ML040150598
| ML040150598 | |
| Person / Time | |
|---|---|
| Site: | Mcguire, McGuire  |
| Issue date: | 01/14/2004 |
| From: | NRC/NRR/DLPM/LPD2 |
| To: | |
| References | |
| TAC MB8361, TAC MB8362 | |
| Download: ML040150598 (15) | |
Text
SLs 2.0 2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.1.1 Reactor Core SLs In MODES 1 and 2, the combination of THERMAL POWER, Reactor Coolant System (RCS) highest loop average temperature, and pressurizer pressure shall not exceed the limits specified in the COLR for four loop operation; and the following SLs shall not be exceeded:
2.1.1.1 The departure from nucleate boiling ratio (DNBR) shall be maintained
>1.14 for the WRB-2M CHF correlation.
2.1.1.2 The peak fuel centerline temperature shall be maintained <5080 degrees F, decreasing 58 degrees F for every 10,000 MWd/mtU of fuel burnup.
2.1.2 RCS Pressure SL In MODES 1, 2, 3, 4, and 5, the RCS pressure shall be maintained < 2735 psig.
2.2 SL Violations 2.2.1 If SL 2.1.1 is violated, 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 />.
2.2.2 If SL 2.1.2 is violated:
2.2.2.1 In MODE 1 or 2, 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 />.
2.2.2.2 In MODE 3, 4, or 5, restore compliance within 5 minutes.
McGuire Units 1 and 2 2.0-1 Amendment Nos.2 1 9 & 201
Reactor Core SLs B 2.1.1 BASES APPLICABLE The fuel cladding must not sustain damage as a result of normal SAFETY ANALYSES operation and A0Os. The reactor core SLs are established to preclude violation of the following fuel design criteria:
- a. There must be at least 95% probability at a 95% confidence level (the 95/95 DNB criterion) that the hot fuel rod in the core does not experience DNB; and
- b. The hot fuel pellet in the core must not experience centerline fuel melting.
The Reactor Trip System setpoints (Ref. 2), in combination with all the LCOs, are designed to prevent any anticipated combination of transient conditions for Reactor Coolant System (RCS) temperature, RCS Flow Rate, Al, pressure, and THERMAL POWER level that would result in a departure from nucleate boiling ratio (DNBR) of less than the DNBR limit and preclude the existence of flow instabilities.
Automatic enforcement of these reactor core SLs is provided by the appropriate operation of the RPS and the steam generator safety valves.
The SLs represent a design requirement for establishing the RPS trip setpoints identified previously. LCO 3.4.1, 'RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits,' or the assumed initial conditions of the safety analyses (as indicated in the UFSAR, Ref. 2) provide more restrictive limits to ensure that the SLs are not exceeded.
SAFETY LIMITS The Figure provided in the COLR shows the loci of points of Fraction of Rated Thermal power, RCS Pressure, and average temperature for which the minimum DNBR is not less than the safety analyses limit, that fuel centerline temperature remains below melting, that the average enthalpy in the hot leg is less than or equal to the enthalpy of saturated liquid, and that the exit quality is within the limits defined by the DNBR correlation.
The reactor core SLs are established to preclude violation of the following fuel design criteria:
- a. There must be at least 95% probability at a 95% confidence level (the 95/ 95 DNB criteria) that the hot fuel rod in the core does not experience DNB; and
- b. There must be at least a 95% probability at a 95%
confidence level that the hot fuel pellet in the core does not experience centerline fuel melting.
The reactor core SLs are used to define the various RPS functions such that the above criteria are satisfied during steady state operation, normal Revision No. 51 McGuire Units McGuire and 2 1 and Units 1 2 B 2.1.1-2 B 2.1.1-2 Revision No. 51
Reactor Core SLs B 2.1.1 BASES SAFETY LIMITS (Continued) operational transients, and anticipated operational occurrences (AOOs).
