ML20236L715

Final Rule 10CFR50, Mod of GDC 4 Requirements for Protection Against Dynamic Effects of Postulated Pipe Ruptures. Rule Allows for Removal of Pipe Whip Restraints & Jet Impingement Barriers

ML20236L715
Person / Time
Issue date:	10/22/1987
From:	Chilk S NRC OFFICE OF THE SECRETARY (SECY)
To:
References
FRN-52FR26393, RULE-PR-50 NUDOCS 8711110036
Download: ML20236L715 (34)

Text

_

@H flVM3ER

,7 1

PROPOSED RULE b-db' 1

~

1 Y

(S/ fL)6MB)

+

'00LKETED USMC

[7590-01, Ti OCT 23 40 $

NUCLEAR REGULATORY COMMISSION.

4FFICE 0' '.Mr iAs V 10 CFR Part 50

$$CKEi g E W.

' Modification of General Design Criterion 4 Requirements

'for. Protection Against Dynamic' Effects of Postulated Pipe Ruptures j

'l AGENCY:

Nuclear Regulatory Commission.

. ACTION:

Final Rule.

SUMMARY

The Commission is amending its regulations to broaden the scope 'of a-recent modification'to General Design Criterion 4 (GDC-4). The amendment would allow the removal of numerous pipe whip restraints and jet impingement barriers as well as other related changes in all reactor types.- Implementation of the amendment will increase safety since inadvertent restriction of pipe growth due

' to thermal expansion and associated stresses leading to pipe cracking is avoided.

Also, the duration of. inservice inspection will be reduced, yielding subst' ntially less occupational radiation exposures.

a EFFECTIVE DATE:

November 27, 1987 ADDRESSES: Copies of the written public comments are available for public in-spection and copying for a fee at the NRC Public Document Room at 1717 H Street, NW, Washington, DC.

J (vtj/{,h N'N 8711110036 871022 gg PDRm A L _

m

w' o

. m;

.

4 5;

[ q;

' ~

'l u

i.h b '

\\

L

[7590-01]

l i.'

l

[

t FOR FURTHER INFORMATION CONTACT: John A. 0'Brien, O'ffice of Nuclear Regulatory tResearch, U.S. Nuclear Regulatory Comission. Washington. DC 20555, Telephone ~

l 1

i (301)443-7854.

l i

1

~

SUPPLEMENTARY INFORMATION:

l Table of; Contents

.I.

Background

II. : Scope of Rulemaking III. Final l Rule.

1 LIV. Acceptance Criteria

/

1 V.

Invitation -'to Coment i

VI.

Issues Analysis VII. ' Availability of Documents VIII.

Finding of No.Significant Environmental Impact: Availability IX. Paperwork Reduction Act Statement-X. ' Regulatory and Backfit Analyses XI.. Regulatory Flexibility Att Certification XII. List.of Subjects in 10 CFR Part 50

.r

On July 23, 1986, the Comission published a proposed rule on the broad scope y,

modification to General Design Criterion 4 of Appendix A,

10 CFR Part.50 1

m (51 FR 26393).

This proposed rule contained a sumary of the acceptance criteria which the Comission had developed.

A 60-day public comment period

{

j.

was allowed. Twenty-eight written comments were received from utilities, 2

_ =_=

u-h

3 4

3 (7590-01]

reactor vendors, architect-engineering companies, industry groups, consulting firms and a citizens group.

There was no overt opposition to the proposed rule; each commenter supported the proposed rule or its intent either in part or entirely.

However, the citizens group expressed certain legal reservations which are addressed below in issues 20 and 21. A compilation of the twenty-one issues raised as.a result of public coments and the accompanying Comission response. is given under issues Analysis.

The text of the final rule is identical to the ' text of the proposed rule.

.The final rule should be applied consistently with the guidance in this Supplementary Information.

BACKGROUND Background to this rulemaking can be found in the limited scope modification to q

GDC-4 published as a proposed rule in the Federal Register on July 1, 1985 (50 FR 27006).

Research performed by the NRC and industry, coupled with operating experience, has indicated that safety can be negatively impacted by the place-ment of protective devices such as pipe whip restraints near certain piping.

The Comission adopted a two-step approach to the modification because safety improvements could be ouickly realized without extensive and time consuming review and discussion if the scope were initially limited to the primary main loop piping of PWRs.

The Commission decided not to defer the limited i

application of leak-before-break technology while the detailed provisions of the acceptance criteria were being reviewed and approved.

Many near term l

operating license (NT0t.) nuclear pnwer plant units and operating nuclear l

power plant units had requested exeeptions from the requirements of G00-4 l

and could benefit from the limited scope rule.

A broader application of 3

L_

p?.!. '

l

'}

~y j

'- y f

l

.[7590-01]'

leak-before-break 1 technology'. requires adoption.of the general - criteria pub-

]

~

lished.in. NUREG-1061, Volume 13, Chapter 5, November.1984, entitled " Report of.

1 the U.S.. Nuclear Regulatory Commission Piping Review Comittee, Devaluation of Potential for Pipe Breaks".

o j

SCOPE OF RULEMAKING l

This'rulemaking modifies General Design Criterion 4'to the extent that. dynamic effects of pipe ruptures in nuclear power units may be excluded from the design basis provided it is demonstrated that the probability of pipe rupture is ex-i tremely low under conditions consistent with the design basis for the piping,.

s Dynamic effects of pipe rupture covered by this rule are missile generation, l

7 pipe whipping, pipe break reaction forces, jet impingement forces,'deccmpres-sion waves within the ruptured pipe.and dynamic or nonstatic pressurization in y

cavities, subcompartments and compartments.. Hcwever, cavities, subcompartments J

and compartments necessary to the containment function are not affected by this 1

modification.

l h

l

. To retain high safety margins, the application of leak-before-break tec no ogy S

to various piping systems should not decrease the capability of containments to i

perform their ' function of isolating the cutside environment from potential leaks, breaks, or malfunctions within the containment.

Containments will con-

'tinue to be designed to accommodate loss of coolant accidents resulting from l

breaks in the reactor coolant pressure boundary up to and including a break l

t equivalent in size to the double-ended rupture of the largest pipe in the j

1 f.

l 4

)

i

WT

. )

m V.; g L.

_[7590-01]

reactor coolant system. 'Also, the functional design for emergency core cooling systems still retains nonmechanistic pipe rupture.

Environmental; qualification of electrical and mechanical equipment is discussed under issue 4 below.

I q

l This amendment-to' GDC-4 allows exclusion from th'e design basis of dynamic ef.-

. fects associated with high energy pipe rupture by application of. leak-before-break technology.

