ML033510806
| ML033510806 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 12/10/2003 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| Download: ML033510806 (67) | |
Text
INDEX SAFFTY I TMITC ANn I TMTTNA AFFTY Y;TFM FTTTNnS SECTION PAGE 2.1 SAFETY LIMITS 2.1.1 REACTOR CORE .... . . . . . . . . . . . . . . . . . . . . . . 2-1 2.1.2 REACTOR COOLANT SYSTEM PRESSURE . . . . . . . . . . . . . . . . . 2-1 FIGURE 2.1-1 REACTOR CORE SAFETY LIMIT ..... ...... . . . . . 2-2 FIGURE 2.1-2 DELETED . . . . . . . . . . . . . . . . . . . . . . . . . 2-3 2.2 LIMITING SAFETY SYSTEM SETTINGS 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS . . . . . . . . . . 2-4 TABLE 2.2-1 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS . . . . 2-5 RASFS SECTION PAGE 2.1 SAFETY LIMITS 2.1.1 REACTOR CORE ......... .. ... ... .. .. ... . B 2-1 2.1.2 REACTOR COOLANT SYSTEM PRESSURE . . . . . . . . . . . . . . . . B 2-2 2.2 LIMITING SAFETY SYSTEM SETTINGS 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS . . . . . . . . . B 2-3 MILLSTONE - UNIT 3 . .
Amendment No. 217 0959
INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.0 APPLICABILITY . . . . . . . . . . . . . . . . . . . . . . . 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL Shutdown Margin - MODES 1 AND 2 . . . . . . . . . . . . 3/4 1-1 Shutdown Margin - MODES 3, 4, AND 5 LOOPS FILLED . . . 3/4 1-3 FIGURE 3.1-1 REQUIRED SHUTDOWN MARGIN FOR MODE 3 . . . . . . . . 3/4 1-4 I FIGURE 3.1-2 DELETED.. .. 3/4 1-5 I FIGURE 3.1-3 REQUIRED SHUTDOWN MARGIN FOR MODE 4 . . . . . . . . 3/4 1-6 FIGURE 3.1-4 REQUIRED SHUTDOWN MARGIN FOR MODE 5 WITH RCS LOOPS FILLED . . . . . . . . . . . . . . . 3/4 1-7 Shutdown Margin - Cold Shutdown -
Loops Not Filled ...................
3/4 1-8 FIGURE 3.1-') REQUIRED SHUTDOWN MARGIN FOR MODE 5 W ITH RCS LOOPS DRAINED . . . . . . . . . . . . . . . 3/4 1-9 Moderator Temperature Coefficient . . . . . . . . . 3/4 1-10 Minimum Temperature for Criticality . . . . . . . . 3/4 1-12 3/4.1.2 BORATION SYSTEMS DELETED . . . . . . . . . . . . . . . . . . . . . . . . 3/4 1-13 DELETED . . . . . . . . . . . . . . . . . . . . . . . . 3/4 1-14 DELETED . . . . . . . . . . . . . . . . . . . . . . . . 3/4 1-15 DELETED . . . . . . . . . . . . . . . . . . . . . . . . 3/4 1-16 DELETED . . . . . . . . . . . . . . . . . . . . . . . . 3/4 1-17 DELETED . . . . . . . . . . . . ... . . . . . . . . . . 3/4 1-18 3/4.1.3 MOVABLE CONTROL ASSEMBLIES Group Height . . . . . . . . . . . . . . *
- 3/4 1-20 TABLE 3.1-1 ACCIDENT ANALYSES REQUIRING REEVALUATION IN THE EVENT OF AN INOPERABLE FULL-LENGTH ROD . . . . . . . 3/4 1-22 Position Indication Systems - Operating . . . . . . . 3/4 1-23 MILLSTONE - UNIT 3 0959 iv Amendment No. , fp, ii, 797, 217
INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE DELETED .................................................. 3/ 1-24 4
Rod Drop Time ............................................ 3/4 1-25 Shutdown Rod Insertion Limit ............................. 3/4 1-26 Control Rod Insertion Limits ............................. 3/4 1-27 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE .................................... 3/4 2-1 3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR - F(Z)..................... 3/4 2-5 3/4.2.3 RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR ................................................... 3/4 2-19 3/4.2.4 QUADRANT POWER TILT RATIO ................... 3/4 2-24 3/4.2.5 DNB PARAMETERS ........................ 3/4 2-27 TABLE 3.2-1 DNB PARAMETERS ........................................ 3/4 2-28 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION . ............. 3/4 3-1 TABLE 3.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION ................... 3/4 3-2 TABLE 3.3-2 DELETED TABLE 4.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS ............................................. 3/4 3-10 3/4.3.2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION .......................................... 3/4 3-15 TABLE 3.3-3 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION .......................................... 3/4 3-17 TABLE 3.3-4 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS ........................... 3/4 3-26 MILLSTONE - UNIT 3 v Amendment No. y0, Po, gy, i, 0959 707, 217
INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE TABLE 3.3-13 DELETED TABLE 4.3-9 DELETED 3/4.3.4 DELETED 3/4.3.5 SHUTDOWN MARGIN MONITOR . . . . . . . . . . . . .... . . .. 3/4 3-82 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION Startup and Power Operation . . . . . . . . . ... . . . ... 3/4 4-1 Hot Standby ................. ... . . .... 3/4 4-2 Hot Shutdown . . . . . . . . . . . . . . . . ... . . .... 3/4 4-3 Cold Shutdown - Loops Filled . ... . . . . . ... . . .... 3/4 4-5 Cold Shutdown - Loops Not Filled . . . . . . ... . . . ... 3/4 4-6 Loop Stop Valves . . . . . . . . . . . . . . .....
. . . . . .. 3/4 4-7 I Isolated Loop Startup .................. . 3/4 4-8 3/4.4.2 SAFETY VALVES. . 3/4 4-9 DELETED . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 4-10 3/4.4.3 PRESSURIZER Startup and Power Operation . . . . . . . . . . . . . . . . 3/4 4-11 FIGURE 3.4-5 PRESSURIZER LEVEL CONTROL . . . . . . . . . . . . . . . 3/4 4-11a Hot Standby ....................... .3/4 4-lib 3/4.4.4 RELIEF VALVES. .3/4 4-12 3/4.4.5 STEAM GENERATORS .3/4 4-14 TABLE 4.4-1 MINIMUM NUMBER OF STEAM GENERATORS TO BE INSPECTED DURING INSERVICE INSPECTION . . . . . . . . . . . . . . . .3/4 4-19 TABLE 4.4-2 STEAM GENERATOR TUBE INSPECTION . . . . . . . . . . . *3/4 4-20 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems ................ .3/4 4-21 Operational Leakage ................... .3/4 4-22 TABLE 3.4-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES . . .3/4 4-24 3/4.4.7 DELETED. .3/4 4-25 TABLE 3.4-2 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS . . . . . . . *3/4 4-26 TABLE 4.4-3 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS SURVEILLANCE REQUIREMENTS . . . . . . . . . . . . . . . . . . . . . . .3/4 4-27 3/4.4.8 SPECIFIC ACTIVITY .................... *3/4 4-28 MILLSTONE - UNIT 3 vii Amendment No. Xfq, 0960 XnY, I7, 09,
INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE FIGURE 3.4-1 DOSE EQUIVALENT I-131 REACTOR COOLANT SPECIFIC ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE REACTOR COOLANT SPECIFIC ACTIVITY >Ci/gram DOSE EQUIVALENT I-131 ................ 3/4 4-30 TABLE 4.4-4 REACTOR COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM . . . . . . . . . . . . . . . . . . . . . . . 3/4 4-31 3/4.4.9 PRESSURE/TEMPERATURE LIMITS . . . . . . . . . . . . . 3/4 4-33 FIGURE 3.4-2 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS -
APPLICABLE UP TO 10 EFPY . . . . . . . . . . . . . . . 3/4 4-34 FIGURE 3.4-3 REACTOR COOLANT SYSTEM COOLDOWN LIMITATIONS -
APPLICABLE UP TO 10 EFPY . . . . . . . . . . . . . . . 3/4 4-35 TABLE 4.4-5 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM -
WITHDRAWAL SCHEDULE ................. 3/4 4-36 Pressurizer . . . . . . . . . . . . . . . . . . . . . 3/4 4-37 Overpressure Protection Systems . . . . . . . . . . . 3/4 4-38 FIGURE 3.4-4a High Setpoint PORV Curve For the Cold Overpressure Protection System .................. 3/4 4-40 FIGURE 3.4-4b Low Setpoint PORV Curve For the Cold Overpressure Protection System ................ 3/4 4-41 I 3/4.4.10 DELETED 3/4 4-42 3/4 .4.11 DELETED . . . . . . . . . . . . . . . . . . . . . 3/4 4-43 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS . . . . . . . . . . . . . . . . . . . . 3/4 5-1 3/4.5.2 ECCS SUBSYSTEMS - Tavg GREATER THAN OR EQUAL TO 350°F 3/4 5-3 3/4.5.3 ECCS SUBSYSTEMS - Tavg LESS THAN 350*F . . . . . . . 3/4 5-7 3/4.5.4 REFUELING WATER STORAGE TANK . . . . . . . . . . . . 3/4 5-9 3/4.5.5 pH TRISODIUM PHOSPHATE STORAGE BASKETS . . . . . . . . . 3/4 5-10 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT Containment Integrity . . . . . . . . . . . . . . . 3/4 6-1 Containment Leakage . . . . . . . . . . . . . . . . 3/4 6-2 Containment Air Locks . . . . . . . . . . . . . . . 3/4 6-5 Containment Pressure . . . . . . . . . . . . . . . . 3/4 6-7 MILLSTONE - UNIT 3 - viii Amendment No. d, F7, , 17d, 0960 70f, 217
INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE Air Temperature . . . . . . . . . . . . . . . . . . . . 3/4 6-9 Containment Structural Integrity . . . . . . . . . . . . 3/4 6-10 Containment Ventilation System . . . . . . . . . . . . . 3/4 6-11 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS Containment Quench Spray System . . . . . . . . . . . . 3/4 6-12 Recirculation Spray System . . . . . . . . . . . . . . . 3/4 6-13 3/4.6.3 CONTAINMENT ISOLATION VALVES . . . . . . . . . . . . . . 3/4 6-15 3/4.6.4 COMBUSTIBLE GAS CONTROL Hydrogen Monitors . . . . . . . . . . . . . . . . . . . 3/4 6-16 Electric Hydrogen Recombiners . . . . . . . . . . . . . 3/4 6-17 3/4.6.5 SUBATMOSPHERIC PRESSURE CONTROL SYSTEM Steam Jet Air Ejector . . . . . . . . . . . . . . . . . 3/4 6-18 3/4.6.6 SECONDARY CONTAINMENT Supplementary Leak Collection and Release System . . . . 3/4 6-19 Secondary Containment . . . . . . . . . . . . . . . . . 3/4 6-22 Secondary Containment Structural Integrity . . . . . . . . . . . . . . . . . . 3/4 6-23 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE Safety Valves 3/4 7-1 TABLE 3.7-1 MAXIMUM ALLOWABLE POWER RANGE NEUTRON FLUX HIGH SETPOINT WITH INOPERABLE STEAM LINE SAFETY VALVES . . . 3/4 7-2 1 TABLE 3.7-2 DELETED . . . . . . . . . . . . . . . . 3/4 7-2 MILLSTONE - UNIT 3 ix Amendment No. By, 7, R7, Py, II0, 0960
. 77P 217}. XM,
INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.9.6 REFUELING MACHINE . . . . . . . . . . . . . . . . . . . . 3/4 9-6 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE AREAS . . . . . . . . . 3/4 9-7 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION High Water Level .................. . . 3/4 9-8 Low Water Level ................... . . 3/4 9-9 3/4.9.9 CONTAINMENT PURGE AND EXHAUST ISOLATION SYSTEM . . . *
- 3/4 9-10 3/4.9.10 WATER LEVEL - REACTOR VESSEL . . . . . . . . . . . . . . 3/4 9-11 3/4.9.11 WATER LEVEL - STORAGE POOL . . . . . . . . . . . . . . . 3/4 9-12 3/4.9.12 FUEL BUILDING EXHAUST FILTER SYSTEM . . . . . . . . . . . 3/4 9-13 3/4.9.13 SPENT FUEL POOL - REACTIVITY . . . . . . . . . . . . *
- 3/4 9-16 3/4.9.14 SPENT FUEL POOL - STORAGE PATTERN . . . . . . . . . . . . 3/4 9-17 FIGURE 3.9-1 MINIMUM FUEL ASSEMBLY BURNUP VERSUS NOMINAL INITIAL ENRICHMENT FOR REGION 1 4-OUT-OF-4 STORAGE CONFIGURATION . . * *3/4 9-18 FIGURE 3.9-2 REGION I 3-OUT-OF-4 STORAGE FUEL ASSEMBLY LOADING SCHEMATIC . . . . . . . . . . . . . . . . .3/4 9-19 FIGURE 3.9-3 MINIMUM FUEL ASSEMBLY BURNUP VERSUS NOMINAL INITIAL ENRICHMENT FOR REGION 2 STORAGE CONFIGURATION . . . . .3/4 9-20 FIGURE 3.9-4 MINIMUM FUEL ASSEMBLY BURNUP AND DECAY TIME VERSUS NOMINAL INITIAL ENRICHMENT FOR REGION 3 STORAGE CONFIGURATION . . . . . . . . . . . . . . . . . . . . .3/4 9-21 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOWN MARGIN .................... 3/4 10-1 3/4.10.2 GROUP HEIGHT, INSERTION, AND POWER DISTRIBUTION LIMITS 3/4 10-2 I 3/4.10.3 PHYSICS TESTS . . . . . . . . . . . . . . . . . . . . . 3/4 10-4 3/4.10.4 REACTOR COOLANT LOOPS .... 3/4 10-5 3/4.10.5 POSITION INDICATION SYSTEM - SHUTDOWN . . . . . . . . . 3/4 10-6 3/4.11 DELETED 3/4.11.1 DELETED 3/4.11.2 DELETED 3/4.11.3 DELETED MILLSTONE - UNIT 3 xii Amendment 7g, py, Ipp, ;py, 0961 7-7,217
2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS REACTOR CORE 2.1.1 The combination of THERMAL POWER, pressurizer pressure, and the highest operating loop coolant temperature (Tavg) shall not exceed the limits shown in Figure 2.1-1. I APPLICABILITY: MODES 1 and 2. ACTION: Whenever the point defined by the combination of the highest operating loop average temperature and THERMAL POWER has exceeded the appropriate pressurizer pressure line, be in HOT STANDBY within 1 hour. REACTOR COOLANT SYSTEM PRESSURE 2.1.2 The Reactor Coolant System pressure shall not exceed 2750 psia. APPLICABILITY: MODES 1, 2, 3, 4, and 5. ACTION: MODES 1 and 2: Whenever the Reactor Coolant System pressure has exceeded 2750 psia be in HOT STANDBY with the Reactor Coolant System pressure within its limit within 1 hour. MODES 3, 4 and 5: Whenever the Reactor Coolant System pressure has exceeded 2750 psia, reduce the Reactor Coolant System pressure to within its limit within 5 minutes. MILLSTONE - UNIT 3 2-1 Amendment No. J7,217 0962
680 660 LL 640 0 w
-< 620
%r-) 600 580 560 0 0.2 0.4 0.8 1.0 1.2 FRACTION OF RATED THERMAL POWER FIGURE 2.1-1 REACTOR CORE SAFETY LIMIT I MILLSTONE - UNIT 3 2-2 Amendment No. #P,217 0962
This page intentionally left blank. MILLSTONE - UNIT 3 2-3 Amendment No. PO,217 0962
TABLE 2.2-1 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS NOMINAL FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUE
- 1. Manual Reactor Trip N.A. N.A.
- 2. Power Range, Neutron Flux
- a. High Setpoint 109% of RTP** < 109.6% of RTP** I
- b. Low Setpoint 25% of RTP** < 25.6% of RTP**
- 3. Power Range, Neutron Flux, 5% of RTP** with < 5.6% of RTP** with High Positive Rate a time constant a time constant
> 2 seconds > 2 seconds
- 4. Deleted
- 5. Intermediate Range, 25% of RTP** < 27.4% of RTP**
Neutron Flux
- 6. Source Range, Neutron Flux 1 X 10+ 5 cps < 1.06 x 10+5 cps
- 7. Overtemperature AT See Note 1 See Note 2 I
- RTP = RATED THERMAL POWER
TABLE 2.2-1 (Continued) REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS NOMINAL FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUE I
- 8. Overpower AT See Note 3 See Note 4 I
- 9. Pressurizer Pressure-Low 1900 psia > 1897.6 psia
- 10. Pressurizer Pressure-High 2385 psia < 2387.4 psia
- 11. Pressurizer Water Level-High 89% of instrument < 89.3% of instrument span span
- 12. Reactor Coolant Flow-Low 90% of loop > 89.8% of loop design flow* design flow*
- 13. Steam Generator Water 18.1% of narrow > 17.8% of narrow Level Low-Low range instrument range instrument span span
- 14. General Warning Alarm N.A. N.A.
- 15. Low Shaft Speed - Reactor 92.4% of rated > 92.2% of rated Coolant Pumps speed speed
- Minimum Measured Flow Per Loop = 1/4 of the RCS Flow Rate Limit as listed in Section 3.2.3.1.a I
TABLE 2.2-1 (Continued) REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS NOMINAL FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUE
- 16. Turbine Trip
- a. Low Fluid Oil Pressure 500 psig > 450 psig
- b. Turbine Stop Valve 1% open > 1% open Closure
- 17. Safety Injection Input N.A. N.A.
from ESF
- 18. Reactor Trip System Interlocks
- a. Intermediate Range 1 x 10.10 amp > 9.0 x 1O" amp Neutron Flux, P-6
- b. Low Power Reactor Trips Block, P-7
- 1) P-10 input (Note 5) 11% of RTP** < 11.6% of RTP**
- 2) P-13 input 10% RTP** Turbine < 10.6% RTP** Turbine Impulse Pressure Impulse Pressure Equivalent Equivalent
- c. Power Range Neutron 37.5% of RTP** < 38.1% of RTP**
Flux, P-8
- RTP = RATED THERMAL POWER
o __3 (0 TABLE 2.2-1 (Continued) O, A C-, REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS 0 MI NOMINAL FUNCTIONAL UNIT TRIP SETPOINT Al.LOWABLE VALUE
-I C,
- d. Power Range Neutroln 51% of RTP** < 51.6% of RTP**
Flux, P-9
- e. Power Range Neutroin 9% of RTP** -> 8.4% of RTP**
Flux, P-10 (Note 6
- 19. Reactor Trip Breakers N.A. N.A.
- 20. Automatic Trip and Interlock N.A. N.A.
Logic 0,
- 21. DELETED I 3
3 0.
=3 CD 0 -I Im **RTP = RATED THERMAL POWER
TABLE 2.2-1 (Continued) TABLE NOTATIONS NOTE 1: OVERTEMPERATURE AT ( AT ) ( 1+.1S) K1 - 4 (2' - T) -K3 (P -P') - fi (AI) AO(1+.r2 s) 2 +rs T Where: AT is measured Reactor Coolant System AT, OF; ATo is loop specific indicated AT at RATED THERMAL POWER, F; (1+'rls) (1+2S) is the function generated by the lead-lag compensator on measured AT; r, and r2 are the time constants utilized in the lead-lag compensator for AT, r1 > 8 sec. r 2 < 3 sec; K1 < 1.20 K2 > 0.02456/F; (1+X 4 s) (1+. s) is the function generated by the lead-lag compensator for Tavg; T4 and are the time constants utilized in the lead-lag compensator for Tavg 74 > 20 sec, 5 < 4 sec; T is measured Reactor Coolant System average temperature, F; T' is loop specific indicated Tavg at RATED THERMAL POWER, < 587.1OF; K3 > 0.001311/psi P is measured pressurizer pressure, psia; P' is nominal pressurizer pressure, > 2250 psia; s is the Laplace transform operator, sec';
3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1.1 BORATION CONTROL SHUTDOWN MARGIN - MODES 1 AND 2 LIMITING CONDITION FOR OPERATION 3.1.1.1.1 The SHUTDOWN MARGIN shall be greater than or equal to 1.3% k/k. APPLICABILITY: MODES I and 2*. ACTION: With the SHUTDOWN MARGIN less than 1.3% Ak/k, immediately initiate and con-tinue boration at greater than or equal to 33 gpm of a solution containing greater than or equal to 6600 ppm boron or equivalent until the required SHUTDOWN MARGIN is restored. SURVEILLANCE REQUIREMENTS 4.1.1.1.1 The SHUTDOWN MARGIN shall be determined to be greater than or equal to 1.3% Ak/k:
- a. Within 1 hour after detection of an inoperable control rod(s) and at least once per 12 hours thereafter while the rod(s) is inoperable.
