ML20235C003

From kanterella
Jump to navigation Jump to search
Forwards Answers to Outstanding Questions from 781117 Wh Zimmer ACRS Subcommittee Meeting,Per 781219 Request
ML20235C003
Person / Time
Site: 05000000, Zimmer
Issue date: 02/13/1979
From: Flynn J
CINCINNATI GAS & ELECTRIC CO.
To: Savio R
Advisory Committee on Reactor Safeguards
Shared Package
ML20234A777 List: ... further results
References
FOIA-87-40 NUDOCS 8707090270
Download: ML20235C003 (26)
v • d • e


Text

,_

l r

i S

SB 3/??&?a2/6 s

w l

=

THE CINCINNATI GAS & ELECTRIC COMPANY d

cmciwNm osio 4eron 5

1 February 13, 1979 tic,.

f, l t at e

, q,

.-Um Dr. Richard Savio U.S. Nuclear Regulatory Commission

,bLd j 6 @g Advisory Committee on Reactor Safeguards

[

[

J gy J

Washington, D.C. 20555 f

' q,5.V t.

RE: OUTSTANDING QUESTIONS FROM THE NOVEMBER 17, 1978 WM. H. ZIMMER ACRS SUBCOMMITTEE MEETING

Dear Dr. Savio:

l This is in reply to your letter of December 19, 1978. Attach-ment A is the list of fcur outstanding questions from the November 17 Subcommittee Meeting which were not resolved at the November 28-30, 1978 fluid dynamics ACRS Subcommittee Meeting. Attachment B is correspondence from Dr. Ivan Catton to you concerning the Zimmer ACRS Subcommittee Meeting of November. While some of the questions in Attachment B overlap those of Attachment A t there are two questions which do not. These are the Fuel Bundle Lift and Control Rod Tubing Location questions. Attachment C contains our responses to Attachment A and those two questions from Attachment B.

As you requested, we are responding in writing in ' order to expedite the Subcommittee Hearings. It is our understanding that both the questions and the answers will become part of the record. Also, as you requested, these responses are being mailed directly to the Subcommittee Members and Consultants which you have specified.

Very truly yours, THE CINCINNATI GAS & ELECTRIC COMPANY

,/l

,.?19ly 4 9/-jf>;,w By f

j JAMES D. FLYNN, Manager Licensing and Environmental Affairs JDFvdew cc: With Enclosures Mr. Myer Bender Dr. Ivan Catton Mr. Harold Etherington l

Mr. I. A. Feltier (NRC)

Dr. Milton Plesset Dr. Zenons Zudans geg/g g (p% N Pl?",

W A A% F'>

A "., g i, o "h

SS K B707090270 870610 l

- tW;-

,.6*

i.".s r" ad PDR FDIA l

/

L Q. h I W d 4.c > 1 'S C E THOMASB7-40 PDR 0h O

MONO ikb o

l ATTACHL

~T "A" g.

ACRS SUBCOMMITTEE MEETING NOVEMBER 17, 1978 OUTSTANDING ITEMS REQUIRING RESP 0NSE FROM APPLICANT PRIOR TO THE JANUARY 17, 1979 g

SUBCOMMITTEE MEETING r

a*

sy i

1.

Transcript pages 13-15 Questions were raised regarding fuel bundle lift potential and

(

the dynamic behavior of the downcomer tubes during blowdown.

A These questions are addressed in more detail in the enclosed h.l 1etter dated November 30, 1978 Catton to Savio, and on page 135 of the meeting transcript.

1 l

2.

Transcript pages 45 and 154 l

A question was raised regarding the film coefficient used in ten i the drywell heat transfer analysis, regarding situations where G '

different types of conservatism would be desirable.

k 3.

Transcript pages 156-160 i.

The suppression pool time temperature history during blowdown L

was discussed. Questions were raised concerning relationship'p p.

between pool temperature during the worst (from the standpoint

[,

of maximum pool temperature) LOCA and the test used to establish T

chugging loads. The enclosed letter dated November 30, 1978

(

also addresses this.

[.

[,

4.

Transcript page 164

[

Questions were raised regarding the differences between single downcomer tests and the conditions existing with the multiple h

downtomers in the' actual suppression pool. The enclosed letter

[i dated November 30, 1978 also addresses this, i

/

L, ee_

fC

.. _.$% r '.; *y ~ <.. v' ~..~ ~~ ; " ::

.w

..~. ~

~

s o-7 ATTAu1 MENT "B"

November 30, 1978 CT- /D77

=ema.

