ML020310364
| ML020310364 | |
| Person / Time | |
|---|---|
| Site: | Kewaunee  |
| Issue date: | 01/03/2002 |
| From: | Webb T Nuclear Management Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| Download: ML020310364 (18) | |
Text
NRC-02-003 Kewaunee Nuclear Power Plant Point Beach Nuclear Plant N490 Highway 42 6610 Nuclear Road Kewaunee, WI 54216-9511 Two Rivers, WI 54241 920.388.2560 920.755.2321 Committed to Nuclear Excellence Kewaunee / Point Beach Nuclear Operated by Nuclear Management Company, LLC January 3, 2002 10 CFR 50.36 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C. 20555 Ladies/Gentlemen:
Docket 50-305 Operating License DPR-43 Kewaunee Nuclear Power Plant BASES REVISIONS TO THE KEWAUNEE NUCLEAR POWER PLANT TECHNICAL SPECIFICATIONS Nuclear Management Company (NMC), licensee for the Kewaunee Nuclear Power Plant (KNPP),
hereby submits a revision to the Bases for the Technical Specifications (TS). The changes are as follows:
" TS Basis Section 3.1 and 3.4 - Changed word processing software from WordPerfect to Word and incorporated minor administrative changes for clarity and uniformity.
" TS Basis Sections 3.1 and 3.4 -Updated the TS Basis for TS 3.1.c and TS 3.4.d to describe the intent of the specification and referenced the KNPP Updated Safety Analysis Report (USAR) for the actual analysis details.
These changes have been screened for evaluation pursuant to the requirements of 10 CFR 50.59 in accordance with approved KNPP procedures and were determined to be acceptable.
Attached is a copy of the revised TS Bases pages for your controlled Technical Specifications.
Sincerely, Thomas J. Webb Site Licensing Director GOR Attachments cc - NRC Regional Administrator ] )
NRC Resident Inspector Electric Division, PSCW
BASIS - Operational Components (TS 3.1.a)
Reactor Coolant Pumps (TS 3. .a. 1)
When the boron concentration of the Reactor Coolant System is to be reduced, the process must be uniform to prevent sudden reactivity changes in the reactor. Mixing of the reactor coolant will be sufficient to maintain a uniform boron concentration if at least one reactor coolant pump or one residual heat removal pump is running while the change is taking place. The residual heat removal pump will circulate the equivalent of the primary system volume in approximately one-half hour.
Part 1 of the specification requires that both reactor coolant pumps be OPERATING when the reactor is in power operation to provide core cooling. Planned power operation with one loop out-of service is not allowed in the present design because the system does not meet the single failure (locked rotor) criteria requirement for this mode of operation. The flow provided in each case in Part 1 will keep DNBR well above 1.30. Therefore, cladding damage and release of fission products to the reactor coolant will not occur. One pump operation is not permitted except for tests. Upon loss of one pump below 10% full power, the core power shall be reduced to a level below the maximum power determined for zero power testing. Natural circulation can remove decay heat up to 10%
power. Above 10% power, an automatic reactor trip will occur if flow from either pump is lost."'
The RCS will be protected against exceeding the design basis of the LTOP system by restricting the starting of a RXCP to when the secondary water temperature of each SG is < 100°F above each RCS cold leg temperature. The restriction on starting a reactor coolant pump (RXCP) when one or more RCS cold leg temperatures is < 200'F is provided to prevent a RCS pressure transient, caused by an energy addition from the secondary system, which could exceed the design basis of the low temperature overpressure protection (LTOP) system.
Decay Heat Removal Capabilities (TS 3.1.a.2)
When the average reactor coolant temperature is < 3"50'F a combination of the available heat sinks is sufficient to remove the decay heat and provide the necessary redundancy to meet the single failure criterion.
When the average reactor coolant temperature is < 200 0 F, the plant is in a COLD SHUTDOWN condition and there is a negligible amount of sensible heat energy stored in the Reactor Coolant System. Should one residual heat removal train become inoperable under these conditions, the remaining train is capable of removing all of the decay heat being generated.
1
'IUSAR Section 7.2.2 TS B3.1-1 12/17/01
The requirement for at least one train of residual heat removal when in the REFUELING MODE is to ensure sufficient cooling capacity is available to remove decay heat and maintain the water in the reactor vessel < 140'F. The requirement to have two trains of residual heat removal operable when there is < 23 feet of water above the reactor vessel flange ensures that a single failure will not result in complete loss-of-heat removal capabilities. With the reactor vessel head removed and at least 23 feet of water above the vessel flange, a large heat sink is available. In the event of a failure of the operable train, additional time is available to initiate alternate core cooling procedures.
