ML20081L790
| ML20081L790 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 06/28/1991 |
| From: | OMAHA PUBLIC POWER DISTRICT |
| To: | |
| Shared Package | |
| ML20081L781 | List: |
| References | |
| NUDOCS 9107080089 | |
| Download: ML20081L790 (19) | |
Text
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F ATTACHMENT A Ng 9107080089 910628 PDR ADOCfc 0* >00028*i
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P PDR' 1 1
l TECllNICAL SPECIFICATIONS TAllLE OF CONTENTS PAGE DEFINITIONS 1
1.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 11 1.1 Safety Limits - Reactor Core 1-1 1.2 Safety Limit, Reactor Coolant System Pressure 1-4 1.3 Limiting Safety System Settings, Reactor Protective System 1-6 2.0 LIMITING CONDITIONS FOR OPERATION 2-0 2.0.1 General Requirements 2-0 2.1 Reactor Coolant System 2-1 2.1.1 Operable Components 2-1 2.1.2 lieatup and Cooldown Rate 23 2.1.3 Reactor Coolant Radioactivity 2-8 2.1.4 Reactor Coolant System Leakage Limits 2-11 2.1.5 Maximu.n Reactor Coolant Oxygen and IIalogens Concentrations 2-13 2.1.6 Pressurizer and hinin Steam C Mem Safety Vaives 2-15 2.1.7 Pressurizer Operability 2-16a 2.1.8 Reactor Coolant System Vents 2-16h 2.2 Chemical and Volume Control System 2 17 2.3 Emergency Core Cooling System 2-20 2.4 Containment Cooling 2-24 2.5 Steam and Feedwater Systems 2-28 2.6 Containment System 2-30 2.7 Electrical Systems 2-32 2.8 Refueling Operations 2-37 2,9 Radioactive Efiluents 2-40 2.9.1 Liquid and Gaseous Effluents 2-40 2.9.2 Solid Radioactive Waste 2-47a 2.10 Reactor Core 2-48 2.10.1 Minimum Conditions for Criticality 2-48 2.10.2 Reactivity Contro Systems and Core Physics Parameter Limits 2-50 2.10.3 In-Core Instrumentation 2-54 2.10.4 Power Distribution Limits 2-56 2.11 Containment Building and Fuel Storage Building Crane 2-58 i
Amendment No. 32,38,52,- 4,67,67, 5
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2.0 LIMITING CONDITIONS FOR OPERATION 2.1 Reactor Coolant System (continued) 2,1.-
Pressurizer and Main Steam System Safety Valves Applicability Applies to the status of the pressurizer and main steam system safety valves.
Objective To specify minimum requirements pertaining to the pressurizer and main steam system safety valves.
Specifications To provide adequate overpressure protection for the reactor coolant system and steam system, the following safety valve requirements shall be met:
(1)
The reactor shall not be made critical unless the two pressurizer safety valves are operable with their lift settings adjusted to ensure valve opening between a12500 psia +/-l % and 2545 psia il%.*
(2)
Whenever there is fuel in the reactor, and the reactor vessel head is installed, a minimum of one operable safety valve shall be installed on the pressurizer.
However, when in at least the cold shutdown condition, safety valve nozzles may be open to containment atmosphere during performance of safety valve tests or maintenance to satisfy this specification.
(3)
Whenever the reactor is in power operation, eight of the ten main steam safety valves shall be operable with their lift settings between 1000 psia and 1050 psia with a tolerance of 14 +3/-2% of the nominal nameplate setpoint values.*
(4)
Both pressurizer power-operated relief valves (PORV's) shall be operable during scheduled heatup and cooldown to prevent violation of the pressure-temperature limits designated by Figures 2-1 A and 2-1B. One PORV may be inoperable for up to 7 days, provided the remaining PORV is operable. If the above conditions of this paragraph cannot be met, the primary system must be depressurized and vented.
