ML20148Q395
| ML20148Q395 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 04/06/1988 |
| From: | CAROLINA POWER & LIGHT CO. |
| To: | |
| Shared Package | |
| ML20148Q380 | List: |
| References | |
| NUDOCS 8804120450 | |
| Download: ML20148Q395 (7) | |
Text
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t ENCIDSURE 1 BRUNSVICK STEAM ELECTRIC PIANT, UNITS 1 AND 2 NRC DOCKETS 50-325 6 50 324 OPERATING LICENSES DPR-71 & DPR-62 BASES PAGE CORRECTION l
TECHNICAL SPECIFICATION PAGES 8804120450 DR 880406 ADOCK 05000324 PDR
I (BSEP-1-126) 3/4.4 REACTOR C00LAh~r SYSTEM BASES 3/4.4.1 RECIRCULATION SYSTEM Operation with a reactor core coolant recirculation loop inoperable is restricted until an evaluation of the performance of the ECCS during one loop operation has been performed, evaluated, and determined to be acceptable.
An inoperable jet pump is not, in itself, a sufficient reason to declare a recirculation loop inoperable, but it does piasent a hazard in case of a design basis accident by increasing the blowdown area and eliminating the capability of reflooding the core. Thus, the requirement for shutdown of the facility with a jet pump inoperable.
In ord.r to prevent undue stress on the vessel nozzles and bottom head region, the recirculation loop temperatures should be within 50 F of each other prior to start-up of an idle loop.
Since the coolant in the bottom of the vessel is at a lower temperature than the water in the upper regions of the core, undue stress on the vessel would result if the temperature difference were greater than 145 F.
Neutron flux noise limits are established to ensure early detection of limit cycle neutron flux oscillations. BWR cores typically operate with neutron flux noise caused by random boiling and flow noise. Typical neutron flux noise levels of 1 to 12% of rated power (peak-to peak) have been reported for the range of low to high recirculation loop flow during both single and dual recirculation loop operation. Neutron flux noise levels significantly larger than these values are considered in the thermal / mechanical fuel design and are found to be of negligible consequence. In addition, stability tests at operating BWR's have demonstrated that when stability related neutron flux limit cycle oscillations occur they result in peak-to peak neutron flux limit cycles 5 to 10 times the typical values. Therefore, actions taken to reduce neutron flux noise levels exceeding three (3) times the typical value are I sufficient to ensure early detection of limit cycle neutron flux oscillations.
Data to establish baseline APRM and LPRM neutron flux noise values is obtained at a point below the 100% rated rod line. A minimum of two detectors 1 of one LPRM string per core octant and two detectors of one LPRM string near the center of the core should be monitored. Detectors used for monitoring should be selected to provide core wide representation. Substitutions are permitted for inoperable LPRM detectors.
These specifications are based on the guidance of General Electric SIL #380, Rev. 1, 2-10-84.
3/4.4.2 SAFETY / RELIEF VALVES The reactor conlant system safety valve function of the safety-relief valves operates to prevent the system from being pressurized above the Safety Limit of 1325 psig. The system is designed to meet the requirements of the BRUNSWICK - UNIT 1 B 3/4 4-1 Amendment No.
. (BSEP-1-136)
REACTOR COOLANT SYSTEM BASES SAFETY / RELIEF VALVES (Continued)
ASME Boiler and Pressure Vessel Code Section III for the pressure vessel and ANSI B31.1, 1975 Code for the reactor coolant system piping.
3/4.4.3 REACTORC00LAhiTSYSTEdLEAKACE 3/4.4.3.1 LEAKACE DETECTION SYSTEMS The RCS leakage detection systems required by this specitication are provided to monitor and detect leakage from the Reactor Coolant Pressure Boundary. These detection systems are consistent with the recommendations of Regulatory Guide 1.45, "Reactor Coolant Pressure Boundary Leakage Detection Systems."
3/4.4.3.2 OPERATIONAL LEAKAGE The allowable leakage rates of conlant from the reactor coolint system have been based on the predicted and experimentally observed behavior of cracks in pipes. The normslly expected background leakage due to equipment design and the detection capability of the instrumentation for determining system leakage was also considered. The evidence obtained from experiments suggests that for leakage somewhat greater than that specified for unidentified leakage, the probability is small that the imperfection or crack associated with such leakage would grow rapidly, However, in all cases, if the leakage rates exceed the values specified or the leakage is located and l known to be PRESSURE BOUNDARY LEAKACE, the reactor will be shut down to allow I further investigation and corrective action.
3/4.4.4 CHEMISTRY i l
The reactor water chemistry limits are established to prevent damage to the reactor materials in contact with the coolant. Chloride limits are specified to prevent stress corrosion cracking of the stainless steel. The effect of chloride is not as great when the oxygen concentration in the coolant is low; thus, the higher limit on chlorides is permitted during full power operation. During shutdown and refueling operations, the temperature necessary for stress corrosion to occur is not present.
