ML19340B048

Forwards Requests for Info Required for Consideration of Effects of Piping Sys Break Outside Containment.Requests Provision of Analyses & Other Info Needed to Determine Consequences of Event.Press Release Encl

ML19340B048
Person / Time
Site:	Dresden
Issue date:	12/15/1972
From:	Anthony Giambusso US ATOMIC ENERGY COMMISSION (AEC)
To:	Brian Lee COMMONWEALTH EDISON CO.
References
NUDOCS 8010170759
Download: ML19340B048 (14)

Text

_ _ - -

t I

Docket No. 50-10 Comonwealth Edison Company I'

ATTN; Mr. Byron Lee, Jr.

Assistant to the President Post Office Box 767 Chicago, Illinois 60690 Gentlemen:

l The Regulatory staff's continuing review of reactor power plant safety indicates that the consequences of postulated pipe failures outside l

of the containment structura, including the rupture of a rMn stcan or feedwater line, need to be adequately documented and analyzed by licensees and applicants, and evaluated by the staff as soon as possible.

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

" Structures, systems, and components important to safety i

shall be designed to accomodate the effects of end to i

be compatible with the environmental conditions associated I

with normal operation, maintenance, testing and postulated accidents, including loss-of-coolant accidents.

These I

atructures, systems, and components shall be appropriately j

proteeted against dynanic eff acts, including the effeets i

of missiles, pipe whipping, and discharging fluids, that may result from equipment f ailures and from events and conditions outside the nuclear power unit."

Thus, a nuclear plant should be designed so that the reactor can be shut down t.nd maintained in a safe shutdown condition in the event of a postulated rupture, outside containment, of a pipe containing a high energy fluid, including the double-ended rupture of the largest pipe i

in the main ateau and feedwater systems. Plant s tructures, sys tens,

and corponents important to safety should be designed cnd located in the facility to acco=odate the eff cets of such a postulated pipe failure to the extent necessary to assure that a sde chutdown condit. ion of the reactor can be accomplished and naintain3d.

Based on the information va presently have available to us on Dresden Unit 1, we understand that generally the plant design appears to be l

i such as to withstand the effects of a postulated rupture in accam or m

fer'imter 1 Sen. A_possible execrtien to this conclucien is the locntica t

i

I I

l Cornonucalth Edison Company....,

,. ::s I

of the feed. tater line in close proximity to the tain stccu line without an intervening barrier for a short distance. Your analyses show that, at present, Unit 1 can meet the AEC Interim ECCS criteria only by taking credit for continued operation of the feedwater systea for a partial spectrum of breaks. It is not clear whether a stcan line break could rupture the feedvater line or whether a steam line break l

would require feedwater addition.

Depending on the analycis of this problem, some modification of the facility may be necessary.

We request that you provide us with annlyses and other relevant information needed to determine the consequences of such an event, using the guidance provided in the enclosed Eeneral infor=ction request.

l The enclosure represents our basic information requirements for plants nov being constructed or operating. You should determine the applicability, for Dresden Unit 1, of the items listed in the enclosure.

I If the results of your analyses indicate that changes in the design of j

structures, systems, or components are necessary to assure safe reactor j

shutdown in the event this postulated accident situation should occur, l

please provide information en your plans to revise the design of your i

facility to accocanodate the postulated failures described above. Any l

design modifications proposed should include appropriate consideration of the guidelines and requests for information in the enclosure.

We vill also need, as soon as possible, estimates of the schedule for design, fabrication, and it.stallation of any modifications found to be necessary. Please inform us within seven days af ter receipt of this letter when we asy expect to receive an amendment with your analysis of this postulated accident situation for Dresden Unit 1, a description I'

of any proposed modifications, and the schedule estimates described above. Sixty copies of the amendment should be provided.

t A copy of the hesion's press announcement on this matter is also enclosed for your information.

Sincercly, e

Q.y A. Giambusco, Deputy Directer

+

for Reactor Projects Directorate of Licensing Enclosures and cc:

See next page.

M-

i 1

-l i, -

1 I

Co:=c::vealth Edison Company -

Lnclosures:

1.

General Infor >:ation Request

[

2.

Press Reicase dtd 12/13/72 cc v/ enclosures:

.ohn V. Rowe, Esquire Isha=, Lincoln & Beale Cot:nselors at Law One First National Plaza Chicago, 1111acis 60670 l

1 Distribution' Docket File P D.

l RP Read tg L Reading i

Branch Reading JRBuchanan, ORNL DJSkovholt, L:0R i

AGia:busso, L:RP RS:y d, L: Zh'R TJCarter, L:0R AC?S (16) l R0 (3)

JScinto, OGC DLZie= ann, L:GRB #2 1

RDSilver, L:0RB #2 FG iggs, L:0RB #2 4

1 i

9 i

p.

