ML20094G784
| ML20094G784 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 02/28/1992 |
| From: | TEXAS UTILITIES ELECTRIC CO. (TU ELECTRIC) |
| To: | |
| Shared Package | |
| ML20093C604 | List: |
| References | |
| TXX-92116, NUDOCS 9203030383 | |
| Download: ML20094G784 (30) | |
Text
__.
Attachsent 3 to TXX-92116 Page 1 of 13 4
i MARKED-UP TECHNICAL SPECIFICATIONS PAGES (NUREG 1399)
, to TXX-92116 3/4 3-1 through 3/4 3-12 1
2 i
)
9203030383 920228 fDR ADOCK 05000445 PDR i
I-..
Attachmont 3 to TXX 92116 NO CHANGE Pago 2 of 13 ON THIS PAGE 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.1 As a minimum, the Reactor Trip Systtm instrumentation channels and interlocks of Table 3.3-1 shall be OPERABLE.
APPLICABILITY:
As shown in Table 3.3-1.
ACTIO_N:
As shown in Table 3.3-1.
SURVEILLANCE REQUIREMENTS 4.3.1.1 Each Reector Trip System instrumentation channel and interlock and the automatic trip logic shall be demonstrated OPERABLE by the performance of the Reactor Trip System Instrumentation Surveillance Requirements specified in Table 4.3 '.
4.3.1.2 The REACTOR TRIP SYSTEM RESPONSE TIME of eatr Reactor trip function shall be demonstrated to be within its limit at least ace per 18 months.
Each test shall irclude at least one train such that t:th trains are tested at least once per 36 months and one channel per function +ach that all channels are testec at least once every N times 18 months where N is the total number of redandant channels in a specific Reactor trip function as shown in the
" Total No. of Chanrels" column of Table 3.3-1.
I COMANCHE PEAK - UNIT 1 3/4 3-1
l t
1 Al_lt. i 3.3-1 n
RtAC10R 1 RIP SYSTIM INSIRUMENIATION kh n
[f MINIMUM
,7 101Al NO.
CllANNL t S CliANNELS APPtICABLE
- s R
l'UNCT IONAL UNil Of CHANNilS 10 IRIP OPERABLE MODES ACil0N gy n
1.
1 2
1, 2 1
o c-2 1
2 3, 4',
5 9
8 d
k e
w 2.
Power Range, Neutron flux a.
High Setpoint 4
2 3
1, 2 2
b.
Low Setpoint 4
2 3
l',
7 2
3.
Power Range, Neutron flux 4
2 3
1, 2 2
liigh Positive Rate x"
4.
Power Range, Neutron flux, 4
2 3
1, 2 2
liigh Negative Rate n
C 5.
Intermediate Range, Neutron Ilux 2
1 2
1,2 3
6.
Source Range, Neutron flux A<c---Reactor Trip and Indication a..
4cStartup 2
1 2
2 4
b.
)
Shutdown 2
1 2
3,4,5 5
A D
(b.
Boron Dilution Flux Doubling 2
1 2
3
,4, 5 5
/
Overtemperature N-16 4
2 3
1, 2 12 8.
Overpower N-16 4
2 3
1, 2 12 d
9.
Pressurizer Pressure--low 4
2 3
I 6"
10.
Pressurizer Pressure liigh 4
2 3
1, 2 6
1ABit 3.3-1 (Continued) 8 x
E RfAC10R IHIP SYSIEM INSTRUMLNTATION 2#
9
~
2 ::
MINIMUM u9 101AL NO.
CilANNELS CilANNELS APPLICABLE o5 m
3E FUNCTIONAL UNIT Of CilANNLLS TO TRIP OPERABLE MODES AC110N
- 5 d
e c
11.
Pressurizer Water Level--ilig~h 3
2 2
1 6
ro z
12.
Reactor Coolant flow--Low a.
Single Loop 3/ loop 2/ loop in 2/ loop l'
6 7
any loop
~
5 b.
Two Loops 3/ loop 2/ loop in 2/ loop 19 6
any two loops 13.
Steam Generator Water 4/stm. gen.
2/stm. gen.
3/stm. gen.
1, 2 6"
w1 Level--Low-Low in any stm.
gen.
w d
9, 14.
Undervoltage--Reactor Coolant 4-1/hus 2
3 1
6 Pumps d
15.
Underfrequency--Reactor Coolant 4-1/ bus 2
3 I
6 Pumps 16.
Iurbine Trip g
a.
Low Fluid Oil' Pressure 3
2 2
1; 6
h.
Turbine Stop Valve Closure 4
4 4
1 10 i
17.
Safety Injection Input 2
1 2
1, 2 8
from ESFAS E5 29
,R g
1 m
9 i
l 1ABIl
- 3. 3-1 (Continued) no 7>
E RIACIOR 1 RIP SYSTEM INSIRilMENTAll0N
=
=.
n x
MINIMUM mg
,1 101Al NO.
CilANNIL5 CHANNLt.S APPIICABtL oo
??
IUNCIIONAL UNIT 01 CilANNII5 10 1 RIP OPERABLE MODE S ACI10N
- A ww h
18.
Reactor Irip System Interlot.ks g
5 a.
Intermettiate Range 2
1 2
2,,
/
y Neutron Ilux, P-6 0
-4 ti.
Low Power Reactor O
Irips Block, l'-7 g
i 1)
P-10 Input 4
2 3
1,2
/
2)
P-13 loput 2
1 2
I
/
c.
Power Range Neutron
'4 2
3 I
/
flux, P-8 m
D i
d.
Power Range Neutron 4
2 3
1
/
w1 Ilux, P-9 e.
Power Range Neutron 4
2 3
1, 2
/
Flux, P-10 19.
Reactor Trip Breakers 2
1 2
1, 2
- 8. 11 a
a 2
1 2
3, 4, 5" 9
20.
Automatic Trip and Interlock 2
1 2
1, 2 8
Logic 2
1 2
3, 4", S'd 9
d 25 2Q
,R E
i i
_ _ _ _ _ _ _ _ _ to VXX-92116 Page 6 of 13 TABLE 3.3 2 (L nti nued]
TABLE NOTATIONS aOnly if the reactor trip breakers happen to be in the closed position and the Control Rod Drive System is capable of rod withdrawal.
DBelow the P-6 (Intermediate Range Neutron Flux Interlock) Setpoint.
cBelow the P-10 (Low Setpeint Power Range Neutron Flux Interlock) Setpoint.
dAbove the P-7 (At Power) Setpoint
'The applicable MODES and ACTION statements for these channels noted in Table 3.3-2 are more restrictive and therefore, applicable.
- .sbove the P-8 (3-loop flow permissive) Setpoint.
9Above the P-7 and below the P-8 Setpoints.
he boron dilution flux doub W signals may be blocked during reactor startupd i Above the P-9 (Reactor trip on Turbine trip Interlock) Setpoint.
hDelet ACTION STATEMENTS ACTION 1 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, resture me inoperable channel to OPERABLE status withir 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or :- 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 />.
ACTION 2 - With the number of OPERABLE channels ore less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed proviced the following cnnditions are satisfied:
a.
