Text

Forwards NRC Final Evaluations of SEP Topics II-1.B, Population Distribution & III-4.D,site Proximity Missiles. Site Conforms to Current Licensing Criteria
	ML17309A193
Person / Time
Site:	Ginna
Issue date:	07/21/1981
From:	Crutchfield D; Office of Nuclear Reactor Regulation
To:	Maier J; ROCHESTER GAS & ELECTRIC CORP.
References
	TASK-02-01.B, TASK-03-04.D, TASK-2-1.B, TASK-3-4.D, TASK-RR LS5-81-7-70, LSO5-81-07-070, LSO5-81-7-70, NUDOCS 8107240143
	Download: ML17309A193 (97)
	v • d • e

July 21',

1981 Docket No. 50-244 LS05-81-07-07O Mr. John E. Maier, Vfce President Electric and Steam Production Rochester Gas 4 Electric Corporation 89 East Avenue Rochester, New York 14649

Dear Hr. Maier:

SUBJECT:

SEP TOPICS II-1.B, POPULATION DISTRIBUTION AND III',SITE PROXIMITY MISSILES. - R E

GINNA Enclosed are the staff's ffnal evaluations of SEP:Topfcs II-l.B and III-4.0 for the R. K. Ginna Nuclear Power Plant.

These evaluations are based on our review ot your topic safety issessment reports sub-mftted by letters dated April 15, 1981 and April 16, 1981, respectively.

You wfll note that. we have revised your calculated population density which fs more properly obtained by dfvfdfng the total population within a gf ven distance by the total area of the complete cfrcle (fncludfng both level.and water) whose radfus fs the distance of interest.

This completes our evaluation of Topics II-1.8 and III-4.D.

These evaluations will be a basic input to the integrated;safety,'assess-ment for your facility unless you identify changes, needed.to. refleW the as-built conditions at your facility.

These, assessments may~,be revised in the future ff your facility desfgn fs changed. or ff=BRC.-.,

criteria relating to this subject are modified before the integrated assessment is completed.

Sincerely, 8107240143 810721 PDR ADGCN, 0 000244 P

POR

Enclosure:

~PAL)- 7+g Dennis M. Crutchffeld, Chief Operatfng Reactors Branch ho.

5 Division of Licensig

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Nr. John E. Maier CC Harry H. Voigt, Esquire LeBoeuf, Lamb, Leiby and MacRae 1333 New Hampshire Avenue, N. M.

Suite 1100 Mashington, D. C.

20036 Nr. Michael Slade 12 Trailwood Circle Rochester, New York 14618 Ezra Bialik Assistant Attorney General Environmental Protection Bureau New York State Department of Law 2 World Tr ade Center New York, New York 10047 Jeffrey Cohen New York State Energy Office Swan Street Building Core 1, Second Floor Empire State Plaza Albany, New York, 12223 Director, Bureau of Nuclear Operations State of New York Energy Office Agency Building 2

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Empire State Plaza

Albany, New York 12223 Rochester Public Library 115 South Avenue Rochester, New York 14604 Supervisor of the Town of Ontario 107 Ridge Road West

Ontario, New York 14519 Resident Inspector R. E. Ginna Plant c/o U. S.

NRC 1503 Lake Road

Ontario, New York 14519 Nr. Thomas B. Cochran Natural Resources Defense Council, Inc.

1725 I Street, N. M.

Suite 600 Washington, D. C.

20006 U. S. Environmental Protection Agency Region II Office ATTN:

EIS COORDINATOR 26 Federal Plaza New York, New York 10007 Herbert Grossman, Esq.,

Chairman Atomic Safety and Licensing Boar d U. S. Nuclear Regulatory Comnission Mashington, D. C.

20555 Dr. Richard F. Cole Atomic Safety and Licensing Board U. S. Nuclear Regulatory Commission Mashington, D. C.

20555 Dr. Emmeth A. Luebke Atomic Safety and Licensing Board U-S. Nuclear Regulatory Cooxnission

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GINNA SYSTEMATIC EVALUATION PROGRAM TOPIC I.

INTRODUCTION The safety objective of this topic is to ensure that the previously-established low population zone and population center distance specified for the site are compatible with the current population distribution, and are in accordance with the guidelines of 10 CFR Part 100.

II.

REVIEM CRITERIA Sections 100.10 and 100.11 of 10 CFR Part 100, "Reactor Site Criteria" provides the site evaluation factors which should be considered.

when evaluating sites for nuclear power reactors.

These sections include guidelines for determining the exclusion area, low population zone and population center distance.

III.

RELATED SAFETY TOPICS Topic II-l.A, reviews the licensee's control over the exclusion area.

Various other topics will evaluate the capabiity of the plant to meet the dose criteria of 10 CFR Part 100 at the exclusion area boundary and low population zone.

The adequacy of emergency preparedness planning for the area surrounding the plant including. the low population zone is being assessed by the Commission in a separate review effort.

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REVIEM GUIDELINES The review has been conducted in accordance with Standard. Review Plan (SRP) Section 2. 1.3, "Population Oistribution."

Y.

EVALUATION The R.

E. Ginna site is in the township of Ontario, in the northwest corner of Mayne County, New York, on the north shore of Lake Ontario about 20 miles ENE of the center of the City of Rochester and 40 miles WSM of Oswego.

The land surrounding the site is primarily of an ag a i'ature and sparsely, populated.

There are no substantial population centezs, industrial complexes, tzansporta-on aztezials, pa ks, or other recreational facilities within a Mee mile, radius oZ the Ginna site.

The Ci~y of Rochester is laziest population cent within a 50 mile radius of the site (24~,539 people, with 702.,745 'in the met opolitan area ).

The ne zest comuniiy with a population of X,,OOO or more is the Town of Ontario wiZ its center located about 3~~ m'les fzcm the site.

The prelmina

~ es~~ated

~980 census

=or the Town or. Onta=io 's To develop the Mayne County and,'monroe County Radiological Emergency

Response

Plans for the R.

E.

Ginna Nuclear Power Station, a r cent survey of the population within a five-mile radius was completed.

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=rom the l!ayne County Radiological

Response

Plan, reproduced as Figure 1

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of this evaluation, details the population whin 5 miles of Ginna, based on preliminary 1980 population estimate.

RGSE estimates that.

10,864 persons reside within five miles of the plant, a density of 138 persons per square mile averaged over the entire area.

(It should be noted that this figure compares favorably with the 1980 population projection of 10,934 persons shown in Figure 2.4-2 of the Ginna FSAR, which was published in 1968).

Other than the residents of the area, there are no large groups of transients within five miles of the site.

The only parks near the site are Mebster Beach Park in Monroe County, approximately 6 miles west of the plant site, and B. Forman Park in Mayne County, approximately 8 miles'ast of the plant site.

There are no federal recreational facilities in the area.

There are no state parks, public campsites, or special use areas within ten miles of the plant.

Mayne County does have a migrant labor population, primarily for apple picking, during the June-October season.

Approximately 115 farmworker camps of five or more persons are scattered throughout Mayne County

, with a total population of about 4400 migrants.

Information from Rural New York Farmworker Opportunities shows that there are only 12 camps, with 10 about 130 migrants, located in the vicinity of the Ginna site.

The nearest population center to the Ginna site containing more than 25,000 residents is the "Rochester urbanized area,"

whose eastern boundary is about ten miles from the site.

The only other.population center of more than 25,000 persons is the City of Auburn (population 32,442),

located more than 40 miles SE of the site.

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The low population zone specified for the Ginna site is the area within a 3 mile (4,827 meter) radius of the plant.

A review of current population estimates and projected growth estimates indicate that the population growth in the area since the plant received an operating license in 1969 has been

modest, and this trend is expected to continue.

No population center of 25,000 residents has developed, or appears likely to develop, closer than the eastern boundary of the Rochester urbanized area.

VI.

CONCLUSION The staff concludes that the low population zone and, population center distances specified for the Ginna site is in conformance with the require-ments of 10 CFR Part 100 in that the population center distance is more than one and one-third times the distance from the reactor to the outer boundary of the low population zone (10 miles vs.

3 miles).

Me further conclude that the site conforms to the current licensing criteria.

This completes the evaluation of SEP Topic II-1.8 for the Ginna

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YII.

REFERENCES I

Rochester Gas and Electric Corporation, Robert ~~ett Girja Nuclear Power Plant Unit No.'

- Final Facili r Descript'on and Sa ety Analysis Report (PS'),

Sections 2.2 and 2.4.

2.

Rochester Gas and H.ect~ c Corporat" on, R. =. Ginna Nucleaz Pcwez Plant Unit No. 1, Environmental Repo&,

Volume 1, Sections 2.1 and 2.2.'

Nuclear Regulatory Commission HUREG-75/087, Standard Revie~

Plan, Section 2.I...3',

September 1975.

4.

Code of FedezaI. Regulations,.

Section 10, Pazt 100 (LQ KR IOO).

5.

Payne County Radiological emergency

Response

Plan, Dra t Rev.

3, November 1980.

6.

Monroe County RadioloqicaZ. Rnergency

Response

Plan Draft, Rev.

B, November 1980.

7.

Pzel~ary Report, 1980 Census of Population and Housing, New'ork, published by the Bu eau of the Census, U. S. Depazt-t of Commerce,

=ehzua~

1981.

4 8.

Conversation via Ze New Yor'e State =eall DepaMen~,

Ap il 1981.

9 Safety Evaluation hy the Divis'on of Reactor Z,icensing, U.

S.

Atomic Energy Comm'ss on in Ze Matter of Rochester Gas and Electric Corporation Robert 2nmett Ginna Nuclear Power Pla-t Un t No. 1, Docket No. SO-244 (S~), Section 2.1, June 19, 1969.

1 Q Z,et er, Thomas J. Barris, RNZ=-O, to George Wrohel,

RGH, April LO~

198 Rochester Gas and =-1ectric Corporation, Ginna Nuclear Station RacLiat~ on Emergency

P1an, Proposed Janu~

1981.

New York Stat Radiological. Zmergenc.r Preparedress

P1an, December 1980.

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0 xx IJAO I'OPULATIONESTIMATES FIGORE j.

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r July 21, 1981 Docket No. 50-244 LS05-81<7-070 Hr. John E. Mafer, Yfce President E1ectric and Steam Production Rochester Gas h Electric Corporation'9 East Avenue Rochester, New York 14649

Dear Hr. Mafer:

SUBJECT:

SEP TOPICS II-1 B, POPULATION DISTRIBUTION AND III',SITE PROXIMITY MISSILES - R E

GINNA

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Enclosed are the staff's final evaluations of SEP Topfcs II-l.B and III-4.D for the R. E. Gfnna Nuclear Power Plant These evaluations are based on our review of your topic safety assessment reports sub-mitted by letters dated April 15, 1981 and April 16, 1981, respectively.

You wfll note that we have revfsed your calculated populatf on density

, which fs more properly obtained by dividing the total population within a given distance by the total area of the complete circle (including both level and water) whose radius fs the distance of fnterest-This coapletes our evaluation of Topfcs II-l.B and III-4.D.

These evaluations will be a basic input to the integrated safety assess-ment for your facility unless you identify changes needed to reflect the as-built conditions at your facility.

These assessments may be revised fn the future ff your facility design is changed or if 4RC criteria relating to this subject are modfffed before the integrated assessment is c'ompleted.

Sfnc rely, 8107240143 810721 PDR ADGCK 08000244 P

PDR

Enclosure:

Dennis M. Crutchffeld, Chief Operating Reactors Branch ho.

5 Division of Licensf

R. E.

GINNA YET itC~

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INTRODUCTION The safety objective of this topic is to ensure that the integrity of the safety-related structures,'ystems and components would not be jeopardized due to the potential for a site proximity missile.

!I.

REYIEM CRITERIA General Design Criterion 4, "Environmental and Missile Design Basis."

of Appendix A, "General Design Criteria for Nuclear Power Plants,"

to 10 CFR Part 50, "Licensing of Production and Utilization Facilities,"

requires that nuclear power plant structures, systems and components important to safety be appropriately protected against events and conditions that may occur outside the nuclear power plant.

III.

RELATED SAFETY TOPICS Topic Il-l.C, "Potential Hazards or Changes in Potential Hazards Due to Transportation, Institutional, Industrial and Military Facilities" provides a description of the potential missile hazards.

IY.

REY IEM GUIDELINES The review was conducted in accordance with the guidance given in Standard Review Plan (SRP) Section 2.2.3, "Evaluation of Potential Accidents," 3.5.1.5, "Site Proximity Missiles (except Aircraft),"

and 3.5.1.6, "Aircraft Hazards."

Y.

EYALUATION The potential for hazazdous activities in the vicinity of the Ginna plant has been addressed in SEP topic ZZ-1.C, "Potential Eazazds due to Zndustzial, Transportation, Znstitutional and Kilita~ Facilities".

As indicated therein, the e is Li~tle industrial activity near the plant.

The distances to the nearest land transpo~mtion routes aze such (about 1700 feet to the nearest

highway, and 3 1/2 miles to the nearest railroad) that the risk associated with potentiaL missiles from transportation accidents on these routes are within the SRP 2.2.3 guidelines-.

