ML20128C272
| ML20128C272 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 11/23/1992 |
| From: | Dyer J Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20128C278 | List: |
| References | |
| NUDOCS 9212040319 | |
| Download: ML20128C272 (7) | |
Text
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4p u n,o, UNIT ED STATES l '
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'n NUCLE AR REGULATORY COMMISSION
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t W A EHINGTON, D. C. 20555 r
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COMMONW1ALTH EDISON COMPANY QQ[KET fl0, 50-237 Q.RISMN NQCLEAR POWER STATIDik URlLZ t
AMENDMENT TO FAClllTY OPERATING LICENSE Amendment No. 121 License No. DPR-19 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by the Commonwealth Edison Company (the licensee) dated September 2, 1992, complies with the standards and requirements of the Atomic Energy Att of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter 1:
B.
The facility will operate in conformity with the applici. tion, 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; 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.
2.
Accordingly, the license is amended by changes to the Technical Specifi-cations as indicated in the attachment to this license amendment and paragraph 2.C.(2) of facility Operating License No. DPR-19 is hereby amended to read as follows:
9212040319 921123 PDR ADOCK 05000237 p
i (2)
Jechnical Specifintioni The Technical Specifications contained in e nendix A, as revised through Amendment No. 121, are hereby inco., orated it the license.
The licensee shall operate the facility in accordance with the Technical Specifications, 3.
This license amendment is effective prior to startup from the next refueling outage (Cycle 14).
FOR THE NUCLEAR REGULATORY COMhlSS10N (Ni W
J.nes E. Dyer, Dir,ector P oject Directorate 111-2 Division of Reactor Projects - lil/IV/V Office of Nuclear Reactor Regulation-
Attachment:
Changes to the Technical Specifications Date of Issuance:
November 23, 1992
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ATTACHMENT TO LICENSE AMENDMENT N0. 1?1
[ACILLTY OPERATING LICENSE NO. DPR-19 DOCKET 40. 50-237 Revise the Appendix A Technical Specifications by removing the pages identified below and inserting the attached pages.
The revised pages are j
identified by the captioned amendment number and contain marginal lines
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indicating the area of change.
REMOVE INSERT 1/2.1-1 1/2.1-1 B 1/2.1-7 B 1/2.1 B 1/2.1-8 B 1/2.1-8 6-19 6-19 F
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l DRESDEN 11 DPR-19 i
Amendment No. 121 1.1 SAFETY LIMIT 2.3 LIMITING SAFETY SYSTEM SETTING FUEL CLADDING INTEGRITY rVEL CLADDING INTEGRITY Applicability:
Applicability:
The $afety Limits established The Limiting Safety System Set-to preserve the fuel claddirg tings apply to trip settings of integrity apply to these the instruments and devices variables which monitor the which are provided to prevent j
fuel thermal behavior.
the fuel cladding integrity Safety Limits from being exceeded.
Objective:
Objective:
The objective of the Safety
-The objective of the Limiting Limits is to establish limits Safety System Settings is to below which the integrity of define the level of the process the fuel cladding is preserved, variables at which automatic protective action is initiated to prevent the-fuel cladding integrity Safety Limits from being exceeded.
Specifi;ations:
Specifications:
A.
Reactor Pressure greater than A.
Neutron Flux Trip Settings 800 psig and Core Flew greater The limiting safety system than 10% of Rated, trip settings shall be at t
The existence of a minimum critical power ratio (MCPR) less than 1.08 shall constitute 1.
APRM Flux Scram Trip violation of the MCPR fuel Setting (Run Mode) cladding integrity safety When the reactor mode limit.
switch is in the run position, the APRM flux scram setting shall be:
When in Single Loop Operation, S less than or equal to
[.58W 62] during Dual the MCPR safe *,y limit shall be increased by 0.01.
LoopOp+erationorSless than or equal to [.58Wn+
58.5] during Single LoDp Operation with a maximum setpoint of 120% for core flow equal to
-98 x 106 lb/hr and greater, where:-
s - setting in percent of rated thermal power, 1/2.1-1
DRESDEN II DPR-19 Amendment No.121 1.1 SAFETY LIMIT BASES (Cont'd.)
power ratio (CPR) which is the ratio of the bundle power which would pro-duce the onset of tran:; tion boiling divided by the actual bundle power.
The minimum value of this ratio for any bundle in the core is the Minimum Critical Power Ratio (MCPR).
It is assumed that the plant operation is controlled to the nominal protective setpoints via the instrumented variables.
(Figure 2.1-3).
The MCPR Fuei Cladding Integrity Safety Limit assures sufficient conserva-tism in the operating MCPR limit that in the event of an anticipated operational occurrence from the limiting condition for operation, at least 99.9% of the fuel rods in the core would l'e expected to avoid boiling transition.
The margin between calculated boiling transition (MCPR=1.00) and the MCPR Fuel Cladding Integrity Safety Limit is based on a detailed statistical procedure which considers the uncertainties in monitoring the core operating state.
