ML19317G778
| ML19317G778 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 10/07/1977 |
| From: | Reid R Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19317G757 | List: |
| References | |
| NUDOCS 8004010604 | |
| Download: ML19317G778 (24) | |
Text
_
/
[>'A Gli\\
UNITED STATES g
NUCLEAR REGULATORY COMMISSION j '4
?lg'tyj y 'o ijf,i}513!S/!
WASritNGTON, D. C. 20655 i
W, v.
...a ]
r SACRAJ' ENTO MUNI,CIPAL UTILITY DISTRICT DOCKET NO. 50-312 RANCHO SECO NUCLEAR GENERATING STATION AMENDNENT TO FACILITY OPEPATING LICENSE Amendment No.14
. License No. DPR-54 E'
l.
The Nuclear Regulatory Commission (the Commission) has found that:
- y A.
The applications for amendment by Sacramento Municipal Utility District (the licensee) dated June 24, 1977, t
and August 5, 1977, comply with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the applications, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activit';s authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compli-ance with the Coanission's regulations; D.
The issuance.of this amendment will not be inimical to the common defense and security or t~ the health and safety of the public; and E.
The. issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
8004010hOf
_ 3.
2-
),
k.
2.
Accordingly, the license is amended by changes to the. Technical i
Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-54 is hereby maended to read as follows:
(2) Technical _ Specifications 7
The Technical Specifications contained in Appendices A and B, as revised through Amendment No.14, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specificatior.s.
3.
This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
],,,
h [.L Robert W. Reid,. Chief
/
1 Operating ReactoFs Branch #4 Division of Operating Reactors
'e Attachnent:
Changes to the Technical Specifications Date of Issuance: October 7, 1977 t
8 i
n 9
+
. n
e r 1.r ATTACHMENT TO LibENSE AMENDMENT NO. la l
FACILITY OPERATING LICENSE NO. DPR-5A i
DOCKET NO. 50-312 Revise Appendix A as follows:
I Re, move Paaes Insert Paces l
2 2-2a 2.1 2.1-3 Figures 2.1 2.1 -3 Fiqures 2.1 2.1 -3 2-3 & 2-4 2-4 2 2-9 2 2-9 Figures 2.3-1 & 2.3-2 Figures 2.3-1 & 2.3-2 3 3-33b 3 3-33b o
[-
Figures 3.5.2 3.5.2-3 Figures 3.5.2 3.5. 2-7 i
6-3 6-3 8
'I i
Changes on the revised pages are shown by marginal lines.
Pages 2-4 and 2-8 are unchanged and are included for convenience only.
a l
I l
l i
l i
l RANCHO SECO UNIT 1
[
I TECHNICAL SPECIFICATIONS I
Safety Limits and Limiting' Safety System Settings l
2.
SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMlTS, REACTOR CORE Apf,licabi l i t y Applies to reactor thermal power, reactor power imbalance, reactor coolant system pressure, coolant temperature, and coolant flow during power operation of the plant.
t m
m D
D I d
Obj ect Ivo Q v E.
To maintain the integrity of the fuel cladding.
p"g-}
~
v
-1 Specification 2.1.1 The combination of the reactor system pressure and co:.,lant 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 tne safety limit is exceeded.
2.1.2 The combination of reactor thermal power and reactor, power imbalance w-(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 IP 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 flux The safety limits gesented have been generated using(BAW-2 critical heat (CHF) correlation I and the actual measured flow rate 2).
This development is i
discussed in the F.ancho Seco Unit 1, Cycle 2 Reload Report, referenca (2).
The 6 lbm/hr) based flow rate utilized is 104.9 percent of the design f. low (137.8 x 10 on four pump ope ra t ion. (2)
To traintain the integrity of the fuel cladding and to prevent fission product releese, it is necessary to prevent overheating of the cladding under normal operating conditions.