DYtsokt committi on H Actos sArEccAtas s.s. s n To:

Dr. Rf savio From:

Ivan tton

Subject:

Zimer Subcommittee Meeting,16-17 November 197Tt t tMtW%546 8$

k Copies To / M. Bender, M.

S.. Plesset, J. Ebersole

~

I I would like to add a few coments to several of the topics discussed W.

rt,A at the Zimer Subcommittee Meeting. They are more generic than specific in 6

nature.

1 I*

Reactor vessel Supports. One of the contributors to reactor vessel

{

cupport loading is non-uniform pressure in the annulus around the vessel.

The annulus pressures are calculated using a one-dimensional code when the 2.:!' '

M;l phenomena is clearly two-dimensional. Large pressure variattuns exist and i

.the flow being two-phase results in sonic conditions being predicted to exist l

between volumes specified for analysis. This is accommodated by the analyst i

~

using a Moody coefficient approach. It is not clear to me that the predicted pressures are realistic or conservative.

N p..

Fuel Bundle Lift. Fuel bundle lift potential during the blowdown phase hk i of a LOCA, a concern of Mr. Ebersole, is not fully addressed in any document

[.$

f I have access to.

As far as I can tell, the upward force on the fuel bundle

)

{

is based on end of channel life friction with end of life being defined as f '-

{

f".

j f

the time at which the friction between the control rod and the assembly t 'r b,.

wall is high enough to impede withdrawal of the control rod. With this t

[e.

definition the upward force ambunts to 107 lb. The downward force, corrected l

f for water buoyancy, is 300 lb. The net is then about 200 lb in the downward f

f p,.e P"

'l f 0 N f #~~f A 4n i

in nr I

a.w,.

.....~:...,.

._.... 9,...,...,...

Qs,i. *.: - uus..:

.......s.*.>-

~..

'::~. ^: -'-

+

" ~

~

_ _ L ___

l

  • 'irectien. During blowdown, t.,. pres;ura in the bypass region'.11s fester d

than in the fuel bundle causing a decrease in the bundle to bundle gap,.

It i

is not inconceivable that a several fold increase in friction factor night

~

f_

~

result and the 2 lb margin might disappear. It is possible that I have f

missed an importait aspect of the problem.

If so, I would like to review

' us e

the documents clarifying how the fuel bundle lift question is put to rest.

Suppression Pool Downecmers and 1.ateral Loads. The downcomers are over

{

thirty feet long and have no lateral restraints. The magnitude of the 7s e

lateral loads depends strongly on the suppression pool temperature and f*

mm l downcomer mass flux. Asaresulttheloadsvarywi[htime.

If the pool temperature remains low enough, large lateral loads do not occur and my f

concerns are not well founded.

To assess the potential for large lateral loads, the pool temperature and downcomer mass flux time histories are h

needed. The pool temperature must properly account for stratification.

Some consideration should also be given to the Eeometric arrangement of the

\\

sighty or so downcomers and possible interactions as the data base is pri-marily the 4T tests. The submergence is ten feet and restraints, if needed, rg could be submerged eliminating concern about pool swell impact loads.

Control Drive Tube Location. The control rod drive tubes are located

'{.

e-very close to the recirculation line with half of the tubes being on each h^

q side of the vessel. The tubes are routed so that if half are lost, due to b

t.

o pipe break, the remaining can still scram the reactor. It seems to me

[.. '

that the possible loss of half your control drive tubes is an important consideration and maybe they ought to be less exposed.

[

t-i Mi b

+

+

,,,a.

. - ~.

%g M$ % eg&.

a4*

[#

T

'49-O

  • I v

t'*,Ib I*

,sg 4*

5*

e -

p-gp 9,

iTTACHMENT "C" Question No. 1 (Transcript Pages 13 to 15) y The downc er piping is analyzed for static and dynamic loads.

y The loadi cases were obtained from the DFFR and are identified in detail n Subsection 4.3.1.1 of DAR, Amendment 1, October 1976.

The loading combinations are per Table 6-1 of the DFFR and are explained in Subsection 4.3.1.2 of DAR, Amendment 1.

The design limits are identified in Subsection 4.3.1.3 and the analytical results are presented in Subsection 4.3.1.4 of DAR Amendment 1.

Tables 4.3-1 and 4.3-2 of DAR Amendment 1 are attached herewith showing maximum stresses in downcomers.

Since the stresses are within allowable limits, there is no need for providing a bracing system.

ZPS-1-MARK II DAR AMENDMENT 1 OCTOBER 197b f

TABLE 4.3-1 MAXIMUM STRESSES IN DOWNCOMER 8

t[.