To ensure that he RPS precludes the violation of the above criteria, additional criteria are applied to the Over Temperature and Overpower AT reactor trip functions. That is, it must be demonstrated that the average enthalpy in the hot leg is less than or equal to the saturation enthalpy and that the core exit quality is within the limits defined by the DNBR correlation. Appropriate functioning of the RPS ensures that for variations in the THERMAL POWER, RCS Pressure, RCS average temperature, RCS flow rate, and Al that the reactor core SLs will be satisfied during steady state operation, normal operational transients, and AOOs.
APPLICABILITY SL 2.1.1 only applies in MODES 1 and 2 because these are the only MODES in which the reactor is critical. Automatic protection functions are required to be OPERABLE during MODES 1 and 2 to ensure operation within the reactor core SLs. The steam generator safety valves or automatic protection actions serve to prevent RCS heatup to the reactor core SL conditions or to initiate a reactor trip function, which forces the unit into MODE 3. Setpoints for the reactor trip functions are specified in LCO 3.3.1, "Reactor Trip System (RTS) Instrumentation." In MODES 3, 4, 5, and 6, Applicability is not required since the reactor is not generating significant THERMAL POWER.
SAFETY LIMIT If SL 2.1.1 is violated, the requirement to go to MODE 3 VIOLATIONS places the unit in a MODE in which this SL is not applicable.
The allowed 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 bringing the unit to a MODE of operation where this SL is not applicable, and reduces the probability of fuel damage.
REFERENCES 1. 10 CFR 50, Appendix A, GDC 10.
- 2. UFSAR, Section 7.2.
I McGuire Units 1 and 2 B 2.1.1-3 Revision No. 51
RTS Instrumentation 3.3.1 Table 3.3.1-1 (page 5 of 7)
Reactor Trip System Instrumentation Note 1: Overtemperature AT The Overtemperature AT Function Allowable Value shall not exceed the following NOMINAL TRIP SETPOINT by more than 4.3 % of RTP.
T(1+vS))(r T(+rTs) 1
)"ATO K1-K 2 ( +r (1+r 5 4 s)T s) L (1+rV s)
-T +KT(P-P')-f (A/)
Where: AT is measured RCS AT by loop narrow range RTDs, IF.
ATo is the indicated AT at RTP, OF.
s is the Laplace transform operator, sec-1.
T is the measured RCS average temperature, 'F.
T' is the nominal Tavg at RTP, < the value specified in the COLR.
P is the measured pressurizer pressure, psig P' is the nominal RCS operating pressure, = the value specified in the COLR.
K1 = Overtemperature AT reactor NOMINAL TRIP SETPOINT, as presented in the COLR, K2 = Overtemperature AT reactor trip heatup setpoint penalty coefficient, as presented in the COLR, K3 = Overtemperature AT reactor trip depressurization setpoint penalty coefficient, as presented in the COLR, T1 J2 = Time constants utilized in the lead-lag controller for AT, as presented in the COLR, T3 = Time constants utilized in the lag compensator for AT, as presented in the COLR,
¶4, T5 = Time constants utilized in the lead-lag controller for Tavg, as presented in the COLR, T6 = Time constants utilized in the measured Tavg lag compensator, as presented in the COLR, and, f1(Al) = a function of the indicated difference between top and bottom detectors of the power-range nuclear ion chambers; with gains to be selected based on measured instrument response during plant startup tests such that:
(i) for q, - qb between the "positive" and "negative" f 1(Al) breakpoints as presented in the COLR; f 1(AI) = 0, where q, and qb are percent RATED THERMAL POWER in the top and bottom halves of the core respectively, and qt + qb is total THERMAL POWER in percent of RATED THERMAL POWER; (continued)
McGuire Units 1 and 2 3.3.1-1 8 Amendment Nos.2 19 & 201
RTS Instrumentation 3.3.1 Table 3.3.1-1 (page 6 of 7)
Reactor Trip System Instrumentation (ii) for each percent imbalance that the magnitude of qt - qb is more negative than the f1(AI) "negative" breakpoint presented in the COLR, the AT Trip Setpoint shall be automatically reduced by the f1(AI) 'negative* slope presented in the COLR; and (iii) for each percent imbalance that the magnitude of qt - qb is more positive than the f 1(Al) "positive" breakpoint presented in the COLR, the AT Trip Setpoint shall be automatically reduced by the f1(Al)
"positive" slope presented in the COLR.