(5)
Two power-operated relief valves (PORV's) and their associated block valves shall be operable in Modes 1,2, and 3.
2-15 Amendment No. 39,4-7,54
=
s 2.0 I,IMITING CONDITIONS FOR OPERATION 2.1 Reactor Coolant System (continued) 2.1.6 Pressuri7er and Main Steam System Safety Valves (continued) a.
With one or more PORV(s) inoperable, within I hour either restore the PORV(s) to operable status or close the associated block valve (s); otherwise, be in at least HOT STANDBY within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
b.
With one or more block valve (s) inoperable, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> either restore the block valve (s) to operable status or close the block valve (s). Otherwise, be in at least HOT STANDBY within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Basis The highest reactor coolant system pressure reached in any of the accidents analyzed was 2480--psia-end resulted from a complete loss of turbine generator load without simultaneous reactor trip while operating at 1500 MWt.m This pressure was less than the 2750 osia safety limit and the ASME Section III upset pressure limit of 10% creater than the design pressureJD The reactor is assumed to trip on a "High Pressurizer Pressure" trip signal.
The power-operated relief valves (PORV's) operate to relieve RCS pressure below the setting of the pressurizer code safety valves. These relief valves have remotely operated block valves to provide a positive shutoff capability should a relief valve become inoperable. The electrical power for both the relief valves and the block valves is capable of being supplied from an emergency power source to ensure the ability to seal this possible RCS leakage path.
To determine the maximum steam flow, the only other pressure relieving system assumed operationalis the main steam system safety valves. Conservative values for all systems parameters, delay times and core moderator coefficients are assumed. Overpressure protection is provided to portions of the reactor coolant system which are at the highest pressure considering pump head, flow pressure drops and elevation heads.
If no residual heat were removed by any of the means available, the amount or steam which could be generated at safety valve lift pressure would be less than half of the capacity of one safety valve. This specification, therefore, provides adequate defense against overpressurization when the reactor is subcritical.
2-15a Amendment No. 54
s 2.0 LIMITING CONDITIONS FOR OPERATION 2.1 Reactor coolant System (continued) 2.1.6 &csurizer and Main Steam System Safety Valves (continued)
Performance of certain calibration and maintenance procedures on safety valves requires removal from the pressurizer. Should a safety valve be removed, either operability of the other safety valve or maintenance of at least one nozzle open to atmosphere will assure that sufficient relief capacity is available. Use of plastic or other similar material to prevent the entry of foreign material into the open nozzle will not be construed to violate the "open to atmosphere" provision, since the presence of this material would not significantly restrict the discharge of reactor coolant.
The total relief capacity of the ten main steam system safety valves is 6.54 x 10' lb/hr.
If. following testing. the as found setooints are outside +/-l % of nominal nameplate values. the_ valves are set to within the +/-l % tolerance. The main steam safety valves were analy7ed for a total loss of main feedwater now while operating at 1500 MWtm to ensure that the ocak secondary pressure was less than 1100 psia. the ASME Section III upset pressure limit of 10% greater than the design pressure. At the power of 1500 MWt, sufGcient relief valve capacity is available to prevent overpressurization of the steam system on loss-of load conditions.*
The power-operated relief valve low setpoint will be adjusted to provide sufficient margin, when used in conjunction with Technical Specification Sections 2.1.1 and 2.3, to prevent the design basis pressure transients from causing an overpressurization incident. Limitation of this requirement to scheduled cooldown ensures that, should emergency conditions dictate rapid cooldown of the reactor coolant system, inoperability of the low temperature overpressure protection system would not prove to be an inhibiting factor.
Removal of the reactor vessel head provides sufficient expansion volume to limit any of the design basis pressure transients. Thus, no additional relief capacity is required.