~

STRESS (PSI)

LOAD (1)

VENT CONDITION (

CASE DESCRIPTION TYPE EMPTY FULL A

Dead Weight Static 436 436 B

LOCA Pressure ( }

Static 470 470 C

OBE Vibration Dynamic 7,428

)

OBE Pool Slosh Dynamic 523 D

SSE Vibration Dynamic 13,927 g)

SSE Pool Slosh Dynamic 981 E-1 LOCA Blowdown Dynamic 15,337 NA(3)

E-2 LOCA Vibration Dynamic 137 NA E-3 LOCA Pool Swell Dynamic 600 NA O

F SRV Bubble Oscillations Dynamic NA 6,917 SRV Vibration Dynamic NA 4,262 (Resonant Sequential 1

Sysnetric Discharge)

(1) As identified in Subsection 4.3.1.1 (2)

For Analyticci Convenience, LOCA pressure was assumed to act at all times.

(3) NA - not applicable (4 ) For Analytical Convenience, the larger " full" load case was used.

P I

4.3-8 J

b I

"PS-1-(IARK ii DAll AMENDMENT 1 OCTOBER 1976 TABLI:

.3-2 DOW'? COMER PIPING MAXIMUM COMBINED STRESSES *

/^

f k

EQUATION 9 ALLOWABLE STRESS STRESS LOAD COMBINATION CLASSIFICATION (psi)

(psi)

I Subsection 4.3.1.2.1 (a)

Upset 9,031 18,000 II Subsection 4.3.1.2.1 (b)

Emergency 11,957 27,000 y

III Subsec tion 4. 3.1. 2.1 (c)

Faulted 17,089 45,000 f

IV Subsection 4.3.1.2.2 (b)

Faulted 21,685 45,000 l

l O

  • The maximum stress always occurred at the drywell floor where the downcomer is anchored.

4.3-9

(

Question No. 2 3.

A.

Transc&ipt Page 45 f

M i

In calculating the pressure response in the drywell during a postulated loss-of-coolant accident, it is assumed that 1

there is no heat transfer to the drywell walls.

As stated in Zimmer FSAR Subsection 6.2.1.3.2.1, the heat transfer coefficient between the drywell atmosphere and drywell liner was assumed to be zero in order to obtain the most conservative approach; whereby appropriate film coefficients were considered in the temperature analysis for drywell concrete under various boundary conditions.

B.

Transcript Pages 154 and 155 l

Regarding the analysis about the temperature of the concrete, the film coefficients for convection at drywell walls and floors are shown on Table 2-1.

The drywell and suppression pool temperature transient responses after LOCA were used as the boundary temperature changes.

A typical temperature transient response in drywell is shown cn Table 2-2.

Figure 1 shows the sketch of primary containment with the area of con-cern.

The temperature gradients for drywell walls and floors are shown on Figures 6, 10, 11, and 12.

4 e

i l

i t

l 1

1 l

l l

i i

1

1 *

(

r f

TABLE 2-1 e+

p:

Heat Transfer Coefficient for Convection #

At 0.02 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after LucA occurred Location Film Medium h During Transient BTU /(HR) ( F) (FT)

.\\.

E,H,M,G Air (af ter 0.1 hr) 1.03 I

Water 92.56 P

Water 85.44 J

Water & Air 35.60 l

K Water 87.11 L

Water 43.55 E,H,M,G Condensed Steam 2000.00 l

I l

l

\\

l 4

Reference:

" Heat Transmission", by William H.

McAdams, Third Edition, McGraw-Hill, 1954 1

c' 1

_______________a

I TABLE 2-2 n.

y-Boundary Temperature s

Drywell Suppression Pool Time Temperature Temperature Hr.

@ Boundaries

@ Boundaries

'P

'F 0.0000 135 90 0.00003 137 90 0.00005 155 90 0.00010 170 90 0.00050 230 92

~

0.00100 250 93 0.00200 265 95 0.00400 283 115 0.00600 285 125 0.00800 285 128 0.01000 285 130 0.02000 285 132 0.04000 27L 135 0.06000 l 260 140 0.08000 230 144 0.10000 205 146 0.20000 185 148 0.40000 180 155 1

0.60000 183 159

)

0.80000 185 162 1.00000 187 165 2.00000 190 172 4.00000 200 195 6.00000 205 204 l

8.00000 207 208 I

10.00000 209 211 20.00000 206 216 40.00000 190 200 60.00000 175 180 80.00000 170 165

)

100.00000 155 360 200.00000 135 152 a

If 1

i 3

i i

1 P.icarc 1.

The Sketch of Primary Containment g 1

p,/

with the Area of Concern l

C i

l s/...,.

