ML19326B883
| ML19326B883 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 03/17/1978 |
| From: | Reid R Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19326B881 | List: |
| References | |
| NUDOCS 8004180613 | |
| Download: ML19326B883 (22) | |
Text
~
[
- g useersorTAr S f
NUCLEAR REGULATOP.Y COMMisslON
.i f ^I
/
waswenarow. o.c.2asas
% +.... /
UtKANSAS POWER & LIGHT COMPANY DOCKET NO. 50-313
~
ARKANSAS NUCLEAR ONE - UNIT NO.1 i
p'ENOMENT TO FACILITY OPERATING LICENSE f
Amendment No. 31
~
License No. DPR-51 i
l 1.
The Nuclear Regulatory Comission (the Comission) has found that:
I A.
The application for amendment by Arkansas Power & Light Company (the licensee) dated December 28,1977, as supplemented by letters dated January 17 and 30, 1978, and March 3, 1978, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Comission; C.
There is reasonable assurance (i) 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 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.
1 4
l 8 004180 f fg,
_2 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. OPR-51 is hereby amended to read as follows:
(2) Technical Soecifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 31, 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 the date of its issuance.
FOR THE NUCLEAR REGU Y COMMISSION ly)$.
Robert W. Reid, Chief Operating Reactors Branch #4 Division of Operating Reactors
Attachment:
Changes to the Technical Specifications Date of Issuance:
March 17,1978 I
9 9
e e
4
ATTACHMENT TO LICENSE AMENDf1ENT NO. 31 FACILITY OPERATING LICENSE NO. DFR-51 DOCKET NO. 50-313 Revise the Appendix A Technical Specifications :: follows:
Remove Pages Insert Pages 9
9 w 9b 9b 12 12 14b 14b 30 & 30a 30 47 & 48 47 & 48 48b 48b 48bb 48bb 48bbb 48bbb 48c 48c 48cc 48cc 48ccc 48ccc 48d 48d 48dd 48dd 48ddd 48ddd 48f - 48h New pages and changes in the revised pages are identified by marginal lines.
.m m
Using a local quality limit of 22 percent at the point of minimen DNBR as a basis for curve 3 of Figure 2.1-3 is a conservative criterion even though the quality at the exit is higher than the quality at the point of minimum DNBR.
The DNBR as calculated by the SAW-2 correlation continually increases from point of minimum GiBR, so that the exit DNBR is always higher and ir a functi,on of the pressure.
The maximum thermal power for three pump operation is 85.6 percent due to a power level trip produced by the flux-flow ratio (74.7 percent flow x 1.060 79.1 percent power) plus the maximum calibration and instrumentation error.
.The maximum thermal power for other reactor coolant pump conditions is pro-duced in a similar manner.
For each curve of Figure 2.1-3, a pressure-temperature point above and to the le ft of the curve would result in a DNBR greater than 1.3 or a local quality at the point of minimum DNBR less than 22 ' percent for that particular reactor coolant pump situation. Curves 15 2 of Figure 2.1-3 are the most restrictive because any pressure / temperature point above and to the left of this curve will be above and to thi left of the other curve.
REFERENCES (1) Correlation of Critical Heat Flux in a Bundle Cooled by Pressurized Water, BAW-10000A, May, 1976.
(2) FSAR, Sect iun 3. 2. 3.1.1.c AmendmenrNo. M 31
Thermal Power Level, 5 UNACCEPTABLE OPERATION
( 0 '_I l 2 )
112 (2s,112) 1 ACCEPTABLE 4 PUNP OPERAT10N
.J 00 85.6
(-70,so)
(-so,ss.s)
(26 s5.6 2
" 80 (5s so)
ACCEPTABLE 3 & 4 PUNP OPERATION
(-64,54) ss.6 60
(-4. ss. s)
(2s,5s.s (55,55) 3
(-48,98)
ACCEPTA8LE 2,3 & 4 PUNP OPERATION
_. 40
('3'*3)
.. 20 i
-80
-60
-40 20 0
20 40 60 80 Power imoalance, 5 CURVE REACTOR COOLANT FLOW (GPN) 3 374,880 2
280,035 3
184,441 ARKANSAS POWER ANJ I.lGHT CORE PROTECTION SAFETY LIMITS
,)d,31 Amendment No.
