ML14183A135
| ML14183A135 | |
| Person / Time | |
|---|---|
| Site: | Robinson  |
| Issue date: | 01/09/1989 |
| From: | Adensam E Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML14183A136 | List: |
| References | |
| DPR-23-A-121 NUDOCS 8901240062 | |
| Download: ML14183A135 (12) | |
Text
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RUNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 CAROLINA POWER & LIGHT COMPANY DOCKET NO. 50-261 H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 121 License No. DPR-23
- 1. The Nuclear Regulatory Commission (the Commission) has found that:
A. The application for amendment by Carolina Power & Light Company (the licensee), dated July 26, 1988, as supplemented August 26, and November 1, 1988, 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 riot 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 arid all applicable requirements have been satisfied.
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- 2. Accordingly, the license is amended by changes to the Technical Specifications, as indicated in the attachment to this license amendment; and paragraph 3.8 of Facility Operating License No. DPR-23 is hereby amended to read as follows:
- 3. This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Elinor G. Adensam, Director Project Directorate II-1 Division of Reactor Projects I/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: January 9, 1989 OFC :LA:PD'1 R PM:PD21:D R:CSB
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OFFICIAL RECORD COPY
ATTACHMENT TO LICENSE AMENDMENT NO. 121 FACILITY OPERATING LICENSE NO. DPR-23 DOCKET NO. 50-261 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages. The revised areas are indicated by marginal lines.
Remove Pages Insert Pages 2.1-3 2.1-3 2.3-2 2.3-2 2.3-4 2.3-4 2.3-5 2.3-5 2.3-6 2.3-6 2.3-6a 3.1-11 3.1-11 4.1-3 4.1-3 4.1-5 4.1-5
(HBR-27) are set to preclude bulk boiling at the vessel exit.
An arbitrary upper safety limit of 118% thermal power is shown.
This limit is based on the high flux trip including all uncertainties.
Radial power peaking factors consistent with the limit on FAH given in Specification 3.10.2.1 have been employed in the generation of the curves in Figure 2.1-1.
An additional heat flux factor of 1.03 has been included to account for fuel manufacturing tolerances and in-reactor densification of the fuel.
The safety limit curves given in Figure 2.1-1 are based on a minimum RCS flow of,97.3 x 106 lbm/hr. These curves would not be applicable in the case of a loss of flow transient. The evaluation of such an event would be based upon the analysis presented in Section 15.3 of the FSAR. The minimum RCS flow is 99.8 x 106 lbm/hr, which is the minimum thermal design flow of 97.3 x 106 Ibm/hr with a 2.6% allowance added for instrument uncertainty associated with the precision calorimetric flow measurement.(3)
The Reactor Control and Protection System is designed to prevent any anticipated combination of transient conditions for Reactor Coolant System temperature, pressure, and thermal power level that would result in a DNB ratio of less than 1.17(2) based on steady state nominal operating power levels less than or equal to 100%, steady state moninal operating Reactor Coolant System average temperatures less than or equal to 575.4*F, and a steady state nominal operating pressure of 2235 psig. Allowances are made in initial conditions assumed for transient analyses for steady state errors of
+2% in power, +4*F in Reactor Coolant System average temperature, and +/-30 psi in pressure. The combined steady state errors result in the DNB ratio at the start of a transient being 10 percent less than the value at nominal full power operating conditions.
Reference (1) XN-NF-711(P) Rev. 0, "XNB Addendum for 26 Inch Spacer."
(2) FSAR Section 15.
(3) WCAP-11889, "RTD Bypass Elimination Licensing Report for H. B. Robinson, Unit 2" 2.1-3 Amendment No. 87, 98 121
(HBR-50) where:
ATo
= Indicated AT at rated thermal power, OF; T
= Average temperature, OF; P
= Pressurizer pressure, psig; K1
< 1.1365; K2
= 0.01228; K3
= 0.00089; 1 +
S 1 + T S
= The function generated by the lead-lag controller for T 2
dynamic compensation;
&l 2
= Time constants utilized in the lead-lag controller for Tavgv r1 = 20 seconds, T2 = 3 seconds; T'
= 575.4*F Reference Tavg at rated thermal power; P1
= 2235 psig (Nominal RCS Operating Pressure);
S
= Laplace transform operator, sec ;
and f(AI) is 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 start-up tests such that:
- 1)
For (q. -
qb) within +12% and -17%, where qt and qb are percent power in the top and bottom halves of the core, respectively, and q
+ qb is total core power in percent of rated power (2300 Mwt), f(AI) = 0. For every 2.4% below rated power (2300 Mwt) level, permissible positive flux difference range is extended by +1 percent.
For every 2.4% below rated power (2300 Mwt) level, the permissible negative flux difference range is extended by -1 percent.
