ML19317G882
| ML19317G882 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 06/24/1977 |
| From: | SACRAMENTO MUNICIPAL UTILITY DISTRICT |
| To: | |
| Shared Package | |
| ML19317G879 | List: |
| References | |
| NUDOCS 8004020542 | |
| Download: ML19317G882 (24) | |
Text
'.
RANCHO SECO UNIT 1
- h TECHNICAL SPECIFICATIONS Safety Limits and Limiting Safety System Settings 2.
SAFETY LIMITS AND LlHITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS, REACTOR CORE AppiIcabi1Ity Applies to reactor thermal power, reactor power imbalance, reactor coolant system pressure, coolant temperature, and coolant flow during power operation of the plant.
Obj ect ive To maintain the integrity of the fuel cladding.
Spe :l fi cat ion 2.1.1 The combination of the reactor system pressure and coolant temperature shall not exceed the safety limit as defined by the locus of points established in Figure 2.1-1.
If the actual pressure / temperature point is within the restricted region the safety limit is exceeded.
2.1.2 The combination of reactor thermal power and reactor power imbalance (power in the top half of the core minus the power in the bottom half of the core expressed as a percentage of the rated power) shall not exceed the safety limit as defined by the locus of points (solid line) for the specified flow set forth in Figure 2.1-2.
If the actual-reactor-thermal power / reactor power-imbalance point is above the line for the specified flow, the safety limit is exceeded.
Bases The safety limits (gesented have been generated using(BAW-2 critical heat flux (CHF) correlation and the actual measured flow rate 2).
This development is discussed in the Rancho Seco Unit 1, Cycle 2 Reload Report, referenca (2). The 49 flow rate utilized is 104.9 percent of the design flow 037.8 x 106 lbm/hr) based on four pump operation. (2)
To maintain the integrity of the fuel cladding and to prevent fission product release, it is necessary to prevent overheating of the cladding under normal operating conditions.
This is accomplished by operating within the nucleate bolling regime of heat transfer, wherein the heat transfer coefficient is large enough so that the clad surface temperature is only slightly greater than the coolant temperature. The upper boundary of the nucleate bolling regime is termed " departure from nucleate boiling" (DNB ).
At this point, there is a sharp reduction of the heat transfer coefficient, which would result in high cladding temperatures and the possibility of cladding failure.
Although DNB is not an observable parameter during reactor operation, the observable pt.rameters of neutron power, reactor coolant flow, temperature, and p essure R004020 N
2.1-1 Proposed Amendment
- o. 49
..n
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RANCHO SECO UNIT 1 4
TECHNICAL SPECIFICATIONS Safety Limits end Limiting Safety. System Settings can' be related to DNB through the use of the BAW-2 correlation (1).
The BAW-2 correlation has been develcped to predict DNB and the location of DNB for
- axially uniform and non-uniform heat flux distributions.
The local DNB ratio (DNBR), defined as the ratio of.the heat flux that would.cause DN8 at a particular core location to the actual. heat flux, Is Indicative of the margin to DNB. The' minimum value of the DNBR, during steady-state operation, normal operational transients, and anticipated transients is ilmited to 1.30.. A DNBR of 1 30 corresponds to a 95 percent probability at a 95 percent confidence t
]
- level that DNB will not occur; this is. considered a conservative margin to l
DNB for all operating conditions. The difference between the actual core outlet pressure and the Indicated reactor coolant system pressure has.been considered in determining the core protection safety limits.
The difference in these two pressures is nominally 45 psi; however, only a 30 psi drop was I
assumed in reducing the pressure trip setpoints to correspond to the elevated location where the pressure is actually measured.
4 The curve presented in Figure 2.1-1 represents the conditions at which a 49 i
minimum DNBR of 1 30 is predicted for the maximum possible thermal power i
(112 percent) when four reactor coolant pumps are operating (minimum reactor coolant flow is 104.9 percent of 137.8 x 106 lbs/h r.).