Note 2: Overpower AT The Overpower AT Function Allowable Value shall not exceed the following NOMINAL TRIP SETPOINT by more than 2.6% of RTP.
AT_(__+_, 1 J*ATO{K4 -K 5 1T7S(1+ JT-K 6 [T 1+r 6 s ] (
Where: AT is measured RCS AT by loop narrow range RTDs, OF.
ATo is the indicated AT at RTP, IF.
s is the Laplace transform operator, sec-1.
T is the measured RCS average temperature, 'F.
T is the nominal Tavg at RTP, < the value specified in the COLR.
K4 = Overpower AT reactor NOMINAL TRIP SETPOINT as presented in the COLR, K5 = The value specified in the COLR for increasing average temperature and the value specified in the COLR for decreasing average temperature, K6 = Overpower AT reactor trip heatup setpoint penalty coefficient as presented in the COLR for T > T and K6 = the value specified in the COLR for T < T.,
T1 , T2 = Time constants utilized in the lead-lag controller for AT, as presented in the COLR, T3 = Time constants utilized in the lag compensator for AT, as presented in the COLR, T6 = Time constants utilized in the measured Tavg lag compensator, as presented in the COLR, T7 = Time constant utilized in the rate-lag controller for Tavg, as presented in the COLR, and f2(AI) = a function of the indicated difference between top and bottom detectors of the power-range nuclear ion chambers; with gains to be selected based on measured instrument response during plant startup tests such that:
(continued)
McGuire Units 1 and 2 3.3.1 -19 Amendment Nos. 2 1 9 & 201
RTS Instrumentation B 3.3.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued) conditions and take corrective actions. Additionally, low temperature overpressure protection systems provide overpressure protection when below MODE 4.
- 9. Pressurizer Water Level-High The Pressurizer Water Level-High trip Function provides a backup signal for the Pressurizer Pressure-High trip and also provides protection against water relief through the pressurizer safety valves.
These valves are designed to pass steam in order to achieve their design energy removal rate. A reactor trip is actuated prior to the pressurizer becoming water solid. The setpoints are based on percent of instrument span. The LCO requires three channels of Pressurizer Water Level-High to be OPERABLE. The pressurizer level channels are used as input to the Pressurizer Level Control System. A fourth channel is not required to address control/protection interaction concerns. The level channels do not actuate the safety valves, and the high pressure reactor trip is set below the safety valve setting. Therefore, with the slow rate of charging available, pressure overshoot due to level channel failure cannot cause the safety valve to lift before reactor high pressure trip.
In MODE 1, when there is a potential for overfilling the pressurizer, the Pressurizer Water Level-High trip must be OPERABLE. This trip Function is automatically enabled on increasing power by the P-7 interlock. On decreasing power, this trip Function is automatically blocked below P-7. Below the P-7 setpoint, transients that could raise the pressurizer water level will be slow and the operator will have sufficient time to evaluate unit conditions and take corrective actions.
- 10. Reactor Coolant Flow-Low
- a. Reactor Coolant Flow-Low (Single Loop)
The Reactor Coolant Flow-Low (Single Loop) trip Function ensures that protection is provided against violating the DNBR limit due to low flow in one or more RCS loops, while avoiding reactor trips due to normal variations in loop flow.