Pressurizer Safety Valves (TS 3.1 .a.3)
Each of the pressurizer safety valves is designed to relieve 325,000 lbs. per hour of saturated steam at its setpoint. Below 350'F and 350 psig, the Residual Heat Removal System can remove decay heat and thereby control system temperature and pressure. If no residual heat were removed by any of the means available, the amount of steam which could be generated at safety valve relief pressure would be less than half the valves' capacity. One valve therefore provides adequate protection against overpressurization.
Pressure Isolation Valves (TS 3.1 .a.4)
The Basis for the Pressure Isolation Valves is discussed in the Reactor Safety Study (RSS),
WASH-1400, and identifies an intersystem loss-of-coolant accident in a PWR which is a significant contributor to risk from core melt accidents (EVENT V). The design examined in the RSS contained two in-series check valves isolating the high pressure Primary Coolant System from the Low Pressure Injection System (LPIS) piping. The scenario which leads to the EVENT V accident is initiated by the failure of these check valves to function as a pressure isolation barrier. This causes an overpressurization and rupture of the LPIS low pressure piping which results in a LOCA that bypasses containment.!2 )
PORVs and PORV Block Valves (TS 3.l.a.5)
The pressurizer power-operated relief valves (PORVs) operate as part of the Pressurizer Pressure Control System. They are intended to relieve RCS pressure below the setting of the code safety valves. These relief valves have remotely operated block valves to provide a positive shutoff capability should a PORV become inoperable.
The pressurizer PORVs and associated block valves must be OPERABLE to provide an alternate means of mitigating a design basis steam generator tube rupture. Thus, an inoperable PORV (for reasons other than seat leakage) or block valve is not permitted in the HOT STANDBY and OPERATING MODES for periods of more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
(2)Order for Modification of License dated 4/20/81 TS B3.1-2 12/17/01
Pressurizer Heaters (TS 3.1 .a.6)
Pressurizer heaters are vital elements in the operation of the pressurizer which is necessary to maintain system pressure. Loss of energy to the heaters would result in the inability to maintain system pressure via heat addition to the pressurizer. Hot functional tests(3 ýhave indicated that one group of heaters is required to overcome ambient heat losses. Placing heaters necessary to overcome ambient heat losses on emergency power will ensure the ability to maintain pressurizer pressure Surveillance tests are performed to ensure heater OPERABILITY.
Reactor Coolant Vent System (TS 3.1 .a.7)
The function of the High Point Vent System is to vent noncondensible gases from the high points of the RCS to ensure that core cooling during natural circulation will not be inhibited. The OPERABILITY of at least one vent path from both the reactor vessel head and pressurizer steam space ensures the capability exists to perform this function.
The vent path from the reactor vessel head and the vent path from the pressurizer each contain two independently emergency powered, energize to open, valves in parallel and connect to a common header that discharges either to the containment atmosphere or to the pressurizer relief tank. The lines to the containment atmosphere and pressurizer relief tank each contain an independently emergency powered, energize to open, isolation valve. This redundancy provides protection from the failure of a single vent path valve rendering an entire vent path inoperable.
A flow restriction orifice in each vent path limits the flow from an inadvertent actuation of the vent system to less than the flow capacity of one charging pump.(4)
(3 Hot functional test (PT-RC-3 1)
(4) Letter from E. R. Mathews to S. A. Varga dated 5/21/82 TS B3.1-3 12/17/01
Heatup and Cooldown Limit Curves for Normal Operation (TS 3.1.b)
Fracture Toughness Properties (TS 3.1 .b. 1)
The fracture toughness properties of the ferritic material in the reactor coolant pressure boundary are 5
determined in accordance with the ASME Boiler and Pressure Vessel Code," ) and the calculation methods of Footnote.(6) The postirradiation fracture toughness properties of the reactor vessel belt line material were obtained directly from the Kewaunee Reactor Vessel Material Surveillance Program.
Allowable pressure-temperature relationships for various heatup and cooldown rates are calculated using methods derived from Appendix G in Section III of the ASME Boiler and Pressure Vessel Code, and are discussed in detail in Footnote.(7)
The method specifies that the allowable total stress intensity factor (K,) at any time during heatup or cooldown cannot be greater than that shown on the KIR curve for the metal temperature at that time. Furthermore, the approach applies an explicit safety factor of 2.0 on the stress intensity factor induced by the pressure gradient. Thus, the governing equation for the heatup-cooldown analysis is:
2 KIM + Kit <ý KIR (3.1b-l) where KIm is the stress intensity factor caused by membrane (pressure) stress K1 , is the stress intensity factor caused by the thermal gradients KIR is provided by the Code as a function of temperature relative to the RT.DT of the material.
(5)Section III and XI of the ASME Boiler and Pressure Vessel Code, Appendix G, "Protection Against Non-ductile Failure."
(6) Standard Method for Measuring Thermal Neutron Flux by Radioactivation Techniques, astm designation E262 86.
(7) WCAP-14278, Revision 1, "Kewaunee Heatup and Cooldown Limit Curves for Normal Operation," T. Laubham and C. Kim, September 1998.