This curve is based on the combination of nuclear power peaking factors, with potential effects of fuel densification and rod bowing, which result in a more conservative DNBR than any other shape that exists during normal operation.
The curves of Figure 2.1-2 are based on the more restrictive of two thermal limits and include the effects of potential fuel densification and rod bowing:
i 1.
The 1 30 DNBR limit produced by the combination of the radial peak, axial peak and position of the axial peak that yields no less than a 1.30 DNBR.
2.
The combination of radial and axial peak that causes central fuel melting at the hot spot.
The limit is 20.4 KW/ft.
Power peaking is not a directly observable quantity. and 'therefore limits have been established on the bases of the reactor power imbalance produced by the i
power peaking.
i The specified flow rates for Curves 1, 2, and 3 of Figure 2.1-2 correspond to the expected minimum flow rates with four pumps, three pumps, and one pump in each' loop respectively.
-i
. The ' curve of Figure 2.1-1 is.the most restrictive of all possible reactor coolant pump-maximum. thermal power combinations sitown in Figure 2.1-3 The maximum thermal power for three pump operation is 84.6 ' percent due to a power level trip produced by. the flux-flow; ratio 74.4 percent-flow x 1.050 =
- 78.1 percent : power plus -the maximum calibration ~ an'd instrument error. The
' 49 maximum thermal power for other coolant pump conditions is' produced in a similar manner.
2.1 ' Proposed Amendment No.~49 s
Vg g
g
~.
-e
~ -.,,
.m-
--.e
,,_r w
- E. -, e ra 4.-*
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4 RANCHO SECO UNIT:1 TECHNICAL SPECIFICATIONS Safety Limits and Limiting Safety System Settings For' Figure 2.1-3, a pressure-temperature point above and 'to the lef t of the 49 curve would result In'a DNBR greater than 1 30.
The 1 30 DNBR curve for four-pump operation is more _ restrictive than any other reactor coolant pump situation because any pressure /terperature. point above and to the lef t of the four-pump curve will be above and to the lef t of the other curves.
References (1)
Correlation" of Critical Heat Flux in a Sundle Cooled by Pressurized Watt.
BAW-10000, March, 1970.
49 (2)
Rancho Seco Unit 1, Cycle 2 Reload Report BAW.
~
2.1-3
-Proposed Amendment No. 49
Figure 2.1-1.
Core Protection Safety Limit, Pressure Vs Temperature 2400 no 2200 5c.
k a
m 0 2000 n.
U U
8 e
j 1800 1600 t
i i
560 5P0 600 620 640 i
Reactor Outlet Temperature. F 1
i
Figure 2.1-2.
Core Protection Safety Limits, Reactor Power Imbalance Thermal Power i.evel, %
120 (112)
(-28, 112)
- - 110 Acceptable
- - 100 Four-Pump operation
(+50, 95)
- 90
-28, 84.6)
(84.6)
(+16, 84.6)
(-f 0,80)
-- 80 Acceptable Three-and Four-
-- 70 Pump Operation
(+50, 67.6)
(57.4) -- 60
(+16, 57.4)
(-2 8, 5 7.4,
(-50, 52.6)
- - 50
'N (+50, 40.4)
-- 40 Acceptable Two,
Three, and Four-Pump Operation
(-50, 25.4)
-- 20
- - 10 t
i I
f
-60
-40
-20 0
20 40 60 Reactor Power Imbalance, %
Reactor coolant flow, Curve 106 lb/h 1
143.9 2
107.0 3
69.8 l
Figure 2.1-3.
Core Protection Safety Bases 2400 Of 2200 2
3 E.
8 1s s
o g;
2000 t
0 8
6 1800 1600 I
I I
i 560 580 600 620 640 Reactor Outlet Temperature, F l
- Power, Pumps operating Curve flow, 106 lb/h (type of limit) 1 143.9 (100%)
112 Four (DNBR Limit) 2 107.0 (74.4%)
87 Three (DNBR Limit) 3 69.8 (48.5%
59 One in each loop (quality limit) l l
l
RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Safety Limits and Limiting Safety System Settings 2.3 LIMITING SAFETY SYSTEM SETTINGS, PROTECTIVE INSTRUMENTATION Applicability a
Applies to instruments monitoring reactor power, reactor power imbalance, reactor coolant system pressure, reactor coolant outlet temperature, flow, number of pumps in operation, and high Reactor Building pressere.