~0nly high energy piping in nuclear power units that meets

y rigorousL acceptance ' criteria is covered.

High energy piping is defined as those systems having pressures exceeding 275' psig or temperatures exceeding'

.i

..0 200 F.

i i

Studies completed by Lawrence Livermore National Laboratory under contract to

.c the NRC-indicate that adverse safety implications can' result from requiring protective devices to resist the dynamic effects associated with postulated pipe rupture. The placement of pipe whip restraints degrades plant safety when thermal growth is inadvertently restricted, reduces the accessibility for.and effectiveness of inservice inspection, increases inservice inspection radiation dosages and adversely affects construction and maintenance economics.

i FINAL RULE l

1 The final rule consists of a substitute sentence at the end of GDC a (replacing j

the sentence' introduced by the limited scope rule) permitting the use of analyses to exclude dynamic effects of pipe ruptures in all high energy piping j

in all-nuclear power units.

A deterministic fracture mechanics evaluation is mandatory.

Evaluations of the potential for water hamer, corrosion, creep j

5

p,,

ej i;

k ',

'j it.

'[7590-01]

4

\\

, damage, fatigue, erosion, environmental conditions, indirect failure mechanisms A'

and' other degradation ~ sources which could lead to pipe ' rupture are..also required. 'In order to' demonstrate-that the probability of fluid system piping rupture 11s extremely Tlow, applicants and licensees may.. follow procedures and acceptance criteria developed.by the' staff, b

The supporting safety analysis must demonstrate from the results of a fracture mechanics analysis that a substantial range of stable pipe crack Lsizes can 7

exist for an. extended period which provides detectable leaks, and that..~the-

~

fluid systems piping.will not rupture under these' conditions consistent with the design' basis for the piping.'

The language of the rule specifies " conditions consistent with the design' basis for the piping."

Th'e design basis'for. piping means those c6,nditions specified

~

in the FSAR, as amended,'and may include 10 CFR Part 50 (especially the General Design Criteria in Appendix A to Part 50),' applicable sections of the Standard Review Plan, Regulatory Guides and industry standards such' as the ASME Boiler and Pressure Vessel Code.

i The' term " extremely low" is-used in this amendment to GDC-4 with reference to' the probability of fluid system pipe rupture. For reactor coolant loop piping, a representative value which would qualify as " extremely low" would be of the I

crder of IE-6 per reactor year when all rupture locations are considered in the fluid. system piping or portions thereof.

For other piping, representative 4

values will be developed consistent with this definition as the need arises.

Alternatively, a deterministic evaluation with verified design and fabrication, 6

7 w:)$

o,,

y-

~.n ; +

+

. 'l'

.[7590-01]L l

s fin addition to adequate.. inservice -inspection, can meet. the ' extremely e low.

j probability - criterion.

The ' deterministic evaluation

.is-based 'on the

~ requirement. that. structures' and components are correctly engineered to meet the '

1. applicable regulations and NRC-endorsed industry codes.-

(This rulemak' ng will ~ introduce an. inconsistency into the design basis : by i

J excluding the ' dynamic effects of postulated pipe ruptures while still' retaining o

nonmechanistic; pipe. rupture; for emergency core cooling systems, containments, s

and environmental qualification' (see issue 4. below for additional information 1

oni potential.. relaxations with respect to environmental qualification).

The

' Comission recognizes the need te address whether and to what - extent leak.-

.before-break analysis techniques may be used to modify present. requirements re-

.lating :to other features. of facility design.

However, this is a longer term-evaluation..

For the present, the rule allows the removal plant hardware which it is believed negatively affects plant performance and safety, while not

'l

~

affecting emergency core cooling systems, containments, and environmental qualification.

The Commission's primary justification for this rulemaking j

i rests with its statutory respons'ibility to ensure an adequate. level of protection of the public health and safety.

This action also rests upon

'a'dvances in the state of knowledge and technology that allow the Comission to

'better focus 'its regulatory requirements so as to improve safety of plant k

The Commission decided to cuantify the degree to which overall personnel.

safety was'_ improved by this action, and to set forth those economic impacts as'sociated with these safety benefits. These are highlighted below.

7

_1

' ~ -

t-1 p.y.;

, e q

g,'

p.

w~

(

q-ts.

<[7590-01.1 U

L[

R For existing-PWRs, consideringip'rimary ' coolant. loops only, cos.t savings'.of }186 Jmillion :and reductions"of 34,000 man-rem are _ estimated for a: population.of 85' PWRs.

These figures. do not include savings L resulting from l redesigniof heavy.

component supports.

One licensee taking advantage.of ?the limited' scope; modi.

J n

fication> of GDC-4 has. estimatedia per plant costLsavings of $20 'million and' ireduced worker exposures. of 'about 2000 man-rem. associated with a redesign of i'

. reactor. coolant pump. supports.

The' above-mentioned ivalue-impacts were realized,under the-already:. published limited scope amendment c to GDC 4.

Additional ~ benefits which.can. be achieved H

under this. broader amendment ar'e discussed below.

~

For existing BWRs, considering only recirculation loop piping, cost savings' of i

y $30:million and ' reductions of 8,600 man-rem are -estimated for a population of

38. plants'.

In existing.PWRs and BWRs, 'offsite' risk is estimated 'to be i significantly ~im-i pacted, or. if. credit is taken for improved inservice inspection and enhanced safety ~, to be reduced by an unquantified amount.

The Comission has not quantified situations in existing plants other than those ' discussed above; however, it is believed that other high energy piping will also indicate favorable value-impacts.

Value-impacts resulting from this rule are greatest for future plants, where estimated costs can be reduced approximately $100 million per unit.

Of this 8

i n

n y

-. s 9'.

[7590-01):

y sum, about - $30 million are direct costs and the balance stems from. reduced financing costs ~and improved scheduling.

Reduction in worker radiation ex-t posures -varies from plant to plant, but is in the range of 300 to 800 man-rem.-

')

Offsite risk is believed to decrease by an.unquantified amount due to improved effectiveness of inservice inspection and enhanced safety.

'The above ouoted figures are based primarily on the elimination of pipe ~ whip restraints and jet impingement barriers'and do not treat other facility-changes that coula result from this rule.

l a

ACCEPTANCE CRITEPTA The Commission developed a new Standard Review Plan Section 3.6.3 which gives mora details on how applicant and licensee submittals will be evaluated. This document has been issued for public comment (52 FR 32626) prior to being adopted by the Commission. The Comission may also develop at some future time a Regulatory Guide after experience is gained with the use of SRP 3.6.3.

INVITATION TO COMMENT I

Comment was invited on the following topics in the proposed broad scope amend-s ment to GDC-4.

1.