L::RB #2

.L : OR3,, i:2

..... L : O.. R,B...J',2.. !L:0R

!L:Bk, h g..

L:RP i

la,, ?

Ih.,/i f,j0.. lL e)Q 3

r.

I'

\\

?.;5 il cer : < j'c FpTig'gs DLZienann ~. DJSlp,$'hcit REcyd

!.Gianbt:s s c-12ila/72

,'12/14/72

,12 //.f-/ 72

,.12/(f/72 jl2/g/72

,12/ _/72

. u..,.

....,u...w

=.

. ~

l J

General Information _ Required for Consideration cf the Ef fects of a Piping System Breck Outside Centainter.t 4

j 1he following is a general list of information required for AEC review of t i.. - ef fects of a pipfng system break outside containment, including i.

the donhle ended rupture of Lthe largest pipe in the main steam and feed-wate r sys tems, and for Ar.C review of any proposed design changes i

that may be f ound ne cessary.

Since piping layouts are substantially different from plant to plant, applicants and licensees should deter-ine on an individual plant basis the applicability of each of the following i

Items for inclusion in their submittals.

1.

'he systems (or portions of systems) for which protection against pipe i

whip is required should be identified.

Protection from pipe whip need not he provided L f any of the f ollowing conditions will exis t:

(a) lioth of the following piping Fystem cc*.ditions are met:

(1) the service temperature is less than 200* F; and (2) the design pressure is 275 psig or less: or (h) The piping is physically separated (or isolated) from structures, s ys tems, or components important to safety by protective barriers,

a p

or res trained f rom whipping by plant design features, such as concrete encasement: or (c)

Following a single break, the unrestrained pipe movement of either end of the ruptured pipe in. any' possible direction about a plastic

' hin ge formed at the nearest pipe whip restraint cannot impact any s t ructure, sys tem, or component imp or t an t to safety; or f

4 mm m

0 4

.S a

=

we n

. (d) The internal energy level associated with the whipping pipe enn be demonstrated to be insufficient to impair the safety function of any structure, system, or component to an unacceptable level.

2.

The criteria used to determine the design basis piping break locations in the piping syste=s should be equivalent to the following:

(a) ASME Section III Code Claqs I piping breaks should be postulated to occur at the following locations in each piping run or branch run:

4 1

(1) the terminal ends; (2) any intermediate locations between terminal ends where the primary plus secondary stress intensities S (circum-ferential or longitudinal) derived on an elastica 11y 1 The internal fluid energy level associated with the pipe break reaction may take into account any line restrictions (e.g., flow limiter) between the pressure source and break location, and the effects of either single-I ended or double-ended flow conditions, as applicable.

The energy level in a whipping pipe may be considered as insufficient to rupture an i=pacted pipe of equal or greater no=inal pipe size and equal or heavier wall thickness.

9

' Piping is a pressure retaining component consisting of straight or curved pipe and pipe fittings (e. g., elbows, tees, and reducers).

3A piping run interconnects components such as pressure vessels, pu=ps, and rigidly fixed valves that may act to restrain pipe movement beyond that required for design ther=al displacement. A branch run differs from a piping run only in that it originates at a piping intersection, as a branch of the main pipe run.

- l

i calculated basis under the loadings associated with one -

i j

half safe shutdown carthquake and operational plant 4

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

(3) any intermediate locations between terminal ends where the cumulative usage factor (U) derived from the piping fatigue analysis and based on all normal, upset, and l

testing plant conditions exceeds 0.1; and

]

(4) at intermediate locations in addition to those determined by (1) and (2) above, selected on a reasonable basis as necessary to provide protection. As a minimum, there i

should be two intermediate locations for each piping run f

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 l

run or branch run:

(1) the terminal ends; 1

l 4

j Operational plant conditions include nor=al reactor operation, upset conditions (e.g., anticipated operational occurrences) cnd testing conditions.

5S is the design stress intensity as specified in Section III of the ASME Boiler and Pressure Vessel Code, " Nuclear Plant Co=ponents."

i 4

6 i

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

1 I

1

. -(2) any intermediate locations between terminal ends where either the circumferential or longitudinal stresses derived on an elastically calculated basis under the loadings associated with seismic events and operational plant i

conditions exceed 0.9 (Sh+S) r the expansi n stresses A

exceed 0.8 S ; and g

(3) intermediate locations in addition to these determined by (2) above, selected on reasonable basis as necessary to provide protection. As a minimum, there should be two intermediate locations for each piping run or branch run.