Conductivity measurements are required on a continuous basis since changes in this parameter are an indication of abnormal conditions. When the conductivity is within limits, the pH, chlorides, and other impurities affecting conductivity must also be within their acceptable limits. With the conductivity outside the limits, additional samples must be examined to ensure that the chlorides are not exceeding the limits.
The surveillance requirements provide adequate assurance that concentrations in excess of the limits will be detect (3 in sufficient time to take corrective action.
BRUNSWICK - UNIT 1 B 3/4 4-2 Amendment No.
(BSEP-1-126)
REACTOR COOLANT SYSTEM BASES 3/4.4.5 SPECIFIC ACTIVITY The limitations on the specific activity of the primary coolant ensure
' hat the 2-hour thyroid and whole body doses resulting from a main steam line failure outside the containment during steady state operation will not exceed small fractions of the dose guidelines of 10 CFR 100. Permitting operation to continue for limited time periods with higher specific activity levels accommodates short-term iodine spikes which may be associated with power level changes, and is based on the fact that a steam line failure during these short time periods is considerably less likely. Operation at the higher activity levels, therefore, is restricted to a small fraction of the unit's total operating time. The upper limit of coolant iodine corcentration during short-term iodine spikes ensures that the thyroid dose from a steam line failure will not exceed 10 CFR Part 100 dose guidelines.
Information obtained on iodine spiking will be used to assess the parameters associated with spiking phenomena. A reduction in frequency of isotopic analysis following power changes may be permissible if justified by the data obtained.
Closing the main steam line isolation valves prevents the release of activity to the environs should the steam line rupture occur. The surveillance cequt-ements provide adequate assurance rSat excessive specific activity levels in tne reactor coolant will be detected in sufficient time to take corrective action.
3/4.4.6 PRESSURE / TEMPERATURE LIMITS l
All components in the Reactor Coolant System are designed to withstand the effects of cyclic loads due to system temperature and pressure changes. These i cyclic loads are introduced by normal load transients, reactor trips, and !
start-up and shutdown operations. The various categories of load cycles used for design purposes are provided in Section 4.2 of the PSAR. During start-up and shutdown, the rates of temperature and pressure changes are limited so that the maximum specified heatup and cooldown rates are. consistent with the design assumptions and satisfy the stress limits for cyclic operation.
During heatup, the thermal gradients in the reactor vessel wall produce thermal stresses which vary from compressive at the inner wall to tensile at the outer wall. These thermal-induced compressive stresses tend to alleviate l
the tensile stresses induced by the internal pressure. Therefore, a pressure- l temperature curve based on steady sta;e conditions, i.e., no thermal stresses, I represents a lower bound of all similar curves for finite heatup rates when l the inner wall of the vessel is treated as the governing location.
l l
BRUNSWICK - UNIT 1 B 3/4 4-3 Amendmer.t No.
1 (BSEP-2-133) 3/4.4 REACTOR COOLANT SYSTEM BASES 3/4.4.1 RECIRCULATION SYSTEM Operation with a reactor core coolant recirculation loop inoperable is restricted until an evaluation of the performance of the ECCS during one loop operation has been perforaed, evaluated, and determined to be acceptable.
An inoperable jet pmp is not, in itself, a sufficient reason to declare a recirculation loop inoperable, but it does present a hazard in case of a design basis accident by increasing the blowdown area and eliminating the capability of reflood.ng the core. Thus, the requirement for shutdown of the facility with a jet ptmp inoperable. I In order to prevent undue stress on the vessel nozzles and bottom head region, the recirculat un loop temperatures should be within 50 F of each other prior to start-up of an idle loop.
l 1
Since the coolant in the bottom of the vessel is at a lower temperature than the water in the upper regions of the core, undue stress on the vessel would result if the temperature difference were greater than 145 F.
1 Neutron flux noise limits are established to ensure early detection of i limit cycle neutron flux oscillations. BWR cores typically operate with neutron flux noise caused by random boiling end flow noise. Typical neutron flux noise levels of 1 to 12% of rated power (peak-to peak) have been reported for the range of low to high recirculation loop flow during both single and j dual recirculation loop operation. Neutron flux noise levels significantly larger than these values are considered in the thermal / mechanical fuel design and are found to be of negligible consequence. In addition, stability tests at operating BWR's have demonstrated that when stability related neutron flux limit cycle oscillations occur they result in peak-to peak neutron flux limit I cycles 5 to 10 times the typical values. Therefore, actions taken to reduce I neutron flux noise levels exceeding three (3) times the typical value are I sufficient to ensure early detection of limit cycle neutron flux oscillations. 1 Data to establish baseline APRM and LPRM neutron f'.ux noise values 3.s obtained at a point below the 100% rated rod line. A minimum of two detectors of one LPRM string per core octant and two detectors of one LPRM string near the center of the core should be me:titored. Detectors used for monitoring should be selected to provide core wide representation. Substitutions are permitted for inoperable LPRM detectors.