Value-impacts associated with this expanded modification to GDC 4, with particular reference to experience with the use of pipe whip restraints and jet impingement shields near nuclear reactor piping.

(The value-impact analysis prepared by Lawrence Livermore National Laboratory is 9

~

{

i t

s t

y,

[7590-01]

a Javailable.for inspection and copying for a fee in the NRC Public Document j

.Roomi1717LHStreetNW, Washington,D.C.)

+l 2n f.The scope..of. piping ! which could or< shoule be 'affected, supported b'y technical. justification.

)

x 3.

'Th decision'.to limit impacts 'of this modification'n 'of, GDC-4 to only dynam--

ic effects associated with pipt. rupture'.'

1 i 4.' -

The acceptance criteria which the Comission proposes to use to evaluate' whether leak-before-break technology is applicable to specific situations...

.,l 5.

L Acceptable-allowables for pipe-connected component ' supports which would

. provide. adeouate assurance that con'ponent support failure would not be a source 'of the pipe rupture loads being eliminated from the, design basis.

j 6.

The imposition of a temperature limitation as a way of avoiding concerns with creep-damage.

ISSUES ANALYSIS Issue 1.

Margins for leak detection should not be rigidly fixed but

'should be based on uncertainties for each particular situation.

g i

Commission Respopse:

J I;

}

10 L_=_ _ _ _ _

p f

g 3

[7, i

s.

.,s

?

[7590-01]'

. W j;

.1

'. k' l 4

i

l Commissi6n's proposedL acceptance; criteria,. the. postulated' j

In /the r

thoughwall;cracklused in 1;he deterministicifracture ' mechanics evaluation.~is y2 7

yHP Sbased onfajdetection '. margin Lof ten with respect. to ithe leakage from the 4

postblated crack.

'The Commission agrees that ' the selection -'of. the ' margin 7 j k,

-should:beidehived from the uncertainties : involved.. LAs? noted in:issueL7 of the-l

.i g

h, >

L finalf.limi ted ' scbpe LGDC 4 2 ruleL (51 FR 12502, April:11',1986), the Comission -

]

Lrecognizes that-the measurement'or determination of leakage'from i system under.

L J

' pressure; involves uncertain, ties for which margins are needed.

Comenters L 0

' suggesting relaxation Ein.. the detection margin cited only. limited ' sources,of Juncertaintyi such as: material properties and calculated flow rates through. a.

y q

- crack. n0ther sodrces. of uncertainty not mentioned include plugging of the.-

.crackswith--particulate material over time,- stresses and number of cycles, and i

uncertainties associated with personnel and instruments used to detect' leakage, For:the present, the -Commission will retain the. leak detec' tion margin of ten E

'u'nless. detailed evidence can be presented for other values. The Commission may

{

require administrative controls.to enforce adequate implementation of leakage o

detection and monitoring. Additionally, the Commission may undertake recurring j

inspections to verify that leakage detection. and monitoring satisfy

' leak-before-break requirements.

1 1

Issue 2.

Margins on loads and leakage crack sizes used in the determinis-tic fractur'e mechanics evaluation should be relaxed.

l Commission Response:

The Comission acknowledges that there are many situations where the margin is not required on loads resulting from the design basis piping I

l 11 L-

1

)

l

[7590-01)

.!,l analyses.

However, thare are situations where the uncertainty in the total procedure, including stress analyses and fracture mechanics evaluations, warrants some margin (see issue 7 below).

Applicants or licensees must maintain the. margin on loads at 1.4, except when the deadweight, thermal expansion, pressure, seismic inertial and seismic anchor motion-loads are combined based on individual tbsolute values.

In this case, the margin on loads may be reduced to 1.0.

The evaluatior of seismic anchor motion loads at SSE conditions may be omitted when these loads are -shown to be small at OBE I

conditions.

The Commission believes that, because of uncertainties associated with, flaw geometry and the different analytical procedures, the margin between the i

leakage crack size and critical crack size stated in the proposed rule should not be reduced below the value of two.

Issue 3.

The acceptance criteria should refer to " crack detection" rather than " leakage detection".

Commission Response:

The fracture mechanics evaluation outlined in the Commission's acceptance criteria examines a postulated throughwall flaw which may grow under service I

and earthquake loads.

The size of the postulated flaw for fracture mechanics evaluation purposes depends on the ability to detect the presence of the flaw during service with an adequate margin for detection.

The standard methods to q

detect the throughwall flaw during service depend on the magnitude of flow or 12

~

+

. m,,.

p1

p 3_

.[7590-01]

l

leakage through 'the flaw.

Therefore, the methodology has to be ' based upon leakage' detection rather.than crack detection.

i I

Issue 4.

Leak-before-break.' technology should be extended to relax. pipe I

l rupture requirements'for.c'ontainment design, emergency core cooling systems and:

.i q

environmental qualification of electrical and mechanical. equipment.1 j

j l

Comission Response:

This was addressed as ' issue 3 in. the ' final limiten scope.GDC-4 : rule (51 FR 12502).

The 'Comission plans to. consider whether environmental quali-fication requirements. can be. modified based upon leak-before-breakLtechnology'..

The Comission does not intend to consider near-term changes to emergency core-i cooling system and containment design bases as discussed in the FINAL RULE q

section of this Supplementary Information.

When leak-before-break technology is applied to dynamih effects design bases, these effects are reduced to zero; there are no replacement dynamic ef-i

.fects postulated.

However, environmental qualification design bases cannot be reduced to zero when leak-before-break technology is applied to piping.

The postulated pipe rupture has served as a convenient and conservative umbreita

[

covering many sources of environmental qualification design bases, such as breaches in the fluid system pressure boundary from failed pump seals, leaking

{

valve packings, flanged connections, bellows, manways, rupture disks and i

throughwall cracks.

Thus, in applying leak-before-break technology to en-vironmental' qualification, the Comission faces the task of developing a replacement environmental qualification design basis.

13

9%/,aQ V W,;}y[n,

.L

-;-).

/

m,.

[7590-01].

c The Comission ~1s not ' prep'ared at this time to propose: new environmental y

p ;g Edesign criteria for temperature, pressure, humidity'and eflooding..If it' can.be?

g a

L shownf thatOittisi-beneficial to apply: leak-before-break ~ technologyito 2 eni te ivironmental qualification, another modification'to GDC-4 would'bel proposed... In the interim,. the Comission ' recoanizes that ' situations 1 may arise' where M

f justif'icationi can :be. developed byl the: industryL for. alternative - equipment

+

qualific' tion' requirements. ' Such justifications, if accepted by the' Comission.

a

. pursuantf to: the': existing exemption' process,. would allow 'a limited number.o'f -

case-by-case' relaxations. iin ; environmental qualification: requirements.