This is accomplished by operating within the nucleate boiling regime of heat t r..ns fer, wherein the heat transfer coefficient is large enough so thct the clad surface temperature is only slightly greater than the coolant temperature.
The upper boundary of the nucleate boiling regine is terned " departure from nutleate boiling" (DNB ).
At this point, there is a sharp reduction of the heat t ransfer coef ficient, which would result in high cladding terperaturcs and the possibility of cladding failure.
Although DNB is not an observable parameter during reactor operation, the observable parameters of neut ron power, reactor coolant flow, temperature, and pressure 2.1-1
.1%. -
n
M ANCHO SECO UNIT 1 l
T uHNICAL SPECIFICATIONS Safety Limits and Limiting Safety System.Settirigs ccn be related to DNB through the use of the BAW-2 correlation (1).
The BAV-2 correlation has been developed to predict DNB and the location of DNB for axially uniform and non uniform heat flux distributions.
The local DNB ratio (DfER), defined as the ratio of the heat flux that would cause DNB 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 limited to 1.30.
A DNBR of 1.30 corresponds to a 95 percent probability at a 95 percent confidence level that DNB will not occur; this is considered a conservative margin to 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 assumed in reducing the pressure trip setpoints to correspond to the elevated location where the pressure is actually measured.
The curve presented in Figure 2.1-1 represents the conditions at which a minimum DNBR of 1.30 is predicted for the maximum possible the mal power (112 percent) when four reactor coolant pumps are operating (minimum reactor coolant ficw is 104.9 percent of 137.8 x 106 l bs/h r. ).
This curve is based on the combination of nuclear porter peaking factors, with potential effects of fuel densification and rod bowing, which result in a more conservative DNBR than any cther shape that exists during normal operation.
1 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:
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 cent ral fuel melting at the hot spot.
The limit is 20.4 KW/ft.
Power peaking is not a directly observable quantity and theref ore limits have been established on the bases of the reactor power imbalance produced by the t
power peaking.
The specified flow rates for Curves 1, 2, and 3 of Figure 2.1-2 correspond to the expected minimum flow rates wi th four pumps, three pumps, and one pump in each loop rcspectively.
1 The curve o' Figure 2.1-1 is the most restrictive of all possible reactor coolant puup-maxiram thernal poveer combinations shown in Figure 2.1-3 l
The r3ximum 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 perccnt poacr plus the maximum calibration and instrument error.
The maxinu'm thermal power for other coolant pump conditions is produced in a similar e ner.
i dim 9
DJ 2.1-2 u1 fcend.nent No.
14 g
g
,[
- u, 11
. d 1_
_a
RANCHO SECO UNIT 1 l
- ltCHNICAL SPECIFICATIONS I
Safety Limits and Limiting Safety System Settings For Figure 2.1-3, a pressure-temperature point above and to 'the lef t of the curve would result in a DNBR greater than 1.30.
The 1.30 DNBR curve for four-pu.np operat ion is more restrictive than any other reactor coolant pump situation because any pressure /tenperature point above and to the lef t of the four pun.p curve will be above and to the left of the otner curves.
Re fe ren,c es (1)
Correlation of Critical Heat Flux in a Bundle Cooled by Pressurized Water, BAV-10000, March, 1970.
(2)
Rancho Seco Unit 1, Cycle 2 Reload Report BAW.
I 4
4 5
9 2.1-3 g
g-}
l vE-O
- A:andment fio. '14
I I
Tigure 2.1-1.
Core Procedefon Safety Limit, Pressure Vs Temperature
~
2 /.00 oc 2200
-e rm O
g m
8 2000 t_.
a 6
e 1500 o
U 1600
____y i
560 550 600 620 640 Reactor Outlet Temperature, F I
l l
D l
O o A
<D _ 1[ 4 Anandrant No.
14 3
m t
Figure 2.1-2.
Core Protection Safety Limits, Reactor Power Imbalance T1.crr.al Power i.evel, %
(112)
(+16, 112)
(-28, 112)
,. - ~.
