ML20127B171
| ML20127B171 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 06/04/1985 |
| From: | Stolz J Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20127B151 | List: |
| References | |
| NUDOCS 8506210371 | |
| Download: ML20127B171 (25) | |
Text
[ ,'g UNITED STATES y ' g NUCLEAR REGULATORY COMMISSION
- . j wassincron. o. c. 20sss
\....+/ SACRAMENTO MUNICIPAL UTILITY DISTRICT DOCKET NO. 50-312 RANCHO SEC0 NUCLEAR GENERATING STATION AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 69 License No. DPR-54
- 1. The Nuclear Regulatory Comission (the Commission) has found that:
A. The application for amendment by Sacramento Municipal Utility District (the licensee) dated December 17, 1984, as supplemented by letters dated March 14, 1985, and April 9, 1985, complies with the-standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I; B. The fadility will operate in confonnity with the application, the
_the provisions of the Act, and the rules and regulations of the Comission; C. There is reasonable assurance (1) that the activities 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 compliance with the Comission's regulations; D. The issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public; and r
l E. The issuance of this amendment is in accordance with 10 CFR Part 51 l of the Comission's regulations and all applicable requirements have been satisfied.
- 2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-54 is hereby amended to read as follows:
t j B506210371 DR 850604 ADOCK 05000312 p
O 2
Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 69 , are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
- 3. This license amendment is effective as of its date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION oh . Stolz, Chief Op ating Reactors Branch #4 ision of Licensing
Attachment:
Changes to the Technical Specifications Date of Issuancc: June 4,1985 l
e e
l
ATTACHMENT TO LICENSE AMENDMENT NO. 69
- FACILITY OPERATING LICENSE NO. DPR-54 DOCKET NO. 50-312 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages as indicated. The revised pages are identified by Amendment number and contain vertical lines indicating the area of change.
Remove Insert xi xi xii xii 2.1-1 2-1 2-2 2-2 2-3 2-3 Figure 2.1-2 Figure 2.1-2 Figure 2.1-3 Figure 2.1-3 2-7 2-7 Figure 2.3-2 Figure 2.3-2 3-17 3-17 3-18 3-18 3-33a 3-33a 3-33b 3-33b Figure 3.5.2-1 Figure 3.5.2-1 Figure 3.5.2-2 Figure 3.5.2-2 Figure 3.5.2-3 Figure 3.5.2-3 Figure 3.5.2-4 Figure 3.5.2-4 Figure 3.5.2-5 Figure 3.5.2-5 Figure 3.5.2-6 ~
Figure 3.5.2-6 Figure 3.5.2-7 Figure 3.5.2-8 Figure 3.5.2-9 Figure 3.5.2-9 Sf'-~ Figure 3.5.2-10 --
Figure.3.5.2-11 --
Figure 3.5.2-12 --
~l
.. . . _ _ _ = _.
RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS LIST OF FIGURES Figure 2.1-1 Core Protection Safety Limit, Pressure vs. Temperature i 2.1-2 Core Protection Safety Limits, Reactor Power Imbalance 2.1-3 Core Protective Safety Bases 2.3-1 Protective System Maximum Allowable Setpoints, Pressure vs Temperature 2.3-2 Protective System Maximum Allowable Setpoints, Reactor Power Imbalas;ce
., 3.1.2-1 Reactor Coolant System Pressure-Temperature Limits for Heatup for the First 5 EFPY 3.1.2-2 Reactor Coolant System Pressure-Temperature Limits for Cooldown for the First 5 EFPY 3.1.2-3 Inservice Leak and Kydrostatic Test (5 EFPY) Heatup and Cooldown 3.1.2-4 Reactor Coolant System, Emergency / Faulted Condition-Cooldown Limitations, Applicable for 5 EFPY I
3.1.9-1 . Limiting Pressure vs. Temperature for Control Rod Drive Operation 3.5.2-1 Rod Index vs. Power Level for Four-Pump Operation, O to 40 EFPD qp,__
3.5.2-2 Rod Index vs. Power Level for Four-Pump Operation, after 30 EFPD 3.5.2-3 Rod Index vs. Power Level for Feur-Pump Operation, after 300 EFPD '
with APSRs Withdrawn l 3.5.2-4 Rod Index vs. Power Level for Three-Pump Operation, O to 40 EFPD 3.5.2-5 Rod Index vs. Power Level for Three-Pump Operation, after 30 EFPD 3.5.2-6 Rud Index vs. Power Level for Three-Pomp Operation, after 300 I EFPD with APSRs Withdrawn l .,
=I.