The inoperable channel is placed in the tripped condition within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, b.
The Minimum Channels OPERABLE requiremtnt is met; however, the inoperable channel may be bypassed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for surveillance testing of other channels per Specification 4.3.1.1, and c.
Either, THERMAL POWER is restricted to less than or equal to 75% of RATED THERMAL POWER and the Power Range Neutron Flux Trip Setpoint is reduced to less than or equal to 85% of RATED THERMAL POWER within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />; or, the QUADRANT POWER TILT RATIO is monitored at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> per Specification 4.2.4.2.
COMANCHE PEAK - UNIT 1 3/4 3-5
Attachoent 3 to TXX-92116 Page 7 of 13 TABLE 3.3-1 (Continued)
ACTION STATEMENTS (Continued)
ACTICN 3 - With the numoer of cnannels OPERABLE one less than the Minimum Channels OPERABLE requirement and with the THERMAL POWER level:
a.
Below the P-6 (Intermediate Range Neutron Flux Interlock)
Setpoint, restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER above the P-6 Setpoint, b.
Above the P-6 (Intermediate Range Neutron Flux Interlock)
Setpoint but below 10% of RATED THERMAL POWER, restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER above 10% of RATED THERMAL POWER.
ACTION 4 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, suspend all operations involving positive reactivity changes.
ACTION 5 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel 4
to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or within the next hour open the reactor trip breakergs + suspend all operations involving positive reactivity changes. land verify either valve 1C5-B455 or valves 1C5-8560, FCV-111B, 1C5-8439, 1C5-8441, and 105-8453 are closed and secured in position, and verify this oositian at least once per 14 days thereafter.j With no channels OPERABLE complete @he above actions within 4 nours.jand verify the Iposiuons or tne above valves at least once per 14 days thereafter.
ACTICN 6 - with the number of OPERABLE channels oqi less than the Total NumDer of Channels, STARTUP and/or POWER OPERATION may proceed proviced the following conditions are satisfied:
a.
The inoperable channel is placed in the tripped condition within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and b.
The Minimum Channels OPERABLE requirement is met; however, the inoperable channel may be bypassed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for surveillance testing of other channels per Specification 4.3.1.1.
ACTION 7 - With less than the Minimum Number of Channels OPERABLE, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> deteimine by observation of the associated permissive i
annunciator window (s) that the interlock is in its required state l
for the existing plant condition, or apply Specification 3.0.3.
COMANCHE PEAK - UNIT 1 3/4 3-6
~. -. -
_ to TXX-92116 NO CHANGE
'9' O I I3 ON THIS PAGE TABLE 3.3-1 (Continued)
ACTION STATEMENTS (Continued)
ACTION 8 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, be in at least HOT STANDSY l
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 surveiliance testing per Specification 4,3.1.1 or maintenance, provided the other channel is OPERABLE.
ACTION 9 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status ivithin 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or open the reactor trip breakers within the next hour.
ACTION 10 - With the number of OPERABLE channels less than the Total Number of Channels, operation may continue provided the inoperable channels are placed in the tripped condition within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, ACTION 11 - With one of the diverse trip features (undervoltage or shunt t
trip attachment) inoperable, restore it to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or declare the breaker inoperable and apply ACTION 8.
The breaker shall not be bypassed while one of the diverse trip features is inoperable except for the time required for performing maintenance to restore the breaker to OPERABLE status, during which. time ACTION 8 applies.
ACTION 12 - With the number of OPERABLE channels o*4 less than the Total Number of Channels, STARTUP and/or POWEi OPERATION may proceed provided the following conditions are iltisfied; The inoperable channel is placed the-tripped condition a.
within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and b.
The Minimum Channels OPERABLE requirement is met; however, the. inoperable 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 Specifications-4.3.1.1 or 4.2.5.4.
I l
l COMANCHE PEAK - UNIT 1 3/4 3-7
x A..
n TABLE 4.3-1
.E %
- g U
REACTOR 1 RIP SYSifM INSTRUMENTAL 10N SURVEltLANCE REQUIREMENTS eg 9
TRIP o,,
- m A
ANALOG ACTUATING MODES IOR CHANNEL DEVICE WHICH ww CHANNEL CilANNI L OPERAT10NAL OPERATIONAL ACTUATION SURVEItLANCE
. *l FUNCTIONAL UNIT CHECK CAlIHRAil0N TEST TEST LOGIC TESI 15 REQUIRID, c-y x !
d d
d I
H 1.
Manual Reactor Trip H.A.
N.A.
N.A.
R(14)
N.A.
1, 2, 3, 4 5
O 2.
Power Range, Neutron Flux a.
liigh Setpoint 5
0(2, 4 ),
.Q N.A.
N.A.
1, 2 M( 3, 4 ),
Q(4. 6),
R(4, 5) b.
-Low $etpoint 5
R(4 )
5/U(1)
N.A.
N.A.
I',
2 g
3.
Power Range. Neutron Flux, N.A.
R(4)
Q N. A.
N.A.
1, 2 a
High Positive Rate 4.
Power Range, Neutron Flux. N.A.
R(4)
Q N.A.
N.A.
1, 2 High Negative Rate C
5.
Intermediate Range, 5
R(4, 5) 5/U(1)
N.A.
N.A.
I,2 Neutron Flux b
6.
Source Range, Neutron Flux 5 R(4, 13)
S/U(1), Q(9)
(12 N.A.
2.34,3 7.
Overtemperature N-16 5-D(2, 4)
Q N.A.
N.A.
I, 2 M(3, 4)
Q(4, 6)
R(4, 5) 8.
Overpower N-16 5
0(2,.4 )
Q N.A.
N.A.
1, 2 R(4, 5) l d
9.
Pressurizer Pressure--Low "5 R
Q(8)
N.A.
N.A.
I 10.
Pressurizer Pressure--High 5 R
Q N.A.
N.A.
1, 2
_= -..
?
1 TABLE 4.3-1 (Continued) n Sc E
REACTOR TRIP SYSTEM INSTRUMLNTATION SURVEILLANCE REQUIREMENTS
,y
= c.
n x
TRIP ANALOG ACTUATING MODES FOR 5i A
CHANNEL DEVICE WHICH o@
3 R-CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE *
CHECK CALIBRATION TEST TEST LOGIC TEST 15 REQUIRED t 'd
[
FUNCTIONAL UNIT
'+
L c-O t
.z d
~
- 11.
Pressurizer Water Level--
5 R
Q N.A.
N.A.
1 y
M High 4
d i
12.
Reactor Coolant Flow--Low 5
R Q
N.A.
M.A.
1 O
i 5
r
- 13. _ Steam Generator Water Level--
5 R
Q(8)
N.A.
N.A.
1, 2 Low-Low-d 14.
Undervoltage - Reactor Coolant. N.A.
R N.A.
Q(10)
N.A.
I w
-)
Pumps d
w i
1-a
.15.
Underfrequency - Reactor N.A; R
N.A.
Q N.A.
I Coolant Pumps i
I 16.
Turbine Trip a.
Low fluid Oil Pressure N.A.