-- 110 Acceptable
-- 100 Four-Pump Operation
(+50, 95)
-- 90
-28, 64.6)
(84.6)
(+16, 84.6)
(-Y),80)
-- 80 Acceptable Three-and Four-
. -- 70 Pump Operation
(+50, 67 o)
(57.4) -- 60
(-20, 57.4
(+16, 57.4) l
- 50
(+50, 40.4) a
- 40 Acceptable Two,
Three, and Four-
[
Pump Operation
/
-- 30
/
l-50, 25.4)
- 20
- 10
,.,.~J...__
- l...
-- L I
-60
-40
-20 0
20 40 60 Reactor Power 1r. balance, %
Reactor coolant flow, Curve 106 lb/h 1
143.9 2
107.0 m
m 3
69.8 D
DJ l
I ov 1LUL
- m. tent tb.
14 l
- ~ ~ ~ - - ~
{
Figure 2.1-3.
Core Protection Safety Btses 2400
"-~
f 2200 2
3 2.
8 1
I2l og 2000 U
O 5
/
1800 l_
1 1600 _
_580_
600 620
__ 640 560 Reactor Outlet Temperature, F Reactor coolant
- Power, Pumps operating
,ge_o f _ limit)
Curve flow 10E
(
lb/_h, n
1 143.9 (100%)
112 Tour (DNBR Limit) 2 107.0 (74.4%)
87
'Ihree (DNBR Limit) 3 69.8 (48.5%
59 One in each loop
, quality limit)
(
D ovI IHIL Amendeent No.
14
"' RAIJCHO SECO UNIT 1 l
e cCHNICAL SPECIFICATIONS Safety Limits and Limiting Safety System Settings i
2,2 SAFETY LIMITS, REACTOR SYSTEM PRESSURE Arplicability Applies to the limit on reactor coolant systeet pressure.
Objective To ciaintaf n the integrity of the reactor coolant system and to prevent the relaase of significant amounts of fission product activity.
.}pecification 2.2.1 The reactor coolant system pressure shall not exceed 2750 psig when there are fuel assemblies in the reactor vessel.
2.2.2 The nominal setpoint of the pressurizer code safety valves shall be less than or ec,uel to 2500 psig.
Bases The reactor coolant system (1) serves as a barrier to prevint radionuclides 'in the reactor coolant from reaching the acciosphere.
In the event of a fuel cloaling failure, the reactor coolant system is a barrier against the release of fission products.
Estab1'.shing a system pressure limit helps to assure the int egrity of the react.or coolant systect. -The maximum transient pressure allewchle in the reactor coolant system pressure vessel
-w & the ASME code,Section III, is 110 percent of design pressure.(2) The' >
.um transient pressuer allowable in the reactor coolant system piping, valves, and fittings under ASSI Section E31.7 is 110 percent of design pressure.
Thus, the safety limit of 2750 nsis, (110 percent of the 2500 psig design pressure) has been establichM.(2) The settings for the reactor high pressure trip (2355 psig)
(3) have been established and' t.he p e ssurizer code safety valves (2500 psig) to assure that the reactor coelant system pressure safety limit is not exceeded.
The initial hydrost4 tic test was conducted at 3125 psig (125 per-of det.ign pressure) to verify the integrity of the reactor coolant system.
cent Addition 21 assurance that the reactor coolant system pressure does not exceed the safety limit is govided 'ey setting the pressurizer electromatic relief w.1ve at 2255 psig.C>'
F r r' ESSES (1)
F C U., section 4 e
on 0
DJ (2)
FSid., paragraph 4.3.8.1 (3) FSAR, paragraph 4.2.4 (4)
I'S.C, t a b i t. 4.2-8 2-4
x RANCHO SECO UNIT 1 llE TECHNICAL SPECIFICATIONS
'l
. Safety Limits' and Limiting i
+
Safety System Settings 1
5.