.The-s.

3 Commission encourages the ! development c 'of generic ' alternative; equipment-D

&g lq'alification. design bases by the industry.

This could support' future; u

amendments: to : GDC-4 and other affected requirements addressing environmental.

~

Lq' qualification.

Issue 5.

Can minor modifications of piping systems not related to the exclusion of dynamic effects be made without examining impacts on the_ original

. leak-before-break evaluation?

Comission Response:

g The' original leak-before-break evaluation must be applicable for the life b-

.of.the plant.

Changes in configuration. or operating conditions must be examined toLdetermine impacts on the validity of the original bak-before-break Devaluation,- particularly as to how stresses are influenced.

The Commission believes'that many minor modifications, such as changing piping insulation, can be made without..affecting the original analysis.

owever, modifications which

.are not' minor in scope, 'as for example when the number or type of pipe supports A

14

>fp Q

he-g lif, j

p

+

g.

'A

'h

'V{

'.}

,h :

j

}

w n

'L,.

l

[7590-011 j>

/

3

,.y,

y

~

((~

'4F

^

are. changed extensively,. require an evaluatipp of,thejapplicability of the "y

j' uriginalileak-befor'e-break analysis.

l

,s

,;;p Issue 6'..l.eak-befge-break should bel mandatory for plants' which have not; 4

il

. yetirecei ved 'thei r cons truhtion L permi t.

i

../J

+

Comission Response:

{

1

'The' Comission believes that economic'and operational considerations will d

motivate. manyjutilities to.' apply leak-before-break technology.

'While. it fis f

/

.r; estimated tpat?.n unquantified reduction inlpublic risk results from this rule,.

?

the actuals scope of piping contrib ting to the! reduction.wille vary from plantl 3

1 x

n.

to plantL The Comission encourages theLuse'of,high onlity piping which doe's-C Jo,t'~ reh71re pipe whip restraints and jet impingement barriirs. For any new ap-j

"}

v plication, the Comission would permit the applicant' to decide 'whether or not,'

y, to,asFl leak-before-break technology.

4 i

s.

f l.

I/

V 'j. Issue 7.., Leak-beforefbreak technology should be applicable to discrete

/

Ll6 cati ns. f'Tjiere shodd b$ nq r'e'quiremenbthat' leak-before-break technology be l

'j t

J

.p appl 1patile only to In entire piping' ' system or analyzable portion 'thereof /

(~

I

~

.' V ry a fl Comissien itecponsja:-

4, Star,dard,Dview Plen (SRP) Sectior/ 3.642 of NUREG-0800 has been used for

[

~

< ' /..

./

q more than e, decade t)o postulate the num'oer ~jand location of pipe ruptures in

)

nuclear power plants.

SPP 3.6.? ignores or'tieqts indirectly many facpors,,

i

.such as material properties, potential corrosion, and the potential for water i

haaimer,,wFlch actually determine where and wnether pipe rupture, will occur.

15 s

e

xp y

~'

.?

t *: y j

1 1

[7590-011 x

Leak'-before-break procedures explicitly treat these factors.

The Commission i

will not commingle SRP 3.6.2 with.more advanced. leak-before-break ' methodology.

Leak-before-break' is intended 'to be a substitute for SRP 3.6.2 only when all I

i breaks in:a. fluid' system piping are eliminated.

This avoids consideration of 1

synergistic effects, that is, th.e effects of a pipe break at one location on another potential break location.

1 I

Additionally, the Commission, through long term and extensive piping re-search programs, has become aware that differences exist between analytically j!

' calculated Stresses and actual stresses occurring at discrete locations in piping.. The differences between calculated and actual stresses ' usually stem.

i from difficulties ' in: modeling pipe supports tmder dynamic and static i

environments. ' Leak-before-break will'be applied only to an entire fluid system j

1 piping or analyzable portion thereof.

Issue 8.

Creep-should not be an issue in applying leak-before-break i

technology.

Commission Response; I

This rule gives guidance for reactors other than light water reactors.

Creep can be an issue for gas and metal cooled reactors.

Normally, creep damage is not an important cor.cern in light water reactors.

i l

Issue 9.

Extensive materials testing requirements should be relaxed. The use of generic materials procerties should be permitted.

16 WL

rys

-T w

~

n.

f~.t d

f cpp

,y/ Y 1

r

.g

>[

^

") '-

-~i'i

[(

y

[7590-011-d R

j hz.,

?

l l

1

' Commission Response:-

u 1

LThe ductile piping fract'ure mechanics analysis t'echriiques 'that are app 1'ied.

q inLthe. leak-before-break 1 assessment are strongly dependent.on the material ten-Li 1

'sile properties an'd material resiitance to crack! extension. The materia 14 test.

ing. requirements are.necessaryl to provide. reliable ia'ssessments. of margins M

'against ' unstable flaw extension. when case-by-case leak-before-break analyses q

l

=are performed 1

However,. if : archival ; materials are not available or; if E actual ' plant-1 m

a material. properties cannot be. defined practically, generic plant specific or.

l industry. wide, materia.1 data. bases can be assembled 'and used x to. define ' the required material' tensile and toughness properties..- To provide an. acceptable level of reliability..' plant specific generic data bases must be reasonable -

]

lower bounds :for-sets of; comp' tible material tensile and toughness properties.

j a

associated'with' actual' materials at the' plant.. Any industry generic data base l

y must be a reasonable ' lower bound for the population of material tensile and toughness. properties associated with any individual material specification (e.g., A106 - Grade B), material type (e.g.,

austenitic. steel), or welding i

i procedure.

Except as indicated in the Commission response to issue 13, f

industry generic data bases for the range of piping materials in light water i

. reactors have not been assembled and proposed for leak-before-break analyses.

~

Industry groups are encouraged to assemble and use reliable generic data bases

)

so that analyses and evaluations can be performed efficiently and effectively.

l l

l l

17

"x

4-i

's

[7590-01)-

-l I

? Issue 10.-

.The temperature limitation of 750 F should.not be adopted for I

evaluation of creep' damage.

"Comission Response:

0 The temperature limitation of 750 F is revised.as follows:

for-ferritic steel piping, the temperature limitation will be 700 F; for austenitic steelc piping'the temperature limi.tation will be 800 F. -These values more accurately reflect the creep performance of piping and are.in accord with the ASME Code.

0 Recent experience in-fossil fuel plants operating at temperatures over 1000 F 1

has ' indicated that creep-related ruptures in large diameter piping may not be low - probability'. events,_ and suggests deficiencies in creep design standards.

after a. service life of ten years or greater.

