ML19339A872

Requests Analyses & Relevant Info to Determine Effects of Postulated Steam Line Rupture & Necessary Mods to Facility. General Info for Consideration of Outside Containment Piping Sys Break Effects Encl

ML19339A872
Person / Time
Site:	Yankee Rowe, Humboldt Bay
Issue date:	12/18/1972
From:	Anthony Giambusso US ATOMIC ENERGY COMMISSION (AEC)
To:	Vandenburgh D YANKEE ATOMIC ELECTRIC CO.
References
NUDOCS 8011050622
Download: ML19339A872 (12)

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18 DEC Docket Ib. 50-29 Yankee Ato:::ic Electric Ccc:pany A7fd: Donald E. Vandenburgh Vice President 20 Curnpike Foad i

Westboro, Fassachusetts 01581 1

1 Gentlecen:

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'Ibe Regulatory staff's continuing review of reactor power plcnt ' safety indicates that the consequences of postulated pipe failures cutside of

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the contaiment structure, includirs the rupture of a r.ain steam or i

feedhater line, need to be adequately doeuraented and analyzed by licensees and applicants, and evaluated L7 the staff as soon as possible.

Criterion No. 4 of the Ccxmission's General Design Criteria, listed in Jgendix A of 10 CFR Part 50 requires that:

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" Structures, systecs, and cceponents icportant to safety shall l

be desigled to acecr:nodate the effects of and to be ccupatible with the envircrrental conditions associated with norml operation, r.nintenance, testing and postulated accidents, in-(

cluding loss-of-coolant accidents. 'Ihese structures, systers, ard ccuponents shall be appropriately protected against cyrrnic effects, including the effects of I: Ass 11es, pipe Lt11ppirc, ard dischergirc fluids, that tay result from equirrent failures and frca events and corditions outside the nuclear power unit."

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'Ihus, a nuclear plant should be designed so that the reactor can be shut-i dovm and mintained in a safe shutdcwn cordition in the event of a postulated 1

rupture, outside contairrent, of a pipe containing a high energy fluid, 4

including the double ended rupture of the largest pipe in the min steam and feedwater systens. Plant structures, syste:as, and ccaponents inpartant j

to safety should be designed and located in the facility to accomodate t!m j

effects of such a postulated pipe failure to the extent necessary to assure i

that a safe shutdown condition of tle reactor can be accorrplished ard mintained.

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Docket File '

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PDR TJCarter hj 1

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Local PDR AGiambusso @1g 5

RP Reading RBoyd (16)hATeets i

Yankee Atomic Electric Ccepany L Reading ACRS Branch Reading R0 (3)

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A JRBuchanan, ORNL OGC RJSchemel Based on the infomation we presently have available to us on the Yankee p

Nuclear Power Station, we understand that steam and feedwater lines out-f side the vapor container are outside buildings that house vital equip-g ment but that the steam lines run close to the wall of the control room.

Ig' From this it appears that although the control room wall is of heavy f

construction it may be susceptible to damage from the dynamic effec'cs of a postulated steam line rupture and that scoe modification of the

' bj facilit/ may be necessary.

t We request that you provide us with analyses and other relevant infor-matio:4 needed to determine the consequences of such an event, using the guidance pIovided in the enclosed Eeneral information request. 'lhe enclosure f

mpresents our basic infomation require: rents for plants now being con-structed or operating. You should detemine the applicability, for the y

Yankee Nuclear Pcwer Station, of the itern listed in the enclosure.

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' W If the results of your analyses indicate that changes in the design of j t shutdown in the event this postulated accident situation should occur, l.k[

structures, systems, or ccrponents are necessary to assum safe reactor

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please provide infomation on your plans to revice the design of your w$

facility to acccroodate the postulated failures described above. Aqy design rodifications proposed should include appropriate consideration of IE$

the guidelines and requests for infomation in the enclosure.

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py 4 e will also need, as soon as possible, estimates of the schedule for design, fabrication, and instn11ntion of any modifications found to be 3

necessary. Please infom us within 7 days after receipt of this letter 4P when we may expect to receive an amendment with your analysis of this

. f postulated accident situation for the Yankee Nuclear Power Station, a description of any proposed modifications, and the schedule estimates C

described above. Sixty copies of the amentent should be provided.

