ML20114A757
| ML20114A757 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 08/13/1992 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20114A758 | List: |
| References | |
| NUDOCS 9208210085 | |
| Download: ML20114A757 (16) | |
Text
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d 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS 1
THERMAL POWER, Lew Pressure or Low Flow 2.1. 4 1HERMAL POWER shall not exceed 25% of RATED THERMAL POWER with the reactor vessel steam dome pressure less inan 785 psig or core flow less than 10% of rated flow.
N APPLICABILITY:
OPERATIONAL CONDITIONS 1 and 2.
ACTION:
With THERMAL POWER exceeding 25% of RATED THERMAL POWER and the reactor vessel steam dome pressure less than 785 psig or core flow less than 10% of rated flow, be in at least HOT SHUTOOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirements of Specification 6.7.1.
THERMAL POWER, High Pressure and High Flow 2.1.2 The MINIMUM CRITICAL POWER RATIO (MCPR) shall not be less than 1.06 during both two loop operation and 1.07 during single loop operation with the reactor vessel steam dome pressure greater than 785 psig and core flow greater than 10% of rated flow.
APPLICABILITY:
OPERATIONAL CONDITIONS i and 2.
ACTION:
With MCDR less than the above limits and the reactor vessel steam dome pressure greater than 785 psig and core flow greater than 10% of-rated flow, te in at least HOT SHUTDOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirements of Specifi-cation 6.7.1.
REACTOR COOLANT SYSTEM PRESSURE 2.1.3 The reactor coolant system pressure, as measured in the reactor vessel steam come, shall not exceed 1325 psig.
APPLICABILITY:
OPERATIONAL CONDITIONS 1, 2, 3 and 4.
ACTION:
With the reactor coolant system pressure, as measured in the reactor vessel steam dome, above 1325 psig, be in at least HOT SHUTDOWN with reactor coolant system pressure less than or equal to 1325 psig within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirements of Specification 6.7.1.
9208210085 920s13 hDR ADOCK 05000416 PDR GRAND GULF-UNIT 1 2-1 Amendment No. 7), 99 i
9
_ 21 SAFETY LIMITS B_ASES
2.0 INTRODUCTION
The fuel cladding, reactor pressure vessel and primary system piping are the principal barriers to the release of radioactive materials to the environs.
Safety Limits are established to protect the integrity of these barriers during normal plant operations and anticipated transients. The fuel cladding integrity Safety Limit is set such that no fuel damage is calculated to occur if the limit is not violated.
Becauie fuel damage is not directly observable, a step-back approach is used to establish a Safety Limit for the MCPR.
MCPR greater than the applicable Safety Limit represents a conservative margin relative to the conditions required to maintain fuel cladding integrity.
The fuel _ cladding is one of the physical barriers which separate the radioactive materials from the environs.
The integrity of this cladding barrier is related to its relative freedom from perforations or cracking.
Although some corrosion or use related cracking may occur during the life of the cladding, fission product migration from this source is incrementally cumulative and continuously measurable.
Fuel cladding perforations, however, can result from thermal stresses which occur from reactor operation significantly above design condi-tions and the Limiting Safety System Settings.- While fission product migration from cladding perforation is just as measurable as that from use related cracking, the thermally caused cladding perforations signal a threshold beyond which still greater thermal stresses may cause gross rather than incremental cladding deterioration.
Therefore, the fuel cladding Safety Limit is defined with a margin to the conditions which would produce onset of transition boiling, MCPR of 1.0.
These conditions represent a significant departure from the condition intended by design for planned operation.
2.1.- 1 THERMAL POWER, Low Pressure or Low Flow The Siemens Nuclear Power Corporation (SNP) ANFB critical power correla-tion is applicable to the SNP core.
The applicable range of the ANFB correla-tion is for pressures above 585 psig and bundle mass flux greater ' Nan 0.25M1bs/
d 2
br-ft.