Amendment flo. 21., gg, 2), pg, );. 69 XI
i
- i RAldCHO SECO UNIT 1 TECHN1 CAL SPECIFICATIONS LIST OF FIGURES (Continued)
Figure 3.5.2-7 Core Imbalance vs. Power Level O to 40 EFPD 3.5.2-8 Core Imbalance vs. Power Level, after 30 EFPD 3.5.2-9 Core Imbalance vs. Power Level, after 300 EFPD with APSRs Withdrawn 3.5.2-10 Deleted 3.5.2-11 Deleted 3.5.2-12 Deleted 3.5.4-1 Incore Instrumentation Specification Axial Imbalance Indication 3.5.4-2 Incore Instrumentation Specification Radial Flux Tilt Indication 3.5.4-3 Incore Instrumentation Specification 3.18-1 General Layout of Site 4.13-1 Main Steam Inservice Inspection 4.13-2 Main Feedwater Inservice Inspection 4.13-3 Main Steam Dump Inservice Inspection 6.2-1 SHUD Organization Chart 6.2-2 Plant Organization Chart i
i
= .
Enend,nnt Fn. 7,6, 72, 73, !?,,st, 69 xit
RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Safety Limits and Limiting Safety System Settings
- 2. SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS, REACTOR CORE i Applicability Applies to reactor thermal power, reactor power imbalance, reactor coolant system pres:ure, coolant temperature, and coolant flow during power operation of the plant. .
Objective To maintain the integrity of the fuel cladding.
- Specification j
2.1.1 Tne 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 themal 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-reacto'r-themal-power / reactor-power-imbalance point is above the line
.. for the specified flow, the safety limit is exceeded.
f Bases .
The safety limits cresented have been generated using g BAW-2 and BWC CHF correlations (I' 41 and the actual measured flow rate . The flow rate utilized i l gpm) based on four-pump operation j2,ogg .. percent of the design flow (369,600 To maintain the integrity of the fuel cladding and to prevent fission product release to the primary coolant system, it is necessary to prevent overheating l of the cladding unoer nomal operating conditions. This is accomplished by operating within the nucleate boiling region of heat transfer, wherein the heat transfer coefficient is large enougn so that the clad surface temperature l . is only slightly greater than the coolant temperature. The upper boundary of the nucleate boiling region is temed " departure from nucleate boiling" (DNB). At this point, there is a sharp reduction of the heat tmnsfer coefficient, which would result in high cladding teuperatures and the i possibility of cladding failure. Although DNB is not an observable parameter l during reactor operation, the observable parameters of neutron power, reactor coolant flow, temperature, and pressure Amendmentflo.//,'69 2-1
RANCHO SECO UN3T 1 TECHN! CAL SPEC 2FICAT3DNS Saf ety Limits anc , Limiting
, Safety System Settings
~ "
can be related to DNS through the use of the CHF correlation Ile 4). The BAW-2 anc SWC correlations have been developed to preciet DNB and the location of DNS for axially uniform anc non-uniform heat flux distributions.- The local
~
DNS ratio (DNBR), defined as tne ratio of the heat flux that would cause DNS at a particular core location to the actual heat flux, is indicative of the margin to DNS. The. minimum value of sne DNBR, during steady-state operation, s nomal operational transients, and anticipated transients is limitec to 1.30 A DNBR of 1.30 (BAW-2) or 1.18 (BWC) corresponds to a (SAW-2) or 1.18 (BWC).
