ML20081F434
| ML20081F434 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 03/16/1995 |
| From: | Kalman G NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20081F439 | List: |
| References | |
| GL-88-16, NUDOCS 9503220134 | |
| Download: ML20081F434 (15) | |
Text
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t UNITED STATES j
j NUCLEAR REGULATORY COMMISSION t
WASHINGTON, D.C. 2006& 4 001
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ENTERGY OPERATIONS INC.
DOCKET NO. 50-313 ARKANSAS NUCLEAR ONE. UNIT NO. 1 AMEIOMENT TO FACILITY OPERATING LICENSE Amendment No. 178 License No. DPR-51 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Entergy Operations, Inc. (the licensee) dated August 30, 1994, as supplemented by letter dated March 9,1995, complies 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 application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance:
(1) that the activities authorized by this amendment etn be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this license amendment will not be inimical to the common defense and security or to 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.
9503220134 950316 PDR ADOCK 05000313 p
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Accordingly, the licte,se 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-51 is hereby amended to read as follows:
2.
Technical Soecifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 178, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications.
3.
The license amendment is effective as of the date of issuance to be implemented within 30 days.
FOR THE NUCLEAR REGULATORY COMMISSION George Kal enior Project M Project Directorate IV-1 Division of Reactor Projects - III/IV Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: flarch 16. 1995 1
ATTACHMENT TO LICENSE AMENDMENT NO.178 FACILITY OPERATING LICENSE NO. DPR-51 DOCKET NO. 50-313 Replace the following pages c' the Appendix "A" Technical Specifications with the attached pages. The revised pages are identified by Amendment number and contain vertical lines indicating the area of change.
REMOVE PAGES INSERT PAGES 7
7 8
8 9b 9b 11 11 12 12 14b 14b 15 15 16 16 31 31 46 46 47 47 142 142
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2.
SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS, REACTOR CORE
^
Applicability Applies to reactor thermal power, reactor power imbalance, reactor coolant system pressure, coolant temperature, and coolant flow when the reactor is
-critical.
i Object ~ we To maintain the integrity of the fuel cladding.
Speci fication 1
2.1.1 The maximum local fuel pin centerline temperature shall be s 5080 - (6.5 x 10~3 x (Burnup, MWD /MTU)*F) for TACO 2 applications
~,
and 54642 -- (5.8 x 10'3 x (Burnup, mwd /MTU)'r for TACO 3 applications.
operation within this limit is ensured by compliance with the Axial Power Imbalance protective limits preserved by Table 2.3-1 " Reactor Protection System Trip Setting Limits," as specified in the COLR.
2.1.2 The departure from nucleate boiling ratio shall be maintained greater than the lindts of 1.3 for the RAN-2 correlation and 1.18 for the BWC correlat.on. Operation'within this limit is ensured by compliance with Specification 2.1.3 and with the Axial Power Imbalance protective limits preserved by Table 2.3 !
" Reactor Protection System Trip Setting Limit's," as specified in the COLR.
2.1.3 Reactor Coolant System (RCS) core outlet temperature and pressure shall be maintained above and to the left of the Safety Limit -
shown in rigure 2.1-1.
Bases 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 boiling' 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 boiling regime is termed departure from nucleate boiling l
(DNB). At this point there is a sharp reduction of the heat transfer coefficient, which could result in high cladding temperatures and the possibility of cladding failure. Although DNB is not an observable parameter during reactor operation, the observable parameters of neutron power, reactor cociant flow, temperature, and pressure can be related to r
DNB through the use of a critical heat flux (CHT) correlation.
The-RAN-2 (1) and BWC(2) correlations have been developed to predict DNB and the.
location of DNB for axially uniform and non-uniform heat flux distributions. The RAN-2 correlation applies to Mark-B fuel and the BWC correlation applies to Mhrk-BZ fuel. The local DNB ratio (DNBR), defined e
as the ratio cf 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 1
operational transients, and anticipated transients is limited to 1.30 (EAN-2) and 1.18 (BWC).
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Amendment No. M,4H,178 7
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A DN!R cf 1.30 (EAN-2) or 1.18 (BWC) estrorpands to o 95 perecnt probability at a 95 percent confidence level that DNB will not occurs 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 for the allowable RC pump combination has been ' considered in determining the core protection safety limits.
The curve presented in Figure 2.1-1 represents the conditions at which the DNBR is greater than or equal to the minimum allowable DNBR for the limiting combination of therwal power and number of operating reactor coolant pumps. This curve is ba.ed on the following nuclear power peaking factors (3) with potential fuel densification effects:
ry=2.83; r[H=1.71;FN = 1.65.
The Axial Power Inhalance Protective Limits in the COLR are based on the more restrictive of two thermal limits and include the effects of potential fuel densification:
1.