References (1) Article 9 of the 1968 ASME Boiler and Pressure Vessel Code, Section 111 (2) USAR FSAR, Section 14.9 (3) USAR Section 14.10 (4) US AR FSAR, Sections 4.3.4,4.3.9.5 2-16 Amendment No. 39,47,5480,M,86
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s TECilNICAL SPECIFICATIONS TABLE OF CONTESTS PAGE DEFINITIONS I
1.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 1-1 1.1 Safety Limits - Reactor Core 11 1.2 Safety Limit, Reactor Coolant System Pressure 1-4 1.3 Limiting Safety System Settings, Reactor Protective System 16 2.0 LIMITING CONDITIONS FOR OPERATION 2-0 2.0.1 General Requirements 2-0 2-1 2.1 R actor Coolant System 2.1.1 Operable Components 2-1 2.1.2 lleatup and Cooldown Rate 2-3 2.1.3 Reactor Coolant Radioactivity 2-8 J
2.1.4 Reactor Coolant System Leakage Limits 2-11 2.1.5 Maximum Reactor Coolant Oxygen and Italogens Concentrations 2-13 2.1.6 Pressurizer and Main Steam Safety Valves 2-15 l
2.1.7 Pressurizer Operability 2-16a 2.1.8 Reactor Coolant System Vents 216b 2.2 Chemical and Volume Control System 2-17 2.3 Emergency Core Cooling System 2-20 2.4 Containment Cooling 2-24 2.5 Steam and Feedwater Systems 2-28 2.6 Containment System 2-30 2.7 Electrical Systems 2 32 2.8 Refueling Operations 2-37 2.9 Radioactive Effluents 2-40 2.9.1 Liquid and Gaseous Effluents 2-40 2.9.2 Solid Radioactive Waste 2-47a 2.10 Reactor Core 2-48 2.10.1 Minimum Conditions for Criticality 2-48 2.10.2 Reactivity Control Systems and Core Physics Parameter Limits 2-50 2.10.3 In-Core Instrumentation 2 54 2.10.4 Power Distribution Limits 2-56 2.11 Containment Building and Fuel Storage Building Crane 2-58 l
i Amendment No. 32,38,52,54,57,67,
i 2.0 1,IMITING CONDITIONS FOR OPERATION 2.1 Ecactor coolant System (continued) 2.1.6 Pressuri7er and Maln Steam Safety Valves l
Applicability Applies to the status of the pressurizer and main steam safety valves, j
Objective To specify minimum requirements pertaining to the pressurizer and main steam safety l valves.
Specifications To provide adequate overpressure protection for the reactor coolant system and steam system, the following safety valve requirements shall be met:
(1)
The reactor shall not be made critical unless the two pressurizer safety valves are operable with their lift settings adjusted to ensure valve opening at 2500 psia 11% and 2545 psia 11%.*
(2)
Whenever there is fuel in the reactor, and the reactor vessel head is installed, a minimum of one operable safety valve shall be installed on the pressurizer.
However, when in at least the cold shutdown condition, safety valve nozzles may be open to containment atmosphere during performance of safety valve tests or maintenance to satisfy this specification.
(3)
Whenever the mactor is in power operation, eight of the ten main steam safety l valves shall be operable with their lift settings between 1000 psia and 1050 psia with a tolerance of +3/-2% of the nominal nameplate setpoint values.*
{
(4)
Both pressurizer power-operated relief valves (PORV's) shall be operable during scheduled heatup and cooldown to prevent violation of the pressure-temperature limits designated by Figures 2-1 A and 2-1B. One PORV may be inoperable for up to 7 days, provided the remaining PORV is operable. If the above conditions of this paragraph cannot be met, the primary system must be depressurized and vented.
(5)
Two power-operated relief valves (PORV's) and their associated block valves shall be operable in Modes 1, 2, and 3.
2-15 Amendment No. 39,4,54 7
2.0 LINilTING CONDITIONS FOR OPERATION 2.1 Reactor Coolant System (continued) 2.1.6 Pressurizer and Main Steam Safety Valves (continued) l a.