9b Figure 2.1 2
.s m
The power level trip set point produced by the power-to-flow ratto provides bnth high powcr !cvel and low flow protection in the event the reactor power level ine rcases or the react or coolant, flow rate dec reas es. The power level trip set point produced by the power to flow ratio provides overpower DNB protection for all modes of pump operatton.
For every flow rate there is a maximum pemissible power levei, and for every power level there is a minimum pemissible low flow rate. Typical power level and low flow rate comoinations for the pump situations of Table 2.3-1 are as follows:
1.
Trip would occur when four reactor coolant pEmps are operat ing if power is 106.0 percent and reactor flow rate is 100 percent or flow rate is 94.3 percent and power level is 100 percent.
2.
Trip would occur when three reactor coolant pumps are operating if power is 79.1 percent and reactor flow rate is 74.7 percent or flow rate is 70.7 percent and power level is 75 percent.
\\
3.
Trip would occur when one reactor coolant pump is operating in each loop (total of two pumps operating) if the power is 52.3 percent and reactor flow rate is 49.2 persent or flow rate is 46.2 percent and the power Icvel is 49.0 mercent.
l The flux / flow rat ios account for the maximum calibration and instrinnentation errors and the maximum variat ion from the aserage value of the RC flow signal in such a manner that the reactor protective system receives a conservative indication of the RC flow.
lie penalty in reactor coolant flow through the core was taken for an open co re vent valve because of the core vent valve survet!!ance program during each refueling outal,e.
For safety analysis calculations the maximtsn cali-bration and instrmentation errors for the power level were used.
The power-imbalance boundartes are established in order to prevent reactor themal limits from being exceeded.
Diese thermal limits are either power pesking kW/ft limits or DNBR limits. The reactor power imhalance (power in top half of core minus power in the bottom half of core) - i tuces the power level trip produced by the power-to-flow ratio so that the boundaries of Figure 2.3-2 are produced. The power-to-flow ratio reduces the power, level trip associated reactor power-to-reactor power imbalance boundaries by 1.060 percent for a 1 percent flow reduction.
B.
Pump raonitors In conjunction with the power imbalance / flow trip, the pt.mp moni-itors prevent the minimum core DNBR from decreasing below 1.3 by trip-ping the reactor due to the loss of reactor coolant pump (s). The pump monitors also restrict the power level for the number of pumps in operation.
C.
RCS Pressure During a startup accident from low power or a slow rod withdrawal from high power, the system high pressure trip set point is reached before the nuclear overpower trip set point. The trip setting limit i
12 Amendment'No. % 0*.
-3 Thermal Power Level, 5 120
,0
(-19.106) j gg (14.106)
+
100- -
+
(-26.99)
ACCEPTABLE 4 y
P'JMP OPERATION
' #,2,
(-19,79.1) 80.. yg.3 (42.80)
.( 4,79, )
/
I
(-26,72.1)
, ACCEPTABLE 3&4 PUMP OPERATION 6 0._
, (-19,52.1) 52.1 (14,52.1)
,(42,53.2)
(-26,45.1),
40._
ACCE P T ABL E 2,3 44 PUMP OPERATION
,(42,26.1) 20 n.
i
+
+
11 il 18 gg
~
=
=
I 9
1 A
t f
t i
g g
-50
-40
-30
-20
-10 0
10 20 30 40 50 Power imoalance, 5 ARKANSAS POWER AND LIGNT COMPANY UNIT I PROTECTIVE SYSTEM MAXINUN ALLOTABLE SETPOINTS Amendment Nor,
31 14b l
i
.s
- 7.,
3 3.1.7 Moderator Temcerature Coefficient of Reactivity Specification 3.1.7.1 The nederator tmperature ocefficient (MIC) shall be ncn-pesitive whenever theral pcwer is >95% of rated thermal power and shall be less positive than 0.5 x 10'4 ak/k/*F whenever themal pcwer is <95% of rated themal power and the reactor is not shutdcc.