\\.

.. \\...

1 7

i J

s DRYWELL COI.'I'AIINENT WALL S

i...

..t t

?-

CORE

+-

3 p-

-. v..:nt..<:r-C

, J.

s,i. s. '.

u.-.

STEEL LIXER PLATE I

.-.i

. J.

(,_ _.'s,

( $ACRinCML (E10 LOGICAL)

F cs.,...,

.. '.\\ t SHIEi.D

)

I N o..'.

7..,..,l

\\,

..g

..'s...

y(.,....,

e

..J

_..'_.. -. Ga'_ v M.._._._. _. _. _.._.H._'".'

s

,n 2* O.-

-h.

5 J

fA'[, - PROPOSED OPENING

  • s,

.--.\\

....M-4; I

u. -

y u2

.. ;-,3

.,..,,. 2, _ _ ?

q..

.y.

f. -l [....

p

.a

(

f-.[

DRYW M FLOOR J :.'...

Q 1 >54

/ SUPPRESSIO.'I CIWGTR smrrL WALL s

N _-

. - _L. _

= QlIE1..I

+E 5.:

f#

.21. '

I

'fe

=-,._.F PLATE,

J. t'

.6 v

(L

.. '; R

-~ \\

T T

, '.. ~wn.,.

K 1

w

__rvs;..

.... j 1

~...,

s......

.j

.,, i., 7 ;....r,

....,5 g...

i

.,,..,s-S

~t

=s e

BASE P.AT~

.' y

_.___________m.

(

2 OO F.1carc 6.

..47 ter.d;/-5tato r.nl Trantient Tecperaturc Distribution within Dryvell F1mr (f.rea M 200 -

k g

viii.h c i u.r., = c e y s c e s ir. 6 c e< in : Tr e.,s ie n,

~ ~ ~ ~

'.9 '.;,

o o,.

tr of 30 hr j

,\\

N r

[

]

180

_0

(

l

\\

30

.\\

he i

\\

i r

i s

(

l.

160 N',

('!

e l

20,

\\

he ~.,__,,.'

m

,he 4

\\

s t

\\

3

\\N 2

10 3

140 "

l g j,

=

hr F<

',, g4r

!a N

i e

N 04 fg

\\

o.o

/

,/

N s.. _

conc n tit I

I 100 O

10 20 30 36 40 WA LL TH 1 CK N b S.S, X. I N ')

4.

~

s.

)

1 ae 0

e 6

f

(

hl tl ia wW nt on ie t m un 0

bi i a 5

r, t

  • n so i C D

l el 0

re 1

uw t y e

ar r

rD u

e 0

g pn i

mo 4

F ei Tl E

T p

E t c R

nu C

ec N

is sA O

n C

a n ro Ti ta 0

s 3

ne dno C

0 2

4

%(.N 3

4Nx g o.-

0 s-1 2

N xS ia s,

i c"i Rf L

E f m f..

AfUA E

i l

t

$LP 1

,l 5l hL f g,)

0I 1

o il l

l 51l i,5il i'l*

] 1. a E

,e=3 0

0 0

0 5

O 0

0 0

S 0

5 3

2 2

I 1

^NVE. N :a k e ;,i-e c

)

l

4 4

5

'I re 4

4 IoisJno 1

4 v g a

f O

m

~

N g

/

d.

c 3

y i:,

V t

3

.'/

.t: s

+ '.

  • 1.,

lr :c A

/f y-'

C

.)

o

.O

=.-

se C

O a

. i,

%^

v

.Y

,o...

I el d J

o L e2 e,

C L1 o e,

h 1

T, s

i;

1.,..

.i l'

l*

k 4

m

  • $ 54 D

aj em b

L' b N

Q

~

c v

B

$, c

?.O ll O

a *:'

m y

~

td li c (g

/

d

.I O

t.

//

w u ::

/

a ec o.<

0 N $

~

o C'

r L'

C ir

=

"'1 i

C k

I f

o

.c U

O V

/

k

/

aC b

O o

n

/ o a*1 1

s

.c O

N,-

~

- as a

y g

s

  • c-f s _s e e ss

/

p'$

1 g..

,, m __

.e

= - - - -

__.-g_,,-

o_

AQtSNI O

O o

O o

o o

m o

M d

N

( d ) 1 '3M1F63dN_I '.

o

(

s

,g 300

' E-.

?

g r

!4 c

Figure 12.