If the inoperable control rod is immovable or untrippable, the above required SHUTDOWN MARGIN shall be verified acceptable with an increased allowance for the withdrawn worth of the immovable or untrippable control rod(s);
- b. When in MODE 1 or MODE 2 with Kff greater than or equal to 1 at least once per 12 hours by verifying that control bank withdrawal is within the limits of Specification 3.1.3.6;
- c. When in MODE 2 with Keff less than 1, within 4 hours prior to achieving reactor criticality by verifying that the predicted critical control rod position is within the limits of Specification 3.1.3.6;
- d. Prior to initial operation above 5% RATED THERMAL POWER after each fuel loading, by consideration of the factors of Specification 4.1.1.1.2, with the control banks at the maximum insertion limit of Specification 3.1.3.6; and
- See Special Test Exceptions Specification 3.10.1.
MILLSTONE - UNIT 3 3/4 1-1 Amendment No. fo, 17,217 0967
3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL SHUTDOWN MARGIN - MODES 3. 4 AND 5 LOOPS FILLED LIMITING CONDITION FOR OPERATION 3.1.1.1.2 The SHUTDOWN MARGIN shall be greater than or equal to the limits shown in Figures 3.1-1, 3.1-3 and 3.1-4.* APPLICABILITY: MODES 3, 4 and 5 ACTION: With the SHUTDOWN MARGIN less than the required value, immediately initiate and continue boration at greater than or equal to 33 gpm of a solution containing greater than or equal to 6600 ppm boron or equivalent until the required SHUTDOWN MARGIN is restored. SURVEILLANCE REQUIREMENTS 4.1.1.1.2.1 The SHUTDOWN MARGIN shall be determined to be greater than or equal to the required value:
- a. Within 1 hour after detection of an inoperable control rod(s) and at least once per 12 hours thereafter while the rod(s) is inoperable.
If the inoperable control rod is immovable or untrippable, the above required SHUTDOWN MARGIN shall be verified acceptable with an increased allowance for the withdrawn worth of the immovable or untrippable control rod(s); and
- b. At least once per 24 hours by consideration of the following factors:
- 1) Reactor Coolant System boron concentration,
- 2) Control rod position,
- 3) Reactor Coolant System average temperature,
- 4) Fuel burnup based on gross thermal energy generation,
- 5) Xenon concentration, and
- 6) Samarium concentration.
4.1.1.1.2.2 Valve 3CHS-V305 shall be verified closed and locked at least once per 31 days.
- Additional SHUTDOWN MARGIN requirements, if required, are given in Specification 3.3.5.
MILLSTONE - UNIT 3 0968 3/4 1-3 Amendment No. 9, }X}, Jf9,217
o-am cor r, q Y C(A C-4 - I a re (2050,3.495) 2 3.5 I/ (2500,3.495) d 0 3 ~~4.---., _ ..., _ z ~ ~ ~ ~ 4 I ~ - - 0 2.5 cc
<2 "olo z
¢ 1.5 (0,l .3XO) (7X,1l.30().
a 0; 1 -. I . . U= 0.5 0 2,000 2,500 0 50X 1,000 '1 ,500 RCS CRITICAL BORON CONCENTRATION (ppm) FIGURE 3.1-1 REQUIRED SHUTDOWN FOR MODE 3 I
This page intentionally left blank. MILLSTONE - UNIT 3 3/4 1-5 Amendment No. gY, y0, 0968
- 90. 7. 194. 2.1 7
REACTIVITY CONTROL SYSTEMS LIMITING CONDITION FOR OPERATION ACTION (Continued) c) A power distribution map is obtained from the movable incore detectors and F(Z) and FN6H are verified to be within their limits within 72 hours; and d) THERMAL POWER level is reduced to less than or equal to 75% of RATED THERMAL POWER within the next hour and within the following 4 hours the High Neutron Flux Trip Setpoint is reduced to less than or equal to 85% of RATED THERMAL POWER.
- c. With more than one rod trippable but inoperable due to causes other than addressed by ACTION a. above, POWER OPERATION may continue provided that:
- 1. Within 1 hour, the remainder of the rods in the bank(s) with the inoperable rods are aligned to within +12 steps of the inoperable rods while maintaining the rod sequence and insertion limits of Specification 3.1.3.6. The THERMAL POWER level shall be restricted pursuant to Specification 3.1.3.6 during subsequent operation, and
- 2. The inoperable rods are restored to OPERABLE status within 72 hours.
- d. With more than one rod misaligned from its group step counter demand height by more than +12 steps (indicated position), be in HOT STANDBY within 6 hours.
SURVEILLANCE REQUIREMENTS 4.1.3.1.1 The position of each full-length rod shall be determined to be within the group demand limit by verifying the individual rod positions at least once per 12 hours except during time intervals when the rod position deviation monitor is inoperable, then verify the group positions at least once per 4 hours. 4.1.3.1.2 Each full-length rod not fully inserted in the core shall be determined to be OPERABLE by movement of at least 10 steps in any one direction at least once per 92 days. MILLSTONE - UNIT 3 3/4 1-21 MAmendment No. P, pp, X97, 217 0969
REACTIVITY CONTROL SYSTEMS POSITION INDICATION SYSTEMS - OPERATING I IMuTINs CNfTIAN FR PFRATTQN 3.1.3.2 The Digital Rod Position Indication System and the Demand Position Indication System shall be OPERABLE and capable of determining the control rod positions within +/-12 steps. APPLICABILITY: MODES I and 2. ACTION:
- a. With a maximum of one digital rod position indicator per bank inoperable:
- 1. Determine the position of the nonindicating rod(s) indirectly by the movable incore detectors at least once per 8 hours and immediately after any motion of the nonindicating rod which exceeds 24 steps in one direction since the last determination of the rod's position, or
- 2. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 8 hours.
- b. With a maximum of one demand position indicator per bank inoperable:
- 1. Verify that all digital rod position indicators for the affected bank are OPERABLE and that the most withdrawn rod and the least withdrawn rod of the bank are within a maximum of 12 steps of each other at least once per 8 hours, or
- 2. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 8 hours.
SURVEILLANCE REQUIREMENTS 4.1.3.2.1 Each digital rod position indicator shall be determined to be OPERABLE by verifying that the Demand Position Indication System and the Digital Rod Position Indication System agree within 12 steps at least once per 12 hours except during time intervals when the rod position deviation monitor is inoperable, then compare the Demand Position Indication System and the Digital Rod Position Indication System at least once per 4 hours. 4.1.3.2.2 Each of the above required digital rod position indicator(s) shall be determinded to be OPERABLE by verifying that the digital rod position indicators agree with the demand position indicators within 12 steps when exercised over the full-range of rod travel at least once per 24 months. MILLSTONE - UNIT 3 3/4 1-23 Amendment No. 0, , g, 0970 707,217
3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE I 1 MITING pAnNnlTiflN F R PFRAT N 3.2.1.1 The indicated AXIAL FLUX DIFFERENCE (AFD) shall be maintained within:
- a. The limits specified in the CORE OPERATING LIMITS REPORT (COLR) for Relaxed Axial Offset Control (RAOC) operation, or
- b. Within the target band about the target flux difference during base load operation, specified in the COLR.
APPLICABILITY: MODE I above 50% RATED THERMAL POWER*. ACTION:
- a. For RAOC operation with the indicated AFD outside of the applicable limits specified in the COLR,
- 1. Either restore the indicated AFD to within the COLR specified limits within 15 minutes, or
- 2. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 30 minutes and reduce the Power Range Neutron Flux--
High Trip setpoints to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours.
- b. For base load operation above APLND with the indicted AFD outside of the applicable target band about the target flux differences:
- 1. Either restore the indicated AFD to within the COLR specified target band within 15 minutes, or
- 2. Reduce THERMAL POWER to less than APLND of RATED THERMAL POWER and discontinue base load operation within 30 minutes.
- c. THERMAL POWER shall not be increased above 50% of RATED THERMAL POWER unless the indicated AFD is within the limits specified in the COLR.
- See Special Test Exception 3.10.2 MILLSTONE - UNIT 3 3/4 2-1 Amendment No. "j, P0,217 0971
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POWER DISTRIBUTION LIMITS 3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR - FA(Z) LIMITING CONDITION FOR OPERATION I 3.2.2.1 F(Z) shall be limited by the following relationships: FRTP FO(Z) < F2--- K(Z) for P > 0.5 P F RTP FO(Z) < F.RT- K(Z) for P < 0.5 0.5 FRTP = the F limit at RATED THERMAL POWER (RTP) provided in the core operating limits report (COLR). THERMAL POWER Where: P = RATED THERMAL POWER K(Z) = the normalized F(Z) as a function of core height specified in the COLR. APPLICABILITY: MODE 1. ACTION: With F(Z) exceeding its limit:
- a. For RAOC operation with Specification 4.2.2.1.2.b not being satisfied or for base load operation with Specification 4.2.2.1.4.b not being satisfied:
(1) Reduce THERMAL POWER at least 1%. for each 1% FQ(Z) exceeds the limit within 15 minutes and similarly reduce the Power Range Neutron Flux-High Trip Setpoints within the next 4 hours; POWER OPERATION may proceed for up to a total of 72 hours; subsequent POWER OPERATION may proceed provided the Overpower AT Trip setpoints have been reduced at least 1% for each 1% FQ(Z) exceeds the limit, and MILLSTONE - UNIT 3 3/4 2-5 Amendment No. 9, Po, Yp, J77, 0972 J70,217
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POWER DISTRIBUTION LIMITS 3/4.2.3 RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR LIMITING CONDITION FOR OPERATION 3.2.3.1 The indicated Reactor Coolant System (RCS) total flow rate and FH shall be maintained as follows:
- a. RCS total flow rate > 371,920 gpm, and
- b. FNH < FRT [3.0 + PFAH (1.0 - P)]
Where:
- 1) P THERMIAL POWER RATED THERMAL POWER
- 2) Fa,, = Measured values of FH obtained by using the movable incore detectors to obtain a power distribution map. The measured value of F should be used since Specification 3.2.3.1b.
takes into consideration a measurement uncertainty of 4% for incore measurement,
- 3) FLTP = The FN limit at RATED THERMAL POWER in the CORE OPERATING LIMITS REPORT (COLR),
- 4) PFAH - The power factor multiplier for F provided in the COLR, and
- 5) The measured value of RCS total flow rate shall be used since uncertainties of 2.4% for flow measurement have been included in Specification 3.2.3.1a.