Pump monitors
~
1he pump 'r-6r i cors prevent the min!:au;a core DNBR' from decreasing below I.3 by. tripping the reactor due te (a) the loss of two. reactor coolant pumps.
In one reactor coolant loop, and (b) loss of one or two reactor coolant pumps due.ng two purrp operation.
The pump monitors also' restrict the power level to 55 percent for one reactor coolant pump operation In
+-
each loop.
i j'
C.
11cactor coolant system pressure During a startup accident from low power or a slow rod withdrawal.from F.
high power, the system high pressure trip set point is rdached before the nuclear overpower trip set point.
The trip setting limit shown in j
figure.2.3-1 for high reactor coolant system pressure (2355 psig) has been esteblished to maintain the system pressure below the safety limit i
(2750 psig) for any de:Ign transient. (le i
i The low pressure (1900 psig) and variable low pressure (12.96 T,,g - 5834).
l trip set point shown in figure 2.3-1 have been established to 1
I maintain the DNB ratio greater than or equal to 1 3 for those design accidents tl.at result in a pressure reduction. (2,3)
Due to the calibration and' instrumentation errors the safety analy-sis used a variabic low reactor coolant system pressure trip value
]
of (12 96 T
- 5684).
]
out 1
D.
Ccolant outlet temperature The high reactor coolant outlet temperature trip setting limit (619 F) shown in figure 2.3-1 has been established to prevent f
excessive core coolant tcmperatures in the operating range.
Due to calibration cd instrc~.entation errors, the safety analysis used a tr!p set point of 620 F.
~
E.
Reactor Building pressurc The high Reactor Building pressure trip setting limit (4 psig) providec positive assurance that a reactor trip will occur in the unlikely event of a'stee;m line failure in the Reactor Building or a i
los s-o f-cco l arit accident, even-in the absence of a low reactor coolant systeu pressure trip.
F.
Shutdown bypass i
.in f.,rder to provide for control rod drive tests, zero power physics testing, and startup procedures, there is provision for bypassing cer:ain sec;mnts -of the reactor protection system.
The reactor pmtcetior, system ngmcats which can be byp*assed are shown in m)
D D
2-7 1
9 ih
.k Ub
. ll
. h.
AncnUnt flo.
14.
=_
y.
RANCHO SECO UNIT I' TECHNICAL SPECIFICATIONS f
S?fety 1.imits'and Limiting-Safety 5ystem Settings I
f table 2.3.1.
Two conditions are imposed when the bypass is used:
1.
By administrative control the nuclear overpower trip set point must be reduced to a value <5.0 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 nonnial 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 befor,e the bypass is initiated.
The overpower trip set point of <5.0 percent prevents any significant s cactor pov.er 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.
P.EFERENCES (1)
F5AR, paragraph 14.1.2.2.
(2)
FSAP., paragraph 14.1.2 7 j
(3)
FSAR, paragraph 14.1.2.8 (4)
FSAR, parograph 14.1.2.3 (5)
FSAR, paragraph 14.1.2.6 0
{p D
2-8 j g {1 U
P
'g"ii n
S AL JJ
. JL A
1 3
Ci TABI.E 2.3-1 l
T 5
.4EACT0ft TR0TECTl0N SYSTEM TRIP SETTING t.13fTS S
l
_s One Reactor Coolant Pump rour Reactor Coolant Pumps Three Reactor Coolant Purps
- I I"
OP h'
Operating (Nominal Operating (Nominal (Nomina peratIng Bypass Operating Power - 100% )
Operating Power - 75%)
P wer - 43%)
I.
Nuclear pewer, 2 cf rated, nax.
105.9 1C5.5 105.5 5.0 I3I 2.
Nuclear power based on flcw 1.05 times flow minus 1.05 times flow minus 1.05 tires flow minus Bypassed and larbilance, t of rated, max.
reduction due to reduction due to reduction due to isnbalance(s)
Imbalance (s)
Imbalance (s) 3.