For low pressure and low flow conditions, a THERMAL POWER safety limit of 25% of RATED THERMAL POWER for reactor pressure below 785 psig anet below 10%
RATED CORE FLOW was justified for Grand Gulf cycle 1 operation based on ATLAS test data and the GEXL correlation.
The use of the GEXL correlation is not valid for all critical power calculations at pressures below 785 psig or core flows less than 10% of rated flow.
Therefore, the fuel cladding integrity Safety Limit was established by other means.
This was done by establishing a limiting condition on core THERMAL POWER with the following basis.
Since the pressure drop in the bypass region is essentially all elevation head, the core pressure drop at low power and flows will always be greater than 4.5 psi.
GRAND GULF-UNIT 1 B 2-1 Amendment No. 7), gg
SAFETY-LIMITS BASES 2.1.2 THERMAL POWER, High Pressure and High Flow The onset of transition boiling results in a decrease in heat transfer from the clad, elevated clad temperature, and the possibility of clad failure.
However, the existence of critical power, or boiling transition, is not a di-rectly observable parameter in an operating reactor.
Therefore, the margin to boiling transition is calculated from plant operating parameters such as core power, core flow, feedwater temperature, and core power distribution.
The mar-gin for each fuel assembly is characterized by the critical power ratio (CPR),
which is the ratio of the bundle power which would produce onset of tran.cition boiling divided by the actual bundle power.
The minimum value of this rc.wo for any bundle in the core is the minimum critical power r3tio (MCPR).
The Safety Limit MCPR assures sufficient conservatism such that, iri the event of a sustained steady state operation at the MCPR safety limit, at least 99.9% of the fuel rods in the core would be expected to avoid Loiling transi-tion.
The margin between calculated boiling transition (MCPR = 1.00) and the Safety Limit MCPR is based on a detailed statistical procedure which considers
.the uncertainties in monitoring the core operatirig state and includes the effects associated with channel bow.
One specific uncertainty included in the safety limit is the uncertainty inherent in the ANFB critical power correlation.
SNP report ANF-524 (P)(A), Rev. 2, " Advanced Nuclear Fuels Corporation Critical Power Methodology for Boiling Water Reactors," April 1989, including supplements, describes the methodology used in determining the Safety Limit MCPR.
The ANFB. critical power correlation is based on a significant body of practical test data, providing a high degree of assurance that the critical power as evaluated by the' correlation is within a small percentage of the
-actual critical pe-:r being estimated.
The assumed reactor conditions used in defining the safety limit introduce conservatism into the limit because bound-ing radial power factors and bounding flat local peaking distributions are used-to estimate the number of rods in boiling transition.
Still further con-servatism is induced by the tendency of the ANFB correlation to overpredict the
-number of rods-in boiling transition.
These conservatisms and the inherent accuracy of the ANFB correlation provide assurance that during sustained opera-tion at the Safety Limit MCPR there would be essentially no transition boiling in the core.
I J
GRAND GULF-UNIT I B 2-2 Amendment No. 77, 99
3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANA 7 LINEAR HEAT GENERATION RATE LIMITING CONDITION FOR OPERATION 3.2.1 Ouring two loop o,eration, all AVERAGE PLANAR LINEAR HEAT GENERATION RATES (APLHGRs) for each type of fuel as a function of AVERAGE PLANAR EXPOSURE shall not exceed the limits shown in Figure 3.2.1-1.
During single loop operation, the APLHGR for each type of fuel as a function of AVERAGE PLANAR EXPOSURE shall not exceed the limit shown in Figure 3.2.1-1 t itiplied by 0.86.
l APPLICABILITY:
OPERATIONAL CONDITION 1, when' THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER.
ACTION:
During tw's loop operation or single loop operation, with an APLHGR exceeding the limits of Figure 3.2.1-1 as corrected by the appropriate multiplication factor, initiate corrective action within 15 minutes and restore APLHGR to within the required limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
SURVEILLANCE REQUIREMENTS 4.2.1 All APLH3Rs shall be verified to be equal to or less than the required limits:
a.
At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, b.
Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER, and c.