4 3.

The criteria used to determine the pipe break orientation at the break 4

locations as specified under 2 above should be equivalent to the following:

l (a) Longitudinal breaks in piping runs and branch runs, 4 inches nominal pipe size and larger, and/or 7S is h

e stress ca culated by the rules of NC-3600 and ND-3600 for Class 2 and 3 components, respectively, of the ASMI Code Section III Winter 1972 Addenda.

1 5

is the allowable stress range f or expansion stress calculated by the rules of NC-3600 of the ASME Code,Section III, or the USA Standard Code for Pressure Piping, ANSI B31.1.0-1967.

E Longitudinal breaks are parallel to the pipe axis and oriented at any point around the pipe circumferance. The break area is equal to the effective cross-sectional flot

a. a upstream of the break le,7ation.

Dynamic forces resulting from e.

S breaks are assumed to cause lateral pipe movements in the direction normal to r.he ti'e axis.

s i

+

,-,e,

-m+- -.- -,,

--w-

-.ee

-,r-,,

,wtw

~

l :

)

(b) Circumferential breaks in piping runs and branch runs exceeding 1 inch nominal pipe size.

4.

A summary should be provided of the dynamic analyses applicable to the 4

design of Category I piping and associated supports which determine l

the resulting loadings as a result of a postulated pipe break including:

(a) The locations and number of design basis breaks on which the i

dynamic analyses are based.

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

and/or longitudinal break (s), for each postulated design basis break location.

1 (c) A description of the forcing functions used for the pipe whip dyna =lc analyses including the direction, rise time, magnitude, duration and initial conditions that adequately represent the a

jet stream dynamies and the system pressure difference.

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

I (e) A sum =ary of,the analyses which demonstrates that unrestrained motion of ruptured lines will r.ot damage to an unacceptable degree, structure, syste=s, or components important to safety, a

such as the control room.

9Circumferential breaks are perpendicular to the pipe axis, and the break area is equivalent to the internal cross-sectional area of the ruptured pipe.

Dynamic forces resulting from such breaks are assumed to separate l

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

J f

i 5.

A destript' ion should be provided of the ceasures, ao cpplicable, to protect against pipe-whip, blowdown jet and reactive forces including:

( a.)

Pipe restraint design to prevent pipe whip impact; (b) Protective provisions for structures, systems, and co=ponents l

required for safety against pipe whip and blowdown jet and reactive forces; (c) Separation of redundar.: features; (d) Provisions to separate physically piping and other co=ponents of redundant features; and i

(e) A description of the typical pipe whip restraints and a su==ary l

of number and location of all restraints in each system.

.1 i

6.

The procedures that will be used to evaluate the structural adequacy of Category I structures and to design new seismic Category I structures should be provided including:

(a) The method of evaluating stresses, e.g.,

the working stress cethod and/or the ultimate strength method that will be used; (b) The allowable design stresses and/or strains; and (c) The load factors sad the lead co=binations.

7.

The design loads, including the pressure and te=perature transiente, the dead, live and equipment loads; and the pipe and equip =ent static, thermal, and dynamic reactions should be provided.

..m-

~ _ - _ - _ - -

. ~

a

.7-4 I

S.

Seismic Category I structural elements such as floors, interior 1

vc118, exterior walls, building penetrations and the buildings as a whole should be analyzed for eventual reversal of loads duc j

to the postulated accident.

~

n

/

9.

If new openings are to be provided in existing structures, the capabilities of the modified structures to carry the design loads I

should be demonstrated.

1 10.

Verification that failure of any structure, including nonscismic i

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

1

]

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

(a)

Mitigation of the consequences of the accidents; and I

j (b) Capability to bring the unit (s) to a cold shutdown condition.

a j

11.

Verification that rupture of a pipe carrying high energy fluid will not i

directly or indirectly result in:

1 (a) Loss of redundancy in any portion of the protection system I

(as defined in IEEE-279), Class IE electric system (as defined in IEEE-308), engineered safety feature equipment, cable pene-I trations, or their interconnecting cables required to mitigate the consequences of the steam line break accident and place the 4

I reactor (s) in a cold shutdown condition; or 4

4

.--y

.------,._y

.-7,

~

i,

(b) Loss of the ability to cope with accidents due to ruptures of pipes other than a steam line, such as the rupture of pipes causing a steam or water leak too small to cause.a reactor 1

accident but large enough to cause electrical failure.

i 12.

Assurance should be provided that the control room will be habitable i

and its equipment functional after a steam line or feedwater line break or that the capability for shutdown and cooldown of the unit (s) q will be available in another habitable area.

13.