95 percent probability at a 95 percent confidence level that DNS will not occur; this is consicered a conservative margin to DNB for all operating ~
conditions. The difference between the actual core outlet pressure and thi indicated reactor coolant system pressure has been considerec in determining the core protection safety limits. The difference in these two pressures in ncminally 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 measurec. t The curve presented in Fiqure 2.1-1 represents the conditions at which a DNBR.
equal to or greater than the correlation limit is predicted for the maximum i possiele thermal power (112 percent) when four reactor coolant pumps are operating (minimum reactor coolant flow is 104.9 percent of 369,6n0 gpm).
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 tnan any other shape that exists during nomal operation.
The curves of Figure 2.1-2 are based on the more restrictive of two themal limits and incluoe tne effects of potential fuel censification and rod bowing.
- 1. The combinations of the radial peak, axial peak and position of the axial peak that yielos a DNBR no less than the CHF correlation limit.
- 2. The combination of radial and axial peak that causes central fuel melting '
at the hot spot. The limit is 20.4 KW/ft.
l 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 power peaking.
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.
- The curve of Tigure 2.1-1 is the most restrictive of all possible reactor l coolant pump-maximum themal power combinations shown in Figure,2.1-3.-t.
For eacn curve of Figure 2.1-3, a pressure-temperature point above and to the left of the curve would result in a DNSR greater than the CHF correlation limit.or a local quality at the point of minimum DNBR 1ess than 22 percent for that particular reactor coolant pump situation.
Amendment No. 25, )$, 63 22 l
l
RANCHO SECO Ud2T 1
- TECHN3 CAL SPEC 2F3 CATIONS Safety Limits and Limitine Safety System Settings The maximum permitted thermal power for three-pump operation depicted in Figure 2.1-2 is 87.8 percent due to a power level trip produced hy the flux-flow ratio 1.06 times 74.4 pertent design flow . 78.86 percent power plus
-he absolute value of the maximum calibration and instrumentation error. The l maximum thermal power for other coolant pump conditions is produced in a similar manner. The actual maximum power levels are calculated by the RFS anc will be directly proportional to the actual flow during partial pump operation.
l REFERENCES (1) Cerrelation of Critical Heat Flux in a Suncle Cooled 4y Pressurizec Wa er, BAW-10000A, May 1976.
(2) Rancho Seco Unit 1, Cycle 2 Reloac Report, BAW-1460, June 1977.
(S) Rancno Seco Unit 1 Cycle 3 Reload Report, EAW-3499, September 1978.
(4) Correlation of 15x15 Geome ry 2ircaloy Grid Roc Buncle CHR Data With the SWC Correlation, BAV-10143P, Part 2, Babcock and Wilcox, Lynchburg, Virginia, August 13cl.
l-i
=#,
=t.
Amendment No. RJ, pp, );. 89 2-3
RANCHO SECO UN1T 1 TECHNICAL SPECIFICATIONS Safety Limits and Limiting Safety System Settings Figure 2.1-2 Core Protection Safety Limits, Reactor Power Imbalance, Rancho Seco 1, Cycie 7 THERMAL POWER LEVEL, %
- 120
(-32.4.112 33.4.112)
C""** I "110 ACCEPTABLE 4 PUMP OPERATION ( 4,101.9) 00
(-48,90.7) (-32.4,87.8) (33.4,87.8)
-90
~
ACCEPTABLE 4 Curve 2
& 3 PUMP -
80 (44,77,7)
OPERATION
(-48,66.5) 70
(-32.4,60.6) (33.4,60.6)
ACCEPTABLE 4
'" Curve 3 3 & 2 PUMP OPERATION -
-50 (44,50.5)
(-4P,,39.3 ) ,. -
40 UNACCEPTABLE UNACCEPTABLE OPERATION
-20 OPERATION
-10 l l l : l l l : l l l 50 30 10 0 10 20 30 40 50 60 Reactor Power Imbalance, %
Curve Reactor Coolant flow, % Design
- =
t 1 104.9 ..
