ML19326C649
| ML19326C649 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 07/09/1975 |
| From: | ARKANSAS POWER & LIGHT CO. |
| To: | |
| Shared Package | |
| ML19326C646 | List: |
| References | |
| NUDOCS 8004250468 | |
| Download: ML19326C649 (18) | |
Text
[j i
i liegulator) D;chet File' 1-
.q.q-16 ATTACIBENT 1
~
]
This. listing denotes those pages to Appendix A, Technical Specifications which are revised as a result of re-evaluation of ECCS performance in t
conformance with the provisions of 10CFR 50.46 and of new ejected rod worth - calculations.
f*El 30 47 k.
48.
48a 48b 48c s
48d 48e 48f 48g t
l l
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e p
t F
8004250 % [
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fr 3.1.7.
Moderator Temperature Coefficient of Reactivity Specification
%e moderator temperature coefficient shall not be positive at power levels above 95% of rated power.
Bases A non-positive moderator coefficient at power levels above 9'5% of rated power is specified such that the maximum clad temperatures will not exceed the Final Acceptance Criteria based on LOCA analyses.
Below 95% of rated power the Final Acceptance Criteria will not be exceeded with a positive modera-tor temperature coefficient of +0.5 x 10-4 ok/k/F corrected to 95*6 of rated power. All other accident analyses as reported in the FSAR have been per-formed for a range of moderator temperature coefficients including +0.5 x 10-4 ok/k/F.
hhen the hot zero-power value is corrected to obtain the hot 95*. value, the following _ corrections will be applied.
1.
Uncertainty in isothermal measurerent - The measured moderator temperature coefficient will contain uncertainty owing to (a) 3 2 F in the AT of the base and perturbed conditions, 4 0
and (b) uncertainty in the reactivity measurement of 20.1 x 10-Ak/k.
Proper corrections will be added for these conditions to provide a conservative moderator coefficient.
2.
Doppler coefficient at hot zero power - During measurement of the isothermal moderator coefficient at hot zero power, the fuel tem-perature will increase by the same amount as for the moderator.
He measured temperature coefficient must be increased by 0.16 x 10-4 (ak/k)/ F to obtain a pure moderator temperature coefficient.
3.
Moderator temperature change - n e hot zero-power measurement must be reduced by 0.08 x.10-4 Ak/k/ F.
His corrects for the differ-ence in water temperature from zero power (532F) and 15'6 power (580F). Above this power, the average moderator temperature re-mains 580F.
4.
Fuel temperature interaction (power effect) - The moderator coef-ficient must be adjusted to account for the interaction of an aver-age moderator temperature with increasing fuel temperatures (as power increases).
His correction is 0.0022 x 10 aam /a% power.
It adjusts the moderator coefficient at 15% power to the coef-ficient at any power 1evel above 15% coefficient to 100% power is
~
(0.0022 x 10-4)(100% - 15*6) = 0.187 x 10-4 aam.
30
6.
-If a control rod in the regulating or axial power shaping groups 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 rods in the group are positioned such that the rod that was declared inoperable is main-tained within allowable group average position limits of Specifi-cation 4.7.1.2 and the withdrawal limits of Specification 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.
3.5.2.4 Quadrant tilt:
1.
Except for physics tests, if quadrant tilt exceeds 4%, power shall be reduced immediately to below the power level cutoff (see Fig-ures 3.5.2-1A, 3.5.2-1B, and 3.5.21C). Moreover, the power level l
cutoff value shall be reduced 2% for each 1% tilt in excess of the thermal power allowable for the reactor coolant pump combin-ation for each 1*. tilt in excess of 4%.
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 re-duced to less than 4%, except for physics tests, or the following adjustments in setpoints and limits shall be made:
a.
The protection system maximum allowable setpoints (Figure 2.3-2) shall be, reduced 2% in power for each 1% tilt.
b.
The control rod group withdrawal limits (Figures 3.5.2-1A, 3.5.2-1B, and 3.5.2-1C) shall be reduced 2% in power for each l
1% tilt in excess of 4%.
c.
The operational imbalance limits (Figure 3.5.2-3) shall be reduced 25 in power for each 1% tilt in excess of 4%.
i 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 condition. Diagnostic testing during power operation with a' quadrant 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.
i 3.5.2.5 - Control rod nasitions :
- 1.. Technical Specification 3.1.3.5 (safety rod withdrawal) does not prohibit the exercising of individual safety rods as required by Table 4.1-2 or apply to inoperable safety rod limits in Tech-nical Specification 3.5.2.2.
i 2.
