ML18142C069

R. E. Ginna - 10/30/1975 Letter Transmittal of Request for Exemption from Certain Provisions of Appendix J to 10 C.F.R. Part 50

ML18142C069
Person / Time
Site:	Ginna
Issue date:	10/30/1975
From:	LeBoeuf, Lamb, Leiby & MacRae, Rochester Gas & Electric Corp
To:	NRC/SECY
References
Download: ML18142C069 (29)

Text

a NRC DISTRIBUTION FOR PART 50 DOCKET MATERIAL (TEMPORARY FORM)

CONTROL NO:

FILE' ROM. LeBoeu f, Lamb, Leiby-(('acR eDATE OF DOC DATE R EC'D LTR TWX RPT OTHER Washing ton,',D, C.

Lamb Leib 6

'eBoeuf MacR e '0-30-75 12-3-75 TO: ORIG CC OTHER NONE .. 0 SENT LOCAL PDR CLASS UNCLASS PROP INFO "

INPUT NO CYS REC'D DOCKET NO:

1 50-244' ESC R I PTI ON: ENCLOSURES:

Ltr. trans the following. ~ ~ "Request for )Exemption",.,'with'ttachment "A"

'nd."B", and Pij'. 1,, '+.' " Notarized 10-2%5 ~

Certificate'f Service .entitled "Request'or

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Bxempadon" '10-30-73,aexvad upon S'afety. 6 Lic. Board Panel;:U.S Chadmaan,~'tomic

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Rec'd'ion 12-3-75-(1 'cy' En'cl, Rec'd-) "

PLANT NAME RE Ginna 8 1 Ree'd no Orig, copy<<-

a FOR ACTION/INFC:RMATION VCR 12-3<<75 BUTLER (L) SCHWENCER (L) ZIEMANN (L) REGAN (E) REXD (L)

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Secretary U.S. Nuclear e3ulatory Commission

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Washington, D.C. 20555 Re Rochester Gas and Electric Corporation R. E. Ginna Nuclear Power Plant, Unit No. 1 Docket No. 50-244

Dear Sir:

Pursuant to Section 50.12 of the regulations of the Nuclear Regulatory Commission, we hereby transmit on behalf of Rochester Gas and Electric Corporation an original of a document entitled "Request for Exempt:ionN together with At-tachment:s A and B. By this request, RG&E seeks relief from cert:ain provisions of.'Appendix J to 10 C.F.R. Part 50. Two additional copies of t:his document are also transmitted for your convenience.

I A Cert:ificate of Service showing service of these documents upon the persons listed therein is also enclosed.

0 Very truly yours,

(>>(L( (, r,.i'gL;;~g, hLL>g '<i((C I<I.L LeBoeuf, Lamb, Leiby 6 LlacRae At:torneys for Rochester Gas and Electric Corporation Enclosures

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13EFORH THE UNXTED STATES cc.".'; '. '.yrJ )'1 NUCLEAR REGULATORY COMMISSION

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GAS AND ELHCTRXC Docket No. 50-244

..ROCHESTER

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CORPORATION (R. E. Ginna )

Nuclear Power Plant, Unit' No. 1) )

CERTII XCATH OP SHRVXCE X hereby certify that I have served a document entitled "Request for Exemption" by mailing copies there=-

of first class, postage prepaid, to each of the following persons this 30th day of Octobel ( 1975 Atomic Safety C. John Clemente,

'hairman, Hsg.

and Licensing Board Panel New York State Department U. S. Nuclear Regulai ory of Commerce Commission 99 A<zshington Avenue Washington, D.C. 20555 Albany, New York 12210 Atomic Safety and Licensing L. Dow.Davis, IV, Esq.

Appeal Board Office of the Executive U.S. Nuclear Regulatory Legal Director Commission t U.S. Nuclear Regulatory Washington, D.C. 20555 Commission Washington, D.C. 20555 tlr, Nichael Slade 1250 Crown Point Drive Edward Luton, Esp.

Webster, New York 14580 Atomic Safety and Licensing Board Panel Warren B. Rosenbaum, Hsq. U.S. Nuclear Regulatory One Hain Street East Commission 707 Wilder Building Washington, D.C. 20555 Rochester, New York 14614

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Thomas N. Reilly, Esca.

