ML20246K224
| ML20246K224 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 05/05/1989 |
| From: | Black S Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20246K229 | List: |
| References | |
| NUDOCS 8905180006 | |
| Download: ML20246K224 (69) | |
Text
_ _ - - _ _ _
h o
UNITED STATES g
.[
g NUCLEAR REGULATORY COMMISSION E
WASHINGTON. D. C. 20555 5
?
hy * * * * *,o L
7 1
J
)
TENNESSEE VALLEY AUTHORITY I
D00)(ET NO. 50-327 0
SEQUOYAH NUCLEAR PLANT, UNIT 1 i
AMENDMENT TO FACILITY OPERATING LICENSE
~
Amendment No. 114 License No. DPR-77 1.
The Nuclear Regulatory Commission (the Comission) has found that:
A.
The application for amendment by Tennessee Valley Authority (the licensee) dated June 24, 1987, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act),
and the Commission's rules hnd regulations set forth in 10 CFR Chapter I; B.
The facility will operate in confor;nity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with tne Commission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
h$0"$$$$ $$
7 P
9 2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 2.C.(2) of Facility Operating License No. DPR-77 is hereby amended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 114, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of its date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION i
Suzanne istant Director for Projects TVA Projects Division Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: May 5, 1989 l
1 ATTACHMENT TO LICENSE AMENDMENT NO. 114 FACILITY OPERATING LICENSE NO. DPR-77 DOCKET NO. 50-327
> Revise the Appendix A Technical Specifications by removing the pages identified below and inserting the enclosed pages. The revised pages are identified by the captioned amendment number and contain marginal lines indicating the area of change. Overleaf pages* are provided to maintain document completeness.
REMOVE INSERT 2-7 2-7 2-9 2-9 8 2-1 B 2-1 B 2-2 B 2-2*
3/4 1-13 3/4 1-13*
3/4 1-14 3/4 1-14 3/4 1-21 3/4 1-21 3/4 3-13 3/4 3-13 3/4 3-56 3/4 3-56 3/4 3-73 3/4 3-73 3/4 7-1 3/4 7-1 3/4 7-2 3/4 7-2*
3/4 7-5 3/4 7-5*
3/4 7-6 3/4 7-5 3/4 7-9 3/4 7-9*
3/4 7-10 3/4 7-10 3/4 7-37 3/4 7-37 3/4 8-4 3/4 8-4 3/4 8-5 3/4 8-5 3/4 8-6 3/4 8-6*
3/4 11-12 3/4 11-12 3/4 12-1 3/4 12-1 3/4 12-2 3/4 12-2 3/4 12-10 3/4 12-10 B 3/4 6-3 8 3/4 6-3 D 3/4 6-3a B 3/4 6-4 8 3/4 6-4*
5-1 5-1*
~
5-2 5-2 5-5 5-5 5-6 5-6 l
1 l
l 1
es l t
}
un
)
pe I
S ml A
E I a
(
s U
v D
e D
y c
L ei ER DR l
ER f
e A
nu TE EE b
TE s
V i q AW TW a
RO 2
E r
P RP i
P
)
L ue f
l f
B Tr oL fL p
oL
=
P A
u A
oA p
A l
?
W
%s
~%M M
A
%M P
I, O
1 s 6R
%R 1R
(
L 1 e 3E 9E t
5E 3
T L
r H
H o
H K
A A
<F
<T
>T N
<T S
+
r T
o N
]
f I
O
'T r
P o
T
)
t E
b a
S 4
s e
I n
P s
1 e
I l t p
R un m
)
T pe 1
o d
ml
(
c e
N I a T
R u
O v
D D
e D
[
g E n
I T
ei ER ER l
ER
)
a W i
T N
nu TE TE b
TE b b l
O t
A I
iq AW AW a
AW 2 3 P
n T
O bE RO RO c
RO T
T e
o N
P r.
P P
i P
A h L C
E T
ue f
f l
f
- +
t A
(
M E
Tr oL oL p
oL d
M U
S u
A A
p A
1 e
n R 1
R
%s
%M
%M A
%M
(
r i E T
P 0s 5R 0R 0R N
u H
2 S
I 1 e 3E 1E t
5E O
2 s
d T N
R r
H H
o H
I K
a e
2 I
T
<F
<T
>T N
<T T
e z D A
m i E E
M T
l T
L E
O y
n i
A B
T N
K o
t R A
S
{
u T
Y r
t S
s g
o s a p
r r
T t
t P
i o
,o A
a n T I
r wt et s
a A R
T oc t c n
t T
L a aa e
s d r
e
)i e
)
p n e R
eo
)R 0dR j
S m
o t O
rt 8
1 e S e y
o c a T
uc d
- m)
- n T
c c
C sa P n P rt Pi 1
e i
A se
( a
( en (b
+
g m d E
eR ti r
a i n R
r
- no
- ur l
L T
I P r w
I p T e
(
e xo x
t x
w
=
=
=
rw ul u
,e uro T
eo l F l es l oP A
3 bP F
F c Ff m
p rw d
e aw no nuo npe r
h o oo ool oit u
1.
CL rL rS(
rra t
1T t
t 4
tTR a
1 T A eo ut uf e u
r st en eog P
er%
e Na N
n No0 p
l ut l
k a p
t5 m
pu eo ecR i
ec e
T mp go go r
gaw t
e I
I n7 nC nl r T
neo r
r N
I -
a aB e aRl e
e U
e P
R r R
w r
R e
v h
n) o eo o
sB O
w L
i3k rt rl Ps t
rk A
b1 c ecp eb p
c ecp N
r-o wai wadi a
woi O
uPl oer onnr e
ol r I
PBT 1
T C
E N
T U
1 2
3 4
5 O
F 2
2 2
2' 2
N yE85x ' Eq -
7" N "'
P
j l
R R
e E
E g
- W W
a tO O
}
r eP tP
)
e s
e I
v c
L sL O
a i
pA A
(
m iM pM 2
g a
rR iR f
n n
tE rE i
y H
tH s
d TT T
a A
T D
AD 9
e g
r v
eE E
T c
a hT eT e
T tA hA d
,R tR r
)
r o
tt
,t 5
o f
na ta 4
f e
n r
ce ee 1
0 e
ru cu
+
l el rl d
l pa ea 1
n o
v pv
(
a r
S 3
T t
T 2s 5s
[
e n
N
- t
+t 6
r o
I i
i K
u c
O s
s t
P df df a
g T
eo eo T
r a
E e
e e
l S
ct ct
)
p xn xn 5
m e
P ee ee 4
e t
I c
c t
a R
) r
) r r
)
T N
N
+
e d
qp qp g
e e
N
)
1 a
h u
O d
0 6
(
r t
n I
e 5
8 e
i T
u b
b
)
v y
t A
n q1 qO S
a b
n T
i
(
(
S o
N t
y y
T g
d C
E n
fb fb 1
1 1
n e
(
M o
o o
+
i t
U C
d d
e e
e s
a 1
R
(
ee ee 1
t t
t a
r T
dc dc
(
o o
o e
e 2
S N
uu uu N
N N
r n
N O
td td 5
c e
2 I
I ie ie K
n n
n n
g T
nr nr i
i i
i E
M A
g g
nn L
E T
ay ay d
d d
re oo B
T O
ml ml 4
e e
e or ii A
S N
l l
K n
n n
f u tt T
Y ea ea
{
i i
i t
ca S
hc hc f
f f
Fa ns ti ti g
e e
e 7
r un P
t t
T d
d d
8
/e f e I
ta ta A
0 2p p
R am am s
s s
- 0. e em m
T h o h o 5
a a
a 1
h o tt tt 0t Tc R
u u
)
O ta ta S
=
=
=
5
=
T n
n y
=
C ee ee t
A cb cb 1
E r
r
+
3 3
R el el 1
5 pl pl l
T S
a a
(
1 S
hh hh g
I cs cs T
7 T 4
5 a
a A
y 1
A K
K y
)
et et d
n n
r e
ri ri e
u oo oo w
n f p f p o
i p
e t
)
)
r r
n i
i e
e o
i i
v h
C
(
i O
W
(
(
1 2
E E
T T
O O
N N
gmy5x ' E-[
e
>2 &g y 5.
a l'
1 L
r L
j
\\
2.1 SAFETY LIMITS BASES 2.1.1 REACTOR CORE i
The restrictions of this safety limit prevent overheating of the fuel and possible cladding perforation which would result in the release of fission products to the reactor coolant. ' Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is i
slightly above the coolant saturation temperature.
Operation above the upper boundary of the nucleate boiling regime could result in excessive cladding temperatures because of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coefficient.
DNB is not a directly measurable parameter during operation and.
therefore THERMAL POWER and Reactor Coolant Temperature and Pressure have been related'to DNB through the W-3 correlation.
The W-3 DNB correlation has been developed to predict the DNB flux and the location of DNB for axially uniform and non-uniform heat flux distributions.
The local DNB heat flux ratio, DNBR, defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux, is indicative of the margin to DNB.
The minimum value of the DNBR during steady state operation, normal operational transients, and anticipated transients is limited to 1.30.
This value corresponds to a 95 percent probability at a 95 percent confidence level that DNB will not occur and is chosen as an appropriate margin to DNB for all operating conditions.
The curves of Figure 2.1-1 show the loci of points of THERMAL POWER, Reactor Coolant System pressure and average temperature for which the minimum DNBR is no less than 1.30, or the average enthalpy at the vessel exit is equal to the enthalpy of saturated liquid.
The curves are based on an enthalpy hot channel factor, F f 1.55 and H,
a reference cosine with a peak of 1.55 for axial power shape.
An allowance is included for an increase in F at reduced power based on the expression:
H F H = 1.55 [1+ 0.3 (1-P)]
where P is the fraction of RATED THERMAL POWER SEQUOYAH - UNIT 1 B 2-1 Amendment No. 19, 114 l
e SAFETY LIMITS BASES These limiting heat flux conditions are higher than those calculated for the range of all control rods. fully withdrawn to the maximum allowable control rod insertion assuming the axial power imbalance is within the limits of the f) (Delta I) function of the Overtemperature Delta T trip.
When the axial power imbalance is not within the tolerance, the axial power imbalance effect on the Overtemperature Delta T trips will reduce the setpoints to provide protection consistent with core safety limits.
2.1.2 REACTOR COOLANT SYSTEM PRESSURE The restriction of this Safety Limit protects the integrity of the Reactor Coolant System from overpressurization and thereby prevents the release of radionuclides contained in the reactor coolant from reaching the containment atmosphere.
The reactor pressure vessel and pressurizer are designed to Section III of the ASME Code for Nuclear Power Plant which permits a maximum transient pressure of 110% (2735 psig) of design pressure.
The Reactor Coolant System piping, valves and fittings, are designed to ANSI B 31.1 1967 Edition, which permits a maximum transient pressure of 120% (2985 psig) of component design pressure.
The Safety Limit of 2735 psig is therefore consistent with the design criteria and associated code requirements.
The entire Reactor Coolant System is hydrotested at 3107 psig, 125% of design pressure, to demonstrate integrity prior to initial operation.
2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETP0INTS The Reactor Trip Setpoint Limits specified in Table 2.2-1 are the values at which the Reactor Trips are set for each functional unit.
The Trip Setpoints have been selected to ensure that the reactor core and reactor coolant system are prevented from exceeding their safety limits during normal operation and design basis anticipated operational occurrences and to assist the Engineered Safety Features Actuation System in mitigating the consequences of accidents.
Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint and the Allowable Value is equal to or less than the drift allowance assumed for each trip in the safety analyses.
SEQUOYAH - UNIT 1 B 2-2 Revised 08/18/87
3 REACTIVITY CONTROL SYSTEMS SURVEILLANCE ~ REQUIREMENTS
'4.1.2.6 Each borated water source shall be demonstrated OPERABLE:
a.
At least.once per 7 days by:
1.
Verifying the boron concentration in each water source',
2.
Verifying the contained borated water volume of each water source, and 3.
Verifying the boric acid storage system solution temperature when it is the source of borated water.
b.
At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by verifying.the RWST temperature.
l l
l l
l SEQUOYAH - UNIT 1 3/4 1-13
i REACTIVITY CONTROL SYSTEMS l
3/4.1.3 MOVA8LE CONTROL ASSEMBLIES GROUP HEIGHT LIMITING CONDITION FOR OPERATION 3.1.3.1 All full length (shutdown and control) rods shall be OPERABLE and positioned within i 12 steps (indicated position) of their group step counter demand position.
APPLICABILITY:
MODES 1* and 2*
ACTION:
With one or more full length rods inoperable due to being immovable a.
as a result of excessive friction or mechanical interference or known to be untrippable, determine that the SHUTOOWN MARGIN requirement of Specification 3.1.1.1 is satisfied within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and be in HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
b.
With more than one full length rod inoperable or misaligned from the group step counter demand position by more than i 12 steps (indicated position), be in HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
c.
With one full length rod inoperable due to causes other than addressed by ACTION a, above, or misaligned from its group step counter demand height by more than 12 steps (indicated position), POWER OPERATION may continue provided that within one hour either:
1.
The rod is restored to OPERABLE status within the above alignment requirements, 2.
The remainder of the rods in the group with the inoperable rod are aligned to within i 12 steps of the inoperable rod within one hour while maintaining the rod sequence and insertion limit of Figure 3.1-1; the THERMAL POWER level shall be restricted pursuant to Specification 3.1.3.6 during subsequent operation, or 3.
The rod is declared inoperable and the SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is satisfied.
POWER OPERATION may then continue provided that:
- See Special Test Exceptions 3.10.2 and 3.10.3.