Operating rod group overlap shall be 25% +5 between two sequen-tial groups, except for physics tests.
~
i 47
m' 3.
Except for physics tests or exercising control. rods, 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 Figure 3.5.2-2 for l
three or two pump operation.
If the 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 oper-ation at rated power and asymptotically approaching stability.
3.5.2.6 Reactor Power Imbalance shall be monitored on a frequency not to ex-ceed two hours during power operation above 40 percent rated power.
~ Except for physics tests, imbalance shall be maintained within the envelope defined by Figure 3.5.2-3.
If the imbalance is not within the envelope defined by Figure 3.3.2-3, corrective measures shall bc
.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.
Bases The power-imbalance envelope defined in Figure 3.5.-2-3 is based on LOCA an-alyses which have defined the maximum linear heat' rate (see Figure 3.5.2-4) such that the maximum clad temperature will not exceed the Final Acceptance g
Criteria. Corrective measures will be taken immediately should the indicated quadrant tilt, rod powition,' 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 improbable because all of the power l distribution parameters (quadrant tilt, rod position, and imbalance) must be at. their limits.while simultaneously all other engineering and uncertainty factors are also lat their-limits.* Conservatism is introduced by application of:
a.
Nuclear uncertainty factors l
b.
Thermal calibration c.
Fuel densification effects -
- d. 'llot rod manufacturing tolerance factors The 25' percent !5 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
. s t roke. 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 lindts is defined in plant operating procedures.
-48 l
Group Function 1
Safety 2
Safety 3
Safety 4
Safety 5
Regulating 6
Regulating 7
APSR (axial power shaping bank)
The rod position limits are based on the most limiting of the following three criteria: ECCS power peaking, shutdown margin, and potential ejected rod worth. As discussed above, compliance with the ECCS power peaking criterion.is ensured by the rod position limits.
The minimum available rod worth, consistant with the rod position limits, provides for achieving hot shutdown by reactor trip at any time, assuming the highest worth control rod that is withdrawn remains in the full out posi-tion _ (1).
The rod position limits also ensure that inserted rod groups will not contain single rod worths greater than 0.65% Ak/k at rated power.
These values h.tve been shown to be safe by the safety analysis (2) of the hypothetical rod. ejection accident.
A maximum single inserted control rod worth of 1.0% ak/k is all >wed by the rod positions limits at hot zero power.
A single inserted :ontrol rod worth of 1.0% ak/k at beginning of life, hot, zero power would result in a lower transient peak thermal power and, therefore, less severe environmental consequences than a 0.65% Ak/k ejected rod worth at rated power.
- Control rod groups are withdrawn in sequence beginning with group 1.
Groups 5, 6, and 7 are overlapped 25%.
The normal position at power is for groups 6 and 7 to be partially inserted.
The quadrant power tilt limits set forth in Specification 3.5.2.4 have been established within the thermal analysis design base using the definition of quadrant power tilt given in Technical Specifications,. Section 1.6.
1hese limits in conjunction with the control rod position limits in Specif-ication 3.5.2.5.3 ensure that design peak heat rate criteria are not exceeded during normal operation when including the effects of potential fuel densi-fication.
The quadrant tilt and axial imbalance monitoring in Specifications 3.5.2.4.6 and 3.5.2.5.4, respectively, will normally be performed in the plant com-puter.
The two hour frequency for monitoring these quantities will provide adequate surveillance when the computer is out of service.
During the physics testing program, the high flux trip setpoints are adminis-
- tratively set as follows to ensure that an additional safety margin is pro-
-vided:~
~
48a-
RCD POSITION LIMITS FOR 4 PUMF OPERATION APPLICABLE DURING THE PERIOD FROM 100 EFPD TO 250110 EFPO (PRIOR TO CONTROL ROD INTEFCHANGE),
(t77.4.t02) 100 (222.3.102)
POWER LEVEL CUT 90 OPERATION IN THIS REGION OFF IS NOT ALLOWED 80
[
5 (300.78)
- 70 U
- s w
s 5
O g 60 PERMISSIBLE OPERATING E
[
REGION eO (1o8.8,48) a-40 RESTRICTED
+
REGION N
30 20 (M.15) 10 0
0 50 100 150 200 250 300 Rod Index. % Withdrawn 0
25 50 75 100 0
25 50 75 100 i
t f
f I
I I
f f
f Group 5 Group 7 0
25 50 75 100 i
i i
i Group 6 Rod inder is the percentage sum of the withdrawl of Groups 5.6 and 7.
