ML20045C681
| ML20045C681 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 06/16/1993 |
| From: | Marsh L Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20045C682 | List: |
| References | |
| NUDOCS 9306240174 | |
| Download: ML20045C681 (9) | |
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UNITED STATES i
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j NUCLEAR REGULATORY COMMISSION f
WASHINGTON D.C. 20005 0001 4
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.s CONSUMERS POWER COMPANY DOCKET NO. 50-255 PALISADES PLANT AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.156 License No. DPR-20 l.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Consumers Power Company (the licensee) dated January 29,-1993, as supplemented April-20, 1993, 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 (i) 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 Commission's regulations; D.
The issuance of this amendment will not be inimical to the common-defense and security or to the health and safety of the public; 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.
2.
Accordingly, the license.is amended by changes to'the Technical Specifications as indicated-in the attachment to the license amendment:and-Paragraph 2.C.(2) of Facility Operating License No. DPR-20 is hereby amended to read as follows:
9306240174 930616 ADOCK0500g5 PDR
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 156, and the Environmental Protection Plan contained in Appendix B are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
This license amendment is effective as of the date of issuance.
FOR THE NUCLEAR RE ULA ORY COMMISSION M
8 Ledyard B. Ma h, Director Project Directorate III-l Division of Reactor Projects - III/IV/V Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: June 16,1993 1
ATTACHMENT TO LICENSE AMENDMENT N0.156 FACILITY OPERATING LICENSE NO. DPR-20 i
-DOCKET NO. 50-255 Revise Appendix A Technical Specifications by removing the pages identified below and inserting the attached pages. The revised pages are identified by amendment number and contain vertical lines indicating the areas of change.
REMOVE INSERT B 2-1 B 2-1 B 2-5 B 2-5 3-3 3-3 3-67 3-67 3-107 3-107 3-111 3-111 p..
2.0 BASIS - Safety Limits and Limitino Safety System Settinos 2.1 Basis - Reactor Core Safety limit To maintain the integrity of the fuel cladding and 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" (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 parameters of thermal power, primary coolant flow, temperature and pressure, can be related to DNB through the use of a DNB Correlation. DNB Correlations have been developed to predict DNB and the location of DNB for axially uniform and nonuniform heat flux distributions.
The local DNB ratio (DNBR) 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 DNB correlation safety limit. A DNBR equal to the DNB correlation safety limit corresponds to a 95% probability at a 95% confidence level that DNB will not occur which is considered an appropriate margin to DNB for all operating conditions.
The reactor protective system is designed to prevent any anticipated combination of transient conditions for primary coolant system temperature, pressure and thermal power level that would result in a DNBR of less than the DNB correlation safety limit.
The Palisades safety analyses uses two DNB correlations. The XNB correlation discussed in References 1 and 2 determines the safety limit for those fuel assemblies initially loaded prior to Cycle 9.
The ANFP correlation discussed in References 4 and 5 determines the safety limit for those fuel assemblies initially loaded in Cycle 9 and later.
Fuel assemblies initially loaded prior to Cycle 9 are of a different construction than later assemblies which utilize a High Thermal Performance design.
The minimum DNBR analyses are in accordance with Reference 6.
References 1
XN-NF-621(P)(A),Rev1 2
XN-NF-709 3
Updated FSAR, Section 14.1.
, May 1989 4
ANF-1224 (P A) January 1990 5
ANF-89-192L 6
XN-NF-82-21
), Revision 1 Amendment No. JJ, /J, JJE, JJJ, JES, 156 B 2-1
2.0 SASIS - Safety Limits and Limitina Safety System Settinas 2.3 Basis - Limitina Safety System Settinas (continued) 5.
Low Steam Generator Water Level - The low steam generator water level reactor trip protects against the loss of feed-water flow accidents and assures that the design pressure of the primary coolant system will not be exceeded. The specified set point assures that there will be sufficient water inventory in the steam generator at the time of trip to allow a safe and orderly plant shutdown and to prevent generatordryoutassumingminimumauxiliaryfeedwatercapacity.geam The setting listed in Table 2.3.1 assures that the heat transfer surface (tubes) is covered with water when the reactor is critical.
6.
Low Steam Generator Pressure - A reactor trip on low steam generator secondary pressure is provided to protect against an excessive rate of heat extraction from the steam generators and subsequent cooldown of the primary coolant.
The setting of 500 psia is sufficiently below the rated load operating point of 739 psia so as not to interfere with normal operation, but still high enough to provide the required protection in the event of excessi was used in the accident analysis. gly high steam flow.
This setting 7.
Containment Hiah Pressure - A reactor trip on containment high pressure is provided to assure that the reactor is shutdown befoge the initiation of the safety injection system and containment spray. '
References EMF-92-178, Table 15.0.7-1 l
Updated FSAR, Section 7.2.3.3.
EMF-92-178, Section 15.0.7-1 l
XN-NF-86-91(P)
ANF-90-078, Section 15.1.5 ANF-87-150(NP), Volume 2, Section 15.2.7 Updated FSAR, Section 7.2.3.9.
ANF-90-078, Section 15.2.1 Amendment No JJ, EJ, JJE, JJ/, JEE, 156 B 2-5
V-3.1 PRIMARY COOLANT SYSTEM (Cont'd)
Basis (Cont'd) measurement 10.06 for ASI measurement; 150 psi for pressurizer pressure; 17'F for iniet temperature; and 3% measurement and 3% bypass for core l
fl ow.
