ML20064E246
| ML20064E246 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 11/09/1978 |
| From: | Gilleland J TENNESSEE VALLEY AUTHORITY |
| To: | Varga S Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 7811150152 | |
| Download: ML20064E246 (18) | |
Text
.
TENNESSEE VALLEY AUTHORITY CH ATTANOOG A. TENN ESSEE 374ot 830 Power Building November 9, 1978 Director of Nuclear Reactor Regulations Attention:
Mr. S.
A.
Varga, Chief Light Water Reactors Branch No. 4 Division of Project Management U.S. Nuclear Regulatory Commission Washington, DC 20555
Dear Mr. Varga:
[
t In the Matter of the Application of
)
Docket Nos.
0-327 /
Tennessee Valley Authority
)
Sol d Enclosed is TVA's draft response to question 8 transmitted in your letter to Godwin Williams, dated 3by 10, 1977. This response revises our response to Q6.54, as submitted in Amendment 49 to Sequoyah Nuclear Plant Final Safety Analysis Report. This response will be incorporated into Amendment 58.
Very truly yours,
/E a1
' J. E. Gil eland Assistant >bnager of Power Enclosure 781115 oI5 1 78111L 40 An Ecuat Opportunity Ereioyer
SQNP Q6.54 During long term cooling following a steamline break, feedwater line break, or small LCCA, the operator must control primary system pressure to preclude w
overpressurizing the pressure vessel after it has been cooled off.
a.
Describe the instructions given the operator to perform long term cooling.
b.
Indicate and justify the time frame for performing the required action.
c.
List the instrumentation and components needed to perform this action and confirm that these components meet safety grade standards, d.
Discuss the safety concerns during this period and the design margins available.
This should include potential adverse hydraulic conditions leading to inadequate cooling or mechanical damage.
e.
Provide temperature, pressure,-and FCS inventory graphs that would show the important features during this period.
The above discussion should account for the following:
a.
operator error or single f ailure c.
small LOCA's may occur in the ccid leg or in the hot leg / pressurizer.
d.
small LCCA's may result in nitrogen blanketing of the steam generators.
e.
long term cooling for a small LOCA may depend on alternating forced convection and vaporization depending on the treak location and size.
Fesponse:
The response to this question as submitted on the C.C.
Cook Unit 2 docket is an appropriate approach to the generic issues which have been raised.
See D.
C.
Cook FSAR amendment 78.
To address the issue of reactor vessel repressurization a' fracture mechanics study on the integrity of the 6.54-1
SCNP Sequoyah Units 1 and 2 reactor vessel beltline under f aulted conditions was performed.
The faulted conditions evaluated were the large steamline treak (LSB) and the small loss-of-coolant accient (LCCA).
These analyses supplement previous studies done for normal, upset, emergency, and faulted conditions as described in 13AR section 5.2.
The LSB transients used for this analysis are generic transients for a UHI four loop plant that have been modified to approximate the impact of pressurizer thick metal heat of the Sequoyah Units.
Two LSB transients were evaluated:
a case which assumes a loss of of f-site power that causes the rain reactor coolant pumps to stop (pumps tripped case) and a case where offsite power remains available that allows the main reactor coolant pumps to continue operation (pumps running case).
The FCS response for the LSB is shown in Figures Q6.54-1 through 06.54-5.
The transients used in the analysis of small LOCAs were taken from work for the British performed by Kestinghouse in 1974.
The PCs responses for the 2, 3, 4, and 6 inch dimater small LOCAs are given in Figures 06.54-6 through Q6.54-13.
These RCS responses were used to determine the temperature, thermal stress, and pressure stress profiles through the vessel wall in the beltline region as a function of tire.
These profilcs were then used in perforaing the fracture nechanics analyses.
In these analyses the following raterial properties were used for a longitudinal flaw in the base material:
Copper Content = 0.15 weight percent Phosphorus Content = 0.011 weight percent Initial RT NDT = 730 F l
For a circumferential flaw, the follcwing material l
properties of the core region circumferential weldment were used:
0.33 weight percent Copper Content
=
Phosphorus Content = 0.021 weight percent Initial RTgg7 = -400F These properties were obtained from the vessel fahrication raterial test certification for the Sequoyah l
Units.