With one or more PORV(s) inoperable, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> either restore the PORV(s) to operable status or close the associated block valve (s); otherwise, be in at least HOT STANDBY within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
b.
With one or more block valve (s) inoperable, within I hour either restore the block valve (s) to operable status or close the block valve (s). Otherwise, be in at least HOT STANDBY within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
D2 sis The highest reactor coola'.t system pressure reached in any of the accidents analyzed l resulted from a complete loss of turbine generator load without simultaneous reactor trip while operating at 1500 MWt.S This pressure was less than the 2750 psia safety limit snd the ASME Section III upset pressure limit of 10% greater than the design pressure.*
The reactor is assumed to trip on a "High Pressurizer Pressure" trip signal.
The power-operated relief valves (PORV's) operate to relieve RCS pressure below the setting of the pressurizer code safety valves. These relief valves have remotely operated bk ik valves to provide a positive shutoff capability should a relief valve become inoperable. The electrical power for both the relief valves and the block vr.lves is -
capable of being supplied from an emergency power source to ensure the ability to seal this possible RCS leakage path.
To determine the maximum steam flow, the only other pressure relieving system assumed operational is the main steam safety valves.
Conservative values for all systems l parameters, delay times and core moderator coefficients are assumed. Overpressure protection is provided to portions of the reactor coolant system which are at the highest pressure considering pump head, flow pressure drops and elevation heads.
If no residual heat were removed by any of the means available, the amount or steam which could be generated at safety valve lift pressure would be less than half of the capacity of one safety valve. This specification, therefore, provides adequate defense against overpressurization when the reactor is suberitical.
l 2-15a-
-- Amendment No. 54 l
2.0 LINilTING CONDITIONS FOR OPERATION 2.1 Reactor Coolant System (continued) 2.1.6 Pressurizer and Main Steam Safety Valves (continued) l Performance of certain calibration and maintenance procedures on safety valves requires removal from the pressurizer. Should a safety valve be removed, either operability of the other safety valve or maintenance of at least one nozzle open to atmosphere will assure that sufficient relief capacity is available. Use of plastic or other similar material to prevent the entry of foreign material into the open nozzle will not be construed to violate the "open to atmosphere" provision, since the presence of this material would not significantly restrict the discharge of reactor coolant.
The total relief capacity of the ten main steam safety valves is 6.54 x 10 lb/hr. If, following testing, the as found setpoints are outside +/-l % of nominal nameplate values, the valves are set to within the +/-l% tolerance. The main steam safety valves were analyzed for a total loss of main feedwater Gow while operating at 1500 MWt* to ensure that the peak secondary pressure was less than 1100 psia, the ASME Section III upset pressure limit of 10% greater than the design pressure. At the power of 1500 MWt, sufficient relief valve capacity is available :o prevent overpressurization of the :; tea.n system on loss-of-load conditions.*
The power-operated relief valve low setpoint will be adjusted to provide sefGeient margin, when used in conjunction with Technical Speci0 cation Sections 2.1.1 and 2.3, to prevent the design basis pressure transients from causing an overpressurization incident. Limitation of this requirement to scheduled cooldown ensures that, should emergency conditions dictate rapid cooldown of the reactor coolant system, inoperability of the low temperature overpressure protection system would not prove to be an inhibiting factor.
Removal of the reactor vessel head provides sufficient expansion volume to limit any of the design basis pressure transients. Thus, no additional relief capacity is required.
References (1) Article 9 of the 1968 ASME Boiler and Pressure Vessel Code,Section III (2) USAR, Section 14.9 (3) USAR Section 14.10 (4) USAR, Sections 4.3.4, 4.3.9.5 2-16 Amendment No. 39,4-7,5480,M,86
ATTACHMENT B i
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9 JUSTIFICATION. DISCUSSION. AND NO SIGNIFICANT HAZARDS CONSIDERATIONS FOR REEISION OF THE MAXIMUM ALLOWABLE SETPOINT DRIFT FOR THE MAIN STEAM SAFETY VALVE SETPOINTS The proposed amendment to the Technical Specifications would increase the maximum allowable setpoint drift for the main steam safety valve setpoints from 1% to +3% / -2%.