SEQUOYAH - UNIT 1 3/4 1-14 Amendment No.ll4
)
l
REACTIVITY CONTROL SYSTEMS CONTROL R00 INSERTION LIMITS LIMITING CONDITION FOR OPERATION 3.1.3.6 The control banks shall be limited in physical insertion as shown in Figure 3.1-1.
APPLICABILITY:
MODES 1* and 2*#.
ACTION:
With the control banks inserted beyond the above insertion limits, except for surveillance testing pursuant-to Specification 4.1.3.1.2, either:
a.
Restore the control banks to within the limits within two hours, or b.
Reduce THERMAL POWER within two hours to less than or equal to that fraction of RATED THERMAL POWER which is allowed by the group position using the above figure, or l
c.
Be in H0T STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
SURVEILLANCE REQUIREMENTS 4.1.3.6 The position of each control bank shall be determined to be within the insertion limits at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> except during time intervals when the Rod Insertion Limit Monitor is inoperable, then verify the individual rod positions at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
^See Special Test Exceptions 3.10.2 and 3.10.3.
- With Keff greater than or equal to 1.0.
SEQUOYAH - UNIT 1 3/4 1-21 Amendment No. 41,114
TABLE 4.3-1 (Continued)
NOTATION With the reactor trip system breakers closed and the control rod drive system capable of rod withdrawal.
(1)
If not performed in previous 7 days.
(2)
Heat balance only, above 15% of RATED THERMAL POWER.
Adjust channel if absolute difference greater than 2 percent.
Compare incore to excore AXIAL FLUX DIFFERENCE above 15% of RATED l
(3)
THERMAL POWER.
Recalibrates if the absolute difference greater than or equal to 3 percent.
(4)
Manual ESF functional input check every 18 months.
(5)
Each train or logic channel shall be tested at least every 62 days on a STAGGERED TEST BASIS.
The test shall independently verify the OPERABILITY of the undervoltage and automatic shunt trip circuits.
l (6)
Neutron detectors may be excluded from CHANNEL CALIBRATION.
(7)
Below P-6 (Block of Source Range Reactor Trip) setpoint.
(8)
Logic only, each startup or when required with the reactor trip system breakers closed and the control rod drive system capable of rod withdrawal if not performed in previous 92 days.
The CHANNEL FUNCTIONAL TEST shall independently verify the operability l
(9) of the undervoltage and shunt trip circuits for the manual reactor trip function.
(10) -
Local manual shunt trip prior to placing breaker in service.
Each train shall be tested at least every 62 days on a STAGGERED TEST l
BASIS.
(11) -
Automatic and manual undervoltage trip.
SEQUOYAH - UNIT 1 3/4 3-13 Amendment No. 54, 114
mi 1
l^
i 1
7-de f
v o
omc1.
~
ei rft t n
~
rce oem tpe dSl r
.p ehp r r r pau o o o o yS t t t t
o a a a n
euo r r r a
v qt e e e e r
l e n n n n d
aSe i e e e a
v s
l g g g u
l n SE q
ei o MLL m m m m e e e atp UEB a a a a v v v e
ns MNA e e e e l
l l
r ue INR t t t t a a a o
o r
NAE s s s s v v v c
td IHP
/ / / /
/ //
/
eA MCO 1 1 1 1 1 1 1 1 1 1 0 1 1 1 1 2 1 rtV enT we ome peh r r r lt o o o hp t t t t
tmn a a a n
.iii r r r a
nw N
e e e e r
o eo O
OS n n n n d
inbt I
NL i e e e a
to T
E l
g g g u
iitd A
DN q
stoe T
EN m m m m e e e oi n t N
RA a a a a v v v e
ps t
E IH e e e e l
l l
r oli 1
UC t t t t a a a o
d pl m tU Q
s s s s v v v c
e i m R
EF
/ / / /
/ / /
/
sdwo T
RO 2 2 2 2 2 2 2 1 1 1 1 2 2 2 2 4 2 oe c
0 S
l st 1
N cons I
l ea 3
ecm G
h em 3
N tere I
hit s
E R
r ntus r
L O
o r
i qy o
B T
t o
nes t
A I
i t
si r i
T N
n a
i s
n O
o c
m dei o
M M
i e
eech m
d t
vint T
n n
s lf a c
N i
I y
ail f i
E
)
g S
vrl o t
D
)
e r
n ei s
I e
)
g a
- o
)
n kvee u
C g
e n M r i e
o c
vs o
C
)
l n
g a o t g
i osru c
A
)
e
)
e a n R g t i n
t li u a
e g e v R a
n a s a
a b
se g n g e R w i c o r R t
e h
e n a n L e
)
o l
i P o n
dvdt h
a R a d e e r e o d t
e e
el n t
R R k i g d r t o n e a d m
t aar e
n W n i a a c I v c i u
av o
e e d e a
(
a W N R b l
i W
r i
nf b
d i d T R
( (
u n ad
(
t cko W
i e
w S o V n s
ocid l
iW ( W e r e o
i I
l n
sot e l
k e
I sl ap a
(
(
g u d l Mt d
a s i l
l F
i c n v s abco h
t l e r s W e e s o o or o e
(
e v v r -
o 1
'o e e l
il s
L l
e eef e H Cu t r r e e e yP 9 t p v hhiv l
T T s S P l
u L L t S r u e ttce e
s e s a
V i e o L ed n
t t e r t v s r r w t R e '
t c ff p n
n n r e n e e o o d n O F c a o l
ii S e a
a a P t a L r t t e a P P W m e r
h l
l a
l P a a e l
e r s eel a c
o o t W o r r r F o r r v t e s ll a o o n o e e e e o e e l
n h e
bbcs e
C C e g C z n n n yC z z a e T V aain
. n m n i i e e r i i V rr ccno7 o r r n i r r L G G a r r r n e r iihi3 T
o o i l
o u i o u u yi r o l l ct7 t N
t t a e t s m m ml t
s s t a o t ppec0 s E
c c t u c s a a a i c s s e t C c p pT u -
a M
a a n f
a e e e e x a e e f n
a aa rG e U
e e o e e r t t t u e r r a o n e stE l
R P R C R R P S S S A R P P S C I
R tti sR T
oohnU t S
NNTI N A N
. 0 1 2 3 4 5 6 7 I
1 2 3 4 5 6 7 8 9 1 1 1 1 1 1 1 1 M@SE. ci]
$ YE y E%g" p- @; 1 %
i' l'
lilll!
l
4
. j i
' I'[.
TABLE 4.3-8 (Continued) 1 TABLE NOTATION as' During liquid additions to tne tank.
(1) The CHANNEL FUNCTIONAL TEST shall also demonstrate that automatic isolation of this pathway and control room alarm annunciation occurs if any of the following conditions exists:
1.
Instrument indicates measured levels above the alana/ trip setpoint.
2.
Circuit failure.
3.
Downscale failure.
(2) The CHANNEL FUNCTIONAL TEST shall also demonstrate that control room alarm annunciation occurs if any of the following conditions exists:
1.
Instrumer t indicates measured levels above the alarm setpoint.
2.
Circuit failure.
3.
Downscale failure.
(3) The initial CHANNEL CALIBRATION shall be performed using one or more of-the reference standards certified by the National Bureau of Standards or using standards that have been obtained from suppliers that participate in measurement assurance activities with NBS.
These standards shall permit calibrating the system over its intended range of energy and measurement range.
For subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration shall be used.
(4) CHANNEL CHECK shall consist of verifying indication of flow during periods of release.
CHANNEL CHECK shall be made at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> on days on which continuous, periodic, or batch releases are made.
l (5) The CHANNEL FUNCTIONAL TEST shall also demonstrate that automatic isola-tion of this pathway and control room alarm annunciation occurs if any of the following conditions occur:
1.
Instrument indicate;.aeasured levels above the alarm / trip setpoint.
2.
Circuit failure.
The CHANNEL FUNCTIONAL TEST shall also demonstrate that control room annunciation occurs if the following cnndition occurs:
1.
Downscale failure.
SEQUOYAH - UNIT 1 3/4 3-73 Amendment No. 13, 57, 114 i
m yu
l
.4 3/4.~7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE SAFETY VALVES i
l LIMITING CONDITION FOR OPERATION
\\
1' l
3.7.1.1 All main steam 'line code safety valves associated with each steam generator shall be OPERABLE with lift settings as specified in Table 3.7-4.
APPLIC, ABILITY:
MODES 1, 2 and 3.
ACTION:
a.
With 4 reactor coolant loops and associated steam generators in operation and with one or more main steam line code safety valves inoperable, operation in MODES 1, 2 and 3 may proceed provided, that within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, either the inoperabic valve is restored to OPERABLE status or the Power Range Neutron Flux High Setpoint trip is reduced per Table 3.7-1;.othorwise, be in at least HOT STANDBY within the next 6. hours and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
b.
With 3 reactor coolant loops and associated steam generators in operation and with one or more main steam line code. safety valves associated with an operating loop inoperable, operation in MODE 3 may proceed provided, that within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, either the inoperable valve is restored to OPERABLE status or the Power Range Neutron Flux High Trip Setpoint trip is reduced per Table 3.7-2; otherwise, be-in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
c.
The provisions of Specification 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.7.1.1 No additional Surveillance Requirements other than those required by Specification 4.0.5.
SEQUOYAH - UNIT 1 3/4 7-1 Amendment No.114
)
R
.E e
W gtO nnP ai M
R oL A
pA E
rtM T
eeR S
wSE o
H E
PhT L
g B
eiD 7
5 3
A lHE 8
6 4
R b
T E
axA P
wuR O
ol N
l Ff I
l o
An-H ot TN mrn IO ut e WI muc T
ier TA xNe NR a
P IE M
(
OP PO TEP SOO HL
-G 1
I4 H
7 G
XN 3
UI LR E
FU L
D B
N A
OS T
RE TV UL EA r
NV o
t EY a
aG R
S EE m
WN ea OI l e PL bt aS E
r L
eg B
pn A
oi 1
2 3
W nt O
I a L
r L
f e A
op O
M r
U ey M
b n I
mA X
u A
Nn M
o mus me ivxl aa MV mEggx a c$
R.
~
l l
~..
PLANT SYSTEMS AUXILIARY FEECWATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.1.2 At least three independent steam generator auxiliary feedwater pumps and associated flow paths shall be OPERABLE v'th:
i a.
Two motor-driven auxiliary feedwater pumps, each capable of being i
powered from separate shutdown boards, and b.
One steam turbine-driven auxiliary feedwater pump capable of being powered from an OPERABLE steam rupply syste.n.
APPLICABILITY: MODES 1, 2 and 3.
ACTION:
a.
With one auxilirey feedwater pump inoperable, r? store the required auxiliary feedwater pumps to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STAND 8Y within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
b.
With two auxiliary feedwater pumps inoperable, be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
c.
With three auxiliary feedwater pumps inoperable, immediately initiate corrective action to restore at least one auxiliary feedwater pump to OPERABLE status as soon as possible.
SURVEILLANCE REQUIREMENTS 4.7.1.2 In addition to the requirements of Specification 4.0.5 each auxiliary feedwater pump shall be demonstrated OPERABLE by :
a.
Verifying that:
1.
each motor-driven pump develops a differential pressure of greater than or equal to 1397 psid on recirculation flow.
2.
the steam-turbine driven pump develops a differential pressure of greater than or equal to 1183 psid on recirculation flow when the secondary steam supply pressure is greater than 842 psig.
The provisions of Specification 4.0.4 are not applicable for entry into MODE 3.
SEQUOYAH - UNIT 1 3/4 7-5 Amendment No. 12
PLANT SYSTEMS SURVEILLANCE b !REMENTS (Continued)
~
3.
at least once per 31 days, each automatic control valve in the l
flow path is OPERABLE whenever the auxiliary feedwater system is placed in automatic control or when above 10% of RATED THERMAL POWER.
b.
At least once per 18 months during shutdown
- by:
1.
Verifying that each automatic valve in the flow path actuates to its correct position upon receipt of an auxiliary feedwater actuation test signal and a low auxiliary feedwater pump suction pressure test signal.
2.
Verifying that each auxiliary feedwater pump starts as designed automatically upon receipt of an auxiliary feedwater actuation test signal.
c.
At least once per 7 days by verifying that each non-automatic valve in the auxiliary feedwater system flowpath is in its correct position.
i
- The provisions of Specification 4.0.4 are not applicable for entry into MODE 3 for the turbine-driven Auxiliary Feedwater Pump.
SEQUOYAH - UNIT 1 3/4 7-6 Amendment No. 12,77, 114
L Ot JL ;
4 TABLE 4.7-2 SECONDARY COOLANT SYSTEM SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM TYPE OF MEASUREMENT-SAMPLE AND ANALYSIS AND ANALYSIS FREQUENCY 1.
Gross Activity Determination
- At least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> 2.
Isotopic Analysis for DOSE a) 1 per 31 days, when-EQUIVALENT I-131' Concentration ever the gross activity determination indicates iodine concentrations greater than 10% of the allowable limit.
b) 1 per 6 months, when-ever the gross activity determination indicates iodine concentrations below 10% of the allow-able limit.
SEQUOYAH - UNIT 1 3/4 7-9
c..
PLANT SYSTEMS MAIN STEAM LINE ISOLATION VALVES LIMITING CONDITION FOR OPERATION j
3.7.1.5 Each main steam line isolation valve shall be OPERABLE.
APPLICABILITY:
MODES 1, 2 and 3.
I ACTION:
MODE 1 - With one main steam line isolation valve inoperable, POWER OPERATION may continue provided the inoperable valve is either restored to i
OPERABLE status or closed within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />; Otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in I
HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
MODES 2 - With one main steam line isolation valve inoperable, subsequent and 3 operation in MODES 2 or 3 may proceed provided:
l a.