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A copy of the Comission's press announcement on this matter is also y.%q enclosed for your infomation.

Sincerely,

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Orighalsigneg y Roger S. Bcyd f

t A. Gia=busso, Deputy Director for Reactor Projects Directorate of Licensirg; h3 W

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12/15/72 l-Form AEC-318 (Rev. 9-53) AECM 0240 e u 5 GOVERwted MMG o**cr 1972-466-983

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g aw (A,y General Information Required for Consideration g

of the Ef fects of a Piping System Break Outside Containment 2%

Y The following is a general list of information required for AEC review g

of the ef fects of a piping system break outside containment, including W

tc the double ended rupture of the larges t pipe in the main steam and feed-i"(

water systems, and for AEC review of any proposed design changes that may be f ound necess ary.

Since piping layouts are substantially i+g.

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di f fe rent from plant to plant, applicants and licensees should determine on an individual plant basis the applicability of each of the following items for inclusion in their submittals.

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~he systems (or portions of sys tems) for which protection against pipe g

whip is required should be identified.

Protection from pipe whip need W

g-y not he provided if any of the following conditions will exist:

3$i (a) lloth of the following pipin, sy ;om conditions are met:

(1) the service temperature is less than 200* F; and k

&4 (2) the design pressure is 275 psig or less; or (h) The piping is physically separated (or isolated) from s tructures,

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v.y sys tems, or components important to safety by protective barriers, k

is-or res trained f rom whipping by plant design features, such as concrete encasement; or XQ; (c)

Following a single break, the unrestrained pipe movement of either j@F Fen end of the rupt'ared pipe in any possible direction about a plastic p

g;n hinge formed at the nearest p!pe whip restraint cannot impact any NQ structure, system, or component imp or t an t.

to safety; or Q

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(d) The internal energy level associated with the whipping pipe

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vt can be demonstrated to be insufficient to impair the safety 3

lt function of any structure, system, or cotspos. nt to an

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t L ;r unacceptable level.

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2.

The criteria used to determine the design basis piping break locations

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(a) ASME Section III Code Clags I piping breaks should '.,e Iy; f

postulated to occur at the following locations in each piping run or branch run:

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,3 (1) the terminal ends; y

(2) any intermediate locations between terminal ends where t-4 the primary plus secondary stress intensities S (circum-D m

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'+1 ferential or longitudinal) derived on an elastically

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The internal fluid energy level associated with the pipe break reaction k

may take into account any line restrictions (e.g., flow limiter) between the pressure source and break location, and the effects of either single-g ended or double-ended flow conditions, as applicable. The energy level d

in a whipping pipe may be consicered as insufficient to rupture an impacted pg pipe of equal or greater nominal pipe size and equal or heavier wall

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,m Piping is a pressure retaining co=ponent consisting of straight or curved W

pipe and pipe fittings (e.g., elbows, tees, and reducers).

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3 Lk A piping run interconnects components such as pressure vessels, pumps, and g]

rigidly fixed valves that may act to restrain pipe movement beyond that g )l required for design thermal displacement. A branch run differs from a b

piping run only in that it originates at a piping intersection, as a branch of the main pipe run.

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calculated basis under the loadings associated with one -

half safe shutdown earthquake and operational plant ri 3

conditions exceeds 2.0 S for ferritic, steel, and 2.4 S for austenitic steel; 7

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(3) any intermediate locations between terminal ends where i

the cumulative usage factor (U)6 derived from the piping

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W fatigue analysis and based on all normal, upset, and E

4 testing plant conditions exceeds 0.1; and j

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(4) at intermediate locations in addition to those determined is 3

by (1) and (2) above, selected on a reasonable basis as Gy f.;

necessary to provide protection. As a minimum, there should be two intermediate locations for each piping run

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or branch run.

(b) ASME Section III Code Class 2 and 3 piping breaks should be postulated to occur at the following locations in each piping

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'..;e run or branch run:

(1) the terminal ends; Operational plant conditions include nor=al reactor cperation, upset

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4 conditions (e.g., anticipated operational occurrences) and testing

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conditions.

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5S is the design stress intensity as specified in Section III of the l'

,x ASME Boiler and Pressure Vessel Code, " Nuclear Plant Components."