Trancient Tc=perature Distribution vith

~

j Condonnr. tion Assumption-F.0cctor Supprortir3 litll(Area'3)'

u 2.50 b

I l'

n.

h,

~

e t

e d

D

- P<'

200 e

CQm c

l ct 4

4 aw 3

.3-z b

-N

. *w-o ll C

.c (I

g k

f, t.: -

fr 250

)

P

" 40 hr p

20,

p s

.em n.

- - - - = = -

I

  • -~ O m' n 1

CONCRETE

)

100 i

i 0

to 20 30 4o 48 s 1 O

M WALL THICNNESS, X (IN) 1

~ _

(

l QUESTIONS 3 AND 4 The main vent downcomer loading conditions are described in FSAR 4

)

Section 1.2.3.5 (page I.2.3-8 cf Appendix I), Section 3.3.1

)

(page 3.3-I;), Section 4.3.1 (page 4.3-1) of the Wm.

H.

Zimmer i

Design Assessment Report.

The design lateral load of 8800 lbf applied at the tip of each downcomer represents an upper bound I

load obtained from foreign licensee data reported in NEDO-21018 for maximum pool temperature and negligible air content.

l j

)

Consideration of the grouping of downcomers in the specific l

Wm.

H. Zimmer Plant geometry arrangement are also discussed in i

the Wm.

H. Zimmer Design Assessment Report in the sections i

referenced previously.

A comparison of the dynamic loading conditions measured on the downcomer in the GE-4T tests with the design load of 8800 lbf l

is presented in Section I.2.3.5 of Appendix I in the l

FSAR.

It is concluded in all cases that the design load as l

l defined in the Wm.

H. Zimmer Design Assessment Report is a conservative upper bound loading condition for design assessments.

Additional items raised by the ACRS regarding the pool tem-l perature, pool temperature stratification, and downcomer L.

mass-flux time-histories are addressed in the following para-graph.

i i

Steam mass-flux time-histories, pool temperatures, and air con-tent are shown in Figures 1 through 3 for the Wm. H. Zimmer Station and also for a typical 4T test run.

The following data is shown in these Figures:

l Figdre 1.

Flow Regime Map for Steam Injection:

T vs m-Recirculation Line Break for Zimmer Figure 2.

T vs m for 4T Typical Run and Recirculation Line Break for Zimmer Figure 3.

% Air vs m - 4T Typical Run and Recirculation Line Break for Zimmer Mass Flux Titre Histories CONTEMPT /LT 026 was run to predict the necessary mass-flux time histories.

The recirculation blowdown transient was taken from the Wm. H.

Zimmer FSAR.

Heat transfer on drywell walls was ignored in this calculation.

The pool temperature was maximized by requiring 100% liquid entrainment in the vent flow.

The air content is calculated as the ratio of air mass to steam mass (% of gaseous portion of blowdown).

The results compare well with the results given in the Wm. H.

l Zimmer FSAR.

l l

e______-__

(

i QUESTIONS 3 AND 4 (Cont'd)

Since detailed data on the 4T l' lowdown and pressure response L

were not abailable, blowdown rates were scaled from Figure 5-33 of NEDE-13468-?- (Phase II & III 4T Test Report).

Results were compared to Figure 5-28, the pressure response for a similar run.

The differential pressure (driving force for vent flow) compared well to test results.

The 4T transient is somewhat different from the DBA because the 4T break was a vapor blowdown rather than a liquid blowdown and the break flow decayed more rapidly.

However, the 4T test did give a full range of steam

. flow rates at similar conditions to the Wm. H. Zimmer Plant.

Although the pool was not heated as much by the 4T blowdown as the actual Mark II pools are expected to be, the runs done at the 1200 F initial pool temperature will bound the Wm. H.

Zimmer Plant condition.

These results are-summarized in Figures 1 and 2.

The air content for the Wm.

H.

Zimmer Plant may be compared with the 4T test results shown in Figure 3.

It.is seen that the range of air content is approximately the same in the test as in the Wm.

H.

Zimmer Plant.

Pool Temperature Stratification Pool temperature stratification data is given in Figure 5-26 (NEDE-13468-P).

This figure shows a temperature profile (as elevation varies) in the pool at various times after the beginning of blowdown.

A large temperature gradient is seen after 20 seconds.

The gradient is much less at 40 seconds and later which is due to the excellent mixing promoted by chugging in the 4T test facility.

This is supported by traces which show the chuggin% beginning at about 26 seconds.

However, even if one applies the maximum stratification observed in the 4T test to the entire Zimmer transient (clearly extremely conservative) local temperatures do not exceed 2050 F.