In addition, transient biases were included in the derivation of the following equation for limiting reactor inlet temperature:
)
T s 542.99 +.0580(P-2060) + 0.00001(P-2060)**2 + 1.125(W-138) -
.0205(W-138)* 2 q
The limits of validity of this equation are:
1800 s pregsure s 2200 psia 100.0 x 10 s Vessel Flow s 150 x 10' lb/h ASI as shown in Figure 3.0 With measured primary coolant system flow rates > 150 M lbm/hr, limiting the maximum allowed inlet temperature to the T *her PCS flow rates *.
l LCO at 150Mlbm/hrincreasesthemargintoDNBforhYg The Axial Shape Index alarm channel is being used to monitor the ASI to ensure that the assumed axial power profiles used in the development of the inlet temperature LCO bound measured axial power profiles. The signal power and the Delta-T pow (Q).is the auctioneered higher of the neutron flux representing core power er The measured ASI calculated from the excore detector signals and adjusted for shape annealing (Y) and the core power i
constitute an ordered pair (0,Y,)daries specified in Figure 3.0.An alarm signal is activ the ordered pair exceed the boun The requirement that the steam generator temperature be s the PCS temperature when forced circulation is initiated in the PCS ensures that an energy addition caused by heat transferred from the secondary system to the PCS will not occur.
This requirement applies only to the initiation of forced circulation (the start of the first primary coolant pump) when the PCS cold leg temperature is < 430'F.
However, analysis shows that under limited conditions when the Shutdown Cooling (Reference 6)
System is isolated from the PCS, forced circulation may be initiated when the steam generator temperature is higher than the PCS cold leg temperature.
References Updated FSAR, Section 14.3.2.
Updated FSAR, Section 4.3.7.
Deleted EMF-92-178 Section 15.0.7.1 l
ANF-90-078 Consumers Power Company Engineering Analysis EA-A-NL-89-14-1 3-3 Amendment No. 71, 5), JJ1 115, J)J, 111, 117, Jf7, 156
_ _ _ _ _ =.. _.
4
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3.12 MODERATOR TEMPERATURE COEFFICIENT OF REACTIVITY Acolicability Applies to the moderator' temperature coefficient of reactivity for the Core.
Ob.iective To specify a limit for the positive moderator coefficient.
Specifications The moderator (emperature coefficient (MTC) shall be less positive than +0.5 x 10~ Ap/*F at s 2% of rated power, biti The limitations on moderator temperature coefficient (MTC) are provided to ensure that the assumptions used in the safety analysis")
remain valid.
Reference (1)
EMF-92-178, Section 15.0.5 l
t i
w k
2 t
t i
t 3-67 Amendment No. JJE, JJJ,.JfJ, 156 4
(next page is 3-69)
TABLE 3.23-1 LINEAR HEAT RATE LIMITS No. of Fuel Rods Assembly 208 216 Peak Rod 15.28 kW/ft 15.28 kW/ft TABLE 3.23-2 RADIAL PEAKING FACTOR LlHITS, Ft Peaking Factor No. of Fuel Rods in Assembly 216 216 216 208 Reload M Reload N Reload 0 AssemblyFj 1.48 1.57 1.66 1.76 PeakRodFi 1.92 1.92 1.92 2.04 TABLE 3.23-3 POWER DISTRIBUTION MEASUREMENT UNCERTAINTY FACTORS LHR/ Peaking Factor Measurement Measurement Measurement Parameter Uncertainty
Uncertainty
- Uncertainty")
LHR 0.0623 0.0664 0.0795 F$
0.0401 0.0490 0.0695 F}
0.0455 0.0526 0.0722 (a)
Measurement uncertainty for reload cores using all fresh incore detectors.
(b)
Measurement uncertainty for reload cores using a mixture of fresh and once-burned incore detectors.
(c)
Measurementuncertaintywhenquadrantpowertilt,asdetermip)edusing incore measurements and an incore analysis computer program', exceeds 2.8% but is less than or equal to 5%.
3-107 Amendment No. JE, JJE, Jf), Jff,156
POWER DISTRIBUTION LIMITS 3.23.2 RADIAL PEAKING FACTORS LIMITING CONDITION FOR OPERATION A
T The radial peakin$mes the fofiowing Qu. shall be less than or equal to the value factors F and F in Table 3.23-2 t antity. The for P 2.5 and the quantity is 1.15 for P <
.5. quantity is [1.0 + 0.3 (1 -P is the core th P ] fraction of rated power.
i APPLICABILITY:
Power operation above 25% of rated power.
ACTION:
1.
For P < 50% of rated with any radial peaking factor exceeding its -
limit, be in at least hot shutdown within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
2.
For P 2 50% of rated with any radial peaking factor exceeding its limit, reduce thermal power within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> to less than the lowest value of:
[1
- 3.33(
r - 1) ] x Rated Power F
L A
T Where F is the measured value of either F, or F and F r
r L
is the Earresponding limit from Table 3.23 z.
Basis A
T The limitations on F, and F are provided to ensure that assumptions-used.in the analysis for estEblishin5 DNB margin, LHR and the thermal margin / low-pressure and variable high-power trip set points remain valid during operation.
Data from the incore detectors are used for determining the measured radial peaking factors. The periodic surveillance requirements for determining the measured radial peaking factors provide assurance that they remain within prescribed limits. Determining the measured radial peaking factors after each fuel loading prior to exceeding 50% of rated power provides additional assurance that the. core is properly loaded.
To ensure that the design margin of safety is maintained, the determination of radial peaking factors takes into account the appropriate measurement uncertainty factors"' given in Table 3.23-3 References (1)
FSAR Section 3.3.2.5 3-111 Amendment No.JE, JJE, JJ7, Jf), J44,156 7,.
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