Objective To provide automatic protection action to prevent any combination of process variables from exceeding a safety limit.
Specification 231 The reactor protection system trip setting limits and the perm ssible 8
bypasses for the instrument channels shall be as stated in table 2 3 and figure 2 3-2.
Bases T
The reactor protection system consists of four instrument channels to monitor each of several selected plant conditions which will cause a reactor trip if any one of these conditions deviates from a' pre-selected operating range to the degree that a safety limit may be reached.
The trip setting limits for protection system instrumentation are listed in table ~2.3-1.
The safety analysis has been based upon these protection system instrumentation trip set points plus calibration and instrumentation errors.
Nuclear Overpower A reactor trip at high power-level (neutron flux) is provided to prevent damage, to the fuel cladding from reactivity excursions too rapid to be detected by pressure and temperature measurements.
During normal plant operation with all reactor coolant pumps operating, reactor trip is initiated when the reactor power level reaches 105.5 percent 2-5
' Proposed-Amendment No. 49
RANCHO SECO UNIT I TECHNICAL SPECIFICATIONS Safety Limits and Limiting Safety System Settings of rated power. Adding to this the possible variation in trip set points due to calibration and instrument errors, the maximum actual power at which a trip would be actuated ceuid be 112 percent, which was used in the safety analysis. (4)
A.
Overpower trip based on flow and imbalance The power level trip set point produced by the reactor coolant system flow is based on a power-to-flow ratio which has been estab-lished to accommodate the most severe thermal transient considered in the design the loss-of-coolant flow accident from high power.
The analysis in section 14 demonstrates the adequacy of the specified power-to-flow ratio.
The power level trip set point produced by the power-to-flow ratio provides both high power level and low flow protection in the event the reactor power level increases or the reactor coolant flow rate decreases.
The power level trip set point produced by the power to flow ratio provides overpower DNB protection for all modes of pump operation.
For every flow rate there is a maximum permissible power level. and for every power level there is a minimum permissible low flow rate.
Typical power level and low flow rate combinations for the pump situations of table 2.3-1 are as follows:
1.
Trip would occur when four reactor coolant pumps are operating if power is 105 percent and reactor flow rate is 100 percent, or flow rate is 95 percent and power level is 100 percent.
2.
Trip would occur when three reactor coolant pumps are operating if power is 78 percent and reactor flow rate is 74.4 percent-or flow rate is 71 percent and power level is 75 percent.
3 Trip wculd occur when one reactor coolant pump is operating in each loop (total of two pumps operating) if the power is 51 percent and reactor flow rate is 48.5 percent or flow rate is 47 percent and the power level is 49 percent.
For safety analysis calculations the maximum calibration and instrumentation errors for the power fevel were used.
The power-imbalance boundaries are established in order to prevent reactor thermal limits from being exceeded. These thermal limits are either power peaking kW/ft limits or DNBR limits. The reactor power imbalance (power in the top half of core minus power in the bottom half of core) reduces the power level trip produced by the power-to-flow ratio so that the boundaries of figure 2 3-2 are produced. The power-to-flow ratio reduces the power level trip and associated reactor power reactor-power-imbalance boundaries by 1.05 percent for a 1 percent flow reduction.
2-6 Proposed Amendment No. 49
RANCHO SECO UNIT 1
-TECHNICAL SPECIFICATIONS Safety Limits and Limiting Safety System. Settings B.
Pump monitors The pump monicorf prevent the minimua core DNBR from decreasing below 1.3 by tripping the reactor due to (a) the loss of two reactor coolant pumps in one reactor coolant loop, and (b) loss one one or two reactor coolant 49 pumps dui.ng two pump operation.