S~larly, the nearest large gas pipelines are about sU miles

<<om the plant, and do not pose a missile t>>eat to the plant Major Lake Ontario shipping out s a e also su<ficien 'y far away (about 23 miles) so as not to presen a c edible missile haza d

f=om lake t a f'c.

The e are no milit~g acil'ties or activities ne z the plant w? 'ch

-ould c ea e a missi' hazard.

review of S~ Topic ZZ-1.C also evaluated the potentiaL for ai<<c=af becoming a m'ssile haza d, both in connec

~on w'th the ope ation of the Williamson clying Club A'zpor

, wh'ch 's abou ten miles =SZ of the plant, and due to commezcial aiz traffic in a d out of Rochester v'a =ede al airways V?'f and V2, which a e

2 L/2 and 10 miles f=om t"e plant, si e.

As evaluated in Topic ZI-1.C, it was determined that, since the Williamson Flying Club Airport expected.

a meum of only 5000 ope ations per year, and is about 10 miles from the site, the crit ia in IZZ.3.a and IZZ.3.b of SBP 3.5.1.6 were met, and there 's no eed to detezm-'ne the probability of an a'zcra t crash into the plant.

Fu~~~er, the hazard to the plant =om commercial

-8 a'zcraft use of ai ways V2 and. V?9 was shown to be only 5.1 x 10 and 1 4 x 10 pe year, respectively.

No danger to the plant from

-8 comme cial ai line traffic is thus expected.

Conclusion Since current regulatory cr'teria are met'ith respect to SV, Topic III-4.D,. "Si e Proximity Missiles", it can be concluded that Ms topic is complete for the R.

L'. G~a site.

'.fo additional zev'ew foz's topic is zecu"'d during the SV integrated asses sment.

VI.

REFERENCES L.

2.

3.

Rochester Gas and Flectric Corporation, Robert Ramett Ginna Nuclear Power Plant Unit No.

1 - Final Facility Desc iption and Safety Analysis Report (FSAR), Sections 2.2 and 2.5.

Roches Gas and Electric Corporation, R. =. Ginna Nuclea Powe Plant Unit No. L, Knvi onmental Repo~,

Volume 1, Sections 2.L and 2.2.

Nuclear Regulatory Commission NUR=-G-75/087, Standa=d Review PLan, Sections 2.2.1, 2.2.2, 2.2.3, and 3.5.1.&, September 1975

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Code of FederaL RecpQ.ations, Section 10, Part 100 (10 C:-R LOO).

St ling Power Project Nuclear Unit No. 1, P elimiz}ary Safety Analysis Report Addendum, Roches r Gas and =-lectr'c, Volume 1, Sections-2.1 and 2.2.

U.S. Nuclear Regulatory Commission Re@Ra ory Guide L.91, Rev.

1, February 1978.

Letter, John>>.

Maie>>,

RG"=, to Dennis M. Crutchf e'd,

NRC, SW Topic ZI-L.C, "Potential

=-a ards Due o

ransporta cn, indus -ial, Xns

~ tutional and Milita~ Fac'1'"'es",

ApriL 15, L981.

..~gS RECy, fp UNITED STATES NUCLEAR REGULATORY COMMISSlON WASHINGTON, O C. 20555 February 22, 1982 Docket No; 50-244 LS05 02-091 Mr. John E. Maier Vice,President Electric and Steam Pr.oduction Rochester Gas 8 Electric Corp.

89 East Avenue

. Rochester, New York 14649

Dear Mr. Maier:

SUBJECT:

GINNA - SEP TOPIC III-5.A, EFFECTS OF. PIPE BREAK ON SYSTEMS STRUCTURES AND COMPONENTS INSIDE CONTAINMENT 4

By 'letter dated June 30, 1981, the staff issued a draft. safety evaluation on SEP Topic III-S.A which identified ten open items fm. further consid-eration.

Your letter of October 1, 1981, provided responses to the above items.

Based on our review of these letters, we conclude that although several of the items have been resolved, additional information is needed to close out the remaining open items.

Encl.osure 1 discusses each of the open items and their status.

Enclosure

.2 summarizes the information that you are requested to provide.

Enclosure 3 is the revised safety evaluation report for Topic III-5.A, including staff guidelines for resolution'of high energy pipe break locations where remedial modifications are impractical.

This safety evaluation will be a

basic input to the integrated plant safety assessment for your facility.

Resolution of the open items will be addressed in the integrated assessment.

You are requested 'to provide your schedule for completion,of;the items iderrtified in Enclosure 2 within 30 days of receipt of this letter, The reporting and/or recordkeeping requirements contained in this letter affect fewer than ten respondents; therefore, 0MB clearance is not required under P.L.96-511.

Sincerely,

Enclosures:

As stated cc w/enclosures:

See next page Dennis i1. CrutcnTie d, C i.ef Operating Reactors Branch No.

5 Division of Licensing k

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Ginna Docket No: 50-244 Rev. 2/8/82 r'r.

John E. Maier CC Harry H. Voigt, Esquire

LeBoeuf, Lamb, Leiby and MacRae 1333 Hew Hampshire

Avenue, N.

W.

Suite 1100.

Washington, D. C.

20036 Mr. Michael Slade 12 Trailwood Circle Rochester, Hew York 14618 Ezra Bialik A'ssistant Attorney General Environmental Protection Bureau New York State Department of Law World Trade Center

'ew York, New York 10047 Resident Inspector

- R. E. Ginna Plant c/o U. S.

HRC 1503 Lake, Road

Ontario, New York 14519 U. S. Environmental. Protection Agency Regibn II Office ATTN:

Regional Radiation Representative 26 Federal Plaza New York, Hew York 10007

.Herbert Grossman, Esq.,'Chairman Atomic Safety and Licensing Board U. S. Nuclear Regulatory Commission Washington, D. C.

20555 Ronald C. 8aynes, Regional Administrator Nuclear Regulatory. Commission, Region I Office of Inspection and Enforcement 631 Park Avenue King of Prussia, Pennsylvania'9406 Director, Bureau.of Nuclear Operations State of Hew York Energy Office Agency Building 2

'Empire State Plaza

Albany, Hew York 12223 Rochester Public Library 115 South Avenue Rochester, Hew York 14604 Supervisor of the Town of Ontario 107 Ridge Road West

Ontario, Hew York 14519

'Or.

Emmeth A. 'Luebke Atomic Safety and Licensing Board U. S. Nuclear Regulatory. Commission Washington, D. C.. 20555 Dr. Richard F. Cole Atomic Safety and Licensing Board U. S. Nuclear Regulatory Commission

. Washington, D. C.

20555

Enclosure 1'TATUS OF OPEN ITEMS FROM DRAFT SER

References:

(1)

Letter, D. Crutchfield (NRC) to J. Maier (RGKE), dated June 30, 1981 (2) 'Letter,'. Maier (RGKE) to D. Crutchfield (NRC), dated October 1, 1981.

The open items identified by the staff and the licensee responses are provided in R'eferences 1

and 2 respectively.

The present status of each item, keyed to the. numbering in the references, is given below.

The first open item was concerned with the general, assumptions made in thi5 an'alysis.

One of the basic assumptions of this topic assessment was that a check valve in an incoming line would prevent primary system blowdown in the event of a pi pe break upstream of the valve.

This is true provided the check valve closes.

Adequate assurance must be demonstrated that the'se normally open check valves will fulfilltheir assumed. isolation function 2.

On a mechanistic basis, the postulated break locations in the main steam line would not impact the containment wall.

For the feedwater.'lines, the.licensee provi ded an analysis of the structural integrity of the contai nment.

As a bounding analysis, the steam 'line break thrust force was used.

The results show that the containment remains intact, even neglecting the containment 'liner plate.

This issue is considered to be.

'resol ved.

3.

4, The licensee has provided the piping stress results for the "8" steam line.

None of the locations exceeded the stress criteria of 0.8 (1.2Sh+SA).

Accordingly, breaks were postulated at the terminal ends and at the. two highest-stressed intermediate locations.

None of these breaks would cause the crane.to fall.

Therefore, this item is resolved.

For the "A" accumulator line a mechanistic evaluation was performed.,

The stresses in this line were all below the criteria, so'reaks were postulated at terminal ends and at the two intermediate locations. of highest stress..

One of these points was inside the loop. compartment where no adverse interactions would occur.

The second point is located just on the reactor side of the (normally locked open) motor-operated valve.

At this loca'tion no adverse pipe whip inter-actions will occur.

Adequate protection from jet impingement effects must be provided. If remedial measures to provide this protection can be shown to be impractical, fracture mechanics evaluations can be performed

~to establish that conditions that could lead tb a double-ended rupture do not exist as discussed in the guidance provided in the Attachment to Enclosure 3.

The effect of a break in the two inch accumulator.level taps on nearby instrument circuits is still under review by the licensee.

I 5.

For the pressurizer surge line, since some jets could affect safety-related equipment, analyses similar to those described in item 4 above should be provided.

6.

A mechanistic evaluation of the 'pressurizer spray line was performed.

Since the calculated stresses did not exceed -the criteria, breaks were

. postulated at the terminal ends and at the two highest-stressed locations.

None of these break locations would pr'event operation of the sump valves and therefore, this item is resolved.

7.

For the letdown line, licensee evaluation of the effects on cables and cable trays is continuing.

Adequate protection for instrumentation'hould be provided, The situation for the steam generator blowdown lines is similar to item 7 for the instrumentation; With respect to the fan coolers, this size break is not limiting with respect to containment pressure/temperature reduction capability.

The containment spray system would be available for contain-ment cooling.

As for item 7 above, final resolution will occur. after the effects on the cable trays are evaluated:

9.

The licensee has provided the requested references to the subcompartment analyses performed for Reactor'oolant System (RCS) guillotine breaks..

In addition,. the analysis discussed above in item 2 showed that a

30 inch steam line would ret penetrate a

30 inch concrete wall. 'ince the pipes under consideration in the compartment are 10 inch diameter lines, we consider that this concern is resolved, 10.

Pipe breaks were not postulated in the primary loop on the basis of,.the work done'under TAP A-2.

We concur with this approach.

However, the SEP branch intends to evaluate the effects on safety-related equipment

'of the jet loads resulting from the crack sizes associated with these

.analyses

Enclosure 2

t

, RE UEST FOR ADDITIONAL INFORNATION 1., Please provide your basis for assurance that check valves relied upon to prevent primary. system blowdown will fulfilltheir function.

1

. 2.

The following breaks are still under evaluation by you:

(a) accumulator level taps (b) letdown (c) steam generator blowdown Please provide your proposed schedule for these further"evaluations.

Adequate protection for, instrumentation ci rcuits should be provided.

3.

For the 10 inch accumulator'ine and the 10 i nch pressurizer surge line, provi de your planned resolution for possible jet impingement inter actions.

If remedial modifications are impractical, the guidance in the attachment to Enclosure 3 may be used to provide reasonable assurance that mitigation of pipe break effects for these lines is unnecessary.

ENCLOSURE 3

SEP TOPIC III-5.A EFFECTS OF PIPE BREAK ON STRUCTURES, SYSTEMS AND AND COMPONENTS INSIDE CONTAINMENT R.

E.

GINNA (FEBRUARY 1982)

z I.

IHTRODUCTIOH.

TABLE OF CONTENTS II.

REVIEW CRITERIA III.

RELATED SAFETY TOPICS AND INTERFACES IV.

REVI EM GUIDELINES V'ISCUSSION

. A.

BACKGROUND B.

ANALYSIS ASSUMPTIOHS C-SAFETY RELATED EgUIPNENT

, VI.

EVALUATION A.

ASSUMPTIONS AND CRITERIA B.

INTERACTION STUDIES VII.

CONCLUSIONS VIII. REFcRcNCES

II

p INTRODUCTION.

The safety objective of. Systematic Evaluation Program (SEP) Topic.

III-5.A,."Effects of Pipe Break on Structures, Systems and Components, Inside'ontainm nt," is to assure that pipe breaks would not cause the loss of needed function of "safety-related"

systems, structures and components and to assure that the'lant can be safely shutdown in the event of such breaks.

The needed functions of "safety-related" systems

're those functions'equired to mitigate the effects of the pipe break and safely shutdown the reactor plant.

II.

REYIEM CRITERIA The current criteria for review of pipe breaks inside containnent are contained in Standard Review Plan 3.6.2,,

"Determination of Break Locations and Dynamic Effects Associated with the Postulated Rupture of Piping,"

'ncluding its attached Branch Technica1

Position, Mechanical Engineering Branch 3-1

{BTP HEB 3-1).

III.

Background

In a letter dated February 9, 1979 (Reference 2),

RGEE submitted a list of high energy lines i nside contai nm nt.

Representatives of the HRC and RGEE staff met at the Ginna site on Narch 13 and

~ 14, 1979, to discuss the analyses done by the licensee on this topic.

As a result of this meeting, the licensee submitted on September 12, 1979 a report (Reference 3)'n the effects of breaks in these lines on safety-related equipmont.

This review. utilized the effects-oriented approach for the Q gh energy line breaks

analyzed, In this approach, breaks were postulated at any location along the line, and were chosen to produce the greatest jet impingemont or pipe whip loadi ngs on essential equipment.