Until creep behavior after long service intervals is: better understood, the Commission will ' retain the 1

\\

temperature limitations cited above.

1 Issue'11.

Delete the words " reviewed and approved by the Commission"

]

from the text of the rule.

- Comission Response:

This coment is rejected. Leak-before-break technology is applicable only i

to high quality piping which is maintained in a high cuality condition. Since much of the plant's piping is custom designed, the Commission would have to undertake detailed case-specific review to determine that acceptable standards of quality are achieved and maintained, and that the analyses meet the Commission's requirements.

Cetailed reviews are especially needed in piping other than PWR primary coolant loops to assure that failure mechanisms such as i

18

[

[7590-01) 1

)

water hammer, corrosion, erosion, fatigue, and creep are not significant contributors to the potential for pipe rupture.

Additionally, factors such as leakage detection, material properties and environmental conditions are more variable ontside 'PWR primary coolant

loops, and possible misuse of leak-before-break ~ technology can occur, unless careful review and evaluation of these aspects are performed by the Commission.

Consequently, the words

" reviewed and approved by the Commission" were added specifically to ensure that a careful evaluation enforcir.g the Commission's rigorous acceptance criteria would be performed for each individual request from licensees and applicants.

l 1

i

.The adopted revision of GDC-4 requires NRC review and approval of the analyses on which the elimination of dynamic effects are based.

As reflected j

in the limited scope rule (51 FR 12502), which is replaced by the adopted broad scope rule, the NRC has previously reviewed and approved the application of leak-before-break technology for eliminating design basis dynamic effects of postulated ruptures in PWR primary loop piping.

No additional review and approval by the Commission in these cases is required under the adopted broad scope rule for elimination of design basis dynamic effects of postulated ruptures in PWR primary loop piping provided the conditions set forth in the Supplementary Information accompanying the rule (51 FR 12502) are satisfied.

The proposed broad scope amendment ($1 FR 26397, July 23, 1986) also stated that " Modifications of the licensed plant design of operating plants may involve an unreviewed safety question under 10 CFR 50.59.

A simple removal of pipe whip restraints and jet impingement barriers would not involve L_

i

[7590-01]

an unreviewed safety question."

The meaning of this last sentence is that after_ analyses reviewed and approved by_ the Commission demonstrate that the probability of. fluid system piping rupture is extremely low, then, without prior approval, pipe whip restraints and jet impingement barriers may be removed.

Pipe whip restraints and jet impingement barriers cannot be removed, however, without condecting an appropriate leak-before-break evaluation, submitting the evaluation for Commission review and obtaining Commission approval.

Moreover, removal of a pipe whip restraint which also serves as a seismic restraint would not be a " simple" removal of a pipe whip restraint and, therefore, would involve an unreviewed safety question.

Issue 12.

How is the demonstration of extremely low probability made for indirect sources of pipe rupture?

Commission Response:

Indirect sources of pipe rupture, as discussed in the plant FSAR, are in-i vestigated by applicants and utilities.

These include seismic events and sys-tem overpressurizations due to accidents resulting from human error, fires or flooding which cause electrical and mechanical control systems to malfunction.

The analysis of indirect sources should also confirm that snubber failure rates l

l are maintained at a low rate.

Compliance with the snubber surveillance requirements of the technical specifications can be used to demonstrate that snubber failure rates are low.

Missiles from equipment, damage from moving equipment and failures of systems or components in close proximity to the piping are investigated as well.

The results of prior analyses conducted to 20

ym s,

m.g,

4F' t

q r

'7 4I, q

e

.[7590-01].

~'

+

.3 L'

^

,yc 'd 1

J how-complia'nce E with Comission. regulations canl be applicable L to potential-E s

4

sources.of indirect pipe rupture.

qw Issu'e 13.

'It.is' recommended that adequate. material toughness bei demon-d strated when limit load analysis is applied and that the margin of three be on the'~ applied force and moment combined rather than just on moment.

The limit loadlanalysis1 procedures in ASME Code,Section XI,' Appendix C, Winter 1986

.c

~ Addenda,-should be allowed.

l t

Comission Response:-

d The. Commission' is revising its requirements on. limit load analysis pro.

j cedures as stated 'in the proposed amendment to GDC-4 (51' FR 26393).

The new

. requirements do not contain the arbitrary margin of three on applied moment, but instead 'ar'e-based on an ' experimentally verified ASME approved procedure, w.

1 q

During preparation of NUREG-1061, Volume 3,. there was significant un-i

~

certainty. associated with reliable application of limit load analysis for i

austenitic steel, especially in the case of submerged arc welds (SAW) and i

shielded metal arc welds (SMAW).

This uncertainty led to restrictions on the use of limit load analysis and application of methods originally used as the basis for IWB-3640 in Section XI of the ASME Code.

o l

'Recently the ASME' Code approved revised evaluation procedures for aus-ttenitic steel piping (see Appendix C of Section XI); these procedures incorporate methods to account for reduced toughness associated with SAW and

-SMAW. The Commission has conclu6sd that the evaluation method in Appendix C of 21 j

a e

l c

~[7590-01]-

q Section XI (including the tensile and toughness properties. defined for base metal and-welds) is acceptable when performing -leak-before-break analyses for.

1 austenitic steel piping provided the margins. presented in the. Commission a

response _to issues 1.and 2.are met.

The value of flow stress used.with.this j

i method will.'be evaluated by the Commission.

Because generic evaluation procedures. and materials properties - have not yet been approved for ferritic.

- piping by: the ASME : Code or the NRC, leak-before-break analysis for fe'rritic piping will continue case-by-case until approved Code procedures are available.

I Issue '1'4.

Leak-before-break should not be limited'to high~ energy piping..

Comission Response:-

i The Comission's rules require postulated pipe ruptures ' only in high

energy piping.

There is no reason to apply leak-before-break technology _ in f

a moderate. energy piping because there are no postulated pipe ruptures in such

]

fluid system piping.

l l

l l

' Issue 15.

Strict adherence to Regulatory Guide 1.45 should not be re-

- quired outside the containment.

1 Commission Response, j

The Commission does not require and did not intend to suggest the need for 1

strict adherence to Regulatory Guide 1.45 outside the containment.

The pro-posed rule stated only that "... leakage detection requirements equivalent to Regulatory Guide 1.45 must be satisfied for all piping within the scope of this 22 L--_-__'nx_-_am. -.. __-._

a

y% '3..

s1 N

L'.

t

??

i,

[

-[7590-01]

M rule." : Scheduled: operator walkdowns can be used as a means~ of leak'detectioni

,outside1the~containme'nt.

.g 1

Issue.16.

Older. operating plants should not be. held oto the requirement:

.that.heavys component ' supports should meet ASME-Code.allowables as a condition for? applying l leak-before-break..

