ML032580501

Technical Specifications Associated with Pressure/Temperature Limits and Overpressure Protection Systems

ML032580501
Person / Time
Site:	Seabrook
Issue date:	09/11/2003
From:	NRC/NRR/DLPM/LPD1
To:
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Download: ML032580501 (16)

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INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PALGE FIGURE 3.4-1 DOSE EQUIVALENT 1-131 REACTOR COOLANT SPECIFIC ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE REACTOR COOLANT SPECIFIC ACTIVITY > 1IiCi / gram DOSE EQUIVALENT 1-131 .............................................................................. 3/4 4-28 TABLE 4.4-3 REACTOR COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM....................................................................................................... 314 4-29 314A.9 PRESSUREITEMPERATURE LIMITS General............................................................................................................. 3/4 4-30 FIGURE 3.4-2 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS -

APPLICABLE UP TO 20 EFPY........................................................................ 3/4 4-31 l FIGURE 3.4-3 REACTOR COOLANT SYSTEM COOLDOWN LIMITATIONS -

APPLICABLE UP TO 20 EFPY ........................... _ 3/4 4-321 Pressurizer...................................................................................................... 3/4 4-33 Overpressure Protection Systems ............................. 3/4 4-34 FIGURE 3.4-4 RCS COLD OVERPRESSURE PROTECTION SETPOINTS .................... 3/4 4-36 314A.10 STRUCTURAL INTEGRITY.... 3/4 4-37 3/4.4.11 REACTOR COOLANT SYSTEM VENTS .i 3/4 4-38 314.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS Hot Standby, Startup, and Power Operation .................... 314 5-1 Shutdown .......................................................................... 3/4 5-3 3/4.5.2 ECCS SUBSYSTEMS - Tan GREATER THAN OR EQUAL TO 350F.... 3/4 5-4 3/4.5.3 ECCS SUBSYSTEMS - T.a LESS THAN 350F . 3/4 5-8 ECCS SUBSYSTEMS - T., Equal To or Less Than 200F . 3/4 5-10 314.5.4 REFUELING WATER STORAGE TANK. 314 5-11 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT Containment Integrity....................................................................................... 314 6-1 Containment Leakage ...................................................... 3/4 6-2 SEABROOK - UNIT I vi Amendment No. X0, 89

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 314.12.2 (THIS SPECIFICATION NUMBER IS NOT USED) ........................................ 314 12-3 3/4.12.3 (THIS SPECIFICATION NUMBER IS NOT USED) ........................................ 3/4 12-5 3.0/4.0 BASES 314.0 APPLICABILITY ......................................................... B 314 0-1 314.1 REACTIVIlY CONTROL SYSTEMS 314.1.1 BORATION CONTROL ................. ........................................ B 314 1-1 3/4.12 BORATION SYSTEMS ............................................B 314 1-2 314.1.3 MOVABLE CONTROL ASSEMBLIES....................................................................... B 314 1-4 314.2 POWER DISTRIBUTION LIMITS1. ........................................................................... B 3/4 2-1 314.2.1 AXIAL FLUX DIFFERENCE ... .......... B 314 2-1 3/4.2.2 and 3142.3 HEAT FLUX HOT CHANNEL FACTOR AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR ......................................... B 314 2-2 3/42A QUADRANT POWER TILT RATIO...............................................................B 3/4 2-3 3/4.2.5 DNB PARAMETERS ......................................................... B 314 2-4 3/4.3 INSTRUMENTATION 314.3.1 and 314.3.2 REACTOR TRIP SYSTEM and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION ............B 314 3-1 3/4.3.3 MONITORING INSTRUMENTATION .............. .................. ............ B 3/4 3-3 3/4.3.4 (THIS SPECIFICATION NUMBER IS NOT USED) ..................... ............ B 3/4 3-6 3/4.4 REACTOR COOLANT SYSTEM 3/4.A.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION B 314 4-1 314A.2 SAFET'Y VALVES . ... B 314 4-1 3/4.4.3 PRESSURIZER . .. B 314 4-2 314.4.4 RELIEF VALVES . ... ;.B 3/4 4-2 3/M .4.

314A.5 RELE FAM VE R....................................................................................

