ML20205F963
| ML20205F963 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 04/01/1999 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20205F969 | List: |
| References | |
| NUDOCS 9904070071 | |
| Download: ML20205F963 (3) | |
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BASES (continued)
. APPLICABILITY SL 2.1.2 applies in MODES 1,2,3,4, and 5 because this SL could be approached or exceeded in these MODES due to overpressurization events. The SL is not applicable in MODE 6 because the reactor vessel head closure bolts are not fully tightened, making it unlikely that the RCS can be pressurized.
SAFETY LIMIT If the RCS pressure SL is violated when the reactor is in MODE 1 or 2,.
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VIOLATION the requirement is to restore compliance and be in MODE 3 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
Exceeding the RCS pressure SL may cause immediate RCS failure and create a potential for radioactive releases in excess of 10 CFR 100,
" Reactor Site Criteria," limits (Ref. 4).~
The allowable Completion Time of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> recognizes the importance of reducing power level to a MODE of operation where the potential for challenges to safety systems is minimized.
If the RCS pressure SL is exceeded in MODE 3,4, or 5, RCS pressure l
must be restored to within the SL value within 5 minutes. Exceeding the l
RCS pressure SL in MODE 3,4, or 5 is more severe than exceeding this l
SL in MODE 1 or 2, since the reactor vessel temperature may be lower and the vessel material, consequently, less ductile. As such, pressure must be reduced to less than the SL within 5 minutes. The action does not require reducing MODES, since this would require reducing temperature, which would compound the problem by adding thermal gradient stresses to the existing pressure stress.
Per 10CFR50.36, if a Safety Limit is violated, operations must not be resumed until authorized by the Commission.
REFERENCES 1.
10 CFR 50, Appendix A, GDC 14, GDC 15, and GDC 28.
2.
ASME, Boiler and Pressure Vessel Code, Section Ill, Article NB-7000.
(continued) 9904070071 9904o1 yDR ADOCK 05000445 PDR COMANCHE PEAK - UNITS 1 AND 2 B 2.0-7 Amendment No. 64 i
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Coro Rsactivity B 3.1.2 i
BASES (continued)
SURVEILLANCE SR 3.1.2.1 REQUIREMENTS Core reactivity is verified by periodic comparisons of measured and predicted RCS boron concentrations. The comparison is made, considering that other core conditions are fixed or stable, including control rod position, moderator temperature, fuel temperature, fuel depletion, xenon concentration, and samarium concentration. The Sunreillance is performed prior to entering MODE 1 as an initial check on core conditions and design calculations at BOC. The SR is modified by a Note. The Note requires that the normalization of predicted core j
reactivity to the measured value must take place within the first I
60 effective full power days (EFPD) after each fuel loading. However, if the deviation between measured and predicted values is within the associated measurement and analytical uncertainties, it is not necessary to normalize the predicted core reactivity. This allows sufficient time for core conditions to reach steady state, but prevents operation for a large fraction of the fuel cycle without estaF ling a benchmark for the design calculations. The required subsequent Frequency of 31 EFPD, following the initial 60 EFPD after entering MODE 1, is acceptable, based on the slow rate of core changes due to fuel depletion and the presence of oth3r indicators (OPTR, AFD, etc.) for prompt indication of an anomaly.
REFERENCES 1.
10 CFR 50, Appendix A, GDC 26, GDC 28, and GDC 29.
2.
FSAR, Chapter 15.
COMANCHE PEAK - UNITS 1 AND 2 8 3.1-12 Amendment No. 64 i
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- r OPTR B 3.2.4 V
. B 3.2 ' POWER DISTRIBUTION LIMITS B 3.2.4 QUADRANT POWER TILT RATIO (OPTR)
BARES BACKGROUND
' The OPTR limit ensures that the gross radial power distribution remains consistent with the design values used in the safety analyses. ' Precise radial power distribution measurements are made during startup testing,
- after refueling, and periodically during power operation.
The power density at any point in the core must be limited so that the fuel design criteria are maintained. Together, LCO 3.2.3, " AXIAL FLUX DIFFERENCE (AFD)," LCO 3.2.4, and LCO 3.1.7, " Control Rod insertion Limits," provide limits on process variables that characterize and control the three dimensional power distribution of the reactor core. Control of these variables ensures that the core operates within the fuel design criteria and that the power distribution remains within the bounds used in the safety analyses.
APPLICABLE This LCO precludes core power distributions that violate the following SAFETY fuel design criteria:
ANALYSES a.
During a large break loss of coolant accident, the peak cladding temperature must not exceed 2200"F (Ref.1);
b.
During a loss of forced reactor coolant flow accident, there must be at least 95% probability at the 95% confidence level (the 95/95 departure from nucleate boiling (DNB) criterion) that the hot fuel rod in the core does not experience a DNB condition; c.
During an ejected rod accident, the average fuel pellet enthalpy at the hot spot must not exceed 280 cal /gm (Ref. 2); and d.
The control rods must be capable of shutting down the reactor with a minimum required SDM with the highest worth control rod stuck fully withdrawn (Ref. 3).
(continued)
COMANCHE PEAK - UNITS 1 AND 2 B 3.2-27 Amendment No. 64
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