Above the P-8 setpoint, which is approximately 48% RTP, a loss of flow in any RCS loop will actuate a reactor trip. The setpoints are based on the minimum flow specified in the McGuire Units 1 and 2 B 3.3.1-17 Revision No. 51
RTS Instrumentation B 3.3.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)
COLR. Each RCS loop has three flow detectors to monitor flow. The flow signals are not used for any control system input.
The LCO requires three Reactor Coolant Flow-Low channels per loop to be OPERABLE in MODE 1 above P-8.
In MODE 1 above the P-8 setpoint, a loss of flow in one RCS loop could result in DNB conditions in the core. In MODE 1 below the P-8 setpoint, a loss of flow in two or more loops is required to actuate a reactor trip (Function 1O.b) because of the lower power level and the greater margin to the design limit DNBR.
- b. Reactor Coolant Flow-Low (Two Loops)
The Reactor Coolant Flow-Low (Two Loops) trip Function ensures that protection is provided against violating the DNBR limit due to low flow in two or more RCS loops while avoiding reactor trips due to normal variations in loop flow.
Above the P-7 setpoint and below the P-8 setpoint, a loss of flow in two or more loops will initiate a reactor trip. The setpoints are based on the minimum flow specified in the COLR. Each loop has three flow detectors to monitor flow.
The flow signals are not used for any control system input.
The LCO requires three Reactor Coolant Flow-Low channels per loop to be OPERABLE.
In MODE 1 above the P-7 setpoint and below the P-8 setpoint, the Reactor Coolant Flow-Low (Two Loops) trip must be OPERABLE. Below the P-7 setpoint, all reactor trips on low flow are automatically blocked since power distributions that would cause a DNB concern at this low power level are unlikely. Above the P-7 setpoint, the reactor trip on low flow in two or more RCS loops is automatically enabled. Above the P-8 setpoint, a loss of flow in any one loop will actuate a reactor trip because of the higher power level and the reduced margin to the design limit DNBR.
McGuire Units 1 and 2 B 3.3.1-18 Revision No. 51
RCS Pressure, Temperature, and Flow DNB Limits 3.4.1 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.1 RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits LCO 3.4.1 RCS DNB parameters for pressurizer pressure, RCS average temperature, and RCS total flow rate shall be within the limits specified in Table 3.4.1-1.
APPLICABILITY: MODE 1.
NOTE----------------------------
Pressurizer pressure limit does not apply during:
- a. THERMAL POWER ramp > 5% RTP per minute; or
- b. THERMAL POWER step > 10% RTP.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Pressurizer pressure or A.1 Restore DNB parameter(s) 2 hours0.0833 days <br />0.0119 weeks <br />0.00274 months <br /> RCS average to within limit.
temperature DNB parameters not within limits.
B. RTS total flow rate > B.1 Reduce THERMAL 2 hours0.0833 days <br />0.0119 weeks <br />0.00274 months <br /> 99%, but < 100% of the POWER to < 98% RTP.
limit specified in the COLR. AND B.2 Reduce the Power Range 6 hours0.25 days <br />0.0357 weeks <br />0.00822 months <br /> Neutron Flux - High Trip Setpoint below the nominal setpoint by 2% RTP.
(continued)
McGuire Units 1 and 2 3.4.1 -1 Amendment Nos.2 19 & 201
RCS Pressure, Temperature, and Flow DNB Limits 3.4.1 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME C. RCS total flow rate C.1 Restore RCS total flow rate 2 hours0.0833 days <br />0.0119 weeks <br />0.00274 months <br />
< 99% of the value to > 99% of the value specified in the COLR. specified in the COLR.
OR 2 hours0.0833 days <br />0.0119 weeks <br />0.00274 months <br /> C.2.1 Reduce THERMAL POWER to < 50% RTP.
AND 6 hours0.25 days <br />0.0357 weeks <br />0.00822 months <br /> C.2.2 Reduce the Power Range Neutron Flux - High Trip Setpoint to < 55% RTP.