!uclear power based en pump 24A NA 55 Bypassed l
r.ani tur s, 2 of ra ted
.nax.
f g
, i. Hiv. reactor coolant m
,2355 2355 2355 18'20II system pressure, psig, max.
5.
I.ew r.uctor coolant system 1900 1900 1900 Bypassed pressure, psig, min.
6.
Variable low reacto. coolant 12.96 T 5834 Bypassed l
~
system pressure, psig, aln.
out out out 7.
Reactor coolant temp.
F., max.
619 619
. 619 619 8.
High Reactor Building 4
4 4
pressure, psig, max.
(I)
T is in degrees Fahrenhele; (F).
ut (2)
F.cactor coolant system flow, %.
t (3)
Administratively controlled reduction set only during reactor shutdown.
(4)
Automatically set when other segments of the RPS ( s specified) are bypassed.
(5)
The purrp monitors also produce a trip on: (a) loss of two reactor coolant pumps in one reactor coolant loop, s
and (b) loss of one or two reactor coolant pumps during two-pump oper'ation.
O m
0 L) 0 v
n
,D E.
k
.,, 3...
't.
Figure 2.3-1.
Protectiva Systec lbximum Allowsble Setpoints, Pressurc Vs Tenperature 2600 N
E.
2400
___ P = 235 M i,g E
y T = 619 F
~
Eg Acceptable Operation 2200
.I Y
e
.5 4
?
Unacceptable g
i O
Operation i
w oY W
2000 p*
E 4
P = 1900 psig 1800.-- -
J.
__1 I
I 540 560 580 600 620 640 Reactor Outlet Temperature, F r
E
.=
0 1 D gd l
n Q
b gF D Id u k Ll h.;endwant No.
14 1
m i
F Figurc 2.3-2.
Protective System liaximum Allowable Setpoints, Reactor Power Imbalance Tierm1 Pover Levci, %
120
(-19,105)
- (+9, 105)
(105) 4
-100
'O. 2 Acceptable
(+30, 94)
[
Tour Pump s'
Operation - -90 o
e p
-80 (7q)
(-40, 76)
Acceptable -70 Three-and-(+30, 67.1)
Four-Pump Operation
-60
(-40, 49.1)
' ~ (51) 2 - 50 Acceptable 1.
Tuo,Three-
-40
(+30, 39.9) and Four -
Pump Opera-tion
-30
(-40, 21. 9)
-20 9
9 o
3 E
9 d
-10 m
l 8
n a
a e
C O _ _ l_
1
-60
-40
-20 0
20 40 Power Imbalance, %
9 7
0 D
Ld r
nyy T' D uuSj u la
_a L3 Ie.
g
RANCHO SECO 8tNIT 1 I
TECHNICAL SPECI. TIONS I
~
Limiting Conditions for Operation F.
If a control rod in the regulating or axial power shaping groups is' declared inoperable por Specification' 4.7 1.2, operation above 60% of rated power may continue provided the 4
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.
3523 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 physics tests, the quadrant t!!t shall be reduced to less than 3 percerc within two hours or the following actions shall be taken:
(1)
If four reactor coolant pumps are in operatlon, the allowable thermal power shall be reduced by 2 percent of maximum' allowable power for cach 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 level cut off mar apply during transient xenon operation.
i (2)
If less than four reactor coolant pumps are in operation, the allowable thermal power shall be reduced by 2 percent e
of maximum allowable power for each I 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 chall be brought to the hot shutdown condition within four hours
'i if the quadrant tilt is not reduced to less than 3 l
percent af t'cr 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 reduced 2 percent of the maximum allowable flux for the RC pump coSination for cach I percent tilt in excess of 3 percent.
II.
If the quadrant tilt exceeds 3 urcent and there is simulta-~
l neces indication of'a mitaligned 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 reacter coolant pump conbination.