The isolation valve is maintained closed; b.
The provisions of Specification 3.0.4 are not applicable.
{
\\
Otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and'in j
HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
SURVEILLANCE REQUIREMENTS 4.7.1.5 Each main steam line isolation valve shall be demonstrated OPERABLE by verifying full closure within 5 seconds when tested pursuant to l
Specification 4.0.5.
l l
l
)
SEQUOYAH - UNIT 1 3/4 7-10 Amendment No.ll4
PLANT SYSTEMS FIRE HOSE STATIONS LIMITING CONDITION FOR OPERATION 3.7.11.4 The fire hose stations shown in Table 3.7-5 shall be OPERABLE.
APPLICABILITY:
Whenever equipment in the areas protected by the fire hose stations is required to be OPERABLE.
ACTION:
a.
With one or more of the fire hose stations shown in Table 3.7-5 inoperable, route an additional equivalent capacity fire hose to the unprotected area (s) from an OPERABLE hose station within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> if the inoperable fire hose is the primary means of fire suppression; otherwise route the additional hose within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Restore the fire hose station to OPERABLE status within 14 days or, in lieu of any other report required by Specification 6.6.1, prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within the next 30 days outlining the action taken, the cause of the inoperability, and the plans and schedule for restoring the station to OPERABLE status.
b.
The provisions of Specifications 3.0.3 and 3.0.4 are not applicablec SURVEILLANCEREQUIPJMENTS 4.7.11.4 Each of the fire hose stations shown in Table 3.7-10 shall be demonstrated OPERABLE:
a.
At least once per 31 days by visual inspection of the stations accessible during plant operations to assure all required equipment is at the station.
b.
At least once per 18 months by:
1.
Visual inspection of the stations not accessible during plant operations to assure all required equipment is at the station, 2.
Removing the hose for inspection and re-racking, and 3.
Inspecting all gaskets and replacing any degraded gaskets in the couplings.
c.
At least once per 3 years by:
1.
Partially opening each hose station valve to verify valve OPERABILITY and no flow blockage.
2.
Conducting a hose hydrostatic test at a pressure of 150 psig or at least 50 psig above maximum fire main operating pressure, whichever is greater.
SEQUOYAH - UNIT 1 3/4 7-37 Amendment No. 36,114
_A
4
. =,
ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)-
4.
Simulating a loss of offsite power by itself, and:
a)
Verifying de-energization of the shutdown boards and load shedding from the shutdown boards.
b)
Verifying the diesel starts on the auto-start signal, energizes the shutdown boards with permanently connected loads within 10 seconds, energizes the auto-connected shutdown loads through the load sequencers and operates for greater than or equal to 5 minutes while its generator is loaded with the shutdown loads.
After energization, the steady state voltage and frequency of the shutdown boards shall be maintained at 6900 690 volts and 60 1 1.2 Hz during this test.
5.
Verifying that on a ESF actuation test signal (without loss of offsite power) the diesel generator starts on the auto-start signal and operates on standby for greater than or equal to 5 minutes.
The generator voltage and frequency shall be 6900 i 690 volts and 60 i 1.2 Hz within 10 seconds after the auto-start signal; the steady state generator voltage and fre-quency shall be maintained within these limits during this test.
6.
Simulating a loss of offsite power in conjunction with an ESF actuation test signal, and a)
Verifying de-energization of the shutdown boards and load shedding from the shutdown boards, b)
Verifying the diesel starts on the auto-start signal, energizes the shutdown boards with permanently connected loads within 10 seconds, energizes the auto-connected emergency (accident) loads through the load sequencers and operates for greater than or equal to 5 minutes while its generator is loaded with the emergency loads.
After energization, the steady state voltage and frequency of the emergency busses shall be maintained at 6900 690 volts and 60 1 1.2 Hz during this test.
c)
Verifying that all automatic diesel generator trips, except engine overspeed and generator differential, are automatically bypassed upon loss of voltage on the shutdown board and/or safety injection actuation signal.
7.
Verifying the diesel generator operates for at least 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
During the first 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of this test, the diesel generator shall be loaded to greater than or equal to 4840 kw and during the remaining 22 hours2.546296e-4 days <br />0.00611 hours <br />3.637566e-5 weeks <br />8.371e-6 months <br /> of this test, the diesel generator shall j
be loaded to greater.than or equal to 4400 kw.
1 SEQUOYAH - UNIT 1 3/4 8-4 Amendment No. 49, 64 114
i.
ELECTRICAL POWER SYSTEMS
' SURVEILLANCE REQUIREMENTS (Continued)
The generator voltage and frequency shall be 6900 690 volts and 60 1 1.2 Hz within 10 seconds after the start signal; the steady state generator voltage and frequency shall be maintained within these limits during this test.
Within 5 minutes after completing this 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> test, perform Specification 4.8.1.1.2.d.4.b.
8.
Verifying that the auto-connected loads to each diesel generator do not exceed the continuous rating of 4400 kW.
9.
Verifying the diesel generator's capability to:
a)
Synchronize with the offsite power source while the genera-tor is loaded with its emergency loads upon a simulated restoration of offsite power, b)
Transfer its loads to the offsite power source, and I
c)
Be restored to its shutdown status.
10.
Verifying that the automatic load sequence timers are OPERABLE with the setpoint for each sequence timer within + 5 percent of its design setpoint.
11.
Verifying that the following diesel generator lockout features prevent diesel generator starting only when required:
a)
Engine overspeed b) 86 GA lockout relay e.
At least once per 10 years or after any modifications which could affect diesel generator interdependence by starting the diesel generators simultaneously, during shutdown, and verifying that the i
diesel generators accelerate to at least 900 rpm in less than or equal to 10 seconds, f.
At least once per 10 years
- by:
1.
Draining each fuel oil storage tank, removing the accumulated sediment and cleaning the tank using a sodium hypoclorite solution, and i
2.
Performing a pressure test of those portions of the diesel fuel oil system design to Section III, subsection ND of the ASME Code at a test pressure equal to 110 percent of the system design pressure.
- These requirements are waived for the initial surveillance.
SEQUOYAH - UNIT 1 3/4 8-5 Amendment No. 52,64,109, 114
?.'
ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 4.8.1.1.3 The 125-volt D.C. distribution panel,125-volt D.C. battery bank and associated charger for each diesel generator shall be demonstrated OPERABLE:
a.
At least once per 7 days by verifying:
1.
That the parameters in Table 4.8-la meet the Category A limits.
2.
That the total battery terminal voltage is greater than or
. equal to 124-volts on float charge.
b.
At least once per 92 days by:
1.
Verifying that the parameters in Table 4.8-la meet the Category B limits, 2.
Verifying there is no visible corrosion at either terminals or i
connectors, or the cell to terminal connection resistance of these items is less than 150 x 10 6 ohms, and 3.
Verifying that the average electrolyte temperature of 6 connected cells is above 60 F.
c.
At'1 cast once per 18 months by verifying that:
1.
The cells, cell plates and battery racks show no visual indication of physical damage or abnormal deterioration.
2.
The battery to battery and terminal connections are clean, tight and coated with anti-corrosion material.
3.
The resistance of each cell to terminal connection is less than or equal to 150 x 10 6 ohms.
4.8.1.1.4 Reports - All diesel generator failures, valid or non-valid, shall be reported to the Commission pursuant to Specification 6.9.2.2.
SEQUOYAH - UNIT 1 3/4 8-6 Amendment No. 52
TABLE 4.11-2 (Continued)
TABLE NOTATION a.
The LLD is defined, for the purp.oses of these specifications, as the smallest concentration of radioactive material in a sample that will yield a net count above system background that will be detected with 95% proba-bility with only a 5% probability of falsely concluding that a blank-observation represents a "real" signal.
For a particular measurement system (which may include radiochemical separation):
4.66 sb bb0
- E V
2.22x10*
Y exp (-A.at)
Where:
LLD is the "a priori" lower limit of detection as defined above in microcurie per unit mass or volume, s
is the standard deviation of the background counting rate or of h
tne counting rate of a blank sample as appropriate (as counts per I
minute),
E is the counting efficiency as counts per disintegration, V is the sample size in units of mass or volume, 2.22 x 106 is the number of disintegrations per minute per microcurie, Y is the fractional radiochemical yield (when applicable),
A is the radioactive decay constant for the particular radionuclides, and at is the elapsed time between midpoint of sample collection and time of counting (midpoint).
It should be noted that the LLD is defined as an a priori (before the fact) limit representing the capability of a measurement system and not as an a postet>ori (after the fact) limit for a particular measurement.
SEQUOYAH - UNIT 1 3/4 11-12 Amendment No. 42, 114
k j
L 3/4.12 ' RADIOLOGICAL ENVIRONMENTAL MONITORING
{
3/4.12.1 MONITORING PROGRAM LIMITING CONDI1 ION FOR OPERATION I
3.12.1 The radiological environmental monitoring program shall be conducted-as specified in Table 3.12-1.
APPLICABILITY:
At all times.
ACTION:
I a.
With the radiological environmental monitoring program not being con-ducted as specified in Table 3.12-1, in lieu of a LER, prepare and submit to the Commission, in the Annual Radiological Environmental Operating Report, a description of the reasons for not conducting the program as required and the plans for preventing a recurrence.
b.
With the level of radioactivity in an environmental sampling medium exceeding the reporting levels of Table 3.12-2 when averaged over any calendar quarter, in lieu of any other report required by Specifica-tion 6.9.1, prepare and submit to the Commission within 30 days from the end of the affected calendar quarter a Report pursuant to Speci-fication 6.9.2, a special report that identifies the cause(s) for ex-ceeding the limit (s) and defines the corrective actions to be taken to reduce radioactive effluents so that the potential annual dose to a member of the public is less than the calendar year limits of Specifications 3.11.1.2, 3.11.2.2, and 3.11.2.3.
When more than one of the radionuclides in Table 3.12-2 is detected in the sampling medium, this report shall be submitted if:
concentration (1), concentration (2) + *> 1*0 limit level (1) limit level (2) i When radionuclides other than those in Table 3.12-2 are detected and are the result of plant effluents, this report shall be submitted if the potential annual dose to a member of the public is equal to or greater than the calendar year limits of Specifications 3.11.1.2, 3.11.2.2 and 3.11.2.3.
This report is not required if the measured level of radioactivity was not the result of plant effluents; however, in such an event, the condition shall be reported end described in the Annual Radiological Environmental Operating Report.
c.
With milk or fresh leafy vegetable samples unavailable from one or j
more of the sample locations required by Table 3.12-1, identify loca-tions for obtaining replacement samples and add them to the radio-logical environmental monitoring program within 30 days.
The specific locations from which samples were unavailable may then be deleted from the monitoring program.
In lieu of a licensee event report (LER) i and pursuant to Specification 6.9.1.7, identify the cause(s) of the unavailability of samples and identify the new locations for obtaining SEQUOYAH - UNIT 1 3/4 12-1 Amendment No. 42, 134
RADIOLOGICAL ENVIRONMENTAL MONITORING LIMITING CONDITION FOR OPERATION (Continued) replacement samples in the Annual Radiological Environmental Operat-ing Report.
A revised figure (s) and table (s) for the ODCM reflecting the new location (s) shall be included in the next semiannual radio-active effluent release report pursuant to Specification 6.9.1.9.
d.
The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.12.1 The radiological environmental monitoring samples shall be collected pursuant to Table 3.12-1 from the locations given in the table and figure in the ODCM and shall be analyzed pursuant to the requirements of Table 3.12-1 and the detection capabilities required by Table 4.12-1.
1 l
l l
SEQUOYAH - UNIT 1 3/4 12-2 Amendment No. 42, 114
i RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.2 LAND USE CENSUS LIMITING CONDITION FOR OPERATION 3.12.2 A Land Use Census shall be conducted and shall identify within a dis-tance of 8 km (5 miles) the location in each of the 16 meteorological sectors of the nearest milk animal the nearest residence and the nearest garden
- of 2
d greater than 50 m (500 ft ) producing fresh leafy vegetables.
APPLICABILITY:
At all times.
[CTION_:
a.
With a Land Use Census identifying a location (s) that yields a calcu-lated dose or dose commitment 20% greater than the values currently being calculated in Specification 4.11.2.3, identify the new loca-tion (s) in the next Semiannual Radioactive Effluent Release Report pursuant to Specification 6.9.1.9.
l b.
With a Land Use Census identifying a location (s) that yields a calculated dose or dose commitment (via the same exposure pathway) 20% greater than at a location from which samples are currently being obtained in accordance with Specification 3.12.1, add the new l
location (s) within 30 days to the Radiological Environmental Moni-toring Program given in the ODCM, if samples are available.
The sampling location (s), excluding the control station location, having the lowest. calculated dose or dose commitment (s), via the same expo-sure pathway, may be deleted from this monitoring program after October 31 of the year in which this Land Use Census was conducted.
Pursuant to Specification 6.14, submit in the next Semiannual Radio-active Effluent Release Report documentation for a change in the ODCM including a revised figure (s) and table (s) for the ODCM reflect-ing the new location (s) with information supporting the change in sampling locations.
c.
The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.12.2 The Land Use Census shall be conducted during the growing season at least once per 12 months using that information that will provide the best results, such as by a door-to-door survey, mail survey, telephone survey, aerial survey, or by consulting local agriculture authorities.
The results of the Land Use Census shall be included in the Annual Radiological Environmental Operating Report pursuant to Specification 6.9.1.7.
l
- Broad leaf vegetation sampling of at least three different kinds of vegetation may be performed at the SITE B0UNDARY in each of two different direction sectors with the highest predicted D/Qs in lieu of the garden census.
Speci-fications for broad leaf vegetation sampling in Table 3.12 1.4c. shall be i
followed, including analysis of control samples.