R N A.
5/U(1, 10)
N.A.
l' b.
Turbine Stop Valve N.A.
v N.A.
S/U(1, 10)
N.A.
I'
' Closure i
j
- 17. ' Safety Injection Input.from.
N A.
N.A.
N.A.
R N.A.
1, 2 i
E5FAS i-1 18.
Reactor Trip System Interlocks
'a.
Intermediate Range N.A.
R(4)
.R N.A.
N.A.
2 EE
,n Neutron Flux,lP-6 xx e
2-r 1
f
lABLE 4.3-1 (Continued) n Sy REACTOR TRIP SYSTEM INSTRUMLNTATION SilRVEILLANCE REQUIREMENTS JR Q
NC' TRIP g.
,7, ANALOG ACTUATING MODE 5 10R
~8 y
CHANNEL DEVICE witICH
%3 CilANNI t CliANNL t OPERAI10NAL OPERATIONAL ACTUATION SURVfiliANC[
yy FUNCTIONAL UNIT ClifCK CAIIHRATION TEST TEST LOGIC IESI 15 REQUIRtp w
o x
-4 18.
Reat. tor Trip System.!nterlocks (Continued) g e
b.
Low Power Reactor Trips Block, P-7 c.
- 1) Power Range Neutron N.A.
R(4)
R N.A.
H.A.
1, 2 Flux P-10 gz hh y
- 2) Turbine First Stage N.A.
R R
N.A.
N.A.
I 4-Pressure P-13 mz T
?
g c.
Power Range Neutron N. A.
R(4)
R N.A.
N.A.
I g
Flux, P-8.
d.
Power Range Neutron N.A.
R(4)
R N.A.
N.A.
1 Flux, P-9 d.
Power Range Neutron N.A.
14 1 )
R N.A.
N.A.
1, 2 flux, P-10 8
d 19.
Reactor Trip Breaker _
N. A.
N.A.
N.A.
M(7, 11)
N.A.
1, 2, 3, 4 5
8 d
20.
Automatic Trip and Interlock N. A.
N.A.
N.A.
N.A.
M(7) 1, 2, 3, 4, $
Logic 8
21.
Reactor Trip Bypass Breaker N.A.
_N.A.
N.A.
M(IS), R(16)
H.A.
1, 2, 3, 4, 5 J
' to TXX 92116 Page 12 of 13 TABLE 4.3-1 (Continued)
TABLE NOTATIONS
'Only if the reactor trip breakers happen to be in the closed position and the Control Rod Drive System is capable of rod withdrawal.
Below P-6 (Intermediate Range Neutron Flux Interlock) Setpoint.
CBelow P-10 (Low Setpoint Power Range Neutron Flux Interlock) Setpoint.
dAbove the P-7 (At Power) Setpoint.
'Above the P-9 (Reactor trip on Turbine trip Interlock) Setpoint.
(1)
If not performed in previous 31 days.
(2) Comparison of calorimetric tu evcore power and N-16 power indication above 15% of RATED THERMAL POWER.
Adjust excore channel and/or N-16 channel gains consistent with calorimetric power if absolute difference of the respective channel is greater than 2%.
The provisions of Specification 4.0.4 are not applicable for entry into MODE 1 or 2.
(3) Single point comparison of incore to excore AXIAL FLUX DIFFERENCE above 15% of RATED 1HERMAL POWER.
Recalibrate if the absolute dif ference is greater than or equal to 3%.
For the purpose of these surveillance requirements, "M" is defined as at least once per 31 EFPD.
The provisions of Specification 4.0.4 are not apr'icable for entry into MODE 1 or 2.
(4) Neutron and N-16 detectors may be excluded from C*ANNEL CALIBRATION.
l (5) Detector plateau curves shall be obtained and evaluated.
For the l
Intermediate Range Neutron Flux, Power Range Neutron Flux and N-16 channels the provisions of Specification 4.0.4 are not applicable for entry into MODE 1 or 2.
(6)
Incore - Excore Calibtation, above 75% of RATED THERMAL POWER.
For the purpose of these surveillance requirements "Q" is defined as at least once per 92 EFPD, The provisions of Specification 4.0.4 are not applic-able for entry into MODE 1 or 2.
(7) Each train shall be tested at least every 62 days on a STAGGERED TEST BASIS.
l (8) The MODES specified for these channels in Table 4.3-2 are more restrictive and therefore applicable.
l 8
a 8
(9) Quarterly surveillance in MODES 3, 4, and 5 shall also include verifica-tion that permissives P-6 and P-10 are in their required state for exist-ing plant conditions by observation of the permissive annunciator window.
Quarterly surveillance shall include verification of the Boron Dilution g Alarm Setpoint of less than or equal to an increase of twice the count rate within a 10-minute period.
COMANCHE PEAK - UNIT 1 3/4 3-11
Attachatnt 3 to TXX-92116 Page 13 of 13 TABLE 4.3-1 (Continuea)
TABLE NOTATIONS (Continued)
(10) Setpoint verification is not applicable.
(11) The TRIT ACTUATING DEv!CE OPERATIONAL TEST shall indeoendently verify the OPERABILITY of the undervoltage and shunt trip attacnments of the reactor trip breakers.
(12) At least once per 18 mor,ths during shutdown, verify that on a simulated Boron Oilution Flux Doubling test signal the normal CVCS discharge valves close and the centrifugal charging pumps suction valves from the RWST open.
l (13) With the high voltage setting varied as recommended by the manufacturer, an initial discriminator bias curve shall be measured for each detector, Subsequent discriminator bias curves shall ba obtained, evaluated and compared to the initial curves.
(14) The TRIP ACTUATING DEVICE OPERATIONAL TEST shall independently verify the OPERABILITY of the undervoltage and shunt trip circuits for the Manual Reactor Trip Function, The test shall also verify the OPERABILITY of the Bypass Breaker trip circuit (s).
(15) local manual shunt trip prior to placing breaker in service, (16) Automatic undervoltage trip.
(I2) D e l e+ul COMANCHE PEAK - UNIT 1 3/4 3-12
.-__m.-.._.._..
_ m _.. _...
4
}
d 4
t 4
WPT-14385, " NOTIFICATION-0F-NONCONSERVATIVE BORON DILUTION ANALYSIS INPUT' ASSUMPTIONS" FEBRUARY 12, 1992 to TXX-92116.
I d
v=r
-is-...e--, - - - -
wwc r
n-- w w-e t -e war.
=- r e mw e e,-
-ee,c-s=w,+-+e
--e-+=1r.
w-=-rw+
-w----er+--+
e---+*e-g
-re,<em v-,5,w-3
+-
e,-
,- ~
er-i--r'y=*,wy-c, e,w,-
E di L-l-11-92-095 Westinghouse Energy Systems
$3y
,,,, g 33.g333 Electric Corporation l
Mr. W. J. Cahill, Jr., Executive Vice President February 21, 1992 Nuclear Engineering & Operations TU Electric Company S.0. No. TBX/TCX-4708 P. O. Box 1002 Glen Rose, Texas 76043 (No Response Required)
Attention:
S. M. Maier TV ELECTRIC COMPANY COMANCHE PEAK STEAM ELECTRIC STATION UNIT NUMBER 1 & 2 NOTIFICATION Of NONCONSERVATIVE BORON DILUTION ANALYSIS INPUT ASSUMPTIONS
References:
1.