C.
Except for phyr.ics tm tt, if quadrant tilt exceeds 8 percent, l
the re actor shall be brought to the hot shutdown conc ip y q' m h
within four hours.
DGI 3-32 p
g-t
/,r.nndment f10.[.[
14 g [
]_ {
U 3
RANCHO SECO tmlT 1 l-TECHNICAL SPECIFICATIONS l
Limiting Conditfons for Operation D.
Whenever the ' reactor is brought to hot shutdown pursuant to 3.5.2.4a(3) or 3.5.2.4c above, subsequent reactor e;;eration is permitted for the purpose of measurement, testing and correc-Live 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 allowabic power for each I percent tilt.
E.
Quadrant power tilt shall be monitored on a minimum frequency of once every 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.an,d axial.powe,r shaping control, rods.
Except for physius tests or exercising cont rol rods, the regulating control rod Insertion / withdrawal Ilmits are specified on Figures 3.5.2-1 and 3 5.2-2.
Also excepting physics tests or exercising control rods, the axial power shaping control rod insertion / withdrawal limits are speci fied on Fic,ures 3.5.2-3 and 3 5 2-4.
If any of these i
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 phys ics test, power shall not be increased above thc power level cut-off of 92 percent of the maximum allowable power level unleis one of the following conditions is satis-fled:
(1)
Xenon reactivity is within 10 percent of the equilibriu.
val'oe for operation at the maximum allowable power level and asymptotically approaching stability.
(2)
Except for xenon f ree startup, when 3.5 2.5D(1) applies,
the reactor had operated within a range of 87 to 92 percent of the maximum allowable power for a perlod 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.
O M
'OSs@f upAJJai 3-u yy.
A J
,.u.neent no./,j./
1, a
' RANCHO SECO UlilT 1 Tt.CHNICAL SPECIFICATIONS-l Limiting Conditions for Operation l
l 3 5 2.6 Reactor power-imbalance shall be monitored on a frequency not to exceed two hoors during power operation above 40 percent rated power.
Except for physics test, iribalance shall be maintained within the envelope defined by Figures 3.5.2-5 and 3 5.2-6.
If the imbalance is not within the envelope defined by Figures 3 5.2-5 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 cont rol 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 cnalyses which have defined the maximum linear heat rate (see Figure 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 Cr.i.teria only if.the control rods are at the position limits as defined by Figures 3.5.2-1,'3.5;2-2,
~
3.5.2-3, and 3 5.2-4 and if a 3 percent quadrant power tilt exists.
Additional conservatism is int roduced by applicat ion of:
A.
Nuclear uncertainty factors.
i B.
Thermal calibration uncertainty.
C.
Fuel densification effects.
D.
Hot rod manufacturing tolerance factors.
The conservative applicat ;nr of the above peaking augmentetion factors compen-s a t et, fer the potential peaking penalty due to Fuel rod bo.v.
The 25 percent overina between successive control rod groups is allowed since the worth of a rod i<, lower at the upper and lower part of the stroke.
Control roos are arranged in groups defined as follows:
Grpup Function p
i Safety D
Q Q
DJ 2
Safety U
U 1 3
Safety D
T 4
Safety O f 5
Regulating
'] [
a 6
Regulating v
7 Regblating 8
APSR (axial power shaping bank)
Co. trol red r,;roups are withdrawn in sequcnce beginning with Group 1.
Group 5 is overlapped 25 percent with Group 6, and Group 6 is overlapped 25 porcent with Group 7 The normal pos' tion at pouc-is for Group 7 to be partially inserted.
AmendentNo./,f4', 14
_333 3
r--<
DANCHOSECOUNIT1
~
TECHNICAL SPECIFICATIONS t
Limiting Conditions for Operation
{
The minimum available rod worth provides for achieving hot shutdown by reactor I',
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 ek/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 af 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 7 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 1imits in coaj unction with the cont rol rod position 1imits in Specification 3 5.2.5c ensure that design peak heat rate criteria are not exceeded during m.,rmal operation when including the ef fects of potential fuel densification.