Assuming that all steam condensation injected into the pool creates mixing and that chugging is a significant mechanism for mixing, it is clear 0 F and 1500 F) 4T test will bound that the high temperature (120

.all expected in-plant conditions.

)

]

l l

m

,i{l1l!II s - y g,32yog*.f;o-MMM M 5

g CCC C 0

P 0

.,f 4j 5

5 ;t1; 804 4,

, );

5 0.3 5

_9 5

)

5422 y

r

= =

)'

c t

)

a dDL8 a

S 4

l

(

(

l

)

e e

)

i lo p

b c t

m s

a o

a 4

O m

P h

(

i S

_O 0

i x

r o

3 e

t h

l a e W

t a

J p r

a p

d h

t.

i S

A c

o e

s e

y f

p

- )

n e

l i

o l

nr n

s l

C b

oo O

e l

(

a if p -

E t

t c

a

(

r

) S ck c

k ea A

5 3

s t

a c

_5 je c?

E a

e 4

G nr J

r

(

e s

IB e e m

B p*

iS t i a

ne e

e eh r -

a2 t

n J S an ei d

S i

tL n F L

e a/

n S

o c

n cM n

S o

C e

ro s.

L

)

(

oi i

ft e

3 t

2 a

r-xR

(

s' a

0 v,

r

,j l

n pl 0

/

u n n u

g ne au g

5 el o i c

K ot Mcr P, F i a r

ia t r i

sW er. ic ws a T c

x s

/

i e o s l

e u

eh l

a rt8 l

e R

9R l

F N F

px7 i

l e

c b.

r pi9 Raf a

s b

e s

uS1 C

)

ea O u S

x ws#

s 1

t e t.

(

B r

a ov" M

ea9 FT2

st e

(

1 l

S t

l Z

r 5

4 t

x b

m ud6 b

~

1 i

se E

u

']

aa str B

t ene 1

r S

reb Psm a

e n.

ee a

nrv e

c

)

iPo i

N g

r.

d

\\

s.

i a

a F

e

)

dA o

c D

aLg L

2 oCn o

t

(

LUi l

0 S

t' t

a

~g a

a-c 5

e n

1

. e

.j i

c t

mlM d

i a

g r aa a

f o

g ny.n e

r L

u o

V e

h

)

Dti a

et t '

i C

e

)

a 2

Y a

n c

r g

(

o,c r

2 e

nr

(

t a

s soo y

g g-,

a L

g etf r

(

L u

i t n.

c h

n h

daI d uC M

i e

=-

C a

g g

t y

l g

r l

S

t l

u a

a eg i

h n

tf r C.Q~

c r

nau O'.

s e

eesb O

t.

cm s

r nnrr I

eioe r a t t.

et ct MijlIi>lf I

fnai

'I 1

,.I-

.t i i.: [ii If eoea

~

RCRG 0

0 O

0 0

2 f-3 0

1 o

2.m73% Lyn a

i

<l

\\l\\l!lllll

k aer F

B C

e G

2 n

L 0 1

i L

3 r

e n

u o

t i

a t

n re

~

la wo g

u d

w c

e N'C r

o T

l ic e

B 4

l e

cn T

o R

o e

4 P

r i

Z l

e l

m, "d

a a

nr c

i

~

1 i

ao p

t 2

f i

n y

uk T

n _-

N I

R a 3

c l r

,o aS c

i e

- t pn TL

~

yi Tn 4 o i

rt oa n c -

-, g fl u i

r vc

~

si e

TR 2

er d

u F

g O

i F

C 7

er u

t

~

a n r we a p

~

d m w e oT l

Bl o

7 o 4P l l a a ci i t pi yn T, I i!-

{ ii.

1 l

.i i.

J 6

0 S

1 3

o t

1 L

u.

cs; % s.g$ -0 of

,(,1I ii!

b

,I l

e

. o m.

k s

n a

w e

c r

s d

B w

s o

e l

n S

i L

T4 no c

l i

a t

a. 1 c

a r

i l

p u

A y

c dw T

r n

i ai t

c Z

f n

/

e R

ur Ro r

f e

l ak n

0 ca t

Z 2

i A.

ic pr T

yS e

7n 1l

,1-l 4

<e 4i L r

~

on fo i

. eta s

sl a.

e vuc rr ii Ac e

%R 3

e o

r i.

u s

g i

F g

o

~

.l it!

I i-

~

G o.

0 0

O 5

9 3

2 1

0 E R <* :.8 CO u u k p l'i 1

s i

11

(

CONTROL ROD DRIVE PIPING LOCATION _

During the November 17, 1978 Wm. H.