Environmental qualification should be demonstrated by test for that 3

electrical equipment required to function in the steam-air environ-ment resulting from a steam line or feedwater line break.

The in-formation required for our review should include the following:

1

,1 (a) -Identification of all electrical equipment necessary to meet requirements of 11 above. The time after the accident in which they are required to operate should be given.

(b)

The test conditions and the results of test data showing that the systems will perform their intended function in the environ-ment resulting from the postulated accident and time interval of the accident. Environmental conditions used for the tests should i

be selected from a conservative. evaluation of accident conditions.

(c) The results of a study of steam systems identifying locations where barriers will be required to prevent steam jet imping =ent from dis-abling a protection system. The design criteria fer the barriers should be stated and the capability of the equipment to survive withic :ne protected environment should be described.

4 y

_9 (d)

An evaluation of the capability f or safety related

ectrical eq ui p men t in the control room to function in the environment that nav exist following a pipe break accident should be p rovi de d.

Environmental conditions used for the evaluation should be selected f rom conservative calculations of accident conditions.

(e)

An evaluation to assure that the onsite pcwer distribution sys tem and onsi te sources (diesels and batteries) will remain operable th rougho ut the ever t.

14 Design diagrams and drawings of the steam and feedwater lines including branch lines showing the routing f rom containment to the turbine building should be provided.

The drawin gs should show elevations and include the location relative to the piping runs of safety related equipnent including ventilation equipment, intakes, and ducts.

15.

A discussion should be provided of the potential for floeding of safety related equipment in the event of failure of a feedwater line or any other line carrying high energy fluid.

16.

A description should be provided of the quality control and inspection p ro grams that will be required er have been utilized for riping systems o u ts i de containment.

17.

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

a discussion of its capabilities should he provided.

i

. I 15.

A summary should be provided of the emergency procedures that would be followed a,*er a pipe break accident, including the cutomatic and manual ope-stions required to place the reactor unit (s) in a cold shutdown condition. The estimated times following the accident for all equipment and personnel operational actions should be included in the proceSure summary.

19.

A description should be provided of the seis=le and quality classi-fication of the high energy fluid piping systems including the stea=

and feedwater piping that run near structures, syste=s, or ce=ponents important to safety.

20.

A description should be provided of the assu=ptions, methods, and results of analyses, including steam generator blowdown, used to calculate the pressure and temperature transients in cc=partments, pipe tunnels, intermediate buildings, and the turbine building following a pipe rupture in these areas. The equipment assumed to function in the analyses should be identified and the capability of systems required to function to meet a single active ec=ponent failure should be described.

21.

A description should be provided of the methods or analyses performed te demonstrate that there will be no adverse' effects on the pri=sry and/or secondary containment structures due to a pipe rupture outside these structures.

1 l

-. j

.u.

7_

e:2= _ ;.

^J

.'i '

r, "" ker.

f

. :n +:

e N

~

.,sr4r 4.r.:. Mm }. -y r : i; gd. _c_y3 1?".

.. :L y g)

'q,;

s - $

.,)

N Ab lm

' M:$ 0 $5 h 0h v

p p.,4 Q

c; ;w,1;.

ga;..q y

yJ, No.

P-429 FOR DiMEDI ATE RELEASE

Contact:

Frank Ingram (Wednesday, December 13, 1972)

Tel.

301/973-7771 AEC PIGULATORY STAFF REQUESTS DATA ON PIPE BREAKS IN NUCLEM. PLANTS The Atomic Energy Commission's Regulatory Staff is asking all utilities that operate nuclear power plants or have applied for operating licenses to assess the effects on essential auxiliary systems of a iaaj or break of the largest main steam or feedwater line.

These lines carry steam f rom inside the reactor containment building to the main turbine in the turbine building, and hot feedwater back from the turbine condenser.

The utility assessments will be evaluated by the AEC's Regulatory Staff.

The probability of a steam-line rupture is low.

Nonetheless it will have to be considered in the AEC's safety evaluation.

The reviev of the pipe break problem has been under way for several weeks.

It was started after the Advisory Com-mittee on Reactor Safeguards received a letter raising questions about the location of pipes in the two-unit Prairie Island plant in Minnesota.

The Regulatory Staff has reviewed the Northern States-Power Company application to operate Prairie Island, and on the basis of data availabic it has concluded that design changes will be required at Prairie Island.

Based on the new information--to be submitted by utili ~

ties as soon as possibic--the Staf f will determine what corrective action, if any, is necessary in each case.

The changes could include such steps as relocating piping. pro-viding venting of compartments, the addition of piping restraints, and, in some cases, structural strengthening.

i 0

D*9

• lD~33h D

n e %2 Xilk -