STEAM GENERATORS......................... B B 314 314 4-2 4-2 3/4.4.6 REACTOR COOLANT STELAKG.B 3/4 4-3 3/44.7 CHEMISTRY . . . .B 3/4 4-5 3/4.4.8 SPECIFIC ACTIVITY...................................................................................... B 3/4 4-5 314.A.9 PRESSURE/TEMPERATURE LIMITS............................................................ B 3/4 4-7 FIGURE B 354A-1 (THIS FIGURE NUMBER IS NOT USED) . . ....................

B 3/4 4-9 FIGURE B 3/4.4-2 (This figure number not used). ........................................................... B 3/4 4-10 SEABROOK - UNIT I x Amendment No. rag, 66, 74 89

INDEX BASES SECTION PAGE TABLE B 314.A-1 (THIS TABLE NUMBER IS NOT USED) ............................................ B 3/4 4-11 314.A.10 STRUCTURACL. INTEGRITY'.................................................... B 3/4 4-16 314.4.11 REACTOR COOLANT SYSTEM VENTS . B 3/4 4-16 314.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS .... ................................................ B 3/4 5-1 3/4.5.2 and 314.5.3 ECCS SUBSYSTE M.S.............................................................. B 3/4 5-1 3/4.5.4 REFUELING WATER STORAGE TANK ........................................... B 3/4 5-2 3/4.6 CONTAINMENT SYSTEMS 314.6.1 PRIMARY CONTAINMENT .B 3/4 6-1 314.6.2 DEPRESSURIZATION AND COOLING SYSTEMS .B 3/4 6-3 3/4.6.3 CONTAINMENT ISOLATION VALVES .. B 3/4 6-3 314.64 COMBUSTIBLE GAS CONTROL . .B 3/4 6-3 3/4.6.5 CONTAINMENT ENCLOSURE BUILDING . .B 3 64 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE .B 3/4 7-1 3/4.7.2 STEAM GENERATOR PRESSURE/TEMPERATURE LIMITATION. B 3/4 7-3 3/4.7.3 PRIMARY COMPONENT COOLING WATER SYSTEM .B 3/4 7-3 3/4.7.4 SERVICE WATER SYSTEMI.... B 3/4 7-3 3/4.7.5 ULTIMATE HEAT SINK . . .B 3/4 7-3 314.7.6 CONTMROL ROOM S UBSYSTE[M S.....0.... . .. B 3/4 7-4 3J4.7.7 SNUBBERS . . .B 34 7-4 314.7.8 SEALED SOURCE CONTAMINATION . . .B 34 7-5 314.7.9 (THIS SPECIFICATION NUMBER IS NOT USED) . .B 314 7-5 314.7.10 (THIS SPECIFICATION NUMBER IS NOT USED) . .B 3/4 7-5 3/4.8 ELECTRICAL POWER SYSTEMS 314.8.1, 3/4.8.2, and 3/4.8.3 A.C. SOURCES, D.C. SOURCES, and ONSITE POWER DISTRIBUTION .B 314 8-1 314.84A ELECTRICAL EQUIPMENT PROTECTIVE DEVICES ................................. B 3/4 8-3 314.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION .. B 3/4 9-1 314.9.2 INSTRUMENTATION .. B 34 9-1 314.9.3 DECAYTIME .. B 3/4 9-1 3/4.9.A CONTINMENT BUILDING PENETRATONS. . .B 3/4 9-1 314.9.5 COMMUNICATIONS .B 3/4 9-1 3/4.9.6 REFUELING MACHINE ......................................... B 3/4 9-1 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE BUILDING . .B 314 9-2 SEABROOK - UNIT I xi Amendment No. 6, 63, 89

MATERIAL PROPERTY BASIS Urftng materlal: LOWER SHELL PLATE R-18061 Umnllng AfT values at 20 EFPY1 4T. I 0gF 3J4T, 880F Curves appicable for tfirst 20 EFPY and contaln nsargbs of 20'F and 10O pslg for possibb Istrument errors 2800 I* a

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0 1Q0 2o0 Boo 400 Boo RCSTE3MPERATUJRE (Deg.F.20 Deg.F PER DlVISlIOM FGURE .4-2 REACTOR COOLANr SYSTEM HEATUP UMITATIONS - APPLICABLE UP TO 20 EFPY SEABROOK - UNIT I 3/44-31 Amendment No.4, 89

MATERIAL PROPERTY BASIS ULmiting mateodal: LOWER SHELL PLATE R-1806-1 Umling ART values at 20 tFPYiI4T, 109OF 314T, 6ESF Cuvesapplicableforthe first 20 EFPY and Contalh nargins of 20F and 100 pslq for possible hstrument errOrs 2800 -. . . . . . . . . . . . . . . . * . . . .