Zimmer ACRS Subcommittee Meeting, the question was raised regarding the closeness of the Recirculation System piping to the Control Rod Drive (CRD)

Insert & Withdraw piping and given a break in the Recirculation piping what effects would result and would the reactor still

~

achieve scram?

Design Considerations The CRD insert and withdraw lines (3/4" - SCH 80 piping) are routed such that half of the lines are on either side of the reactor vessel.

Appropriate design considerations were given to the effects of a postulated recirc pipe rupture which would lead to pipe whip and/or jet impingement:

1.

Pipe Whip Restraints The Zimmer Plant Design considered the possibility of pipe whip and for the recirculation piping a pipe whip restraint system is provided.

The design provisions and criteria used to assure that the reactor and all essential equipment within primary containment are adequately protected against pipe whip are discussed in detail in FSAR Section 3.6.

High energy piping systems were analyzed for dynamic potential.

Both longitudinal and circumferential breaks were postulated.

Design criteria required pipe in any possible direction about a plastic hinge formed at the nearest pipe whip restraint cannot impact any structure system, or component important to safety.

FSAR Figures, 3.6-1 and 3.6-2 locate the pipe whip restraints as well as various breaks considered for the recirculation system.

FSAR Table 3.6-2 summarizes the results of the above analysis.

From this table, we confirm that no CRD piping is located close enough to the recirculation piping to be contacted during pipe whip.

2.

Jet Impingement The CRD piping is 3/4" - SCH 80 pipe which is routed in two groups on either side of the reactor 1800 a part.

Postulating a recirculation system pipe break the piping restraint system will restrict pipe movement (see FSAR Sections 3.6.4 and 3.6.5) but a resultant fluid jet force could possibly l

impinge on some of the CRD insert and withdraw lines.

(It is noted that the CRD piping is located on either side of the l

fl 4

~

e CONTROL ROD DRIVE TUBING LOCATION (Cont'd) recirculation pump and only some of the CRD piping would be affected by a break in the pump suction piping and conversely only some affected by a break in the discharge line.)

However given that small pipes have high crush strength properties, the jet impact velocities will have an insignificant effect in compressing the CRD lines.

Additionally,

- there is no structure directly behind the impacted lines to support direct reaction loads.

Consequently, most of the impact energy imposed will either be transferred into kinetic energy or absorbed as strain energy in bending.

Therefore, compression of the piping is not of concern.

On the subject of crimping of small piping, experience tells us that a 3/4"

- SCH 80 pipe could be completely bent around a 24" pipe (this is not possible with a jet impingement force but helps to demonstrate a point) and would not be crimped closed (i.e.

minimal cross-sectional flattening).

A minimum of 3 gpm is required to accomplish scram and the piping would have to be completely sealed to prevent flow.

Thus, it is physically impossible to prevent scram by jet impingement or pipe whip against the CRD piping.

3.

CRD Piping Rupture The CRD design is such that, if CRD piping should rupture, reactor pressure will act upon the drive piston causing rod insertion.

Neither jet impingement or pipe, whip (becausa of restraints) could cause a pipe rupture however.

e F

t F27 4

a 0

270 F26 g

F22

\\

F2i y so F2 o

h F19

).

F4 F15 54 F14

  1. j (Fia N@ J K.

s' e "% "'

e D7 C

ss Ab

- RHR-C UMFERENTIAL BREAK 3 !"A!a'#2$'rktu A1 S'

L ANY ANGLE PARALLEL TO PtPE (

s' O BREAKID g,

+

O i

V@

e,,

v c,

d-

\\wN v>

e@

sia WM, H IlMMER NUCLE AR POWER ST ATION. UNIT 1 FIN AL SAFETT AN AL Y Sl6 REPORT FIGURE 3.6-1 RECIRCULATION LOOP-A WITH POSTULATED BREAVS L -- ---- ------- - ---

1

\\

F27 N

F28 1

pg i

a :: 3 ::1. w AD p(

t (f

F13 L6he c

O INDICATES POSTUL ATED RHR CIRCUMFERENTI AL BRE AK L NG TUDINAL BRE A AT

\\

ANY ANGLE PARALLEL TO PIPE (

O BR E AK 10 8

i vD e

7@

hJ g

W of?e l

v G

s1o l

8 NOTE: This figure is identical to 3.6-1 WM. H. ZlM M ER NUCLE AR POWER ST ATION, UNIT 1 FtN AL SAFETf A N AL f $i5 REPCRT FIGURE 3.6-2 RECIRCULATION LOOP-B WITH POSTULATED BREAKS