3.1.7.2 The MIC shall be determined to be within its limits by confi=atory measurments pricr to initial operation above 5% of rated theral power after ea'ch fuel loading. MIC measured values shall be extra-polated and/or ecmpensated to pemit direct carparisen with the limits in 3.1.7.1 above.
Bases A non-positive mcderator coefficeint at power levels above 95% of rated power is specified such that the maximum clad terperatures will not exceed the Final Acceptance Criteria based on ILCA analyses. Below 95% of rated power the Final Acceptance Criteria will not be exceeded with a positive moderator temperature coefficient of +0.5 x 10-4 ok/k/'F corrected to 95%
of rated pcwer. All other accident analyses as reported in the FSAR have been perfomed for a range of nederator te:perature coefficients including
+0.5 x 10-4 ak/k/'F.
e m
G Amendment 'no.l,,.,
30 r
s -
s
- 6. If a control rod in the regulating or axial power sheping grcups is declared inoperable per Specification 4.7.1.2. operation above 60 percent of the thermal power allowable for the reactor coolant pump combination may continue provided the rmis in the group are positioned such that the rod that was declared inop.erable is con -
tained within allowable group average position limits of Specifica-tion 4.7.1.2 and the withdrawal limits of Specificat ion 3.5.2.5.3.
- 3. 5. 2. 3 The worth of single inserted control rods during criticality are limited by the restrictions of Specification 3.1.3.5 and the Control Rod Position Limits defined in Specification 3.5.2.5.
i 3.5.2.4-Quadrant tilt:
1.
Except for physics tests, if quadrant tilt exceeds 4.92% power shall be reduced immediately to below the power level cutoff Isce Figures 3.5.2-1A and 5.5.2-181 kloreover, the power level cutoff value shall be reduced 2% for each 1% tilt in excess of 4.92% tilt. For less than 4 pump operation, thermal power shall be reduccJ 2% of the thermal power allowable for the reactor coolant pump combin-ation for each 1% t ilt in excess of '4.92%.
2.
Within a period of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. the quadrant power tilt shall be reduced tu less than 4.9'2% except for physics tests, or the following adjust-l ments in.<etpoints and limit s shall be made:
The protection system maximwn allowable setpoints (Figure a.
2.3-2) shall be reduced 2% in power foreach 1% tilt,
b.
The control red group and ApSR withdrawal limits shal'1 be reduced 2% in power for each 1% tilt ' n excess of 4.92%.
i The operational imbalance limits shall be reduce ~d 2% in power c.
for each li tilt in excess of 4.92%.
3.
If quadrant tilt is in excess of 25%, except for physics tests or diagnostic testing, the reactor will be placed in the hot shutdown i
condition. Diagnostic testing during power operation with a quad-rant power tilt is permitted provided the thermal power allowable for the reactor coolant pump combination is restricted as stated in 3. 5.2. 4.1 above.
4.
Quadrant tilt shall be monitored on a minimum frequency of once every two hours during power operation above 15% of rated power.
3.5.2.5 Cent rol rod positions:
1.
Technical Specification 3.1.3.5 (safety rod withdrawall does not prohibit the exercising of individual safety rods as required by Table 4.1-2 or apply to inoperable safety rod limits in Technical Speci fi ca t ion 3. 5. 2. 2.
2.
Operating rod group overlap shall be 25) 15 between two sequential groups, except for physics tests.
Ainendment No. J. [ 3, 47 j
e
.O 3.
Except for physics tests or exercising control rods, a) the control rod withdrawal limits are specified on Figures 3.5.2-1A, 3.5.2-1B and 3.5.2-1C for four pump operation and.on Figures 3.5.2-2A, 3.5.2-2B and 3.5.2-2C for three or two pump operation I
and b) the axial power shaping control rod withdrawal limits are specified on Figures 3.5.2-4A, 3.5.2-4B and 3.5.2-4C. If any of these control rod position limits are exceeded, corrective measures shall be taken immediately to achieve an acceptable control rod position. Acceptable control rod positions shall be attained within four hours.