SEQUOYAH - UNIT 1 3/4 12-10 Amendment No.42, II4
CONTAINMENT SYSTEMS BASES 3/4.6.1.8 EMERGENCYGASTREATMENTSYSTEM(EGTSJ The OPERABILITY of the EGTS cleanup subsystem ensures that during LOCA conditions, containment vessel leakage into the annulus will be filtered through the HEPA filters and charcoal adsorber trains prior to discharge to the atmosphere.
This requirement is necessary to meet the assumptions used in i
l the accident analyses and limit the site boundary radiation doses to within the limits of 10 CFR 100 during LOCA conditions.
Cumulative operation of the system with the heaters on for 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> over a 31 day period is sufficient to reduce the buildup of moisture on the absorbers and HEPA filters.
ANSI N510-1975 l
will be used as a procedural guide for surveillance testing.
3/4.6.1.9 CONTAINMENT VENTILATION SYSTEM Use of the containment purge lines is restricted to only one pair (one supply line and one exhaust line) of purge system lines at a time to ensure that the site boundary dose guidelines of 10 CFR Part 100 would not be exceeded in the event of a loss of coolant accident during purging operations.
The analysis of this accident assumed purging through the largest pair of lines (a
- 24. inch inlet line and a 24 inch outlet line), a pre-existing iodine spike in the reactor coolant and four second valve closure times.
3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS 3/4.6.2.1 CONTAINMENT SPRAY SYSTEM l
The OPERABILITY of the containment spray system ensures that containment depressurization and cooling capability will be available in the event of a LOCA.
The pressure reduction and resultant lower containment leakage rate are consistent with the assumptions used in the accident analyses.
3/4.6.2.2 CONTAINMENT C0OLING FANS The OPERABILITY of the lower containment vent coolers ensures that ade-quate heat removal capacity is available to provide long-term cooling following a non-LOCA event.
Postaccident use of these coolers ensures containment tem-peratures remain within environmental qualification limits for all safety-related equipment required to remain functional.
3/4.6.3 CONTAINMENT ISOLATION VALVES The OPERABILITY of the containment isolation valves ensures that the containment atmosphere will be isolated from the outside environment in the event of a release of radioactive material to the containment atmosphere or pressurization of the containment.
Containment isolation within the time limits specified ensures that the release of radioactive material to the environment will be consistent with the assumptions used in the analyses for a LOCA.
By letters dated March 3, 1981, and April 2, 1981, TVA will submit a report on the operating experience of the plant no later than startup after the first refueling.
This information will be used to provide a basis to re evaluate the adequacy of the purge and vent time limits.
SEQUOYAH - UNIT 1 B 3/4 6-3 Amendment No.
67, 114
e,--,,-----e,-,
c i
CONTAINMENT SYSTEMS BASES 3/4.6.4 COMBUSTIBLE GAS CONTROL The OPERABILITY of the equipment and systems required for the detection and control of hydrogen gas ensures that this equipment will be available to maintain the hydrogen concentration within containment below its flammable limit during post-LOCA conditions.
Either recombiner unit or the hydrogen mitigation system, consisting of 68 hydrogen ignitions per unit, is capable of controlling the expected hydrogen generation associated with 1) zirconium-water reactions, 2) radiolytic decomposition of water and 3) corrosion of metals within containment. These hydrogen control systems are designed to mitigate the effects of an accident as described in Regulatory Guide 1.7,
" Control of Combustible Gas Concentrations in Containment Following a LOCA",
revision 2 dated November 1978.
The hydrogen mixing systems are provided to ensure adequate mixing of the containment atmosphere following a LOCA.
This mixing action will prevent localized accumulations of hydrogen from exceeding the flammable limit.
The operability of at least 66 of 68 ignitors in the hydrogen mitigation system will maintain an effective coverage throughout the containment.
This system of ignitors will initiate combustion of any significant amount of hydrogen released after a degraded core accident.
This system is to ensure burning in a controlled manner as the hydrogen is released instead of allowing it to be ignited at high concentrations by a random ignition source.
3/4.6.5 ICE CONDENSER The requirements associated with each of the components of the ice con-denser ensure that the overall system will be available to provide sufficient pressure suppression capability to limit the containment peak pressure tran-sient to less than 12 psig during LOCA conditions.
3/4.6.5.1 ICE BED The OPERABILITY of the ice bed ensures that the required ice inventory will 1) be distributed evenly through the containment bays, 2) contain suffi-cient boron to preclude dilution of the containment sump following the LOCA and 3) contain sufficient heat removal capability to condense the reactor system volume released during a LOCA.
These conditions are consistent with the assumptions used in the accident analyses.
The minimum weight figure of 1200 pounds of ice per basket contains a 10%
conservative allowance for ice loss through sublimation which is a factor of 10 higher than assumed for the ice condenser design.
The minimum weight figure of 2,333,100 pounds of ice also contains an additional 1% conservative allowance to account for systematic error in weighing instruments.
In the Amendment 4, 5 SEQUOYAH - UNIT 1 B 3/4 6-4 Revised 08/18/87
c a-6.
1 5.0 DESIGN FEATURES 5.1 SITE
-EX,CLUSION AREA' 5.1.1 The exclusion area shall be as shown in Figure 5.1-1.
LOW POPULATION ZONE 5.1.2 The low population zone shall be as shown in Figure 5.1-2.
I SITE BOUNDARY FOR GASE0US EFFLUENTS 5.1.3 The site boundary for gaseous effluents shall be as shown in Figure 5.1-1.
SITE B0UNDARY FOR LIQUID EFFLUENTS 5.1.4 The site boundary for liquid effluents shall be as shown in Figure 5.1-1.
5.2 CONTAINMENT CONFIGURATION 5.2.1 The shield building is a reinforced concrete building of cylindrical shape, with a dome. roof around a free standing steel containment and having the following design features:-
a.
Nominal inside diameter = 125 feet.
b.
Nominal inside height = 175 feet.
c.
Minimum thickness of concrete walls = 3 feet.
d.
Minimum thickness of concrete roof = 2 feet.
e.
Minimum thickness of concrete floor pad = 9 feet, f.
Minimum thickness of steel containment = 0.5 inches at the spring line and 0.25 inches at the bottom liner plate.
g.
Net free volume = 375,000 cubic feet between the steel contain-ment and the shield building.
DESIGN PRESSURE AND TEMPERATURE 5.2.2 The steel containment is designed and shall be maintained for a maximum l
internal pressure of 12 psig and a temperature of 250 F.
l 1
SEQUOYAH - UNIT 1 5-1
1I v
es
)
b e
y e
i r
s t
e a
c es n
a e
d e
e t e n
dt a
au ee e
=
eo nt t
e e
rB dP n
o A
ee s
e s
e ea d t ee ei rl Ae tS
.t c
f:l 4,$V-l et i e t e ee rd uu ti tl i
ss ef cf l
et Ra R
Y ru nO 1
U(
/g
~
1 AD
~
3 N
4 e
U r
O es B
i 2
F W
E E
T 2
a W
N I
S O
.U" 3 Z
\\
~
E SA
~
E E
N L
E O
N
.e e
O Z
1 I
E E
T
.e S
M A
u.
O U
L e
I Z
C l
6 E
E L
R S
A A
C I
l E
m*
\\
i a '
O t
S is D
W-I i
2-A BD
(
[
r@
w eg
,ac
\\
t e
a g r.
e.
t n
f F
t 03 Y
L 1
i 1
6 e
A
.2 2
C t1
[4 e.
t I
1 s
G l
e e
\\
t a O
e L
w t
t
\\b ot e
e O
A he v
.n R
nv a
ea h
OR a
h n
EE t n e
e hs ee
\\
TW T) e t
e re vt e
e EO r
td s
o MT t
ea ge v
n e
rr ov e
Ko r
g t
s j
t dg e
e g
st l a
,g s
Ae it d
A L ae 1 ud 2 i 3 t t A bI t
e nN nl p T t
T e T A
a E yu L.
b k,
a C e
I rb
'a
'C l
e sv O e e o e s E oo F 8e P c F s n
i C eb 1t i
e n
u R t a E
t e E vL d A
q Nce s
S st S r S e
t A e A uh A e A a n
L I AS E S t c o
3 t
n 1
u t
t 1
rt 1
s e 5 ee R
2 R
c.
r t v t
f te c
u ul xe o
oe s
n(
mxheo
'hm (
$8SE.
2ak$3" zP g
4 l<
l!1
4 DESIGN FEATURES 5.6 FUEL STORAGE CRITICALITY - SPENT FUEL 5.6.1.1 The spent fuel storage racks are designed for fuel enriched to 4.0 weight percent U-235 and shall be maintained with:
a.
Ak equivalent to less than 0.95 when flooded with unborated eff water, which includes a conservative allowance of 1.42% delta k/k for uncertainties.*
b.
A nominal 10.375 inch center-to-center distance between fuel assemblies placed in the storage racks.
CRITICALITY - NEW FUEL 5.6.1.2 The new fuel pit storage racks are designed and shall be maintained with a nominal 21.0 inch center-to-center distance between new fuel assemblies l
such that k,ff will not exceed 0.98 when fuel having an enrichment of 4.5 weight percent U-235 is in place and optimum achievable moderation is assumed.
New fuel enrichment is limited to 4.0 weight percer; as noted in 5.3.1 and 5.6.1.1.
DRAINAGE 5.6.2 The spent fuel pit is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 722 ft.
CAPACITY 5.6.3 The spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than 1386 fuel assemblies.
5.7 COMPONENT CYCLIC OR TRANSIENT LIMIT 5.7.1 The components identified in Table 5.7-1 are designed and shall be maintained within the cyclic or transient limits of Table 5.7-1.
(
"For some accident conditions, the presence of dissolved boron in the pool water may be taken into account by applying the double contingency principle which requires two unlikely, independent, concurrent events to produce a criticality accident.
SEQUOYAH - UNIT 1 5-5 Amendment No. 13, 60, 114
'0 o
li W
e t
as O
r 0
cn P
e 2
o R
i o f
t E
r L
f 5
m e
W t e A
i o
l F OR ch M
d m
r c
PE et r
R g
g g
o f
y0 W
l o
E e
i i
i t
H r
s s
s r
c5 LO eg f
g 6
AP nm c
T u
p p
p v
n M
.ie a
t.
g a
w>-
RL C yt e
D aF 5
7 6
v TF o
.l s
r E
r 8
0 5
a d m HM Apy T
e0 4
1 3
T 0
l o TR pS e
A p6 2
3 1
0 or E
eu n
R m5 e2 of DH t sl o
e o
o o
f t5 t
t t
l c
ET i
a ET e
c5 s
T sec yp o
.y re AD f ci l m rd d
d d
LN l
CE cF co er RE f rr nu en e
e e
YI y
t zu T
out op 0
t a z
z z
CS c0 n i t.
fA oc a
i i
i N
5 wF raF oR f se ft o
wF r
r r
NA p5 o ur o
l on t
u u
u GR u
d0 se0
%f rE a
y0 s
s s
IT t>T5 s p0 5 o se sl a2 s
s s
S a
o5 em2 1
swF so 0
r3 e
e e
ER eoo re ooS oo 0
p r
r r
DO HtC>
Pt i
>0 L pE L c 1
S>
P P
P S
T I
M I
L TN E
I S
1
.N p
A i
e 7
R r
s t r n
T h
e ut o
5
/
l o
R F
c h r e
n s
s E
O y
t o t
i t
t L
0t c
i t i
s s
B C
0 a wc s
w s
e e
A I
1 n
a f
o e
t t
T L
s w
,e f
l s
l C
o sr o
f e
c e
e Y
l d
e r
l y
r r
C t c l
l r f e f p c
c u
u ay o
co ow oo y
s s
T c
o y
o o
c n
s s
N T
s c
ce sp sl o
e e
E I
en n
s s
p i r
r N
M w
r di ol ot i
t p
p l
O RI co e
ab l a l n r
a P
OL yd zr or c
a t
u s
c c
M cl ih l u fi fl t
t i
i O
CT o
r/
t or oo r
c s
t t
C IN por uF f
t o
o a
e a
a LE uch s
oe sc sc t
t t
t CI t
/
s0 t
ee e
c y
s s
YS a0F e0 sa l l l r a
a k
o o
CN e0 r2 si ce co e
r a
r r
A h20 p
od y
yt r
p e
d d
R 0
i l e c
cc s
l y
y T
0d1 0
m C
a 0
h h
0 n 0t 0 m 0
0 e 0
2 0
2 a<
2 a 8i 4A 8r 4
1 5
5 5
me tsy S
m t
e n
t a
s l
y o
S o
T C
y N
r E
r a
N o
d O
t n
P c
o M
a c
O e
e C
R S
vEg$: rig""
[o[3 5, ll
... parnog
][
k UNITED STATES g
NUCLEAR REGULATORY COMMISSION -
5 j
WASHINGTON, D. C. 206f8
.]
T.,...../
i 1
7 TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-328 SEQUOYAH NUCLEAR PLANT, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE i
Arendment No.104 License No. DPR-79 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Tennessee Valley Authority (the licensee) dated June 24, 1987, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act),
and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the i
provisions of the Act, and the rules and regulations of the 1
- Comission; t
C.
There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D.
The issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public; and i
E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.
i i
. l 2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 2.C.(2) of Facility Operating License No. DPR-79 is hereby i
amended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 104, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of its date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION L&&nL M
/
Suzanne ack, Assistant Director for Projects TVA Projects Division Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: May 5, 1989 i
l-
ATTACHMENT TO LICENSE AMENDMENT N0. 104 FACILITY OPERATING LICENSE NO. DPR-79 DOCKET NO. 50-328 Revise the Appendix A Technical Specifications by removing the pages identified below and inserting the enclosed pages.