2 -78.0 i 3 50.9 Amendment No.16, 27, pp, 33, pg, 69 l
~
RANCHO SECO UN37 i TECHi?ICAL SPEClFICATIOl45 Safety Limits and Limiting Safety System Settings Figure 2.1-3 Core Protective Safety Bases, Rancho Seco I, Cycle 7 2400 m 2200 - 12 3
~
m c.
5 E
5 2000 -
t 8
2 0
1800 -
y '
1600 560 580 600 620 640 Reactor Outlet Temperature, F Reactor coolant Pumps operating Curve flow, % design Power, % (type of limit) 1 104.9 112 Four (DNBR limit) 2 78.0 87.8 Three (DNBR limit) 3 50.9 60.6 One in each loog I
- l (quality limit)
Amendment flo. gg, pp, 23, 69 l
I RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS f Safety Limits and Limiting
! Safety System Settings B. Pump Honitors
- The pump nonitors prevent the minimum core DNBR from decreasing below l the CHF correlation licit by tripping the reactor due to (a) the loss l of two reactur coolant pumps in one reactor coo! ant loop, and (b) loss i of one or two reactor coolant pumps during two-pump operation. The pump monitors also restrict the power level to 55 percent for one reactor coolant pump operation in each loop.
4 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 (2300 psig) has i been established to maintain the system pressure below the safety limit (2750 psig) for any design transient (1) and minimize the challenges to i the EMOV and code safeties.
The low pressure (1900 psig) and variable low pressure
, (12.96 Tout - 5834) trip set point shown in figure 2.3-1 have been established to maintain the DNB ratio greater than or equal to the CHF correlation limit for those design ' accidents that result in a pressure
- reduction. (2,3)
Due to the calibration and instrumentation errors the safety analysis used a variable low reactor coolant system pressure trip value of (12.96 Tout - 5884). ,
-- D. Coolant outlet temperature f The high reactor coolant outlet temperature trip setting limit (618 F) shown in figure 2.3-1 has been established to prevent excessive core coolant temperatures in the operating range. Due to calibration and 4 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 -
3 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 reactet l
protection system segments ubich can be bypassed are shown in Amendment No. JJ, JJ, ;), 09 2~7
RANCHO SECO UNIT 1 TECHNfCAL SPECIFICATIONS Safety Limits and Limiting Safety System Settings Figure 2.3-2 Protective System Maximum Allowable Setpoints, Reactor Power Imbalance, Rancho Seco 1, Cycle 7 THERMAL POWER LEVEL, %
-- 120 T 110 f19,106)
(-19,106) p
'- 100 p. M2 * 'I*II9 M3 = 1.577 ATION 1 -
- 90 l/ (31,92.56)
(19,7846)
(-34.6,81.39) ( 59,78.86) .. 80 i CCEPTABLE c.
$&3 PUMP- 70 l *,
OPERATION g .
I I -
- 60
(-34.6,54.25) ($9,51.4) (19,51.1) ,
j CCEPTABLf' #
bM - 40 g> (31'37.96) o OPERATION
- 30 l D
(-34.6.26.79) l-l I
- . gg
- I I A " 10 e -
J Jl 7l 7 4 i #l E j: #
l l : ! : . ! :
50 30 10 0 10 20 30 40 50 60 Reactor Power Imbalance, %
Curve Reactor Coolant Flow,1 Desien ;g 1 104.9 --
2 78.0 3 50.9 Amendment flo. 73, 27, Ap, );, pp,49 i
l
. RANCHO SECO UNIT 1 TECHN8 CAL SPECIFICAT!ONS Limiting Conditions for Operation l
3.2 HIGH PRESSURE INJECTION AND CHEMICAL ADDITION SYSTEMS Applicability Applies to the operational status of high pressure injection and chemical l aedition systems.