These specifications are based on the guidance of General Electric SIL #380, Rev. 1, 2-10-84.
3/4.4.2 SAFETY / RELIEF VALVES The reactor coolant system safety valve function of the safety-relief valves operate to prevent the system from being pressurized above the Safety Limit of 1325 psig. The system is designed to meet the requirements of the BRUNSWICK - UNIT 2 B 3/4 4-1 Amendment No.
- (BSEP-2-333)
REACTOR COOLANT SYSTEM BASES SAFETY / RELIEF VALVES (Continued)
ASME Boiler and Pressure Vessel Code Section III for the pressure vessel and ANSI B31.1, 1967, code for the reactor coolant system piping.
3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE 3/4.4.3.1 LEAKACE DETECTION SYSTEMS The RCS leakage detection systems required by this specification are provided to monitor and detect leakage from the Reactor Coolant Pressure Boundary. These detection systems are consistent with the recommendations of Regulatory Guide L.45, "Reactor Coolant Pressure Boundary Leakage Detection Systems."
3/4.4.3.2 OPERATIONAL LEAKACE The allowable leakage rates of coolant from the reactor coolant system have been based on the predicted and experimentally observed behavior of cracks in pipes. The normally expected background leakage due to equipment design and the detection capability of the instrumentation for determining system leakage was also considered. The evidence obtained from experiments suggests that for leakage somewhat greater than that specified for unidentified leakage, the probability is small that the imperfection or crack associated with such leakage would grow rapidly, However, in all cases, if the leakage rates exceed the values specified or the leakage is located and known to be PRESSURE BOUNDARY LEAKACE, the reactor will be shut down to allow further investigation and corrective action. ,
i 3/4.4.4 CHEMISTRY The reactor water chemistry limits are establiched to prevent damage to I the reactor materials in contact with the coolant. Chloride limits are specified to prevent stress corrosion cracking of the stainless steel. The 1 effect of chloride is not as great when the oxygen concentration in the I coolant is low; thus, the higher limit on chlorides is permitted during full I power operation. During shutdown and refueling operations, the temperature I necessary for stress corrosion to occur is not present. )
l Conductivity measurements are required on a continuous basis since changes 1 in this parameter are an indication of abnormal conditions. When the l conductivity is within limits, the pH, chlorides, and other impurities ;
affecting conductivity must also be within their acceptable limits. With the l conductivity outside the limits, additional samples must be examined to ensure l that the chlorides are not exceeding the limits. I The surveillance requirements provide adequate assurance that I concentrations in excess of the limits will be detected in sufficient time to take corrective action.
BRUNSWICK - UNIT 2 B 3/4 4-2 Amendment No.
(BSEP-3-133)
REACTOR COOLANT SYSTEM i BASES 3/4.4.5 SPECIFIC ACTIVITY The limitations on the specific activity of the primary coolant ensure that the 2-hour thyroid and whole body doses resulting from a main steam line failure outside the containment during steady state will not exceed small I fractions of the dose guidelines in 10CFR100. Permitting operation to continue for limited time periods with higher specific activity levels accomodates short-term iodine spikes which may be associated with power level changes, and is based on the fact that a steam line failure during these short time periods is considerably less likely. Operation at the higher activity levels, therefore, is restricted to a small fraction of the unit's total operating time. The upper limit of coolant iodine concentration during short-term iodine spikes ensures that the thyriod dose from a stema line failure will not exceed 10 CFR Part 100 dose guidelines.
Information obtained on iodine spiking will be used to assess the !
parameters associated with spiking phenomena. A reduction in frequency of isotopic analysis following power changes may be permissible, if justified by the data obtained. l l
Closing the main steam line isolation valves prevents the release of l activity to the environs should the steam line rupture occur. The l survelliance requirements provide adequate assurance that excessive specific i activity levels in the reactor coolant will be detected in sufficient time to take corrective action. 1 i
3/4.4.6 PRESSURE / TEMPERATURE LIMITS I All components in the Reactor Coolant System are designed to withstand the i effects of cyclic loads due to system temperature and pressure changes. These i cyclic loads are introduced by normal load transients, reactor trips, and start-up and shutdown operations. The various categories of load cycles used for design purposes are provided in Section 4.2 of the FSAR. During start-up and shutdown, the rates of temperature and pressure changes are limited so that the maximum specified heatup and cooldown rates are consistent with the design assumptions and satisfy the stress limits for cyclic operation.
During heatup, the thermal gradients in the reactor vessel vall produce thermal stresses which vary from compressive at the inner wall to tensile at the outer wall. Th2se thermally induced compressive stresses tend to alleviate the tensile stresses induced by the internal pressure. Therefore, a pressure-temperature curve based on steady state conditions, i.e., no thermal stresses, represents a lower bound of all similar curves for finite heatup rates when the inner wall of the vessel is treated as the governing location.
l l
l BRUNSWICK - UNIT 2 B 3/4 4-3 Amendment No.
l