The quadrant tilt and axial imbalance monitoring in Speci fications 3.5.2.4 and 3 5 7.6, respectively, norrr. ally will be performed in the process computer.
The two-hour f requency for monitoring these quantities will provide adequate surveillance when the computer is out of service.
Alicuance 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 reduction of power taken.
Operating restrictions are included in Technical Specifications 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 opercted in ti.e range of 87? to 9^t c the maximum allowable power for a period exceeding two hours in the soluble poison control rrode so that the transient peak is burned out at,a lower powe r 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:
4 Test plateau Level %
Overpower Tript
<5 0
e 15 50 Lo 50 D"
75 95 V
90 100
[Q
~ Q
~
"Y
~
105.5 100 RtFfREr4CES
. s 11 k
2 (1)
F MP., Paragr aph 3. 2.2.1.:
(2)
FSAP., Paragraph 14.2.2.4 kondmtmt _ No.((
14
Figure 3.5.2-1 Rod Index'Vs Power Level (0 to 160 U PD)
.I (206.2, 102).
(283.8. 102) g(300,102) gg_
l Operation Ibt Allowed In Thic Regica (283.8, 92) 90f
/
c 80
[
(270.1, 80)
Shutdown
/
Lint j
u 7C E
I M
Restricted 60 Region n
4 b
5e (149.7. 50)
(248.2, 50) 5 y,
4C 30 Permissible Operating 20 Region (90, 15)
(0, 8.6) 10
--p 0
I 8
8 I
I f
I f
I I
I I
l 0
20 40 60 80 100 120 140 160 ISO 200 220 240 250 260 300 i
Rod Index, I Withdrawn 0
25 50 7,5 87.5 10p.
Bank 5 0
25 50 75 37.5 100 I
I I
I I
, u,,, 6 o
2s 50 75 100 y
Bank 7 EED
~
u:
n
o Figure 3.5.2-2.
Rod Index Vs Power Level (150 to 300 EFFD)
_______.__., ~.._.__...-
1 c.
g 199 L (249.1, 102)7 (280.6, 102)[
j(300,102)
I
==
P 1
(200.6, 92) g 901 -
g
,f' l
{
Shutdcwn l
Limit 80 Operation Not Allowed (267.8, 80)
In This Region 7U
~
Restricted p"
Region mE 60 i
N 50 (186.0, 50)
(248.2, 50) i Permissible
$ 40 A
Operating Region 30 20 (131.5, 15) 10. (0, 6'. 7)
(a-1 I
I I
I 1
I I
I I
I I
I J
d 20 40 60 8G 100 120 140 160 180 200 220 240 26te 28'O 3d0 cp Rod Index, % Withdrawn E
a a
a a
0 25 50 75 87.5 100 i
i Bank 5 $
$5 50
[5 bl.5 100
. Bank 6 0
25
$0
[5 100 i
M Bank 7 EED g
f
-s Figure 3.5.2-3 APSR Withdrawal Vs Power Level (0 to 160 EPPD) 110 100'-
(22.3, 102)
(22.3, 92) 90 ~
T 80 (35.2, 80)
Permissible 70 Operating Restricted Region Region y
'A 60 -
w 50 (60, 50) h
- n 2
40 -
~
30 t
20 t
10 0._
i t
i I
I I
~
O 10 20 30.
40 50 60 70 S0 APSR Withdrawal, %
D Pn A B
. U g
PS'g 7
i all_f.J_1 e
f A,.cnd:r.ent tio. 14
m.