APPLICABILITY: MODE 1. ACTION: With the RCS total flow rate or F H outside the region of acceptable operation:
- a. Within 2 hours either:
- 1. Restore the RCS total flow rate and FH to within the above limits, or
- 2. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER and reduce the Power Range Neutron Flux - High Trip Setpoint to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours.
MILLSTONE - UNIT 3 3/4 2-19 Amendment No. 7, XP, P9, 717, 217 0973
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TABLE 3.2-1 DNB PARAMETERS PARAMETER LIMITS Indicated Reactor Coolant System Tvg < 591.1*F Indicated Pressurizer Pressure > 2218 psia*
- Limit not applicable during either a THERMAL POWER ramp in excess of 5% of RATED THERMAL POWER per minute or a THERMAL POWER step in excess of 10% of RATED THERMAL POWER.
TABLE 3.3-1 I-3 0r REACTOR TRIP SYSTEM INSTRUMENTATION 4 MINIMUM zrm TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION C-I
- 1. Manual Reactor Trip 2 1 2 1, 2 1 z
(a 2 1 2 3*, 4*, 5* 11
- 2. Power Range, Neutron Flux
- a. High Setpoint 4 2 3 1, 2 2
- b. Low Setpoint 4 2 3 1###, 2 2
- 3. Power Range, Neutron Flux 4 2 3 1, 2 2 High Positive Rate
- 4. Deleted
- 5. Intermediate Range, Neutron Flux 2 1 2 1###, 2 3 v 6. Source Range, Neutron Flux W a. Startup 2 1 2 2## 4 r6 b. Shutdown 2 1 2 3*, 4*, 5* 11
- 7. Overtemperature AT 4 2 3 1, 2 6 I
- 8. Overpower AT 4 2 3 1, 2 6 I
¢ 9. Pressurizer Pressure--Low 4 2 3 1** 6 (1)
- 3. 10. Pressurizer Pressure--High 4 2 3 1, 2 6 (1) a
(. 11. Pressurizer Water Level--High 3 2 2 1** 6
~eR
TABLE 3.3-1 (Continued) o3
- m rACTnV ro-_ur - TDTD rIer.
QVYTIM
- lull I TNTDIIMrNTATTnm
~llIUI'II;ItI^
- I I I-(A MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE C FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION
-I
- 12. Reactor Coolant Flow--Low La
- a. Single Loop (Above P-8) 3/loop 2/loop 2/loo0p 1 6 1
- b. Two Loops (Above P-7 and 3/loop 2/loop in 2/l oop 1 6 1 below P-8) two oper-ating loops
- 13. Steam Generator Water 4/stm. gen. 2/stm. gen. 3/stm. gen. 1, 2 6 (1)
Level--Low-Low I
- 14. Low Shaft Speed--Reactor I Coolant Pumps 4-1/pump 2 3 1** 6 W
(A)
- 15. Turbine Trip La Ia
- a. Low Fluid Oil Pressure 3 2 2 1*** 12
- b. Turbine Stop Valve Closure 4 4 4 1*** 6
- 16. Deleted
- 17. Reactor Trip System Interlocks
- a. Intermediate Range Neutron Flux, P-6 2 1 2 8 ED 0.
- b. Low Power Reactor Trips Block, P-7 P-10 Input 4 2 3 1 8 0.
t or P-13 Input 2 1 2 1 8 IQ
o 3-1 TABLE 3.3-1 (Continued)
-4 r-r- RFACT TP
\l.rrv
.. .. . l
. vim l #\s s
... SYSTEM INSTRUMENTATION 0
m MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE C FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION 1-4
-4 (A) 17. Reactor Trip System Interlocks (Continued)
- c. Power Range Neutron Flux, P-8 4 3 1 8
- d. Power Range Neutron 4 2 3 1 8 Flux, P-9
- e. Power Range Neutron Flux, P-10 4 2 3 1,2 8 W
- 6) 18. Reactor Trip Breakers(2 ) 2 1 2 1, 2 10, 13 2 1 2 3*, 4*, 5* 11
- 19. Automatic Trip and Interlock 2 1 2 1, 2 13A Logic 2 1 2 3*, 4*, 5* 11
- 20. DELETED I 3
- 21. DELETED CD
=
to I..
TABLE 3.3-1 (Continued) TABLE NOTATIONS
*When the Reactor Trip System breakers are in the closed position and the Control Rod Drive System is capable of rod withdrawal. **Above the P-7 (At Power) Setpoint.
- Above the P-9 (Reactor Trip/Turbine Trip Interlock) Setpoint.
##Below the P-6 (Intermediate Range Neutron Flux Interlock) Setpoint.
- Below the P-10 (Low Setpoint Power Range Neutron Flux Interlock) Setpoint.
(1) The applicable MODES and ACTION statements for these channels noted in Table 3.3-3 are more restrictive and, therefore, applicable. (2) Including any reactor trip bypass breakers that are racked in and closed for bypassing a reactor trip breaker. ACTION STATEMENTS ACTION 1 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours or be in HOT STANDBY within the next 6 hours. ACTION 2 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:
- a. The inoperable channel is placed in the tripped condition within 6 hours,
- b. The Minimum Channels OPERABLE requirement is met; however, the inoperable channel may be bypassed for up to 4 hours for surveillance testing of other channels per Specification 4.3.1.1, and
- c. Either, THERMAL POWER is restricted to less than or equal to 75% of RATED THERMAL POWER and the Power Range Neutron I Flux Trip Setpoint is reduced to less than or equal to 85%
of RATED THERMAL POWER within 4 hours; or, the QUADRANT I POWER TILT RATIO is monitored at least once per 12 hours per Specification 4.2.4.2. MILLSTONE - UNIT 3 3/4 3-5 Amendment No. 7, p, i, 0976 9, 217
TABLE 4.3-1 (Continued) REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS TRIP ANALOG ACTUATING MODES FOR CHANNEL DEVICE WHICH CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST IS REQUIRED
- 18. Reactor Trip Breaker N.A. N.A. N.A. M(7, 11) N.A. I1) 2* 3*p 4~
- 19. Automatic Trip and N.A. N.A. N.A. N.A. M(7) 1, 2, 3*,
Interlock Logic 4*, 5*
- 20. DELETED I
- 21. Reactor Trip Bypass N.A. N.A. N.A. MR7 15) N.A. 1* 2t*3*,
Breaker R 16)
- 22. DELETED
TABLE 3.3-3 (Continued) ENGINFERFD SAFFTY FEATURES ACTUATTON SYSTEM NSTRIMNTATTON MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION
- 5. Turbine Trip and Feedwater Isolation
- a. Automatic Actuation 2 1 2 1, 2 25 Logic and Actuaion Relays
- b. Steam Generator 4/stm. gen. 2/stm. gen. 3/stm. gen. 1, 2, 3 20, 21 Water Level -- in each in any oper- in each High-High (P-14) operating ating loop operating loop loop
- c. Safety Injection 2 1 2 1, 2 22 Actuation Logic
- d. Te Low Coincident 1 Tave/l OoP 1 Tve in I TaVe in 1, 2 20 with P-4 any two any three loops loops
TABLE 3.3-4 o 'Os-
- ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS Ir (A -4i NOMINAL 0
M FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUE
- 5. Turbine Trip and Feedwater Isolation
- a. Automatic Actuation Logic N.A. N.A.
Actuation Relays
- b. Steam Generator Water 80.5% of narrow < 80.8% of narrow Level--High-High (P-14) range instrument range instrument span. span.
- c. Safety Injection Actuation See Item 1. above for all Safety Injection Trip Logic Setpoints and Allowable Values.
- d. Tave Low Coincident with 5640F > 563.6*F Reactor Trip (P-4) I
- 6. Auxiliary Feedwater
- a. Manual Initiation N.A. N.A.
- b. Automatic Actuation Logic N.A. N.A.
and Actuation Relays
- c. Steam Generator Water Level--Low-Low
- 1) Start Motor-Driven 18.1% of > 17.8% of narrow Pumps narrow range range instrument span.
instrument span.
INSTRUMENTATION 3/4.3.5 SHUTDOWN MARGIN MONITOR LIMITING CONDITION FOR OPERATION 3.3.5 Two channels of Shutdown Margin Monitors shall be OPERABLE
- a. With a minimum count rate as designated in the CORE OPERATING LIMITS REPORT (COLR), or
- b. If the minimum count rate in Specification 3.3.5.a cannot be met, then the Shutdown Margin Monitors may be made operable with a lower minimum count rate, as specified in the COLR, by borating the Reactor Coolant System above the requirements of Specification 3.1.1.1.2 or 3.1.1.2. The additional boration shall be:
- 1. A minimum of 150 ppm above the SHUTDOWN MARGIN require-ments of Figure 3.1-1 (Mode 3), or
- 2. A minimum of 350 ppm above the SHUTDOWN MARGIN require-ments of Figure 3.1-3 (Mode 4), Figure 3.1-4 (Mode 5 - RCS loops filled) and Figure 3.1-5 (Mode 5 - RCS loops drained).
APPLICABILITY: MODES 3*, 4, and 5. ACTION:
- a. With one Shutdown Margin Monitor inoperable, restore the inoperable channel to OPERABLE status within 48 hours.