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UP~~tl~-l SEABROOK - UNIT I 3J4 4-32 Amendment No. 49, 89

REACTOR COOLANT SYSTEM PRESSUREITEMPERATURE LIMITS OVERPRESSURE PROTECTION SYSTEMS LIMITING CONDITION FOR OPERATION 3.4.9.3 The following Overpressure Protection Systems shall be OPERABLE:

a. In MODE 4 when the temperature of any RCS cold leg Is less than or equal to 290 0F; and In MODE 5 and MODE 6 with all Safety Injection pumps Inoperable at least one of the following groups of two overpressure protection devices shall be OPERABLE when the RCS Is not depressurized with an RCS vent area of greater than or equal to 1.58 square Inches:

1) Two residual heat removal (RHR) suction relief valves each with a setpoint of 450 psig +0,-3 %; or

2) Two power-operated relief valves (PORVs) with lift setpoints that vary with RCS temperature which do not exceed the limit established in Figure 3.4-4, or

3) One RHR suction relief valve and one PORV with setpoints as required above.

b. In MODE 5 and MODE 6 with all Safety Injection pumps except one Inoperable:

1) The Reactor Coolant System (RCS) depressurized with an RCS vent area equal to or greater than 18 square Inches, or

2) The RCS in a reduced inventory condition*.

APPLICABILITY: MODE 4 when the temperature of any RCS cold leg Is less than or equal to 2900F; MODE 5 and MODE 6 with the reactor vessel head on and the vessel head closure bolts not fully detensioned.

ACTION:

a) In MODE 4 with all Safety Injection pumps inoperable and with one of the two required overpressure protection devices Inoperable, either restore two overpressure protection devices to OPERABLE status within 7 days or within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> (a) depressurize the RCS and (b) vent the RCS through at least a 1.58-square-Inch vent.

• Areduced Inventory condition exists whenever reactor vessel (RV) water level Islower than 36 Inches below the RV flange.

SEABROOK - UNIT I 314 4-34 Amendment No. 3, 5,46, Z4, 89

VALID FOR THE FIRST 20 EMPY, SEPOW CONTAINS MARGIN OFOF FORYTRSE EEECrS Ts 200.00F P ' 561.0 PSIG; 20.00 F < TS 2305F, P- 12.1*(T-200.0) + 926.0 PSIG; 230.50F < TS 255.0 0F, P - 23.15S(T-230.5) + 1295.05 PSIG; T > 255ffFp P = 34.5*r-N255.0) + 186225 PSIG 50 100 150 200 250 300 350 RCS TEMPERATE (DEG. F)

FIGURE 3.4-4 RCS COLD OVERPPESSRE PROTECTON SETPOIMT SEABROOK - UNIT I 12I4 36 Amendment No. 89

REACTOR COOLANT SYSTEM BASES 3/4.4.9 PRESSUREITEMPERATURE LIMITS The temperature and pressure changes during heatup and cooldown are limited to be consistent with the requirements given In the ASME Boiler and Pressure Vessel Code, Section Xl, Appendix G. Reference (1):

1. The reactor coolant temperature and pressure and system heatup and cooldown rates (with the exception of the pressurizer) shall be limited in accordance with Figures 3.4-2 and 3.4-3 for the service period specified thereon:

a. Allowable combinations of pressure and temperature for specific temperature change rates are below and to the right of the limit lines shown. Umit lines for cooldown rates between those presented may be obtained by Interpolation; and

b. Figures 3.4-2 and 3A-3 define limits to assure prevention of non-ductile failure only. For normal operation, other Inherent plant characteristics, e.g., pump heat addition and pressurizer heater capacity. may limit the heatup and cooldown rates that can be achieved over certain pressure-temperature ranges.