AND 24 hours1 days <br />0.143 weeks <br />0.0329 months <br /> C.2.3 Restore RCS total flow rate to > 99% of the value specified in the COLR.
D. Required Action and D.1 Be in MODE 2. 6 hours0.25 days <br />0.0357 weeks <br />0.00822 months <br /> associated Completion Time not met.
McGuire Units 1 and 2 3.4.1-2 Amendment Nos2 19 & 201
RCS Pressure, Temperature, and Flow DNB Limits 3.4.1 Table 3.4.1-1 (page 1 of 1)
RCS DNB Parameters PARAMETER INDICATION No. LIMITS OPERABLE CHANNELS
Average meter 3 <The limit specified in the COLR.
Temperature computer 4 < The limit specified in the COLR.
computer 3 < The limit specified in the COLR.
- 2. Indicated meter 4 > The limit specified in the COLR.
Pressurizer meter 3 > The limit specified in the COLR.
Pressure computer 4 > The limit specified in the COLR.
computer 3 > The limit specified in the COLR.
McGuire Units 1 and 2 3.4.1-4 Amendment Nos. 2 1 9 & 201
RCS Pressure, Temperature, and Flow DNB Limits B 3.4.1 BASES APPLICABLE SAFETY ANALYSES (continued) assessed for their impact on the acceptance criteria. A key assumption for the analysis of these events is that the core power distribution is within the limits of LCO 3.1.6, 'Control Bank Insertion Limits"; LCO 3.2.3, "AXIAL FLUX DIFFERENCE (AFD)"; and LCO 3.2.4, "QUADRANT POWER TILT RATIO (QPTR)."
The pressurizer pressure limits and the RCS average temperature limits specified in the COLR correspond to analytical limits used in the safety analyses, with allowance for measurement uncertainty.
The RCS DNB parameters satisfy Criterion 2 of 10 CFR 50.36 (Ref. 2).
LCO This LCO specifies limits on the monitored process variables-pressurizer pressure, RCS average temperature, and RCS total flow rate-to ensure the core operates within the limits assumed in the safety analyses. The numerical limits of these variables are contained in the COLR to provide operating and analysis flexibility from cycle to cycle. However, the minimum RCS flow, based on previously analyzed maximum steam generator tube plugging, is retained in the TS LCO. Operating within these limits will result in meeting the acceptance criteria, including the DNBR criterion.
RCS total flow rate contains a measurement error based on the performance of past precision heat balances and using the result to calibrate the RCS flow rate indicators. Sets of elbow tap coefficients, as determined during these heat balances, were averaged for each elbow tap to provide a single set of elbow tap coefficients for use in calculating RCS flow. This set of coefficients establishes the calibration of the RCS flow rate indicators and becomes the set of elbow tap coefficients used for RCS flow measurement.
Potential fouling of the feedwater venturi, which might not have been detected, could have biased the result from these past precision heat balances in a nonconservative manner. Therefore, a penalty for undetected fouling of the feedwater venturi raises the nominal flow measurement allowance for no fouling.
The numerical values for pressure and average temperature specified in the COLR are given for the measurement location with adjustments for the indication instruments.
APPLICABILITY In MODE 1, the limits on pressurizer pressure, RCS coolant average temperature, and RCS flow rate must be maintained during steady state operation in order to ensure DNBR criteria will be met in the event of an unplanned loss of forced coolant flow or other DNB limited transient. In McGuire Units 1 and 2 B 3.4.1-2 Revision No. 51
RCS Pressure, Temperature, and Flow DNB Limits B 3.4.1 BASES APPLICABILITY (continued) all other MODES, the power level is low enough that DNB is not a concern.
A Note has been added to indicate the limit on pressurizer pressure is not applicable during short term operational transients such as a THERMAL POWER ramp increase > 5% RTP per minute or a THERMAL POWER step increase > 10% RTP. These conditions represent short term perturbations where actions to control pressure variations might be counterproductive.