One specific uncertainty included in the safety limit is the uncertainty inherent in the NRC-approved critical power l
correlation.
Refer to Specification 6.6.A.4 for the methodology used in determining the MCPR Fuel Cladding Integrity Safety Limit.
The NRC-approved critical power correlation is based on a significant body l
of practical test data, providing a high degree of assurance that the criti-cal power as evaluated by the correlation is within a small percentage of the actual critical power being estimated.
The assumeu reactor conditions used in defining the safety limit introduce conservatism into the limit because boundingly high radial power peaking factors and boundingly flat local peaking distributions are used to estimate the number of rods in boiling transition.
Still further conservatism is induced by the tendency of the NRC-approved l
correlation to overpredict the number of rods in boiling transition.
These conservatisms and the inherent accuracy of the NRC-approved correlation i
provide a reasonable degree of assurance that during sustained operation at the MCPR Fuel Cladding Integrity Safety Limit there would be no transition boiling in the core.
If boiling transition were to occur, however, there is reason to believe that the integrity of the fuel would not necessarily be compromised.
Significant test data accumulated by the U.S. Nuclear Regulatory Commission and private organizations indicate that the use of a boiling transition limitation to protect against cladding failure is a very conservative approach; much of the data indicates that. LWR fuel can survive for an extended period in an environment of transition boiling.
During Single Loop Operation, the MCPR safety limit is increased by 0.01 to conservatively account for increased uncertainties in the core flow and TIP measurements.
l l.
B 1/2.1-7
4 ORESDEN II.
OPR-19 Amendment No.121 1.1 SAFE 1Y LIMIT BASES (Cont'd.)
If the reactor pressure should ever exceed the limit of dpplicability of the NRC-approved critical power correlation as defined in the NRC approved l
methodology listed in Specification 6.6.A.4, it would be assumed that the MCPR Fuel Cladding Integrity Safety Limit had been violated.
This applicability pressure limit is higher than the pressure safety limit specified in Specification 1.2.
Fuel design criteria f. ave been established to provide protection against fuel centerline melting and 1% plastic cladding strain during transient overpower conditions throughout the life of the fuel.
To demonstrate compliar.ce with these criteria, fuel rod centerline temperatures are determined at 120% over-power conditions as a check against calculated centerline melt temperatures.
FDLRC is incntporated to protect the abov'e criteria at all power levels consid-ering event, which cause the reactor power to increase to 120% of rated thermal power.
B.
Core The*tmal Power Limit-(Reactor Pressure less than 800 psia)
At pressures below 800 psia, the core elevation pressure drop (0 power, O flow) is greater than 4.56 psi.
At low powers and flows this pressure differential is maintained in the bypass region of the core.
Since the pressure drop in the bypass region is essentially all elevation head, the core pressure drop at low powers and ficws willalwgysbegreaterthan4.56 psi.
Analyses show that with a flow of 28x10 lbs/hr. bundle flow, bundle pre uure drop is-nearly inde-pendent of bundle power and has a value of 2.5 psi.
Thus, the bundle flow i
l l
B 1/2.1-8
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DRESDEN II DPR-19 Amendment No.121
- 6. 0 ADMINISTRATIVE CONTROLS (Cont'd.)
4.
Core Operating Limits Report a.
Core operating limits shall be established and documented 11 the Core Operating Limits Report before each reload cycle or any remaining part of a reload cycle for the following:
1)
The Control Rod Withdrawal Block Instrumentation for Table 3.2-3 of Specification 3.2.C.
2)
The Averaga Planar Line:- Heat Generation Rate (^PLHGR)
Limit and associat.ed APoiGR multipliers for Specifi-cations 3.5.I, 3.5.D.2, and 3.6.H.3.f..
3)
The L cal Steady State Linear Heat Generation Rate (LHGR) for Specification 3.5 J.
4)
The Local Transient Linear Heat Generation Rate (LHGR) for Specification 3.5.K.
5)
The Minimum Critical Power Operating Limit for Specification 3.5.L.
This includes rated and off-rated flow conditions.
b.
The analytical niethods used to determine the core operating limits shall be those previously reviewed and approved by the NRC in the latest approved revision or supplement of the topical reports describing the methodology.
For Dresden Unit 2, the topical reports are:
1)
ANF-1125(P)(A), " Critical Power Correlation - ANFB."
l 2)
ANF-524(P)(A), "ANF Critical Power Methodology for Boiling l
Water Reactors."
3)
XN-NF-79-71(P)(A), " Exxon Nuclear Plant Transient Methodology for Boiling Water Reactors".
4)
XN-NF-80-19(P)(A), " Exxon Nuclear Methodology for Boiling Water Reactors".
5)
XN-NF-85-b7(P)(A), " Generic Mechanical Design for Exxon Nuclear. Jet Pump Boiling Water Reactors Reload Fuel".
6)
XN-NF-81-22(P)(A), " Generic Statistical Untertainty.
Analysis Methodology".
7)
'ANF-913(P)(A),"COTRANSA2:
A Computer Program for Boiling Water Reactor Transient Analyses."
6-19