'Co'mmission' Response:

I The'.use -of ASME Code allowables for heavy component supports' of older.

operating plants.as' a condition for applyingileak-before-break? technology is L

Lnot required. : However, when-heavy component supports are. redesigned excluding b Lthe dynamic effects of : pipe rupture, current industry codes'(such as the ASME 1

~

or AISC code) may be' required.

Additionally, current NRC criteria. for calcu--

I ating seismic loads (coupled with the 'already. existing SSE) may-also be re-L l

quired..For example, a simple replacement of high strength fasteners with more ductile. fasteners of' lower yield ctrength would not require changes 'in the in-'

-dustry codes or seismic criteria from that used..in the original design. On the other hand, modification of the heavy component supports that involves. redesign and ; removal of snubbers in early vintage. plants would require use of current

. industry codes and NRC seismic criteria.

Dynamic effects from pipe ruptures in branch connections must be' considered -if the branch connections do not qualify for. leak-before-break.

In heavy component support

redesign, improved l

i functional reliability must be demonstrated for any changes made.

Structural l

' capacity associated' with the original steel and concrete, including struts.

l V

l t

columns, pedestals, hangers, trusses and skirts cannot be diminished in the E,

Q support system of operating plants or plants under construction.

Redesigns f

C

= _ _ _ _

]

23

W

'i, m

a

[7590-01]

I wil1"be' limited to replacing high strength fastener material and reducing' the-j

- number and capacity-ofisnubbers.

Applicants and licensees undertaking heavy -

f component supportL redesign, with dynamic ' effects of pipe rupture eliminated,

.j i

.s ould 'use independent design and fabrication - verification procedures to h

minimize design and construction errors.

Displacements and rotations resulting~

from potential failure of redesigned lateral (horizontal)' supports should.not I

lead 'to '.the ruptureJ of piping connected to - the reactor coolant loop heavy

~

components.

1 l

Issue 17.

Additional' guidance 'is needed on the acceptability of remedial

. stress enhancement programs such' as induction'. heating as:it pertains to stress.

]

- corrosion cracking, residual stress states and sensitization.

i Commission Response:

.The rule precludes leak-before-break evaluations for systems that have

.j

- materials that are. susceptible to intergranular _ stress' corrosion cracking-4 (IGSCC).

The Commission recognizes that remedial residual stress improvement treatments are effective in reducing susceptibility to IGSCC.

However, re-medial stress improvement treatments of nonconforming materials alone do not i

provide'a sufficient basis to support leak-before-break evaluations in the con-text of this rule.

The Comission would, however, review such evaluations case-by-case if hydrogen water chemistry were used as an adjunctive measure l

with the remedial stress improvement treatments.

Practices with regard to j

facility water chemistry would be an additional factor considered in the re-i view. Nonconforming piping with any planar flaws in excess of the standards in

'IWB 3514.3 of Section XI of the ASME Code would not be permitted to use 24

- 4 o

g[e,,

$D t

. g;

,t; e.

L _~ l '

U :;g

'6*

s

[7590-01]-~

j

l deak-before-break L ana' lyses t '

However, -nonconforming" piping: that 'has been;

.e

treated f by (twoi mitigating methods oma'y 1

qualify for" leak-before-break ifithe -

piping, contains lno flaws larger-than j those permitted ;by IWB 3514.3' of Section -

m XI U of the ' ASME 1 Code.

If piping has been~ repaired by weld ' overlays,-

leak-before-breakitechnology cannot be applied.

^

c - Issue 18'.

The. fracture. mechanics. approach 'should 'not require that the -

location of highest stress utilize;the poorest' material properties.

}

3 r

TComission Response:

1 i

The proposed. rule stated that,7 n : conducting the ' determinist c fracture.

mechanics evaluation,' investigators would " identify.the location (s) at which

~

~

8 the highest stresses coincident with peorest materials properties ' occur...".

~

This; sentence should have read " identify the location (s) which have the least

' favorable combination of high stress and. poor. material-properties...". The Comission did not-intend to combine the highest stresses-at one location.with the poorest material properties atianother location.

The critical rupture lo-cations depend on stress.and material. properties, among other things, and in-vestigators may need to examine several locations to decide which is - the -

/

! controlling case.

Issue 19.

The decision that leak-before-break technology is not appli-l 6

cable to materials subject to cleavage type. fracture should be reconsidered.

r I;

l l

'Co,nission Response:

l l

r l

25 1

w.

Ch_____m _ _ _ _. _ _ _. _. _. _ _ _ _

"{ $7 4

[7590-01]

ii This comment is rejected.

The Comission will allow leak-before-break 7

technology' only. to' materials which are ductile under the full range of system

- i joperating temperaturesLin order to avoid sudden brittle piping failures.

Issue 20.

The. clause 'in' the rule requiring Comission review and approval of analyses demon'strating piping integrity will. preclude litigation-over the. scope of_the piping affected and the. adequacy of the analyses.

This would amount ' to a de facto illegal removal of a material issue from an operating license amendment proceeding.

4 L Comission' Response:

The comentericites in support ~ of this proposition Cleveland Electric Illuminating Company -(Perry Nuclear Power Plants, Units 1 and ?), ALAB-841, (July 25,'1986), reconsideration denied ALAB-844, (Augus.t 18, 1986). The 1

Licensing Board case affirmed in ALAB-841 is Cleveland Electric Illuminating 1

{

_ Perry Nuclear Power ' Plant, Units 1 and 2), LBP-85-35, 22 NRC 514

)

(

Company (1985).

]

i A careful reading of these cases shows that they do nct stand for the com-L I

I menter's proposition.

At issue was the scope and adequacy of the applicant's preliminary hydrogen control analysis required by 10 CFR 50.4d(c)(3). One of the criteria for this analysis is that it "use accident scenarios that are ac-cepted ' by the NRC staff." [50.46(c)(vi)(B)(3)] The Licensing Board did not hold that the staff's approval of the applicant's analysis was binding and thus precluded a challenge to the scope and adequacy of the analysis.

Rather, the L

L Board permitted such a challenge on a number of issues.

22 NRC at 533-548.

26

o.

c q

[7590-01]-

'However, the Board. did. not allow Lthe. intervenor to - raise other issues under this contention'which:went: beyond the.' scope of the hydrogen control ruleTit-self. 122 NRC at 548-S49.. The Board's views on these matters were upheld 'on appeal.

h

' A edirect-application of this case to the. GDC-4 context 'shows that the commenter's' conclusion is incorrect.