The fluence used in these analyses was that l
calculated for a generic four loop vessel similar to the Sequoyah Units and satisfactorily approximates the I
fluence levels of these units.
l t
l l
l l
6.54-2 i
SCNP The irradiation demage of the material is correlated by trend curves.
These curves were develcped by Westinghouse to relate the magnitude of the shift of RT to the amount of neutron fluence and are a yp func[ ion cf copper content.
The final RT values are NDT then used to calculate the plane strain fracture toughness (K IR) and the reference fracture toughness (K Ig) as a function of the fractional depth through the vessel wall.
A two dimensional combined flaw analysis that is an approximation of a three dimensional flaw is used.
The results of this fracture mechanics analycis are presented in Table Q6.54-1.
These results are presented in terms of the maximum number of calendar years
(. 8 load factor is assumed) the plant will conform to the following criteria:
Minimum critical flaw is greater than 0.1156 a/t (1.0 inch) or flaw arrest is within 75 percent of the vessel wall thickness.
From Table Q6.54-1 it can he seen that for the two dimensional flaw method the vessel integrity can te shown for only about 30 years of plant operating life for two cases.
All other cases indicated vessel integrity is assured for at least 40 years.
For the 4 and 6 inch small LOCAs, the maximum number of calendar years the plant will conform to the vessel integrity criteria is 27 and 28 years respectively.
TVA is reviewing plans to perform a 10 CFR 50 Appendix G analysis of the Sequoyah vessels pricr to the one quarter service life surveillance.
This more realistic analysis is fully expected to verify reactor vessel integrity for the full 40 year plant life.
Based on the anticipated cutcome of the upcoming Appendix G analysis and the analyses already performed showing vessel integrity for nearly 30 years, vessel integrity is assured with adequate margin for the first 10 years of l
its service life.
l To provide guidance to the operator to be alert to the l
potential for vessel repressurization after an accident and also to he able to respond quickly, the plant operating precedures provide explicit instructions.
The l
operator is instructed to he continucusly aware of primary system pressure and temperature comparing them to 10 CFP 50 Appendix G pressure-temperature curves for l
Sequoyah which are provided in the Technical Instructions.
The procedures also identify the qualified instruments necessary for this monitoring l
action.
I 6.54-3 l
t
e Table Q6.54-1 Sequoyah Small LOCA and LSB Vessel Integrity Analyses 2 Dimensional Fracture Mechanics Analysis PLANT LIFE (Years)
BS-LG BS-LG
'n'D-C F Cu-0.15%
Cu*0.13%
Cu=.38(.33)%
Case P=0.011%
P=0.015%
P=0.021%
RT.DT1=40*F RTNDT1=7 F RTNDTl=- 0*F 3
2 Inch Small 40 40 40 LOCA 3 Inch Small 40 40 40 LOCA 4 Inch Small 40 27 40 LOCA
') Inch small 40 28 40 LOCA ISB with Reactor 40 40 40 Coolant Pumps P.unning LSB
{
with Reactor 40 40 40 l
Coolant Pu=ps t
Tripped J.
i-l l
BS-Base M.1terial LG-Longitudinal Flaw i
b'D-b* eld Material CF-Circumferential Flaw I_
Q6. 54-4 1
i L
Figure Q6.54-1 Large Steamline Break with Reactor Coolant Pumps Running Temperature Versus Time 600 400 C
L c
u 3
E 8.
!H 200 1
t 1
I I
0
'1000 2000 3000 TIME (Seconds)
~;--e. ~
/
Q6.34 5
l Figure Q6.54-2 Large Steamline Break with Reactor Coolant Pumps Running Pressure Versus Time 2400 1600 N
7 56 E
E Eu De 800 t
I i
I I
O 1000 2000 3000 Time (Seconds) ii&: -
T.
Q6.54-6 l
2 Figure Q6.54-3 T.arge St.eamline Break With Feactor Coolant Pumps Tripped s.*
Temperature Versus Time 600 400 C
L o
L 3
Eec.