This amendment was initiated-in response to LER 87-003 which resulted from safety valve setpoint drifts in excess of that allowed by Technical Specification 2.1.6(3).
To support this amendment, the NRC-approved CESEC-III transient analysis code and analysis methods were used to determine the accentable setpoint drift for the main steam safety valves. Acceptable setp > int drift is defined as the maximum allowable drift in the main steam safecy valve setpoints which would ensure that the peak secondary pressure does not exceed the design basis acceptance criteria (i.e.110% of design) of 1100 psia, as specified in the Updated Safety Analysis Report, Section 14.10. The-wording in the Basis for Technical Specification 2.1.6 has been revised and no longer contains the highest primary pressure calculated in the transient analyses. The revised wording specifies that the highest primary 3ressure reached in any of the accidents analyzed is less than 2750 psia, w1ich is the Safety Limit for the Reactor Coolant System Pressure as specified in Technical Specification 1.2.
Future transient analyses will continue to demonstrate that the highest reactor coolant system pressure is below 2750 psia, but a facility license change will no longer be necessary to update this value.
Based on past experience and review of the NRC approved OPPD reload licensing methodology, the Loss of Load and Loss of Feedwater Flow events were found to be most limiting.
The loss of Feedwater Flow event was analyzed as the most limiting transient for determining the maximum steam generator pressure. While the analysis was performed with Cycle 11 conditions, the results bound Cycle 13 as well.
LOSS OF FEEDWATER FLOW EVENT The loss of Main Feedwater Flow event was reanalyzed using the same assumptions are Reference 1, except that the primary safety valves were modeled as inoperable and the allowable drift for the main steam safety valves was changed from +1% to +5%. This was a bounding case used in the analysis, however, an allowable drift of +3% is proposed for this amendment to ensure margin requirements are maintained. The Loss of.Feedwater Flow analyais was initiated at the conditions, and for the combination of parameters, show in' Table 1-1 to maximize the calculated peak RCS pressure and peak secondary pressure and resulted in a high pressurizer pressure trip signal at 34.9 seconds. At 37.9 seconds, the primary pressure reaches its maximum value of 2515 psia. -The increase in secondary pressure is limited by the opening of the main steam safety valves, which open at 38.7 seconds.
The secondary pressure reaches its maximum value of 1099.6 psia at 42.5 seconds after initiation of the event.
Table 1-2 summarizes the secaence of events for this transient.
Figures 1-1 through 1-4 show the transient behavior of power, RCS pressure, RCS coolant temperature and steam generator pressure, i
MAN' VAL REACTOR TRIE A manual reactor trip case with no setpoint drift was analyzed to determine the maximum allowable negative setpoint drift that would not challenge any of the safety systems. A -4% drift was determined to be the maximum allowable, however, an allowable drift of -2% is proposed for this amendment to ensure adequate margin is maintained to prevent challenge of the ESF System or components.
Based on these analyses, the enclosed application for amendment to the Technical Specifications requests revision of the allowable setpoint drift for the main steam safety valves from 1% to +3% / -2%.
Basis for No Sianificant Hazards Considerations The proposed amendment to the Technical Specification does not involve a significant hazards consideration because the operation of the Fort Calhoun Station in accordance with this change would not:
(1)
Involve a significant increase in the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report.
The proposed revision to the allowable setpoint drift was conservatively analyzed using the NRC approved transient analysis methodology and computer code (CESEC-III).
The results demonstrate that during a severe transient the peak steam generator pressure would fall significantly below the Safety Limit and design basis acceptance criteria of 1100 psia, as specified in the Updated Safety Analysis Report Section 14.10.
Since the main steam safety valves function to control transient events, revision of the allowable setpoint drift would not increase the probability of occurrence of such events.
