ML052230403

Technical Specification Pages Extension of Reactor Trip System and Reactor Trip Device Functional Test Intervals

ML052230403
Person / Time
Site:	Three Mile Island
Issue date:	08/10/2005
From:	Office of Nuclear Reactor Regulation
To:
References
TAC MC4903
Download: ML052230403 (5)

Text

Bases (Cont'd)

The 600 ppmb limit in Item 4, Table 4.1-3 is used to meet the requirements of Section 5.4. Under other circumstances the minimum acceptable boron concentration would have been zero ppmb.

Calibration Calibration shall be performed to assure the presentation and acquisition of accurate information.

The nuclear flux (power range) channels amplifiers shall be checked and calibrated if necessary, every shift against a heat balance standard. The frequency of heat balance checks will assure that the difference between the out-of-core instrumentation and the heat balance remains less than 4%.

Channels subject only to "drifts errors induced within the instrumentation itself can tolerate longer intervals between calibrations. Process system instrumentation errors induced by drift can be expected to remain within acceptance tolerances if recalibration is performed at the intervals of each refueling period.

Substantial calibration shifts within a channel (essentially a channel failure) will be revealed during routine checking and testing procedures.

Thus, minimum calibration frequencies set forth are considered acceptable.

Testing On-line testing of reactor protection channels is required semi-annually on a rotational basis. The rotation scheme is designed to reduce the probability of an undetected failure existing within the system and to minimize the likelihood of the same systematic test errors being introduced into each redundant channel (Reference 1).

The rotation schedule for the reactor protection channels is as follows:

a) Deleted b) Semi-annually with one channel being tested every 46 days on a continuous sequential rotation.

The reactor protection system instrumentation test cycle is continued with one channel's instrumentation tested every 46 days. The frequency of every 46 days on a continuous sequential rotation is consistent with the calculations of Reference 2 that indicate the RPS retains a high level of reliability for this interval.

Upon detection of a failure that prevents trip action in a channel, the instrumentation associated with the protection parameter failure will be tested in the remaining channels. If actuation of a safety channel occurs, assurance will be required that actuation was within the limiting safety system setting.

The protection channels coincidence logic, the control rod drive trip breakers and the regulating control rod power SCRs electronic trips, are trip tested quarterly with one channel being tested every 23 days on a continuous sequential rotation. Calculations have shown that the frequency of every 23 days maintains a high level of reliability of the Reactor Trip System in Reference 4. The trip test checks all logic combinations and is to be performed on a rotational basis.

Discovery of a failure that prevents trip action requires the testing of the instrumentation associated with the protection parameter failure in the remaining channels.

For purposes of surveillance, reactor trip on loss of feedwater and reactor trip on turbine trip are considered reactor protection system channels.

4-2a Amendment No. 78,167,181, 200, 216, 255

Bases (Cont'd)

The equipment testing and system sampling frequencies specified in Tables 4.1-2, 4.1-3, and 4.1-5 are considered adequate to maintain the equipment and systems in a safe operational status.

REFERENCE (1) UFSAR, Section 7.1.2.3(d) - 'Periodic Testing and Reliability" (2) NRC SER for BAW-10167A, Supplement 1, December 5, 1988.

(3) BAW-10167, May 1986.

(4) BAW-101 67A, Supplement 3, February 1998.

4-2b Amendment No. 4184,226, 255

TABLE 4.1-1 INSTRUMENT SURVEILLANCE REQUIREMENTS CHANNEL DESCRIPTION CHECK TEST CALIBRATE REMARKS

1. Protection Channel NA Q NA I Coincidence Logic

2. Control Rod Drive Trip NA a NA (1) Includes independent testing of shunt Breaker and Regulating trip and undervoltage trip features. I Rod Power SCRs

3. Power Range Amplifier D(1) NA (2) (1) When reactor power is greater than 15%.

(2) When above 15% reactor power run a heat balance check once per shift. Heat balance calibration shall be performed whenever heat balance exceeds indicated neutron power by more than two percent.

4. Power Range Channel S S/A M(1)(2) (1) When reactor power is greater than 60% verify imbalance I using incore instrumentation.

(2) When above 15% reactor power calculate axial offset upper and lower chambers after each startup if not done within the previous seven days.

5. Intermediate Range Channel S(1) P S/U NA (1) When in service.

6. Source Range Channel S(1) P S/A NA (1) When in service.

7. Reactor Coolant Temperature S S/A F Channel I

TABLE 4.1-1 (Continued) z CD CHANNEL DESCRIPTION CHECK TEST CALIBRATE REMARKS

8. High Reactor Coolant S S/A R I I

Pressure Channel 0

,3 9. Low Reactor Coolant S S/A R I Pressure Channel

10. Flux-Reactor Coolant Flow S S/A F I Comparator

11. Reactor Coolant Pressure-Temperature S S/A R I Comparator

12. Pump Flux Comparator S S/A R I

13. High Reactor Building S S/A F I Pressure Channel

14. High Pressure Injection NA Q NA Logic Channels

15. High Pressure Injection Analog Channels

a. Reactor Coolant S(1) M R (1) When reactor coolant system is pressurized Pressure Channel above 300 psig or T .. is greater than 2000F

16. Low Pressure Injection NA Q NA Logic Channel

17. Low Pressure Injection 0 Analog Channels

a. Reactor Coolant S(1) M R (1) When reactor coolant system is pressurized Pressure Channel above 300 psig or T., 0 is greater than 200 0F

18. Reactor Building Emergency NA Q NA Cooling and Isolation System Logic Channel

TABLE 4.1-1 (Continued)

3 CHANNEL DESCRIPTION CHECK TEST CALIBRATE REMARKS

3 3B. OTSG Full Range Level W NA R 0 39. Turbine Overspeed Trip NA R NA

40. BWST/NaOH Differential NA NA F Pressure Indicator

41. Sodium Hydroxide Tank NA NA F Level Indicator

-J 42. Diesel Generator NA NA R Protective Relaying

43. 4 KV ES Bus Undervoltage Relays (Diesel Start)

a. Degraded Grid NA M(1) A (1) Relay operation will be checked by local test pushbuttons.

b. Loss of Voltage NA M(1) R (1) Relay operation will be checked by local test pushbuttons.

44. Reactor Coolant Pressure S(1) M R (1) When reactor coolant system is DH Valve Interlock Bistable pressurized above 300 psig or Tae is greater than 200 0F.

45. Loss of Feedwater Reactor Trip S(1) S/A(1) R (1) When reactor power exceeds 7% I power.

46. Turbine Trip/Reactor Trip S(1) S/A(1) F (1) When reactor power exceeds 45% I power.

47. a. Pressurizer Code Safety Valve S(1) NA F (1) When Tave is greater than 5250 F.

and PORV Tailpipe Flow Monitors

b. PORV - Acoustic/Flow NA M(1) R (1) When Tave Is greater than 525 0F.

48. PORV Setpolnts NA M(1) R (1) Per Specification 3.1.12 excluding valve operation.