The pump monitors also restrict the power level to 55 percent for one reactor coolant pump operation in each loop.
C.
Reactor coolant system pressure During a startup accident from low power or a slow rod withdrawal from high power, the system high pressure trip set point is reached before the nuclear overpower trip set point.
The trip setting limit shown in figure 2 3-1 for high reactor coolant system pressure (2355 psig) has been established to maintain the system pressure below the safety limit (2750 psig) for any design transient. (li The low pressure (1900 psig) and variable low pressure (12.fo T
- 5834) l49 out trip set point shown in figure 2.3-1 have been established.o maintain the DNS ratio greater than or equal to 1.3 for those design accidents that result in a pressure reduction. (2,3)
Due to the calibration and instrumentation errors the safety analy-sis used a tariable low reactor coolant system pressure trip value of (12.96 T
- 5884).
l49 out D.
Coolant outlet temperature The high reactor coolant outlet temperature trip setting. limit (619 F) shown in figure 2.3-1 has been established to prevent excessive core coolant temperatures in the operating range.
Due to calibration and instrumentation errors, the safety analysis used a trip set point of 620 F.
E.
Reactor Building pressure The high Reactor Building pressure trip setting limit (4 psig) provides positive assurance that a reactor trip will occur in the unlikely event of a steam line failure in the Reactor Building or a loss-of-coolant accident, even in the absence of a low reactor coolant system pressure trip.
F.
Shutdown bypass
.in order to provide for control rod drive tests, zero power physics testing, and startup procedures, there is provision 'for bypassing certain. segments of the reactor protection system.
The reactor protection system segments which can be bypassed are shown in 2-7 Proposed Amendment No. 49
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RANCHO SECO UNIT I TECHNICAL SPECIFICATIONS' Safety Limits and Limiting Safety System Settings table 2.3.1.
Two conditions are imposed when'the bypass is used:
I.
- By administrative control the-nuclear overpower trip set point must be reduced to a value 15.0 percent of rated power during reactor shutdown.
2.
A high reactor coolant system pressure trip set point of 1820 psig is automatically imposed.
The purpose of the 1820 psig high pressure trip set point is to prevent-normal operation with part of the. reactor protection system bypassed.
This high pressure trip set point is lower than the normal low pressure trip set point so that the reactor must be tripped before the bypass is initiated. The overpower trip set point of 15.0 percent prevents any significant reactor power-from being produced when performing the physics tests.
Sufficient natural circulation (5) would be available to remove 5.0 percent of rated power if none of the reactor coolant pumps were operating.
REFERENCES (1)
FSAR, paragraph 14.1.2.2.
(2)
FSAR, paragraph 14.1.2.7 (3)
FSAR, paragraph 14.1.2.8 (4)
FSAR, paragraph 14.1.2.3 (5)_
FSAR, paragraph 14.1.2.6 4
't A
2-8 Proposed Amendment No. 49
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TA8tE 2 3-1 REACTOR PROTECTION SYSTEM TRIP SETTING LIMITS One Reactor Coolant Pump Four Reactor Coolant Pumps Three Reactor Coolant Pumps Operating in Each Loop Shutdown Operating (Nominal Operating (Nominal (Nominal Operating Bypass Operatirig Power - 100% )
Operating Power - 75%)
Power - 49%)
I3}
1.
Nuclear power, t of rated, max.
105.,
105.5 105.5 5.0 3.
Nuclear power based on flow l.05 times flow minus 1.05' times flow minus 1.05 times flow minus Bypassed and imbalance, % of rated, max.
reduction due to reduction due to reduction due to Imbalance (s)
Imbalance (s)
Imbalance (s) 3.
Nuclear power based on pump NA NA 55 Bypassed _.
l 49
- moni tors, 2 of ra ted, max.
I%)
4.
High reactor coolant 2355 2355 2355 1820 system pressure, psig. ma).