Atomic Safety and Licensing

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A. Dixon Callihan Union Carbide Corporation Board Panel P. O. Box X U.S..Nuclear Regulatory Oak Ridge, Tennessee 37830 Commission Washington, D.C. 20555 Mr. Robert N. Pinkney Supervisor, Town of Ontario Dr. Franklin C. Daiber 107 Ridge Road Nest Department of Biological Ontario, New Xork 14519 Sciences University of Delaware Newark, Delaware 19711 Dr. Emmeth A. Luebke Atomic Safety and Licensing Board Panel U.S. Nuclear Regulatory Commission Nashington, D.C. 20555 Hope M. Babcock LeBoeuf, Lamb, Leiby 6 MacRae Attorneys for Rochester Gas a'nd Electric Corporation

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9 OCT30 IS75 k Qilgo oi li.o Soccccary i~ociccii C Socolco Sociioo UNXTED STATES 01 AI4ERICA NUCLEAR REGULATORY COI1I4ISSXON Xn the matter of ).

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ROCIIESTER GAS AND ELECTRXC CORPORATION ) Docket Ho. 50-244 (R. E. Ginna Nuclear Power Plant, )

Unit No. l) ).

REQUEST FOR EXHIIPTXON Pursuant to Section 50.$ 2 of the regulations of the Nuclear Regulatory Commission, Rochester Gas and Electric Corpora-tion ("RGGE"), holder of Provisional Oporati'ng License No. DPR 18, hereby requests that it. be exempteil from certain provisions of Appendix J to 10 CFR Part 50. The specific exemptions requested are set forth in Attachment A to thi application. A safety c

evaluation which demonstrates thai. the proposed exemptions will not endanger life and property or the common defense and security and are otherwise in the public interest is set forth in Attach-ment B. The proposed exemptions would not authorize any change in the types or .any, increase in the amounts of normal plant effluents or any change in the..authoriz'ed power level of the facility.

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e NHLREFORE', Applicant respectfully requests that it be exempted from Appendix J to 10 CPR Part 50 as set forth in I

Attachment A..

ROCHESTER GAS AND ELECTRIC CORPORATXO~il eon D. Vlhite, Jr..

Vice President, Electric and Steam Production Subscribed and sworn to before me this 2 r>issins Expires I.'.arch 30, 19.77.

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ATTACHMENT A In 1969 Rochester Gas and Hlectri:c Corporation performed type A preoperat'onal containment leak rate .testing at Ginna Station. The results of that testing at 60 psig and at a reduced pressure of 35 psig are 'shown on figure 1. Given that Ltm and Lam are the preoperational reduced pressure and full pressure test leakage rates respectively, and that La is the maximum allow-able leakage rate, current 10 CFR 50 Appendix J regulations require

, that the acceptance criteria for subsequent, reduced pressure test-tm ~ 0.7. In the event that ing be L Ltm Lam 1

, provided that am ..

Ltm ~ 0.7, the subsequent acceptance criteria is to be La p a

Lam or the maximum allowable leakage rate times the square root of the ratio of the test pressures.

As seen on figure 1, our 1969 reduced pressure test yielded a negative leakage value. The value is small and its error band includes positive values as expected ur..der nearly 4

all circumstances for a valid test. Literal interpretation of the regulations, however, would require all of our successive reduced pressure tests to show a negative leakage result. Since this is clearly impossible, a more realistic approach to deter-mine an acceptance criterion is to reduce the maximum al'lowable leakage rate by a linear factor derive'd from the slope of the Page 1 of 7

~ g l -t il between the p

'ine perational test data po' with no regard 1 e ~

to their absolute'.e value (see,figure l). Positive .values for successive te t " would then be permis .iblo. For-RG&H's case, the resulting acceptance criteria would be approximately equal.

to that calculated using the ratio of the test pres ures formula.

Therefore, RGGH requests that an exemption from p'aragraph XIX. A.

'4. (a) (l) (iii) of Appendix J to l0 CPR Part 50 be granted which will allow use of the ratio of the test pressures acceptance for.

mula, La t 'here L is 0.2 weight percent per day/

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• is 60 psig, and Pt is the gauge test pzessure. This relationship vill alloO positive leakage rates for successive tests but. still

. vill maintain acceptable of .site accident do es as shown in prior safety analyses.