CPSES-9204091, "CPSES Units 1 (sic) ICRR vs. Boron Concentration Data for Modes 3, 4, and 5 Boron Dilution Event Analysis," February 4,1992 2.
Standard Review Plan, NUREG-0800, Section 15.4.6, Rev. 1 - July 1981 3.
" Inadvertent Boron Dilution Events,"
January 31, 1985 4.
NS-TMA-2273,
" Boron Dilution Concerns at Cold and Hot Shutdown,' July 8,1980
Dear Mr. Cahill:
Introduction Based on recent conversations between Westinghouse Transient Analysis and Texas Utilities Electric (TVE), various nonconservatisms have been identified related to the input assumptions / boundary conditions (ICRR data ano flux-multiplication setpoint) in the analyses of the licensing-basis boron dilu", ion event.
The Modes 3, 4 and 5 licensing-basis boron dilution event analyses for the Comanche Peak plant utilizes inverse-count-rate-ratio (ICRR) deta and a flux-doubling (26) alarm to determine both the time of the 24 alarm and the time from the alarm to loss of plant shutdown margin following an inadvertent dilution event.
Based on information received from TUE, the Comanche Peak licensing-basis boron l
dilution event analyses (FbAR Section 15.4.6) are no longer bounding.
In addition, the safety analysis value for the flux-doubling signal / alarm in the plant Technical Specifications may no longer be valid with the current analysis methods.
FEB 21 '92 15:21 PAGE.002
Mr. U. J. Cahill 2
February 21, 1992 Summarv of the Issue The ICRR data used in the Comanche Peak analyses of the inadvertent boron dilution event is based on plant data which was the most limiting of all the data available in the late 1970s.
This data, along with a nominal 26 flux-multiplication setpoint, provided the basis for the detection mechanism following an inadvertent boron dilution event.
The flux-doubling signal actuates the protective function of the boron dilution mitigation system (BDMS) to isolate the dilution source and ini;iate a reboration of the RCS.
TVE has provided Westinghouse with Comanche Peak Unit 1 plant-specific ICRR data (Reference 1) that is nct bounded by the most-limiting known data from the 1970s.
Furthermore, the methodology is no longer conservative with respect to the 24 setpoint which includes no instrumentation uncertainties that ought to be applied to produce an equivalent
" trip setpoint" as presented in the Technical Specifications.
The effect on the Comanche Peak licensing basis is that the inadvertent boron dilution analyses for Modes 3, 4 and 5 are no longer bounding since Ccmanche Peak Unit I has demonstrated c worse characteristic for the ICRR data than was used in the ana'y. sis to determine the times of the flux-doubling signal.
Preliminary analyses using the Comanche Peak plant-specific ICRR data and a revised, more-conservative flux-multiplication setpoint do not yield acceptable results.
Justification for Continued Ooeration (JCO)
The purpose of the FSAR Section 15.4.6 analyses results is to show that the acceptance criterion delineated in the Standard Review Plan (Reference 2) is satisfied. The acceptance criterion is that the reactor should not be diluted to the point of total loss of plant shutdown margin. With respect to the generic implications of this issue, it should be noted that it has already been detcrmined that this is not a safety problem. The specific concern is that of a return to criticality in Mode 3, 4 or 5 and the resulting potential challenge to the integrity of the RCS due to a pressure increase. Thus, safety analyses have traditionally been performed to show that plant shutdown margin is not lost and that criticality does not occur. The NRC has reiterated the position (Reference 3) that the criteria in the Standard Review Plan (Reference 2) remain valid.
Based on safety analysis performed by the NRC in which a return to criticality was modeled, the noted generic letter (Reference 3) was written to address Generic Issue 22, Inadvertent Boron Dilution Events. This generic letter states that power excursions are possible due to a boron dilution event but that the excursion should be self limiting.
The NRC analysis which supports the information in the generic letter indicate that the self-limiting boron dilution power excursions are not expected to exceed the overpressurization criterion
, delineated in Reference 2.
While the specifics of the current concern of the BDMS at TUE differ from those which eventually resulted in the generic letter (Reference 3), the underlying concern (plant criticality) and the conclusions delineated by the generic letter still apply to TUE's concern today. The use of the BDMS at Comanche Peak serves to provide a means to detect a dilution event and automatically initiate corrective action in most instances. While not specifically documented, there are possible situations in which a flux-doubling signal would not detect an FEB 21 '92 15:22 PAGE.003
l Mr. W. J. Cahill 3
February 21, 1992 l
inadvertent dilution; e.g., some small dilution flowrate in which the flux increases so slowly that the BDMS microprocessor could not produce a 24 signal, and operator awareness is required for preventing criticality.
- However, it should be noted that even if criticality does occur, the BDMS will have actuated automatic reboration and the plant will return to a subcritical condition event with ng operator action.
The Comanche Peak BDMS provi?
a diverse mitigation feature for the boron dilution event as compared to plants not equipped with such an automatic system.
Available Compensatino Actions Since the Comanche Peak BDMS may not detect a dilution event and automatically initiate corrective action in all instances to prevent criticality, eManced awareness by the operator is recommended during activities where a potential dilution situation could exist.
The high-flux-at-shutdown alarm (HFSA) provides an indication of the amount of neutrons being emitted into containment at any given time.
By lowering the setpoint on the HFSA instrumentation, an early warning of an unexpected flux increase could be available to the operators when in the shutdown modes of plant operation. The operators need to be cognizant of the fact that electronic noise affects the HFSA instrumentation and spurious alarms at a reduced setpoint (from the normal 54 setting) could occur.
Increased surveillance may be considered on those valves in the CVCS which are along the potential dilution paths. Awareness of the sources and possible paths of an inadvertent boron dilution would greatly enhance the operator's chances of early mitigation of the event.
Likewise, the locking out of the valves along possible dilut4n paths when planned RCS dilution is not occurring would greatly decrease the chances of an inadvertent dilution.
Administrative controls may be implemented which place limits on the assumed dilution flowrate and RCS boron concentration such that there is sufficient time for operator action following an inadvertent dilution event. Originally developed in 1920 in response te an NRC concern for a boron dilution event while operating the plant on RHR (Modes 4 and 5), the operating procedures (Reference 4) have been implemented in instances where plant-specific and mode-specific analyses have not been performed and documented in the FSAR.
Recently revised, the procedures allow greater Mode 4 and Mode 5 operational flexibility. The key improvements in the
- revised procedures over the original ones (Reference 4) are summarized below.
1.
The revised curves are more likely to remain unchanged as a result of the fuel cycle reload verification process due to l
a.
removal of the dependency on a boron worth assumption, and l
b.
removal of the dependency on a specific range of RCS or critical boron concentrations through the introduction of the ratio of the two.
2.
The revised curves are specified as a function of RR flowrate and, therefore, support greater Modes 4 and 5 operational flexibility.