5.
Low ructor coolant system 1900 1900 1900 Bypassed pressure, psig. max.
C 6.
variable low reactor coolant 12.96 T
-b
' '8
- 5834 12 96 T
- 5834 Bypassed l49-out out out system F
.?are, psig, min.
7.
Reactor coolant temp.
F., max.
619 619 619 619 8.
High Reactor Building 4
4 4
4 pressure, psig, max.
(1)
T 9'***
- "" ' ' E out (2)
Reactor coolant system flow, %.
(3)
Administratively controlled reduction set only during reactor rhutdown.
(4)
Automatically set when other segments of the RPS (as specified) are bypassed.
(5)
The pump monitors also produce a trip on: (a) loss of two reactor coolant pumps In one reactor coolant loop, and (b) loss of. one or two reactor coolant pumps during two pump operation.
Proposed Amendment No. 49
Figure 2.3-1.
Protective System Maximum Allowable Setpoints, Pressure Vs Temperature 2600 N
E.
2400 -
P = 2355 psig 8
g T = 619F
- g Acceptable Operation 2200 U
@Y 3
Unace.eptable o
e u
Operation oI l:j 2000 Q*
e 4
P = 1900 psig 1800 l
I I
I 540 560 580 600 620 640 Reactor Outlet Temperature, F
Figure 2.3-2.
Protective System !!aximum Allowable Setpoints, Reactor Power Imbalance Thermal Power Level. %
120
(-19,105)
- (+9, 105)
(105) g
'O.y2
-100 Acceptable
[
Four Pump
(+30, 94)
Y Operation - -90 o
V (78)
- -80
(-40, 76)
Acceptable Three-and- -70
(+30, 67.1)
Four-Pump Operation
-60 (51)
(-40, 49.1)
- -50 Acceptable Two,Three-
-40
(+30, 39.9) and Four -
Pump Opera-tion
-30
(-40, 21.9)
- -20 o
o
-10 m
8 a
e a
a t
1
-60
-40
-20 0
20 40 Power Imbalance, %
j l
l
/
-RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 352 CONTROL R0D GROUP AND POWER DISTRIBUTION LIMITS Applicability-This specification applies to power distribution and operation of control rods during power operation.
Objective To assure an acceptable core power distribution during power operation, to set a Ilmit on potential reactivity insertion from a hypothetical control rod ejection, and to assure core subcriticality after a reactor trip.
Specification 3 5.2.1 The available shutdown margin shall be not less than 1 percent Ak/k with the highest worth control rod fully withdrawn.
3 5.2.2 Operation with inoperable rods:
A.
Operation with more than one inoperable rod as defined in Specification A 7.1 and 4.7.2.3 in the safety or regulating rod banks shall not be permitted.
B.
If a control rod in the regulating and/or safety rod banks is declared inoperable in the withdrawn position as defined in Specification paragraph 4.7.1.1 and 4.7.1.3, an evaluation shall be initiated immediately to verify the existence of I percent Ak/k hot shutdown margin.
Boration may be initiated to increase the available rod worth either to compensate for the worth of the inoperable rod or until the regulating banks are fully with-drawn, whichever occurs first.
C.
If within one hour of determination of an inoperable rod as defined in Specification 4.7.1, it is not determined that a 1 percent Ak/k hot shutdown margin exists combining the worth of the Inoperable rod with each of the other rods, the reactor shall be brought to the hot standby condition until this margin is established.
D.
Following the determination of an inoperable rod as defined in Specification 4.7.1, all rods shall be exercised by a movement until indication is noted but not exceeding 2 inches within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and exercised weekly until the rod problem is solved.
E.
If a control rod in the regulating or safety rod groups Is declared inoperable per 4.7.1.2, power shall be reduced to 60 percent of the thermal power allowable for the reactor coolant pump combination.
3-31 Proposed Amendment No. 49
RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation.
F.