Also, since they represent transients initiated from power levels
< 100% RTP, an increased DNBR margin exists to offset the temporary pressure variations.
The DNBR Limit is provided in SL 2.1.1, "Reactor Core SLs." The conditions which define the DNBR Limit are less restrictive than the limits of this LCO, but violation of a Safety Limit (SL) merits a stricter, more severe Required Action. Should a violation of this LCO occur, the operator must check whether or not an SL may have been exceeded.
ACTIONS A.1 Pressurizer pressure and RCS average temperature are controllable and measurable parameters. With one or both of these parameters not within LCO limits, action must be taken to restore parameter(s).
The 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Completion Time for restoration of the parameters provides sufficient time to adjust plant parameters, to determine the cause for the off normal condition, and to restore the readings within limits, and is based on plant operating experience.
B.1 and B.2 RCS total flow rate is not a controllable parameter and is not expected to vary during steady state operation. If the indicated RCS total flow rate is >
99%, but < 100% of the limit specified in the COLR, then THERMAL POWER may not exceed 98% RTP. THERMAL POWER must be reduced within 2 hours0.0833 days <br />0.0119 weeks <br />0.00274 months <br />. The completion time of 2 hours0.0833 days <br />0.0119 weeks <br />0.00274 months <br /> is consistent with Required Action A.1. In addition, the Power Range Neutron Flux - High Trip Setpoint must be reduced from the nominal setpoint by 2% RTP within 6 hours0.25 days <br />0.0357 weeks <br />0.00822 months <br />. The Completion Time of 6 hours0.25 days <br />0.0357 weeks <br />0.00822 months <br /> to reset the trip setpoints recognizes that, with power reduced, the safety analysis assumptions are satisfied and there is no urgent need to reduce the trip setpoints. This is a sensitive operation that may inadvertently trip the Reactor Protection System.
McGuire Units 1 and 2 B 3.4.1-3 Revision No. 51
RCS Pressure, Temperature, and Flow DNB Limits B 3.4.1 BASES ACTIONS (continued)
C.1, C.2.1, C.2.2. and C.2.3 If the indicated RCS total flow rate is less than 99% of the valve specified in the COLR, then RCS total flow must be restored to greater than or equal to 99% of the value specified in the COLR within 2 hours0.0833 days <br />0.0119 weeks <br />0.00274 months <br /> or power must be reduced to less than 50% RTP. The Completion Time of 2 hours0.0833 days <br />0.0119 weeks <br />0.00274 months <br /> is consistent with Required Action A.1. If THERMAL POWER is reduced to less than 50% RTP, the Power Range Neutron Flux - High Trip Setpoint must also be reduced to
- 55% RTP. The Completion Time of 6 hours0.25 days <br />0.0357 weeks <br />0.00822 months <br /> to reset the trip setpoints is consistent with Required Action B.2. This is a sensitive operation that may inadvertently trip the Reactor Protection System. Operation is permitted to continue provided the RCS total flow is restored to greater than or equal to 99% of the value specified in the COLR within 24 hours1 days <br />0.143 weeks <br />0.0329 months <br />. The Completion Time of 24 hours1 days <br />0.143 weeks <br />0.0329 months <br /> is reasonable considering the increased margin to DNB at power levels below 50% and the fact that power increases associated with a transient are limited by the reduced trip setpoint.
D.1 If the Required Actions are not met within the associated Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 2 within 6 hours0.25 days <br />0.0357 weeks <br />0.00822 months <br />. The Completion Time of 6 hours0.25 days <br />0.0357 weeks <br />0.00822 months <br /> is reasonable to reach the required plant conditions in an orderly manner.