The DNBR limit produced by a nuclear power peaking factor of Fl
= 2.83 or the combination of the radial peak, axial peak and position of the axial peak that yields no less than the DNBR limit.
2.
The combination of radial and axial peak that prevents central fuel melting at the hot spot as given in the COLR.
Power peaking is not a directly observable quantity and therefore lindts have been established on the basis of the reactor power imbalance produced by the power peaking.
The flow rates for curves 1, 2, and 3 of Figure 2.1-3 correspond to the expected minimum flow rates with four pumps, three pumps, and one pump in each loop, respectively.
The curve of rigure 2.1-1 is the most restrictive of all possible reactor coolant pump naximum thermal power cenbinations shown in Figure 2.1-3.
The curves of rigure 2.1-3 represent the conditions at which the DNBR limit is predicted at the maximum possible thermal power for the number of reactor coolant purps in operation.
If the actual pressure / temperature point is below and to the right of the pressure / temperature line the Safety Limit is exceeded. The local quality at the point of ndnimum DNBR is less than 22 percent (BAW-2 ) (1) or 26 percent (BWC) (2 ).
i Amendment No. M, M,M,+H,178 8
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.e, THIS PAGE INTENTIONALLY LEFT BLANK Amenchent No. 6,M, M,4, W,4, 9b 4, M,4M,4U,178
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.2.3LIMITIN3 SAFETY SYSTEM SETTINZS, PROTECTIVE INSTRUMENTATION Applicability 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 pressure.
Objective To provide automatic protection action to prevent any combination of process variables from exceeding a safety limit.
Specification 2.3.1 The reactor protection system trip setting limits and the permissible bypasses for the instrument channels shall be as stated in Table 2.3-1 and the Protection System Maximum Allowable Setpoints for Axial Power Imbalance as given in the COLR.
Bases 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 setpoints plus calibration and instrumentation errors.
l Nuclear Overpower i
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 104.9 percent of rated power. Adding to this the possible variation in trip setpoints due to calibration and instrument errors, the maximum actual power at which a trip would be actuated could be 112%, which is the value used in the safety analysis.
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 established to accommodate the most severe thernal transient considered in the design, the loss-of-coolant-flow accident from
)
high power. Analysis has demonstrated that the specified i
power-to-flow ratio is adequate to prevent a DNBR of less than 1.30 i
(BAN-2) or 1.18 (BWC) should a low flow condition exist due to any
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electrical malfunction.
1 Amendment M,6,&,44,178 11
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Th3 ' power icvel trip cetpoint 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 setpoint 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.
The flux / flow ratios account for the maximum calibration and instrumentation errors and the maximum variation from the average value of the RC flow signal in such a manner that the reactor protective system receives a conservative indication of.the. RC
. flow.
i No penalty in reactor coolant flow through the core was taken for an open core vent valve because of the core vent valve surveillance; program during each refueling outage.
For safety analysis calculations the maximum calibration and instrumentation errors for the power level 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 DNEA limits. The reactor-i power imbalance (power in 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 the Protection system Maximum Allowable Setpoints for Axial Power Imbalance in the COLR are produced. The power-to-flow ratio reduces the power level trip associated reactor power-to-reactor power imbalance boundaries by the value specified in the COLR for a 1 percent flow reduction.
B.
Pump Monitors In conjunction with the power imbalance / flow trip, the pump
-l monitors prevent the minimum core DNBR from decreasing below 1.30 (BAW-2) or 1.18 (BWC) by tripping the reactor due to the loss of reactor coolant l
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l Amendment No. M,M,6, M,G, M, 12
+w.178 l
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a THIS PAGE INTENTIONAI T Y LEFT BLANK l
Amend:nent 6, M, M,43, 63, g,4, M, 14b 443,443,178
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Table 2.3-1 Reactor Protection System Trip Setting Limits One Reactor Coolant Puup Four Reactor Coolant Pumps Three Reactor Coolant Pungs Operating in Each Loop *I l
I Operating (Nominal Operating (Nominal (Nominal Operating Shutdown
_ Operating Power - 100%)
Operating Power, 75%)
Power, 49%)
Bypass Nuclear power, % of 104.9 104.9 104.9 5.0(a) l rated, max Nuclear Power based on Protection System Maximum Protection' System Maximum Protection System Maximum Bypassed flow (b) and imbalance, Allowable Setpoints for Allowable Setpoints for Allowable Setpoints for
% of rated, max Axial Power Imbalance Axial Power Imbalance Axial Power Imbalance envelope in COLR envelope in COLR envelope in COLR Nuclear Power based on NA NA 55 Bypassed pump monitors, % of rated, max (c) l High RC system 2355 2355 2355 1720 *I l
I pressure, psig, max Low RC system 1800 1800 1800 Bypassed pressure, psig. min Variable low RC 13.89 Tout-6766(d) 13.89 Tout-6766(d) 13.89 Tout-6766(d)
Bypassed l system pressure, psig, min RC temp, F, max 618 618 618 618 Nigh reactor building 4(18.7 psla) 4(18.7 psia) 4(18.7 psia) 4(18.7 i
pressure, psig, max psia)
(a) Automatically set when other segments of the RPS (as specified) are bypassed.