If a control rod in the regulating or axial power shaping groups is declared inoperable per Specification 4.7.1.2, operation above 60% of rated power may continue provided the rods in the group are positioned such that the rod that was declared inoperable ~is maintained within allowable group average position limits of Specification 4.7.1.2 and the withdrawal limits of Specification 3.5.2.5.c.
3.5.2 3 The worth of a single inserted control rod shall_not exceed 0.65 percent Ak/k at rated power or 1.0 percent Ak/k at hot zero power except for physics testing when the requirement of Specification 3.1.8 shall apply.
3.5.2.4 Quadrant tilt:
A.
Whenever the quadrant power tilt exceeds 3 percent, except for phys ics tes ts, the quadrant tilt shall be reduced to less 49 than 3 percent within two hours or the following actions shall be taken:
(1)
If four reactor coolant pumps are in operation, the allowable thermal power shall be reduced by 2 percent of maximum allowable power for each I percent tilt in excess of 3 percent.
The allowable thermal power is defined by Figures 3.5.2-1 and 3.5.2-2, where the power 49 level cut off may apply during transient xenon operation.
(2)
If less than four reactor coolant pumps are in operation, the allowable thermal power shall be reduced by 2 percent of maximum allowable power for each 1 percent tilt below the power allowable for the reactor coolant pump combina-tion.
(3)
Except as provided in 3.5.2.4.b, the reactor shall be brought to the hot shutdown condition within four hours if the quadrant tilt is not reduced to less than 3 l 49 percent af ter 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
(4)
The power range high flux set point shall be re..ced 2 percent of the maximum allowable flux for the RC pump combination for each I percent tilt in excess of 3 percent.
B.
If the quadrant tilt exceeds 3 narcent and there is simulta-l 49 neous ir.J: cation of a misaligned control rod per Specification 3.5.2.2, reactor operation may continue, provided power is reduced to 60 percent of the thermal power allowable for the reactor coolant pump combination.
C.
Except for phys ics tes ts, if quadrant tilt exceeds 8 percent, l
49 the reactor shall be brought to the hot shutdown condition within four hours.
3-32
RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation D.-
Whenever the reactor is brought to hot shutdown pursuant o 3 5.2.4a(3) or 3 5.2.4c above, _ subsequent reactor operation is permitted for the purpose of measurement, testing and correc-tive action provided the thermal power and power range high
' flux set point allowable for the reactor coolant pump combina-tion are restricted by a reduction of 2 percent of maximum allowabie power for each I percent tilt.
E.
Quadrant po'.~ r tilt shall be monitored on a _ minimum frequency of once ever two hours during power operation above 15 percent of rated power.
3.5.2.5 Control Rod Positions A.
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 Technical Specification 3.5.2.2.
B.
Operating rod group overlap shall be 25 percent, +5 percent between three sequential groups, except for physics tests.
C.
Position limits are specified for regulating and axial power shaping control rods.
Except for physics tests or exercising control rods, the regolating control rod Insertion / withdrawal limits are specified on Figures 3.5.2-1 and 3 5.2-2.
Also excepting physics tests or exercising control rods, the axial 49 powe-shaping control rod insertion / withdrawal limits are speci fied on Figures 3.5.2-3 and 3 5.2-4.
If any of these control rod position limits are exceeded, corrective measures shall be taken immediately to achieve an acceptable control rod position.
Acceptable control rod positions shall be attained within two hours.
D.
Except for physics test, power shall not be increased above the power level cut-off of 92 percent of the maximum allowable power level unless one of the following conditions is satis-j fled:
~
(1)
Xenon reactivity is within 10 percent of the equilibrium value for operation at the maximum allowable power level and asymptotically approaching stability.
(2)
Except for xenon free startup, when 3.5.2.5D(1) applies,
the reactor has operated within a range of 87 to 92 percent of the maximum allowable power for a period exceeding,
2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in the soluble poison control mode, i
3-33 Proposed Amendment No. 49
RANCHO SECO UNIT 1 TECHNICAL SPECIFICATI0tJS Limiting Conditions for Operation 3 5.2.6 Reactor power imbalance shall be monitored on a frequency not to exceed two hours during power operation above 40 percent rated power.