Also, the assumed plane of motion was that which produced the most adverse effects unless otherwise justified; y(

B.

Analysis Assumptions The'ollowing assumptions were made by the licensee:

1.

High energy fluid systems are systems with operating teoqerature greater. than 200'F or operating pressure greater than 275 psig.

In accordance with Branch Technical Positi'on (BTP)

YiEB 3-1, breaks are not postulated in pipino of systems that qualify as high energy systems for only short operational 'periods (i.e.,

less than 2>> of the time the sys em operates as a.moderate energy system).

Pipes less than one inch (1") in diameter were also elimi'nated in accordance'ith Regulatory Guide 1.46.

2.

Pipe. of a given section modulus will not cause.a loss of func-tion in pipe of equal or larger section modulus as a result of

.,pipe whip or jet impingement.

3.

Pipe whip can only occur in the sec.ion of pipe which is attached to a sustained high energy source.

Credit is.taken for all closed.

or automatically. closed valves (e.g.,

check valves) in the'piping section, that could terminate flow.

4.

The jet impingement force (calculated to be less than 200 pounds) due to breaks in the 2" diameter lines fed by the positive displacement charging pumps will not impair function-ing of equipment.

5.

In addition to the equipment affected by the br'eak, a single independent failure of an active component inside containment is cons idered.

C-.

Safety-Related Equipment Safety-related equipment includes systems needed to mitigate the effects of the line breaks and to bring the reactor to safe shut-down.

Breaks inside cqntainm nt generally result in or have the same effect as loss of coolant accidents or steam/feed line breaks.

Engineered safety features are required to mitigate these breaks.

Other breaks (such's accumulator line breaks) do not result in a loss of inventory or energy from the reactor coolant system and thus require only normal safe shutdown systems such as CVCS.

Systems that are all or partially inside containment a'e:

. 'afety Injection (SI) - two trains one to each cold leg, n'o active components inside containment Low Pressure Safety Injection (LPSI) - two trains which pump water to the injection nozzles on the vessel through motor-operated valves (NOYS) 852A and B, which nust change position on receipt of a safety injection signal Accumulators - di.rected to each cold leg, no active components Containment Spray - two trains to spray headers in containm nt, no.active components. inside containm nt Containment Fan Coolers and Service Rater - four fan coolers which must operate to provide cooling; service water has no active components inside containment Sump Recirculation - two lines from the sump to Emergency Core Cooling System (ECCS)

pumps, no active components inside containment Residual Heat Reroval - one drop line, one return line, each with two tlOVs

r Chemical and Volume Control lCYCS) Charging and Letdown - two physically separated charging paths Standby Auxiliary Feedwater System - ties into main feedwater

lines, no active components inside containment;. auxiliary feedwater

~ system is totally outside containment Essential Instrumentation - pressurizer

pressure, steam generator level.

~

YI.

EVALUATION A.

Assumptions and Criteria As discussed earlier, lines separated from an energy reservoir by a check valve were.not assumed to have~sufficient energy to whip

'or produce jets-For long runs of large piping, the'enero~

stored within'the pipe volume from the break to the valve could be su'ffi-cient to form a jet.

For Ginna,

however, the only pipes for which the check valve separation is utilized to limit interactions are 2"

pipes, so this effect's not expected to be significant.

~

V However, assurance must be provided that the normally'open check valve closes sufficiently so that the dynamic forces from the..

reactor side are not significant.

The staff concurs that use of the pipe section modulus is an appropriate measure of relative strength of pipes.

The licensee has assumed that a pipe of larger section modulus will break a pipe of smaller section

modulus, but a smaller section 'nodulus was not considered to affect a larger section modulus.

In accordance with staff

'ositions transmitted on January 4, 1980 (Reference 4), the effects'f jet impingement loads should be considered and evaluated regardless of the magnitudes of the section modulus of impinged and postulated broken pipes.

Therefore,.the licensee should perform additional'valuations of the effects of jet inqingem nt on equipment and-piping.

~

An acceptable iet model is described in Standard Review Plan Section 3.6;2.

Si'ngle failures of acti ve components i nside containment, such as the fan coolers or LPSI valves were considered by the licensee, Loss of offsice power was not specifically addressed in this study, but the staff has included c'onsideration of the consequences in its.

revie~.

In the safety injection and accumJlator

systems, a loss of offsite pmer, a single failure and a broken injection line would no prevent injection flow into the other loop.

For the con-:

tainment spray

system, a loss of offsite power, single failure of a diesel and a rupture of one spray line could reduce containment heat removal capability below the minimm assumed in the LOCA analysis.

Thus, a break that could affect a containm nt spray line should be. further considered toisee if the remaining systems are adequate.

B.

Interaction Studies For each of the postulated break locations, the licensee evaluated the effects on the essential equipment.

In addition,'he effects on other impacted equipment were considered to ensure that failure of such equipment would not exacerbate the break effects.

Several of the breaks would be confined within one of the loop coin-partments, would not affect the other train of safety injection, or the low pressure safety injection system, and therefore, would not

.prevent safe shutdown.

Most breaks in loop compartments do not affect safety-related electrical equipment since this equipment is not located inside, the compartments.

A hi gh energy.line is assumed to break an irqacted line of smaller section 'aedulus.

If this impacted line is also a high energy line,

. the potential dynamic effects of that break must be concurrently considered.

The check and isolation valves located close to the reactor.connection on aest of the high energy lines assure that.

even if a line is broken by the initiating pipe break, there is insufficient energy to produce other effects from the'second break.

Such a situation arises with the accumulator line (from tank skirt to loop compartment wall).

Breaks. in this line can a feet the Residual Heat Removal (RHR) outlet line-This line is. itself a high energy line within the loop compartment, but is not a high energy line outside the compartm nt due to the two nornallymlosed isolation valves.*

h'ithin loop compartments or within the pressurizer compartment, the potential exists for'high energy lines to impact other high

. energy.lines, such as the RHR in line impacting a charging line.

However, in general, the minim.m engineered safety features (ESF) needed to mitigate these breaks are physically separated from the break and are thus unaffected.

Breaks in the primary loop of reactor coolant system (RCS) were not addressed by the licensee'n this study on the basis of the work performed for USI A-2 (Asymmetric Blowdown Loads on Reactor Primary Coolant System}.

Based on the interaction studies under the effects-oriented

approach, several locations were identified with potentially unacceptable consequences for which further evaluation was necessary:

The staff issued a draft safety evaluation on this topic on June 30, 1981 (Reference 5).

The potentially unacceptable break locations were identified for further review...

The licensee responded to our draft evaluation on October 1, 1981.

(Reference 6).

For some high energy lines that could not be shown to be acceptable on an effects-ori'ented

basis, a mechanistic evaluation was performed (see Reference 1).

In this approach,

'stress analyses were performed to locate the most"highly stressed points, which are the locations most likely to fail.

Breaks must be postulated at all intermediate locations where.the stress for the limiting normal and upset conditions exceeds 0:8 (1.2 Sh+SA) and at terminal ends.'f all stresses are 6elow this criteria, at least two intermediate points, the highest two stresses, must be

'postulated.

Breaks at the highest stress locations did not result in unaccept-able consequences for the main steam line and the pressurizer spray line.

The effect on.the containment wall of whip of the feedwater line was determined to be acceptable based on a bounding analysis o'f steam line impact.

The results of this analysis showed that the penetration of the concrete is less than 1.4 inches, even neglecting the steel containment liner.

Thus; a feedwater line break will not result in loss of structural integrity of the containment, and break consequences are considered to be acceptable.

For three lines, the licensee is continuing his review of jet effects on instrumentation and cable trays:

(a) letdown line (b) steam generator blowdown line (c) accumulator level taps..

Although no adverse pi pe whi p interactions can occur, safety-related equipment must be adequately protected'rom jets resulting from failures of the accumulator line and the pressuri zer surge line.

"p w

7 Pipe ruptu'res in the primary coolant loop were not postulated beaause of the A-2 "leak-before-break" technique, however, the staff will assess the effects of jets f'rom crack sizes determined by that work.

VII.

CONCLUSIONS The staff has reviewed the layout'rawings, analyses and other informa-tion provided by the licensee.

In addition, the staff toured repre-sentative locations on June 1-2, 1981 in the Ginna containment to observe the pipe configurations and proximity to safety related equipment.

Based on these

reviews, we conclude that. the licensee has satisfactorily addressed the pipe whip and jet effects of high energy line breaks inside containment and has demonstrated an adequate level of protection subject to resolution of the following:

l.

A basis must be provided for assuming that the normally open check valves relied upon for'revention of reactor blowdown will close.

2.

The evaluation of effects on instrumentation circuits from breaks in letdown pi ping, steam generator blowdown piping and the accumu-lator level taps is still ongoing.

An adequate level of protection for'he instrumentation must be demonstrated.

3.

For the accumulator line and the pressurizer surge line, adequate protection of safety-related targets must be provided. If remedial

'odifications are shown to be impractical, fracture mechanics evaluations may be performed.

Guidelines for this analysis are provided. in the attachment.

4.

As discussed

above, the staff wi 11 evaluate jets from cracks in the primary coolant loop, VIII. REFERENCES 1.

Letter from 7,

1978.

D. Eisenhut (NRC) to L. D. White (RG&E), dated September 2.

Letter from L.

D. White (RG&E) to D. Ziemann (NRC,. dated February',

1979.

3.

L'etter from

'2, 1979.

4.

Letter from 4,

1980.

L. D. White (RG&E) to D. Ziemann (NRC), dated September D. Ziemann (NRC) to L. D. White (RG&E), dated January 5.

Letter from D. Crutchfield (NRC) to J. Maier (RG&E), dated June 30, 1981.

6.

Letter from J.'aier (RG&E) to D. Crutchfield (NRC), dated October 1,

1981,

p GUIDANCE FOR RESOLUTION OF HIGH A

y r P

Attachment'to Enclosure 3

From the results of reviews conducted to date, the staff has concluded that the relocation of equipment or other modifications to mitigate the consequences of some postulated pipe breaks may be impractical due to physical plant configura-tions. or other considerations.

Therefore, the staff has determined that for specific locations where re'location of equipment or other modifications to mitigate consequences of pipe breaks are shown to be impractical, fracture mechanics evaluation of the piping should be performed to determine if unstable ruptures could occur in piping that contained service induced large undetected flaws.

y(

The. intent of the guidance provided by'the staff is to provide reasonable assurance that the mitigation of pipe breaks are addressed.

"The approach taken is to provide assessment. that condition which could lead to a double ended pipe rupture do not exist thereby making it unecessary for high energy pipe break considerations to mitigate effects of a guillotine rupture.

This.

would be accomplished using a defense in depth approach that is a combinatign of augmented inservice inspection

( ISI), local leak'etection and fracture mech-

'nics evaluations.

Augm nted inservice inspections would be perfor'med with the goal of detecting and limiting any service induced flaws to limits prescribed by the ASME BKPY Code,Section XI, approximately 10~ thru wall.

Should t'e flaws go undetected, a local leak detection system would be provi ded with the requisite sensitivity 'to identify leakage from a through crack, either longitudinal or circumferential, of a length of twice the wall thickness for minimm flow rates associated with normal (Level A) operating conditions.

Fracture mechanics evaluations would be performed to determine that for a circumferential or longitudinal through crack of four wall thickness subjected to maximum ASME design code loads (Level D) that:

(l) substantial crack growth'oes not occur.

P (2) local or. general plastic collapse {instability) does.not occur.

~

(3)

.flow through the crack or the effects of a jet from the crack does riot impair safe system shutdown.

To provide assurance that a double ended rupture could not occur by. unantici-..

pated loads being applied to a large undetected

crack, a fracture mechanics evaluation would be performed to demonstrate that a through crack of a length of four times the wall thickness, 90 total circumferential length, or.a larger crack if justified for system service experience would remain stable for local

fully plastic large deformation bending conditions.

The basis for performance of this more conservative fracture mhchanics evaluation to assure a double ended pipe rupture would not occur is as follows:,

(1) operating experience has shown that unanticipated and undefined loads in access of design can and do occur in piping systems, i,e., water hamner events have failed piping system supports.

'(2). uncertainty in (a) current analysis methods to accurately predict piping loads analysis and (b) prediction of the energy and frequency content of earthquakes and their effect on piping loads.

(3)

SEP criteria for evaluation of structures and system resistance to

'ostulated earthquake loads depend on global structural ductility.

This, assumption is based on the ability to have load redistributions occur..

For unflawed. piping, the necessary local ductility is cer-tainly provided.

However, for'lawed sections of piping the ability to sustain fully plastic behavior without crack instability is required to assure prudently that local ductility is preserved.

The details of the guidance for the combined augmented ISI, leak detectionand fracture mechanics evaluations are attached as Enclosure l.

Attachment 1

ALTERHATIYE SAFETY ASSESSMEHT FOR SELECTED A

A L

This assessment is required only if a L'MR high energy piping system '(i.e.,

275 psi, or higher; or 200 F or higher, etc.) is being considered.

It is only requi.red, if a postulated double ended pipe break would impair safe system shutdown by pipe whip (lacking pipe whip constraints) consequences, or by the consequences of the implied leakage or its jet action.