TS B3.1-4 12/17'01
From equation (3.lb-1) the variables that affect the heatup and cooldown analysis can be readily identified. K,,, is the stress intensity factor due to membrane (pressure) stress. K,, is the thermal (bending) stress intensity factor and accounts for the linearly varying stress in the vessel wall due to thermal gradients. During heatup Ki, is negative on the inside and positive on the outer surface of the vessel wall. The signs are reversed for cooldown and, therefore, an ID or an OD one quarter thickness surface flaw is postulated in whichever location is more limiting. KR is dependent on irradiation and temperature and, therefore, the fluence profile through the reactor vessel wall and the rates of heatup and cooldown are important. The heatup and cooldown limit curves have been developed by combining the most conservative pressure temperature limits derived by using material properties of the intermediate forging, closure head flange, and beltline circumferential weld to form a single set of composite curves. Details of the procedure used to account for these variables are explained in the following text.
Following the generation of pressure-temperature curves for both the steady-state (zero rate of change of temperature) and finite heatup rate situations, the final limit curves are produced in the following fashion. First, a composite curve is constructed based on a point-by-point comparison of the steady-state and finite heatup rate data for each of the limiting materials. At any given temperature, the allowable pressure is taken to be the lesser of the values taken from the curves under consideration. The composite curve is then adjusted to allow for possible errors in the pressure and temperature sensing instruments including the pressure difference between the gage and beltline weld.
The use of the composite curve is mandatory in setting heatup limitations because it is possible for conditions to exist such that over the course of the heatup ramp the controlling analysis switches from the OD to the ID location. The pressure limit must, at all times, be based on the most conservative case.
The cooldown analysis proceeds in the same fashion as that for heatup with the exception that the controlling location is always at the ID. The thermal gradients induced during cooldown tend to produce tensile stresses at the ID location and compressive stresses at the OD position. Thus, the ID flaw is clearly the worst case.
As in the case of heatup, allowable pressure-temperature relations are generated for both steady-state and finite cooldown rate situations for each of the limiting materials. Composite limit curves are then constructed for each cooldown rate of interest. Again adjustments are made to account for pressure and temperature instrumentation error.
The use of the composite curve in the cooldown analysis is necessary because system control is based on a measurement of reactor coolant temperature, whereas the limiting pressure is calculated using the material temperature at the tip of the assumed reference flaw. During cooldown, the 1/4T 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 the AT induced during cooldown results in a calculated higher KIR for finite cooldown rates than for steady-state under certain conditions.
TS B3.1-5 12/17/01
Limit curves for normal heatup and cooldown of the primary Reactor Coolant System have been calculated using the methods discussed above and limited application to ASME Boiler and Pressure Vessel Code Case N-588 to the circumferential beltline weld. The derivation of the limit curves is t8 consistent with the NRC Regulatory Standard Review Plan ) and Footnotei9)
Transition temperature shifts occurring in the pressure vessel materials due to radiation exposure have been obtained directly from the reactor pressure vessel surveillance program. As presented in WCAP 14279, Revision 1,"1°) weld metal Charpy test specimens from Capsule S indicate that the core region weld metal exhibits the largest shift in RTNDT (250'F).
The results of Irradiation Capsules V, R, P, and S analyses are presented in WCAP 8908,(") WCAP 9878,"2) WCAP-12020,*13) WCAP-14279,'14) and WCAP-14279, Revision 1°)respectively. Heatup and cooldown limit curves for normal operation of the reactor vessel are presented in Figures TS 3.1-1 and TS 3.1-2 and represent an operational time period of 33(1] effective full-power years.
The isothermal cooldown limit curve (Figure TS 3.1-2) is used for evaluation of low temperature overpressure protection (LTOP) events. This curve is applicable for 331!1 effective full-power years of fluence (through the end of OPERATING cycle 33[lI). If a low temperature overpressure event occurred, the RCS pressure transient would be evaluated to the limits of this figure to verify the integrity of the reactor vessel. If these limits are not exceeded, vessel integrity is assured and a TS violation has not occurred.
Note:
ý11 Although the curves were developed for 33 EFPY, they are limited to 28 EFPY (corresponding to the end of cycle 28) by WPSC Letter NRC-99-017.
""8 '"Fracture Toughness Requirements," Branch Technical Position MTEB 5-2, Chapter 5.3.2 in Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, LWR Edition, NUREG-0800, 1981.
9)1989 ASME Boiler and Pressure Vessel (B&PV) Code,Section XI, Appendix G, "Fracture Toughness Criteria for Protection Against Failure."
1")C. Kim, et al., "Evaluation of Capsule S from the Kewaunee and Capsule A35 from the Maine Yankee Nuclear Power Reactor Vessel Radiation Surveillance Programs," WCAP-14279, Revision 1, September 1998.