Except for physics test, imbalance shall be maintained within the envelope defined by Figures 3.5.2-5 and 3.5.2-6.
If the imbalance is not within t.ie envelope defined by Figures 3.5.2-5 49 and 3.5.2-6, corrective measures shall be taken to achieve an acceptable imbalance.
If an acceptable imbalance is not achieved within two 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 or his desig-nated representative.
Bases The power-imbalance envelope defined in Figures 3.5.2-5 and 3 5.2-6 is based on LOCA analyses which have defined the maximum linear heat rate (see Figure 49 3.5.2-7) such that the maximum clad temperature will not exceed the Final Acceptance Criteria.
Operation outside of the power imbalance envelope alone does not constitute a situation that would cause the Final Acceptance Criteria to be exceeded 'should a LOCA occur.
The power imbalance envelope represents the boundary of operation limited by the Final Acceptance Criteria only if the control rods are at the position limits as defined by Figures 3 5.2-1, 3.5 2-2, 49 3 5.2-3, and 3.5.2-4 and if a 3 percent quadrant power tilt exists. Additional conservatism is introduced by application of:
A.
Nuclear uncertainty factors.
B.
Thermal calibration uncertainty.
C.
Fuel densification effects.
D.
Hot rod manufacturing tolerance factors.
The conse vatise application c. the above peaking augmentation factors compen-sates fe-the perantial peaking penalty due to Fuel rod bo.v.
49 The 25 percent overlap between successive control rod groups is allowed since the worth of a rod is is e.
at the upper and lower part of the stroke.
Control rods are arranged in groups defined as follows:
Group Function i
Safety 2
Safety 3
Safety 4
Safety 5
Regulating 6
Regulating 7
Regulating 8
APSR (axial power shaping bank)
Control rod groups are wi thdrawn in sequence beginr ing with Group 1.
Group 5 is overlapped 25 percent with Group 6, and Group 6 is overlapped 25 percent with Group 7 The normal position at power is for Group 7 to be partially inserted.
49
RANCHO SECO UNIT I
-TECHNICAL SPECIFICATIONS Limiting Condi tions - for-Operation The minimum available rod worth provides for achieving hot shutdown by reactor
- trip at any time assuming the highest worth control rod remains'in the full out position. (1)
Inserted rod groups during power operation will not contain single rod worths greater than 0.65 percent.Ak/k.
This value has been shown to be safe by the safety analysis of the hypothetical rod ejection accident.
(2)
A single inserted control rod worth of 1.0 percent Ak/k at beginning of life, hot, zero power would result in the same transient peak thermal power and therefore the same environmental consequences as a 0.65 percent ak/k ejected rod worth at rated power.
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.
These limits in conjunction with the control rod position limits in Specification-3.5.2.5c ensure that design peak heat rate criteria are not exceeded during normal operation when including the effects of potential fuel densification.
The quadrant tilt and axial imbalance monitoring in Specifications 3 5.2.4 and 3 5.2.6, respectively, normally will be performed in the process computer.
The two-hour frequency for monitoring these qutr.tities will provide adequate surveillance when the computer is out of service.
Allowance is provided for withdrawal limits and reactor power imbalance limits to be exceeded for a period of two hours without specification violation.
Acceptable rod positions and imbalance must be achieved within the two-hour time period or appropriate action such as a red':ction of power taken.
Operating restrictions are included in Technical Speci fications 3.5.2.5d(l) and 3.5.2.5d(2) to prevent excessive power peaking by transient xenon.
The xenon reactivity must either be beyond the "undershoot" region and' asymptotically approaching its equilibrium value at rated power or the reactor must be operated in the range of 87% to 92% ci the maximum allowable power fo'r a period exceeding two hours in the soluble poison control' mode so that the transient peak is burned out at a lower power level.