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6 O is the cumulative usage factor as specified in Section III of the ASME Boiler and Pressure Vessel Code, " Nuclear Power Plant Components."

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(2) any intermediate locations between terminal ends where SI either the circu=ferential or longitudinal stresses derived

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on an elastica 11y calculated basis under the loadings

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i associated with seismic events and operational plant

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7 conditions exceed 0.9 (Sh+ A rt e expansion stresses j,

exceed 0.8 S ; and g

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intermediate locations in addition to these determined by e

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(2) above, selected on reasonable basis as necessary to k

provide protection. As a minicum, there should be two h'

intermediate locations for each piping run or branch run.

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The criteria used to deter =ine the pipe break orientation at the break k

locations as specified under 2 above should be equivalent to the d;

?t following:

8 (a) Longitudinal breaks in piping runs and branch runs, 4 inches

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no inal pipe size and larger, and/or 3g, 7

Sh is the stress calculated by the rules of NC-3600 and ND-3600 for h

Class 2 and 3 co=penents, respectively, of the ASME Code Section III m

Winter 1972 Addenda.

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ia the allo.rable etress range for expansion stress calculated by the k

rules of NC-3603 of the ASME Code,Section III, or the USA Standard Code hJ for Pressure Piping, ANSI B31.1.0-1967.

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Longitudinal breaks are parallel to the pipe axis and oriented at any 7

point around the pipe circu=ference. The break area is equal to the effective cross-sectional flow area upstream of the break location.

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ents in the direction nor=al to the pipe axis.

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E h (b) Circu=ferential breaks in piping runs and branch runs exceeding h.

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1 inch nominal pipe size.

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A su==ary should be provided of the dynamic analyses applicable to the 4.

! k design of Category I piping and associated supports which determine 7i the resulting loadings as a result of a postulated pipe break including:

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n (a) The locations and nu=ber of design basis breaks on which the t

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'T dynamic analyses are based.

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(b) The postulated rupture orientation, such as a circumferential

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break location.

Xf ic (c) A description of the forcing functions used for the pipe whip we ptN dyna =ic analyses including the direction, rise time, magnitude, duratica and initial conditions that adequately represent the j$

16 stream dynamics and the system pressure difference.

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rv (d) Diagrams of mathematical models used for the dynamic analysis.

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\\p (e) A su= mary of the analyses which demonstrates that IM motien of ruptured lines will not damage to an unacceptable

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degree, structure, systems, or components important to aafety,

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such as the control room.

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tg.y Circu=ferential breaks are perpendicular to the pipe axis, and the break

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to the internal cross-sectional area of the ruptured jy area is equivalent Dyna =ic forces resulting from such breaks are assumed to separate pipe.

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the piping axially, and cause whipping in any direction normal to the j) pipe axis.

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A descriptien should be provided of the measures, as applicable, to I,g k^?

protect against pipe whip, blowdown jet and reactive forces including:

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hN (a) Pipe restraint design to prevent pipe whip impact; y

(I (b) Protective provisions for structures, systems, and components

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required for safety against pipe whip and blowdown jet and 3

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reactive forces; iC I

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Separation of redundant features; (d) Provisions to separate physically piping and other cocponents of redundant features; and

-q (e) A description of the typical pipe whip restraints and a su= nary

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of number and location of all restraints in each system.

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6.

The procedures that will be used to evaluate the structural adequacy fQ dh of Category I structures and to design new seismic Category I structures f.+x; should be provided including:

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$5 (a) The =ethod of evaluating stresses, e.g.,

the working stress

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=ethod and/or the ultimate strength method that will be used; jy

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(b) The allowable design stresses and/or strains; and M

(c) The load factors and the load co=binations.

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The design loads, including the pressure and temperature transients, e

the dead, live and equipment loads; and the pipe and equipment static,

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ther=al, and dynamic reactions should be provided.

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Seismic Category I structural elements such as floors, interior j

X ve118, exterior walls, building penetrations and the buildings d

Pa as a whole should be analyzed for eventual reversal.of loads due g

1 to the postulated accident.

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If new openings are to be provided in existing structures, the j

T capabilities of the modified structures to carry the design loads 7fW 4

should be de=onstrated.