- 2)
For each percent that the magnitude of (qt - q ) exceeds +12% in a positive direction, the AT trip setpoint sha 1 be automatically reduced by 2.4% of the value of AT at rated power (2300 Mwt).
2.3-2 Amendment No.
07, 121
(HBR-27) 2.3.2 Protective instrumentation settings for reactor trip interlocks shall be as follows:
2.3.2.1 The low pressurizer pressure trip, high pressurizer Level trip, and the low reactor coolant flow trip (for two or more loops) may be bypassed below 10% of rated power.
2.3.2.2 The single-loop-loss-of-flow trip may be bypassed below 45% of rated power.
2.3.3 The RCS narrow range temperature sensors response time shall be less than or equal to a 4.0 second lag time constant.
Basis The power range reactor trip low setpoint provides protection in the power range for a power excursion beginning from lower power. This trip value was used in the safety analysis.(1)
In the power range of operation, the overpower nuclear flux reactor trip protects the reactor core against reactivity excursions which are too rapid to be protected by temperature and pressure-protective circuitry. The prescribed set point, with allowance for errors, is consistent with the trip point assumed in the accident analysis.(2)
The source and intermediate range reactor trips do not appear in the specification, as these settings are not used in the transient and accident analysis (FSAR Section 15).
Both trips provide protection during reactor startup. The former is set at about 10+5 counts/sec and the latter at a current proportional to approximately 25% of full power.
The high and low pressure reactor trips limit the pressure range in which reactor operation is permitted. The high pressurizer pressure reactor trip is also a backup to the pressurizer code safety valves for overpressure protection, and is therefore set lower than the set pressure for these valves (2485 psig).
The low pressurizer pressure reactor trip also trips the reactor in the unlikely event of a loss-of-coolant accident.(3) 2.3-4 Amendment No. 87, 121
(HBR-50)
The overtemperature AT reactor trip provides core protection against DNB for all combinations of pressure, power, coolant temperature, and axial power distribution provided only that (1) the transient is slow with respect to transport to and response time of the temperature detectors (about 4 seconds), and (2) pressure is within the range between the high and Low pressure reactor trips.
With normal a i power distribution, the reactor trip limit, with allowance for errors, is always below the core safety limit as shown in Figure 2.1-1.
If axial peaks are greater than design, as indicated by the difference between top and bottom power range nuclear detectors, the reactor trip is automatically reduced according to Specification 2.3.1.2.d.
The overpower AT reactor trip prevents power density anywhere in the core from exceeding 118% of design power density as discussed in Section 7.2.2 of the FSAR and includes corrections for axial power distribution, change in density, and heat capacity of water with temperature, and dynamic compensation for piping delays from the core to the loop temperature detectors.
The specified setpoin meet this requirement and include allowance for instrument errors.
The setpoints in the Technical Specifications ensure the combination of power, temperature, and pressure will not exceed the core safety Limits as shown in Figure 2.1-1.
The low flow reactor trip protects the core against DNB in the event of a sudden loss of power to one or more reactor coolant pumps.
The setpot n specified is consistent with the value used in the accident analysis. 5 The undervoltage and underfrequency reactor trips protect against a decrease in flow caused by low electrical voltage or frequency. The specified setpoints assure a reactor trip signal before the low flow trip point is reached.
The high pressurizer water level reactor trip prote ts the pressurizer safety valves against water relief. Approximately 1150 ft of water correspon9to 92% of span.
The specified setpoint allows margin for instrument error and transient level overshoot beyond this trip setting so that the trip function prevents the water level from reaching the safety valves.
2.3-5 Amendment No.
0/7, 121
(HBR-27)
The low-low steam generator water level reactor trip protects against loss of feedwater flow accidents. The specified set point assures that there will be sufficient water inventory in the steam generators at the time of trip to allow for starting delays for the auxiliary feedwater system.(6)
The specified reactor trips are blocked at Low power where they are not required for protection and would otherwise interfere with normal plant operations. The prescribed set point above which these trips are unblocked assures their reliability in the power range where needed.
Above 10% power, an automatic reactor trip will occur if two reactor coolant pumps are lost during operation. Above 45% power, an automatic reactor trip will occur if any pump is Lost.
This latter trip will prevent the minimum value of the DNB ratio, DNBR, from going below 1.17 during normal operational transients and anticipated transients when only two loops are in operation and the overtemperature AT trip setpoint is adjusted to the value specified for three loop operation.
The turbine and steam-feedwater flow mismatch trips do not appear in the specification, as these settings are not used in the transient and accident analysis.