During physics testing, additional safety margins are provided by administra-tively setting special reactor protection systems limitations.
During.the power ascension testing program, the following high flux trip settings will be set prior to increasing power to the next plateau:
Test Plateau Level %
Overpower Trip %
0
<5 15 50 40 50 75-95 90 100 100' 105.5 i
REFERENCES (1)
FSAR, Paragraph 3.2.2.1.2-(2)
. FSAR,. Pa ragraph 14.2.2.4
'3-33b-Prowsed AmenfhTEfLRA 00
Figure 3.5.2-la.
Rod Index Vs Power Level (O to 160 EFPD)
- 02) -
(300, 102) 100 Operation Not Allowed 90
- 8 #"
80 (270.1, 80)
Shutdown Limt u
70 Restricted 60 Region m
o 50 (149.7. 50)
(248.2, 50) f 40 30 Permissible Operating 20 Region (90, 15)
(0, 8.6) 10
-p (0.0 0
8 t
I I
I I
I I
I I
I i
0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 Rod Index, % Withdrawn 0
25 50 7,5 87.5 10p Bank 5 0
25 50 7f 87.5 1p0 Bank 6 0
25 50 75 100 Bank 7 1
Figure 3.5.2-lb.
Rod Index Vs Power Level (150 to 300 EFPD) 100 (249.1, 102)
(280.6, 102)
(300, 102)
(280.6, 92) 90 Shutdown Limit 80 Operation Not Allowed (267.8, 80)
In This Region y0 Restricted Region S 60 n
se 50 (186.0, 50 (248.2, 50) i8 40 Permissible Operating Region 30 20 (131.5, 15) 10. (0, 6~. 7)
(o-1 I
I I
I I
I I
I I
I I
0 O
20 40 60 80 100 120 140 160 180 200 220 240 260 260 300 Rod Index, I Withdrawn 0
25 50 75 87.5 100 Bank 5 0 25 50
[5 87.'5 100 3
Bank 6 0
25 50 75 100 Bank 7
F Figure 3.5.2-2a.
Core Imbalance Vs Power Level (0 to 160 EFPD) l 110 1
(-12.2, 102)
(12.2, 102) 100 l
90 80
(-21.5, 80)
(24.7, 80) 0 g
Permissible Operating egion S
60 N
w 50
(-30, 50)
(40.0, 50)
U l
40 30 l
20 10 -
0 I
I I
i 1
I l
l
-50
-40.
-30
-20
-10 0
10 20 30 40 50 Core Imbalance, %
Figure 3.5.2-2b.
Core Imbalance Vs Power Level l
(150 to 300 EFPD) l' 110 100
(-12.1, 102)
(14.6, 102)
(-12.1, 92)
(19.2, 92).
90
(-25, 80) 24.7, 80) 80 Permissible Restricted 70 Operating Region Region m
s N
60
'o
(-30, 50)
(40.0, 50) 50 3
8.
40 30 20 10 0
I I
I I
I I
I
-50
-40
-30
-20
-10 0
10 20 30 40 50 Core Imbalance, %
l l
l
Figure 3.5.2-4a.
APSR Withdrawal Vs Power Level (0 to 160 EFPD) i 110 (22.3, 102) 100 (22.3, 92) 90 -
80 (35.2, 80)
Permissible 70 Operating Restricted Region Region Rg 60 -
o 50 (60, 50)
C e
40 30 -
20 -
10 -
0 I
I I
I I
I 0
10 20 30 40 50 60 70 80 APSR Withdrawal, %
Figure 3.5.2-4b.
APSR Withdrawal Vs Power Level (150 to 300 EFPD) 110-(25.5, 102) 100 90 (35.2, 80) 80 70 Restricted E"
Region m
b 60 Permissible Operating o
Region 50 (60, 50) 3 2
40 30 20 10 0
t i
I I
I i
0 10 20 30 40 50 60 70 80 '
l APSR Withdrawal, %
. _..