Objective
, To provide for adequate boration under all operating conditions to assure ,
ability to bring the reactor to a cold shutdown conoition.
Specification The reactor shall not remain critical unless the following conditions are met:
3.2.1 Two pumps capable of supplying high pressure injection are operabic
. . (also see Specification 3.3.2).
3.2.2 The borated water storage tank and its flow path to the reactor for high pressure injection are operable. ,
l -
3.2.3 A source of concentrated boric acid solution in addition to the l ,
borated water storage tank is available and operable. This requirement is fulfilled by the concentrated boric acid storage tank.
Tnis tank shall contain at least the equivalent of 10,000 gallons of l
7 100 ppm boron. System piping and valves necessary to establish a flow path for high pressure injection shall also be operable and shall have at least the same temperature as the boric acid storage tank.
, One associated boric acid pump,is operable. The concentrated boric acid storage tank water shall not be less than 70F, and at least pne channel of heat tracing shall be operable for this tank's associated f p g. The concentrated boric acid storage tank boron concentration s not exceed 8,500 ppm baron.
Ba'ses The makeup and purification system and chemical addition systems provide control of the reactor coolant system boron concentration.' This is normally accomplishco by using aither the makeup pump or one of the two high pressure injection pumps in series with a boric acid pump associated with the concentrated boric acid storage tank. The alternate method of boration will
- be the use of the makeup or high pressure injtetion pumps taking suction directly from the borated water storage tank.
The quantity of boric acid in storage from either of the two aboy-mentioned sources is sufficient to borate the reactor coolant system to a 1 percent subcritical margin in the cold condition (70F) at the worst time ~in core life with a stuck control rod assembly. The maximum required is the equivalent of 9586 gallons of 7100, rpm boron. This requirement is satisfied by requiring a l minimum volume of 10,000 gallons of 7100 ppo in the concentrated borated acia
(
Anendment No. 27,69 3-1 l
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- IN i ,
RANCHO SECO UNIT 1
' TECHNICAL SPECIFICATIONS
<. ~ s Limiting Conditions for Operation l
storage ta'nk during critical operations. The minimum volume for the borated '
water storage tank (390,000 gallons of 1800 ppm boron), as specified in section 3.3, is based o,i refueling volume requirements and easily satisfies the cold shutdown requirement. The specification assures that the two r supplies are available whenever the reactor is critical so that a single failure will not prevent boration to a cold condition. The minimum volumes of boric acid solution given include the boron necessary to account for xenon dec4y. .
The primary method of adding boron to the primary system is to pump the l o concentrated boric acid solution (7100 ppm boron, minimum) into the makeup y' tank using the 50 gpo boric acid pt. cps. Using only one of the two boric acid pumps, the required volume of boric acid can be injected in less than 3.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. The alternate method of addition is to inject boric acid from the bofated Unter storage tank using the high pressure injection pumps.
'?Cencentration of boron in the concentrated boric acid storage tank may be
~ ~ higher 'than the concentration which would crystallize at ambient conditions.
For this reason and to ensure that a flow of boric acid is available when needed, this tank and its associated piping will be kept above 70F (30F above the crystallization temperature for the concentration present). Once in the high pressure injection system, the concentrate is sufficiently well mixed and diluted so that normal system temperatures ensure boric acid solubility. The value of-70F is significantly above the crystallization temperature for a solution containing 12,700 pp,a boron. ,
REFERENCES y '
~~
l lFSAR sdbsections 9.2 and 9.3. ' ~ .
2- ' NSARFigure6.2-1.