(I) S.E. Yanichko, S. L. Anderson, and K. V. Scott, "Analysis of Capsule V from the Wisconsin Public Service Corporation Kewaunee Nuclear Plant Reactor Vessel Radiation Surveillance Program," WCAP 8908, January 1977.
(12) S.E. Yanichko, et al., "Analysis of Capsule R from the Wisconsin Public Service Corporation Kewaunee Nuclear Plant Reactor Vessel Radiation Surveillance Program," WCAP 9878, March 1981.
(3) S.E. Yanichko, et al., "Analysis of Capsule P from the Wisconsin Public Service Corporation Kewaunee Nuclear Power Plant Reactor Vessel Radiation Surveillance Program," WCAP-12020, November 1988.
(4) E. Terek, et al., "Analysis of Capsule S from the Wisconsin Public Service Corporation Kewaunee Nuclear Power Plant Reactor Vessel Radiation Surveillance Program," WCAP-14279, March 1995.
TS B3.1-6 12/17/01
Pressurizer Limits (TS 3.1 .b.3)
Although the pressurizer operates at temperature ranges above those for which there is reason for concern about brittle fracture, OPERATING limits are provided to ensure compatibility of operation with the fatigue analysis performed in accordance with Code requirements. In-plant testing and calculations have shown that a pressurizer heatup rate of 100°F/hr cannot be achieved with the installed equipment.
Low Temperature Overpressure Protection (TS 3.1 .b.4)
The Low Temperature Overpressure Protection system must be OPERABLE during startup and shutdown conditions below the enable temperature (i.e., low temperature) as defined in Branch Technical Position RSB 5-2 as modified by ASME Boiler and Pressure Vessel Code Case N-514.
Based on the Kewaunee Appendix G LTOP protection pressure-temperature limits calculated through 331 effective full-power years, the LTOP System must be OPERABLE whenever one or more of the RCS cold leg temperatures are
- 200'F and the head is on the reactor vessel. The LTOP system is considered OPERABLE when all four valves on the RHR suction piping (valves RHIR- IA, 1B, 2A, 2B) are open and valve RHR-33-1, the LTOP valve, is able to relieve RCS overpressure events without violating Figure TS 3.1-2.
The set pressure specified in TS 3.1 .b.4 includes consideration for the opening pressure tolerance of+/- 3% (+/- 15 psig) as defined in ASME Boiler and Pressure Vessel Code,Section III, Division 1, Subsection NC: Class 2 Components for Safety Relief Valves. The analysis of pressure transient conditions has demonstrated acceptable relieving capability at the upper tolerance limit of 515 psig.
If one train of RHR suction piping to RHR 33-1 is isolated, the valves and valve breakers in the other train shall be verified open, and the isolated flowpath must be restored within five days. If the isolated flowpath cannot be restored within five days, the RCS must be depressurized and vented through at least a 6.4 square inch vent within an additional eight hours.
If both trains of RHR suction are isolated or valve RHR 33-1 is inoperable, the system can still be considered OPERABLE if an alternate vent path is provided which has the same or greater effective flow cross section as the LTOP safety valve (_ 6.4 square inches). If vent path is provided by physical openings in the RCS pressure boundary (e.g., removal of pressurizer safety valves or steam generator manways), the vent path is considered secured in the open position.
Note t'J Although the curves were developed for 33 EFPY, they are limited to 28 EFPY (corresponding to the end of cycle 28) by WPSC Letter NRC-99-017.
TS B3.1-7 12/17/01
Maximum Coolant Activity (TS 3.1.c)
The maximum dose to the thyroid and whole body that an individual may receive following an accident is specified in GDC 19 and 10 CFR 100. The limits on maximum coolant activity ensure that the calculated doses are held to the limits specified in GDC 19 and to a fraction of the 10 CFR 100 limits.
The Reactor Coolant Specific Activity is limited to < 0.2 ptCi/gram DOSE EQUIVALENT 1-131 to ensure the thyroid dose does not exceed the GDC-19 and 10 CFR 100 guidelines. The applicable accidents identified in the USAR!'"5 are analyzed assuming an RCS activity of 0.2 ýtCi/gram DOSE EQUIVALENT 1-131 incorporating an accident initiated iodine spike when required. To ensure the conditions allowed by Figure TS 3.1-3 are taken into account, the applicable accidents are also analyzed considering a pre-existing iodine spike based on Figure TS 3.1-3. The results obtained from these analyses indicate that the control room and off-site thyroid dose are within the acceptance criteria of GDC-19 and a fraction of 10 CFR 100 limits.
The Reactor Coolant Specific Activity is also limited to a gross activity of<9C_ Again the E cc 91 pCi accidents under consideration are analyzed assuming a gross activity of 9= -c The results obtained E cc from these analyses indicate the control room and off-site whole body dose are within the acceptance criteria of GDC-19 and a small fraction of 10 CFR 100 limits.