3.
The revised curves are based on a Mode-5 RCS volume assumption corresponding to the vessel drained down to the hot-leg midloop level, also supporting greater cold shutdown operations' flexibility.
FEB 21 '92 15:22 PAGE 004
~.,
Mr. M. J. Cahill 4
February 21, 1992 4.
The procedure is based on quantities typically known and readily available to the plant operator:
RCS boron concentration and critical boron concentration.
It should be noted that these procedures are designed to allow at least 15 minutes for operator action from the beginning of the dilution event, not from the time of an assumed alarm, until loss of shutdown margin.
In Mode 3, the volume available for dilution is much larger than in Mode 4 or 5; therefore, it is expected that 15 minutes is also available from the beginning of the dilution for the operator to detect the event.
The Reference-4 letter and operating procedures are found in Attachment 1; the revised procedures, recomended for use by TUE, are found in Attachment 2.
Any or all of these actions, as a supplement to the automatic BDMS, lead towards a reduction in the probability of a boron dilution event and/or a reduction in the probability that the reactor would reach criticality should an inadvertent dilution event occur.
TUE should also be aware that Westinghouse has comunicated the boron dilution analysis concerns to the WOG-RRG in a telecon on February 19, 1992.
Westinghouse will continue to have discussions with the WOG on this subject.
If there are any questions please contact Melita Osborne on 412-374-4481.
Very truly yours, WESTINGHOUSE ELECTRIC CORPORATION C 0 a-r v-4--
ri v
~
R. H. Owoc J. L. Vota, Manage Comanche Peak Projects FEB 21 '92 15:23 PAGE.005
~ _..,
Mr. M. J. Cahill 5
February 21, 1992 WPT-14385 ET-l4SL-0PL-II-92-095 cc:
W. J. Cahill, Jr.
- 5. C. Wood IL, lA VETIP Coordinator IL, 1A
- 1. A. Hope IL, lA W. G. Guldemond IL, IA L. Terry ll, IA S. M. Maier IL, lA D. Woodlan IL, lA D. Throckmorton IL, lA FEB 21 '92 15:23 PAGE.006
Letter NS-TMA-2273 - Original (1980) Operating Procedures c
A.
FEB 21 '92 15:23 PAGE.007
y<estinghouse Electric Corporation Power Systems msmesm
_kz es.-
July 8, 1980 NS.TMA-2273
'Mr. Victor Stello U.S. Nuclear Regulatory ComissionOffice of Nuclea Phillips Building 7920 Norfolk Avenue Bethesda, MD 20014
SUBJECT:
Boron Dilution Concerns at Cold and Hot Sh
Dear W. Stallo:
utdown On June 27, 1980 Ed Jordan of your staff concerned the potential for an inadvertent boo shutdewn and operating on the Residual Heat R 9.
ren dilution event while-This notif 1 is the text of the written notification s July 8,1960 which outlines potential W emoval System upplied 4;o our.
Attachment These interim actions are somewhat modifibasis customers on c
address these concerns.
reported.
contact D. W. Call atIf there are any questions regarding the at 412/373-5074 ached, please Very truly yours, sr n~
T. M. Anderson, Manager Nuclear Safety Department W
..s Attachment cc:
E. Jordan R. Woods N@Moq-LW 7ff.
FEB 21 '92 15:24
4 ATTACHMENT 1 t
On June 27,1980, you were notified of certain Westinghouse. concerns and recom-mended actions regarding the potential' for an inadvertent boron dilution event '
at cold or hot shutdown conditions while-on the Residual: Heat Removal System.
This notification was in -accord with Westinghouse determination that these con-cerns constitute an Unreviewed Safety Question under 10CFR Part 50.59.
The NRC-Office of Inspection and Enforcement was also notified on June-27,1980 that these concerns have generic applicability to Westinghouse-supplied nuclear power plants. Further clarification was made to the NRC Office of Inspection and-Enforcement on June-30.1980 that Westinghouse concerns are not applicable while -
the plant is greater than 55 shutdown.
This letter is-intended to fonaally doctment these concerns and to provide ad-ditional relevant information.. This letter also modifies the earlier recommend-ed actions by a more detailed specification.of applicable plant operating conditions.
Inadvertent boron dilution at shutdown has been generally regarded as an event which can be identified and teminated by operator action prior _ to a return to cri tical. Automatic protection ~ has not been a standard feature for Westinghouse plants. Westinghouse has recently been conducting a general investigation of this_ potential event relative to the licensing requirements imposed.on newer plants not yet in operation. - This investigation is sot yet complete.
However,-
it has been detennined that under certain shutdown conditions and.with certain-assumed dilution ratas.. adequate time for operator action to prevent a. return to critical may not be available.
The current Westinghouse evaluations are based'on plant conditions as noted below:
1.
The Reactor Coolant System effective voltme is. limited to the vessel and the active portions of the hot-and cold legs when on:RHR,- i.e.,: steam gen-
~
erator vehsees are not included..
2.
The plant is borated to a shutdown margin greater than:or equal to:l%
ak/k.
3.
Uniform mixing of clean and-borated RCS water is not assumed,1.e.. mixing 1 of the clean. injected water and the affected loop.is assumed but instan :
taneous, unifann mixing with the vessel, hot legs,Eand cold leg volumes-upstream of tne charging lines is. not asstaned.; Thus a " dilution front"-
moves through the cold legs, downcomer, and lower plenum to.the core vol-ume as a single.voltane front.
This results in subsequent decreases in shutdown margin 'due to dilution fronts moving through the active core L
region with a time constant equal-to the -. loop transit time when on RHR.
1 (five to seven minutes).
FEB 21 '92'l'5:24 PAGE.009
_ _.., _ _. _. _., _ _ _, _ _ _ _ ~.. _.., _, _.,. _. _.,., - _.
.. _ _.. If a return to critical occurs as a result of an inadvertent dilution, the fol-lowing potential concerns have been identified:
1.
A rapid, uncontrolled power excursion into the low and intermediate power ranges occurs, resulting in a power / flow mismatch due to the low flow (approximately 1 - 2% of nominal) provided by the RHR pumps.
2.
The potential exists for significant system overpressurization.
Pressure increases above the RiiR cut off head (ap' roximately 600 psig) furthee ac-p centuate the effects of a power / flow mismatch when all RCS (RHR) flow is lost.
An investigation of the adequacy of existing cold overpressurization protection systems is necessary in order to assess the full impact of this potential problem.
3.
The potential exists for limited fuel damage.
This is not currentlyla significant concern.
Preliminary evaluation indicates that the potes tial for exceeding DNB limits is low due to the cold initial operating con [di-tions.
Further investigation of this problem is underway.
The recommended interim actions to preveit or mitigate an inadvertent boron di-lution at shutdown conditions-are detailed in Appendix A.
If no cocked control rods are required, as specified in Figure A-1, the plant operator has fifteen minutes from the initiation of dilution event to ter1minate the event return to critical occurs.
It is the Westinghouse position that a fifteen ein-ute time interval from the initiation of the dilution to the time shutdown gin is lost is sufficient time for operator action.- If cocked control rods are required, the source range reactor trip provides positive indication for immed-iate operator action to tarininata dilution.
