ML20093K612

Forwards Evaluation of Environ Impacts for Requested Exemptions to Apps a & J.No EIS Should Be Prepared Since NSHC Involved

ML20093K612
Person / Time
Site:	Catawba
Issue date:	10/16/1984
From:	Tucker H DUKE POWER CO.
To:	Adensam E, Harold Denton Office of Nuclear Reactor Regulation
References
NUDOCS 8410180078
Download: ML20093K612 (8)

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DUKE POWER GOMPANY P.O. BOX 33189

' CHARLOTTE, N.C. 26242

. HAL B. TUCKER. ntzaanomc 2.~1 October 16, 1984 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.- S. Nuclea'r Regulatory Commission Washington, D. C.20555 Attention: ~Ms. E. G. Adensam, Chief Licensing Branch No. 4 Re: Catawba Nuclear Station Docket Nos. 50-413 and 50-414

Dear Mr. Denton:

On. September 19 and 21, and October 2,1984, Duke Power Company requested that' exemptions to 10 CFR 50, Appendix A, GDC 1, 2 and 4 and Appendix J, Paragraph III.D.2(b)(ii) as granted in Facility Operating License NPF-24 be extended for varying periods of time. In accordance with 10 CFR 51.41, attached is Duke Power's evaluation of. environmental impacts of the requested exemptions.

It is our conclusion that for each requested exemption no~ significant

. hazards consideration is involved, each exemption meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9), and therefore,pursuantto10CFR51.22(b)noenvironmentalimpactstatement or environmental assessment need be prepared in connection with granting of these exemptions.

Very truly yours, t

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al B. Tucker ROS: sib Attachment-cc: . Mr. James P. O'Reilly, Regional Admi11strator U. S. Nuclear Regulatory Commission \

Region II

'101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30323 NRC Resident Inspector Catawba Nuclear Station 8410100078 841006 PDR ADOCK 05000413 P PDR d2[

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Mr. ' Harold R. Dent n, Director.

iPage Two cc:1 Mr. Robert Guild, Esq.

Attorney-at-Law P. O. Box 12097 Charleston, South Carolina 29412 Palmetto Alliance

.21351-Devine Street . .

LColumbia, South Carolina 29205

. Mr. Jesse L. Riley.

Carolina Environmental Study Group 854 Henley. Place Charlotte, North Carolina 28207 0

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m Attachment 1 Duke Power Company Catawba Nuclear Station Proposed Exemption to 10 CFR 50, Appendix J Evaluation of Environmental Impacts Description of Requested Exemption

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On June'25, 1984 Duke requested a change to the Catawba Unit 1 Draft Technical Specification 4.6.1.3.6 related to test requirements for the containment personnel air locks. The requested change involved an exemption to paragraph III.D.2(b)(ii) of 10 CFR 50, Appendix J.

Paragraph III.D.2(b)(ii) of Appendix J states:

" Air locks opened during periods when containment integrity is not required by the plant's Technical Specifications shall be tested at the end of such' periods at not less

.than Pa."

Whenever the plant is in cold shutdown (Mode 5) or refueling (Mode 6),

. containment integrity is not required. However, if an air lock is

. opened during Modes 5 and 6, paragraph III.D.2(b)(ii) of Appendix J requires that an overall air lock leakage test at not less than Pa be conducted prior to plant heatup and startup (i.e., entering Mode 4).

The existing air lock doors are so designed that a full pressure, i.e.,

Pa=14.7 psig(, test of an entire air . lock can only be performed afterstrongback Strongbacks are needed since the pressure exerted'on the inner door during the test is in a direction opposite to that of the accident pres-sure direction. Installing strongbacks, performing the test, and removing strongbacks requires at least 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> per air lock (there are 2 air locks) during which access through the air lock is prohibited.

If.the periodic 6-month test of paragraph III.D.2(b)(i) of Appendix J and the test required by paragraph III.D.2(b)(iii) of Appendix J'are current, no maintenance has been performed on the air lock, ano air lock is properly sealed, there should be no reason to expect the air lock to leak excessively just because it has been opened in Mode 5 or Mode 6.

It isseal the therefore leakage concluded that the test of para proposediii) h III.D.2(b) ap(proach of pressure for the full substituting

. test of paragraph III.D.2(b)(grapii) of Appendix J is acceptable when no maintenance has been performed on an air lock. Whenever maintenance has been performed on an air lock, the requirements of paragraph III.D.2(b)(ii) of Appendix J would still be met.

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-Environmental Impacts The proposed exemption' grants the substitution of an airlock seal test

'for ~an airlock pressure test while the. reactor is in a shutdown or refueling mode. With Appendix J, the increment of environmental impact lis related solely to the potential increased probability of co'ntainment

._ leakage during an accident. .This could-lead to higher offsite and control room doses. However, this potential increase is very small, due to the

, added seal leakage tests and the protection against excessive leakage

. afforded by the other tests required by Appendix J.

As discussed above, performance of an overall air lock leakage test

. requires at least-6 hours per air lock during which access through the

. air lock is prohabited. Any access and agress to lower containment during testing of the lower air lock'will _ involve climbing through the emergency hatch between upper and lower containment. This would result in more contaimination in the upper containment which would usually be cleaner than the lower containment. Similarly access to the upper containment while testing the upper air lock would require passing through lower containment where radiation levels would be higher, thus increasing _radi-

--ation exposure to personnel and increasing contamination in the upper containment.