The staff's acceptance of a GDC-4 anal-ysis will,not preclude litigation of either-the scope of piping included or the'

, adequacy of. the ana' lysis itself.. However, a' challenge on either basis-must be

~

confined Eto the. overall scope of GDC-4 and could not.be used as a~ collateral-challenge to otheriparts of the regulations' or to argue that the rule'itself is.

j inadequate.

Challenges.of this type must be brought. pursuant to 10.CFR 2.758.

l u

1 The staff.'s review and approval of the piping integrity analyses is an indispensable part. of' the implementation of the leak-before-break concept.

Without'such review (for piping other than the PWR primary loop), the staff has no means.to assure itself that the acceptance criteria have been properly ap-plied. The coment is therefore rejected.

i Issue 21.

'The reallocation of resources within the NRC to review piping integrity analyses submitted under the amendment is barred by the Atomic Energy Act, which requires that public safety take precedence over cost savings to licensees.

I Comission Response:

27

7 q

ff ip

[7590-01]

The Regulatory Analys'is performed ' to ' support. this : rulemaking-shows that 1

there. is a ' net ; safety benefit to be realized from proper. application of leak-before-break technology. The Commission has undertaken the rulemaking for,that

.l purpose.

The positive results in~ terms of simplicity of. the plant, ease of inspe.ction, avoidance of improper removal. and reinstallation of unneeded Lsupports and restraints, and the reduction 'of personnel exposures have been shown:to vastly outweigh any additional risk associated viith removing supports and restraints.

Therefore, reallocation of NRC' resources to' ensure that NRC acceptance criteria are rigorously adhered to' is fully justified in terms. of 1

public safety.

.i J

In addition to these issues, the Commission deleted the fatigue crack j

.I

. growth analysis specified in the proposed rule.. This requirement was found to be unnecessary.because it was bounded by the crack stability analysis.

j i

Having considered all of. the above, the Commission has determined that a '

j final rule be promulgated AVAILABILITY OF DOCUMENTS L

1.

Copies of NUREG-1061, Volume 3, may be purchased from the Superintendent

~

of Documents, U.S. Government Printing Office. P.O. Box 37082, Washington, l

D.C., 20013-7082.

Copies are also available from the National Technical i.

Information Service, 5285 Port Royal Road, Springfield, VA 22161.

A j

copy is also available for public inspection and/or copying at the NRC Public Document Room, 1717 H St., NW., Washington, DC.

I 28 l

u-_-

.y a&

w v m-n

+

/

x

1

+Y

's hi

'\\l!q_

y &( p..

h'

' ml -,

^

a (,,

o e

e

~

/

x i,

'A

~

m g

a.,

- t, i;,k

{

,9 M

t

( ' ]v.

L[7590-01]

l,

^

)w%,' s fl g:

m

- 2L,c LCo' pie's? of the ASPE Boiler and PressureIVsssel Code 'may. be obtained from.

i

- (;

~.theLAmerican Society of L Nechanical Engineefs, 345' East 47thi Street, New 7,,

y,

'_,s 9

York. NY, 10017.E ap, 4

~ 3.. --Copies ;of R.egulatoryj Guide 1.451 entitled' "Peactor', Coolant' Pressure. Bound- -

y

'ary Leakage Detection _ Systems'E may bh _obtained by writing to the Division

<(

~

q Lof Technical Information and Document Control, U.S..Nuclearc Regulatory

.o W

Commission, Washington, D.C.,.20555.,

3 e

FINDING'0F NO SIGNIFICANT ENVIRONMENTAL IMPACT: AVAILABILITY n

I TheLCommission'.has determined under.the National. Environmental Policy Act of 1969, as, amended, land lthe Commission'.s regulatio'ns Ein. Subpart A of > 10 CFR

,2 j>

s t[

Y Part 51,' that this rule.. 'if.' adopted, would n'otl be -'a major: Federal action

~

'significantly affecting the quality of'the human environment'and, therefore, an-environmental impact statement is not' required. ; Although certain plant hard-^

+

ware might be removed from the plant, ~ consistent with this ' rule, the, removal would not alter:the environmental'. impact of.the licensed activities as set out

.f in the Final Environmental Impact Statement for each facility.

The environ-

~l mental assessment and finding of no 'significant impact on which this. deter-

!l

^

mination is based are available for inspection at the NRC Public Document Room, 1

1717' H Street, NW, Washington, DC..

Single copies of the environmental assess-x significant impact are available from John A.

j 4

ment and the finding of no 0'Brien, Office of Nuclear Regulatory Research,. U.S. Nuclear Regulatory Com-I l

mission, Washington, DC 20555, telephone.(301)443-7854.

y.

y 3

.q a:

29 j

23

a.;

' i. '

l

]

[7590-01)_

l' t.

g

,,s 3

j' j '

c j

e 1

PAPERWORKREDUCTIONACThSTATEMENT

,]

This final rule does not contain a!new or amended infonnation collection' requirement subject to the Paperwork' Reduction Act of-1980 (44 U.S.C.,3501 et' 1

seq.).

Existing information collection. requirements under 10 CFR Part 50 were-m approved by the' Office of Management and-Budget ~ approval number 3150-0011.

1

{l

~

REGULATORY AND.BACXFIT ANALYSES

l The regulatory analysis.is available for inspection in the NRC'Public Document-

'j R60m,1717 H Street NW, Washington, DC.

Single copies of the analysis may be..

obtained from dohn A. O'Brien, Office of Nuclear Regulatory Research, ' U.S.

Nuclear. Regulatory Comission, Washington,'0C 20555, telephone'(301)t43-7854.

i A backfit - analysis under 10 CFR 50.109 for the purpose of completeness was published in the._ proposed-broad scope GDC-4' modification.(51 FR 26393),

l although. it was not' required because the rule will not require licensees or applicants to make any -changes.

The Comission's primary justification for this rulemaking rests on its statutory responsibility to ensure an adequate

.l level-of protection of the public health and safety.

Economic advantages or

}

disadvantages resulting from this action did not affect such responsibilities.

.The Comission remains mindful of its statutory responsibilities pursuant to j

Union of Concerned Scientists et. al. v. NRC, DCC No. 85-1757, August 4, 1987.

f The Comission has prepared, however, a regulatory analysis to set forth l

clearly the costs and benefits of the impacts of this rule and the examined 1

I alternatives.

I i

30 l

l L-

,.p:

[g y(i Th

,s,

!Y I../

' J. i ~

{

, ; j'J. s

7. -

s

-).il\\

3 -

pl';t, ; E s t 7 <,

l[

l-9 s

g.

s, c jy :Q ',

{

[}

Q u' U,, g 4.'1, 3'

,i;

?,

? REG lJLATORY FLEX 1BILITY. ACT CERTIFICATION

r. L 5, ( ?

