ML20078P615
| ML20078P615 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 02/10/1995 |
| From: | ENTERGY OPERATIONS, INC. |
| To: | |
| Shared Package | |
| ML20078P610 | List: |
| References | |
| NUDOCS 9502170156 | |
| Download: ML20078P615 (12) | |
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i PROPOSED ANO-1 TECHNICAL SPECIFICATION BASES PAGES i
9502170156 950210 PDR ADOCK 05000313 P
RASEse I
The plant is designed to operate with both reactor coolant loops and at least one reactor coolant pump per loop in operation, and maintain DNBR above'l.30 (for the BAN-2 correlation) and 1.18 (for the BNC correlation) l l
during all normal operations and anticipated transients. (1) l whenever the reactor coolant average temperature is above 200*F, single failure considerations require that two loops be operable.
The decay heat removal system suction piping in designed for 300*F thus, the system can remove decay heat when the reactor coolant syst.em is below this temperature.
(2,3)
One pressuriser code safety valve is capable of preventing overpressurization when the reactor is not critical since its relieving capacity is greater than that required by the sum of the available heat sources which are pump energy, pressurizer heaters, and reactor decay heat.
(4)
Both pressurizer code safety valves are required to be in service prior to criticality to conform to the system design relief capabilities.
The code safety valves prevent overpressure for a rod withdrawal accident.
(5)
The pressurizer code safety valve lift setpoint shall be 2,500 psig il percent allowance for error and each valve shall be capable of relieving 300,000 lb/h of saturated steam at a pressure not greater than 3 percent above the set pressure. When testing the pressurizer code safety valves, the "as found" lift setpoint may be 2500 psig +1,
-3 percent.
However, if found outside the il percent tolerance band, they shall be reset to 2500 psig il percent.
The internals vent valves are provided to relieve the pressure generated by steaming in the core following a LOCA so that the core remains sufficiently covered.
Inspection and manual actuation of the internal vent valves (1) ensure operability, (2) ensure that the valves are not open during normal operation, and (3) demonstrate that the valves begin to open and are fully open at the forces equivalent to the differential pressures assumed in the safety analysis.
The reactor coolant vents are provided to exhaust noncondensible gases and/or steam fram the primary system that could inhibit natural circulation core cooling. 15e operability of at least one reactor coolant system vent path from the reactor vessel head, the reactor coolant system highpoints, and the pressurizer steam space ensures the capability exists to perform this function. The valve redundancy of the vent paths serves to minimize the probability of inadvertent actuation and breach of reactor coolant pressure boundary while ensuring that a single failure of a vent valve, power supply, or control system does not preven 4 isolation of the vent path. Testing requirements are covered in Sectf.on 4.0 for the class 2 valves and Table 4.1-2 fcr the vent paths. These are consistent with ASME j
Section XI and Item II.B.1 of NUREG-0737, " Clarification of TMI Action Plan Requirements," 11/80.
i REFERENCES (1)
FSAR,' Tables 9-10 and 4-3 through 4-7 (2)
FSAR, Section 4.2.5.1 and 9.5.2.3 (3)
FSAR, Section 4.2.5.4 (4)
FSAR, Section 4.3.10.4 and 4.2.4 (5)
FSAR, Section 4.3.7 Amendment No. M, H,M 17 REVISED BY NRC LETTER DATED:
OEC ~" ". 15, 1991 i
Minimam volumes (including a 10% safety factor) as specified by Figure 3.2-1 for the boric acid addition tank or an operable borated water storage l
tank (3) will each satisfy this requirement. The specification assures that adequate supplies are available whenever the reactor is heated above 200'F so that a single failure will not prevent boration to a cold condition. The minimum volumes of boric acid solution given include the boron necessary to account for xenon decay.
The principal method of adding boron to the primary system is to pump the concentrated boric acid solution (8700 ppm boron, udnimum) into the makeup tank using the 25 gpm boric acid pumps.
The alternate method of addition is to inject boric acid from the borated water storage tank using the makeup pumps.
Concentration of boron in the boric ac.id addition tank may be higher than the concentration which uould crystallize at ambient conditions. For this reason and to assure a flow of boric acid is available when necded this tank and its associated piping will be kept 10'F above the crystallization temperature for the concentration present. Once in the makeup system, the concentrate is sufficiently well mixed and diluted so that normal system temperatures assure boric acid solubility.
REFEREN'XS 1.
FSAR, Section 9.1; 9.2 2.
FSAR, Figure 6-2 3.
SAR, Section 3.1 l
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r r
Amendment No. 44,44,64 35
370,100 gallcas of b rsted wator cro cupplied for emergsney caro ecoling cnd rosctor building cprey in th3 cv:nt of a loco-of-csolant accid:nt, This amount fulfills requirements for emergency core cooling. Approximately 16,000 gallons of bo, rated water are required to reach cold shutdown. The origitaal nominal borated water storage tank capacity 'sf 380,000 gallons is based on refueling volume ~ requirements.
Heaters maintain the borated water supply at a temperature to prevent crystallization and local freezing of the boric acid. The minimum required BWST boron concentration of 2270 ppa assures that the core will be maintained at least 1 percent Ak/k suberitical at 70'r without any control rods in the core.