4 Except for physics tests, power shall not be increased above the power level cutoff (see figures 3.5.2-1) unless the xenon reactivity is within 10 percent of the equilibrium value for operation at rated power and asymptotically approaching stability.
3.5.2.6 Reactor Power Imbalance shall be monitored on a frequency not 'to exceed two hours during power operation above 40 percent rated power. Except for physics tests, imbalance shall be maintained within the envelopes defined by Figures 3.5.2-3A, 3.5.2-38 and 3.5.2-3C.
If the imbalance is not within the envelopes defined by Figures 3.5.2-3A, 3.5.2-3B and 3.5.2-3C corrective measures shall be taken to achieve an acceptable imbalance.
If an acceptable imbalance is not achieved within four hours, reactor power shall be reduced until imbalance limits are met.
3.5.2.7 The control rod drive patch panels shall be locked at all times with limited access to be authorized by the superintendent.
88M The $ower-imbalance envelopes defined in Figures 3.5.2-3A, 3.5.2-3B and 3.5. -3C are based on 1) LOCA analyses which have defined the maximum linear heat rate (Sco Fig. 3.5.2-4) such that the maximum clad temperature will not exceed the final Acceptance Criteria and 2) the Protective Sys. tem Maximum Allowable SetMints (Figure 2.3-2).
Corrective measures will be taken immediately should the indicated, quadrant tilt, rod position, or imbalance be outside their specified boundary. Operation in a situation that would cause the final acceptance criteria to be approached should a LOCA occur is highly Onprobable because all of the power distribution parameters (quadrant tilt, rod position, and imbalance) must be at their limits while simultaneously all other engineering and uncertainty factors are also at their limits.* Conserva-tism is introduced by application of:
a.
Nuclear uncertainty factors b.
Thermal calibration c.
Fuel densification effects d.
Hot rod manufacturing tolerance factors e.
Fuel rod bowing The 25 25 percent overlap between successive control rod groups is allowed since the worth of a rod is lower at the upper and lower part of the stroke.
Control rods are arranged in groups or banks defined as follows:
' Actual operating limits depend on whether or not incore or excore detectors are used and their respective instrument and calibration errors. The method used to define the operating limits is defined in plant operating procedures.
AmendmentNo?f,g,33 as
m s
,.s 8
.)
~
~
5 e=
==
e..
e e9
=
b t.0 es ada.
A am gu
& M =
, w a e =
a E
=
M.- e
.e m
- w e.8 e z-e
- ei e-e w e a
a.
e.
e ea e e
/e-g. e e
a-=
~.
s.*
al 2 $
2
~
e o
N
- - a
=s e~
~
e a
%.j u ~
e mo g
A
~~
e
_ gu e
e e
d d_
2 ow A
e e.
8
~
~
a e
~
4e m
e a
ea a
.a.*
ee a
e, e
a e
~
e e.
~.
e.
e.
e o~ s e
3 e*
.ft w
=
.e.
a w
e*
e e
- a E
a e =
c
~
- a E-3 u
o e.
e.
e o.
a
=
a M.
- a ue a
" e a
om E
e e
a o
=
a.e 3
3
.o.
e*
u e
E o
e c.
e
~
e o
e e
e e
a e
e a.
R 2
2 m
2 o
e
~
w c
(85 $$$Z id t) Jaa04 t
1 I'
A: lend: lent No.
i
,[, 3 l 48b l
l l
l I
,m g
2 7
~
.as ea o
a.n.
~
=
=.
= _
C3 3
~
a 8-e-.
M C 88'l 2a
. Eo
.a.u u sJ r3 a
2' l
dp
=
~.
=.
tlo g
f.3
~
. ~.
2 4-J. 2 ~a e '-
- s fn
~
m.