UNIT 1 ROD POSITION LIMITS Figure 3.5.2-1A 48b
RCD POSIi JN LIMITS FOR 4 PLM' OPERA..ON APPLICABLE DURING THE PERIOD FROM 250110 EFPD TP 435110 EFPD (ArfER CONTROL ROD TNTE'RC'F ANCFT e
(162.102)
(177.4.102)
(222.3.102) 100 90 OPERATION IN THIS REGION
' POWER LEVEL CUT 0FF (172.3.87)
IS NOT ALLOWED 80 (300.78)
- 70 C
p
- s M
g 60 PERMISSIBLE OPERATING E
8 REGION (162.50)
N 50 h
40 RESTRICTED j
3 REGION 8
N 30 20 g
(121.15) 10 (189.5.0) 0 I
I 0
50 100 150 000 250 300 Rad index. % Withdrawn 0
25 50 75 100 0
25 50 75 100 I
I I
l i
I 1
I Group 5 Group 7 0
25 50 75 100 l
i I
f f
Group 6 Rad index is the percentage sum of the withdrawl of Groups 5.6 and 7.
UNIT 1 ROD POSITION LIMITS 48c Figure 3.5.2-1B
ROD POSITI ' LIMIT FOR 4 PlMP CPERATIr" APPLICABLE DURING THE PERICO AFTER 435t10 EFPD.
(270.102)
(l62.102) 100 -
(173.8.90) 90 -
OPERATION IN THIS REGION (254.3.90) r IS NOT ALLOWED 80 POWER LEVEL
[
CUT OFF
- 70 U
3
- n 1Eg 60 PERMISSIBLE OPERATING N
REGION E
50 (is2.so)
RICTED g
40 3
REGION N
8 30 f'
\\
20 l
(121.15) 10 (189.5.0) 0 l
0 50 100 150 200 250 300 Rod index. % Witndrawn 0
25 50 75 100 0
25 50 75 100 t
i l
l 1
1 1
1 i
f Group 5 Group 7 0
25 50 75 100 1
1 l
i 1
Group 6 Rod index is the percentage sum of tne witndrawl of Groups 5.6 and 7.
UNIT 1 ROD POSITION LIMITS
~
48d F.igure 3.5.2-1C
R00 POSITION LIMITS FOR 2 AND 3 PUNP OPERATION AFTER 250210 EFPD
'"8
2' 100 -
90 OPERATION IN THIS REGION 5
IS NOT ALLOWED 0 80 5
E S
8 70 U
m E
[60 PERMISSIBLE OPERATING REGION
[
(162.50)
O 50 h"
[ RESTRICTED
~
S REGION 8
}
w
~
s 30 8
f y 20 E
(" '6)
(ta.is) 10 (149.s.0) 0 0
50 100 150 200 250 300 Rod Index. % Witnurawn 0
25 50 75 100 0
25 50 75 100 i
t i
i i
1 I
I I
I Group 5 Group 7 0
25 50 75 100 1
I I
f I
Group 6 Rod index is the percentage sum of the withdrawl of Groups 5.6 and 7 UNIT 1 ROD POSITION LIMITS 488 Figure 3.5.2-2
POWER LEVEL
-20.4
+7.1 100 --
-22.9
+14.8 80 __
-35 60 --
+27.8 40 --
20 --
I I
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t t
t
-40
-20 0
+20
+40 CORE IMBALANCE, %
OPERATIONAL POWER IMBALANCE ENVELOPE Figure 3.5 2-3 48f
m 21 20 19 s
u.
18 5
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E 17
/
/
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16
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E 15 E
e 14
.5 5
13 12 0
2 4
6 8
10 12 Axial location of Peak Power From Bottom of Core, ft LOCA LIMITED MAXIMUM ALLCWABLE LINEAR HEAT RATE Figure 3.5.2-4
)
1 4
48g
liUbiUijidu$id{f ATTACHMENT 2
't-9-7f c
Single Failure Analysis Per Branch Technical Position EICSB-18, " Application of the Single Failure Criterion to ~ Manually-Controlled Electrically-Operated Valves," safety systems were analysed to determine if a single failure could result in loss of capability to perform a safety function. Those systems which were re-viewed for the applicability of the single failure criteria are as follows:
~
- 1) Service Water System 2)
Decay Heat Removal System (Low Pressure Injection)
- 4) Makeup and Purification System (High Pressure Injection) 5)
Reactor Building Spray System 6)
Core Flooding System These systems were first reviewed to determine those manually-controlled electrically-operated valves which were present.