SURVEILLANCE SR 3.4.1.1 REQUIREMENTS This surveillance demonstrates that the pressurizer pressure remains within the required limits. Alarms and other indications are available to alert operators if this limit is approached or exceeded. The frequency of 12 hours0.5 days <br />0.0714 weeks <br />0.0164 months <br /> is sufficient, considering the other indications available to the operator in the control room for monitoring the RCS pressure and related equipment status.
The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> interval has been shown by operating practice to be sufficient to regularly assess for potential degradation and to verify operation is within safety analysis assumptions.
McGuire Units 1 and 2 B 3.4.1 -4 Revision No. 51
5.6 Reporting Requirements 5.6.2 Annual Radiological Environmental Operating Report (continued)
The Annual Radiological Environmental Operating Report shall include summarized and tabulated results of the analyses and measurements in the format of the table in the Radiological Assessment Branch Technical Position, Revision 1, November 1979. In the event that some individual results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted in a supplementary report as soon as possible.
5.6.3 Radioactive Effluent Release Report
NOTE--------------------------------------------
A single submittal may be made for a multiple unit station. The submittal should combine sections common to all units at the station; however, for units with separate radwaste systems, the submittal shall specify the releases of radioactive material from each unit.
The Radioactive Effluent Release Report covering the operation of the unit in the previous year shall be submitted prior to May 1 of each year in accordance with 10 CFR 50.36a. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit.
The material provided shall be consistent with the objectives outlined in Chapter 16 of the UFSAR and in conformance with 10 CFR 50.36a and 10 CFR Part 50, Appendix I, Section IV.B.1.
5.6.4 Monthly Operating Reports Routine reports of operating statistics and shutdown experience, including documentation of all challenges to the pressurizer power operated relief valves or pressurizer safety valves, shall be submitted on a monthly basis no later than the 15th of each month following the calendar month covered by the report.
5.6.5 CORE OPERATING LIMITS REPORT (COLR)
- a. Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:
- 1. Illustration of Reactor Core Safety Limits for Specification 2.1.1.
- 2. Moderator Temperature Coefficient BOL and EOL limits and 60 ppm and 300 ppm surveillance limits for Specification 3.1.3, (continued)
McGuire Units 1 and 2 5.6-2 Amendment Nos. 21 9 & 201
5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)
- 3. Shutdown Bank Insertion Limit for Specification 3.1.5, l
- 4. Control Bank Insertion Limits for Specification 3.1.6, l
- 5. Axial Flux Difference limits for Specification 3.2.3, l
- 6. Heat Flux Hot Channel Factor for Specification 3.2.1, l
- 7. Nuclear Enthalpy Rise Hot Channel Factor limits for Specification l 3.2.2,
- 8. Overtemperature and Overpower Delta T setpoint parameter l values for Specification 3.3.1,
- 9. Reactor Coolant System pressure, temperature, and flow departure from Nucleate Boiling (DNB) limits for Specification 3.4.1,
- 10. Accumulator and Refueling Water Storage Tank boron l concentration limits for Specification 3.5.1 and 3.5.4,
- 11. Reactor Coolant System and refueling canal boron concentration I limits for Specification 3.9.1,
- 12. Spent fuel pool boron concentration limits for Specification 3.7.14, l
- 13. SHUTDOWN MARGIN for Specification 3.1.1, and. I
- 14. 31 EFPD Surveillance Penalty Factors for Specifications 3.2.1 l and 3.2.2.
- b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:
- 1. WCAP-9272-P-A, "WESTINGHOUSE RELOAD SAFETY EVALUATION METHODOLOGY," (W Proprietary).
- 2. WCAP-1 0266-P-A, "THE 1981 VERSION OF WESTINGHOUSE EVALUATION MODEL USING BASH CODE," (W Proprietary).
- 3. BAW-1 01 68P-A, "B&W Loss-of-Coolant Accident Evaluation Model for Recirculating Steam Generator Plants," (B&W Proprietary).
(continued)
McGuire Units 1 and 2 5.6-3 Amendment Nos.21 9 & 201