Initially and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor is operating with a LIMITING CONTROL R0D PATTERN for APLHGR.
d.
The provisions of Specification 4.0.4 are not applicable.
GRAND GULF-UNIT 1 3/4 2-1 Amendment No. AG, 99
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3/4.2 POWER DISTRIBUT10h LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE LIMITING CONDITION FOR OPERATION 3.2.1 During two loop operation, all AVERAGE PLANAR LINEAR HEAT GENERATION RATES (APLHGRs) for each type of fuel as a function of AVERAGE PLANAR EXPOSURE shall not exceed the limits shown in Figure 3.2.1-1.
During single loop operation, the APLHGR for each type of fuel as a function of AVERAGE PLANAR EXPOSURE shall not exceed the limit shown in Figure 3.2.1-1 multiplied by 0.86.
l APPLICABILITY:
OPERATIONAL CONDITION 1, when' THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER.
ACTION:
During two loop operation or single loop operation, with an APLHGR exceeding the limits of Figure 3.2.1-1 as corrected by the appropriate multiplication factor, initiate corrective action within 15 minutes and restore APLHGR to within the required limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
SURVEILLANCE REQUIREMENTS 4.2.1 All APLHGRs shall be verified to be equal to or less than the required limits:
-a.
At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />,
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b.
Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER, and c.
Initially and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor is operating with a LIMITING CONTROL R00 PATTERN for APLHGR.
d.
The provisions of Specification 4.0.4 are not applicable.
GRAND GULF-UNIT 1 3/4 2-1 Amendment No, MG, 99
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4 POWER DISTRIBUTION L1 HITS 3/4.2.4 LINEAR HEAT GENERATION RATE L1HITING CONDIT]ON FOR OPERATION 3.2.4 The LINEAR HEAT GENERATION RATE (LHGR) shall not exceed the limits shown in Figure 3.2.4-1 as multiplied by the smaller of either the flow-dependent LHGR factor (LHGRFAC ) of Figure 3.2.4 2, or the power-dependent LHGR factor 9
(LHGRFAC ) of Figure 3.2.4 3.
APPLICACILITY:
OPERATIONAL COND] TION 1, when THERMAL POWER is greater than or equal to 25L of RATED THERMAL POWER.
ACTION:
With the LHGR of any fuel red e.tdeding the limit of Figure 3.2.41, as cor-rected by the appropriate multiplication factor, initiate corrective action within 15 minutes and restore the LHGR to within the limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
SURVEILLANCE REQUIREMENTS 4.2.4 LHGR's shall be
'termined to be equal to or less than their allowable limits; a.
At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, b.
Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER, Initially and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor is c.
operating on a LIMITING CONTROL ROD PATTERN for LHGR, and d.
The provisions of Specification 4.0.4 are not applicable.
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3/4.2 POWER DISTRIBUTION LIMITS B_ASE S The specifications of this section assure that the peak cladding temper-ature following the postulated design basis loss-of coolant accident will not exceed the 2200'F limit specified in 10 CFR 50.46.
3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE This specification assures that the peak cladding temperature following the postulated design basis loss-of-coolant accident will not exceed the limit specified in 10 CFR 50.46.
The peak cladding temperature (PCT) following a postulated loss-of-coolant accident is primarily a function of the avera0e heat generation rate of all the rods of a fuel assembly at any axial location and is dependent only secondarily on the rod to rod power distribution within an assembly.
The Maximum Average P '. s er Linear Heat Generation Rate (MAPLHGR) limits of Figure 3.2.1-1 are appiicable to two loop operation.
For single-loop (peration, a MAPLHGR limit corresponding to the product of the MAPLHGR, Figare 3.2.1-1, and 0.86 can be conservatively use: to ensure l
that the PCT for single loop operation is bounded by the PCT for two loop operation.
The daily requirement for calculating APLHGR when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER is sufficient since power distribu-tion shifts are very slow when there have not been significant power or control i
GRAND GULF-UNIT I B 3/4 2-1 Amendment No.
1(99
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