Upon determining these valves, analysis was performed individually to evaluate the potential con-sequences of these valves failing in an unsafe position.
The determination yielded three valve < which could potentially have serious adverse effects to safety system o ; ration if failure occurred.
These valves are as follows:
1 6 2) Core Flood Tank vent valves (CF-3A and CF-3B)
- 3) Service Water System common return valve to circulating water discharge fiume (CV-3824)
The normal position of the Cois Flood Tank (CFT) vent valves is the closed position.
If single failure w;re to ocr 1r, either through operator error or electrical fault, the valve would fail open.
If these vent valves were
~
to open before or during CFT discharge, the tank discharge flow rate could be less than that used in the ECCS analysis. However, it has been determined also that the time it would take for the tank to vent below Technical Spec-ification limits (575 psig) is more than enoug time for the failure to be recognized and corrected.
The valve is equippeo with a pressure reducing orifice so that pressure would bleed off at a very s'.sw rate. An alarm situation would occur when tank pressure reachad 585 psig.
This would occur within approximately 15 minutes -ioilewing valve failure.
It has been estimated approximately 30 total minutes would elapse before the tank pres-sure would reach 575 psig, allowing more than enough time for the valve to be closed manually. ' For this reason it -is felt no modification is needed.
' 'Ihe closure of the comon return valve to the circulating water discharge flume (CV-3824)(normally open) will result in 'the loss of cooling by both Low Pressure Injection (LPI) _ strings and to all ECCS components and pro-vides no discharge for-service. water.
For this reason, the breaker to-this valve will be locked open and tagged during normal operation.
Based on the above analyses and the proposed corrective action,. it is con-cluded that a single failure or operator error will not result in signifi-cantly adverse consequences to ECCS performance.
d d
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ATTACINENT 3 Submerged Valves Conservative analysis has shown that the depth of water which may accumulate in the reactor building following a LOCA will be S'9".
This yields an upper level of water in the reactor building at Elev. 345'3" (bottom of the reactor building is Elev. 336'6").
However, no motor operated valves are present between Elev. 345'3" and Elev. 357'0", thus a very conservative depth of water of 20.5 ft. may be considered.
Those motor operated -valves which may become submerged f-owing a LOCA, in-cluding that corrective action or modification needed i-ae event of sub-mergence, are as follows:
a) Core Flood Tank (T2A) Outlet Block Valves, CV-2415, Elevatior 338'6".
This valve is normally open and requires a key to unlock the switch which closes the valve. Following a LOCA, the core flood talk would be empty. Therefore, this valve would no longer be function-ally useful for short or long term cooling or containment isolation.
Thus, no modification is deemed necessary.
b) Reactor Coolant Quench Tank (T42) Discharge Isolation Valve, 3
CV-1053, Elevation 337 '3 /4".
This valve is normally closed.
Following a LOCA, this valve would serve no function in short or long term cooling. This valve does receive an ES-5 signal to close and is needed for containment isolation.
No modification is necessary, c)
Steam Generator Letdown Cooler Inlet Valves, CV-1213 6 CV-1215, Elevation 340'2".
CV-1213 is normally open and CV-1215 is normally closed.
Following a LOCA, these valves would not be used for short or long term cooling. The system is isolated by the outlet valves discucced in d) below.
No modification is
- needed, d) Steam Generator Letdown Cooler Outlet Valves, CV-1214 6 CV-1216, Elevation 338'7".
These valves are normally open.
Following a LOCA, these valves would not be used for short or long term cooling.
These valves receive an ES-1 signal to close for system isolation.
No' modification is deemed necessary.
1 e)
Intermediate Cooling Water Supply Inlet Valve to Steam Generator Letdown Coolers,.CV-2216 6 CV-2217, Elevation 342'4".
These i
valves control cooling water to the Letdown Coolers and are not needed for short or long term ECCS functions or containment isolation. No modification is needed.
f) Reactor-Building Sump to Decay Heat Removal System Block Valves, ii CV-1414-6'_CV-1415, Elevation 331'1 7/8" and 330'8", respectively.
These valves are normally open. They are required for long term ECCS functions. _The corrective action proposed is locking open the breakers to prevent closing.
During shutdown, the breakers
\\
ATTACINENT 3 Submerged Valves Conservative analysis has shown that the depth of water which may accumulate in the reactor building-following a LOCA will be 8'9".
This yields an upper level of water in the reactor building at Elev. 345'3" (bottom of the reactor building is Elev. 336'6").