The revised pages are identified by the captioned amendment number and contain marginal ifnes indicating the area of change. Overleaf pages* are provided to maintain document completeness.
REMOVE
_ INSERT 2-7 2-7 2-9 2-9 B 2-1 B 2-1 B 2-2 B 2-2*
3/4 1-13 3/4 1-13*
3/4 1-14 3/4 1-14 3/4 1-21 3/4 1-21 3/4 3-3 3/4 3-3*
3/4 3-4 3/4 3-4 3/4 3-7 3/4 3-7 3/4 3-13 3/4 3-13 3/4 3-14 3/4 3-14*
3/4 3-57 3/4 3-57 3/4 6-4 3/4 6-4 3/4 6-18 3/4 6-18 3/4 7-1 3/4 7-1 3/4 7-2 3/4 7-2*
3/4 7-5 3/4 7-5*
3/4 7-6 3/4 7-6 3/4 7-9 3/4 7-9*
3/4 7-10 3/4 7-10 3/4 7-25 3/4 7-25 3/4 7-49 3/4 7-49*
3/4 7-50 3/4 7-50 3/4 9-1 3/4 9-1 3/4 9-2 3/4 9-2*
3/4 11-9 3/4 11-9 3/4 12-2 3/4 12-2 3/4 12-9 3/4 12-9 5-1 5-1*
5-2 5-2 5-5 5-5*
5-6 5-6
jl)l c
?
e s
lt
}
un
)
pe I
S ml 0
s E
I a f
c U
v D
e D
1 e
L ei ER DR l
ER f
s A
nu TE EE b
TE V
iq AW TW a
AW 2
RO E
r P
RP i
P
)
=
L ue f
l f
B Tr oL fL p
oL
'P y
A u
A oA p
A t
W
%s
%M M
A
%M P
O 1 s 6R
%R 1R
(
L 1e 3E 9E t
5E 3
T L
r H
H o
H K
A A
<P
<T
>T N
<T
+
r S
o T
]
f N
'T r
I O
o P
)
t T
S a
E 4
s S
e T
n s
1 e
P l t
+
p I
un m
R pe 1
o
)
T ml
(
c d
I a T
R e
N v
D D
e D
[
g E
u O
T ei ER ER l
ER
)
a W
n I
N nu TE TE b
TE 5
5 l
O i
T I
iq AW AW a
AW 2
3 P
t A
O bE RO RO c
RO T
T T
e n
T P
r P
P i
P A
h L
o N
T ue f
f l
f
+
t A
C E
E Tr oL oL p
oL d
M
(
M S
u A
A p
A 1
e n
R U
%s
%M
%M A
%M
(
r i
E 1
R P
0s 5R 0R 0R N
u H
T I
1e 3E 1E t
5E O
2 s
d T
2 S
R r
H H
o H
I K
a e
N T
<P
<T
>T N
<T T
e z
D 2
I A
m i
E T
l T
E M
O y
n i
A L
E N
K o
t R
B T
{
u A
S r
t T
Y s
o s
a S
p
,r T
t t
i o
A a
n T
P r
w et s
a A
I T
o t c n
t R
L aa e
s d
T r
)i e
)
p n
e eo
)
0dr
)
S m
o t
R rt 8
1 e 9e y
c a
O uc d
- m)
- n t
c c
1 T
sa Pn P rt Pi 1
e i
5 1
C se
( a
( en (b
+
g m
d 1
0 A
eR ti r
i n
E r
- no
- ur 1
L T
I 1
0 R
Pr w
I p Te
(
e xo x
,t x
w
=
=
=
=
rw ul u
e uro T
eo lF l es l oP A
S bP F
Fc Ff y
m p
rw d
e T
aw no nuo npe r
ho oo ool oit u
+
CL rL rS(
rra t
y T y
2 t
t 4
tTR a
1 t
A K
K eo ut uf e u
r st en eog P
er%
e l
Na N
n No0 p
ut l p k a p
t5 m
pu eoi ecR i
ec e
T mp gor go r
gaw t
e I
I n7 nCT nl r T
neo r
r N
I -
a abe aRl e
e U
e P
R rr R
w r
R e
v h
n) oo eo o
sB O
w L
i3k rtt rl P s t
rk A
b1c ecc eb p
c ecp N
r-o waa wadi a
woi O
uPl oee onnr e
ol r I
T(B PRR PEAT R
PBT 1
T C
E N
T U
1 2
3 4
5 O
F 2
2 2
2 2
N mEgEr C5[
[
- le% y g*
ls-
- ljlll!lll l!ll1 l
R R
e E
E g
- W W
a tO O
I r
eP tP
)
e s
e I
v c
L sL O
a i
pA A
(
m iM pM 2
g a
rR iR f
n n
tE rE y
i H
tH s
d TT T
a A
T D
AD 9
e g
r v
eE E
T c
a hT eT e
T tA hA d
,R tR r
)
r o
tt
,t S
o f
na t a 4
f e
n T
r ce ee 1
0 e
S ru cu
+
l T
el rl d
l N
pa ea 1
n o
I v
pv
(
a r
O 9
T t
P 2 s 5s
[
e n
T
- t
+t 6
r o
E i
i K
u c
S s
s t
df df a
g P
eo eo T
r a
I e
e e
l R
ct ct
)
p
)
T xn xn S
m e
d ee ee 4
e t
e N
)
c c
I t
a u
O d
) r
) r 1
r n
I e
N N
e i
T u
qp qp g
e t
A n
1 a
h n
T i
- 0 6
(
r t
o N
t 5
8 e
C E
n f
b
)
v y
(
M o
ql qO b
a b
U C
(
(
5 1
R
(
y y
b 1
g d
T fb fb S
1 1
1 n
e 2
S N
o o
+
i t
N O
d d
e e
e s
a 2
I I
ee ee 1
t t
t a
r T
d c d c
(
o o
o e
e E
M A
uu uu N
N N
r n
L E
T td td 5
c e
B T
O i e i e K
n n
n n
g A
S N
nr nr i
i i
i T
Y g
g nn S
ay ay d
d d
re oo ml ml 4
e e
e or i i P
l l
K n
n n
f u tt I
ea ea
{
i i
i t
ca R
h c h c f
f f
F a ns T
ti ti g
e e
e 7
r un t
t T
d d
d 8
/e f e R
t a t a A
0 2 p p
O am am s
s s
0 m em T
h o h o 5
a a
1
. e h o C
tt tt 0L T c A
u u
)
E t a t a S
=
=
=
5
=
R n
n y
=
ee ee t
cb cb 1
r r
+
g 3
el el 5
pl pl 1
T b
7 a
a
(
1 5
hh hh g
T cs cs T
1 T 4
5 a
a A
y 1
A K
K y
)
et et d
n n
r e
ri ri e
u oo oo w
n f p f p o
i p
e t
)
)
r r
n i
i e
e o
i i
v h
C
(
i O
W
(
(
1 2
E E
T T
O O
N N
$"cEE E*~
]
f*E&@" 5.
l
2.1 SAFETY LIMITS BASES 2.1.1 REACTOR CORE The restrictions of this Safety Limit prevent overheating of the fuel and possible cladding perforation which would result in the release of fission products to the reactor coolant.
Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat. transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature.
Operation above the upper boundary of the nucleate boiling regime could reselt in excessive cladding temperatures because of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coefficient.
DNB is not a directly measurable parameter during operation and therefore THERMAL POWER and Reactor Coolant Temperature and Pressure have been related to DNB through the W-3 correlation.
The W-3 DNB correlation has been developed to predict the DNB flux and the location of DNB for axially uniform and non-uniform heat flux distributions.
The local DNB heat flux ratio, DNBR, defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux, is indicative of the margin to DNB.
The minimum value of the DNBR during steady state operation, normal operational transients, and anticipated transients is limited to 1.30.
This value corresponds to a 95 percent probability at a 95 percent confidence level that DNB will not occur and is chosen as an appropriate margin to DNB for all operating conditions.
The curves of Figure 2.1-1 show the loci of points of THERMAL POWER, Reac-l tor Coolant System pressure and average temperature for which the minimum DNBR I
is no less than 1.30, or the average enthalpy at the vessel exit is equal to the enthalpy of saturated liquid.
These curves are based on an enthalpy hot channel factor, F of 1.55 andareferencecosinewithapegkof1.55foraxialpowershape.g,Anallowance is included for an increase in F at reduced power based on the expression:
3g F H = 1.55 [1+ 0.3 (1-P)]
where P is the fraction of RATED THERMAL POWER These limiting heat flux conditions are higher than those calculated for the range of all control rods fully withdrawn to the maximum allowable control rod insertion assuming the axial power imbalance is within the limits of the (delta I) function of the Overtemperature Delta T trip.
When the axial power f, balance is not within the tolerance, the axial power imbalance effect on the im Overtemperature delta T trips will reduce the setpoints to provide protection consistent with core safety limits.
SEQUOYAH - UNIT 2 B 2-1 Amendment No. 21,104 Revised 08/18/87
j l
~.
SAFETY LIMITS BASES 2.1.2 REACTOR COOLANT SYSTEM PRESSURE The restriction of this Safety Limit protects the integrity of the Reactor Coolant System from overpressurization and thereby prevents the release of radionuclides contained in the reactor coolant from reaching the containment atmosphere.
The reactor pressure vessel and pressurizer are designed to Section III of the ASME Code for Nuclear Power Plant which permits a maximum transient pressure of 110% (2735 psig) of design pressure.
The Reactor Coolant System piping, valves and fittings, are designed to ANSI B 31.1 1967 Edition, which permits a maximum transient pressure of 120% (2985 psig) of component design pressure.
The Safety Limit of 2735 psig is therefore consistent with the design criteria and associated code requirements.
The entire Reactor Coolant System is nydrotested at 3107 psig, 125% of design pressure, to demonstrate integrity prior to initial operation.
2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS The Reactor Trip Setpoint Limits specified in Table 2.2-1 are the values et which the Reactor Trips are set for each functional unit.
The Trip Setpoints have been selected to ensure that the reactor core and reactor coolant system are prevented from exceeding their safety limits during normal operation and design basis anticipated operational occurrences and to assist the Engineered Safety Features Actuation System in mitigating the consequences of accidents.
Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint and the Allowable Value is equal to or less than the drift allowance assumed for each trip in the safety analyses.
I SEQUOYAH - UNIT 2 B 2-2
REACTIVITY CONTROL SYSTEMS i
SURVEILLANCE REQUIREMENTS 4.1.2.6 Each borated water sourcs ehill be demonstrated OPERABLE:
a.
At least once per 7 days by:
1.
Verifying the boron concentration in the water, 2.
. Verifying the contained borated water volume of the water source, and 3.
Verifying the boric acid storage system solution temperature when it is the source of borated water.
b.
At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by verifying the RWST temperature.
l l
I SEQUOYAH - UNIT 2 3/4 1-13
k REACTIVITY CONTROL SYSTEMS 3/4.1.3 MOVABLE CONTROL ASSEMBLIES GROUP HEIGHT LIMITING CONDITION FOR OPERATION 3.1.3.1 All full length (shutdown and control) rods shall be OPERABLE and positioned within i 12 steps (indicated position) of their group step counter F
demand position.
APPLICABILITY:
Modes 1* and 2*.
ACTION:
a.
With one or more full length rods inoperable due to being immovable as a result of excessive friction or mechanical interference or known to be untrippable, determine that the SHUTDOWN MARGIN require-ment of Specification 3.1.1.1 is satisfied within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and be in HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
b.
With more than one full length rod inoperable or misaligned from the group step counter demand position by more than i 12 steps (indicated position), be in H0T STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
c.
With one full length rod inoperable due to causes other than addressed by ACTION a, above, or misaligned from its group step counter demand height by more than i 12 steps (indicated position), POWER OPERATION may continue provided that within one hour either:
1.
The rod is restored to OPERABLE status within the above alignment requirements, or 2.
The remainder of the rods in the group with the inoperable rod are aligned to within + 12 steps of the inoperable rod while maintaining the rod sequence and insertion limit of Figure 3.1-1; the THERMAL POWER level shall be restricted pursuant to Specifi-cation 3.1.3.6 during subsequent operation, or 3.
The rod is declared inoperable and the SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is satisfied.
POWER OPERATION may then continue provided that:
a)
A reevaluation of each accident analysis of Table 3.1-1 is performed within 5 days; this revaluation shall confirm that the previously analyzed results of these accidents remain valid for the duration of operation under these conditions.
- See Special Test Exceptions 3.10.2 and 3.10.3.
SEQUOYAH - UNIT 2 3/4 1-14 Amendment No.104
REACTIVITY CONTROL SYSTEMS CONTROL R0D INSERTION LIMITS LIMITING CONDITION FOR OPERATION 3.1.3.6 The control banks shall be limited in physical insertion as shown in Figure 3.1-1.
APPLICABILITY:
Modes 1* and 2*#.
ACTION:
With the control banks inserted beyond the above insertion limits, except for surveillance testing pursuant to Specification 4.1.3.1.2, either:
Restore the control banks to within the limits within two hours, or a.
b.
Reduce THERMAL POWER within two hours to less than or equal to that fraction of RATED THERMAL POWER which is allowed by the group position using the above figure, or l
c.
Be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
SURVEILLANCE REQUIREMENTS 4.1.3.6 The position of each control bank shall be determined to be within the insertion limits at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> except during time intervals when the Rod Insertion Limit Monitor is inoperable, then verify the ind'vidual rod positions at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
- See Special Test Exceptions 3.10.2 and 3.10.3.
- With Keff greater than or equal to 1.0.