The following guidance is for a safety assessment that may be permitted as an alternative to other system modifications or alterations for locations where the mitiga-tion of the consequences of high energy pipe break (or leakage) have been shown to be impra'ctical.

C Guidance for.'lternate Safety A'ssessment The suggested guidance. are as follows:

A; Detectability Requirements Provide 'a leak detection system to detect through-.cracks of a length of twice the wall thickness for minimum flow rates associated with normal (level A) ASNE BKPV Code operating conditions.

Both circumferential and longitudinal cracks must be considered for all critical break or leak locations.

Methods for estimation of crack opening areas are attached.

Surface roughness of the crack should be considered.

8; Integrity Requirements (1)

'Loads for Which Level D is Specified (a)

Show that circumferential or longitudinal through-cracks of four wall thicknesses in length subjected to maximum level D loading conditions do not exhibit substantial. monotonic load-ing crack growth (e.'g

, staying below JI or K'y plastic zone corrected lineay~elastic fracture mfchani/( methods or a

suitable alternative Also assure that local or general plastic instability does not occur for these loading conditions and crack sizes.

. T7 For 4t flaws that are calculated to be greater than K Cor JI, con-sideration wiil be given to; (1) fl'aw growth argumenti, (2) Iiostulation

.of small flaws sizes than 4t if ju'stifled by leak detection sensitivity.

(b)

Under conditions in "B.(1)" show that the flow through the crack and the action of the jet through the crack will not impair safe shutdown of the system.

l Acceptable methodology for the estimation of crack opening area for a circumferential through crack in a pipe in tension and bending and for longitudinal cracks subject to internal pressure are attached.

(2)

Extreme Conditions to Preclude a Double-Ended Pipe Break Using elastic-plastic fracture-mechanics or suitable alternative show.

that circumferential through-cracks will remain stable for local fully plastic large-deformation bending conditions under the following addi-tional conditions:

(.a)

Fully plastic bending of the cracked section is to be assumed, unless other load limiting local conditions (such as 'elbow collapse) dictate maximum bending loads, for all critical locations.

(b)

Assume that all system anchors are effective, but that other supports (such as hangers and snubbers) are inoperative unless e'specially justified..

(c)

Other as built displacement limits or constraints may be assumed as especially justified (such as displacement limits of a pipe running through a hole in a sufficiently strong concrete wall or floor, etc. ).

0 (d)

Assume a through-crack size of 4t or 90 total circumferential length whichever is greater; or a larger crack only if especially justified.

~ 'e)

Assume large deformations means deformations proceeding to as built displacement limits or other especially justified limits.

~(3) t'iateri.al Properties Conservative material properties should be used in the analyses.

Sufficient justification est be provided for the properties, both..

weldment and base metal, used in the analyses.

(

C.

Subcritical Crack Development Consideration should be given to the types of subcritical cracks which may be developed at all locations associatediwith this type of analysis.

From prior experience and/or direct analysis it should be. shown that:

(1) there is a positive tendency to develop through-wall cracks.

(2) if there is a tendency to develop long surface cracks in addition to through-wal-1 cracks, then it should be further demonstrated that the long surface crack will remain sufficiently shallow.

D.

Augmented Inservice Inspection Piping system locations for which corrective measures are not practicable should be inspected volumetrically 'in accordance with ASME Code,Section XI for.a Class 1 system regardless of actual system classification.

Acknowl edgement Assistance in developing this guidance have been provided by Dr. Paul C. Paris, Del Research Corporation (and Mashington University, St. Louis, NO) under sub-contract K-8195 in support of technical assistance provided by Idaho National Engineering Laboratory, Idaho Falls, Idaho (FIN A-6456).

ESTIMATION OF STRESS INTENSITY FACTORS AND THE CRACK OPENING AREA OF A CIRCUMFERENTIAL AND A LONGITUDINAL THROUGH-CRACK IN A PIPE by H. Tada 'and P. Paris Del Research Corporation St. Louis, Missouri Introduction Formulas for estimating the crack opening area are developed for a circumferential and a longitudinal through-crack in a pipe subjected to several types of'oading.

For the circumferential crack, estimation for-mulas are presented for axial force and bo nding moment applied to the pipe far from the cracked section and for internal pressure loading.

For the longitudinal crack, an estimation formula for the case of internal pres-sure is presented.

Estimation is based on the method of linear elastic fracture mechanics, which requires the knowledge of the solution of stress intensity factor, K, for each problem.

For the internal pressure loading, K-solutions are readily available for both circumferential and longitudinal cracks as func-tons of a single geometric parameter,

~(= a/i Rt), relating crack size and pipe geometry.

Consequently, the crack opening area formulas are also formulated as functions of this single parameter.

For the case of tension and bendi ng of circumferential

crack, however, the stress intensity factors are not formulated as functions of a single parameter and no simple formula is readily available.

Therefore, in this discussion, a typical valu'e of

P mean radius to thickness ratio, R/t = 10, is specifically selected and for-mulation is made for this. value..

Estimation formulas a'eexpected to yield a slight overestimate. for R/t = 10.

For smaller R/t ratios, degree of overestimate would increase, The: formulas presented here may be used with a l.easonable accuracy when R/t ratio is about 3,0.

Formulas for the crack opening for these cases are not available in simple closed forms, but here moderately long po~er series approximations based directly. on the estimating

.formulas for K

are given..

e A Circumferential Throu h-Crack in Tension and Bendin

(

The K

formulas are first developed here based on'the resul'ts rect.ntly obtained by Sanders.

[1, 2j.

As stated

above, the K

solutio~s for the'se loadings are not expressed as functions of a single geometric parameter.'anders presented approximate formulas for the energy release rate for these

loadings, which are readily converted into K

formulas.

The formulas are," in essence, functions of two geometric parameters for given elastic constan.s, which may be written in either of the following forms.

or

'K,= ei~mRB F(e,-)R" where a

is an applied, stress, 2Re is the total circumferential length of through-crack.

\\

oe C

In this discussion, 8

and R/t are. chosen as geometric parameters and the second form of Eq. (1) is employed for the stress intensity expression.

Approximate K

formulas and the subsequent dstimation formulas for the.

crack opening areas are developed specifically for R/t = 10, which is con-

~

'I

~

sidered to be a typical value, of interest in the present study.

That is, the function F(a) in ths subsequent discussion represents F(e.;10).

Let

'P and M

be the axial tensile force and bending

moment, respec-

.tively, applied to the pipe far from the crack location and let subscripts t

and b

represent respectively tension and bending.

The nominal stresses I

due to tension and bending are defined by P

at 2mRt

(.2)

The stress intensity factors are expressed in the following. forms.

K< = ~ '~Re F<(e)

Rb

= nb~as F>(e)

(,

where Ft(e) and Fb{e) are non-dimensional functions.

The numerical values of the functions Ft(e) and Fb(e) are calculated from Sanders'pproximate formulas for R/t = 10, which are tabulated as follows.

h 5~

~ P

~ ~

(Ft(e) and Fb(e) for R/t = 10) 00 9

18 27 36 45 54 63

~

72 81.

90 99 108 F (e) 1.000 1.039 1.151 1.314 1.505 1.725

.1.987 2.305 2.702

3. 209

3. 872 4.764 6.003 Fb(e)

\\

1.000

~ 1.037 1.140

1. 278

1. 425 1.580 1.747 1.934

2. 154

2. 406

2. 760

3. 209

3. 827 These values represent slight overestimates of F (e) and Fb(e) D,2j.

The following approxima'te expressions of the functions Ft(e) and Fb(e) represent the values of the table with a reasonable accuracy (within a few percent)..

~

3/2 S/2 Vt'2 F (a)

=

1 +

7 5(-)

- 15(-)

+ 33(-)

TT IT 3/2 5/2 7/2 Fb(e)

=

1 + 6.8(-)

- 13.6(-)

+ 20(-)

(4)

(0 8

100')

7

~ P

~t

~ C When the pipe is subjected to axial force and bending moment at the same time, the total stress intensity factor is obtained simply by super-position of these separate

factors, I

total t

b

~ The crack opening'reas due to. tension and bending, At and A

, may be conveniently expressed in the following form.

~

'A

= (iR ) I (e) t E

~b Ab =

E

(>R ) Ib(6)

(6) where E

is the Young's modulus, and It(e) and Ib(e) are non-dimensional functions.

The crack opening area for the tensile loading, A, is obtianed by eneroy method (Castigliano's theorem) as follows:

2 1 aVt 6

a Kt A

=

= '2f

'()Rde t

t aa<

roast E

(7) since 2

aUt Kt G

Rt a8 E

where V

is'the total strain energy in the cracked pipe.

Combining t

Eqs. (3), (6) and (7), the functions It(e)-is'obtained as follows:

h

V

,J e

e (e(e)

= 4j e(Fe(e)}

de 0

Substituting F (e) given by Eq. (4),

It(8) eis written as Ie(e)

= 2e 1+

()

(8.6 -13.3(-)

+ 24()

I 83<'2 8

82

+

(

) {22,5 - 75() + 205.7(-)

8 3

e 2

(I0) 247.5(-)

+ 242() f 8

3 8

(0 9 < 100')

The crack opening-area for bending load, Ab, however, can not be obtained as readily because the "crack absent stress distribution" is not uniform along the, crack (direct application of the energy method is difficult).

~

Therefore,

'Ab or Ib(e) will be e'stimated in the following way.

First, comparison of the crack absent stress dist'ributions for 'tensile and bending loads, the following bounds are imposed on Ab'.

At( t ='~bcose)

< Ab ~b) < At(~t

~b (cose)It(e)

< Ib(e) < It(e) h'here A (a

) is the crack opening area by bendi ng, and At(at = a cosa) and b

b A '(a

= a

)

are the crack opening area due to axial force with tension stress t

b b

c cose and a

respectively.

The first approximation would be to take the b

b' average uniform stress between these extremes and or b( b) t( b 2')

= t( b( o 2)

)

r Ib(e)

= (cos 2) It(e)

(12)

Since the function Ib(e) given by Eq.

(12) may yield underestimated values of the crack opening by bending, the stress intensity factors K

and Kb t

are compared in a similar manner.

Corresponding to Eq. (11), it is obvious that or Kt t

= abcos8)

~

Kb b

b ~t

(0< X< j)

(2I.)

(1 <

X'< 5).

Y I Applying the energy method again, the crack opening area, A, is readily obtained as follows A

= (2mRt) '

(x) p E

P (22) where G () ) is given by an integral G.(x)'

2( x(F (x))

d).

0 Corresponding to Eq. (21);

G (X) is evaluated as 2

4 G (x)

= x

+ 0.16>

P

= 0.02

+ 0.81K

+ 0.30K

+ 0.03K 3.

4.

(0< X< 1)

(23)

(1< )<<, 5)

The effect of yielding near the crack tip may be simi1arly incorporated using the effective (plastic zone corrected) crack size which,is calculated from the iterative relation (2O)

Lon studinal Throuoh-Crack Sub'ected to Internal ressure For a pipe subjected to internal pressure, p, the hoop stress, i, is estimated by

, ~ >

Rt (25)

The stress intensity factor for a longitudinal through-crackof length 2a is given by K = a.ha.F()

)

(26) where again X = a//Rt The geometric factor F(a) can be empirically expressed over the range of interest by

>/z F(~)

= (1 + 1.25~

)

= 0.6 + 0,9X

.(0< ~<1)

(1<

X< 5) f27)

-Eq.

(27) provides a aood approximation for the shell factor F(X) with accuracy of the order of one percent

[3, 4, 5, 6j.

The crack opening area, A

, can be obtained by the method in the previous di.scussion.

A = (2-Rt) ~ G(X )

E t

where G(A.) corresponding to Eq.

(27) is given by (2S)

G(x)

= z

+ 0.625~

. (0<

X< 1)

(29)

= 0.14 + 0. 36K

+ 0.72K

+ 0.405K (1 <

X < 5)

'teration with a plastic zone correction similar to'q.(24) can be applied to accoont for the yielding effect near the crack tip.

,~ P References I

[1j J.

L. Sanders, Jr., "Circumferential Through-Cracks in Cylindrical Shells Under Tension,

" to be published in Journal of Applied Mechanics.

'[2j J.