3 Technical Specification 3.3.
i s
..\
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J G 'r A g 3-18 O
RANCHO SECO UN8T 1 TECHN8 CAL SPECIF! CATIONS 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 tests, imbalance shall be naintained within the envelope oefined by Figures 3.5.2-7 through 3.5.2-9. If the imbalance is not within the envelope defined by Figures 3.5.2-7 through 3.5.2-9, corrective measures shall be taken to achieve an acceptable imbal ance. If an acceptable imbalance is not achieved within two nours, 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 designated representative. .
Bases ,
The power-imbalance envelope defined in Figures 3.5.2-7 through 3.5.2-9 are l based on LOCA analyses which have defined the maximum ifnear heat rate such that the pximum clad temperature will not exceed the Final Acceptance Criteria. Corrective measures will be taken should the indicated quadrant tilt, rod position, or imbalance be outside their specified boundry.
. 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 oower distribution parameters (quadrant tilt, rod position, and imbalance) must be
~
at their limits while simultaneously all other engineering and uncertainty factors are also at their limits.**
- a. Nuclear uncertainty factors
- b. Thermal calibration uncertainty
- c. Hot rod manufacturing tolerance factors
- d. Fuel densification effects The conservative application of the above peaking augmentation factors compensates for the potential peaking penalty due to Fuel rod bow. .
The 25%
- overlap between successive control rod groups is allowed since the worth of a rod is lower at the upper and lower part of the stroke. Control rods are arranged in groups or banks defined as follows:
Group Function I sareTy .
2 Safety 3 Safety 4 Safety 5 Regulating 6 Regulating 7 Regulating 8 APSR (axial power silaping group)
- Actual operating limits depenu on whether or not incore or excore detectors are used and their respective instrument Calibration errors. The metnod used to cefine the operating limits is defined in piant operating procedures.
Amendment No. U , 69 3-33a
RAliCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS l Limiting Conditions for Operation 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. Therefore, compliance with the ECCS power peaking criterion is ensured by the rod position limits. The minimum available rod worth, consistent with the rod position limits, provides for achieving hot shutdown by reactor trip at any time, assuming the ghest worth control rod that is withdrawn remains in the full out position. The rod position limits also ensure that inserted rod groups will not contain single rod worths greater than 0.65% ak/k l at rated power. These values have been shown t l analysis of hypothetical rod ejection accident.p2ge A maximum safesingleby the safety inserted control roo worth of 1.0 %Ak/k is allowed by the rod position limits at hot zero power. A single inserted control rod worth of 1.0% ak/k at beginning-of-life, hot zero power would result in a lower transient peak thennal power and, therefore, less severe environmental consequences than an 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 percent. The nonnal position at power is for.
Group 7 to be partially inserted. ,
The Quadrant Power Tilt limits set forth in Specification 3.5.2.4 have been
. established to prevent the linear heat rate peaking increase associated with a positive quadrant power tilt during normal power operation from exceeding 7.36%. The limits in Specification 3.5.2.4 are measurement system
~
independent. The actual operating limits, with the appropriate allowance for
- observability and instrumentation errors, for each measurement system are defined in the station operating procedures. ,,
\ .
The Quadrant Tilt and axial imbalance monitoring in Specifications 3.5.2.4F and 3.5.2.6 respectively, nonaally will be performed in the process computer. The two-hour frequency for monitoring these quantities will provide y 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 reduction of power taken.
Operating restrictions are included in Technical Specifications 3.5.2.5.D.(1) and 3.5.2.5.D.(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 % of the maximum allowable power for a period exceeding two hours in the soluble poison control mode so that the transient peak is burned out at a lower power level. , g REFERENCES -
(1) F5AR, Section 3.2.2.1.2 (2) FSAR, Section 14.2.2.4 (3) BAW-1850, October 1984, page 7-5 l l
Amendment *!o. 7$, 69 3-33b
RANCHO SECO UNIT 1 TECHN1 CAL SPEClFICAT20NS Limiting Conditions for Operation Figure 3.5.2-1 Rod Index Vs Power Level 'for Four-Pump Operation, O to 40 EFPD
-- Rancho Seco 1, Cycle 7 110 (229,102) ~
3co_ ,
(285,102) (300,102) 90- OPERATION NOT (275,92)
ALLOWED 80- '
SHUT 00WN MARGIN
, 70- LIMIT g RESTRICTED
~ 60-E (140,50) t 50-225,50) w .
