ML19308B494
| ML19308B494 | |
| Person / Time | |
|---|---|
| Site: | Maine Yankee |
| Issue date: | 12/05/1979 |
| From: | Maine Yankee |
| To: | |
| Shared Package | |
| ML19257A270 | List: |
| References | |
| NUDOCS 8001030507 | |
| Download: ML19308B494 (16) | |
Text
,
Page 1 of 3 Attachment A Proposed Technical Specification Modifications Technical Item #
Specification Description of Change Reason for Change 1.
2.1.lb Modify coefficients for Thermal The new coefficients reflect Cycle 5 page 2.1-1 Margin / Low Pressure Trip as follows:
power distributions. Refer to Section A = 2004.3 6.0 of YAEC-1202 Il = 17.9 C = -10053 2.
Figure 2.1-la Replace Figure 2.1-la with the attached The revised At function of the TM/LP page 2.1-4 revised Figure 2.1-la trip reflects Cycle 5 power distribu-tions.
Refer to Section 6.0 of YAEC-1202.
3.
Figure 2.1-lb Replace Figure 2.1-lb with the attached The revised QRt function of the page 2.1-5 revised Figure 2.1-lh TM/LP trip reflects the Cycle 5 power distributions. Refer to Section 6.0 of YAEC-1202.
4.
Figure 2.1-2 Replace Figure 2.1-2 with the attached The revised Symmetric Offset trip and page 2.1-6 revised Figure 2.1-2.
pre-trip reflects Cycle 5 power distributions. Refer to Section 6.3 of YAEC-1202.
5.
3.10.A.3 Modify the req ui rement for available Results of the EOC-5 low power steam page 3.10-1 shutdown margin when critical from 2.9%
line break. Refer to Section 5.4.1 to 3.2%.
of YAEC-1202.
6.
3.10A Add following NOTE and asterisks to This change re-instates the flexi-h Technical Specification 3.10A:
bility to insert control rods 3" to r
withdrawn when positioned such that rods wear.
C3 N
are inserted within 4 steps f rom their C
upper electrical limit.
fs (Jh O
N u,
,1 W
r Yc n
[]
o Page 2 of 3 Attachment A (cont.)
Technical Item
- Specification Description of Change Reason for Change 7.
3.10.B.1 Modify 3.10.b.1 to read as follows:
Reflects Cycle 5 ECCS evaluation.
page 3.10-1 and 1.
the peak linear heat rate with Refer to Section 5.4.5 of YAEC-1202.
3.10-2 a pp ropri ate consideration of normal flux peaking, measurement calculational uncer-tainty (8%), engineering f ac tor (3%),
increase in linear heat rate due to axial fuel densification and thermal expansion (0.3% for Types E, G, H6 1 only), and power measurement uncertainty (2%) shall not exceed:
X Type J 13.5 kw/ft g ) 0.50 and CAB g 792 MWD /MTU X
14kw/ftg>0.50 and CAB > 792 MWD /MTU X
16 kw/ft 7 < 0.50 X
Types E,G,H,61:
14.0 kw/ft 7 > 0.50 16.0 kw/ft E < 0.50 L-X w ere g is fraction of core height and CAB is cycle average bu rn up.
8.
3.10. B.2 page 3.10-2, Delete 3.10.B.2 and re-number remaining Flux peaking augmentation factors are 3.10-3 and 3.10-4 sections of 3.10.B only applied in determining symmetric offset trip limits. Refer to Section 5.4.5 and 6.0 of YAEC-1202.
I
4 b
s t
Page 3 of 3 Attactmient A (cont.)
Technical Item
- Specification Desc ription of Change Reason for Change 9.
Fi gure 3.10-2 Delete Figure 3.10-2 Same as 8 above, page 3.10-10 10.
Figure 3.10-3 Replace Figure 3.10-3 with attached Reflects Cycle 5 ECCS evaluation.
page 3.10-11 revised Figure 3.10-3 Refer t o Sec tion 5.4. 5 of YAEC-12 0 2.
11.
Figure 3.10-4 Replace Figure 3.10-4 with attached Reflects Cycle 5 power distributiens.
page 3.10-12 revised Figure 3.10-4.
Refer to Section 4.3 of YAEC-1202.
12.
Figure 3.10-5 Replace Figure 3.10-5 with attached Reflects assumptions in the genera-page 3.10-13 revised Figure 3.10-5 tion of Cycle 5 RPS setpoint. Refer to Section 6.0 of YAEC-1202.
13.
3.15 Modify last sentence under Specifi-Existing speci fication refers to cat ion to read:
inappropriate section.
....with Technical Specification 5.9.1.6.
2.1 LIMITING SAFETY SYSTEM SETTING - REACTOR PROTECTION SYSTEM Applicability:
Applies to reactor trip settings and bypasses for the instrument channels monitoring the process variables which influence the safe operation of the plant.