2. These limit lines shall be calculated periodically using methods provided below,

3. The secondary side of the steam generator must not be pressurized above 200 psig If the temperature of the steam generator Is below 700F,

4. The pressurizer heatup and cooldown rates shall not exceed 100°Fh and 2000F1h, respectively. The spray shall not be used if the temperature difference between the pressurizer and the spray fluid Is greater than 3200F, and

5. System preservice hydrotests and inservice leak and hydrotests shall be performed at pressures in accordance with the requirements of ASME Boiler and Pressure Vessel Code, Section Xl.

Operation within the limits of the appropriate heatup and cooldown curves assures the integrity of the reactor vessel's ferritic material against fracture Induced by combined thermal and pressure stresses. As the reactor vessel Is subjected to increasing fluence, the toughness of the limiting beltline region material continues to diminish, and consequently, even more restrictive pressure/temperature (PIT) limits must be maintained. Each PIT limit curve defines an acceptable region for normal operation during heatup or cooldown maneuvering as pressure and temperature indications are monitored to ensure that operation Iswithin the allowable region. A heatup or cooldown Isdefined as a temperature change of greater than or equal to I 0F In any one-hour period.

SEABROOK - UNIT I B 314 4-7 Amendment No. -9, 89

REACTOR COOLANT SYSTEM BASES 314.4.9 PRESSURE/TEMPERATURE LIMITS (Continued)

The PIT limits have been established In accordance with the requirements of ASME Boiler and Pressure Vessel Code Section Xl, Appendix G. as modified by ASME Code Case N 641, Reference (2), and the additional requirements of I OCFR50 Appendix G, Reference (3). The heatup and cooldown P/T limit curves for normal operation, Figures 3.4-2 and 3.4-3 respectively, are valid for a service period of 20 effective full power years (EFPY).

The technical justification and methodologies utilized In their development are documented In WCAP-15745, Reference (4). The PIT curves were generated based on the latest available reactor vessel Information and latest calculated fluences.

Heatup and Cooldown limit curves are calculated using the adjusted RTwrT (reference nfl-ductility temperature) corresponding to the limiting beitline region material of the reactor vessel. The adjusted RTNDT of the limiting material In the core region of the reactor vessel Is determined by using the unirradiated reactor vessel material fracture toughness properties, estimating the radiation-induced ARTNDT, and adding a margin.

RTNDT Increases as the material Is exposed to fast-neutron radiation. Therefore, to find the most limiting RTHDT at any time period In the reactor's life, ARTNom due to the radiation exposure associated with that time period must be added to the unirradiated RTNDT (IRTi,=). The extent of the shift In RTNDT Is enhanced by certain chemical elements (such as copper and nickel) present in reactor vessel steels. The Nuclear Regulatory Commission (NRC) has published a method for predicting radiation embrittlement in Regulatory Guide 1.99, Revision 2, Reference (5). Regulatory Guide 1.99, Revision 2, Is used for the calculation of Adjusted Reference Temperature (ART) values (IRTNDT + ARTHOT + margins for uncertainties) at the 114T and 314T locations, where T Is the thickness of the vessel at the belfline region.

The reactor vessel materials have been tested to determine their Initial RTN.T Reactor operation and resultant fast neutron (E greater than I MeV) Irradiation can cause an Increase in the RTNDT. Therefore, an adjusted reference temperature, based upon the fluence, best estimate copper and nickel content of the limiting beltline material, can be predicted using surveillance capsule data and the value of ARTNx computed by Regulatory Guide 1.99, Revision 2. Surveillance capsule data, documented In Reference (6), Is available for two capsules (Capsules U and Y) having already been removed from the reactor vessel. This surveillance capsule data was used to calculate chemistry factor (CF) values per Position 2.1 of Regulatory Guide 1.99, Revision 2. It also noted that Reference (6) concluded that all the surveillance data was credible as the beltline material was behaving as empirically predicted. The heatup and cooldown limit curves of Figures 3.4-2 and 3.4-3 include predicted adjustments for this shift In RTNDT at the end of 20 EFPY as well as adjustments for possible errors in the pressure and temperature sensing instruments.