'?

5 i

s

>l,.

i 1

As required bycthe Regulatory Flexibility Act of 1980, '(5 U.S.C. 605(b)),,

g the ; Commission certifies that this ' rule will. net ' hive a. significant economic impact on a substantial' number-of small entit'ies.

This rule affects only the 1

' licensing and operation of. nuclear power plants.- The companies that own these -

U; plants-do. not. fall within' the ' scope of. the definitions of "small entities" set '

D forth in the Regulatory Flexibility Act or. thel Small Business LSize Staddards!

set out in. regulations issued by the Shall Business Administration at 13 CFR i:

-Part 121..

i i

.L'IST OF SUBJECTS IN 10 CFR PART 50 i

1

'l a

I Antitrust, Classified ;information, Fire prevention,. Incorporation by.

1 reference, Intergovernmental relations, Nuclear power plants and reactors, j

Penalty, Radiation protection, Reactor ' siting

criteria, Reporting and j

.recordkesping requirements.

1 For - the reasons set out in the preamble and under the authority of the i

Atomic Energy Act of 1954, as amended, the Energy Reorganization Act of 1974, as amended, and 5 U.S.C. 553, the NRC is adopting the following amendments to 1

10 CFR Part 50.-

I

'PART 50 - DOMESTIC LICENSING 0F PRODUCTION AND UTILIZATION FACILITIES 1.

The authority citation for Part 50 continues to read as follows:

I l

31

.y,

x y

lQr ', L,

g

e zm jik..fE

p j

s

,~

D'

'N

[7590-01]

q

+

AUTHORfTY:Secs. 102, 103. 104, 105, 161, ( 182, 183 ', 186, 189,168; S tat.

+

%k 936,2937, 948, 953,'954, 955,.956,..as - amended,?sec. 234, 83 Stat. 1244, Las a.

Ea' mended (42.U.S.C.2132,~2133,"2134,2201,.2232,l2233,2236,2239..2282);. secs.'

j

~

]j

,W W'

l201, as ' amended, - 202, 206,.88 Stat. 1242, as amended, 1244,1246,-(42:U.S.C.

5841,5842,.5846).

3.,

y 1

Vli Section 50.7 also~ issued under Pub.'L.95-601, sec. 10,' 92 Stat. 2951.(42 l

e, M

U.S.C. 5851).

Section 50.10 'also issued under ' secs.101,185,- 68. Stat. 936,-

o a

1955', as amended (42 U.S.C. 2131, 2235); sec. 102,. Pub. L.91-190, 83 Stat. 853 j

m (42.. U.S.C., 4332).

Sections 50.23,: 50.53, 50.55, and 50.56. also'. issued ' under c

!sec. 185, 68 Stat. 955 (42 U.S.C. 1235).

Sections 50.33a,;50.55a and Appendix.

1

/

0 also ' issued under 'sec.102, Pub. L.91-190, - 83 Stat. ~ 853 -(42 U.S.C. / 4332).

a Sections 50I34-and 50.54 a'lso ' issued under sec. 204,. 88 Stat.1245 '(42 U.S.C.

5844).

Sections 50.58, 50.91, and 50.92 -also issued under Pub. L.97-415, 96

Stat. 2073 (42 U.S.C. 2239).

Section 50.78 Lalso issued -der sec. 122, 68 Stat.'939 (42'U.S.C. 2152).

Sections 50.80-50.81 also-issuco under sec.'184,

.j l

.68' Stat. 954, as amanded (42 U.S.C. 2234).

Section 50.103 also issued under j

sec. 108, 68 Stat. 939, as amended (42 U.S.C. 2138).

Appendix F also issued q

undersec.187,68 Stat.955(42U.S.C.2237).

.For the purposes of sec. 223, 68 Stat. 958, as amended (42 U.S.C. 2273);

6550.10(a), (b), and (c), 50.44, 50.46, 50.48, 50.54, and 50.80(a) are issued

.under sec. 161b, 68 Stat. 948, as amended (42 U.S.C. 2201(b)); 5650.10(b) and

;p i

(c), and 50.54 are issued under sec. 1611, 68 Stat. 949, as: amended (42 U.S.C.

2201(1)); 'and - 5s50.55(e), 50.59(b), 50.70, 50.71, 50.72, 50.73, and r0.78 are j

'issued under sec. 161o, 66 Stat. 950, as amended (42 U.S.C. 2201(o)).

I I

32 l

V

.D... w i

m y~, ;, e

-n

i; s ; n x -- TT --- 77, (c

=

8 gv

.g._

c

. w.

. a._

m fI) lr

")

i, n.

+, '

1 x$4 ' '

'.[7590-01]i f-

, ~.

12. a. In: Appendi.x. A, General LDesignf Criterion /4Ris Jrevised l to readiast L

followsr b

.).

[

1 k'.

,.' APPENDIX Al ' GENERAL DESIGN CRITERIA-FOR NUCLEAR POWER. PLANTS 3

y 1-L. ;

3]

1 j. ;;'

L CRITERIA-

'm 4

Y i

l I.

Overallt Requirements; 1

L

'j-i '

.. Criterion 4'- Environmental and' dynamic. effects' design bases.

' Structures, systems,.and-components important to safety 'shalll beidesigned. to.

! :j 1

accommodate the effects of and to. be-compatible with the environmental condi--

tions associated with normal operation, maintenance. testing, and' postulated-3 accidents, including loss-of-coolant. accidents. These structures, systems, and components.shall be appropriately protected against dynamic effects, including l q

.the effects of missiles, pipe' whipping, and discharging fluids, that may result' l

I from equipment failures and from events and conditions outside the nuclear j

power-unit.

However,. dynamic effects ' associated with postulated pipe ruptures in nuclear power units may be excluded from the design basis

-l ll when analyses reviewed and approved by the Commission demonstrate 3

l i

~l H

i

['

l i

i y,

33 q

$\\Wy$:.',

,n

-.c

.l

..s j _,

a r

w.

,,m e

a

)l.,

Y

[7590-01.1' Li q

9.-

._L i-i m

_.y:.

a

~

i i:

a: ;' that! the probability 'off fluid' system piping,' _ rupture. is - extrernely.- low - <

d

-,j

1 L --

f under conditions ' ' consistent with the ' design basis for. the. piping; p

c

,s t-i I'-

..i'.

1 Dated at Washington,'O.C. this.

22nd day'of-october

., 1987,

)

1 i

l

)

' For-the Nuclear Regulatory Comission o

s-i d

y d

r

('

. Samuel.J. Chilk, Secretary of the Comission, q

r l

1

'l i

i 1.

a

/

34 a

.:._-__--_________..