. .t 0-I OPERATION g, . ALLOWED e
(70,15) 10- -
(0,7,8) .
O s s , .
s > , , , , e n ,
0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 s
. Rod Index , , , , ,
0 25 50 75 200 0 25 50 75 100 g BANK 5 BANK 7 6 b 2S 50 /5 lbo BANK 6 e
. . . ! i t
- = I
! )
i Amendment No. 2E. ?.9,.2.3, Ag, 69
RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Ooeration Figure 3.5.2-2 Rod Index Vs Power Level for Four-Pump Operation After 30 EFPD
-- Rancho Seco 1, Cycle 7 110 (229,102) (275,102) (300,102)
(260,92) 90-OPERATION NOT 80- ALLOWED (225,80)
- SHUTDOWN E 70- MARGIN R LIMIT Z 60-50- (140,50) RESTRICTED (200,50) s.
g 40-c.
30- OPERATION ALLOWE0 20-p (70,15) 10-(0.7.8) 0 2b 4'O 60 8'O Id0 Ik0 l401501$0 2d0 2k0 240 2$0 2$0 300 ,,
Rod index , , , ,
0 25 50 75 100 0 25 50 75 100 BANK 5 BANK 7
' f f i f a
0 25 50 75 100 BANK 6 t
- = .
Amendment no. R$, 19, pp, pp, 59
- ~
RANCHO SECO UN3T 1 TECHN! CAL SPECIF.ICATIONS Limiting Conditions for Ooeration Figure 3.5.2-3 Rod Index Ys Power Level for Four-Pump Operation Af ter 300 EFPD With APSRs Withdrawn -- Rancho Seco 1, Cycle 7.
110 (229,102) (280,102)
(300,10e,)
200-OPERATION NOT (270,92 90- ALLOWED (250,80) o SHUTDOWN E 70 MARGIN u LIMIT R
N 60-RESTRICTED
- w 50- (225,50)
, C (140,50)
[
40-m.
30- OPERATION ALLOWED
. 20-D 10-(70,15)
(0.7.8) 0 2O A0 6'O 8'O 16012'O 1d0 ISO 1$0 200 250 2do 250 2$0 300 Rod Index , , , , ,
0 25 50 75 300 0 25 50 75 100 BANK'S BANK 7 2 I t f f I O 25 50 75 100 BANK 6
- =
Amendment No. R$, 27, 3), Jg,.69 l
l
RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 1 Figure 3.5.2-4 Rod Index Ys Power Levei for Three-Pump Operation, .0 to 40 EFPD -- Rancho Seco 1, Cycie 7 110 100-OPERATION NOT 90- ALLOWE0 (229,77)
(247.5,77) 80- (300,77)
- a E 70-E SHUT 00WN E 60- MARGIN
- - LIMIT 50- RESTRICTE(225,50)
C l
c-40-(140,38) 30-OPERATION
~-
LOWED 20- .
10- -(70,11.75) 0 (0.6.4) , , . . . . . .
0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300
, , , , , Rod Index , , , , ,
0 25 50 75 100 0 25 50 75 100 t
BANK 5 BANK 7 e i I t e 0 25 50 75 100 BANK 6 l . . .I l
- - n ,. n . n , n . n . m 6
l l - , , - . . _ .
RANCHO SECO UNIT 1 TECHN! CAL SPECIFICATIONS Limiting Conditions for Operation Figure 3.5.2-5 Rod Index Vs Power Level 'for Three-Pump Operation Af ter 30 EFPD -- Rancho Seco 1, Cycie 7 110 100-90-80' (229,77)
- (300,77)
E 70-
~
% OPERATION NOT R 60- ALLOWED w SHUTDOWN 50- MARGIN (200,50)
LIMIT L
g 40-
- n. (140,38) 30-OPERATION ALLOWE0 20-10- (70,11.75)
(0.6.4) 0 . . . . , , , , . . . . . .