The action of reducing average reactor coolant temperature to < 500'F prevents the release of activity should a steam generator tube rupture occur since the saturation pressure of the reactor coolant is below the lift pressure of the main steam safety valves. The surveillance requirements provide adequate assurance that excessive specific activity levels in the reactor coolant will be detected in sufficient time to take corrective action.
(I) USAR Section 14.0 TS B3.1-8 12/17/01
Leakage of Reactor Coolant (TS 3.1.d) 6 )
TS (TS 3.1.d. 1)
Leakage from the Reactor Coolant System is collected in the containment or by the other closed systems. These closed systems are: the Steam and Feedwater System, the Waste Disposal System and the Component Cooling System. Assuming the existence of the maximum allowable activity in the reactor coolant, the rate of 1 gpm unidentified leakage would not exceed the limits of 10 CFR Part 20. This is shown as follows:
If the reactor coolant activity is 91/E pCi/cc (E = average beta plus gamma energy per disintegration in Mev) and I gpm of leakage is assumed to be discharged through the air ejector, or through the Component Cooling System vent line, the yearly whole body dose resulting from this activity at the SITE BOUNDARY, using an annual average X/Q = 2.0 x 106 sec/m3 , is 0.09 rem/yr, compared with the 10 CFR Part 20 limits of 0.1 rem/yr.
With the limiting reactor coolant activity and assuming initiation of a I gpm leak from the Reactor Coolant System to the Component Cooling System, the radiation monitor in the component cooling pump inlet header would annunciate in the control room. Operators would then investigate the source of the leak and take actions necessary to isolate it. Should the leak result in a continuous discharge to the atmosphere via the component cooling surge tank and waste holdup tank, the resultant dose rate at the SITE BOUNDARY would be 0.09 rem/yr as given above.
Leakage directly into the containment indicates the possibility of a breach in the coolant envelope.
The limitation of 1 gpm for an unidentified source of leakage is sufficiently above the minimum detectable leak rate to provide a reliable indication of leakage, and is well below the capacity of one charging pump (60 gpm).
Twelve hours of operation before placing the reactor in the HOT SHUTDOWN condition are required to provide adequate time for determining whether the leak is into the containment or into one of the closed systems and to identify the leakage source.
TS 3.1.d.2 Limiting the leakage through any single steam generator to 150 gpd ensures that tube integrity is maintained during a design basis main steam line break or loss-of-coolant accident. Remaining within this leakage rate provides reasonable assurance that no single tube-flaw will sufficiently enlarge to create a steam generator tube rupture as a result of stresses caused by a LOCA or a main steam line break accident within the time allowed for detection of the accident condition and resulting commencement of plant shutdown. This operational leakage rate is less than the condition assumed in design basis safety analyses and conforms to industry standards established by the Nuclear Energy Institute through its NEI 97-06, "Generic Steam Generator Program Guidelines."
(6) USAR Sections 6.5, 11.2.3, 14.2.4 TS B3.1-9 12/17/01
TS 3.1 .d.3 When the source of leakage has been identified, the situation can be evaluated to determine if operation can safely continue. This evaluation will be performed by the plant operating staff and will be documented in writing and approved by either the Plant Manager or his designated alternate.
Under these conditions, an allowable Reactor Coolant System leak rate of 10 gpm has been established. This explained leak rate of 10 gpm is within the capacity of one charging pump as well as being equal to the capacity of the Steam Generator Blowdown Treatment System.
TS 3.1.d.4 The provision pertaining to a non-isolable fault in a Reactor Coolant System component is not intended to cover steam generator tube leaks, valve bonnets, packings, instrument fittings, or similar primary system boundaries not indicative of major component exterior wall leakage.
TS 3.l.d.5 If leakage is to the containment, it may be identified by one or more of the following methods:
A. The containment air particulate monitor is sensitive to low leak rates. The rates of reactor coolant leakage to which the instrument is sensitive are dependent upon the presence of corrosion product activity.
B. The containment radiogas monitor is less sensitive and is used as a backup to the air particulate monitor. The sensitivity range of the instrument is approximately 2 gpm to > 10 gpm.
C. Humidity detection provides a backup to A. and B. The sensitivity range of the instrumentation is from approximately 2 gpm to 10 gpm.
D. A leakage detection system is provided which determines leakage losses from all water and steam systems within the containment. This system collects and measures moisture condensed from the containment atmosphere by fancoils of the Containment Air Cooling System and thus provides a dependable and accurate means of measuring integrated total leakage, including leaks from the cooling coils themselves which are part of the containment boundary. The fancoil units drain to the containment sump, and all leakage collected by the containment sump will be pumped to the waste holdup tank. Pump running time will be monitored in the control room to indicate the quantity of leakage accumulated.
If leakage is to another closed system, it will be detected by the area and process radiation monitors and/or inventory control.