With respect to non-radiological effluents, the personnel air locks play no role in mitigrating the consequences of events leading to non-radiological effluent releases.

Conclusion It is concluded that there are no adverse radiological or non-radiological environmental impacts associated with the requested exemption. A signifi-cant radiological benefit however would accured as a result of lower plant personnel radiation exposures.

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Attachment:

2 JDuke Power Company 6 .

' Catawba Nuclear Station

. Proposed Exemption to'10 CFR 50, Appendix A, GDC-1 Evaluation of Environmental Impacts

- Description of Requested Exemptioni

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_On July.13,i1984, Duke Power Company requested an exemption from certain

1. 1 requirements of.10 CFR 50, Appendix. A,' General Design Criteria (GDC) 1, f as such related Lto fuel load -and precritical testing activities. By (letter dated .0ctober 26, 1983, Duke proposed to upgrade the pressurizer

.PORV's"and steam generator PORV's to safety-related in order to comply

.with 'the . requirements of BTP RSB 5-1, that cold shutdown can be achieved

> from.the Control Room using only safety-related equipment. In order to fallow sufficent time to= implement these modifications, Duke requested

-that the modifications be ins.talled during f.he first refueling outage for the unit.

GDC-1 requires inpart-

" Structures, ~ systems, .and components important to safety

, 'shall-be designed fabricated, erected,-and tested to quality standards commensurate with the importance of the safety 1 functions to be performed."

In order-to mitiga'e the effects of~a~ steam generator tube rupture, diverse

_means=have been ovided to circulate, cool, and depressurize the Reactor

. Coolant' System i ohe Residual Heat Removal (RHR) System entry conditions.

Those functions m.y be achieved through the use of condenser dump valves,

- steam ' generator PORV's,L pressurizer PORV's, high-pressure auxiliary pressurizer spray and/or. nomal pressurizer spray. Not all of.the equipment is currently safety-related.' The steam generator PORV's are safety-grade valves with non safety-grade pneumatic operators,:and can be operated

~-locally to permi_t plant'cooldownc The high-pressure auxiliary pressurizer

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f spray has ~ been upgraded to safety-grade. - The pressurizer. PORV's are

.l safety-grade' valves with non safety-grade' pneumatic operators.

Environmental Impact The consequences of a steam generator tube rupture are discussed in

'Section 15.6.3 of the Catawba FSAR. Therefore operation of Catawba Unit 1 l during the first. cycle is ' addressed by.the current FSAR analysis. Granting of the proposed-exemption would not result in an increase in the amount of

. effluents that may be released ~offsite. The proposed upgrade of the pressurizer PORV's and' steam generator PORV's would facilitate the operator's

- response to a steam generator tube rupture event.

With respect to non-radiological effluents, the pressurizer PORV's and steam

- generator'. PORV's play'no role in mitigating the consequences of events

. leading to non-radiological effluent releases.

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.4 Conclusion' It is' concluded that there are' no adverse radiological or non-radiological environmental .-impacts -associated with the ~ requested exemption.

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Attachment 3 Duke Power Company Catawba Nuclear Station Proposed Exemption to 10 CFR 50, Appendix A, GDC-2 and -4 Evaluation of Environmental Impacts Description of Requested Exemption On July 11, 1984, Duke requested a partial exemption from the requirements of 10 CFR 50, Appendix A, General Design Criteria (GDC) 4 for the fuel load and precritical phase of the Catawba Unit 1 startup because all seismic qualification issues had not been resolved for the solid-state protection system. The Staff reviewed this request and determined that exemption to GDC-2 and GDC-4 were required.

By letter dated October 2, 1984, Duke proposed to take the following actions:

1. Prior to exceeding 5% power-a) The bolt through the existing electrical isolation one-piece fiberglass bushing will be tightened to

" snug tight" (as defined in the AISC Manual) and then tightened an additional'b turn.

b) The field mounting will be revised to ensure a

" snug fit", i.e., suitable shims will be placed between the bottom of the cabinet and the electrical isolation glastic pad to eliminate gaps and provide a uniform bearing condition.

2. A seismic test will be performed utilizing a generic mounting scheme with a glastic pad and fiberglass bushing for electrical isolation to verify the acceptability of the existing mounting.

This test would be completed by July,1985.

Therefore, final resolution of the qualification of the solid-state protection system involves a confirmatory seismic test and an exemption to GDC-2 and GDC-4 was requested until the test is completed.

Environmental Impact Based on the proposed cabinet mounting scheme discussed above, there is reasonable assurance that the solid-state protection system will be capable of performing its intended safety function in the event of a seismic occurance. The proposed exemption would not result in a change to any accident currently evaluated in the Catawba FSAR.

With respect to non-radiological effluents, the solid-state protection system plays no role in mitigating the consequences of events leading to non-radiological effluent releases.

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, Conclusion m

.It is concluded that there' are-~no adverse radiological' or non-radiological environmental-impacts associated with the requested exemption.

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