SEQUOYAH - UNIT 2 3/4 1-21 Amendment No. 33, 104
_J
NO I
TC #
g A
7 7
7 7
6 6
E L'
BS AE CD 2
2 IO LM 1
1 P
1 1
1 1
P A
l rl e
wnoe wn
- p
- p
- p v
oi v
oi SE nro nro nro e
l pe l
p MLL i eo ieo i eo ldfholdfho 3
3 UEB pl pl pl n-co-n-
co MNA
'po po po p a pt l p a pt l INR o
g o
g o
g o
oa o
oa NAE oh n oh n oh n o
omeo cme N
IHP l ci l ci l ci l
l sml sm 1
O MCO
/ at
/ at
/at
/
/i a/
/ia I
2 ea 2 ea 2 ea 1
A 2 ms2 1ms T
T
)
N l
d E
S s
e wn e
M LP p
p p
vt oi u
U EI n-o n-o n-o en l
p n
R NR i ra i ro i ro l ehfho i
T NT e
el el dt - co 2
2 t
S A
p p' pp pp pi i p t l n
N HO oog oog oog ocwoa o
I CT o
n o
n o
n on ome C
l yi l oi l
yi li l sm
(
M
/nt
/ wt
/nt
/ o
/ia E
2aa 2t a 2aa 1 c 1ms 1
T S
3 Y
l S.
S e
wn 3
.L v
oi P
OE e
l p
E I
NN l
fho L
R N
d - co s
s B
T LA p
p p
p n pt l u
u A
AH o
o o
oaoa b
b T
R TC o
o o
o ome
/
/
O O
l l
l l
l sm 1
1 T
TF
/
/
/
/
/ia C
O 3
3 3
2 2 ms 4
4 A
E R
tn t
a n
l p
a o
o
)
l o
o s8 o
C L
p-o oP m
C r
e o
r al o
l L w e
wee r
t g
o t
ot v o
c n
ol a
l S e t
a we W
F L
c e
i S
Tb w
a R
r ror e
- d ow eL e R
y n
t o t
t c
w wa aL ad a e
n pu o) o r-wnW g
e il T
l 8 7
ew da a
u rF l
I F -
F -
no e
r t
q T
N P
P eL eh o l
e U
f f
G -
F ct o
r eo oe oe
/t a v
f nL L
v v
ml mar rs rs i
A so so ae ame ep ep b
N sb sb ev esn d m d m r
O oA oA t e ti e nu nu u
I L(
L(
SL SMG UP UP TAB T
C N
U 2
3 4
5 6
7 8
F 1
1 1
1 1
1 1
mmjYz c% "
y# w0
51 NO I
2 26 26 a
b c
e d
b 4
T 1
11 11 8
8 8
8 8
8 1
CA 5
5 d
d d
n n
n E
a a
a L
d BS n
AE 2
24, 24, a
2 2
CD IO LM 1
13 13 2
1 1
1 1
1 1
P P
A SE MLL UEB MNA INR NAE 2
22 22 2
3 3
3 3
2 2
N IHP O
MCO I
TAT
)
N d
E S
e M
LP u
U EI n
R NR i
T NT 1
11 11 1
2 2
2 2
1 1
t S
A n
N HO o
I CT C
(
M E
1 T
S 3
YS S
3
.L P
OE E
I NN L
R N
B T
LA 2
22 22 2
4 4
4 4
2 2
A AH T
R TC O
O T
TF C
O A
E R
n n
o o
i i
s t
t k
a a
c r
r r
o e
e e
l b
p p
r m
O O
e n
n n
n a
t e o
o o
o h
t r
r n g6 r
r r
r C
4 u
se e
I n-t t
t t
p rw cw aP u
u u
u e
P n
eo io mR e
e e
e s3 I
kP gP e
N N
N N
l 1 a
o t ex u-p n
ed Ld st u e
e e
e pP i
o rn n
yal g
g g
g0 m
i B a pa SiF n7 n8 n9 n1 I
T t
n i
n d
a-a-
a-a-
e c
e i uo T uo imo nu T
j rtd td rrr r,
r r,
r, i s t
I n
T rt crt Tet ex ex ex ex bs c
N IF au i au t u wu wu wu wu re a
U S
rth tth rne ol ol ol ol ur e
yE oSS ASS oIN PF PF PF PF TP R
L t
t m
t A
em c
o c
N f o a
t a
O ar e..
u..
e I
C D
E F
G T
C N
U 9
0 1
2 F
1 2
2 2
hS$
ci * "
R* {
F@&@ &
- e.
- i' oR 2 3 g-" %g@ E ?r o
l
l l
TABLE 3.3-1 (Continued)
ACTION 8 - With less than the Minimum Number of Channels OPERABLE, declare the interlock inoperable and verify that all affected channels of the functions listed below are OPERABLE or apply the appro-priate ACTION statement (s) for those functions.
Functions to be evaluated are:
a.
Source Range Reactor Trip.
b.
Reactor Trip Low Reactor Coolant Loop Flow (2 loops) l',dervol tage Underfrequency Pressurizer Low Pressure Pressurizer High Level c.
Reactor Trip Low Reactor Coolant Loop Flow (1 loop) d.
Reactor Trip Intermediate Range Low Power Range Source Range e.
Reactor Trip Turbine Trip ACTION 9 - Deleted ACTION 10 - Deleted ACTION 11 - Deleted ACTION 12 - With the number of OPERABLE channels one less than required by the Minimum Channels OPERABLE requirement, be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; however, one channel may be bypassed for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for surveillance testing per Specification 4.3.1.1.1 provided the other channel is OPERABLE.
1 l
l 1
SEQUOYAH - UNIT 2 3/4 3-7 Amendment No. 46, 99, 104
l l
i TABLE 4.3-1 (Continued)
I NOTATION With the reactor trip system breakers closed and the control rod drive system capable of rod withdrawal.
(1)
If not performed in previous 7 days.
(2) -
Heat balance only, above 15% of RATED THERMAL POWER.
Adjustchannel if absolute difference greater than 2 percent.
Compare incore to excore AXIAL FLUX DIFFERENCE above 15% of RATED l
(3)
THERMAL POWER.
Recalibrates if the absolute difference greater than or equal to 3 percent.
(4)
Manual.ESF functional input check every 18 months.
(5)
Each train or logic channel shall be tested at least every 62 days
.on a STAGGERED TEST BASIS.
The test shall independently verify the OPERABILITY of the undervoltage and automatic shunt trip l
circuits.
(6)
Neutron detectors may be excluded from CHANNEL CALIBRATION.
(7)
Below P-6 (Block of Source Range Reactor Trip) setpoint.
(8)
Logic only, each startup or when required with the reactor trip system breakers closed and the control rod drive system capable of rod withdrawal if not performed in previous 92 days.
.The CHANNEL FUNCTIONAL TEST shall independently verify the l
(9) operability of the undervoltage and shunt trip circuits for the manual reactor trip function.
(10) -
Local manual shunt trip prior to placing breaker in service.
Each train shall be tested at least every 62 days on a l
STAGGERED TEST BASIS.
(11) -
Automatic and manual undervoltage trip.
1 SEQUOYAH - UNIT 2 3/4 3-13 Amendment No. 46, 104
4 I
. INSTRUMENTATION 3/4.3.2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.2 The Engineered Safety Feature Actuation System (ESFAS) instrumentation channels and interlocks showa in Table 3.3-3 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3-4 and with RESPONSE TIMES as shown in Table 3.3-5.
APPLICABILITY:
As shown in Table 3.3-3.
ACTION:
a.
With an ESFAS instrumentation channel or interlock trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3-4, declare the channel inoperable and apply the applicable ACTION requirement of Table 3.3-3 until the channel is restored to OPERABLE status with the trip setpoint adjusted con-sistent with the Trip Setpoint value.
b.
With an ESFAS instrumentation channel or interlock inoperable, take the ACTION shown in Table 3.3-3.
SURVEILLANCE REQUIREMENTS 4.3.2.1.1 Each ESFAS instrumentation channel and interlock shall be demon-strated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations for the MODES and at the frequencies shown in Table 4.3-2.
4.3.2.1.2 The logic for the interlocks shall be demonstrated OPERABLE during the automatic actuation logic test.
The total interlock function shall be demonstrated OPERABLE at least once per 18 months during CHANNEL CALIBRATION testing of each channel affected by interlock operation.
4.3.2.1.3 The ENGINEERED SAFETY FEATURES RESPONSE TIME of each ESFAS function shall be demonstrated to be within the limit at least once per 18 months.
Each test shall include at least one logic train such that both logic trains are tested at least once per 36 months and one channel per function such that all channels are tested at least once per N times 18 months where N is the total number of redundant channels in a specific ESFAS function as shown in the " Total No. of Channels" Column of Table 3.3-3.
SEQUOYAH - UNIT 2 3/4 3-14
t t t t
in a a a n
rce r r r a
oem e e e e r
t pe n n n n d
aSl i e e e a
r l
g g g u
ehp SE q
pau HLL m m m m e e e o yS MEB a a a a v v v e
o IMR A
e e e e l
l l
r euo t t t t a a a o
vqt NAE s s s s v v v c
l e LHP
/ / / /
/ / /
/
aSe NCO 1 1 1 1 1 1 1 1 1 1 0 1 1 1 1 2 1 v
l n eio htp t nsue o
td eA rtV enT we ome peh r r r l t o o o hp t t t t
t mn a a a n
.iii N
r r r a
nw O
e e e e r
o e
I OS n n n n d
inbt T
NL i e e e a
t o A
E l
g g g u
iitd T
EM D
q stoe N
m m m m e e e oint E
RA a a a a v v v e
ps 3
IH e e e e l
l l
r oli UC t t t t a a a o
d pl m tR Q
s s s s v v v c
e i
T EF
////
/ / /
/
sdwo 0
S RO 2 2 2 2 2 2 2 1 1 1 1 2 2 2 2 4 2 oe 1
N l st I
3 cons l ea G
3 N
ecm h
e I
E R
tere hit L
O r
ntus B
T o
r i
q A
I t
o nes T
N i
t sir O
n a
i M
o c
m dei M
i e
eech T
d t
vint N
n n
s lfa E
i I
y ailf D
)
g S
vrl o I
)
e r
n ei C
e
)
g a
- o
)
n kvee C
g e n M r i e
o c
v A
)
l n
g a o t g
i osru
)
e) e a n R g t i n
t li u e g e v R a
n a s a
a b
s g n g e R w i c o r R t
e n a n L e) o l
i P o n
dvdt a R a d e e r e o d t
e e
el n R
R k i g
d r t o n e a d m
t aar n W n i a a c I
v c i u
av me hid a
(
a W N R b l
i W
r i
n T
R
( (
u n a d
(
t cko i W i
e wS o V n s
ocid W ( W e r e o
i I
l n
sote
(
(
g u d l
m t k e
I sl ap d
a s i l
l F e i c n v s abco t
l e r s W e e t s o o e e l
il o or o e
(
e v v r s o l
i L l
e eef e H Cu t r r e e e y P B t p v hhiv T T s S P l
u L L t S i r u e ttce s
e s a
V V s e o L
- e t t e r t v s r r w t R R o t c ff p n n r e n e e o o d n O O P a o l
iiSe a a P t a L r t t e a P P W m e l
l a
l P a a e l
e r s eel a o o t W o r r r F o r r v t e s l l a E
t t a e t s m m ml t s s t a o t ppec0 o o n o e e e e o e e l
n h e bbcs C C e g C z n n n yC z z a e T V a ai n m n i i e e r i i V m ccno7 T
r r n i r r L G G a r r r n e r iihi3 N
o o i l
o u i o u u yi r o l l ct M
c c t u c s a a a i c s s e t C c p pT u U
a a n f a e e e e x a e e f
n a
aa R
e e o e e r t t t u e r r a o n e s
T R R C R R P S S S A R P P S C I
R ttisR SN oohnU hNTIN I
1 2 3 4 5 6 7 8 9 0 1 2 3 4 5 6 7 1 1 1 1 1 1 1 1
- y3#* [zW "
y[h
)
l CONTAINMENT SYSTEMS l
)
I SURVEILLANCE-REQUIREMENTS (Continued) 3.
Valves pressurized with fluid from a seal system.
j
)
The combined bypass leakage rate to the auxiliary building shall be e.
1 determined to be less than or equal to 0.25 L,by applicable Type B l
and C tests at-least once per 24 months except for penetrations which are not individually testable; penetrations not individually testable shall be determined to have no detectable leakage when tested with soap bubbles while the containment is pressurized to P, 12 psig, i
a during each. Type A test, f.
Air locks shall be tested and demonstrated OPERABLE per Surveillance Requirement 4.6.1 3.
g.
Leakage from isolation valves that are sealed with fluid from a seal system may be excluded, subject to the provisions of Appendix J,Section III.C.3, when determining the combined leakage rate provided the seal system and valves are pressurized to at least 1.10 P,13.2 a
psig, and the seal system capacity is adequate to maintain system pressure (or fluid head for the containment spray system and RHR spray system valves at penetrations 48A, 48B, 49A and 498) for at least 30 days.
h.
Type B tests for penetrations employing a continuous leakage monitoring system shall be conducted at P, 12 psig, at intervals no a
greater than once per 3 years, i.
All test leakage rates shall be calculated using observed data converted to absolute values.
Error analyses shall be performed to select a balanced integrated leakage measurement system.
j.
The provisions of Specification 4.0.2 are not applicable.
i SEQUOYAH - UNIT 2 3/4 6-4 Amendment No. 63, 90, 104 (Correction Letter of 7-11-88)
CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 4.6.3.2 Each isolation valve specified in Table 3.6-2 shall be demonstrated OPERABLE during the COLD SHUTDOWN or REFUELING MODE at least once per 18 months by:
a.
