to item 2.1.6.b identified four areas -

control room, remote shutdown panel, sampling stations, and the-Technical Support' Center - as being vital for persannel access after an accident.

Your evaluation to determine the necessary vital areas should include but not be limited to, con ideration of the control room, Technical Support Center, Operational Support Center -(see letter dated April. 25, 1980, D. G. Eisenhut to all power reactor licensees.which allows substitution of an onsite TSC with offsite TSC), sampling and sample analysis areas, motor control centers, instrument panels, emergency power supplies, security center, and radwaste control panels.

Include in your 1

i response pos c-accitient dose rates for all vital areas.

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d.

designation of the codes used for analysis, such as ORIGEN, IS0 SHIELD, QAD or others.

a brief description of the proposed plant modifications resulting 7

e.

from the design review and confirmation that these modifications will

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be complete by January 1,1981, or full power, whichever is later.

These modifications must be sufficient to provide for vital system 1

operation and for occupancy of the control room, TSC, sampling station, I

and sample analysis area..For other modifications, to allow access to l

areas where access would be.useful.but not vital, you should specify the anticipated modification and the scheduled completion date for mod i f ica tion.- Nate that the control room modifications required by our letter of May 7,1930, r ust be ccapleted by January 1983.

- 331.33 Your response to Lessons Learned item 2.1.8.b/II.F.1 of fiUREG-0578 concerning high-range containment radiation monitors is unacceptable.

Our position is the monitors should be~ located in a manner as to provide a reasonable assessment of area radiation conditions inside containment.

They should not be located in areas which are shielded (e.g., in steel sleeves) to such a degree that they are insensitive to low energy gammas (down to 60 Kev). The use of a grab sample system to supplement the containment radiation monitoring system for measurement low energy ymmas is unacceptable.

State how you plan to comply with the above criteria.

In addition, provide the following information:

1) Verification that the monitors will be rea'u. ibly accessible for replacement, maintenance, or calibration.

2) "erification that the monitors will be operable on January 1, 1931, or prior to full power operation, whichever is later.

3) A plant layout drawing showing the location of the monitors.

The nonitors should be widely separated so as to provide independent measurements and should " view" a large fraction of the containment volume.

4) Verification that radiation monitors will be calibrated using R

a calibrated radiation source for all decades below 10 /hr.

331.34 Concerning item 2.1.8.c/III.D.3.3 (Imp. 7ved In-Plant Iodir.e Ins tru-r.:entation) of fiUREG-0578, provide the fcilowing information:

l 1) the n.;mber and type of air samplers and portable SA;1-2's available for accident use.

2) the type of air sampling cartridges (such as silver zeolite) used.

3) a description of training and procedures provided for plant perscnnel to be able to determine the presence of airborne radioiodine in plant areas following an accident.

In addition, specify the location of the low-background, low contamination area used for iodine analysis following an accident. Also describe how you intend to purge sample cartridges of entrapped noble gases prior to cartridge analysis.

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• 331.35 The plant-organizational chart in Figure 13.1-4 shows the Rad /Chen L

Supervisor reporting through the Technical Staff Supervisor to the Administrative Assistant. The Rad /Chen Supervisor is also shown i

reporting on a lower level than the Senior Operating Engineer (ANSI

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equivalent is Operations Manager)..The draft document, " Criteria' for Utility Managenent and Technical Competence", specifies that:

1) the Radiation Protection Manager (RPM) (equivalent to your Rad / Chem Supervisor) should have a clear line of authority to the Plant Manager /

Assistant Plant Manager (or Ad'ninistrative Assist:nt in your organization);

2) the RPM should report at the same level as the Operations Manager (equivalent to your Senior Operating Engineer); 3) the RPM should be-a a member of PORC; and 4) the Radiation Protection Section should be separate from the Chemistry Section.

It is our position that you 5

make the above :hanges and revise your FSAR and proposed Technical-

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Specifications accordingly.

s 331.36 It is our position that the Rad / Chem Supervisor, H.P. Technical Support Personnel, and the Radiation Protection Foreman i

meet the minimum requirements of Regulatory Guide 1.8 (September 1975)." Personnel Selection and Training" which references ANSI 18.1 (1971). Provide updated resumes for the personnel who have i

been chosen to fill these positions with a breakdown of their qualifi-cations corresponding to Regulatory Guide 1.8/ ANSI 18.1 requirements (education, training, experience).. The experience referenced for.all above personnel-must be in the individuals speciality, which in this i.

case would be radiation protection.

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331.37 Based n information contained in the draft document " Criteria for Utiliti Management and Technical Competence", it is our position that L

your organization chain contain a qualified health physicist to provide l

backup in the event of the absence of the Rad / Chem Supervisor. The 3

Decenber 1979 revision of ANSI 3.1 specifies that individuals temporarily 1

filling the RPM position should have a B.S. degree in science or i-engineering, 2 years experience in radiation protect'on,1 year of which j

should be nuclear.pewer plant experience, 6 months of which should be on-site.

It is our position that such experience be professional experience.

Provide an outline of the qualifications of the individual l

who will act as the backup for the RPM in his absence.

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Mr. J. S. Abel Director of Nuclear Licensing Commonwealth Edison Company P. O. Box 767 Chicago, Illinois 60690 cc:

Richard E. Powell, Esq.

Isham, Lincoln & Beale One First National Plaza

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2400 Chicago, Illinois 60670 Dean Hansell, Esq.

Assistant Attorney General State of Illinois 188 West Randolph Street Suite 2315 Chicago, Illinois 60601 Resident Inspector, LaSalle NPS

.. U,.5, tiuclear Regulatory Commission P. O. Box 224 Marseilles, Illinois 61341 4

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