(b) Reactor coolant system flow, %
(c)The pump monitors also produce a trip on (a) loss of two RC pumps in one RC loop, and (b) loss of one or two RC pumps during two-pump operation.
(d)T is in degrees Fahrenheit (F).
out (e) Operation with one Reactor Coolant Punp operating in each loop is limited to 24 hrs. with the reactor critica).
Amendment No. 4,44,43,49,64,44,94,444,444,178 15
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6 3.1 - REACTOR COOLANT SYSTEM'
- Applicability
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,1 Applies to the' operating statur of the reactor' coolant system.
objective To. specify those limiting conditions. for operation of the reactor ' coolant i
system which must be met to ensure safe reactor operations.
3.1.1 Operational Components Specification
' 3.1.1.1 Reactor Coolant Pumps j
A.
Pump combinations permissible for given power levels shall l
be as shown in Table 2.3-1.
Operation with one Reactor coolant Pump operating in each loop is limited to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> with the reactor critical.
B.
The boron concentration in the reactor coolant system shall
'l not be reduced unless at least one reactor coolant pump or one decay heat removal pump is circulating reactor coolant.
l With no reactor coolant pumps or decay heat removal pumps' running, immediately suspend all operations involving a reduction of boron concentration in the reactor coolant 4
system.
3.1.1.2 Steam Generator-A.
Two. steam generators shal1 ~ be ' operable whenever the reactor coolant average temperature is above 280'F.
3.1.1.3 Pressurizer Safety Valves A.
Both pressurizer code safety valves'shall be operable when the reactor is critical. With one pressurizer code' safety valve inoperable, either restore.
the valve to operable status within 15 minutes or be in NOT SHUTDOWN within.12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
B.
When the reactor is suberitical, at least one pressurizer code safety valve shall be operable if all reactor coolant system openings are closed, except for hydrostatic tests in accordance with ASME Boiler and Pressure Vessel Code,Section III. The provisions of Specification 3.0.3 are not applicable.
3.1.1.4 Reactor Internals Vent Valves The structural integrity and operability of the reactor internals vent valves shall be maintained at a level consistent with the acceptance criteria in Specification 4.1.
The provisions of Specification 3.0.3 are not applicable.
3.1.1.5 Reactor Coolant Loops l
A.
With the reactor coolant average temperature above 280*F, the reactor coolant loops listed below shall be operable:
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Amendment No. M,M,M,W,178 16 i
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3.1;8 Low Power Phy91cs T3" ting R9atrictirns Specification' The following special limitations are placed on low power physics testing.
3.1.8.1 Reactor Protective System Requirements j
A.
Below 1720 psig, shutdown bypass trip setting limics shall apply in accordance with Table 2.3-1.
B.
Above 1800 psig, nuclear overpower trip shall be set at less l
than 5.0 percent. Other settings shall be in accordance with Table 2.3-1.
l 3.1.8.2 Startup rate rod withdrawal hold (5) shall be in effect at all times.
i 3.1.B.3 During low power physics testing the minimum reactor coolant temperature for criticality shall be to the right of the criticality limit of Figure 3.1.2-2.
The shutdown margin shall be maintained greater than or equal to that specified in the COLR i
with the highest worth control rod fully withdrawn.
i Base 3, The above specification provides additional safety margins during low power physics testing.
j REFERENCES (1)
FSAR, Section 7.2.2.1.3.
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Amendment No. M,178 31
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3.5.2
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Control Rod Group and Power Distribution Limits
. 1
.Adplicability 1
This. specification applies to power distribution and operation of control' t
rods during power operation.
[
objective
[
To assure are acceptable core power distribution during power operation, to set a limit on potential reactivity insertion from a hypothetical control i
rod ejection, and to assure core suberiticality after a reactor trip.
l Specification 3.*. 2.1 The available~ shutdown margin shall be greater than or equal' to that specified in the COLR with the highest worth control rod fully withdrawn..With the shutdown margin less than that required, immediately initiate and continue boration injection until the required shutdown margin is restored.
t 3.5.2.2 Operation with inoperable rods:
G 1.
Operation with more thar one inoperable rod, as defined in
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Specification 4.7.1 and 4.7.2.3, in the safety or regulating rod groups shall not be permitted.
2.
If a control rod in the regulating or safety rod groups is declared inoperable in the withdrawn position as defined in Specification 4.7.1.1 and 4.7.1.3, an evaluation shall be initiated immediately to verify the existence of an available shutdown margin greater than or equal to that j
specified in the COLR. Boration may be initiated either to the worth of the inoperable rod or unti1~the regulating and transient rod groups are withdrawn to the lindts of
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Specification 3.5.2.5.3, whichever occurs.first.