Therefore, this amendment would not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report.
(2)
Create the possibility for an accident or malfunction of a different type than any evaluated previously in the. safety analysis report. The safety valves function to control transient events.
Analyses of the proposed allowable setpoint drift using the NRC approved transient analysis methodology and computer code (CESEC-III) demonstrates that during the limiting overpressure transient, peak steam generator pressure would be significantly below the design basis acceptance criteria of 1100 psia, as specified in the Updated Safety Analysis Report Section 14.10.
No new or different kind of accident is created because actual operation of the plant remains unchanged.
Therefore, the possibility of an accident or malfunction of a different type than any evaluated previously in the safety analysis report _ would not be created.
(3)
Involve a significant reduction in the margin of safety as defined in the basis for any Technical Specification. This revision only increases the allowable main steam safety valve setpoint drift within design '.imits as demonstrated using the NRC approved transient analysis methodology and computer code (CESEC-III).
Therefore, the margin of safety as defined in the basis for any Technical Specitication is not reduced.
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i Based on the above considerations, this amendment does not involve a significant hazards consideration.
REFERENCES 1.
Fort Calhoun Station, Unit No.1, Amendment No.109 to Facility Operating License No. DPR-40 (TAC Ncs. 64434 and 64486) from NRC (Walter A. Paulson) to OPPD (R. L. Andrews) dated May 4, 1987.
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t Table 1-1 FORT CALHOUN CYCLE 11 KEY PARAMETEP.S ASSUME 9 IN THE LOSS OF FEEDWATER FLOW ANALYSIS Parameter Units Value Initial Core Power Level MWt 1530 (102%)
Initial Core Inlet Temperature
- F-547 Initial Pressurizer Pressure psia 2,053 Initial Steam Generator Pressure psia 815 Initial RCS Flow Rate gpm 196,000 Mooerator Tempe*ature Coefficient 10-4 Ap /*F
+ 0.5 Fuel Temperature Coefficient 10-4 Ap /*F Least negative predicted during core life.
Fuel Temperature Coefficient Multiplier 0.85 l
CEA Time to 100% Insertion (Including Holding coil Delay) sec.
3.1 Scram Reactivity Worth
% Ap
-6.65 Kinetics Parameters 6
.004696 Allowable Primary Safety Valve Setpoint Drift Inoperable A'110wable Secondary Safety Valve
.Setpoint Drift
+5 L
m m
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Table 1-2 FORT CALHOUN CYCLE 11 SEQUENCE OF EVENTS FOR THE LOSS OF FEEDWATER FLOW EVENT TO MAX 1MlZE CALCULATED RCS PEAK PRESSURE Time (seci Event Setooint or Value 0.0 Loss of Secondary Load 34.9 High Pressurizer Analysis Trip 2422 psia Setpoint is Reached 36.2 CEAs Begin to Drop Into Core 37.9 Maximum RCS Pressure 2515 psia 38.7 Steam Generator Safety Valves Open 1066 psia 42.5 Maximum Steam Generator Pressure 1099.6 psia l
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90 80 i
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60 a
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O' 40 E
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2600 i
2500 2400 5
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2100
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2000 1900 1800 0
15 30 45 60 75 TIME, SECONOS LossofFeedwaterFlow OmahaPublicPowerDistrict Figure RCSPressurevs. Time fortCalhounStation-UnitNo.i 1,
I 620 i
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610 I h 600
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tu ave a
O 580
&j 570 IO 560 550 540 O
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lossofFeedwaterFlow OmahaPublicPowerDistrict Figure RCSTemperaturesvs. Time FortCalhounStation-UnitNo.i la
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1 1200 i
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E 1100 uI C
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15 30 45-60 75 TIME, SECONOS-i LossofFee0waterFlow OmanaPublicPowerDistrict Figure SteamGeneratorPressurevs. Time Fort Calhoun Station-Unit No.'i 12 l-i
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