Objective:
To provide automatic protective action in the event that the process variables approach a safety limit.
a Specification?
The reactor protective system trip setting limits and bypasses f or the requi red operable instrument channels shall be as follows:
2.1.1 Core Protection a) Variable Nuclear Overpower
<Q + 10, or 106.5 (whichever is smaller) for 10fQ$100
<20 f or Q<10.
where Q = Percent thermal or nuclear power, whichever is larger.
b) Thermal Margin / Low Pressure 1A QDNB + BTc + C, or 1835 psig (whichever is larger) where T
=c Id leg temperature, F c
A = 2004.3 B = 17.9 Cr -10053 QDNB
- Al x QRg At and QRg are given in Figure 2.1-la and 2.1-lb, respectively.
This trip may be bypassed below 10 percent of rated power.
c) The symmetric offset trip and pretrip function shall not exceed the limi ts shown in Figure 2.1-2, for three loop operation. This trip may be bypassed below 15 percent of rated power.
d) Low Reactor Coolant Flow 4
1 293 percent of 360,000 GPM (3 pump operation) w This trip may be bypassed below 2 percent of rated power.
2.1-1
t n
E
$YE Where: Q
=A X QR
-nm DNB 1
1 and p'a# P = 2004.3 + 17.9 T
-10053 unz vr in n u r.
o
--m T
= Cold leg temperature, F
o m
in
- s 1.30 1.25 d
5n Hu E
1.20 m
r, M
3 N
5' w n <c A
.4839 Y +1.0532
=
m o 1.15 1
I
- Cl*nc
$2 A
> mm 1
m C
8 1.10 g
m m
r.n e."
A =. I172 Y +.9871 o
1.05 g
7
- sn
/
1.00
.5
.4
.3
.2
.1 0
.1
.2
.3
.4
.5 U-L symmetric Offset Y7=A U+L w rs
{"* N D
C ua
Where: A x QR
=Q g
g DNB t fI and p P = 2004.1 Q
& 17.9 T
- 10053 var DNB in T
= Cold leg temperature in e
1.2
- --/
1.0 1, o ___,.
n
. 82 +
- 1. 0 -
.8 A
\\
.70 i
.6 QRt I!
/
f
.4' l
e.25
.2 00
.2
.4
.6
.8 1.0 1.2 Fraction of Rated Thermal Power s
k
'l MAINE YANKEE Thermal Margin / Low Pressure Technical Trip Setpoint Part 2 Figure Specifications (Fraction of Rated Thermal Fewer versus QR )
2.1-lb I
2.1-5
i 110 t
j
/
s
~
100
- ((
\\
j
\\
/
\\
90 t
/
.\\
c0
/
\\
80
\\
a
/
\\
s
'/
\\
3 E
70 l
i T
TRIP LIMIT l-I x
60 f
1-a 8
l-O l
50 l - LCO PRE-TRIP ALARM
\\
I.
,i 40 i
lj l
i{
l j
30 nl t l
l
{
I l
20 I
.6
.5
.4
.3.2
.1 0
.1
.2
.3
.4
.5
.6 Symmetric Offset = A
+B MAINE YANKEE Symmetric offset Function, Three Pump Operation Figure Technical 2.1-2 Specification 2.1-6
3.10 CEA CROUP, POWER DISTRIBUTION, MODERATOR TEMPERATURE COEFFICIENT LIMITS AND COOLANT CONDITIONS Applicability:
Applies to insertion of.CEA groups and peak linear heat. rate during operation.
Objective:
To ensure (1) core subcriticality af ter a reactor trip, (2) limited potential reactivity insertions frau a hypothetical CEA ejection, and (3) an acceptable core power distribution, moderator temperature coefficient, core inlet temperature, and reactor coolant system pressere during power operation.
Specification:
A.
CEA Insertion Limite 1.
When the reactor is critical, except for physics 4
tests and CEA exercises, the shutdown CEA's (Groups
,j A, B and C) shall be fully withdrawn.*
l 2.
When the reactor is critical, except for physics tests and CEA exercises, the regulating CEA groups (1 through 5) shall be no further inserted than the limits shown in Figure 3.10-l* for 3 loop operation.
3.
When the reactor is critical, the available shutdown margin with one CEA stuck out will not be less than 3.2% in reactivity. During low power physics testing at the beginning of a cycle, CEA insertion is permited such that the minimum shutdown margin is no less than 2% in reactivity.
i 4.
Operation of the CEA's in the automatic mode is not 4
permitted.
2 NOTE - CEA's shall be considered fully withdrawn i
when positioned such that rods are inserted within 4 steps from their upper electrical limit.
1 B.
Power Distribution Limits 1.