It is expected that the operator has available the following information for determination that a dilution event is in progress:
1.
Source W Neutron Flux with, i
HigWFlux at Shutdown Alarm set at half a decade above background, a.
b.
Use of the audible count rate indication to distinguish significant changes in flux, i.e., a doubling of the count rate.
Periodic, i.e., frequent surveillance of the Source Range meters per-c.
formed by the operator.
2.
Makeum Water System with, Status indication of the Chemical i
l' FEB 21-'92 15:24 PAGE.010
I
- l Indication of boric acid and blended (total) flow rate, or a.
b.
Indirstion of boric acid and clean uskaup flow rate, c.
CVCS valve position status lights, and d.
Reactor Makeup Water Ptap " running" status light.
The operator action necessary upon detemination that a dilution event is in pro-gress (by High Flux at Shutdown Alarm, Source Range Reactor Trip, "P-6 Available" indication, high indicated or audible count rates, or make up flow deviation alarms) is:
1.
Immediately open the charging /S!
open on receipt of an "S' signal) pump suction valves from the RW For 412 plants these ars LCV-112-D, E.)
Immediately close the charging /SI 2.
close on receipt of an "$" signal) pump suction valves from the VC For 412 plants these are LCV-112-8, C.)
3.
For two-loop plants, itsmediately open the charging suction valves from the RWST.
(For 212 plants these are LCV-113-8 and LCV-112-C. ) Also inusediate-ly close the charging suction valves from the VCT.
(For 212 plants these are LCV-113-A and LCV-112-8.)
Through the use of Appendix A and the above noted operator action requirements, Westinghouse is attempting to minimize the operational burden placed on the plant to prevent or mitigate an inadvertent dilution event while maintaining adequata safety margin.
Our investigation of this event is continuing. A detailed ana-lytical model of the system response to a dilution event at shutdown conditions is being developed and the potential for system overpressurization and fuel fail-ura. will subsequently be assessed.
The Westinghouse investigation is expected to be completed by September 15,1980.
suits of our efforts.
We will-keep you informed as to the re-l FEB 21 ' 3 7, 15:25 PAGE.011
-~
APPENDIX A Figure A-1, attached, provides the shutdown margin requirements as a function of Reactor Coolant Sptem boron concentration and maximum possible dilution flow rate.
Prior to use of this figure, the plant must detennine the maxima dilution flow rate of all charging pumps not rendered inoperable once the plant is placed on RHR.
To cover all modes, it should be assumed that the flow rate is based on pump runout unless there are flow limiting devices in the system (orifices, pip-ing resistances, etc.).
The Reactor Makeup Water piccp capacity may-be limiting in the detennination of the maximum possible dilution flow rate.
Figure A-1 notes areas of acceptable operation of three different dilution flow For a given dilution flow rate, if the RCS boron concentration and shu margin result in a point placed to the left of the flow rate line, no control rod bank withdrawal 1; necessary.
If the results place the plant to the right of the line, then either the shutdown margin must be increased such that the plant is moded to the area of acceptable operation, or 1% Ak/k in control rods must be withdrawn to provide additional shutdown margin.
The tripping of the withdrawn rods provides positive operator indication that a dilution event is in progress and additional time for o rutor termination of'the event.
In all cases, a shutdown nargin of 55 Ak/k (K
< 0.95 ued operation without a require $ Tor con) trol rod bank withdrawal.is corsider Figure A-1 is based on best estimate calculations for the "all rods in" configu-ration.
It is reconnended that the Westinghouse Nuclear Design Report for your plant be used as a reference in determining the RCS boron concentration with the appropriate conservatism to be used in the figure.
The Westinghouse Nuclear Fuel Division is available to provide assistance in meeting the constraints imposed by the Figure A-1 requirements.
Use of Figure A-1 is applicable any time there is boration/ dilution capability from the normal boric acid blending system.
The above procedure is not required if boration and/or makeup during celd and hot shutdown is perfonsed utilizing water from the RWST.
This requires that the nonnal dilution /boration path is isolated from the charging path.
Two means of lockout to isolate the charging path are available:
1.
Lock out Reactor Makeup Water Supply.
1 This is accomplished by valve 8338 for 212 plants, valve 8457 for 312 plants and valve 8455 for 412 plants.
OR:
FEB 21 '92 15:25 PAGE.012
2 2.
l.ock out valves between the boric acid blender and the YCT.
These are FCV-1118. FCV-1106. 8339. 8355. and 8361 for 212 plants; FCV-114A.
FCV-1138, 8454. 8441, and 8439 for 312 plantsi FCV-1118. FCV-1108. 8453.
8441. 8439 for 412 plants.
This reconnendation precludes the occurrence of an inadvertent dilution while borating or making up water frw the RMST inder these conditions.
l' L
FEB 21 PAGE.013
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FEB 21 92 15:27 PAGE.014
Revised (1989) Operating Procedures s
F B 21 '92 15:27 PAGE.015
1 Revised Interio Procedure I.
Procedure Parameters i
1.
Restesal Heat Removal System (RHRS) flow rate, gpe.
Range: 1000 to 6000 2.
Maximum predicted dilution flow rate (q), gpm.
Range: 100 to 300 4
3.
Dilution Factor (DLF). F;
- bc/ Chi Range: 0.6 to 1.0 4.
RCS critical boron conce* Ti91 usAt y ell rods inserted minus the most reactive RCCA stud. ed M the cord (Cbc), pps.
5.
Required RCS boron concentration to enfure 15 minutes are available from the beginning of a cilution event until loss of shutdown margin for a given RHR flow rate and dilution flow rate (Cbi),ppe.
Application Guidelines 1.
Detemine current or minimum intended RHRS flow rate.
2.
Determine maximum predicted dilution flow rate.
Use the closest curve corresponding to a value which is greater than or equal to this.
3.
Calculate RCS critical boron concentration (Cbc).
4 Use revised figure to find OLF = DLF(q,RHR flow).
5.
Calculate Chi - Cbc/0LF.
6.
Ensure RCS boron concentration is 2 Chi or 1% delta k/k in control rods is withdrawn (tripping of withdrawn rods provides positive operator indication of a bor$n dilution event in progress and preyides additional time for operator action). An al tive appeench is to further limit the maximum possible di flow rete.
III. Limits Applicability 1.
Applicable for Mode 4 (hot shutdown) and Mode 5 (cold shutdown) operation.
l 2.
If the calculated value of Chi is less than the RCS boron i
concentration required to meet the Technical Specification minimum shutdown margin requirement, borate to the latter.
FEB 21 '92 15:27 PAGE.016
!!!. Lidts of Applicability (cont.)
3.
Be RHR$ flow rates used for determining Chi should be the lowest projected flowrates during Mode 4 and 5 operation.
4 The maximum predicted dilution flow rate while operating on RHRS may be limited by the maximum capacity Reactor Makeup Water System to deliver unborated water to the charging pump suction, or it could L.e Ilmited by the maximum delivery rate of all charging pumps not rendered inoperable. In any case, the assumed l
flow rate should be based on pump runaut sniens there are flow limiting devices in the system.