- b. With both Shutdown Margin Monitors inoperable or one Shutdown Margin Monitor inoperable for greater than 48 hours, immediately suspend all operations involving positive reactivity changes via dilution and rod withdrawal. Verify the valves listed in Specifica-tion 4.1.1.2.2 are closed and secured in position within the next 4 hours and at least once per 14 days thereafter.** Verify comp-liance with the SHUTDOWN MARGIN requirements of Specification 3.1.1.1.2 or 3.1.1.2, as applicable, within 1 hour and at least once per 12 hours thereafter.
- The shutdown margin monitors may be blocked during reactor startup in accordance with approved plant procedures.
- The valves may be opened on an intermittent basis under administrative controls as noted in Surveillance 4.1.1.2.2.
MILLSTONE - UNIT 3 3/4 3-82 Amendment No. 7, 217 nqrn
3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION STARTUP AND POWER OPERATION I I MTTTNA cnNnTTTAN FR flPFRATTAN 3.4.1.1 Four reactor coolant loops shall be OPERABLE and in operation. I I APPLICABILITY: MODES 1 and 2.* ACTION: With less than the above required reactor coolant loops in operation, be in at least HOT STANDBY within 6 hours. SIIRVFTI I ANF RFQUTRFMFNTS 4.4.1.1 The above required reactor coolant loops shall be verified in operation and circulating reactor coolant at least once per 12 hours.
- See Special Test Exceptions Specification 3.10.4.
MILLSTONE - UNIT 3 .3/4 4-1 Amendment No.217 0981
REACTOR COOLANT SYSTEM LOOP STOP VALVES I I MTTING CONDITION FR PFRATTON 3.4.1.5 Each RCS loop stop valve shall be open and the power removed from the valve operator. I APPLICABILITY: MODES 1, 2, 3 and 4. ACTION:
- a. With power available to one or more loop stop valve operators, remove power from the loop stop valve operators within 30 minutes or be in HOT STANDBY within the next 6 hours and COLD SHUTDOWN within the following 30 hours.
b.IU With one or more RCS loop stop valves closed, maintain the valve(s) closed and be in HOT STANDBY within 6 hours and COLD SHUTDOWN within the next 30 hours. i SURVEILLANCE REQUIREMENTS 4.4.1.5 Verify each RCS loop stop valve is open and the power removed from the valve operator at least once per 31 days. I All required actions of Action Statement 3.4.1.5.b shall be completed whenever this action is entered. MILLSTONE - UNIT 3 3/4 4-7 Amendment N. 217 0982
3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE SAFETY VALVES I TMTTTNR CnNnTTTnN FR PFRATTION 3.7.1.1 All main steam line Code safety valves shall be OPERABLE with lift settings as specified in Table 3.7-3. APPLICABILITY: MODES 1, 2, and 3. ACTION:
- a. With one or more main steam line Code safety valves inoperable, operation in MODES , 2, and 3 may proceed provided, that within 4 hours, either the inoperable valve is restored to OPERABLE status or the Power Range Neutron Flux High Trip Setpoint is reduced per Table 3.7-1; otherwise, be in at least HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours.
1IIRVFTI I ANCE RFQIITRFMFNTN 4.7.1.1 No additional Surveillance Requirements other than those required by Specification 4.0.5. MILLSTONE - UNIT 3 .3/4 7-1 Amendment No. 7. 217 0983
TABLE 3.7-1 MAXIMUM ALLOWABLE POWER RANGE NEUTRON FLUX HIGH SETPOINT WITH INOPERABLE STEAM LINE SAFETY VALVES I MAXIMUM NUMBER OF INOPERABLE MAXIMUM ALLOWABLE POWER RANGE SAFETY VALVES ON ANY NEUTRON FLUX HIGH SETPOINT OPERATING STEAM GENERATOR (PERCENT OF RATED THERMAL POWER) 65 2 46 3 28 TABLE 3.7-2 DELETED I MILLSTONE - UNIT 3 3/4 7-2 Amendment No. 17,217 0983
SPECIAL TEST EXCEPTIONS 3/4.10.2 GROUP HEIGHT, INSERTION, AND POWER DISTRIBUTION LIMITS I TMTTTNn roNnTTTON FP PFPATTQN 3.10.2.1 The group height, insertion, and power distribution limits of Specifications 3.1.3.1, 3.1.3.5, 3.1.3.6, 3.2.1.1, and 3.2.4 may be suspended during the performance of PHYSICS TESTS provided:
- a. The THERMAL POWER is maintained less than or equal to 85% of RATED THERMAL POWER, and
- b. The limits of Specifications 3.2.2.1 and 3.2.3.1 are maintained and determined at the frequencies specified in Specification 4.10.2.1.2 below.
APPLICABILITY: MODE 1. ACTION: With any of the limits of Specification 3.2.2.1 or 3.2.3.1 being exceeded while the requirements of Specifications 3.1.3.1, 3.1.3.5, 3.1.3.6, 3.2.1.1, and 3.2.4 are suspended, either:
- a. Reduce THERMAL POWER sufficient to satisfy the ACTION requirements of Specifications 3.2.2.1 and 3.2.3.1, or
- b. Be in HOT STANDBY within 6 hours.
CllRVF1I I ANUF RFQITPFMFNTS 4.10.2.1.1 The THERMAL POWER shall be determined to be less than or equal to 85% of RATED THERMAL POWER at least once per hour during PHYSICS TESTS. 4.10.2.1.2 The Surveillance Requirements- of the below listed specifications shall be performed at least once per 12 hours during PHYSICS TESTS:
.a. Specifications 4.2.2.1.2 and 4.2.2.1.3, and
- b. Specification 4.2.3.1.2.
MILLSTONE - UNIT 3 3/4 10-2 Amendment No. 217 0984
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2.1 SAFETY LIMITS 2.1.1 REACTOR CORE The restrictions of this Safety Limit prevent overheating of the fuel and possible cladding perforation which would result in the release of fission products to the reactor coolant. Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature. Operation above the upper boundary of the nucleate boiling regime could result in excessive cladding temperatures because of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coefficient. DNB is not a directly measurable parameter during operation and therefore THERMAL POWER and reactor coolant temperature and pressure have been related to DNB. This relation has been developed to predict the DNB flux and the location of DNB for axially uniform and nonuniform heat flux distributions. The local DNB heat flux ratio (DNBR) is defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux and is indicative of the margin to DNB. The DNB design basis is as follows: uncertainties in the WRB-1 or WRB-2 correlations, plant operating parameters, nuclear and thermal parameters, fuel fabrication parameters, and computer codes are considered statistically such that there is at least a 95 percent probability with 95 percent confidence level that DNBR will not occur on the most limiting fuel rod during Condition I and II events. This establishes a design DNBR value which must be met in plant safety analyses using values of input parameters without uncertainties. In addition, margin has been maintained in the design by meeting safety analysis DNBR limits in performing safety analyses. The curves of Figure 2.1-1 show the loci of points of THERMAL POWER, Reactor Coolant System pressure, and average temperature below which the calculated DNBR is no less than the design DNBR value or the average enthalpy at the vessel exit is less than the enthalpy of saturated liquid. These curves are based on an enthalpy hot channel factor, F NH , of 1.70 (includes measurement uncertainty) and a reference cosine with a eak of .55 for axial power shape. An allowance is included for an increase in FA H at reduced power based on the expression: F NH = 1.70 [1 + 0.3 (-P)] where P is the fraction of RATED THERMAL POWER These limiting heat flux conditions are higher than those calculated for the range of all control rods fully withdrawn to the maximum allowable control rod insertion assuming axial imbalance is within the limits of F (delta I) function of the Overtemperature trip. When the axial power imbalance is not within the tolerance, the axial power imbalance effect on the Overtemperature delta T trips will reduce the setpoints to provide protection consistent with core safety limits. MILLSTONE - UNIT 3 B 2-1 Amendment No. Po, 217 0964
2.2 LIMITING SAFETY SYSTEM SETTINGS BASES REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS (Continued) functional capability at the specified trip setting is required for those anticipatory or diverse reactor trips for which no direct credit was assumed in the safety analysis to enhance the overall reliability of the Reactor Trip System. The Reactor Trip System initiates a turbine trip signal whenever reactor trip is initiated. This prevents the reactivity insertion that would otherwise result from excessive Reactor Coolant System cooldown and thus avoids unnecessary actuation of the Engineered Safety Features Actuation System. Manual Reactor Trip The Reactor Trip System includes manual Reactor trip capability. Power Range, Neutron Flux In each of the Power Range Neutron Flux channels there are two independent bistables, each with its own trip setting used for a High and Low Range trip setting. The Low Setpoint trip provides protection during subcritical and low power operations to mitigate the consequences of a power excursion beginning from low power, and the High Setpoint trip provides protection during power operations to mitigate the consequences of a reactivity excursion from all power levels. The Low Setpoint trip may be manually blocked above P-10 (a power level of approximately 10% of RATED THERMAL POWER) and is automatically reinstated below the P-10 Setpoint. Power Range, Neutron Flux, High Positive Rate The Power Range Positive Rate trip provides protection against rapid flux increases which are characteristic of a rupture of a control rod drive housing. Specifically, this trip complements the Power Range Neutron Flux High and Low trips to ensure that the criteria are met for all rod ejection accidents. MILLSTONE - UNIT 3 B 2 - 4 Amendment No. 77P, 1M, 217 0965
LIMITING SAFETY SYSTEM SETTINGS BASES Intermediate and Source Range, Neutron Flux The Intermediate and Source Range, Neutron Flux trips provide core protection during reactor startup to mitigate the consequences of an uncon-trolled rod cluster control assembly bank withdrawal from a subcritical condition. These trips provide redundant protection to the Low Setpoint trip of the Power Range, Neutron Flux channels. The Source Range channels will initiate a Reactor trip at about 105 counts per second unless manually blocked when P-6 becomes active. The Intermediate Range channels will initiate a Reactor trip at a current level equivalent to approximately 25% of RATED THERMAL POWER unless manually blocked when P-10 becomes active. No credit was taken for operation of the trips associated with either the Intermediate or Source Range Channels in the accident analyses; however, their functional capability at the specified trip settings is required by this specification to enhance the overall reliability of the Reactor Trip System. Overtemperature AT The Overtemperature AT trip provides core protection to prevent DNB for all combinations of pressure, power, coolant temperature, and axial power distribution, provided that the transient is slow with respect to piping transit delays from the core to the temperature detectors, and pressure is within the range between the Pressurizer High and Low Pressure trips. The Setpoint is automatically varied with: (1) coolant temperature to correct for temperature induced changes in density and heat capacity of water and includes dynamic compensation for piping delays from the core to the loop temperature detectors, (2) pressurizer pressure, and (3) axial power distribution. With normal axial power distribution, this Reactor trip limit 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 the notations in Table 2.2-1. Although a direction of conservatism is identified for the Overtemperature AT reactor trip function K? and K gains, the gains should be set as close as possible to the values contained in Note 1 to ensure that the Overtemperature AT setpoint is consistent with the assumptions of the safety analyses. Overpower AT The Overpower AT trip provides assurance of fuel integrity (e.g., no fuel pellet melting and less than 1% cladding strain) under all possible overpower conditions, limits the required range for Overtemperature AT MILLSTONE - UNIT 3 0965 B 2-5 Amendment No. fl, gp, J9Z,217