The peak linear heat rate with appropriate consideration of normal flux peaking, measurement-calculational uncertainty-(8%), engineering factor (3%), increase in linear heat rate due to axial'
^
fuel densification and thermal expansion (0.3% for Types E,G,H&I only) and power measurement uncertainty (2%) shall not exceed:
t t
3.10-1 w
b t
np---
r.-
g
t i
4 Type '.1:
l').5 kw/f t f > 0.50 and CAB $ 792 MWD /MTU 14 kw/ft {X > 0.50 and CAB > 792 MWD /MT i
16kw/ftf$0.50 Types E,G,H,6I:
14.0kw/ftf>0.50 16.0kw/ftf$0.50 X
where g is fraction of core height and CAB is cycle average burnup.
I Should either of these limits be exceeded, 5
immediate action will be taken to restore the linear heat rate to within the appropriate limit specifed above.
r T'P 2.
The total radial peaking factor, defined as F =F R R (1+Tq), shall be evaluated at least once a month during powcr operation above 50% of rated full
- power, i
P 2.1 F is the latest avadaMe untoMed ramal g
peak determined from the incore monitoring
]
system for a condition where all CEA's are at or above the 100% power insertion limit.
T is given by the following expression:
i 4
(Pb-Pd)2 i
(Pa-Pc) +
T
=2 l
i 9
(Pa+Pb)
+ Pc+Pd) 1 l
Pi = relative quadrant power determined from incore system for quadrant t 6
a when the incore system is operable and by
[
.[
Specification 3.10.B.4 otherwise.
h 1
i r
T 1
2.2 If the measured value of FR *** **k" value given in Figure 3.10-4, perform one of the following within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
i a) Reduce symmetric offset pre-trip alarm and trip band (Figure 2.1-2), thermal margin / low pressure trip limit (Figure 2.1-1 and Tech.
Spec. 2.1), and Excore LOCA monitoring limit
~
F R measured (Figure 3.10-3) by a factor i F
E"#*
R k
or 3.10-2.
y r
.r y
v
b) Reduce TilERMAL POW'ER at a rate of at least 1%/ hour to bring the combination of TilERMAL power T
and % increase in F E"#
R 3.10-5, while mai ntaining CEA's at or above the 100% power insertion limit; or c) Be in at least 110T STANDBY.
o 3.
Incore detector alarms shall be set at least weekly.
l Alarms will be based on the latest power distribution obtained, so that the peak linear heat rate does not exceed the linear heat rate limit defined in Specification 3.10.B.I.
If four or more coincident alarms are received, the validity of the alarms shall be immediately determined and, if valid, power shall be immediately decreased below the alarm setpoint.
3.1 If the incore monitoring system becomes ino perable,
l perf orm one of the f ol l owi ny, wi t hi n 4 E. F. P.ll.
a) Initiate a power reduction to < P at a rate of at least 1%/ hour where P(% of rated Power) is given by:
P = 0.85 (Linear heat rate permitted by Specification 3.10.B.1) x 100 Latest measured peak linear heat rate corrected to 100% Power while maintaining CEA's above the 100% power insertion limit and monitor symmetric offset once a shift to insure that it remains within + 0.05 of the value measured at the time when the above equation is evaluated. This procedure may be employed for up to 2 effective full power weeks, or b) Comply with the alarm band given in Figure 3.10-3.
If a power reduction is required, reduce power at a rate of at least 1%/ hour.
4.
The azimuthal power tilt, Tq, shall be determined prior to operation above 50% of full rated power af ter each refueling and at least once per day during operation chove 50% of full rated power.
Tq is given by the following expression:
(Da-Dc) (Db-Dd)
Tq = 2 (Da + Db + Dc + Dd)2 3
Di = signal from excore detector channel i. Tq shall not exceed 0.03.
3.10-3
r i
4 4
4.1 If the measured value of Tq 0.03 by <0.10, or nn excore channel is inoperable, assure that the total radial peaking factor (Ff)is within
^
the provisions of Specification 3.10.B.2 once l
l per shift.
4.2 If the measured value of Tq is > 0.10, opera-tion may proceed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> as long as F is maintained within the provisions of Specification 3.10.B.2.
Subsequent operation l
4 for the purpose of measurenent and to identify _
the cause of the tilt is allowable provided:
a) The THERMAL POWER level is restricted to
^
$20% of the maximum allowable THERMAL POWER level for the existing Reactor. Coolant Pump combination, and t
b) Reduce setpoints in accordance with Specifi-cation 3.10.B.2.2.
l 5.
The incore detector system shall be used to confirm power distribution, such that the peaking assumed in the safety analysis is not exceeded, after initial fuel loading and after each fuel reloading, prior to operation of the plant at 50% of rated 2
l power.
I l
6.