TS B3.1-10 12/17/01
Maximum Reactor Coolant Oxygen, Chloride and Fluoride Concentration (TS 3.1.e)
By maintaining the oxygen, chloride and fluoride concentrations in the reactor coolant below the limits as specified in TS 3.1.e.1 and TS 3.1.e.4, 17) the integrity of the Reactor Coolant System is ensured under all OPERATING conditions.(
If these limits are exceeded, measures can be taken to correct the condition, e.g., replacement of ion exchange resin or adjustment of the hydrogen concentration in the volume control tank.""8 Because of the time-dependent nature of any adverse effects arising from oxygen, chloride, and fluoride concentration in excess of the limits, it is unnecessary to shut down immediately since the condition can be corrected. Thus, the time periods for corrective action to restore concentrations within the limits have been established. If the corrective action has not been effective at the end of the time period, reactor cooldown will be initiated and corrective action will continue.
The effects of contaminants in the reactor coolant are temperature dependent. The reactor may be restarted and operation resumed if the maximum concentration of any of the contaminants did not exceed the permitted transient values; otherwise a safety review by the Plant Operations Review Committee is required before startup.
Minimum Conditions for Criticality (TS 3.1.f)
During the early part of the fuel cycle, the moderator temperature coefficient may be calculated to be positive at _<60% RATED POWER. The moderator coefficient will be most positive at the beginning of life of the fuel cycle, when the boron concentration in the coolant is greatest. Later in the fuel cycle, the boron concentrations in the coolant will be lower and the moderator coefficients either will be less positive or will be negative.(9)
The requirement that the reactor is not to be made critical except as specified in TS 3.1 .f.1 provides increased assurance that the proper relationship between reactor coolant pressure and temperature will be maintained during system heatup and pressurization whenever the reactor vessel is in the nil-ductility temperature range. Heatup to this temperature will be accomplished by operating the reactor coolant pumps and by the pressurizer heaters.
The shutdown margin specified in TS 3.10 precludes the possibility of accidental criticality as a result of an increase in moderator temperature or a decrease in coolant pressure.(19)
The requirement that the pressurizer is partly voided when the reactor is < 1% subcritical ensures that the Reactor Coolant System will not be solid when criticality is achieved.
1 USAR Section 4.2 o USAR Section 9.2 (19)USAR Section 3.2.1 TS B3.1-11 12/17/01
The requirement that the reactor is not to be made critical when the moderator coefficient is > 5.0 pcm/°F has been imposed to prevent any unexpected power excursion during normal operation as a result of either an increase in moderator temperature or a decrease in coolant pressure. The moderator temperature coefficient limits are required to maintain plant operation within the assumptions contained in the USAR analyses. Having an initial moderator temperature coefficient no greater than 5.0 pcm/°F provides reasonable assurance that the moderator temperature coefficient will be negative at 60% rated thermal power. The moderator temperature coefficient requirement is waived during low power physics tests to permit measurement of reactor moderator coefficient and other physics design parameters of interest. During physics tests, special OPERATING precautions will be taken. In addition, the strong negative Doppler coefficientl 02 and the small integrated Ak/k would limit the magnitude of a power excursion resulting from a reduction in moderator density.
Suitable physics measurements of moderator coefficients of reactivity will be made as part of the startup testing program to verify analytical predictions.
Analysis has shown that maintaining the moderator temperature coefficient at criticality < 5.0 pcm/°F will ensure that a negative coefficient will exist at 60% power. Current safety analysis supports OPERATING up to 60% power with a moderator temperature coefficient < 5.0 pcm/°F.
At power levels greater than 60%, a negative moderator temperature coefficient must exist.
The calculated hot full power (HFP) moderator temperature coefficient will be more negative than
-8.0 pcm/°F for at least 95% of a cycle's time at HFP to ensure the limitations associated with and 2 2 anticipated transient without scram (ATWS) event are not exceeded. NRC approved methods1 °111) will be used to determine the lowest expected HFP moderator temperature coefficient for the 5% of HFP cycle time with the highest boron concentration. The cycle time at HFP is the maximum number of days that the cycle could be at HFP based on the design calculation of cycle length. The cycle time at HIFP can also be expressed in terms of burnup by converting the maximum number of days at full power to an equivalent bumup. If this HFP moderator temperature coefficient is more negative than -8.0 pcm/°F, then the ATWS design limit will be met for 95% of the cycle's time at HFP. If this HFP moderator temperature coefficient design limit is still not met after excluding the 5% of the cycle bumup with the highest boron concentration, then the core loading must be revised.
The results of this design limit consideration will be reported in the Reload Safety Evaluation Report.
(20) USAR Section 3.2.1 (20) "NRC Safety Evaluation Report for Qualification of Reactor Physics, Methods for Application to Kewaunee,"
dated October 22, 1979.
(21) "NRC Safety Evaluation Report for the Reload Safety Evaluation Methods for Application to Kewaunee." dated April 11, 1988.