SEA13ROOK - UNIT I B 3/44-8 Amendment No. 49, 89

REACTOR COOLANT SYSTEM BASES 314.4.9 PRESSUREITEMPERATURE LIMITS (Continued)

The results from the material surveillance program were evaluated according to ASTM E185. Capsules U and Y were removed In accordance with the requirements of ASTM E185-73 and 10CFR50, Appendix H. The lead factor represents the relationship between the fast neutron flux density at the location of the capsule and the Inner wall of the reactor vessel. Therefore, the results obtained from the surveillance specimens were used to predict future radiation damage to the reactor vessel material by using the lead factor and the withdrawal time of the capsule. The fluence values used to determine the CFs are the calculated fluence values at the surveillance capsule locations. The calculated fluence values were used for all cases. All calculated fluence values (capsule and projections) are documented In Reference (6). These fluences were calculated using the ENDFIB-VI scattering cross-section data set The measured ARTNDT values for the weld data were adjusted for chemistry using the ratio procedure given In Position 2.1 of Regulatory Guide 1.99, Revision 2.

FIGURE B 314.4-1 (THIS FIGURE NUMBER IS NOT USED)

SEABROOK - UNIT I B 3/4 4-9 Amendment No. 89

TABLE B 314.4-1 (THIS TABLE NUMBER IS NOT USED)

SEABROOK- UNIT I B 3/4411 Amendment No. 89

REACTOR COOLANT SYSTEM BASES 314A.9 PRESSURE1TEMPERATURE LIMITS (Continued)

The ASME approach for calculating the allowable limit curves for various heatup and cooldown rates specifies that the total stress Intensity factor, K,, for the combined thermal and pressure stresses at any time during heatup or cooldown cannot be greater than the reference stress intensity factor, K1c, for the metal temperature at that time. K. Is obtained from the reference fracture toughness curve, defined In Code Case N-641, Reference (2).

The Kle curve Is given by the following equation:

K, =33.2+20.734*e" 2 RT. (1)

where, KC = reference stress Intensity factor as a function of the metal temperature T and the metal reference nil-ductility temperature RTNDT This Kc curve Is based on the lower bound of static critical Kg values measured as a function of temperature on specimens of SA-533 Grade B Class 1, SA-508-1, SA-508-2, and SA-508-3 steel.

The governing equation for the heatup-cooldown analysis Is defined in Appendix G of the ASME Code as follows:

C* K,+Kit<K,, (2)

where, Kim = stress Intensity factor caused by membrane (pressure) stress Kit = stress Intensity factor caused by the thermal gradients Kic = function of temperature relative to the RTtmT of the material C = 2.0 for Level A and Level B service limits C = 1.5 for hydrostatic and leak test conditions during which the reactor core Is not critical At any time during the heatup or cooldown transient, Kc is determined by the metal temperature at the tip of a postulated flaw at the 114T and 314T location, the appropriate value for RTNDT, and the reference fracture toughness curve. The thermal stresses resulting from the temperature gradients through the vessel wall are calculated and then the corresponding (thermal) stress Intensity factors, Kit, for the reference flaw are computed.

From Equation 2, the pressure stress intensity factors are obtained and, from these, the allowable pressures are calculated.

SEABROOK - UNIT i B 3/44-12 Amendment No. 89

REACTOR COOLANT SYSTEM BASES 314.4.9 PRESSUREtTEMPERATURE LIMITS (Continued)

COOLDOWN For the calculation of the allowable pressure versus coolant temperature during cooldown, the reference flaw of Appendix G to the ASME Code Is assumed to exist at the Inside of the vessel wall. During cooldown, the controlling location of the flaw Is always at the inside of the wall because the thermal gradients produce tensile stresses at the Inside, which increase with Increasing cooldown rates. Allowable pressure-temperature relations are generated for both steady-state and finite cooldown rate situations. From these relations, composite limit curves are constructed for each cooldown rate of interest.