If the core is operating above 50% of rated power with one excore nuclear channel out of service, then the azimuthal power tilt shall be determined-once per shif t by at least one of the following means:
a) Neutron det'ectors (at least 2 locations per I
1 quadrant).
l b) Core-exit thermocouples (at least 2 thermo-couples per quadrant).
7.
The pre-trip limits of Figure 2.1-2 constitute Limiting Conditions of Operation.
i C.
CEA Drop Times i
i 1.
At operating temperature and 3 pump flow, the I
req ui rement for the maximum drop time of each CEA i
shall be not greater than 2.7 seconds from the time the holding coil is de-energized until the rod
]
reaches 90% of its full insertion.
3.10-4
_.. ~,. _...
r w
e S
This sheet purposely left blank MAINE YANKEE Flux Peaking Figure Technical Augmentation Factors 3.10-2 Specifications 3.10-10
i 100
[
> >792 MWJ 95,
.03 MTU 90 t
90.
.10 90, +.10 92k y
.03 3
S j
80 80,
.28 80,
.17 80, +.20 l
[
80, +.1
.a V
[
70, 3
70.
.40'
/ 70.
.30 70, +.28
.70, +.32 e
%o r
b 4.792 MWD et
~
S 60 MTU s
S l
O C
50 i
1 40
+ - - -.
.'S
.4
.3
.2
.1 0
.1
.2
.3
.4
.5 Symmetric Offset = A -1,,3 U
U+L e
MAINE YANKEE Excore Monitor S.O. Alarm Band for LOCA Limiting Technical Conditions of Operation When incore Monitors are Inoperable Figure Specifications 3.10-3 3.10-11
Note:
1.
This curve includes 10% calEulational uncertainty Ff = F[ x 1.03 2.
.g.
l.
Me.inu t eil Fit r;lusiilit he any.nn'ut eil by nn tmurement uncertainty (8%) before comparlalon to this curve 1.69 a
.._ y =. -
}_
l.68,
c 4
t-
.21.674 1.675 h_2--
__. )) _..
. Q
~g se 1.67.
..r.._..__.
l+i,,
.se 1,66 1.656 t
g.
m t=.e.
T 1.65 1.655 -
- 1. 6 5 0
.r 4 g
e e
1.64
- r : 1.
1.639
.i.
1.638
, - -.... = t:-
e n
. t.:.
g 1,63
-t F
_l.
- g 1.62
. -i
.o 1,61
~
~~
~~~
1.614 g
f 1.60,_
0 2
4 6
8 10 12 Cycle Average F.xposure (KMWD/MT)
MAINE YANKEE A110walde l'u rodded Radial Figure Technical Peak Versus Cycle Average Burnup 3.10-4 Specifications 3.10-12
Note: CEA's are to be maintained at or above the 100% power insertion limit when applying 3.10.B.2.2b l
e 100,
t j
i 90
'~
I i
t-l' i
n.
80 t-ce t
o E
70 a:
t' O
l M
60 50 0
2 4
6 8
10 12 T
% lucrease in F (above Figure 3.10-4)
R 6
MAINE YANKEE Technical Allowable Power Level vs.
Figure Specifications Increase in Unrodded Total Radial Peak 3.10-5 3.10-13
3.15 REACTIVITY ANOMALIES Applicability:
Applies to potent ial reactivity anomalies.
Objective:
To require evaluation of reactivity anomalies within the reactor.
Specification:
Following a normalization of the computed boron concentration a function of burnup, the actual boron concentration of as the reactor coolant shall be periodically compared with the predic ted value.
I f the di f f erence between the observed and predicted steady-state concentrations reaches the equivalent of 1% in reactivity, the Nuclear Regulatory Commission shall be notified and an evalaation as to the cause of the discrep-ancy shall be made and reported to the Nuclear Regulatory Commission in accordance with Technical Specification 5.9.1.6.
Basis:
To eliminate possible errors in the calculations of the initial reactivity of the core and the reactivity depletion rate, the predicted relation between fuel burnup and the boron concentration, necessary to maintain adequate control characteristics, must be adjusted (normalized) to accurately reflect ac ttal core conditions. When full power is reached initially, and with the CEA groups in the desired positions, the boron concent ration is measured and the predicted curve is adjusted to this poi nt. As power operation proceeds, the measured boron concentration is compared with the predicted concentration and the slope of the curve relating burnup and reactivity is compared with that predicted. This process of normalization should be completed af ter about 10% of the total core burnup. Thereafter, actual boron concentration can be compared with prediction and the reactivity status of the core can be continuously evaluated, and its occurrence would be thoroughly investigated and evaluated. The methods employed in calculating the reactivity of the core vs.
burnup, and the reactivity worth of boron vs. burnup, are given in the FSAR.(I)
References:
(1) FSAR, Section 3.4.7 9
1.15-1