TS B3.1-12 12/17/01
In the event that the limits of TS 3.1.f.3 are not met, administrative rod withdrawal limits shall be developed to prevent further increases in temperature with a moderator temperature coefficient that is outside analyzed conditions. In this case, the calculated HFP moderator temperature coefficient will be made less negative by the same amount the hot zero power moderator temperature coefficient exceeded the limit in TS 3.1 .f.3. This will be accomplished by developing and implementing administrative control rod withdrawal limits to achieve a moderator temperature coefficient within the limits for HFP moderator temperature coefficient.
Due to the control rod insertion limits of TS 3.10.d and potentially developed control rod withdrawal limits, it is possible to have a band for control rod location at a given power level. The withdrawal limits are not required if TS 3.1.f3 is satisfied or if the reactor is subcritical.
If after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, withdrawal limits sufficient to restore the moderator temperature coefficient to within the limits of TS 3.1.f.3 are not developed, the plant shall be taken to HOT STANDBY until the moderator temperature coefficient is within the limits of TS 3.11. The reactor is allowed to return to criticality whenever TS 3.1 .f is satisfied.
TS B3.1-13 12/17/01
BASIS - Steam and Power Conversion System (TS 3.4)
Main Steam Safety Valves (TS 3.4.a)
The ten main steam safety valves (MSSVs) (five per steam generator) have a total combined rated capability of 7,660,380 lbs./hr. at 1181 lbs./in.' pressure. The maximum full-power steam flow at 1721 MWt is 7,449,000 lbs./hr. Therefore, the main steam safety valves will be able to relieve the total maximum steam flow if necessary.
While the plant is in the HOT SHUTDOWN condition, at least two main steam safety valves per steam generator are required to be available to provide sufficient relief capacity to protect the system.
The OPERABILITY of the MSSVs is determined by periodic surveillance testing in accordance with the Inservice Testing Plan.
Auxiliary Feedwater System (TS 3.4.b)
The Auxiliary Feedwater (AFW) System is designed to remove decay heat during plant startups, plant shutdowns, and under accident conditions. During plant startups and shutdowns the system is used in the transition between Residual Heat Removal (RHR) System decay heat removal and Main Feedwater System operation.
The AFW System is considered OPERABLE when the components and flow paths required to provide redundant AFW flow from the AFW pumps to the steam generators are OPERABLE. This requires that the two motor-driven AFW pumps be OPERABLE, each capable of taking suction from the Service Water System and supplying AFW to separate steam generators. The turbine-driven AFW pump is required to be OPERABLE with redundant steam supplies from each of two main steam lines upstream of the main steam isolation valves and shall be capable of taking suction from the Service Water System and supplying AFW to both of the steam generators. With no AFW trains OPERABLE, immediate action shall be taken to restore a train.
Auxiliary feedwater trains are defined as follows:
"A" train - "A" motor-driven auxiliary feedwater pump and associated AFW valves and piping to "A" steam generator, not including AFW-1OA or AFW-1OB "B" train - "B" motor-driven auxiliary feedwater pump and associated AFW valves and piping to "B" steam generator, not including AFW-10A or AFW-1OB Turbine-driven Turbine-driven AFW pump and associated AFW valves and piping to both train - "A" steam generator and "B" steam generator, including AFW-10A and AFW-10B TS B3.4-1 12/17/01
In the unlikely event of a loss of off-site electrical power to the plant, continued capability of decay heat removal would be ensured by the availability of either the steam-driven AFW pump or one of the two motor-driven AFW pumps, and by steam discharge to the atmosphere through the main steam safety valves. Each motor-driven pump and turbine-driven AFW pump is normally aligned to both steam generators. Valves AFW-10A and AFW-10B are normally open. Any single AFW pump can supply sufficient feedwater for removal of decay heat from the reactor.
As the plant is cooled down, heated up, or operated in a low power condition, AFW flow will have to be adjusted to maintain an adequate water inventory in the steam generators. This can be accomplished by any one of the following:
- 1. Throttling the discharge valves on the motor-driven AFW pumps
- 2. Closing one or both of the cross-connect flow valves
- 3. Stopping the pumps If the main feedwater pumps are not in operation at the time, valves AFW-2A and AFW-2B must be throttled or the control switches for the AFW pumps located in the control room will have to be placed in the "pull out" position to prevent their continued operation and overfill of the steam generators. The cross-connect flow valves may be closed to specifically direct AFW flow. Manual action to re-initiate flow after it has been isolated is considered acceptable based on analyses performed by WPSC and the Westinghouse Electric Corporation. These analyses conservatively assumed the plant was at 100% initial power and demonstrated that operators have at least 10 minutes to manually initiate AFW during any design basis accident with no steam generator dryout or core damage. The placing of the AFW control switches in the "pull out" position, the closing of one or both cross-connect valves, and the closing or throttling of valves AFW-2A and AFW-2B are limited to situations when reactor power is <15% of RATED POWER to provide further margin in the analysis.