O 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300
, Rod Index , , t . .
0 25 50 75 100 0 25 50 75 100 BANK 5 BANK 7 e i t t t 0 25 50 75 100 BANK 6
' . lt ,
- vndment No. 4 , y , 33, Jg, 69 l
r--- --#e .
RANCHO SECO UNIT 1 TECHN! CAL SPECIF! CAT 10NS
. ' Limiting Conditions for Operation Figure 3.5.2-6 Rod Index Ys Power Level for Three-Pump Operation After 300 EFPD With APSRs Withdrawn -- Rancho Seco 1, Cycle 7 l
110 l
100-90-80- (229,77) (247.5,77)
( 00,77) k 70-R Z 60-
, OPERATION NOT o ALLOWED 50- SHUT 00WN ESTRICTE
. MARGIN 225,50)
LIMIT f
n.
40-30- (140,38)
OPERATION 20- ALLOWED D 10- (70,11.75)
(0,6.4) 0 , , , , , , , , , , , , , , ,
0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 i i , e . Rod Index , , , , ,
0 25 50 75 100 0 25 50 75 100 BANK 5 BANK 7 1 e f f f 0 25 50 75 loo BANK 6
- lt .
AmendmentNo.R$,2),JJ,A),69 l
I .
RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions'for Operation 4
4 Figure 3.5.2-7 Core Imbalance vs Power Level, O to 40 EFPD -- Rancho Seco 1, Cycle 7 110 -
- (-11,102) (22.5,102) 4 100 _
(-16.5,92) (31,92) 90 -
RESTRICTED
, REGION 30 _
(-23,80) 1 70 _
PERMISSIBLE '
OPERATING 60 -
REGION 50 (- 6, 1
f -
40 _
30 -
20 _
10 _
t
= ,
0 _, , , , , , ,
-50 -40 -30 -20 -10 0* 10 20 30 40 50 4
Amendment No. 69
-. - - . ~ _ _. - ._ _ - . _ _ . - . _ _ _ _ , _ - - , . - ___ __ . _ _ _ _ _ _ _ __ _ _ _ _ _ _ _
RANCHO SECO Ul1IT 1 TECHi/ICAL SPECIFICATIONS Limiting Conditions for Operation Figure 3.5.2-8 Core Imbalance Ys Power Leval After 30 EFPD -- Rancho Seco 1, Cycle 7 RESTRICTED 110 REGION
(-21.5.102) (22.5,202) 3gg_
go_ (-27.8,92) (31,92) 80- (-34.6,80)
- a E 70-
~
R
- 60-o
" 50- PERMISSIBLE
- - OPERATING E REGION E 40-c.
30-
~~
20
' y- .
10- -
0, , , s , , , , , , ,
-50 -40 -30 -20 -10 0 10 20 30 40 50 Core Imbalance, l 't
= .
Amendment Ho 69 l
l
RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS
- Limiting Conditions for Operation Figure 3.5.2-9 Core Imbalance Ys Power Level Af ter 300 EFPD With APSRs Withorawn -- Rancho Seco 1, Cycle 7 RESTRICTE0 110 - REGION ,
(-21.5.102) (22.5,102) 200-(-27.8,92) (23*92) 90-80- (-34.6,80) a
.E 70-R S 60-e
- 50-
' PERMISSIBLE
$ 40- OPERATING
' REGION 30-
. 20-D' 30 l
0 , , , , , , , , , ,
l -50 -40 -30 -20 -10 0 10 20 30 40 50 Core Imbalance, %
l l
l
=\ .
Am:ndment No. !$, 27, 33, $5,69 l
. , . , . , _