~.
2
~
-o
~
~
-.=
~
~
a
- w w
o
.s
-Q
. 2
~ ~
2
~
OO E
=.
n_
4
.r
~
~
.t
~
M.
o
~
2
..2,
.=
s.
m
- S" a
=
u 2
o-
. 2
,a
~.
u
- =
~
-o 2
~~
e a
~
==
o
=-
_~
=.
-.=
2
-.=
o u
~
- o
~
- k 2
2 R
3 2
3 3
2 2
2A*
u (sur to u assov
.x
.E t
AmendmentSto.
3i 4Sbb ii
(
f 1
i r
W
=
~
E_
~
.S 3
In
' a rg
.E M,
-d W h 33
.~
.=
73
~.
~
~
=
~.
~
E
.t
~
o -.
3
,.e
.=
-g
=
.~
,4
~
M
~.
~.
~
a a
=
~
~
s.
~
. D w
~
3
. ~
2 g
~
=_
a.
3 f_u g
=
2 3
e 2
o.
2 u
d.
.s g
~
~
k
~
I 2
3
~
e
~
m.*
_o (as asst is s) aaned es Amendmentfo.
~
3g 48bbb
1IO E lit 10115 (113.182)
(lu.;12; (263.182) t
=
.3,....,
(07.05) k
^
se S
-5
?
18
_R
'd 58 J 33. 50)
Pf aul551M E Se a
48 (95.39) 30 28 (0,85) < /
/
Af1IalCI10 IDR 2 & 3 P9se n
8 c"o 0
28 40 60 80 100 120 140 150 180 208 220 240 260 200 300 R00 F05til0N LIBIIS FOR 2 & 3 Puer OPERAIIM IR05 810100 t 10 tips AREAN541 CTCil 3 Roa laces (s so) legase 3 5 2-24 m
a e
a a
0 25 58 15 100 Group 7 m
0 25.
50 75 100 Group 6 0
25 50 IS 100 Esoup 5 o.
4 e
Q O
P.e N
3e
-M M
m M
%.ee M
E tad Gud
==
ns 3
h me O J
- 40 em.
b M Es3=
es ag
- e..
es a b
M na.
C4 ene ta.
G M an em
- H E
=@
a 49 N
3OOe=
b
- O 09 --
G. N 4
is e un e'
a O
O O
M M
em
" ee Pe eu.
d O
N
- na N
S S 3 O
e d O ens er W
ce
.N.
M b
M Pe
= S
=
Po e e-
)
ee M
{
geg*
ind b
g
=
S a
w O
m an
- e d
e a
m es em O EL 3
3 O
e9 w
e O W en 9
as sup GS
.O S
n 6
a N
,n a
\\
m 5
N s o "w
O I
S
^
en g
a_
I t
Pe 1
m O
G En
- 8
- m ed a
es d
4 g
i O
3 i
a A e e
er
==.
53
- A G
d N
m Pe 6
l w
T a
S S
S S
S O
O O
O M
S G
- o e
me art G
=
em A
9 d"t ee OG (310tselly as g) Jased w ome G
Amendment No. a 4
=
31,
=
S
i 1
r-~~
=
u o
N e
==
M 1
4 E
trl.
mas M
G M e.
S.
nae
=3 S
.d 3 u GG h am
=
t3 he.
i=9 M
e6 4 M
N 3 4
E Ild C E w
- E M Q
- = b
- w E"'" has t
I
.d C3
==
G
+t N
- sf1 M sw Q& M 6ee Q
8'*
O w E 4 O
" C l
l M
e am O
O N
el1 pm A
O M
N nas 8
N A
ed
=
m 3
W=
. O =e H
dl N
M S
63 N
==
Nw w
e o
G N
a G
C N
N em O
O N
A Q
W e e W*
mm Q
m 3
w et
- C3
=
art
^
m "w
e W
O W
W e
em SE R
W M
3
=
U S
O C
ur3==
O -
41 er mag ens O
m O
E O
==.