I Several points in the regulations appear to be subject to

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interpretation. As a result, inconsistencies may exist in the regulations or between the regulations and Ginna Station'pproved Technical Specifications. To resolve the following points, exemp-tions aze requested from Appendix J to 10 CE"R Part 50 if the Nuclear Regulatory Commis ion believes such exemptions are required (1) Paragraph XX.'K. of the regulations defines L as the maximum allowable leakage rate at pressure P, as specified for reo erational tests. in the technical speci-fication" or associated bases, and as specified for perio8ic tests in the operating license." (emphasis added)

The value approved for periodic testing and appearing in Ginna Station's Technical Specifications is 0.20 weight pezcent at 60 psig. Ilowever, the acceptance Page 2 of 7

value used in preoperat'nal 'testing was 0.1 weigh't percent at 60 psig. This was established with our supplier to ensure that the .0.20 percent leak-age rate requirement would be conservatively met.

Because our PSST safety analys6s assumed a 0.25 percent leakage, and the staff SER used a leakage rate of 0,20 percent. with acceptable offsite doses resulting from the calculations and because Ginna Station Technical Specifi-cation. have used at lea t 0.20 percent for 'post opei-a-tional tests, we intend to continue to use the 0.20 percent maximum allowable leakage rate appearing in Ginna Station Technical Specifications.

(2) Paragraph XI . N. define leaJ:age rates that are

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'obtained from testing the containment with co>nponents and systems in the state as close as practical to that.

which would exist. under design basis accident conditions."

Paragraph XXI. A. 1. (d), on the other hand, states that ll "fa] vented systems shall be drained of water or other to the extent necessary to assure exposure of the 'luids syst: em containment isolation valves to containment air

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'est pressure ..." This paraar'aph also states that

"[s]ystems that are normally filled with water and operat-I" ing under post-accident condition , such as the containment heat removal system, need not be vented."

Xn view of the differences in interpretation which may be Page 3 of 7

attached to the -e regulations, and with no technical specification covering these point., RGGE intend to pursue the following course of action.

(a) Venting Outside Contairuaent Linc'.s which pone'ate the containment and which are open to th containment atmosphere as in (b),'ill be vented to the atmosphere outside of the contain-meixt. bhere piping configurations outside contain-ment exist such that the fluid in fluid carrying lines does not drain to expose -t.he isolation valves to the atmosphere by opening existing vents and drains, the fluid will be 1eft in the lines.

(b) Ventinq Inside Containment Portions of the fluid system that are part of the reactor coolant pressure boundary and are open directly to the containment atmosphere under post-accident con-ditions and become an extension of the bounda"y of the containment will be .opened or vented to the containment atmo phere priqr to and during the test. Portions of closed systems inside containment that penetrate con-tainment and that also pass inside the primary shield wall near the broken leg, and which are postulated to rupture as a result of a loss of coolant accident will be vented to containment atmosphere. Lines which have t

never been postulated to rupture, consistent wit..h the containment integrity: analysis of section 14.3.4 of the PSM, will not be vented., Where. check valves or Page 4 of 7 .

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piping configurations exi st between the primary shield wall and the containment p'enetration or in .places where damage to the piping system is not postulated to occur as a rcsuli= of a LOCA such that fluid seals are formed as a result of norma3. operation and containment iso-lation, the fluid wi3.1 bo left undisturbed.

Thai is,. those portions of systems not postulated to rupi ure as the result of a LOCA will not be drained unless, they drain unaided to the postulated breaks in the systems.

(c) Isolation Valves Nhere two isolation valves exist in a single line which are either check valves, or valves capab3.e of automatic closure, or a combinai;ion thereof, no attempt will be made to vent to atmosphere from a point be-tween the valves.

I (3) Paragraph III. D. 2. of Appendix J to 10 CPR Part 50 states that "[a]ir locks shall be t=sted at 6 month intervals. However, air locks which are opened during such intervals, shall be tested after each opening."

Ginna Station Technical Specifications, on the other hand, require that "... the personnel air lock seals shall be tested at. 4 month intervals, except when the air locks are Page 5 of 7

not opened during the interval. Xn that case, the test is to be performed after each opening, except that no test interval i" to exceed 12 months."