L. Sande'rs; Jr

, Under Bending, Private Communication,

November, 1981.
[3j D.
P.
Rooke and D. J. Cartwright, "Compendium of Stress Intensity Factors,"
Her Majesty'p Stationary Office', London, 1976.
[4j
,'F:
S. Folias, "An Axial Crack in a Pressurized Cyli'ndrical Shell,"
Int. J. of Fracture Mechanics, Vol. 1, 1965, pp.
104-113.
[53 F.
Erdogan and J. J. Kibler, "Cylindrical and Spherical Shells with Cracks," Int. J. of Fracture Mechanics, Vol.5, 1969, pp. 229-237.
[6]
S.
Krenk, "Influence of Transverse Shear on an Axial Crack in 'a Cylindrical Shell," Int. J. of Fracture, Vol. 14,
1978, pp.
123-143.
gP,S RE0I 4
P0
'+p*~4
(
Docket No. 50-244 LS05-81 49-018 UNITED STATES NUCLEAR REGULATORY COiMibllSSION WASHINGTON, O. C, 20555 September 4,
1981 John E. Haier Vice President Electric and Steam Production Rochester Gas 5 Electric Corporation 89 East Avenue Rochester, New York 14649
Dear Hr. tlaier:
SUBJECT:
SEP TOPIC III-5.B, P IPE BREAK OUTSIDE COttTAII')tlENT R. E.
G INNA
.The staff evaluation of SEP Topic III-5.8 was transmitted to you on June
-"' " "i d f'v~
s "= positio~s
"" '"'c1 2"
'".Ip I ei".'
T 0! Gn scn2Q" Ie w "s '"eq J"-s 'Q ~
I our r2sponse was pcovi Q"c i fl letter dated August 7, 1980.
Each of the five positmns, your responses and staff resolutions are di scussed bel ow-Staff Position 1
Because high and moderate energy line breaks in the Screen House could damage the power supplies to all service water pumps, the licensee must provide protection for these power supplies in accordance with. Standard Review Plan 3.6.1 consistent with the service water system modifications which must be performed in connection with other ongoing SEP reviews'nd the fire protection review.
Hodifications to provide this protection-can be acceptably delayed until the SEP integrated assessment of the plant provided that the diesel generator cooling method described, in the licensee's December 28, 1979 fire protection safe shutdown analysis, is tested to assure its timely availability and its capability to provide adequate cooling.
The results of this testing should be submitted for NRC staff review.
Res onse to Staff Position 1
It is planned to conduct the alternative diesel generator cooling method test by June 1981.
Resolution The alternate diesel generator cooling method depends on installation of hose connections to each diesel generator.
These connections have not yet been installed.
As discussed in the June l980 SEF, protection should be provided for-Buses 17 and 18 and associated cables.
Such modifications should be coordinated in the integrated assessment with Fire protection and other SEP topic concerns.
Staff Position 2
The licensee must provide the means to warn the control room operator that flooding conditions exist in the Intermediate Building sub-basement.
The licensee should provide the implementation schedule for this capa-bi 1 ity.
Response
to Staff Position 2
Based on RGKE's review of tliis scenario, we find the proposed solution to be unnecessary.
Present routine walk-through inspections of the its'
~ I'>~~~ ~el I<i ~ij s ~~
~
c
~
v'~ a or< I': tv i~ ~~i ol s
~ s' A
any danger of flooding safety-related equipment.
If the postulated leak occurred at a level above the sub-basement, leakage into the sub-basement via the floor drains would be obvious during the routine once-per-shift walk-throughs.
And even a large'econdary side break would result in only a 2-foot depth in the sub-basement.
If the leak were in the Service Mater piping located in the sub-basement of the Intermediate Building, there would be a significant time interval between the initiation of the crack-and the floo'ding of safety-related equipment.
The I~termediate Building sub-basement has a volume of approximately 50,000 ft.
With a service water leak rate of about 585 gpm (as calculated on p. 13 of the NRC assessment), it would take over 10 1/2 hours to begin flooding the basement level. It does not seem conceivable that a sizeable leak rate such as this would not be detected, visibly or audibly by personnel during the walk-throughs, or by personnel monitoring the control board (the 585 gpm leak would be a significant fraction -
10% - of the Service Water pump flow).
Resolution The staff has determined from discussions with the licensee during a
site visit on June 2, 1981, that there are two sump pumps in the sub-basement.
Operation of the pumps is alarmed at the water treatment stati'on.
A control room alarm is provided indicating that an alarm condition exists at the water treatment station.
As stated in the topic evaluation, even
~,
I 3w if the basement elevation was flooded, safe shutdown would not be prevented.
Based on this, and the other information provided above, the staff con-cludes that there are adequate means to warn of flooding conditions in the sub-basement and therefore, that no modifications are required.
Staff Position 3
~ ~
Based on our evaluation of Nain Steam (NS) and Nain Feed (NF) line breaks in the Turbine Building and Intermediate Building, the licensee should (1) proceed with the design and installation of jet impingement shielding in the Intermediate Building (as previously committed to by the licensee),
(2) provide protection from the effects of the failure of the Turbine Building/Intermediate Building cir'Cer block wall for the NS atmospher'ic dump valves and assess the need for and provide protection as necessary for the NS safety valves.
The installation of additional jet impingement shielding for the NS bypass valves and associated piping is not necessary since the bypass valves are not required for safe shutdown or pipe break mitigation.
A proposal to accommodate item (2) above should be submitted for staff review.
Response
o S
~, Posi ion 3
Protection from the effects of the Turbine Building/intermediate Building cinder block wall failure on the atmospheric dump valves and main steam safety valves will be integrated into the modification program resulting from RGBE's review of IBE Bulletin 80-11, "Masonry Mall Oesign."
Our initial response to this bulletin is contained in a July 7, 1980 letter from L. O. Mhite, Jr.
(RG8E) to Nr. Boyce H. Grier (NRC Region I Oirector).
Resolution Additional information in response,to IBE Bulletin 80-11 was submitted by the licensee on November 4, 1980 and January 30, 1981.
The SEP review of these
- letters has revealed that pipe break loads were not included in this evaluation of masonry wall design.
Furthermore, since the evaluation against original design criteria showed that the walls would satisfy their intended function, no assessment of effects of cinder block wall failure has been provided. Therefore, the licensee should comply with item 2 above.
Staff Position 4
Since certain moderate energy line breaks (NELB) in the mechanical equipment room could result in flooding both battery
rooms, the licensee must provide protection from the effects of these postulated NELB's in accordance with the acceptance cri teria of Standard Review Plan 3.6.1.
The licensee should provide a schedule for the implementation of this position.
Response
to Staff Position 4
It is presently planned to separate the battery rooms from the mechanical equipment room, where the source of a Service Water leakage exists, by replacing the doorway with a watertight wa11.
This modification should be coapleted by June 1981.
Resolution The modification will be completed shortly.
The licensee also plans to install at the same time a means of removing water from the mechanical equipment room into the turbine building.
The staff concludes that these r odifications will adequately mitigate the effects of these postulated MELB's.
Staff Position 5
To preclude adverse environmental conditions resulting from a heating steam or CVCS letdown break in the Auxiliary Building, the licensee must analyze the adequacy of once-per-shift inspections to prevent the formation of the adverse environment or to provide some other acceptable means of preventing tl e exis';=.nce of the adverse envi~on.-,en..
<<he results of 'h;s analysis (with a co;;~i-.ment to provide the required protec-.ion, if necessary j should be submitted for HRC staff review.
Response
to Staff Position 5
RG&E is performing an evaluation to determine the effects of a CVCS letdown or steam heating line break.in the Auxiliary Building in the vicinity of safety-related equipment.
The results of this study and proposed modifica.-
tions, will be submitted to the NRC for review in January 1981.
Pending the resolution of any noted concerns, present once-per-shift inspections, together with the procedures available f'r isolation of the steam heating line, should provide adequate protection against the effects of significant adverse environment damaging safety-related equipment.
Resolution The environmental effects of these breaks on safety-related equipment are being addressed as part of Unresolved Safety Issues (QSI) "gualifi'cation of Class lE Equipment".
Per the Conmission's Memorandum and Order of May 23, 1980, all safety-related electrical equipment must be. qualified for the ad-verse environments they would experience by June 30,1982.
Therefore, this item will not be further addressed under Topic III-5.8.
The staff now considers this SEP topic to be completed except for comple-tion of the commitments discussed above.'and of modifications necessary to protect equipment in the screen house and Turbine Building/Intermediate Building.
Enclosed is the revised evaluation which will be a basic input to the integrated safety assessment for your facility unless you identify changes needed to reflect the as-built conditions at your facility.
This topic assessment may be revised in the future if your facility design is changed or if NRC criteria relating to this topic are modified before the integrated assessment is completed.
Sincerely,
Enclosure:
As stated
~( - -~ NI (~"~>c4~~~6/
Oenni s tl. Crutchfiel d, Chi ef Operating Reactors Branch Ho.
5 Oivision of Licensing cc w/enclosure:
, See next page
H'r. John E. Miaier CC Harry H. Voigt, Esquire
LeBoeuf, Lamb, Leiby and MacRae 1333 Hew Hampshire
Avenue, N.
W.
Suite 1100 Washington, D. C.
20036 Mr. Michael Slade 12 Trai lwood Circle Rochester, Nevi York 14618 Ezra Bialik Assistant Attorney General Environmental Protection Bureau New'York State Department of Law
'2 World Trade Center New York, New York 10047 J effrey Cohen New York Sta-
~ 'rer gy 0-->~~
Swan Street Building Core 1,
Second Floor Empire State Plaza
Albany, New York 12223 Director, Technical Development Programs State of New York Energy Office Agency Building 2 Empire State Plaza
Albany, New York 12223 Mr. Thomas B. Cochran Natural Resources Defense Council, I'nc.
1T25 I Street, H.
W.
Suite 600 Washington, D. C.
20006 U. S.
Environm ntal Protection Agency Region II Office ATTN:
EIS COORDINATOR 26 Federal Plaza New York, Hew York 10007 Herbert Grossman, Esq.,
Chairman Atomic Safety and Licensing Board U. S. Nuclear Regulatory Coomission Washington, D. C.
20555 Dr. Richard F, Cole Atomic Safe.y nd
~ icensing Bca. C
U. S. Nuclear Regulatory Commission Washington, D. C.
20555 Dr.
Emmeth A. Luebke Atomic Safety, and Licensing Board U. S. Nuclear Regulatory Commission Washington, D. C.
20555 Rochester Public Library 115 South Avenue Rochester, New York. 14604 Supervisor of the Town of Ontario 107 Ridge Road West
Ontario, Hew York 14519 Resident Inspector R. E. Ginna Plant c/o U.
S.
NRC 1503 Lake Road
'ntario, New York 14519
SEP REVIEW PIPE BREAK OUTSIOE CONTAIi'lilEiVT TOPIC
' II-5. B FOR THE R.
6.
GINNA NUCLEAR POWER PLANT
INTRODUCTION The safety objective of Systematic Evaluation Program (SEP) Topic III-5.8, "Pipe Break Outside Containment" is to 'assure that pipe breaks would not cause the loss of needed functions of safety-related
systems, structures and comp" onents and to assure that the plant can be safely shut down in the event of such breaks.
The needed functions of safety-related systems are those functions required to mitigate the effects of the pipe break and safely shutdown the reactor plant.
The current criteria for review of pipe breaks outside contain-ment are contained in Standard Review Plan 3. 6. 1 and 3,. 6. 2 including their attached Branch Technical Positions.
BACKGROUNO In December
1972, the staff sent letters (Reference
1) to all power reactor licensees
'requesting an analysis of the effects of postulated failures of high energy lines outside of containment.
A summary of the criteria and requi rements in this letter is set forth below:
~ (
a.
Protection of equipment and structures necessary to shut down the reactor and maintain it in a safe shutdown condition, assuming a
concurrent and unrelated single active fa'lure of protected equip-
<ent shou d be oro'i'.
ed f"o'a>1 effec s resu
'lc'rom ra~ ures pipes carrying high energy fluid, where the temperature and pressure conditions of the fluid exceed 200'F and 275 psig, respectively, up to and including a double-ended rupture of such pipes.
Breaks should be assumed to occur in those locations specified in the "pipe whip criteria."
The rupture effects to be considered include pipe whi'p, structural (including the effects of jet impingement),
and environmental.
b.
In addition, protection of equipment and structures necessary to shut down the reactor and maintain it in a safe shutdown condition, assuming a concurrent and unrelated single active failure of pro-tected equipment, should be provided from the environmental and structural effects (including the effects of jet impingement) result-ing from a single open crack at the most adverse location in pipes'arrying.
fluid routed in the vicinity of this equipment.