S g a N
ef9 O==
G ufD N
me W
O N
O m
O q=p em W
N Oe E E M
o Q an a
C3 a
ef1 O
m ert w==
O E
De g
Q C3 4 m
W 3
d had E w
3 s=
mne O
"E u
w
" h O =
O g
a
=
l M E nee 2 N
a uf9 E &
C3 N
i i
t O
e O
=
3
=
=.
g
=
m e
N e.
W m a...
, i. u n...
4scec i
Amendment,No.
3i
POWER (5 0F 2568 MWt)
-20.1.102 i
100 - -
-21.0.92
'I' 90 - -
-26.7,80 80 -
> +10.7,80 70 - -
60--
PERMISSLBLE OPERATING REGION RESTRICTED 40 --
REGION REGION 30 --
20 --
10 - -
i i
i i
i i
i
-50 40 -30 20 10 0
10 20 30 40 50 Axial Power imoalance (5)
OPERATION POWER 15 BALANCE ENVELOPE FOR OPERATION FROM 0 TO 100 1 10 EFPD ARKANSAS CYCLE 3 Figure 3.5.2-3A A
Amendment No.
31 48d l
1
POWER (S of 2568 utt)
-16.9.102
^
100 23.7,92
+9.1, 9 2 g
-26.7.80 <
80 -
i +10.7,80 70 - -
60 - -
RESTRICTED PERNISSIBLE REGION O
IN RESTRICTED R
ON REGION 40 30 20 10 1
I I
I 1
L 1
I i
1 50 30
-20 10 0
10 20 30 40 50 Axial Power Imaaionca OPERATIONAL POWER IN8ALANCE ENVELOPE FOR OPERATION FRON 100 t 10 TO 250 t 10 EFP0 ARKANSAS CYCLE 3 Figure 3.5.2-35 A=endment No.
31.
I 48dd M
POWER (5 of 2568 Mut)
-27.3.102 -
100
-33.3,92 90
-36.3,80 80
,+16.6,80 70 60 PERillSSIBL E OPERATING RESTRICTED REGION REGION REGION 40 30 20 10 t
i i
i i
i i
i i
I 40
-30
-20
-10 0
10 20 30 40 50 Axial Power imoalance (5)
OPERATIONAL POWER IMBALANCE ENVELOPE FOR OPERATION AFTER 250 t 10 EFPO ARKANSAS, CYCLE 3 Figure 3.5.2-3C i
l Amendment No.
3 1.'
48ddd i
1
^
100 o
.0,102 MW REGION 90 35.5,92 43.5,80 E
2 m
100,70 60
_o 4t*
50 40 30 20 PERMISSIBLE OPERATING 10 REGION 0
O 10 20 30 40 50 60 70 80 90 100 APSR, *, Witnarawn APSR POSIT 10N LIMITS FOR OPERATION FROM 0 TO 100 1 10 EFPD ANO, CYCLE 3 Figure 3.5.2-4A i
AmendEent No. 31
~48f
-s s
100 29.0.102 RESTRICTED REGION 33.0,92 43.5.80 80 G
=
70
=
g 100,70 a
60
(
50 40 30 20 PERMISSIBLE OPERATING 10 REGION 0
i 0
10 20 30 40 50 60 70 80 90 100 APSR, *, letnarawn APSR POSITION LIMITS FOR OPERATION FRON 100 t 10 TO 250 t 10 EFP0 ANO. CYCLE 3 Figure 3.5.2 48 Amendsknt No.
31 48g
s fs 4-1 100 RESTRICTED REGION 47.5,92 80 61.0,80 E
a 70
==
gj 100,70 I;
60 o
h 50 PERMISSIBLE OPERATING REGION 40 30 20 10 0
i i
i O
10 20 30 40 50 60 70 80 90 100 APSR, 4 Withdrawn APSR POSITION LIMITS FOR OPERATION AFTER 250 i 10 EFPO - AhD, CYCLE 3 Figure 3.5.2-4C Amendment No. 11 48h l
_