0 9
200 f
i I'
i
=
i I
L i
l l
0 1000 3'000 i
3000 t
i Time (Seconds)
'Qf ;
/
v:
Q6.54 7 r
I l
Figure q6.55-4 Large Steamline Break with Reactor Coolant Pumps Tripped
- a, t'
l Pressure Versus Time 2400 1600 7
26 E
5 Ew D.e 800 I
l i
I I
~
O 1000 2000 3000 Time (Seconds) gn-s 7:
q6. 5 4-8
- _ =.. -.
l l
Figure Q6.54-5 4
i 1
Large Steamline Break with 1
Reactor Coolant Pumps Tripped 1
Flow Versus Time l
1.0 i
I Nominal Reactor Coolant Flow 94420 GPM/ Loop 0.8 -
t I
n
-e
.c.e 0.6 z
i C
c
.3 u
u i
n e
0.4 I
i 3
c 1,
I~
h 1
0.2 r
l l
1 1
I 4
1 0
200 400 4
1 Time (Seconds) i
.l "f1 46.54-9 J
t i
wn n.
.,-, we., - - -v-,,,n.,w,
.,n,_.-.......,n.,
nn,,.-,..,._,---...e-,--.n.-n.
-,,, _.. - -. ~,,..
-r,--.=-
-n,
,,c..
m
{
i j
Figure Q6.54-6
.I i
a s
Small LOCA Inch Diameter Break
. 3-s; Temperature Versus Time I
600 t
t 4
?
i 1
i i
I i
400 i
I
^
w O
E s
a u
4 e
1 Q
1 c.
4 Eoe j
200 1
I t
i i
r 1
3
}
t l
l 1
I I
~
l
)
0 2000 4006 6000 i
l Time (Seconds) i dr T.:u s -
f a
r i
Q6. 54-10 t
t i
l T
ir-v %,-ee 3 w-~rn iy, w y,-, r-e n w n w we-
-., ww w wwm--+-=
r w w-,v - w r*,:mw-w-,me
.---.-e--++----.
i Figure Q6.54-7 4
i s
Small LOCA 2 Inch Diameter Break
,e Pressure Versus Time 2400 i
1600 1
O O
w Vs C.
v D
u b
3 5
m (n
G k
4, 3
A 800 W
I I
i I
I O
2000 4000
-6000 Time (Seconds) en v
i qs.54-11 J
l 1
w
,-,.,.__.-.,.,_m,,.
,.,.,_,.,,,,,_.m_
_-.,_,yy.,.,
b i
~' l Figure Q6.54 Small LOCA 3 Inch Diameter Break j
Temperature Versus Time 1
600 i
i i
i i
I 1
I i
ir 400 n
i u.
1 o
i w
2 1
Ad nWv La0H I
i 200 1
i 1
i i
l i
t L
1 I
I i
i 0
2000 4000 6000 TIME (Seconds) sfg.
1 1
6 Q6. 54-12 m____._
Figure Q6.54-9 Small LOCA 3 Inch Diameter Break
+
y-Pressure Versus Time i
2400 1
1600 n
3 m6 E
o mm i
o bid De 800
[
i I
i l
i I
i j
0 2000 40C0 6000 l
TIME (Seconds) 3,*-yr I
/
s Q6. 54-13 l
6 Figure Q6.54-10.
Small LOCA 4 Inch Diameter Break Temperature Versus Time 600
=
4 i
400 C
t 0
g-l l.
0, C.
i eoe 200 i
i l
I i
1 O
1000 2000 3000 TIME (Seconds)
[9 _
Q6.54-14 f
r
..w-,
---.---my-w eiw-- - - <- - + - - - -- - - - - - - - - - - -
- - + - - - - - - - - -
Figure Q6.54-ll 1
l Small LOCA 4 Inch Diameter Break
,x
_s g.
4 Pressure Versus Time 2400 A
i 1600 nn w
u:
l e
b ro vi t'
b.
De 800 i
E I
I I
l l
0 1000 2000 3000 TIME (Seconds)
Fr.f.
/
Q6.54-15 i
Figure Q6.54-12 Small LOCA S
6 6 Inch Diameter Break p,
a.
Temperature Versus Time 600.
L' 400 CA.eD ow 5
~
Oo C.
EUH 200 i
~
)
t I
I l
l l
3 0
200 ou 600 7114E (Seconds) 6.
s T-l Q6. 54-16 l
l i
i
(
r
Figure Q6.54-13 Small LOCA 6 Inch Liameter Break 4
+
P 2400 Pressure Versus Time 1600 A
.3 eW S
E K
800 l
1 I
i 1
0 200 400 600 TIME (Seconds) hp q6.54-17 I
I l