Becdu~e the regulations do not say specifically how the testing is t:o be performed, because extensive testing.

after each opening of the air lock when multiple openings may ta)-e place in short time spans is impractical, and because of the proven reliability of these air locks,. RGGE intends to meet the intent of paragraph XXX. D. 2 of Appendix J to 10 CFR Part 50 by testing as follows.

Tl>e containment air locks are to be tested at inter-vals of no more than 6 months by pressurizing the space between the air loc): doors." (An application for an appropriate amendmcnt to Ginna Station Technical Specifi-

'cation will be submitted at'a later date.) Xn addition, following opening of the air lock door during the interval, a test will be performed by pressurizing between the dual . ".

seals of each door ~opened, within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> of the opening, unless the reactor was in the cold shutdown'ondition at the time of the opening or has been subsequently brought to the cold shutdown condition. Thus, in the event that several .

openings occur within a short period of time, one test within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> of the first opening will satisfy the requirements for lea) testing.

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0 (4) Paragraph XXX.13.1. give those test methods which are acceptable means o . performing periodic Type 8 tests.

One of the methods is to mea ure "the rate of pressure lo s of the test chamber of the containment pene-tration..." The method of testing preferred by RGGH is a measurement by means of a rotometer of the air floe to the test chamber which is required to maintain the chamber at the test pressure. Ne believe that this method produces accurate results and meets the intent of the r gulations.

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3."IGURE 1 43 ~ 33 END. DEI'T. sTAT Io'c: Ginna DATE: 7/7/75 IIDE1 OI. 1, I

Joo: Co>>tainrne>>t I.eak Bate Test INVADE D1:, P. Wilkens CK:

0. 26 0,24 Leakage Rate Assumed in FSAR

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1. ealcage Rate Assumed in SER l

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0. 16 I

0. 14 I s

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0. 06 1972 test

0. 04 (. 042+ . 020) 1969 test 0,02

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~ ~ w Safct i Evaluat1on The requested exemption will present no significant hazard to the health and safety of the public for the following reasons:

The proposed change in reduc d pressure test acceptance criterion does not alter the maximum acceptable leakage. rate.

The leakage rates used in the Staff's Safety Evaluation Report and the FSAR were 0.20 and 0.25 percent per day respectively.

Offsite dose calculations u ing these leakage rates demonstrated P

acceptable public exposures well below 10 C1;"R Part 100 values.

. An appropriate factor is to be applied to re'duce the acceptance criterion for reduced pressure testing. Although it is Qifficult to establish the relationship between 3.eakage rates' at Qiffc.rent test pressures for a specific containment, mass flow

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through orifices wi3.1 generally behave as a function of the square C

root of the differential pressure. Thus, in the absence of ex-

. tensive test data, the square root relationship is believed to be valid and will reduce the maximum allowable leakage rate assumed in

. the Staff's Safety Evaluation Report by, an appropriate. amount for reduced pressure testing. 'L ~ ~

Therefore, the overall effect of the requested exemption is to provide an acceptance criterion which has already been found acceptable in accident analyses. The limit on conf ainment leakage is such that there i" no undue risk to the health and safety of the public.

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0 ~ ~ Pg The proposod method of testing personnel air lock doors within the ix month intervals after an opening of a door will adequately insure the integrity of the doors by detecting damage

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to the seals which may have resulteQ during the opening of the air lock. Testing the air locks by p essurizi>>g between the seals will require approximately 15.'minutes whereas testing by pressurizing the entire access hatch will require approximately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. By not inhibiting entry anQ inspections inside the containment, i'f they are required, the alternate procedure tends to 'augment the afe operation of the plant.

The regulations specify that lines which rupture as the re-A suit of a loss of coolant accident should be vonted to the contain-ment atmosphere prior to testing. The lines which will be vented are selected based upon assumptions made in the .containment integrity analysis.

The proposeQ method of venting and draining systems penetrating the containment is consistent, we believe, with the intent of the regulations.

Thu , the test method" and procedures to be applied to meet the requirements of Appendix J to 10 CPR Part 50 will resuli in a valid test to determine the integrity I

of the reactor con-tainment. The acceptance standards for this test. have been shown to result 'n offsite doses, under postulated accident conditions, we'll within the requirements of 10 CFR Part 100.

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