However, no motor operated valves are present between Elev. 345'3" and Elev. 357'0", thus a very conservative depth of water of 20.5 ft. may be considered.
Those motor operated valves which may become submerged following a LOCA, i,n-cluding that corrective action or modification needed in the event of sub-mergence, are as follows:
a) Core Flood Tank (T2A) Outlet Block Valves, CV-2415, Elevation 338'6".
This valve is normally open and requires a key to unlock the switch which closes the valve.
Following a LOCA, the core flood tank would be empty. Therefore, this valve would no longer be function-ally useful for short or long term cooling or containment isolation.
Thus, no modification is deemed necessary, b) Reactor Coolant Quench Tank (742) Discharge Isolation Valve, 3
CV-1053, Elevation 337'3 /4".
This valve is normally closed.
Following a LOCA, this valve would serve no function in short or long term cooling. This valve does receive an ES-5 signal to close and is needed for containment isolation.
No modification is necessary.
c)
Steam Generator Letdown Cooler Inlet Valves, CV-1213 6 CV-1215, Elevation 340'2".
CV-1213 is normally open and CV-1215 is normally closed. Following a LOCA, these valves would not be used for short or long term cooling. The. system is isolated by the outlet valves discussed in d) below. No modification is needed.
d)
Steam Generator Letdown Cooler Outlet Valves, CV-1214 6 CV-1216, Elevation 338'7".
These valves are normally open.
Following a LOCA, these valves would not be used for short or long term cooling.
These valves receive an ES-1 signal to close for system isolation.
No modification is deemed necessary.
e)
Intermediate Cooling Water Supply Inlet Valve to Steam Generator Letdown Coolers, CV-2216 6 CV-2217, Elevation 342.'4".
These valves control cooling water to the Letdown Coolers and are not needed for short or long term ECCS functions or containment isolation. No modification is needed.
f)
Reactor; Building Sump to Decay Heat Removal System Block Valves, CV-1414 6 CV-1415, Elevation 331'1 7/8" and 330'8", respectively.
These. valves are normally open. They ar'e required, for long term ECCS functions. The corrective action proposed is locking open
.the breakers to prevent closing.
During shutdown, the breakers
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ATTACBIENT 4 1-9-7s Containment Pressure The following is 'a presentation of as-built passive heat sink data for Arkansas Nuclear One-Unit 1.
The overriding aspect of this data which makes the Babcock 6 Wilcox model a very conservative model as compared to ANO-1 is the value of net free volumes used.
It can be seers the 3 of net free volume more than ANO-1.
generic model has nearly 400,000 ft Also, ANO-l's small heat sink area as compared to the generic model con-tributes to a higher containment back' pressure during LOCA.
i j
e 4
l
- )
- i
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r
.a n _. (cont'd)
As-built heat sink data for Arkansas Nuclear One-Unit 1 3
Net: free volume 1,865,590 ft l
a) Reactor Building walls including concrete wall, steel liner and
- anchors :
exposed area 65,000'ft2 paint thickness 0.000564 ft s tec1 ' thickness 0.0208 ft concrete thickness 3.75 ft-b) Reactor _ Building dome including concrete, steel liner and anchors:
l l
exposed area 15,500 ft2 paint thickness 0.0005 ft steel thickness 0.0208 ft concrete thickness 3.25 ft c)
Painted internal steel 2
exposed area 92,024.00 ft paint thickness 0.000717 ft steel thickness 0.0312 ft d)
Unpainted internal steel (stainless and carbon) exposed. area 84,099 ft2 steel thickness.
0.0111 ft e)
Internal concrete exposed area' 94,110 ft 2 paint thickness 0.00169 ft concrete thickness 2.264 ft f) - Ihermodynamic Properties i
- Thermal Conductivity (BTU /hr-ft-F)
Material Value Concrete 0.9 Carbon Steel
- 26 Stainless Steel 10 P aints 0.083 to-1.5 y
4
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7.
~~~
~
3
.. Heat Capacity (B"IU/ft _p) blaterial Value Concrete 30 Carbon steel 56 Stainless steel 55.7 Paints 39.6 to 76.8 lg) - Del'ay - times, L sec.
Reactor Building Coolers 22 sec.
(no loss of off-site power)
Reactor Building Sprays 56 sec.
. (no loss of off-site power).
h) Building initial conditions-Temperature, F Basement 80 Dome 130 pressure, psia' 16.1 relative humidity, %'
basement 40 ceiling 15 i)
Outside ambient temperature (summer), F-flinimum-75-blaximum 95 l
l l
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--