Figure 3.5.2-4 APSR Withdrawal Vs Power Level (150 to 300 EFPD) f 110 &
- - ~ - - - - - - - - - -
r
\\
e 100 C (25.5, 102)
(25.5, 92) 90 80 (35.2, 80) g 7p_
Restricted Region n
k 60 Permissible Operating o
Region 50 -
(60, 50)
J 2
40 -
's.
e e
~
30
~
20,-
10 0..
L-J -__L t
3 0
10 20 30 40 50 60 70 80 APSR Withdrawal, %
py
. S.1_ Ild _a v
f
^
Core 1mbalance Vs Power Level (0 to 160 EFFD)
Figure 3.5.2-5
_ _z _
110-( ~ -
I j
,(12.2, 102)
(-12.2, 102'/
--=
l t
r.
100 i
l
(-12.2, 92)
/11.9, 92) 90 (24. 7, 80-)
j 80j- -
(.21.5, 80) l 70 Ferr.issible
- j Operating
~
60 I
-c s
,(40.0, 50) 30 --(-30, 50)g r
l c[
40' j
30 2C 10 t
O'...._..). m. b..... L
.l_
I-
--. -_I
-I
-- - I-n. = *
-50
-40
-30
-20
-10 0
10 20 30 40 50 Core Imbalance, %
0 D
D i
E' a
g g7 uJU.S.I.lfd_a
m.
j Figure 3.5. 2 6.
Core Itabalance Vs Power Level l
(150 to 300 EFPD) t i
110 m-~ ~ -----
100 -
(-12.1, 102)
(14.6, 102)
(19.2, 92).
(-12.1, 92) 90 -
24.7, 80)
(-25, 80) 80 Perdssible Restricted 70 Operating Region 3
Region m
b 60 mo (40,0, 50) 50
(-30, 50)-
- u S
40 e
30 9
20 o
10 OL _.) _.,L. _ 1
. J.__l, J.
I t
-50
-40
-30
-20
-10 0
10 20 30 40 50 Core Imbalance, %
e B
7 0
1 D
L6 I3 3
g' D g'
g[g[
g
/sendt.:en t. f,'o.
14
?
f 4
21 20 t
I9 R
x Iw
- i N
e
' /y.
-\\.
I$
i7 N
16 5
[
m l
g 15
)
a 14 g
l 13 e
mw i
12 0
2 4
6 8
10 12 Axial location of Peak Power From Bottom of Core. ft LOCA LIMITED MAXIMlhi ALLOWAELE.'
1 LINEAR HEAT RATE Figt d
0 II OOE o'9~T j
v 1. S.1.
_a
".e..
e 0
I
/.DMINISTRAT10E. CONTROLS
=
6.3 FACILITY ST AFF QUALIFICATIONS
- 6. 3.1.
Each member of the operating staff shall meet or exceed the minimum" qualificctions of ANSI N16.1-1971 for comparable positions.
6.3.2 The Chemical-Radiation Supervisor shall-meet or exceed the minimum qualifications of NRC Guide 1.3 September 1975.
6.4 T[flNING 6.4.1 A retraining and renlacenent training program for the operating 5
staf f shall be maintained under_ the direction of the Training Supervisor and shal.1 meet or exceed the requirements an; recom-mendations of Section 5.5 of ANSI N18.i-1971 and Appendix "A" of 10 CFR Part 55 h5 REVIri AND AUDIT 6.5.1 PI. ANT REVIEW COMMITTEE (PRC)
FUt!C T I ON.
6.5.1.1 The Plant Review Committee shall function to advise the Plant Seperintendent on all matters related to nuclear safety.
COMPorlTION 6.5 1.2 The Plant Review Ccemittee shall be composed of the:
Chairnan:
Technical Assistant Ne.be r:
Assistant Superintendent, Nuclear Plant Operations Member:
Assistent Superintendent, Technical Support Member:
Maintenance Supervisor Member:
Chemical and Rqdiation Supervisor.
Other members as the Plant Superintendent may appoint from time to time.
D@@M
@@k i
u G-3 Aw.adment flo.
E.