The size of the cracks should be assumed to be 1/2, the pipe diameter in length and 1/2 the wall thickness in width.
In response to our letter and subsequent requests for additional information, Rochester Gas and Electric (RG8E, the licensee) submitted a report, "Effects of Postulated Pipe Breaks Outside the Containment Building," and several addi-tional letters providing information and schedules for plant modifications.
A complete bibliography of these letters is contained in the NRC Safety Evaluation Repor t (SER) for Amendment No.
29 for the Ginna plant (Ref. 2).
The SER for Amendment No.
29 also provides the NRC staff evaluation of certain facility modifications proposed by the licensee to provide protection from the effects of a postulated pipe break outside containment.
Reference 3 approved the licensee's augmented Inservice Inspection (ISI) Program which is intended to
'nsure a very low probability of pipe breaks at locations in the main steam t
and main feed systems where modifications to mitigate the effects of the
~
~
~
breaks could not be installed.
In addition, the licensee committed to make certain modifications in conjunction with the Systematic Evaluation Program (SEP) reevaluation of the effects of pipe breaks outside containment.
The NRC staff reevaluation of the effects of pipe brhaks outside containment under SEP Topic III-S.B involves the comparison of the Ginna plant with current criteria for pipe breaks outside containment..
The~taff used an "effects ori-ented" approach to determine the acceptability of plant response to pipe breaks, i.e.,
each structure,
system, component, and power supply which must function to mitigate the effects of the pipe break and to safely shutdown the plant was examined to determine its susceptibility to the effects of the postulated break.
Break effects considered were compartment pressurization, pipe whip, jet impingement,
spray, flooding, and environmental conditions of temperature,
pressure, and humidity.
This review complements that of SEP Topic III-12; "Environmental qualification of Safety-Related Equipment."
(The effects of potential missiles generated by fluid system ruptures and rotating machinery will be evaluated under SEP Topic III-4C, "Internal.ly Generated Missiles.")
e, The previous evaluation of pipe breaks outside containment for the Ginna Plant was performed using some methods and criteria which are no longer used by the siaff in tne, eview of cu. rani plants.,
0 exa;. pie, it>> cul ren def iniiion of a high energy fluid sysiem as ore
thaz,
>s maintain d uncer condiiions 4 el e either or both the maximum operating temperature and pressure exceeds 200 F
and 275 psig is different from the definition applied in the previous review where a high energy fluid system was one in which both temperature and pressure exceed 200 F and 275 psig.
The SEP reevaluation of this topic was performed using the criteria extracted from Standard.Review Plan 3.6.1 and 3.6.2 and their attached Branch Technical Positions.
Data for this assessment was gathered during a visit to the Ginna plant on September 25-27, 1979.
The staff issued its draft evaluation of this topic on June 24, 1980 (Reference 14).
The licensee was requested to provide a schedule for resolution of the open items.
The licensee response was transmitted by letter dated August 7, 1980 (Reference 15).
This information has been incorporated into this evaluation.
EVALUATION The results of the SEP reevaluation of pipe breaks outside containment for the Ginna plant are provided in Table 1.
The table lists the zones within the plant which contain systems required for safe shutdown and/or systems required to mitigate the effects of postulated pipe breaks.
These zones are the screen
house, diesel generator
rooms, intermediate building (elev. 293', 278'nd 253'), turbine building (elev. 289', 271',
and 253'), control room, relay room, battery rooms, mechanical equipment
room, and auxiliary building (elev.
271', 253',
and 235').
The safe-shutdown systems which were examined from the standpoint of protection
'rom pipe break effects are identified in the SEP Safe Shutdown Review for
".-. Gihn.(Ref;,'9)..
These-.systems are.
"=-=,
(a)
Reactor. Protection System (b)
Auxiliary Feed System (c)'ain steam safety, isolation, and atmospheric dump valves (d)
Service Water System (e)
Chemical and Volume Control System (f)
Component Cooling Water System (g)
Residual Heat Removal System (h)
Instrumentation for Shutdown and Cooldown (i)
Emergency Power (AC and DC) and control power for the above systems and components.
This section provides additional information used to evaluate certain pipe breaks listed in Table l.
Screen House Service Mater System (SWS) or fire system i~loderate Energy Line Breaks
(~1ELB's) and heating steam line breaks could result in the loss of the SWS by damaging 480V electrical buses 17mnd 18 or their associated electrical cabling.
Loss of the SWS would result in a plant trip because of the loss of several components cooled by the SWS such as the reactor feed pump lube oil systems, circulating water pumps, and the CCM system.
In accordance with current criteria, a pipe break which results in a reactor or turbine trip r'esults, in turn, in a loss of offsite power.
To supply AC power following a loss of offsite power, redundant emergency diesel generators are available;
however, the diesel generators are supplied cooling water by the SWS.
Therefore, the postulated pipe break could cause the total loss of AC power at the plant, and reactor core'decay heat removal would be dependent on the turbine driven auxiliary feed pump which is susceptible to a postulated single active failure.
The licensee has been evaluating the SWS in connection with the ongoing NRC fire protection review and the SEP reviews of flooding and tornado missiles.
To conduct a plant cooldown following a fire which causes a loss of all SWS with no offsite power available, the licensee has developed a procedure which is described in Ref.
4.
The procedure requires the installation of fire hoses from the city hydrant system to provide the diesel generators cooling water and to provide additional water to the auxiliary feed pumps for steam generator makeup water.
While the fire hoses are being installed, the turbine driven auxiliary feed pump is used to add water from the Condensate Storage Tank to the steam gener'ators for decay heat removal.
After a diesel generator is
operable, additional auxiliary feed pumps and the reactor coolant system charging pumps can be operated as required.
According to the procedure, fire hoses and portable pumps would have to be connected to one CCM heat exchanger if a plant cooldown to cold shutdown conditions were required with no SWS flow available.
The proposed procedure could be used for the pipe break case even if the turbine driven auxiliary feed pump is assumed to fail.
Without feedwater
addition, the steam generators can remove decay heat for approximately 50 minutes before they are boiled dry.. This time could be used to makeup the temporary diesel generator cooling connections to start a diesel generator and a motor driven auxiliary feed pump.
The staff's conclusion and position for resolution of these postulated pipe breaks in the Screen House and their associated equipment failures are contained in the CONCLUSIONS section of this report.
Intermediate Buildin Floodin As noted in several places in Table 1, flooding from pipe breaks in the Intermediate Building (IB) would flow via open stairways and hatch gratings to the sub-basement of. the IB.
Sufficient drainage area is available so that no appreciable buildup of water would occur on any floor of the IB except for the sub-basement.
No equipment necessary for safe shutdown or flood mitigation is located on this level; but, if the flooding condition went unchecked, the
~
IB 253'levation could be affected to a depth of about 30 inches.
Equipment on this elevation includes the auxiliary feed pumps and the reactor trip breakers.
If this equipment were flooded, a reactor trip would occur and the auxiliary feed system would be inoperable.
The standby auxiliary feed system, which is not located in the IB, would still be operable even if a loss of offsite power occurred.
Operation o
the sump pumps in the sub-basement is alarmed at the water treatment station.
In the control room an alarm is provided to alert the operator of an alarm condition at the water treatment station.
Intermediate Buildin Main Steam and Main Feed Breaks Postulated Main Steam (MS) and Main Feed (MF) system High Energy Line Breaks (HELB's) in the IB could result in the following:
(a)
The "A" MS line on the 293'levation could damage cable trays 16, 72, and 122 by jet impingement.
At this elevation, these trays contain control and power cables for the containment fan coolers and the containment purge exhaust fans.
These systems are not required to function to mitigate a
MS break outside containment or to shutdown the plant.
(b)
The 30" dia.
"A" MS line on the 293'levation could damage the north IB cinder block wall (whip or impingement),
an interior steel column supporting the IB floors above 293'whip), or the cable trays discussed in (a) above (whip or impingement).
(c)
On the 278'levation of the IB, large MS line breaks could damage both the floor supporting the MS header and MF line "A"; and a break at the juncture of the 36" dia.
and the 30" dia.
steam lines could overstress the anchors which connect the =lines to the IB structure.
(d)
A "B" MF line break on the 278'levation could damage one or more steam safety valves for the "A" steam generator.
"\\
4 g
++%%tW 5j
(e)
IB pressurization by a large HELB was predicted in Ref.
6, for the bounding case of the 36" dia.
MS line break, to. result in fai lure of the cinder block walls and roof beams and decking of the IB although the IB structure was not predicted to be damaged.
(f)
In Reference 6, it is stated that a "8" 30" dia.
MS line break outside the IB at the penetration.to the containment building could damage the control building by means of pipe whip.
Because of the severe consequences of these postulated MS and MF line breaks in the IB and because plant modifications to prevent these consequences were not practical (Ref. 7), the licensee unde>took a two-part program to reduce the vulnerability of the plant to a HELB in the IB.
The first part of the program was an augmented radiographic inspection
program, described in Ref. 8, to provide added assurance that postulated large MS and MF breaks would not occur.
This program was reviewed and accepted by the NRC staff in 1975 (Ref. 3). 'he second part of the licensee's program was to move essential equipment from the IB into locations unaffected by an HELB in the IB.
The intent of this program is to preclude the large (greater than the equivalent of six inch diameter) breaks and acceptably mitigate the small breaks, A
summary of plant modifications installed and equipment relocated is provided in Ref.
2.
I he 1 icensee has co";,m" ted to insta' addi tiona i modi fications in conjunc. ion with the SEP review of this issue.
These modifications would include the ns-tallation in the IB of jet impingement shielding for one steam generator atmospheric dump valve aod all MS safety valves.
In the Intermediate Build-ing, the.licensee committed to install jet impingement protection for the two main steam bypass valves and associated piping.
The staff has concluded that the installation of jet impingement shielding for the MS bypass valves and associated piping is not necessary since the bypass valves are not required for safe shutdown or pipe break mitigation.
Also, modifications to the IB cinder block wall resulting from the analysis of HELB's in the Turbine Build-ing will be made as necessary upon completion of the SEP.
The licensee's commitment is detailed in Ref.
10.
A comparison of the IB pressurization caused by a 6" dia.
HELB provided in Ref.
6 with the design limits of the IB cinder block wall provided in Ref.
11 shows that even this small HELB could fail the cinder block wall.
As a result of this failure, equipment in the Turbine Building could be damaged.
The only equipment which may be of concern from the standpoint-'of plant shutdown are the MF regulating valves and bypass valves on the 270'levation of the Tur-bine Building.
However, even if these valves were
damaged, the Standby Auxiliary Feed System (SAFS) would be available to feed the steam generators and effect a safe shutdown of the plant.
The SAFS was installed to provide steam generator feed in case a pipe break in the IB damaged the Auxiliary Feed System.
Turbine Buildin Main Steam and Main Feed Breaks Postulated MS and MF system HELB's in the Turbine Building (TB) could result in:
(a)
The'4" MS lines could whip into the IB wall at the proper elevation to damage the "B" NS line safety valves, atmospheric dump valve, and steam supply line to the turbine driven auxiliary feed pump.
(b)
NS and NF breaks could pressurize the TB itself.
The following pressures have been calculated:
Bre~
Location 20" MF 8.27.0'4" MS
.8 298'6" HS 8 270'2" MF 8
270'B 298'456 psi
. 589 psi
. 742 psi
.'33 ps i TB 270'nd 243'848 psi
.507 psi 1.26 psi
.259 psi These results are provided in Ref.
11 for the 20" NF break and in Ref.
12 for the other breaks.
The pressurization of the TB could. adversely affec" those areas adjacent to the TB in which safe shutdown or pipe break mitigating equipment is located.
These areas are the control room, diesel generator room, relav room, bat.ery room and he IB.
Again, because of the consequences of these postulated MS and MF line breaks in the TB, the licensee utilized the two-part program to reduce the vulnerability of the plant to these HELB's.