LIMITING SAFETY SYSTEM SETTINGS BASES Reactor Trip System Interlocks (Continued) P-8 On increasing power, P-8 automatically enables Reactor trips on low flow in one or more reactor coolant loops. On decreasing power, the P-8 automatically blocks the above listed trips. P-9 On increasing power, P-9 automatically enables Reactor trip on Turbine trip. On decreasing power, P-9 automatically blocks Reactor trip on Turbine trip. P-10 On increasing power, P0 provides input to P-7 to ensure that Reactor Trips on low flow in more than one reactor coolant loop, reactor coolant pump low shaft speed, pressurizer low pressure and pressurizer high level are active when power reaches 11%. It also allows the manual block of the Intermediate Range trip and the Low Setpoint Power Range trip; and automatically blocks the Source Range trip and deenergizes the Source Range high voltage power. On decreasing power, P-10 resets to automatically reactivate the Intermediate Range trip and the Low Setpoint Power Range trip before power drops below 9%. It also provides input to reset P-7. P-13 Provides input to P-7. MILLSTONE - UNIT 3 B 2-8 Amendment No. PA,217 0966
POWER DISTRIBUTION LIMITS AXIAL FLUX DIFFERENCE (Continued) (2) APLND (for base load operation). Penalty deviation minutes for base load operation are not accumulated based on the short period of time during which operation outside of the target band is allowed. 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR and RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR The limits on heat flux hot channel factor, RCS flow rate, and nuclear enthalpy rise hot channel factor ensure that: (1) the design limits on peak local power density and minimum DNBR are not exceeded and (2) in the event of a LOCA the peak fuel clad temperature will not exceed the 2200'F ECCS acceptance criteria limit. Each of these is measurable but will normally only be determined periodically as specified in Specifications 4.2.2 and 4.2.3. This periodic surveillance is sufficient to ensure that the limits are maintained provided:
- a. Control rods in a single group move together with no individual rod insertion differing by more than +12 steps, indicated, from the group demand position;
- b. Control rod groups are sequenced with overlapping groups as described in Specification 3.1.3.6;
- c. The control rod insertion limits of Specifications 3.1.3.5 and 3.1.3.6 are maintained; and
- d. The axial power distribution, expressed in terms of AXIAL FLUX DIFFERENCE, is maintained within the limits.
FN6H will be maintained within its limits provided Conditions a. through
- d. above are maintained. The relaxation of FNAH as a function of THERMAL POWER allows changes in the radial power shape for all permissible rod insertion limits.
The FNAH as calculated in Specification 3.2.3.1 is used in the various accident analyses where FNAH influences parameters other than DNBR, e.g., peak clad temperature, and thus is the maximum "as measured" value allowed. MILLSTONE - UNIT 3 B 3/4- 2-3 Amendment No. , P'217 0985
POWER DISTRIBUTION LIMITS RASRS 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR and RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR (Continued) Margin is maintained between the safety analysis limit DNBR and the design limit DNBR. This margin is more than sufficient to offset any rod bow penalty and transition core penalty. The remaining margin is available for plant design flexibility. When an FQ measurement is taken, an allowance for both experimental error and manufacturing tolerance must be made. An allowance of 5% is appropriate for a full core map taken with the incore detector flux mapping system and a 3% allowance is appropriate for manufacturing tolerance. The heat flux hot channel factor, F(Z), is measured periodically using the incore detector system. These measurements are generally taken with the core at or near steady state conditions. Using the measured three dimensional power distributions, it is possible to derive F(Z), a computed value of F(Z). However, because this value represents a steady state condition, it does not include the variations in the value of F(Z) that are present during nonequilibrium situations. To account for these possible variations, the steady state limit of F(Z) is adjusted by an elevation dependent factor appropriate to either RAOC or base load operation, W(Z) or W(Z)BL, that accounts for the calculated worst case transient conditions. The W(Z) and W(Z)BL, factors described above for normal operation are specified in the COLR per Specification 6.9.1.6. Core monitoring and control under nonsteady state conditions are accomplished by operating the core within the limits of the appropriate LCOs, including the limits on AFD, QPTR, and control rod insertion. Evaluation of the steady state F(Z) limit is performed in Specification 4.2.2.1.2.b and 4.2.2.1.4.b while evaluation nonequilibrium limits are performed in Specification 4.2.2.1.2.c and 4.2.2.1.4.c. When RCS flow rate and FNAH are measured, no additional allowances are necessary prior to comparison with the limits of the Limiting Condition for Operation. Measurement errors of 2.4% for RCS total flow rate and 4% for FNLH have been allowed for in determination of the design DNBR value. The measurement error for RCS total flow rate is based upon performing a precision heat balance and using the result to calibrate the RCS flow rate indicators. Potential fouling of the feedwater venturi which might not be detected could bias the result from the precision heat balance in a non-conserVative manner. Therefore, a penalty of 0.1% for undetected fouling of the feedwater venturi will be added if venturis are not inspected and cleaned at least once for 18 months. Any fouling which might bias the RCS flow rate measurement greater than 0.1% can be detected by monitoring and trending various plant performance parameters. If detected, action shall be taken before performing subsequent precision heat balance measurements, i.e., either the effect of the fouling shall be quantified and compensated for in the RCS flow rate measurement or the venturi shall be cleaned to eliminate the fouling. MILLSTONE - UNIT 3 MlRR; B 3/4 2-4 Amendment No. 17, PP, 770, 217
POWER DISTRIBUTION LIMITS RASES HEAT FLUX HOT CHANNEL FACTOR and RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR (Continued) The 12-hour periodic surveillance of indicated RCS flow is sufficient to detect only flow degradation which could lead to operation outside the accept-able region of operation defined in Specification 3.2.3.1. 3/4.2.4 QUADRANT POWER TILT RATIO The QUADRANT POWER TILT RATIO limit assures that the radial power distribution satisfies the design values used in the power capability analysis. Radial power distribution measurements are made during STARTUP testing and periodically during power operation. The limit of 1.02, at which corrective action is required, provides DNB and linear heat generation rate protection with x-y plane power tilts. A limiting tilt of 1.025 can be tolerated before the margin for uncertainty in F. is depleted. A limit of 1.02 was selected to provide an allowance for the uncertainty associated with the indicated power tilt. The 2-hour time allowance for operation with a tilt condition greater than 1.02 but less than 1.09 is provided to allow identification and correc-tion of a dropped or misaligned control rod. In the event such action does not correct the tilt, the margin for uncertainty on F is reinstated by reducing the maximum allowed power by 3% for each percent of tilt in excess of 1. For purposes of monitoring QUADRANT POWER TILT RATIO when one excore detector is inoperable, the moveable incore detectors are used to confirm that the normalized symmetric power distribution is consistent with the QUADRANT POWER TILT RATIO. The incore detector monitoring is done with a full incore flux map or two sets of four symmetric thimbles. The two sets of four symmetric thimbles is a unique set of eight detector locations. These locations are C-8, E-5, E-11, H-3, H-13, L-5, L-11, N-8. 3/4.2.5 DNB PARAMETERS The limits on the DNB-related parameters assure that each of the parameters are maintained within the normal steady-state envelope of operation assumed in the transient and accident analyses. The limits are consistent with the initial FSAR assumptions and have been analytically demonstrated adequate to maintain a minimum DNBR greater than the design limit throughout each analyzed transient. The indicated Tg value of 591.1VF MILLSTONE - UNIT 3 B 3/4 2-5 Amendment No. Z77M f'217 0985
POWER DISTRIBUTION LIMITS RASPF DNB PARAMETERS (Continued) and the indicated pressurizer pressure value is 2218 psia. The calculated values of the DNB related parameters will be an average of the indicated values for the operable channels. The 12-hour periodic surveillance of these parameters through instrument readout is sufficient to ensure that the parameters are restored within their limits following load changes and other expected transient operation. Measurement uncertainties have been accounted for in determining the parameter limits. MILLSTONE - UNIT 3 B 3/4 2-6 Amendment No. 1Z, P9,217 0985
INSTRUMENTATION BASES 3/4 3.5 SHUTDOWN MARGIN MONITOR The Shutdown Margin Monitors provide an alarm that a Boron Dilution Event may be in progress. The minimum count rate of Specification 3/4.3.5 and the SHUTDOWN MARGIN requirements of Figures 3.1-1, 3.1-3, 3.1-4, and 3.1-5 ensure that at least 15 minutes are available for operator action from the time of the Shutdown Margin Monitor alarm to total loss of shutdown margin. By borating an additional 150 ppm above the SHUTDOWN MARGIN required by Figure 3.1-1, or 350 ppm above the SHUTDOWN MARGIN required by Figure 3.1-3, 3.1-4, or 3.1-5, lower values of minimum count rate are accepted. Shutdown Margin Monitors
Background:
The purpose of the Shutdown Margin Monitors (SMM) is to annunciate an increase in core subcritical multiplication allowing the operator at least 15 minutes response time to mitigate the consequences of the inadvertent addition of unborated primary grade water (boron dilution event) into the Reactor Coolant System (RCS) when the reactor is shut down (Modes 3, 4, and 5). The SMMs utilizes two channels of source range instrumentation (GM detectors). Each channel provides a signal to its applicable train of SMM. The SMM channel uses the last 600 or more counts to calculate the count rate and updates the measurement after 30 new counts or I second, whichever is longer. Each channel has 20 registers that hold the counts (20 registers X 30 count = 600 counts) for averaging the rate. As the count rate decreases, the longer it takes to fill the registers (fill the 30 count minimum). As the instrument's measured count rate decreases, the delay time in the instrument's response increases. This delay time leads to the requirement of a minimum count rate for OPERABILITY. During the dilution event, count rate will increase to a level above the normal steady state count rate. When this new count rate level increases above the instrument's setpoint, the channel will alarm alerting the operator of the event. Applicable Safety Analysis The SMM senses abnormal increases in the source range count per second and alarms the operator of an inadvertent dilution event. This alarm will occur at least 15 minutes prior to the reactor achieving criticality. This 15 minute window allows adequate operator response time to terminate the dilution, FSAR Section 15.4.6. LCO LCO 3.3.5 provides the requirements for OPERABILITY of the instrumentation of the SMMs that are used to mitigate the boron dilution event. Two trains are required to be OPERABLE to provide protection against single failure. MILLSTONE - UNIT 3 B 3/4 3-7 Amendment No. j0, 217 0986
3/4.4 REACTOR COOLANT SYSTEM RASF.S _ 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION The purpose of Specification 3.4.1.1 is to require adequate forced flow rate for core heat removal in MODES I and 2 during all normal operations and anticipated transients. Flow is represented by the number of reactor coolant pumps in operation for removal of heat by the steam generators. To meet safety analysis acceptance criteria for DNB, four reactor coolant pumps are required at rated power. An OPERABLE reactor coolant loop consists of an OPERABLE reactor coolant pump in operation providing forced flow for heat transport and an OPERABLE steam generator in accordance with Specification 3.4.5. With less than the required reactor coolant loops in operation this specification requires that the plant be in at least HOT STANDBY within 6 hours. In MODE 3, three reactor coolant loops, and in Mode 4, two reactor coolant loops provide sufficient heat removal capability for removing core decay heat even in the event of a bank withdrawal accident; however, in MODE 3 a single reactor coolant loop provides sufficient heat removal capacity if a bank withdrawal accident can be prevented, i.e., the Control Rod Drive System is not capable of rod withdrawal. In MODE 4, if a bank withdrawal accident can be prevented, a single reactor coolant loop.or RHR loop provides sufficient heat removal capability for removing decay heat; but single failure considerations require that at least two loops (any combination of RHR or RCS) be OPERABLE. In MODE 5, with reactor coolant loops filled, a single RHR loop provides sufficient heat removal capability for removing decay heat; but single failure considerations require that at least two RHR loops or at least one RHR loop and two steam generators be OPERABLE. In MODE 5 with reactor coolant loops not filled, a single RHR loop provides sufficient heat removal capability for removing decay heat; but single failure considerations, and the unavailability of the steam generators as a heat removing component, require that at least two RHR loops be OPERABLE. In MODE 5, during a planned heatup to MODE 4 with all RHR loops removed from operation, an RCS loop, OPERABLE and in operation, meets the requirements of an OPERABLE and operating RHR loop to circulate reactor coolant. During the heatup there is no requirement for heat removal capability so the OPERABLE and operating RCS loop meets all of the required functions for the heatup condition. Since failure of the RCS loop, which is OPERABLE and operating, could also cause the associated steam generator to be inoperable, the associated steam generator cannot be used as one of the steam generators used to meet the requirement of LCO 3.4.1.4.1.b. The operation of one reactor coolant pump (RCP) or one RHR pump provides adequate flow to ensure mixing, prevent stratification and produce gradual reactivity changes during boron concentration reductions in the Reactor Coolant System. The reactivity change rate associated with boron reduction will, therefore, be within the capability of operator recognition and control. The restrictions on starting the first RCP in MODE 4 below the cold overpressure protection enable temperature (226°F), and in MODE 5 are provided to prevent RCS pressure transients. These transients, energy additions due to the differential temperature between the steam generator secondary side and the RCS, can result in pressure excursions which could challenge the P/T limits. MILLSTONE - UNIT 3 B 3/4 4-1 Amendment No. 7,PO, , 7}7, 217 0987 I7,
3/4.4 REACTOR COOLANT SYSTEM BASES (Continued) The RCS will be protected against overpressure transients and will not exceed the reactor vessel isothermal beltline P/T limit by restricting RCP starts based on the differential water temperature between the secondary side of each steam generator and the RCS cold legs. The restrictions on starting the first RCP only apply to RCPs in RCS loops that are not isolated. The restoration of isolated RCS loops is normally accomplished with all RCPs secured. If an isolated RCS loop is to be restored when an RCP is operating, the appropriate temperature differential limit between the secondary side of the isolated loop steam generator and the in service RCS cold legs is applicable, and shall be met prior to opening the loop isolation valves. The temperature differential limit between the secondary side of the steam generators and the RCS cold legs is based on the equipment providing cold overpressure protection as required by Technical Specification 3.4.9.3. If the pressurizer PORVs are providing cold overpressure protection, the steam generator secondary to RCS cold leg water temperature differential is limited to a maximum of 50'F. If any RHR relief valve is providing cold overpressure protection and RCS cold leg temperature is above 1500F, the steam generator secondary water temperature must be at or below RCS cold leg water temperature. If any RHR relief valve is providing cold overpressure protection and RCS cold leg temperature is at or below 150'F, the steam generator secondary to RCS cold leg water temperature differential is limited to a maximum of 500F. Specification 3.4.1.5 The reactor coolant loops are equipped with loop stop valves that permit any loop to be isolated from the reactor vessel. One valve is installed on each hot leg and one on each cold leg. The loop stop valves are used to perform maintenance on an isolated loop. Operation in MODES 1-4 with a RCS loop stop valve closed is not permitted except for the mitigation of emergency or abnormal events. If a loop stop valve is closed for any reason, the required actions of this specification must be completed. To ensure that inadvertent closure of a loop stop valve does not occur, the valves must be open with power to the valve operators removed in MODES 1, 2, 3 and 4. The safety analyses performed for the reactor at power assume that all reactor coolant loops are initially in operation and the loop stop valves are open. This LCO places controls on the loop stop valves to ensure that the valves are not inadvertently closed in MODES 1, 2, 3 and 4. The inadvertent closure of a loop stop valve when the Reactor Coolant Pumps (RCPs) are operating will result in a partial loss of forced reactor coolant flow. If the reactor is at rated power at the time of the event, the effect of the partial loss of forced coolant flow is a rapid increase in the coolant temperature which could result in DNB with subsequent fuel damage if the reactor is not tripped by the Low Flow reactor trip. If the reactor is shutdown and a RCS loop is in operation removing decay heat, closure of the loop stop valve associated with the operating loop could also result in increasing coolant temperature and the possibility of fuel damage. The loop stop valves have motor operators. If power is inadvertently restored to one or more loop stop valve operators, the potential exists for accidental closure of the affected loop stop valve(s) and the partial loss of forced reactor coolant flow. With power applied to a valve operator, only the interlocks prevent the valve from being operated. Although operating MILLSTONE - UNIT 3 B 3/4 4-la Amendment No. 7, PO, PY, 77, 0987 }97, V71g 217
3/4.4 REACTOR COOLANT SYSTEM RASFS procedures and interlocks make the occurrence of this event unlikely, the prudent action is to remove power from the loop stop valve operators. The time period of 30 minutes to remove power from the loop stop valve operators is sufficient considering the complexity of the task. Should a loop stop valve be closed in MODES 1 through 4, the affected valve must be maintained closed and the plant placed in MODE 5. Once in MODE 5, the isolated loop may be started in a controlled manner in accordance with LCO 3.4.1.6, "Reactor Coolant System Isolated Loop Startup." Opening the closed loop stop valve in MODES 1 through 4 could result in colder water or water at a lower boron concentration being mixed with the operating RCS loops resulting in positive reactivity insertion. The time period provided in Action 3.4.1.5.b allows time for borating the operating loops to a shutdown boration level such that the plant can be brought to MODE 3 within 6 hours and MODE 5 within 30 hours. The allowed action times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. Surveillance Requirement 4.4.1.5 is performed at least once per 31 days to ensure that the RCS loop stop valves are open, with power removed from the loop stop valve operators. The primary function of this Surveillance is to ensure that power is removed from the valve operators, since Surveillance Requirement 4.4.1.1 requires verification every 12 hours that all loops are operating and circulating reactor coolant, thereby ensuring that the loop stop valves are open. The frequency of 31 days ensures that the required flow is available, is based on engineering judgement, and has proven to be acceptable. Operating experience has shown that the failure rate is so low that the 31 day frequency is justified. Specification 3.4.1.6 The requirement to maintain the isolated loop stop valves shut with power removed ensures that no reactivity addition to the core could occur due to the startup of an isolated loop. Verification of the boron concentration in an isolated loop prior to opening the first stop valve provides a reassurance of the adequacy of the boron concentration in the isolated loop. RCS Loops Filled/Not Filled: In MODE 5, any RHR train with only one cold leg injection path is sufficient to provide adequate core cooling and prevent stratification of boron in the Reactor Coolant System. The definition of operability states that the system or subsystem must be capable of performing its specified function(s). The reason for the operation of one reactor coolant pump (RCP) or one RHR pump is to:
- Provide sufficient decay heat removal capability
- Provide adequate flow to ensure mixing to:
- Prevent stratification
- Produce gradual reactivity changes due to boron concentration changes in the RCS The definition of "Reactor coolant loops filled" includes a loop that is filled, swept, and vented, and capable of supporting natural circulation heat transfer. This allows the non-operating RHR loop to be removed from service MILLSTONE - UNIT 3 B 3/4 4-lb Amendment No. 7, fg, , 797, 217 1i7. 22.