During accident conditions, the AFW System provides three functions:
- 1. Prevents thermal cycling of the steam generator tubesheet upon loss of the main feedwater pump
- 2. Removes residual heat from the Reactor Coolant System until the temperature drops below 300-350'F and the RHR System is capable of providing the necessary heat sink
- 3. Maintains a head of water in the steam generator following a loss-of-coolant accident Each AFW pump provides 100% of the required capacity to the steam generators as assumed in the accident analyses to fulfill the above functions. Since the AFW System is a safety features system, the backup pump is provided. This redundant motor-driven capability is also supplemented by the turbine-driven pump.
TS B3.4-2 12/17/01
The pumps are capable of automatic starting and can deliver full AFW flow within one minute after the signal for pump actuation. However, analyses from full power demonstrate that initiation of flow can be delayed for at least 10 minutes with no steam generator dryout or core damage. The head generated by the AFW pumps is sufficient to ensure that feedwater can be pumped into the steam generators when the safety valves are discharging and the supply source is at its lowest head.
Analyses by WPSC and the Westinghouse Electric Corporation show that AFW-2A and AFW-2B may be in the throttled or closed position, or the AFW pump control switches located in the control room may be in the "pull out" position without a compromise to safety. This does not constitute a condition of inoperability as listed in TS 3.4.b.1 or TS 3.4.b.2. The analysis shows that diverse automatic reactor trips ensure a plant trip before any core damage or system overpressure occurs and that at least 10 minutes are available for the operators to manually initiate auxiliary feedwater flow (start AFW pumps or fully open AFW-2A and AFW-2B) for any credible accident from an initial power of 100%.
The OPERABILITY of the AFW System following a main steam line break (MSLB) was reviewed in our response to IE Bulletin 80-04. As a result of this review, requirements for the turbine-driven AFW pump were added to the Technical Specifications.
For all other design basis accidents, the two motor-driven AFW pumps supply sufficient redundancy to meet single failure criteria. In a secondary line break, it is assumed that the pump discharging to the intact steam generator fails and that the flow from the redundant motor-driven AFW pump is discharging out the break. Therefore, to meet single failure criteria, the turbine-driven AFW pump was added to Technical Specifications.
The cross-connect valves (AFW-I OA and AFW- I OB) are normally maintained in the open position This provides an added degree of redundancy above what is required for all accidents except for a MSLB. During a MSLB, one of the cross-connect valves will have to be repositioned regardless if the valves are normally opened or closed. Therefore, the position of the cross-connect valves does not affect the performance of the turbine-driven AFW train. However, performance of the train is dependent on the ability of the valves to reposition. Although analyses have demonstrated that operation with the cross-connect valves closed is acceptable, the TS restrict operation with the valves closed to <15% of RATED POWER. At > 15% RATED POWER, closure of the cross-connect valves renders the TDAFW train inoperable.
An AFW train is defined as the AFW system piping, valves and pumps directly associated with providing AFW from the AFW pumps to the steam generators. The action with three trains inoperable is to maintain the plant in an OPERATING condition in which the AFW System is not needed for heat removal. When one train is restored, then the LIMITING CONDITIONS FOR OPERATION specified in TS 3.4.b.2 are applied. Should the plant shutdown be initiated with no AFW trains available, there would be no feedwater to the steam generators to cool the plant to 350'F when the RHR System could be placed into operation.
It is acceptable to exceed 350'F with an inoperable turbine-driven AFW train. However, OPERABILITY of the train must be demonstrated within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after exceeding 350°F or a plant shutdown must be initiated.
TS B3.4-3 12/17-/01
Condensate Storage Tank (TS 3.4.c6 The specified minimum water supply in the condensate storage tanks (CST) is sufficient for four hours of decay heat removal. The four hours are based on the Kewaunee site specific station blackout (loss of all AC power) coping duration requirement.
The shutdown sequence of TS 3.4.c.3 allows for a safe and orderly shutdown of the reactor plant if the specified limits cannot be met. (1)
Secondary Activity Limits (TS 3.4.d)
The maximum dose to the thyroid and whole body that an individual may receive following an accident is specified in GDC 19 and 10 CFR 100. The limits on secondary coolant activity ensure that the calculated doses are held to the limits specified in GDC 19 and to a fraction of the 10 CFR 100 limits.
The secondary side of the steam generator's activity is limited to < 0.1 ýtCi/cc DOSE EQUIVALENT 1-131 to ensure the thyroid dose does not exceed the GDC-19 and 10 CFR 100 guidelines. The applicable accidents identified in the USAR121 are analyzed assuming various inputs including steam generator activity of 0.1 ptCi/cc DOSE EQUIVALENT I-131. The results obtained from these analyses indicate that the control room and off-site thyroid dose are within the acceptance criteria of GDC-19 and a fraction of 10 CFR 100 limits.