The licensee's previously approved augmented inspection program has been applied in the TB to MS lines 'larger than 12" dia.
and several locations on the 20" dia.
NF header.
The inspection program limits. the breaks which must,. be considered to a 12" MS or a 20" MF line break which are the largest potential double-ended breaks in locations which are not inspected.
Of these, the 20" MF is more limiting.
To protect the areas adjacent to the TB from the effects of HELB's, the licensee has installed pressure diaphrag~ walls between the TB and the control room, relay room, battery 'rooms, mechanical equipment
room, and diesel generator rooms.
The
. design differential pressure for these walls is 0.7 psi for the control room and 1. 14 psi for the other spaces.
The NRC evaluation of these walls is in
. Reference 2.
The pressure resulting from a 20" MF or 12" HS line beak in the TB is sufficient to cause failure of.the TB/ IB cinder block walls (design pressure
.13 osid).
If.
these walls failed, the following systems and components could be damaged by fal-ling cinder blocks or adverse'enviormental conditions:
one containment purge exhaust fan on the IB 298'lev.,
the auxiliary feed system (AFS) steam supply valves on the IB 278'lev.,
and the AFS turbine driven pump, reactor trip'reak-
ers, and reactor rod control motor generator sets on the IB 253'lev.
The purge exhaust fan is not required to function to mitigate a
HELB outside containment.
The rod control mbtor generators and reactor trip breakers fail safe if damaged and would not prevent a reactor trip (core shutdown).
The AFS function is required for a safe shutdown;,however, the SAFS has been installed
by the licensee to accomplish this function if a HELB disables the AFS.
The turbine driven AFS pump is not specifically required to operate following a postulated HELB.since, even if offsite power were assumed to be lost, the redundant emergency diesel generators w'ould be available to power the two SAFS pumps or the remaining two AFS pumps all of which ave driven by electric motors.
Only one of these four motor driven pumps is required for a plant shutdown and cooldown.
The discussion in the previous paragraph shows that most of the equipment which can be damaged by a failure of the TB/IB block wal 3 is not necessary for HELB mitigation or safe plant shutdown.
However, the MS isolation valves and MS safety and atmospheric relief valves aWe necessary for HELB mitigation and safe shutdown.
Although the safety valves would probably not be rendered inoperable by failure of the TB/IB walls, the licensee will be requested to assess this possibility and consider incorporating protection of the valves with the jet impingement shields to be installed.
Both atmospheric dump
'valves would have to be protected from the effects of the wall failure.
Batter Room/Mechanical E ui ment Room Floodin A SMS or fire system MELB in the mechanical equipment room could flood both battery rooms and result in a loss of all emergency OC power.
A 20" diameter S4S line brea'~ ir, the mechanical equipment room would result in a calculated flooding rate of 585 gpm using the methods of Ref.
5.
Ho sump level or flood alarms are installed in this space or in the battery rooms which are connected to the mechanical equipment room by normally closed non-watertight doors.
The licensee has committed to replace the non-watertight doors by a wall.
Auxiliar Feed S stem Breaks on the 253'levation of the IB The AFS discharge lines from the pumps in the IB (253'lev.) to the "B" MF header run along the north wall of the IB at approximately the 270'levation.
A break in this line, which is a high energy line, could result in pipe whip or jet impingement on cable trays and containment electrical penetrations in that area.
(The steam lines for the turbine driven AFS pump are also in this area but are not considered high energy lines since they are not pressurized during normal plant conditions.)
Reference 4 presents an analysis of plant shutdown capability following an exposure fire in this area which destroys all electrical cables and equipment in the area'.
This condition envelopes the damage which could be done by the AFS HELB.
To provide safe shutdown capabi-lity following the fire, the licensee has proposed methods and identified plant modifications to be installed (Ref.
~ 4).
Upon completion of these modifications and because of previously installed modifications, specifically the standby AFS and relocation of safe shutdown instruments from the IB, the plant will have an acceptable level of protection from the effects of AFS breaks on the 253'levation of the IB.
CONCLUSIONS Based on the information submitted by the licensee and obtained during the site visit to the Ginna plant, we have.'determined that the following review areas have 'not been adequately addressed in previous staff safety evaluations and should be resolved with the SEP:
1.
SWS and fire system MELB's and heating steam line breaks in the screen house could result in the loss of all SWS flow, by damaging Buses 17 and 18, and the loss of all AC power.
The licensee is implementing a method to provide cooling to the onsite emergency diesel generators which is not dependent on the SWS.
The staff position regarding these pipe breaks is that the licensee must.
provide protection for Buses 17 and 18 and their associated cables from the effects of the breaks in accordance with Standard Review Plan 3.6. 1 consistent with the modifications which must be performed on the SWS to accomodate other ongoing SEP reviews, e.g.,
the tornado missile and fire protection reviews.
2.
Based on our evaluation of Hain Steam (HS). and Main Feed (MF) line breaks in the Tu. bine Building and Intermediate Building, the licen-see should provide protection from the effects of the failure of the Turbine Building/intermediate Building cinder block wall for the MS atmospheric dump valves and assess the heed for and provide protec-tion as necessary for the HS safety valves.
The proposal should be submitted for staff review.
TABLE ".
FFFECTS OF PIPE BREAK OUTSIDE COHTAIHMEHT
~
! I I ~
1 'f ~~ rc.8P t
~
<. i~.
Zone Affected Affected Safe Pi e Break Miti atin S stc.m Shutdown S stem Adequacy.-of Pro(ection/Remarks Screen House SWS (MELD)*
or Fire System (MELB)
Hone CW (MELB)
Hone Heating steam Hone (HELD)
~ I SWS Power Supply Bus 17 8
18
SWS, Bus 17 8
18 SWS Power Supply Bus 17 8c 18 SWS or fsre system leak can affect,bot6,!(
and cause loss of all SWS pumps.
.SeII,"~I")'4 "
discussion in EVALUATION section.
'".h!4>;~,.-.
(
Adequate.
Previ ously analyzed iri.bJ.<<y-.'.
flooding evaluation (Ref. 13)..; "i',.
Potentially inadequate.
High tempeV'5",>
ture environment effects on cables,ko.".'g.
Buses 17 and 18 could cause loss of .~..*~;
gjij'.,j~)I'jI 5WS.
See remarks above.
Potentially inadequate.
Spray from a '"~<,",'f>>
=
Diesel SWS (MELD)
Generator
~
or Room lA (253') Fire System (MELD)
Hone Diesel generator 1A
'i"">-'- '.>@@
Adequate.
Spray from MELB may affect':"-.;j(
enerator.or associated electrical-:
',='-: "I !)Il'L I
panels, but redundant diesel gener5toP'>:,.-s'
..-!-)!'lg o and offsste power are available as
<.~.;,gg backup.
Flooding in room is detecteII<~I,,p(
by sump pump alarm in control room,",i.~I'!""
Cable vault below diesel generator, c,'.'.~ "
Heating steam Hone (HELB) l Diesel generator lA room is protected from flooding by,':;-,~If]~".,
waterti ht manhole cover.
k
~'$lgfl Adequate.
Fire protection temperatiii 0"Il:.P!~l!~,'etector warn control room of high
..:4<4'!~"-..
temperature conditions.
No LOP; other'-"-<P~+"".';,>
diesel available.
Diesel SWS (MELB)
Generator
'r Room 1B (253') Fire System" (MELB)
Hone Diesel generator 1B P(pj'slg
<<c~ f~ji>, fbi>>
diesel 1A passes through 1B diesel
=J room but leaka e from a crack brea(.!ih' this SWS line would not be enough to,'.i.>
1 render the lA diesel inoperable th'1'dU)h" loss of cooling water.
Adequate.
See comments above for MELO'-"~.;."-.'~r" in lh diesel room.
SWS supply line '46 <'>>'"c'I"'-.';
A list of abbreviations is provided at the encl of this table.
-:!'@$iigg
'iI>>j)ji'<<>>yl Zone Diesel Heating Steam.
Generator (HELB)
Room 1B (253')
(continued)
None Diesel generator 1B Intermediate Fire system Building (HELB)
(293')
None None HS and HF Various; see (HELB) evaluation section Lcrack breaks] of this report.
Various; see evaluation section of this report.
HS, "A" and "B" headers, HF "B"
s header (HELB)
I.'large breakj
Various, see" evaluation section of this report.
t'arious; see evaluation section of this report.
Affected.
Affected Safe Pi e Break Miti atin S stem Shutdown S stem Adequacy. o f.
Protection/Remarks
~ j>>>>shiit
'\\
.;,:.;p,;.'ding>p Adequat'e.
See remarks above f6r,.";.>:Id/~!/I'",f~
~.
heating steam leak ! n d>>esel room,:<<A:,.Sall', $
'-'s1 I!.he'.t:r'
<<ie>g L s
Adequate.
sae evaiuat>>on sect!on g$,,I
i s'E- '"'"tr'~
Adequate.
Jet impingement from a craCkii. P,.>."-I in "A" h15 line could impact cable,tf'a)5sII ~<j;4
)4.*
W' of this report 16, 72, 122.
Although these trays aI'e '-.'I'Jf,'<~-,
safety related, at the 293'levate)oh,"i>i'";et~5.
they contain no cables needed to m]H>>s,.'>>,i'd%,
s t
sj".ls b;
gate the effects of the break or t0 '1 ";.'",i0~~Q><=,~",
safely shutdown.
Environmental effectS It>>j~4'<,'(.
of MS and MF crack breaks would be s'-,'.;i;,'<>~>.ggi 1
te
~
~ '
I>> ~
'. y'
. )>>
Adequate.
Although a large HS or, HF,;;::,'~
line rupture in the intermediate bu)it[";,"'.
ing has the potential to structural,lp!!','g damage the building, these ruptures:i'it'h<
effectively precluded by the liceliki.e',.:4';
ongoing inservice inspection of thI~
~
g<<~)c' i mg
>>s>>
ff't
=
fi t
",y".'t
'h
-'l
',i s
.>>>>'umts,
~ g
- ~
'r eke.a(p,...
lines.
' i",.'> -.g$'jar f]
s exper>>enced throughout tha>>ntermadihte,,kt!>>g!.p.,
building.
Licensee has modified thh'.j~4'.~f'fj";j',@
plant to withstand these conditions]fn
<I$'@
gas the intermediate building and to $reyetih'8g<Pp.">>
these conditions from spreading to.'the(k auxiliar buildin Refs.
2 and 10
'1"",
Zone Affected Affected Safe Pi e Break Hiti atin S stem Shutdown S stem Adequacy of Prot. ection/Remarks Intermediate Building (278')
(continued)
AFS (HELD)
None HSIVs,Atmospheric HS dump valves, HS safety valves, AFS turbine driven pump steam supply valve I'ntermediate Fire system Building (HELB)
(278')
None None HS and HF Various; see (HELB) evaluation section I.crack breakj of this report.
Various; see evaluation section of this report.
HS and HF (HELB)
)large break]
(
Various; see Various; see evaluation section.
evhluation section.
not render these inoperable.
,Impinge.,
l >>pg>g.
ment on AFS turbine driven pump steam'=;.q',,i)".I~9.'upply valve or either atmospherio.'.dumog'"';v~f"-'i" air control system could render theAg'~ilt~j~y<fq-inoperable;
however, the turbine dHVbh"j(((gg pump is not normally used for safe,khUt"".',-~.".,<il9I down and the function of steam gener4to0i;f"<
(<
makeup can be performed by other AFS~v.Ange'j," )
SAFS pumps, and the atmospheric dumph.c45" j;, @
be manually operated by handwheelsl!e:,t'Clif, CI fll Adequate.
See remarks above for HELB;.':-'."';>14~~"':
at 293'levation.
'Pelf isiifl~!
has protected~ t;ot'-'gttz'j the breaks or.: Kb'.,'.:"-'<,'":(,'4c Adequate.
Licensee moved instrumentatio gate the effects of safely shutdown.
This is further.edit",.;d'i!gg n the evaluation section ogf/it's>>ily~k.,
Adequate.
See remarks for large l(Svlor,, r> (k.,
el evation of-".'.,-;*.';~:A', '~)~~~g<
MF-line break on 293'ntermediate building Adequate.
Jet imPingement Of AFS:Watdl;4ri ji,l~qqq
("BO F) on MSIVs, safety valves wou)d '-rs><<','tf<'.,
\\
m, w
Some protection will be installed in conjunction with the SEP review of the facility.
July 21, 1981 Docket, No. 50-244 LS05-81-07-070 Nr. J ohn E. Nai er, Vice Pres ident Electric and Steam Production Rochester Gas 8 Electric Corporation 89 East Avenue Rochester, New York 14649
Dear. Nr. Ilaier:
SUBJECT:
SEP TOPICS II-l.B, POPULATION DISTRIBUTION AND III-4 D, SITE PROXIMITY MISSILES - R. E.
GINNA CO b
-.s JUL
~toi<
g 9
leS1
~19S92 CO
.~o Enclosed are the staff's final evaluations of SEP Topics II-1.B.and III-4.D for the R. E. Ginna Nuclear Power Plant.
These evaluations are based on our review of your topic safety assessment reports sub-mitted by letters dated April 15, 1981 and April 16, 1981, respectively.
You will note that we have revised your calculated population density which is more properly obtained by dividing the total population within a given distance by the total area of the complete circle (including both level and water) whose radius is the distance of interest.
This completes our evaluation of Topics II-1.B and III-4.D.
These evaluations will be a basic input to the integrated safety assess' ment for your facility unless you identify changes needed to reflect the as-built conditions at your facility.
These assessments may be revised in the future if your facility design is changed or if NRC criteria relating to this subject are modified before the integrate assessment is completed.
Sincerely, Si07240143 Bi072i PDR ADOCK 05000244 P
PDR
Enclosure:
As stated Dennis N. Crutchfield, Chief Operating Reactors Branch No.
5 Division of Licensi A
DL GL inas 7/+/81 UL 30 19M OFFICE/
SURNAME/
BATE/
cc w enc
.""See.next osure:
gage
~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~
SEPB:DL
~22 ~ ~ ~ ~ ~ ~ ~ ~ ~
GC Qa:dk 7//6/81
~ ~
~
~ 1 ~ ~
~ ~ ~ ~ ~ ~ ~ ~ ~ ~
SEPB:D 0 ~ ~ ~ ~ ~ ~ ~ ~ ~ ~
~
~ ~ ~
CBerlinger
?l5(81 SEP
~ ~ ~
)I
~ ~ 0
~ ~ ~
WR ssell 7/4/81
~ ~ ~
~ ~ Q ~ I ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 0 Snaideri-81
~
~ ~ ~ 0 ~ ~
~ 0 ~ ~ ~ ~ ~
ORQBP~I'utchfield
~ ~
~ ~
~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~