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Verification that failure of any structure, including nonseismic e

Category I structures, caused by the accident, will not cause

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failure of any other structure in a manner to adversely affect:

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(a) Mitigation of the consequences of the accidents; and (b)

Capability to br.ng the :; nit (s) to a cold shutdown condition.

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Verification that rupture of a pipe carrying high energy fluid will not directly or indirectly result in:

.g (a) Loss of redundancy in any portion of the protection system (as defined in IEEE-279), Class IE electric system (as defined 2

in IEEE-308), engineered safety feature equipment, cable pene-3 trations, or their interconnecting cables required to mitigate 4

the consequences of the steam line break accident and place the F[?

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24 reactor (s) in a cold shutdown condition; or

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Y 3 hg i&9 kb (b) Loss of the ability to cope with accidents due to ruptures gy

Ka of pipes other than a steam line, auch as the rupture of pipes g,

9 causing a steam or water leak too small to cause a reactor

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accident but large enough to cause electri'.al failure.

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Assurance should be provided that the control room vill be habitable and its equipment functional after a steam line or feedwater line gg break or that the capability for shutdown and cooldown of the unit (s)

(N-4I will be available in another habitable area.

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Environmental qualification should be demonstrated by test for that w

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electrical equipment required to function in the steam-air environ-

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JkT cent resulting from a steam line or feedwater line break. The in-gp(

lE formation required for our review should include the following:

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(a)

Identification of all electrical equipment necessary to meet m5

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Cs M4 requirements of 11 above. The time after the accident in which jgg

k they are recaired to operate should be given.

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The rest conditions and the results of test data showing that is the systems will perform their intended function in the environ-h

5 cer.t resulting from the postulated accident and time interval of

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the accident. Environmental conditions used for the tests should f${ '

be sele 2ted from a conservative evaluation of accident conditions.

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(c) The results of a study of steam systems identifying locations where barriers will be required to prevent steam jet impingment from dis-s5r my

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abling a protection system. The design criteria for the barriers

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4 should be stated and the capability of the equipment to survive a-

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f within the protected environ =ent should be described.

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An evaluation of t'* capability for safety related electrical

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the control room to function.1,n the environnent h

I that may exist following a pipe break accident should be i

A rovided.

Environmental conditions used for the evaluation T

j should be selected from conser/ative calculations of accident 4

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conditions.

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An evaluation to assure that the onsite power distribution

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VF system and onsite sources (diesels and batteries) will remain i.[-

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operable th roughout the event.

14.

Design diagrams and drawings of the steam and feedwater lines N

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including branch lines -howing the routing from containment to the

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74 turbine building should be provided.

The drawings should show Ty 7

elevations and include the location relative to the piping runs of I

r safety related equipment including ventilation equipment, intakes,

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and ducts.

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A discussion should be provided of the potential for flooding of safety 1

related equipment in the event of failure of a feedwater line or any j

J other line carrying high energy fluid.

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A description should be provided of the quality control and inspection

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v p rograms that will be required or have been utilized for piping systems

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$l" outside containment.

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If leak detection equipment is to be used in the proposed modifications,

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rg a discussion of ito capabilities should be provided.

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A sum =ary should be provided of the e.:.ergency procedures that would 2 >:

be followed after a pipe break accident, including the automatic

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and manual operations required to place the reactor unit (s) in a Si cold shutdown condition.

Th-estimated times following the accident I?)

Gk for all equip =ent and personnel operational actions shculd be included dd gj ID in the procedure summary.

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A description should be provided of the seismic and quality classi-

g ihr fication of the high energy fluid piping systems including the steam W@i

.y, and feedwater piping that rua near structures, systems, or components i;ff Y

important to safety.

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A description should be provided of the assu=ptions, methods, and

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3 results of analyses, including steam generator blowdown, used to f

cal'.ulate the pressure and te=perature transients in co=partments, hg w

pipe tunnels, intermediate buildings, and the turbine building IE xx following a pipe rupture in these areas. The equipment assumed to

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i k.l function in the analyses should be identified and the capabiligy u s; of systems required to function to meet a single active component

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failure should be described.

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A description should be provided of the methods or analyses performed

,.h to demonstrate that there vill be no adverse effects on the primary dd.

M and/or secondary containment structures due to a pipe rupture outside 71??

these structures.

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