ZPS-1 TABLE 3.6-2 RESULTS OF DYNAMIC ANALYSIS ON RECIRCULATION SYSTEM PIPE RESTRAINT RESTRAINT DEFLECTION BREAK

  • LOADING PEAK DYNAMIC DEFLECTION AT BREAK ID DIRECTION LOAD (kips)

(inches)

(feet)

REMARKS *

  • 51 R-511 0.805 0.451 1/2-inch clearance S3 T

456 0.340-0.164 1-inch clearance S6 R

605 0.806 0.396 1/2-inch clearance S6 T

693 1.06 0.290 1-inch clearance S9 R

512 0.808 0.297 1/4-inch clearance 510 T

548 0.56 0.485 1-inch clearance D12 R

504 0.791 0.244 1/2-inch clearance D3 R

135 0.239 0.705 1-inch clearance 1-inch D3 T

482 0.395 0.272 clearance D4 T

270 0.151 0.423 Load on each restraint; 1-inch l

clearance D6 T

362 0.181 0.613 1-inch l

clearance D8 R

257 0.448 0.422 Load on each restraint; 1-inch cicarance D9 T

203 0.074 0.362 Load on each restraint; 1-inch clearance F1 T

404 0.473 0.667 1-inch i

clearance

  • Fee Figure 3.6-1.
    • Clearance is the physical gap between the pipe OD and restrcint cable or restraint frame, respectively, as indicated by the loading direction.

3.6-25

(.-

ZPS-1 TAFLE 3.6-2 (Cont'd)

PIPE RESTRAINT RESTRAINT DEFLECTION BREAK

  • LOADING PEAK DYNAMIC DEFLECTION AT BREAK ID DIRECTION L0rtD (kips

_ inches)

(feet)

REMARKS **

l

(

F4 T

157 0.058 0.312 1-inch clearance F6 R

181 0.355 0.498 1-inch clearance F7 T

162 0.059 0.991 Load on each res-traint; 1-inch clearance 4

4

  • See Figure 3.6-1.
    • Clearance is the physical gap between the pipe OD and restraint cable or testraint f r.ame, respectively, as indicated by the loading direction.

3.6-26

.(

i FUEL'BU11DLE LIFT A'GEjLicensing-Topical' Report,"BWR/6 Fuel' Assembly Evaluation of Combined Safe Shutdown Earthquake- (SSE) and~ Loss-of-Coolant Accident (LOCA) Loadings", NEDE-21175-2-P, demonstrates that-

'DWR/6 fuel will:not lift.from its seat even in the event of-combined SSE and LOCA and indicates that the effect of control-rodLforces may be conservatively omitted from the evaluation.

This'is generally applicable to Zimmer.and other BWR/5 reactors also.

The tendency.for the control rod to lift the fuel'as a result-of fuel channel bulge has been considered.

The amount of bulge required'to prevent control rod settle should not be experienced since it would require channel =re-use well beyond design exposures.

Operating. reactor experience indicates that. channels are replaced'so as:not to-require control rod insertion if the control rod does not settle.

Even if the channel were used to the point where control rod settle did not-occur, the drag on each of the'four adjacent fuel assemblies would amount to only 42 pounds.

This_ force would be completely relieved within 0.030 inch of lift as the fuel nose piece moved on its seat since'the calculated channel-blade interference to prevent control-rod r

settling is 0.030 inches and the radial clearance actually provided

.for in the fuel support for the fuel nose piece.is.030 inches (See attached-detail A).

If the peak LOCA pressure differential l

pressure were'to occur at that instant,.it is conservatively calculated.that'the resulting drag would be 75 pounds per fuel assembly.

This load would be resisted by friction in the fuel support wher,e it.is reacted and'the net effect would not be significant.

P l

}

I

- _= : -

a

gr:

(

(

(.,'

1 e

si I

w '1 h

Y ll; E

r\\

)

I t

TUEL ASSEELY l

~

LOWER TIE PLATE i

FUEL TO FUEL

) I l

$UPPORT SEAT s

1 i

s i.

e 0.030" CLEnRENCE

/,, ?-

i

\\

'/

cIy f i

FUEL $UPPORT i

q'/ -

TO GUIDE TUBE 1

2-

$ EAT

's t.

I

[

C~._g,

l T

i I

I FUEL SUPPORT I

l f

I CORE PLATE-f,'

/~ GUIDE Time l

i 1

vs i

1 DETAIL A 1

i i

i d

.f

_ -. _...

Retrieved from "https://kanterella.com/w/index.php?title=ML20235C003&oldid=5475923"