The use of the composite curve in the cooldown analysis Is necessary because control of the cooldown procedure Is based on measurement of reactor coolant temperature, whereas the limiting pressure Is actually dependent on the material temperature at the tip of the assumed flaw. During cooldown, the 114T vessel location Is at a higher temperature than the fluid adjacent to the vessel ID. This condition, of course, Is not true for the steady-state situation. It follows that at any given reactor coolant temperature, the AT developed during cooldown results in a higher value of Kic at the 114T location for finite cooldown rates than for steady-state operation. Furthermore, f conditions exist such that the Increase In Kic exceeds Kit, the calculated allowable pressure during cooldown will be greater than the steady-state value.

The above procedures are needed because there is no direct control on temperature at the I14T location; therefore, allowable pressures may unknowingly be violated Ifthe rate of cooling Is decreased at various Intervals along a cooldown ramp. The use of the composite curve eliminates this problem and assures conservative operation of the system for the entire cooldown period.

HEATUP Three separate calculations are required to determine the limit curves for finite heatup rates. As is done in the cooldown analysis, allowable pressure-temperature relationships are developed for steady-state conditions as well as finite heatup rate conditions assuming the presence of a 1/4T defect at the Inside of the vessel wall. The heatup results In compressive stresses at the inside surface of the wall that alleviate the tensile stresses produced by Internal pressure. The metal temperature at the crack tip lags the coolant temperature; therefore, the Kic for the 114T crack during heatup Is lower than the Kic for the 1/4T crack during steady-state conditions at the same coolant temperature.

SEABROOK - UNIT I B 3/4 4-13 Amendment No. 89

REACTOR COOLANT SYSTEM BASES 314.4.9 PRESSURErrEMPERATURE LIMITS (Continued)

HEATUP (Continued)

During heatup, especially at the end of the transient, conditions may exist such that the effects of compressive thermal stresses and lower K1c values for steady-state and finite heatup rates do not offset each other and the pressure-temperature curve based on steady-state conditions no longer represents a lower bound of all similar curves for finite heatup rates when the 1/4T flaw Is considered. Therefore, both cases have to be analyzed in order to assure that at any coolant temperature the lower value of the allowable pressure calculated for steady-state and finite heatup rates is obtained.

The second portion of the heatup analysis concerns the calculation of pressure-temperature limitations for the case In which a 114T flaw located at the 114T location from the outside surface Is assumed. Unlike the situation at the vessel inside surface, the thermal gradients established at the outside surface during heatup produce stresses which are tensile In nature and thus tend to reinforce any pressure stresses present. These thermal stresses are dependent on both the rate of heatup and the time (or coolant temperature) along the heatup ramp. Since the thermal stresses at the outside are tensile and Increase with Increasing heatup rate, a lower bound curve cannot be defined.

Rather, each heatup rate of Interest must be analyzed on an individual basis.

Following the generation of pressure-temperature curves for both the steady-state and finite heatup rate situations, the final limit curves are produced as follows. A composite curve Is constructed based on a point-by-point comparison of the steady-state and finite heatup rate data. At any given temperature, the allowable pressure Is taken to be the lesser of the three values taken from the curves under consideration. The use of the composite curve Is necessary to set conservative heatup limitations because It Is possible for conditions to exist wherein, over the course of the heatup ramp, the controlling condition switches from the Inside to the outside and the pressure limit must at all times be based on analysis of the most critical criterion.

Finally, the composite curves for the heatup rate data and the cooldown rate data are adjusted for possible errors In the pressure and temperature sensing Instruments by the values Indicated on the respective curves.

SEABROOK - UNIT I B 314 4-14 Amendment No. 89

REACTOR COOLANT SYSTEM BASES 3144.9 PRESSUREiTEMPERATURE LIMITS (Continued)

HEATUP (Continued) 10 CFR Part 50, Appendix G. Reference (3), addresses the metal temperature of the closure head flange and vessel flange regions. This rule states that the metal temperature of the closure flange regions must exceed the material unirradlated RTNDT by at least 120'F for normal operation when the pressure exceeds 20 percent of the preservice hydrostatic test pressure (3106 psi), which In this case Is 621 psig. The limiting unirradiated RTNDT of 30"F occurs in the vessel flange of the reactor vessel, consequently the minimum allowable temperature of this region Is 150°F at pressures greater than 621 psig. This limit is shown as the horizontal lines In Figures 3.4-2 and 3A-3. (NOTE: Figures 3.4-2 and 3A-3 include a compensation of 200F and 100 psIg for possible instrument errors.)