Verifying that on a Phase A containment isolation test signal, each Phase A isolation valve actuates to its isolation position.
b.
Verifying that on a Phase B containment isolation test signal, each Phase B isolation valve actuates to its isolation position.
Verifying that on a Containment Ventilation isolation test signal, c.
each Containment Ventilation valve actuates to its isolation position.
d.
Verifying that on a high containment pressure isolation test signal, each Containment Vacuum Relief Valve actuates to its isolation position.
Verifying that on a Safety Injection test signal that the Normal e.
Charging Isolation valve actuates to its isolation position.
4.6.3.3 The isolation time of each power operated or automatic valve of Table 3.6-2 shall be determined to be within its limit when tested pursuant to Specification 4.0.5.
4.6.3.4 Each containment purge isolation valve shall be demonstrated OPERABLE within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after each closing of the valve, except when the valve is being used for multiple cyclings, then at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, by verifying that when the measured leakage rate is added to the leakage rates determined pursuant to Specification 4.6.1.2.d for all other Type B and C penetrations, l
the combined leakage rate is less than or equal to 0.60 L
- i a
l l
l l
SEQUOYAH - UNIT 2 3/4 6-18 Amendment No. 72, 90, 104 L_________
-4 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE SAFETY VALVES LIMITING CONDITION FOR OPERATION 3.7.1.1 All main steam line code safety valves associated with each steam generator shall be OPERABLE with lift settings as specified in Table 3.7-3.
APPLICABILITY:
Modes 1, 2 and 3.
ACTION:
a.
With 4 reactor coolant loops and associated steam generators in operation and with one or more main steam line code safety valves inoperable, operation in Modes 1, 2 and 3 may proceed provided, that within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, either the inoperable valve is restored to OPERABLE status or the Power Range Neutron Flux High Trip Setpoint is reduced per Table 3.7-1; otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and'in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />, b.
With 3 reactor coolant loops and associated steam generators in operation and with one or more main steam line code safety valves associated with an operating loop inoperable, operation in MODE 3 may proceed provided, that within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, either the inoperable valve is' restored to OPERABLE status or the Power Range Neutron Flux High Trip Setpoint is reduced per Table 3.7-2; otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />, c.
The provisions of Specification 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.7.1.1 No additional Surveillance Requirements other than those required by Specification 4.0.5.
i SEQUOYAH - UNIT 2 3/4 7-1 Amendment No. 104
a I
l.
.o, TABLE 3.7-1 MAXIMUM ALLOWABLE POWER RANGE NEUTRON FLUX HIGH SETPOINT WITH INOPERABLE STEAM LINE-SAFETY VALVES DURING 4 LOOP OPERATION l
Maximum Number of Inoperable Maximum Allowable Power Range
?
Safety Valves on Any Neutron Flux High Setpoint Operating Steam Generator (Percent of RATED THERMAL POWER)
,l' 87-2.
65 3.
43
/ se
,4 t ' S i
SEQUOYAH - UNIT 2 3/4 7-2
PLANT SYSTEMS AUXILIARY FEEDWATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.1.2 At least three independent steam generator auxiliary feedwater pumps and associated flow paths shall be OPERABLE with:
a.
Two motor-driven auxiliary feedwater pumps, each capable of being powered from separate shutdown boards, and b.
.One turbine-driven auxiliary feedwater pump capable of being powered from an OPERABLE steam supply system.
APPLICABILITY:
Modes 1, 2 and 3.
ACTION:
a.
With one auxiliary feedwater pump inoperable, restore the required auxiliary feedwater pumps to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
b.
With two auxiliary feedwater pumps inoperable, be in at least H0T STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTOOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
c.
With three auxiliary feedwater pumps inoperable,.immediately initiate corrective action to restore at least one auxiliary feedwater pump to OPERABLE status as soon as possible.
SURVEILLANCE REQUIREMENTS
- 4. 7.1. 2 In addition to the requirements of Specification 4.0.5 each auxiliary feedwater pump shall be demonstrated OPERABLE by:
a.
Verifying that:
1.
each motor-driven pump develops a differential pressure of greater than or equal to 1397 psid on recirculation flow.
2.
the steam-turbine driven pump develops a differential pressure of greater than or equal to 1183 psid on recirculation flow when I
the secondary steam supply pressure is greater than 842 psig.
The provisions of Specification 4.0.4 are not applicable for entry into MODE 3.
SEQUOYAH - UNIT 2 3/4 7-5 l
l
,j'
(
'\\.
]e
~
PLANT SYSTEMS' SURVEILLANCE REQUIREMENTS (Continued)
.I y
l F
n' 1
E 3.
at least once per 31 days, each automatic control valve in the flow path is OPERABLE whenever the auxiliary feedwater system is placed in automatic control or when above 10% of RATED THERMAL POWER.
b.
At least once per 18 months during ' shutdown
- by:
1.
Verifying that each automatic valve in the flow path actuates to its correct position upon receipt of an auxiliary feedwater
- actuation test signal and a low auxiliary feedwater pump suction pressure test signal.
2.
Verifying that each auxiliary feedwater pump starts as designed automatically upon receipt of an auxiliary feedwater actuation test signal.
c.
At least once per 7 days by verifying that each non-automatic valve in the auxiliary feedwater system flowpath is in its correct position.
- The provisions of Specification 4.0.4 are not applicable for entry into MODE 3 for the turbine-driven Auxiliary Feedwater Pump.
SEQUOYAH - UNIT 2 3/4 7-6 Amendment No.
68, 104
.o TABLE 4.7-1 SECONDARY COOLANT SYSTEM SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM TYPE OF MEASUREMENT SAMPLE AND ANALYSIS AND ANALYSIS FREQUENCY 1.
Gross Activity Determination At least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
2.
Isotopic Analysis for DOSE a) 1 per 31 days, when-EQUIVALENT I-131 Concentration ever the gross activity determination indicates iodine concentrations greater than 10% of the allowable limit, b) 1 per 6 months, when-ever the gross activity.
determination indicates iodine concentrations below 10% of the allow-able limit.
i SEQUOYAH - UNIT 2 3/4 7-9
iv S
,D y
PLANT SYSTEMS MAIN STEAM LINE ISOLATION VALVES i
LIMITING CONDITION FOR OPERATION t
- 3. 7.1. 5 Each main steam line isolation valve shall be OPERABLE.
APPLICABILITY:
Modes 1, 2 and 3.
ACTION:
Modes 1.With one main steam line isolation valve inoperable, POWER OPERATION may continue provided the inoperable valve is either restored to OPERABLE status or closed within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />; Otherwise, be-in HOT-STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
MODES 2 - With one main steam line isolation valve inoperable,' subsequent and 3 operation in MODES 2 or 3 may proceed provided:
a.
The isolation valve is maintained closed.
b.
The provisions of Specification 3.0.4 are not applicable.
Otherwise, be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTOOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
SURVEILLANCE REQUIREMENTS 4.7.1.5 Each main steam line isolation valve shall be demonstrated OPERABLE by verifying full closure within 5 seconds when tested pursuant to Specification 4.0.5.
4 SEQUOYAH - UNIT 2 3/4 7-10 Amendment No. 104
PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) g.
Functional Test Failure - Attached Component Analysis For snubbers (s) found inoperable, an engineering evaluation shall be performed on the components which are restrained by the snubber (s).
The purpose of this engineering evaluation shall be to determine if the components restrained by the snubber (s) were adversely affected by the inoperability of the snubbers (s),.and in order to ensure that the restrained component remains capable of meeting the designed service.
h.
Functional Testing of Repaired and Spare Snubbers Snubbers which fail the visual inspection or the functional test acceptance criteria shall be repaired or replaced.
Replacement snubbers and snubbers which have repairs which might affect the functional test results shall be tested to meet the funtional test criteria before installation in the unit.
These snubbers shall have met the acceptance criteria subsequent to their most recent service, and the functional test must have been performed within 12 months before being installed in the unit, i.
Snubber Service Life Program The seal service life of hydraulic snubbers shall be monitored to ensure that the seals do not fail between surveillance inspections.
The maximum expected service life for the various seals, seal materials, and applications shall be estimated based on engineering information, and the seals shall be replaced so that the maximum expected service life does not expire during a period when the snubber is required to be operable.
The seal replacements shall be l
documented and the documentation shall be retained in accordance wth 6.10.2.n.
Mechanical snubber drag force increases greater than 50 percent of previously measured values shall be evaluated as an indication of impending failure of the snubber.
These evaluations and any associated corrective action, shall be documented, and the documentation shall be retained in accordance with 6.10.2.n.
j.
Exemption From Visual Inspection or Functional Tests Permanent or other exemptions form the surveillance program for individual snubbers may be granted by the Commission if a justifiable basis for exemption is presented and if applicable snubber life destructive testing was performed to qualify snubber operability for the applicable design conditions at either the completion of their l
fabrication or at a subsequent date.
SEQUOYAH - UNIT 2 3/4 7-25 Amendment No. 31,104 (Correction Letter of 6-28-85)
-f TABLE 3.7-5 FIRE H0SE STATIONS LOCATION ELEVATION HOSE RACK #
a.
Reactor Building - Annulus Area Platform 778.0 2-26-1196 Platform 778.0 2-26-1197 Platform 778.0 2-26-1198 Platform 778.0 2-26-1199 Platform 759.0 2-26-1200 Platform 759.0 2-26-1201 Platform 759.0 2-26-1202 Platform 759.0 2-26-1203 Platform 740.0 2-26-1204 Platform 740.0 2-26-1205 Platform 740.0 2-26-1206 Platform 740.0 2-26-1207 Platform 721.0 2-26-1208 Platform 721.0 2-26-1209 Platform 721.0 2-26-1210 Platform 721.0 2-26-1211 Platform 701.0 2-26-1212 Platfonn 701.0 2-26-1213 Platform 701.0 2-26-1214 Platform 701.0 2-26-1215 Platform 679.78 2-26-1216 Platform 679.78 2-26-1217 Platform 679.78 2-26-1218 Platform 679.78 2-26-1219 b.
Reactor Building - RCP & Lower Containment Air Filters Area Reactor Building 679.78 2-26-1220 Reactor Building 679.78 2-26-1221 Reactor Building 679.78 2-26-1222 Reactor Building 679.78 2-26-1223 Reactor Building 679.78 2-26-1224 Reactor Building 679.78 2-26-1225 c.
Control Building Control Building 732 0-26-1186 Control Building 732 0-26-1191 Control Building 706 0-26-1187 Control Building 706 0-26-1192 SEQU0YAH - UNIT 2 3/4 7-49 I
- -_m
Table 3.7-5 (Continued)
FIRE H0SE STATIONS LOCATION.
ELEVATION HOSE RACK #
Control Building 685 0-26-1188 Control Building 685 0-26-1193 Control Building 669 0-26-1189 Control Building 669 0-26-1194.
d.
Diesel Generator Building Corridor 722 0-26-1077 Corridor 740.5 0-26-1080 Air Exhaust'Rm.
740.5 0-26-1082 r
1 e.
Additional Equipment Building - Unit 2 North Wall 740.5 2-26-687 North Wall 706 2-26-686 f.
Auxiliary Building 759 2-26-669 l
749 2-26-664 749 1-26-664 734 2-26-670 734 0-26-684 734 1-26-670 734 0-26-682 734 Siamese Outlet 2-26-671 734 2-26-672 734 2-26-665 714 0-26-660 714 2-26-666 714 0-26-677 706 0-26-658 690 0-26-690 690 0-26-661 690 Siamese Outlet 2-26-674 690 2-26-675-669 2-26-667 669 2-26-668 669 0-26-662 669 0-26-680 653 0-26-663 653 0-26-691 SEQUOYAH - UNIT 2 3/4 7-50 Amendment No.104 i
3/4.9 REFUELING OPERATIONS l
3/4.9.1-BORON CONCENTRATION l
LIMITING CONDITION FOR OPERATION 3.9.1 With the reactor vessel head closure bolts less than fully tensioned or with the head removed, the boron concentration of all filled portions of the Reactor Coolant System and the refueling canal shall be maintained uniform and sufficient to ensure that the more restrictive of the following reactivity conditions is met:
a.
Either a K of 0.95 or less, which includes a 1% delta k/k conservatiNallowanceforuncertainties,or b.
A boron concentration of greater than or equal to 2000 ppm, which includes a 50 ppm conservative allowance for uncertainties.
APPLICABILITY:
MODE 6*
ACTION:
With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS or positive reactivity changes and initiate and continue boration at greater than or equal to 10 gpm of a solution containing greater than or equal to 20,000 ppm boron or its equivalent until K is reduced to less than or equal to 0.95 or the boron concentration is rN(ored to greater than or equal to 2000 ppm, whichever is the more restrictive.
The provisions of Specification 3.0.3 are not applicable.
SURVEILLANCE REQUIREMENTS 4.9.1.1 The more restricti/e of the above two reactivity conditions shall be determined prior to:
a.
Removing or unbolting the reactor vessn1 head, and b.
Withdrawal of any full length control rod in excess of 3 feet from its fully inserted position within the reactor pressure vessel.
- The reactor shall be maintained in MODE 6 whenever fuel is in the reactor vessel with the vessel head closure bolts less than fully tensioned or with the head removed.
SEQUOYAH - UNIT 2 3/4 9-1 Amendment No. 104 i
I
~.
a REFUELING OPERATIONS
)
SURVEILLANCE REQUIREMENTS (Continued) 4.9.1.2 The boron concentration of the reactor coolant system and the refueling canal shall be determined by chemical analysis at least once por 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
- 4. 9.1. 3 One of the following valve combinations shall be verified closed under administrative control at least once per 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />s:
Combination A Combination B Combination C Combination 0 a.