5.
Use of the curves ensures availability of time for operator _
mitigation of an inadvertent boron dilution assuming no cocked rods. Use of the curves is not necess6ry if boration and/or makeup during hot and cold shutdown is performed utilizing water from the Refueling Water Storage Tank. This requires that the normal dilution /boration path is isolated from the charging path which precludes the occurrence of an inadvertent boron dilution.
6.
The minimum RHR flow rate considered is 1000 gpm.
The curves must not be extrapolated below this minimum value.
7.
The maximus RHR flow rate considered is 6000 spo. For any flow rate higher than this, use of the curves should be based on 6000 j
gps.
8.
These curves are derived for a minimum Mode 4 RCS volume assuming the vessel is filled and vented and a uttrimus Mode 5 RCS volume assuming the water level drained to the mid plane of the hot leg.
i i
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- TOTAL PAGE.017 **
92 15123 PAGE.017
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FEB 21
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i D. E. EISENHUT LETTER OF JANUARY 31, 1985, i
GENERIC LETTER 85 05 to TXX-92116 i
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UNITED STATES
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NUCLEAR REGULATORY COMMIS0lON
,e
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f W A$mNGT ON, D. C. 20%5
%, ' wy f t%
......f January 31, 1985 TO ALL PprSStlP!7En WATED REACTOP LICENSEES
- 1
- .g r,entl emen
- 3. P., CLEMENTS SUR.1ECT :
INAhyERTENT RORON DilllTION EVENTS (Generic Letter 85-05)
The purpose of this letter is to inform each licensee of operating pressurized water reactors of the staff position resulting from the evaluation of Generic issue P2, " inadvertent Boron Dilution Events" regarding the need for upgrading the instrumentation for detection of bornn dilution events in operat nig reactors.
A boron dilution event is considered as an anticipated operational occurrence which may necur at mnderate frecuency.
The staff has perfomed analyses of unmitigated bornn dilutinn events for a typical plant for each pressurized water reactor (PWR) vpf r.
The staff determined that while power excursions durino bnrer dilution events are possible if the operator does not take any action and sufficient volume of dilution water is available, the excursion shnuld be self-limiting.
The staff analyses indicate that these type of bornn dilution transients should not exceed the staff's acceptance criteria.
However, nur analyses also show that a few plants may experience slight nyerpressurizatinr in arcass of the 110% overpressure limit in the Desidual Heat Remnval system if the event occurs during a particular mode of operation.
In addition, the staff recogni?es that manv operating plants do not have distinct, positive alams to alert the operatnrs to boron dilution events but raly on nther devices such as audible count rate meters.
Other prnblems include lack of alarm redundancy and lack of technical specifications which would prevent nparatnrs frnm taking alarming devices out of service.
The staff also dnas not consider it prudent to credit operat rs with the ability to recognize a bornn dilutinn event and take the proper mitigative action within specified time limits in the absence of pnsitive bornn dilution alarms.
Cnnsidering all of the abnve factors and possible consaquences of boron dilution events, the staff has enneluded that the criteria in Section 15.4.6 of the Standard Peview plan are adequate and should continue to be anplied to plants currently underonino licensing review.
However, the consecuencas are not severe enough tn.ienpardi7e tha health and safetv of the public and do not warrant backfitting requirements fnr bnrnn dilution events at operatino reactors.
The staff will continue to review the analyses of the Roron Dilution Event in reinad applications tn assure that reasonable confidence is prnvided that nnerators can be expectad to take the right corrective action using the installed systems.
In summary, while the NRC will not reouire operatinn plant backfits +nr boren dilutirn events at this time, the staff would regard an unmitigated boron dilution evert at a sarious breakdown in the licensee's ability to control ite plant and Stronaly urges ecch licensee to assure itself that adeouate prntectim achinst bnron dilutinn avents exists in its plants.
65W we3ev
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2-This generic letter is provided for information only, and does not involve j
l any reportino requirements. Therefora, no clearance from tha Office of j
Panagement and Budget is required.
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irec r Darrell i, Division of itensino Ent bsure:
List of Generic Letters 1
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LIST Of RECENTLY 1550E0 GENERIC LETTERS t
GENERIC
}
LETTER NO.
SUBJECT DATE 84-15 Proposed Staff Actions to Improve and Maintain Diesel Generator Reliability 7/2/84 84-16 Adequacy of On-Shif t Operating Exper-ience for Applicants 6/27/84 84-17 Annual Meeting to Discuss Recent Develop-ments Regarding Operator Training, Qualifications and Examinations 7/3/84 84-18 Filing of Applications for Licenses and i
Amendments 7/6/84 84-19 Availability of Supplement 1 to NUREG-0933 "A Prioritization of Generic Safety Issues" 8/6/84 84-20 Scheduling Guidance for Licensee Submittals of Reloads that involve Unreviewed Safety Questions 8/20/84 84-21 Long Term Low Power Operation in PWR's 10/16/84 84-22 Not used 84-23 Reactor Vessel Water Level Instrumentation in BWRs 10/26/84 84-24 Clarification of Compliance to 10 CFR 50,49 Environmental Qualification of Electrical Equipment Important to Safety for Nuclear Power Plants 12/27/84 85-01 Fire Protection Policy Steering Committee Report 1/9/85 85-03 Clarification of Equivalent Control Capacity 1/28/85 For Standby Liquid Control Systems 85-04 Operator Licensing Examinations 1/29/85 85-05 Inadvertent Boron Dilution Events 1/31/85 i
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NUREG 0797 CPSES SER AND SUPPLEMENTS 1
i t 'to TXX-92116 Pages SER, 15 4 through 15 6 and SSER 23. 15-1 8
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(5) startup of an inactive reactor coolant pump at an incorrect temperature None of these transients are limiting; the most severe in terms of departure from nucleate boiling ratio and system pressure are the excessive load increase events.
Only slight changes in primary system pressure were calculated, and the departure from nucleate boiling ratio did not fall below 1.4.
The staff finds these results acceptable because they do not violate the appropriate limits.
15.2.2 Decreased Cooling Transients The applicant has analyzed the following events which produced decreased primary system cooling:
(1) loss of external electrical load (2) turbine trip (3) inadvertent closure of main steam isolation valves (4) loss of condenser vacuum and other events resulting in turbine trip (5) loss of nonemergency ac power to the station auxiliaries (6) loss of normal feedwater flow (7) partial loss of forced reactor coolant flow None of these transients are limiting; the most severe in terms of primary system overpressurizationistheturbinetriptransient}speakpressureismuchlower which results in a peak RCS pressure of approximately 2550 psia.
Because th than 110% of the RCS design pressure, the staff finds these results acceptable.
15.2.3 increased Core Reactivity Transients 15.2.3.1 Boron Dilution Events The principal means of causing an inadvertent boron dilution are the opening of the primary water makeup control valve and failure of the blend system, either by the controller or mechanical failure.
The chemical volume and control system (CVCS) is designed to limit, even under various postulated failure modes, the dilution rate to values which, will allow sufficient time for automatic or operator response (depending on the mode of operation) to terminate the dilu-tion before the shutdown margin is exhausted.