Although the pressurizer operates Intemperature ranges above those for which there Is reason for concern of nonductile failure, operating limits are provided to assure compatibility of operation with the fatigue analysis performed In accordance with the ASME Code requirements.

References

1. ASME Boiler and Pressure Vessel Code, Section Xl, Appendix G, "Fracture Toughness Criteria for Protection Against Failure!, dated December 1995. through 1996 Addendum.

2. ASME Boiler and Pressure Vessel Code Case N-641, Section Xl, Division 1, "Alternative Pressure-Temperature Relationship and Overpressure Protection System Requirements', dated January 17,2000.

3. 10 CFR Part 50, Appendix G, Fracture Toughness Requirements", U.S. Nuclear Regulatory Commission, Federal Register, Volume 60, No. 243, dated December 19, 1995.

4. Westinghouse WCAP-15745, Revision 0, 'Seabrook Unit 1 Heatup and Cooldown Limit Curves for Normal Operation", dated December 2001.

5. Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials", U.S. Nuclear Regulatory Commission, dated May 1988.

6. Duke Engineering and Services Report DES-NFQA-98-O1, Revision 0, 'Analysis of Seabrook Station Unit I Reactor Vessel Surveillance Capsules U and Y', dated May 1998.

SEABROOK - UNIT I B 3/4 4-14a Amendment No. 891

REACTOR COOLANT SYSTEM BASES 3/4.4.9 PRESSUREITEMPERATURELIMITS (Continued)

COLD OVERPRESSURE PROTECTION (Continued)

The OPERABILITY of two PORVs, or two RHR suction relief valves, or a combination of a PORV and RHR suction relief valve, or an RCS vent opening of at least 1.58 square inches ensures that the RCS will be protected from pressure transients which could exceed the limits of Appendix G to 10 CFR Part 50 when one or more of the RCS cold legs are less than or equal to 2900 F. Either PORV or either RHR suction relief valve has adequate l relieving capability to protect the RCS from overpressurization when the transient is limited to either. (1) the start of an Idle RCP with the secondary water temperature of the steam generator less than or equal to 50 0F above the RCS cold leg temperatures, or (2) the start of a centrifugal charging pump and Its Injection Into a water-solid RCS.

The Maximum Allowed PORV Setpoint for the Cold Overpressure Mitigation System (COMS) Is derived by analysis which models the performance of the COMS assuming various mass Input and heat Input transients. Operation with a PORV Setpoint less than or equal to the maximum Setpoint ensures that Appendix G criteria will not be violated with consideration for. (1) a maximum pressure overshoot beyond the PORV Setpoint which can occur as a result of time delays In signal processing and valve opening; (2) a 500F heat transport effect made possible by the geometrical relationship of the RHR suction line and the RCS wide range temperature indicator used for COMS; (3) Instrument uncertainties; and (4) single failure. To ensure mass and heat input transients more severe than those assumed cannot occur, Technical Specifications require both Safety injection pumps and all but one centrifugal charging pump to be made Inoperable while in MODES 4, 5, and 6 with the reactor vessel head Installed and not fully detensloned, and disallow start of an RCP if secondary coolant temperature Is more than 500F above reactor coolant temperature.

Exceptions to these requirements are acceptable as described below.

Operation above 3500F but less than 375¢F with only one centrifugal charging pump OPERABLE and no Safety Injection pumps OPERABLE Is allowed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. As shown by analysis, LOCAs occurring at low temperature, low pressure conditions can be successfully mitigated by the operation of a single centrifugal charging pump and a single RHR pump with no credit for accumulator Injection. Given the short time duration and the condition of having only one centrifugal charging pump OPERABLE and the probability of a LOCA occurring during this time, the failure of the single centrifugal charging pump Is not assumed.

Operation below 3500 F but greater than 3250F with all centrifugal charging and Safety Injection pumps OPERABLE is allowed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. During low pressure, low temperature operation all automatic Safety Injection actuation signals except Containment Pressure-High are blocked. Innormal conditions, a single failure of the SEABROOK - UNIT 1 B 3144-1 5 Amendment No. 3, 16, 74, 89