Simultaneously a program of exercising the remainary regulating and safety. rods shall be initiated to v : ify operability.
3.
If within one (1) hour of determination of an inoperable rod as defined in Specification 4.7.1, it is not determined that an available shutdown margin greater than or equal to that specified in the COLR exists combining the wort h of the
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inoperable rod with each of the other rods, the reactor.
t shall be brought to the Hot Standby condition until this margin is established.
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4.
Following the determination of an inoperable rod as j
defined in Specification 4.7.1, all remaining rods shall be exercised 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
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the rod problem is solved.
5.
If a control rod in the regulating or safety rod groups is declared inoperable per 4.7.1.2, power shall be reduced to
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60% of the thermal power allowable for the reactor coolant pump combination.
l Arnendment No. M 178 46
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6.
If a cintral rod in th2 regulcting or exici power chtping groups is declared inoperable per specification 4.7.1.2 operation above 60 percent of the thermal power allowable for the reactor coolant pump combination may continue-
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provided_the rods in the group are positioned such that the rod that was declared inoperable is contained within
- allowable group average position limits of specification
-i 4.7.1.2 and the withdrawal limits of Specification 3.5.2.5.3.
-i 3.5.2.3 The worth of single inserted control rods during criticality
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are limited by the restrictions of specification 3.1.3.5 and the.
Control Rod Position Limits defined in specification 3.5.2.5.
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3.5.2.4 Quadrant Power Tilt i
1.
Except for physics. tests, if quadrant power tilt exceeds the tilt limit set in the CORE OPERATING LIMITS REPORT, reduce power so as not to exceed the allowable power level for the existing reactor coolant pump combination less at least 24 -
for each 14 tilt in excess of the tilt limit.
2.
Within a period of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, the quadrant power tilt shall be reduced to less than the tilt limit except for physics tests, or the following adjustments in setpoints and limits shall be mades a.
.The Protection System Maximum Allowable Setpoints for Axial Power Imbalance in the COLR shall be reduceo 26 in power for each 18 tilt in excess of the tilt limit, b.
The control rod group and APSR withdrawal limits shall be reduced 2% in power for each 1% tilt in excess of the tilt lindt.
c.
The reactor poNer imbalance setpoints shall be reduced 26 in power for each 1% tilt in excess of the tilt limit.
3.
If quadrant power tilt is in excess of 25%, except for i
physics tests or diagnostic testing, the reactor will be placed in the hot shutdown condition.
Diagnostic testing.
during power operation with a quadrant power tilt is permitted provided the thermal power allowable for the reactor coolant pump combination is restricted as stated in 3.5.2.4.1 above.
l 4.
Quadrant power tilt shall be monitored on a minimum frequency of once every two hours during power operation above 15% of rated power.
l Amendment No. 4,M,M,44,M,4M, 47 M 4, 9 4, R G.178
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6.12.3 CORE OPERATING LIMITS REPORT 6.12.3.1 The core operating limits shall be established and documented in the CCF3 OPERATING LIMITS REPORT prior to each reload cycle or prior to any remaining part of a reload cycle for the following specifications:
2.1 Safety Limits, Reactor Core -- Axial Power Imbalance protective limits 2.3.1 Reactor Protection system trip setting limits --
Protection System Maximum Allr>wable Setpoints for Axial Power Imbalance 3.1.8.3 Minimum Shutdown Margin for Low Power Physics Testing 3.5.2.1 Allowable Shutdown Margin limit during Power operation 3.5.2.2 Allowable shutdown Margin limit during Power operation with inoperable control rods 3.5.2.4 Quadrant power Tilt limit 3.5.2.5 Control Rod and APSR position limits 3.5.2.6 Reactor Power Imbalance limits 6.12.3.2 The analytical methods used to determine the core operating limits addressed by tne individual Technical specification shall be those previously reviewed and approved by the NRO in Babcock i
s Wilcox Topical Report RAN-10179P-A, " Safety criteria and Methodology for Acceptable Cycle Relead Analyses" (the approved revision at the time the reload analyses are performed). The approved revision nu.-ber shall be identified in the CCRE OPERATING LIMITS REPORT.
6.12.3.3 The core operating limits shall be determined so that all applicable limits (e.g. fuel thermal-mechanical limits, core then=al-hydraulic limits, ECCS limits, nuclear limits 'such as shutdown margin, and transient and acoident analysis limits) of the safety analysis are met.
6.12.3.4 The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements thereto, shall be provided upon issuance for each reload cycle to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.
t Amendment No. 24,G9, SS,H4A4G, 142 (next page is 146)
M e,178 i
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