This dilution rate is indicated by instrumentation.
The applicant has analyzed the boron dilution event for all modes of operation.
Dilution During Refueling Uncontrolled boron dilution cannot occur during the refueling mode because all l
sources of unborated water are isolated in this mode.
15-4 i
. =.
l Oilution During Cold Shutdown In this mode, the dilution rate is limited to a possible maximum of 10 pam/ min.
This corresponds to the unborated primary water flow rate of 150 gpm.
T11s restriction is administrative 1y carried out and is to be implemented in the plant's Technical Specifications.
With the above dilution rate and a shutdown margin of 1% AK/K, a safety grade microprocessor connected to the source range monitors will sound an alarm in the control room, close the normal CVCS discharge valves, and open the centri-fugal charging pumps suction valves from the RWST, when the neutron count rate doubles over any 10 min-period.
This microprocessor continuously monitors then neutron count rate and com) ares it to the neutron count rate 10 min earlier.
When the count rate is dou)1ed, the system initials the valve position changes and sounds an alarm.
These actions by the microprocessor will prevent return to criticality.
Dilution During Hot-Shutdown and Hot-Standby The differences in assumptions between this mode of operation and the preceding case are (1) The shutdown margin is 1.6% AK/K for this case while it is 1% AK/K for the preceding case.
(2) The active RCS volume is 5000 ft3 for this case while it is 3500 f t3 for the preceding case.
Therefore, the dilution rate is limited for this case to 20 ppm / min.
This corresponds to 425 gpm of unborated water.
The microprocessor will provide the same protection for this_ case as for the preceding case.
Dilution During Startup and Power Operation In these modes of operation, the reactor is protected by various trips and rod L
insertion limit alarms such that enough time is afforded the operator to terminate the dilution event and return the reactor under control.
15.2.3.2.
Uncontrolled Rod Cluster Control Assembly (Rod) Bank Withdrawal i
From Zero Power Conditions i
The applicant's discussion of the consequences of an uncontrolled rod cluster i
control assembly bank withdrawal at zero power have been analyzed.
Such a tran-sient can be caused by a failure of the reactor control or rod control systems.
The analysis assumes a conservatively small (in absolute magnitude) negative Doppler coefficient and a positive moderator coefficient.
Further, hot zero power initial conditions with the reactor just critical are chosen because they are known to maximize the calculated consequences.
The-reactivity insertion rate is assumed to be equivalent to the simultaneous withdrawal of the two highest worth banks at maximum speed (45 in./ min).
i Reactor trip is assumed to occur on the low setting of-the power range neutron flux channel at 35% of full power (a 10% uncertainty has been added to the set-15-5
point value).
The maximum heat flux is less than the full power value, and the average fuel temperature increases to a value lower than the nominal full power value.
Evaluation Findings The staff has reviewed the reactivity worths and reactivity coefficients used in the analysis and concludes that conservative values have been used.
The staff has reviewed the calculated consequences of this design transient and concludes that they are acceptable.
Therefore, the staff finds that the requirements of GDC 20, which requires that protection be automatically initiated, and GDC 25 which requires that a single failure of the ]rotection system does not result In violation of specified fuel design limits, 1 ave been satisfied.
15.2.3.3 Rod Cluster Control Assembly Malfunctf ns The applicant has analyzed rod cluster control assembly misalignment incidents including a dropped full-length assembly, a dropped full-length bank, a mis-aligned full-length assembly and the withdrawal of a single assembly while operating at power.
Misaligned rods are detectable by:
(1) asymmetric power-distributions sensed by excore nuclear instrumentation or core exit thermo-couples, (2) rod deviation alarms, and (3) rod position Indicators.
A deviation of a rod from its bank by about 15 in. or twice the resolution of the rod position indicator will not cause power distributions to exceed design limits.
When one or rod position channels are out of service, additional surveillance will be required to ensure rod alignment.
In the event of a dropped assembly, the automatic controller may return the reactor to full power.
Analysis indicates that departure from nucleate boiling will not occur during this event.
For the case of dropped groups, the reactor is tripped by the power-range negative-neutron-flux-rate trip, and the reactor is shut down without violating fuel damage limits.
For cases where a bank is inserted to its insertion limit with a single rod in the bank stuck in the fully withdrawn position, analysis indicates that departure from nucleate boiling will not occur.
The staff has. reviewed the calculated estimates of the expected reactivity and power distribution changes that accompany postulated misalignments of' representative assemblies.
The staff has concluded t1at the values used in this-analysis-conservatively bound the expected values, including calculational uncertainties.
The inadvertent withdrawal of a single assembly requires multiple failures in the control system, multiple operator errors, or deliberate operator actions, combined with a single failure of the control system.
As a result, the single assembly withdrawal is classified as an infrequent occurrence.
The resulting transient is similar to that which results from a bank withdrawal, but the increased peaking factor may cause departure from nucleate boiling to occur in i
the region surrounding the withdrawn assembly.
Fewer than 5% of the rods in the core experience departure from nucleate boiling for such a transient.
15-6
15 ACCIDENT ANALYSIS 15.2 Moderate frequency Transients 15.2.1 Increased Cooling Transients The applicant has analyzed the following events which produced increase system cooling:
(1) decrease in feedwater temperature, (2) increase in feedwater flow, (3) excessive load increase, and (4) opening of system generator relief or safety valve.
None of these transients are limiting.
Since the Technical Spe-i cifications for the Comanche Peak Steam Electric Station (CPSES) require all f reactor coolant loops to be operational in Modes 1 and 2, the applicant stated that there is no need to analyze the event of startup of an inactive reactor coolant pump at an incorrect temperature.
Therefore, the staff concludes that sign limits are not violated.the analytical results are acceptable because the s 15.2.3 Increased Core Reactivity Transients 15.2.3.1 Boron Dilution Events The principal means of causing an inadvertent boron dilution are the ope the primary water makeup control valve and failure of the blend system.
chemical volume and control system (CVCS) is designed to limit the dilution The to terminate the dilution before the shutdown margin is exh to the FSAR makes changes to reflect the plant as-built conditions such as a Amendment 74 dilution flow rate of 167 gallons per minute (gpm) for all modes of operation and 4169 ft3 minimum reactor coolant system volume for dilution during hot shut-down and hot standby.
The staf f concludes that the applicant's analyses of or operator action to terminate the dilution before shutdown margin is exhausted 15.3 Infrequent Transients and Postulated Accidents 1E L '
Reactor Coolant Pump Locked Rotor Accident of one reactor coolant system pump rotor.The locked rotor accident w idly and the reactor would shut down as a result of a low-flow signal,The applicant reanalyzed this accident considering certain modified core values.
The system pressure of 2648 psia, which indicates that the boundary still maintains its integrity.
I the reanalysis of locked rotor accident is acceptable.The staff, therefore, concludes Comanche Peak SSER 23 15-1 v
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PWR BORON DILUTION EVENT ANALYSES CONDUCTfD BY LNAL DATED APRIL 13, 1984
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