3/4.4 REACTOR COOLANT SYSTEM BASES (Continued) while filling and unisolating loops as long as steam generators on the operable reactor coolant loops are available to support decay heat removal. Any loop being unisolated is not OPERABLE until the loop has been swept and vented. The process of sweep and vent will make the previously OPERABLE loops inoperable and the requirements of LCO 3.4.1.4.2, "Reactor Coolant System, Cold Shutdown - Loops Not Filled," are applicable. When the RCS has been filled, swept and vented using an approved procedure, all unisolated loops may be declared OPERABLE. One cold leg injection isolation valve on an RHR train may be closed without considering the train to be inoperable, as long as the following conditions exist:
- CCP temperature is at or below 950 F
- Initial RHR temperature is below 184°F
- The single RHR cold leg injection flow path is not utilized until a minimum of 24 hours after reactor shutdown
- CCP flow is at least 6,600 gpm
- RHR flow is at least 2,000 gpm In the above system lineup, total flow to the core is decreased compared to the flow when two cold legs are in service. This is acceptable due to the substantial margin between the flow required for cooling and the flow available, even through a slightly restricted RHR train.
The review concerning boron stratification with the utilization of the single injection point line, indicates there will not be a significant change in the flow rate or distribution through the core, so there is not an increased concern due to stratification. Flow velocity, which is high, is not a concern from a flow erosion or pipe loading standpoint. There are no loads imposed on the piping system which would exceed those experienced in a seismic event. The temperature of the fluid is low and is not significant from a flow erosion standpoint. The boron dilution accident analysis, for Millstone Unit 3 in MODE 5, assumes a full RHR System flow of approximately 4,000 gpm. Westinghouse analysis, Reference (1), for RHR flows down to 1,000 gpm, determined adequate mixing results. As the configuration will result in a RHR flow rate only slightly less then 4,000 gpm there is no concern in regards to a boron dilution accident. The basis for the requirement of two RCS loops OPERABLE is to provide natural circulation heat sink in the event the operating RHR loop is lost. If the RHR loop were lost, with two loops swept and vented and two loops air bound, natural circulation would be established in the two swept loops. Natural circulation would not be established in the air bound loops. Since there would be no circulation in the air bound loops, there would be no mechanism for the air in those loops to be carried to the vessel, and subsequently into the swept loops rendering them inoperable for heat sink requirements. MILLSTONE - UNIT 3 B 3/4 4-Ic Amendment No. 217 nQR7
3/4.4 REACTOR COOLANT SYSTEM BASES (Continued) The LCO is met as long as at least two reactor coolant loops are OPERABLE and the following conditions are satisfied:
- One RHR loop is OPERABLE and in operation, with exceptions as allowed in Technical Specifications; and Either of the following:
- An additional RHR loop OPERABLE, with exceptions as allowed in Technical Specifications; or
- The secondary side water level of at least two steam generators shall be greater than 17% (These are assumed to be on OPERABLE reactor coolant loops)
When the reactor coolant loops are swept, the mechanism exists for air to be carried into previously OPERABLE loops. All previously OPERABLE loops are declared inoperable and an additional RHR loop is required OPERABLE as specified by LCO 3.4.1.4.2 for loops not filled. When the RCS has been filled, swept, and vented using an approved procedure, all unisolated loops may be declared OPERABLE. ISOLATED LOOP STARTUP The below requirements are for unisolating a loop with all four loops isolated while decay heat is being removed by RHR and to clarify prerequisites to meet T/S requirements for unisolating a loop at any time. With no RCS loops operating, the two RHR loops referenced in Specification 3.4.1.4.2 are the operating loops. Starting in MODE 4 as referenced in Specification 3.4.1.3, the RHR loops are allowed to be used in place of an operating RCS loop. Specification 3.4.1.4.2 requires two RHR loops OPERABLE and at least one in operation. Ensuring the isolated cold leg temperature is within 20'F of the highest RHR outlet temperature for the operating RHR loops within 30 minutes prior to opening the cold leg stop valve is a conservative approach since the major concern is a positive reactivity addition. SR 4.4.1.6.1: When in MODE 5 with all RCS loops isolated, the two RHR loops referenced in LCO 3.4.1.4.2 shall be considered the OPERABLE RCS loops. ISOLATED LOOP STARTUP (Continued) The isolated loop cold leg temperature shall be determined to be within 20OF of the highest RHR outlet temperature for the operating RHR loops within 30 minutes prior to opening the cold leg stop valve. Surveillance requirement 4.4.1.6.2 is met when the following actions occur within 2 hours prior to opening the cold leg or hot leg stop valve:
- An RCS boron sample has been taken and analyzed to determine current boron concentration
- The SHUTDOWN MARGIN has been determined using OP 3209B, "Shutdown Margin" using the current boron concentration determined above
- For the isolated loop being restored, the power to both loop stop valves has been restored MILLSTONE - UNIT 3 B 3/4 4-1d Amendment No. 217 0987
3/4.4 REACTOR COOLANT SYSTEM BASES (Continued) Surveillance 4.4.1.6.2 indicates that the reactor shall be determined subcritical by at least the amount required by Specifications 3.1.1.1.2 or 3.1.1.2 for MODE 5 or Specification 3.9.1.1 for MODE 6 within 2 hours of opening the cold leg or hot leg stop valve. Specification 3.1.1.1.2 requires the SHUTDOWN MARGIN to be as shown in Figure 3.1-2 for three loop operation. Figure 3.1-2 is for three loop operation in MODE 3. The other figures, as used by this specification, require four loop operation, so cannot be used to determine the required SHUTDOWN MARGIN for the MODE 5 loops isolated condition. Specification 3.1.1.2 requires the SHUTDOWN MARGIN to be as shown in Figure 3.1-5 or Figure 3.1-4 with CVCS aligned to preclude boron dilution. This specification is for loops not filled and therefore is applicable to an all loops isolated condition. Specification 3.9.1.1 requires Keff of 0.95 or less, or a boron concentration of greater than or equal to 2,600 ppm in MODE 6. Specification 3.1.1.1.2 or 3.1.1.2 for MODE 5, both require boron concentration to be determined at least once each 24 hours. SR 4.1.1.1.2.1.b.2 and 4.1.1.2.1.b.1 satisfy the requirements of Specifications 3.1.1.1.2 and 3.1.1.2 respectfully. Specification 3.9.1.1 for MODE 6 requires boron concentration to be determined at least once each 72 hours. S.R.4.9.1.1.2 satisfy the requirements of Specification 3.9.1.1.
References:
- 1. Letter NEU-94-623, dated July 13, 1994; Mixing Evaluation for Boron Dilution Accident in Modes 4 and 5, Westinghouse HR-59782.
- 2. Memo No. MP3-E-93-821, dated October 7, 1993.
MILLSTONE - UNIT 3 B 3/4 4-le Amendment No. 217 0987
REACTOR COOLANT SYSTEM BASES PRESSURE/TEMPERATURE LIMITS (continued) Violating the LCO limits places the reactor vessel outside of the bounds of the analyses and can increase stresses in other RCPB components. The consequences depend on several factors, as follows:
- a. The severity of the departure from the allowable operating PT regime or the severity of the rate of change of temperature;
- b. The length of time the limits were violated (longer violations allow the temperature gradient in the thick vessel walls to become more pronounced); and
- c. The existences, sizes, and orientations of flaws in the vessel material.
APPLICABILITY The RCS P/T limits LCO provides a definition of acceptable operation for prevention of nonductile failure of ferritic RCS components using ASME Section XI, Appendix G,.as modified by Code Case -640 and the additional requirements of IOCFR50, Appendix G (Ref. 1). The P/T limits were developed to provide requirements for operation during heatup or cooldown (MODES 3, 4, and 5) or ISLH testing, in keeping with the concern for nonductile failure. The limits do not apply to the Pressurizer. During MODES 1 and 2, other Technical Specifications provide limits for operation that can be more restrictive than or can supplement these P/T limits. LCO 3.2.5, "DNB Parameters"; LCO 3.2.3.1, RCS Flow Rate and Nuclear Enthalpy Rise Hot Channel Factor"; LCO 3.1.1.4, "Minimum Temperature for Criticality"; and Safety Limit 2.1, "Safety Limits," also provide operational restrictions for pressure and temperature and maximum pressure. Furthermore, MODES 1 and 2 are above the temperature range of concern for nonductile failure, and stress analyses have been performed for normal maneuvering profiles, such as power ascension or descent. ACTIONS Operation outside the P/T limits must be corrected so that the RCPB is returned to a condition that has been verified by stress analyses. The Allowed Outage Times (AOTs) reflects the urgency of restoring the parameters to within the analyzed range. Most violations will not be severe, and the activity can be accomplished in this time in a controlled manner. Besides restoring operation within limits, an evaluation is required to determine MILLSTONE - UNIT 3 0988 B 3/4 4-10 Amendment No. 7W7, P7. 217
3/4.7 PLANT SYSTEMS BASES 3/4.7.1 TURBINE CYCLE 3/4.7.1.1 SAFETY VALVES The OPERABILITY of the main steam line Code safety valves ensures that the Secondary System pressure will be limited to within 110% (1305 psig) of its design pressure of 1185 psig during the most severe anticipated system operational transient. The maximum relieving capacity is associated with a Turbine trip from 100% RATED THERMAL POWER coincident with an assumed loss of condenser heat sink (i.e., no steam bypass to the condenser). The specified valve lift settings and relieving capacities are in accordance with the requirements of Section III of the ASME Boiler and Pressure Code, 1971 Edition. The design minimum total relieving capacity for all valves on all of the steam lines is 1.579 X 107 lbs/h which is 105% of the total secondary steam flow of 1.504 X 107 lbs/h at 100% RATED THERMAL POWER. STARTUP and/or POWER OPERATION is allowable with safety valves inoperable within the limitations of the ACTION requirements on the basis of the reduction in Secondary Coolant System steam flow and THERMAL POWER required by the reduced Reactor trip settings of the Power Range Neutron Flux channels. The Reactor Trip Setpoint reductions are derived on the following bases: Hi 4 =(100/Q) (w.hfgN) K where: Hi = Safety Analysis power range high neutron flux setpoint, percent Q = Nominal NSSS power rating of the plant (including reactor coolant pump heat), Mwt K = Conversion factor, 947.82 (Btu/sec) Mwt hg = heat of vaporization for steam at the highest MSSV opening pressure including tolerance ( 3%) and accumulation, as appropriate, Btu/lbm N = Number of loops in plant MILLSTONE - UNIT 3 0989 B 3/4 7-1 Amendment No. IpZ, 217}}