7PP/81
~ ~ T 1%0
~ ~ ~ ~ ~
~ ~ ~ ~
NRC FORM 318118/80) NRCM 8240 OFFICIAL RECORD COPY
USGPOs 198~29.82e
t P
II'5
~
A II H
J
-I t
~
Mr. John E. Maier CC Harry H. Yoigt, Esquire
LeBoeuf, Lamb, Leiby and MacRae 1333 New Hampshire Avenue, N.
W.
Suite 1100 Washington, D. C.
20036 Mr. Michael Slade 12 Trailwood Circle Rochester, New York 14618 Ezra Bialik Assistant Attorney General Environmental Protection Bureau New York State Department of Law 2 World Trade Center New York, New York 10047 Jeffrey Cohen New York State Energy Office Swan Street Building Core 1, Second Floor Empire State Plaza
Albany, New York 12223 Director, Bureau of Nuclear Operations State of New York Energy Office Agency Building 2 Empire State Plaza
Albany, New York 12223 Rochester Public Library 115 South Avenue Rochester, New York 14604 Supervisor of the Town of Ontario 107 Ridge Road West
Ontario, New York 14519 Resident Inspector R. E. Ginna Plant c/o U. S.
NRC 1503 Lake Road
Ontario, New York 14519 Mr. Thomas B. Cochran Natural Resources Defense Council, Inc.
1725 I Street, N.
W.
Suite 600 Washington, D. C.
20006 U. S. Environmental Protection Agency Region II Office ATTN:
E IS COORDINATOR 26 Federal Plaza New York, New York 10007 Herbert Grossman, Esq.,
Chairman Atomic Safety and Licensing Board U. S. Nuclear Regulatory Comnission Washington, D. C.
20555 Dr. Richard F. Cole Atomic Safety and Licensing Board U. S. Nuclear Regulatory Commission Washington, D. C.
20555
'r.
Emmeth A. Luebke Atomic Safety and Licensing Board U. S. Nuclear Regulatory Commission Washington, D. C.
20555
R.
E.
GIHNA SYSTEMATIC EVALUATION PROGRAM TOPIC I.
INTRODUCTION The safety objective of this topic is to ensure that the previously-established low population zone and population center distance specified for the site are compatible with the current population distribution, and are in accordance with the guidelines of 10 CFR Part 100.
II.
REVIEW CRITERIA Sections 100.10 and 100.11 of 10 CFR Part 100, "Reactor Site Criteria" provides the site evaluation factors which should be considered when evaluating sites for nuclear power reactors.
These sections include guidelines for determining the exclusion area, low population-zone and population center distance.
III.
RELATED SAFETY TOPICS Topic II-l.A, reviews the licensee's control over the exclusion area.
Various other topics will evaluate the capabiity of the plant to meet the dose criteria of 10 CFR Part 100 at the exclusion area boundary and low population zone.
The adequacy of emergency preparedness planning for the area surrounding the plant including the low population zone is being assessed by the Conmission in a separate review effort.
IV.
REVIEW GUIDELINES The review has been conducted in accordance vrith Standard. Review Plan (SRP) Section 2.1.3, "Population Distribution."
V.
EVALUATION The R.
E. Ginna site is in the township of Ontario, in the northwest corner of Wayne County, New York, on the north shore of Lake Ontario about 20 miles ENE of the center of the City of Rochester and 40 II miles WSW of Oswego.
The land surrounding the site is primarily of an agrarian nature and sparsely populated.
There are no substantial population centers, industrial complexes, transporta-Cion arterials,
parks, or other recreational facilities within a tMee mile radius of the Ginna site.
The City of Rochester is the larcrest population cente within a 50 mile radius of the site (241,539 people, with 701,745 in the metropolitan area
).
The 7
nearest community with a population of 1,000 or more is the Town of Ontario with its center located about 3~~ miles from the site.
The preliminary estimated 1980 census or the Town of Ontario is 7,452.
To develop the Wayne County and,"',onroe County Radiological Emergency
Response
Plans for the R.
E.
Ginna Nuclear Power Station, a recent survey of the population within a five-mile radius vras completed.
Figure J-2 from the Wayne County Radiological
Response
Plan, reproduced as Figure 1
of this evaluation, details the population whin 5 miles of Ginna, based on preliminary 1980 population estimate.
RG&E estimates that 10,864 persons reside within five miles of the plant, a density of 138 persons per square mile averaged over the entire area.
(It should be noted that this figure compares favorably with the 1980 population projection of 10,934 persons shown in Figure 2.4-2 of the Ginna FSAR, which was published in 1968).
Other than the residents of the area, there are no large groups of transients within five miles of the site.
The only parks near the site are Webster Beach Park in Monroe County, approximately 6 miles west of the plant site, and B.
Forman Park in Wayne County, approximately 8 miles'ast of the plant site.
There are no federal recreational facilities in the area.
There are no state parks, public campsites, or special use areas within ten miles of the plant.
Wayne County 2
does have a migrant labor population, primarily for apple picking, during the June-October season.
Approximately 115 farmworker camps of five or more persons are scattered throughout Wayne County
, with 8
a total population of about 4400 migrants.
Information from Rural New York Farmworker Opportunities shows that there're only 12 camps, with 10 about 130 migrants, located in the vicinity of the Ginna site.
The nearest population center to the Ginna site containing more than 25,000 residents is the "Rochester urbanized area,"
whose eastern boundary is about ten miles from the site.
The only other population center of more than 25,000 persons is the City of Auburn (population 32,442),
located more than 40 miles Sf of the site.
The low population zone specified for the Ginna site is the area within a 3 mile (4,827 meter) radius of the plant.
A review of 9
current population estimates and projected growth estimates indicate that the population growth in the area since the plant received an operating license in 1969 has been modest, and this trend is expected to continue.
No population center of 25,000 residents has deVeloped, or appears likely to develop, closer than the eastern boundary of the Rochester urbanized area.
VI.
CONCLUSION The staff concludes that the low population zone and population center distances specified for the Ginna site is in conformance with the require-ments of 10 CFR Part 100 in that the population center distance is more than one and one-third times the distance from the reactor to the outer boundary of the low population zone (10 miles vs.
3 miles).
Me further conclude that the site conforms to the current licensing criteria.
This completes the evaluation of SfP Topic II-l.B for the Ginna site.
YEL.
REFERENCES 1.
Rocnester Gas and Electric Corporation, Robert Emmett Ginna Nuclear Power Plant Unit No.
1 - Final Facility Desczipt'on and Safety Analysis Report (FSAR), Sections 2.2 and 2.4.
2.
Rochester Gas and Electric Corporation, R. E. Ginna Nuclear Power Plant Unit No. 1, Environmental Report, Volume 1, Sections 2.1 and 2.2.
3.
Nuclear Regulatory Commission NUREG-75/087, Standard Review 4
Plan, Section 2.1.3, September 1975.
Code of Federal Regulations, Section 10, Pazt 100 (10 CFR 100).
5.
Wayne County Radiological Emergency
Response
Plan, Draft. Rev.
B, November 1980.
6.
Monroe County Radiological Emergency
Response
Plan Draft, Rev.
B, November 1980.
7.
Preliminary Report, 1980 Census of Population and Housing,
\\
New York, published by the Bureau of the Census, U.
S. Depart-ment of Commerce, February 1981.
Aaril 13 8.
Conversation with the New York State Heal& Department, April 3,
1981.
9.
Safety Evaluation by the Division of Reactor Licensing, U-ST
'I Atomic Energy Commission in the Matter of Rochester Gas and I
Electric Corporation Robert Emmett Qinna Nuclear Power Plant Unit No. 1, Docket No. 50-244 (SER), Section 2.1, June 19, 1969.
10.
Letter, Thomas J. Harris, RNYFO, to George Wrobel, RG&Z, April 10, 1981.
11.
Rochester Gas and Electric Cornoration, Ginna Nuclear Station Radiation Emergency Plan, Proposed January 1981.
12.
New York State Radiological Emergency Preoaredness
Plan, December 1980.
~ ~
~
~ ~
~ ~
~
I g.AyF'.
tthtIW A
ON7ARIO
'GINhlA. SITE I4 20 Il2 a
I7 l00
~210 - ~"
~ ~N---.
4I5 ttlmlmll ll rP 200 30 PtjjtAAtfg ~ttEt'AAD D~~
206
~255 o
'S 00
'l2 03 30 CI< Cl tt l2I 2SO Xa
~wsw SCt ll,EGEl.
M 8S IEF oK I,I O
.DRKPMt l62 O750 R
I61 CI.I: "ID AO 3
'B 657 0Z III CNYOtI R
672 axe 260 500 At I.E trn l257 w o
SE EY RO
~276 3 Ea>>.m ES VILL/ IV YIDDDS.
RD
~ llI x'x xx I~JOO I'OPULATIONESTIMAfCS
/i@72 I
w CLEVEttGER no 5
S otx O
Q020 SSE S-RIDGE AD ill O
PIIACtt /
Rb
'(
FIGURE j.,
0 I:
.I
) /
0-5 MII.E I~JOO POPULATION ESTIMATES J-5
o e
R. E.
GINNA CTCTRPIATIC ~PRORRAII TOPIC S
I.
INTRODUCTION The safety objective of this topic is to ensure that the integrity of the safety-related structures, systems and components would not be jeopardized due to the potential for a site proximity missile.
II.
REVIEW CRITERIA General Design Criterion 4, "Environmental and Missile Design Basis."
of Appendix A, "General Design Criteria for Nuclear Power Plants,"
to 10 CFR Part 50, "Licensing of Production and Utilization Facilities,"
requires that nuclear power plant structures, systems and components
'mportant to safety be appropriately protected against events and conditions that may occur outside the nuclear power plant.
III.
RELATED SAFETY TOPICS Topic II-l.C, "Potential Hazards or Changes in Potential Hazards Due to Transportation, Institutional, Industrial and llilitary Facilities" provides a description of the potential missile hazards.
IV.
REVIEW GUIDELINES The review was conducted in accordance with the guidance given in Standard Review Plan (SRP) Section 2.2.3, "Evaluation of Potential Accidents," 3.5.1.5, "Site Proximity t/issiles (except Aircraft),"
and 3.5.1.6, "Aircraft Hazards."
V.
EVALUATION The potential for hazardous activities in the vicinity of the Ginna plant has been addressed in SEP topic II-1.C, "Potential Hazards due to Industrial, Transportation, Institutional and Military Facilities".
As indicated therein, there is little industrial activity near the plant.
The distances to the nearest land transportation routes are such (about 1700 feet to the nearest
highway, and 3 1/2 miles to
~ the nearest railroad) that the risk associated with potential missiles from transportation accidents on these routes are within the SRP 2.2.3 guidelines.
Similarly, the nearest large gas pipelines are about six miles from the plant, and do not pose a missile threat to the plant.
Najor Lake Ontario shipping routes are also sufficiently far away (about 23 miles) so as not to present a credible missile hazard from lake traffic.'here are no military facilities or activities near the plant which would create a, missile hazard.
The review of SFP Topic II-1.C also evaluated the potential fbr aircraft becoming a missile hazard, both in connection with the operation of the Williamson Flying Club Airport, which is about ten miles ESE of the plant, and due to commercial air traffic in and out of Rochester via federal airways V2N and V2, which are 2 1/'2 and 10 miles f om the plant site.
As evaluated in Topic XI-1.C, it was determined that, since the Williamson Flying Club Airport expected, a maximum of only 5000 operations per year, and is about 10 miles from the site, the criteria in XXI.3.a and IXI.3.b of SRP 3.5.1.6 were met, and there is no need to determine the probability of an aircraft crash into the plant.
Further, the harard to the plant from commercial
-8 aircraft use of airways V2 and V2N was shown to be only 5.1 x 10 and 1.4 x 10 per year, respectively.
No danger to the plant from commercial airline traffic is thus expected.
Conclusion Since current regulatory criteria are met with respect to SrÃ Topic IIX-4.D,. "Site Proximity Missiles", it can be concluded that this topic is complete for the R.'.
Ginna site.
No additional review for this topic is recruired during the SEP integrated assessment.
YI.
REFERENCES 2.
3.
a.
5.
6.
7.
Rochester Gas and Electric Corporation, Robert Emmett Ginna Nuclear Power Plant Unit No.
1 Final Facility Description and Safety Analysis Repozt (FSAR), Sections 2.2 and 2.5.
Rochester Gas and Electric Corporation, R. E. Ginna Nuclear Power Plant Unit No. 1, Environmental Report, Volume 1, Sections 2.1 and 2.2.'uclear Regulatory Commission NUREG-75/087, Standard Review
Plan, Sections 2.2.1, 2.2.2, 2.2.3, and 3.5.1.6, September 1975.
Code of Federal Regulations, Section 10, Part 100 (10 CFR 100).
Sterling Power Project Nuclear Unit No. 1, Preliminary Safety Analysis Report Addendum, Rocnestez Gas and Electric, Volume 1, Sections-2.1 and 2.2.
U.S. Nuclear Regulatory Commission Regulatory Guide 1.91, Rev.
1, February 1978.
Z,etter, John E. Maier, RG&E, to Dennis M. Crutchfield,
NRC, SFP Topic II-l.C, "Potential Hazards Due to Transportation, Industrial, Institutional and Military Facilities",
April 15, 1981.

ML17309A193

Contents

Text

Dear Hr. Maier:

SUBJECT:

Enclosure:

Response

Response

Response

Response

Dear Hr. Mafer:

SUBJECT:

Enclosure:

Dear Mr. Maier:

SUBJECT:

Enclosures:

References:

Background

Dear Hr. tlaier:

SUBJECT:

Response

Response

Response

Response

Enclosure:

Dear. Nr. Ilaier:

SUBJECT:

Enclosure:

Response

Response

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ML17309A193

Text

Dear Hr. Maier:

SUBJECT:

Enclosure:

Response

Response

Response

Response

Dear Hr. Mafer:

SUBJECT:

Enclosure:

Dear Mr. Maier:

SUBJECT:

Enclosures:

References:

Background

Dear Hr. tlaier:

SUBJECT:

Response

Response

Response

Response

Enclosure:

Dear. Nr. Ilaier:

SUBJECT:

Enclosure:

Response

Response

Response

Response

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