2-81-536 a.
2-81-536 a.
2-81-536 a.
2-81-536 b.
2-62-922 b.
2-62-922 b.
2-62-907 b.
2-62-907 c.
2-62-916 c.
2-62-916 c.
2-62-914 c.
2-62-914 d.
2-62-933 d.
2-62-940 d.
2-62-921 d.
2-62-921 e.
2-62-696 e.
2-62-933 e.
2-62-940 f.
2-62-929 f.
2-62-929 g.
2-62-932 g.
2-62-932 h.
2-FCV-62-128 h.
2-62-696 i.
2-FCV-62-128 SEQUOYAH - UNIT 2 3/4 9-2
)
fD oL(
t()
g i
l 4 8 mnm 2
1 io/
4 4
8 00 4
8 1
1 Lii 11 tC 0
0 0
~
0 0
0 0
rcp 1
1 1
xx 1
1 1
1 ee(
x x
x x
x x
x wt 1
1 1
11 1
1 l
1 oe LD M
A R
GO R
P 9
9 9
9 9
s s
s s
s S
r r
r r
r I
e e
e e
e S
t t
t t
t Y
t t
t t
t L
i i
i i
i A
s m
m m
m m
N i
E E
E E
E A
s y
a a
a a
a)
D l
m m
m m
ms N
a m
m m
mr A
f n a
a a
a ae oA G
G G
G Gh G
t N
ey l
l l
l l O I
pt a
a a
a a
L yi p
p p
p p,
=
P Tv i
i i
i i 1 M
i c
c c
c 1
c3 2A t
n n
n n
3 n1
- S c
i i
3 i3 i
3 1
i -
1 A
r r
r-r rI 1G P
P H
PH P
I P(
1 F
.N 4I R EO LT BI AN e
TO e
t M
y k
g a
sc n
r y
l l
E mi n a
u a
d a d u T
use PT iP D
Woe Wce M S
myu D
cl il A
il q h
h h
rp t p W
nae c
c c
am rm inr a
a ja h a aa S
MAf E
E DE M
CS PS U
O E
S e
A e
l s
s G
y k
g p
u u
gc n
r ym o
o E
nn a
u aa ur ur V
i e T
e P
e DS e
ne ne I
l u P l
i l
l i l il T
pq hb p Phbp hb b p t p t p C
me cam cam ca am nm nm A
ar ara arajar ra oa oa 0
D g
g A
n n
R m
i g
i e
u d
n g
d g
e g
u l
i n
i l
n p
a d
c i
d i
t i
i y
r n
a u
l g
r u
d T
o a
V B
i l,
a B
l t
e u f
P i
e S
t s
r RB u
y u
g d
r B
s ne e
eh a
s eg t
s s t at et t
n at e
a mr n
a n
si s c s d s a is d s G
nu e
Gme ul uiul u sl u l u le iP V
ud ai a v a e a e ei a ea R
e a
ei n h xh rhih nt xh ih tk t
l t o xuxexh x i aux h x s
sn n
biC EAESESE dl AE SE u
aa o.
or ou o
WT C1 2
NT1 2
3 4
I c1 2
esa G
A B
C D
$$57 cz E
9 ]n 5
$* 7E
,5S$5 zP
.u jle
i RADIOLOGICAL ENVIRONMENTAL MONITORING LIMITING CONDITION FOR OPERATION (Continued) l deleted from the monitoring program.
In lieu of a licensee event report (LER) and pursuant to Specification 6.9.1.7, identify the l
cause(s) of the unavailability of samples and identify the new loca-tions for obtaining replacement samples in the Annual Radiological Environmental Operating Report.
A revised figure (s) and table (s) for the ODCM reflecting the new location (s) shall be included in the next semiannual radioactive effluent release report pursuant to Specification 6.9.1.9.
d.
The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.12.1 The radiological environmental monitoring samples shall be collected pursuant to Table 3.12-1 from the locations given in the table and figure in the ODCM and shall be analyzed pursuant to the requirements of Table 3.12-1 and the detection capabilities required by Table 4.12-1.
SEQUOYAH - UNIT 2 3/4 12-2 Amendment No. 34, 104 L_________.__
RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.2 LAND USE CENSUS LIMITING CONDITION FOR OPERATION 3.12.2 A Land Use Census shall be conducted and shall identify within a dis-tance of 8 km (5 miles) the location in each of the 16 meteorological sectors l
the nearest milk animal, the nearest residence and the nearest garden
- of greater than 50 m2 (500 ft2) producing fresh leafy vegetation.
APPLICABILITY:
At all times.
ACTION:
a.
With a Land Use Census identifying a location (s) which yields a calcu-lated dose or dose commitment 20% greater than the values currently being calculated in Specification 4.11.2.3, identify the new locations (s) in the next Semiannual Radioactive Effluent Release Report, pursuant to Specification 6.9.1.9.
b.
With a Land Use Census identifying a location (s) which yields a calcu-lated dose or dose commitment (via the same exposure pathway) 20%
greater than at a location from which samples are currently being obtained in accordance with Specification 3.12.1, add the new loca-tion (s) within 30 days to the Radiological Environmental Monitoring Program given in the ODCM, if samples are available.
The sampling location, excluding the control station location, having the lowest calculated dose or dose commitment (s), via the same exposure pathway, may be deleted from this monitoring program after October 31 of the year in which this land use census was conducted.
Pursuant to Speci-fication 6.14, submit in the next Semiannual Radioactive Effluent Release Report documentation for a change in the ODCM including a revised figure (s) and table (s) for the ODCM reflecting the new loca-tion (s) with information sunporting the change in sampling locations.
c.
The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.12.2 The Land Use Census shall be conducted during the growing season at least once per 12 months using that information that will provide the best results, such as by a door-to-door survey, mail survey, telephone survey, aerial survey, or by consulting local agriculture authorities.
The results of the Land Use Census shall be included in the Annual Radiological Environmental Operating Report pursuant to Specification 6.9.1.7.
- Broad leaf vegetation sampling of at least three different kinds of vegeta-tion may be performed at the SITE BOUNDARY in each of two different direction sectors with the highest predicted D/Qs in lieu of the garden census.
Speci-fications for broad leaf vegetation sampling in Table 3.12-1.4c shall be followed, including analysis of control samples.
SEQUOYAH - UNIT 2 3/4 12-9 Amendment No. 34, 104
- 5. 0 DESIGN FEATURES 5.1 SITE EXCLUSION AREA 5.1.1 The exclusion area shall be as shown in Figure 5.1-1.
LOW POPULATION ZONE 5.1.2 The low population zone shall be as shown in Figure 5.1-2.
SITE BOUNDARY FOR GASEOUS EFFLUENTS 5.1.3 The site boundary for gaseous eff 7uents shall be as shown in Figure 5.1-1.
SITE BOUNDARY FOR LIQUID EFFLUENTS 5.1.4 The site boundary for liquid effluents shall be as shown in Figure 5.1-1.
- 5. 2 CONTAINMENT CONFIGURATION 5.2.1 The shield building is a reinforced concrete building of cylindrical shape, with a dome roof around a free standing steel containment and having the following design features:
a.
Nominal inside diameter = 125 feet.
b.
Nominal inside height = 175 feet.
c.
Minimum thickness of concrete walls = 3 feet, d.
Minimum thickness of concrete roof = 2 feet.
e.
Minimum thickness of concrete floor pad = 9 feet.
f.
Minimum thickness of steel containment liner = 0.5 inches at the spring line and 0.25 inches at the bottom liner plate.
5 g.
Net free volume = 3.75 x 10 cubic feet between the steel containment and the shield building.
l DESIGN PRESSURE AND TEMPERATURE 5.2.2 The steel containment is designed and shall be maintained for a maximum internal pressure of 12 psig and a temperature of 250 F.
SEQUOYAH - UNIT 2 5-1
f eo st
/
rB dh
[
,n e
ae s
e s
e sa d t ee ei rl ae t S R
c net
=
i e t m ee rd ee l/
t i tt ct
~
S s eRU et
=
t u Y
nO R
1 G
U(
A 1
D S
N a
s, U
,y O
w s
B
,p.
a 2
r E
E T
=
N IS O
s Z
E l
S uI 3
A E
E N
L N
E O
k O
Z Z
I t
E T
N A
e S
L O
U C
l j
E L
i t
S A
- <l t
E A
C l;
E E
5
- \\
R O
\\-
D I
m A
s B
p{
p-
__D K
A p
- L "l
AN g
AC t
E CR I
i A
O
}
D
- N g r.
S t.
I t
f j
r t
e3 e
L t
1 A
n
.1 C
t1e I
i G
t e
k e.
t e O
(
L t
s mt O
e R
u eo O
h E
t c e
,.s e
t
~
T
?)
F se E
t M
r t d e
,e a
t oo o
~
0 re P
g esv hOo t
d g e
E st n n i
~
S A e
- i. i A l ue d 1 E a Ll ti i
.~
L n rp T t
T E a t
E Fe h A
R C e
t rb t
sv O e 0
e A
9 t ue P ed F I
t s ob tl R
u a s e E
t e t A
q n s
S
- e. h i $
n t
ae A
A s
L o ce E At E o
s t
L L
t t i st E
L S
u D ee R
S U
t t v L
i ie C
U vl I
D oe E
a n(
mN aZ sm C
U 1
;w m Co g 0 E:
m4 F 0 gg@ +c Z
r l
ll' il
l DESIGN FEATURES 5.6 FUEL STORAGE f
CRITICALITY - SPENT FUEL
- 5. 6.1.1 The spent fuel storage racks are designed for fuel enriched to 4.0 weight percent U-235 and shall be maintained with:
A k,ff equivalent to less than 0.95 when flooded with unborated a.
water, which includes a conservative allowance of 1.42% delta k/k for uncertainties.*
b.
A nominal 10.375 inch center-to-center distance between fuel assemblies placed in the storage racks.
CRITICALITY - NEW FUEL 5.6.1.2 The new fuel pit storage racks are designed and shall be maintained with a nominal 21.0 inch center-to-center distance between new fuel assemblies such that k,77 will not exceed 0.98 when fuel having an enrichment of 4.5 weight percent U-235 is in place and optimum achievable moderation is assumed.
New fuel enrichment is limited to 4.0 weight percent, as noted in 5.3.1 and 5.6.1.1.
DRAINAGE 5.6.2 The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 722 ft.
CAPACITY 5.6.3 The spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than 1386 fuel assemblies.
- 5. 7.1 The components identified in Table 5.7-1 are designed and shall be maintained within the cyclic or transient limits of Table 5.7-1.
I
- For some accident conditions, the presence of dissolved boron in the pool water i~
may be taken into account by applying the double contingency principle which requires two unlikely, independent, concurrent events to produce a criticality accident.
SEQUOYAH - UNIT 2 5-5 Amendment No. 4, 52
0 t
as O
r 0
cn P
e 2
o R
i o f
t E -
r, e L
f 5
m e
W A
i o
lF OR
'c h M
d m
r c
PE et r
R g
g g
o f
y0 W
l o
E e
i i
i r
c5 LO eg t
H r
s s
s f
9 6
AP nm c
T u
p p
p n
M
.i e a
t.
g T"F w>
RL C yt e
D aF 5
7 6
v o
.l s
r E
r 8
0 5
a dm HM Apy T
e0 4
1 3
T 0
l o TR pS e
A p6 2
3 1
0 or E
eu n
R m5 e2 of DH t sl o
e o
o o
l c
ET i
a f
t5 t t
t ET e
c5 s
T -
sec yp o
LN l
.y re AD f ci l
m rd d
d d
CE cF co er RE f rr nu en e
e e
YI y
t zu T
out op 0
t a z
z z
CS c0 n i t.
fA oc a
i i
i N
5 wF raF oR f s e ft o
wF r
r r
NA p5 o u r "0 o
l on t
u u
u GR u
d0 se
%f rE a
y0 s
s s
IT t>T5 s p0 5 o se sl a2 s
s s
S a
o5 em2 1
swF so 0
r3 e
e e
ER eoo re ooS oo 0
p r
r r
DO HtC>
Pt<
>0 L pE L c 1
S>
P P
P p
i e
r s
t r n
h e
ut o
/
l o
F c
h r e
n s
s S
y t o t
i t
t T
0t c
i t i
s s
I 0 a wc s
w s
e e
M 1
n a
f o
e t
t I
s w
,e f
l s
l L
< e o
sr o.
f e
c e
e l
d e
r l
y r
r T
tc l
l r f e f p c
c u
u N
ay o
co ow oo y
s s
E c
o y
o o
c n
s s
I T
s c
ce sp sl o
e e
S I
en n
s s
p i r
r 1
N M
w r
di ol ot i
t p
p l
A RI co e
ab l a l n r
a 7
R OL yd zr or c
a t
u s
c c
T cl ih l u fi fl t
t i
i 5
CT o
r/
t or oo r
c s
t t
R IN por eF f
t o
o a
e a
a E
O LE uch s
oe sc sc t
t t
t L
CI t
/
s0 t
ee e
c y
s s
B C
YS a0F e0 sa l l l r a
a k
o o
A I
CN e0 r2 si ce co e
r a
r r
T L
A h20 p
od y
yt r
p e
d d
C R
0 5
l e c
cc s
l y
y Y
T 0d1 0
m C
a 0
h h
C 0 n 0t 0 m 0
0 e 0
2 0
2 a<
2a 8i 4A 8r 4
1 5
5 5
TN E
N O
P M
O C
me tsy S
m t
e n
t a
s y
l o
S o
T C
y N
r E
r a
N o
d O
t n
P c
o M
a c
O e
e C
R S
$@SE,
ci*
N p@&$" e.
i' l