(FSAR Section 15)
The RCS temperature measurement response time parameters define the time delay between when the OTAT reactor trip conditions are reached and when the control rods are released and free to fall and is based on a sensor lag of 4.0 seconds for the narrow range temperature measurement with a 0.75 second electromechanical delay.(7)(8)(9)
References (1) FSAR Section 15.4 (2) FSAR Section 15.0 (3) FSAR Section 15.6 (4) Deleted (5) FSAR Section 15.3 (6) FSAR Section 15.2 (7) FSAR Section 7.2.2.2.2 2.3-6 Amendment No. $7, 121
(HBR-27)
(8) WCAP-11889, "RTD Bypass Elimination Licensing Report for H. B. Robinson, Unit 2" (9) ANF-88-094, "H. B. Robinson, Unit 2, Chapter 15, OTAT Trip Event Analysis for Elimination of RTD Bypass Piping" 2.3-6a Amendment No. 121
(HBR-50) 3.1.3 Minimum Conditions for Criticality 3.1.3.1 Except during low power physics tests, the reactor shall not be made critical at any temperature, above which the moderator temperature coefficient is greater than:
a)
+5.0 pcm/*F at less than 50% of rated power, or b) 0 pcm/*F at 50% of rated power and above.
3.1.3.2 In no case shall the reactor be made critical above and to the left of the criticality Limit shown on Figure 3.1-la or 3.1-lb (as appropriate per 3.1.2.1).
3.1.3.3 When the reactor coolant temperature is in a range where the moderator temperature coefficient is greater than as specified in 3.1.3.1 above, the reactor shall be subcritical by an amount equal to or greater than the potential reactivity insertion due to depressurization.
3.1.3.4 The reactor shall be maintained subcritical by at least 1% until normal water level is established in the pressurizer.
Basis During the early part of fuel cycle, the moderator temperature coefficient may be slightly positive at low power levels. The moderator coefficient at low temperatures or powers will be most positive at the beginning of the fuel cycle, when the boron concentration in the coolant is the greatest. At all times, the moderator coefficient is calculated to be negative in the high power operating range, and after a very brief period of power operation, the coefficient will be negative in all circumstances due to the reduced boron concentration as Xenon and fission products build into the core. The requirement that the reactor is not to be made critical when the moderator coefficient is more positive than as specified in 3.1.3.1 above has been imposed to prevent any unexpected power excursion during normal operations as a result of either an increase of moderator temperature or decrease of coolant 3.1-11 Amendment No. 87. 773.
121
(HBR-27)
Calibration Calibration is performed to ensure the presentation and acquisition of accurate information.
The nuclear flux (linear level) channels daily calibration against a thermal power calculation will account for errors induced by changing rod patterns and core physics parameters.
For RCS narrow range temperature sensors, verification of response time will be a part of calibration. Cross calibration of RCS narrow range temperature sensors will be performed on a refueling interval.
Other channels are subject only to the "drift" errors induced within the instrumentation itself and, consequently, can tolerate longer intervals between calibration. Process system instrumentation errors induced by drift can be expected to remain within acceptable tolerances if recalibration is performed at intervals of each refueling shutdown.
Substantial calibration shifts within a channel (essentially a channel failure) will be revealed during routine checking and testing procedures.
Thus, the minimum calibration frequencies set forth are considered acceptable.
Testing(1)
Minimum testing frequency is based on evaluation of unsafe failure rate data and reliability analysis. This is based on operating experience at conventional and nuclear plants. An "unsafe failure" is defined as one which negates channel operability and which, due to its nature, is revealed only when the channel is tested or attempts to respond to a bona fide signal.
The minimum testing frequency for those instrument channels connected to the safety system is based on an average unsafe failure rate of 2.5 x 10-6 failure/hr per channel.
4.1-3 Amendment No. 82 121
(HBR-27)
TABLE 4.1-1 MINIMUM FREQUENCIES FOR CHECKS, CALIBRATIONS AND TEST OF INSTRUMENT CHANNELS Channel Description Check Calibrate Test Remarks
- 1. Nuclear Power Range S
D (1)
B/W (2)
(1)
Thermal Power calculations M* (3) during power operations R* (3)
(2)
Signal to AT; bistable action (permissive, rod stop, trips)
(3)
Upper and lower chambers for symmetric offset:
monthly during power opera tions. When periods of re actor shutdown extend this interval beyond one month, the calibration shall be performed immediately fol lowing return to power.
- 2. Nuclear Intermediate Range S (1)
N.A.
S/U (2)
(1)
Once/shift when in service (2)
Log level; bistable action (permissive, rod stop, trip)
- 3. Nuclear Source Range S (1)
N.A.
S/U (2)
(1)
Once/shift when in service (2)
Bistable action (alarm, trip)
- 4. Reactor Coolant Temperature S
R (4)
B/W (1)(2)
(1)
Overtemperature -
AT rD (2)
Overpower - AT R
(3)
(3)
Narrow range RTD response time (4)
To include narrow range RTD cross calibration a
- 5. Reactor Coolant Flow S
R M
- 6. Pressurizer Water Level S
R K
- 7.
Pressurizer Pressure S
R M
- 8. 4 Kv Voltage N.A R
M Reactor Protection circuits only
- By means of the movable in-core detector system