Specification 3.3.2 assures that above 350*r two high pressure injection pumps are also available to provide injection water as the energy of the reactor coolant system is increased.
l Specification 3.3.3 assures that above 800 psig both core floeding tanks are operational. Since their design pressure is 600 i 25 psig, they are not brought into the operational state until 800 psig to prevent spurious injection of borated water. Both core flooding tanks are specified as a single core flood tank has insufficient inventory to reflood the core. (1)
Specification 3.3.4 assures that prior to going critical the redundant train of reactor building emergency cooling and spray train are operable.
The spray system utilizes common suction lines with the low pressure injection system.
If a single train of equipment is removed from either system, the other train must be assured to be operable in each system.
Reference 6 provides an assessment of the impact of level indicator instrument error on the allowed NaOH tank level variation. Note that the indicated level variation of 33.2 + 1.8 feet includes an allowance for instrument loop error.
When the reactor is critical, maintenance is allowed per specification 3.3.5.
Operability of the specified components shall be based on the results of testing as required by Technical Specification 4.5.
The maintenance period of up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is acceptable if the operability of equipment redundant to that removed from service is demonstrated within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to removal.
Exceptions to Specification 3.3.6 permit continued operation for seven days if one of two BWST level instrument channels is operable or if either the pressure or level instrument channel in the crT instrument channel is operable, r
In the event that the need for emergency core cooling should occur, functioning of one train (one high pressure injection pump, one low pressure injection pump, and both core flooding tanks) will protect the core and in the event of a main 9
coolant loop severance, limit the peak clad temperature to less than 2200'r and 7
the metal-water reaction to that representing less than 1 percent of the clad.
The service water system consists of two independent but interconnected, full capacity, 100% redundar.t systems, to ensure continuous heat removal.(4) i one service water pump is required for normal operation. The normal operating requirements are greater than the emergency requirements following a loss-of-coolant accident.
5 l
1 Amendment No. 444,146,164 39 REVISED BY NRC LETTER DATED:
9/31/?l; ?/15/?
m _. _
3.
Exc:pt for physics tests or sxsreising centrol rods, ths control rod position setpoints are sp2cified in ths CORE OPERATING LIMITS REPORT for 4, 3, AND 2 pump operation.
If the applicable control rod position setpoints are exceeded, corrective measures shall be taken immediately to achieve an acceptable control rod position. Acceptable control rod positions shall be attained within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
4.
Except for physics tests or exercising axial power shaping rods (APSRs), the limits for APSR position are specified in the CORE OPERATING LIMITS REPORT.
With the APSRs outside the specified limit provided in the CORE OPERATING LIMITS REPORT, corrective measures shall be taken immediately to achieve the correct position.
Acceptable APSR positions shall be attained within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
t 3.5.2.6 Reactor Power Imbalance:
1.
Reactor power imbalance shall be monitored on a frequency not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> during power operation above 40% rated power.
2.
Except for physics tests, reactor power imbalance shall be maintained within the envelope defined by the CORE OPERATING LIMITS REPORT.
3.
If the reactor power imbalance is not within the envelope defined by the CORE OPERATING LIMITS REPORT, corrective measures shall be taken to achieve an acceptable reactor power imbalance.
4.
If an acceptable reactor power imbalance is not achieved within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, reactor power shall be reduced until reactor power imbalance setpoints are met.
3.5.2.7 The control rod drive patch panels shall be locked at all times with limited access to be authorized by the Superintendent.
Bases The reactor power-imbalance envelope defined in the CORE OPERATING LIMITS REPORT is based on either LOCA analyses (which have defined the maximum linear l
heat rate (see CORE OPERATING LIMITS REPORT), such that the maximum cladding temperature will not exceed the Final Acceptance criteria) or loss of forced reactor coolant flow analysis (such that the hot fuel rod does not experience a departure from nucleate boiling condition). Corrective measures will be taken immediately should the indicated quadrant power tilt, control rod position, or reactor power imbalance be outside their specified boundaries.
Operation in a situation that would cause the Final Acceptance Criteria to be approached should a LOCA or loss of forced reactor coolant flow occur is highly improbable because l
all of the p3wer distribution parameters (quadrant power tilt, rod position, and reactor power imbalance) must be at their limits while Amendment No. 6,M,M,4,M,M6, 48 4M, M4, Me v
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l 4.7 REACTOQ CONTROL ROD SYSTEM TESTS 4.7.1 Control Rod Drive System Functional Tests l
Applicability Applies to the surveillance of the control rod system.
Objective i
To assure operability of the control rod system.
Specification 4.7.1.1 The control rod trip insertion time shall be measured for each control rod at either full flow or no flow conditions following each refueling outage prior to return to power. The maximum control rod trip insertion time for an operable control rod drive mechanism, except for the Axial Power Shaping Rods
( APS Rs ), from the fully withdrawn position to 3/4 insertion (104 inches travel) shall not exceed 1.66 seconds at reactor l
coolant full flow conditions or 1.20 seconds for no flow conditions.
For the APSRs it shall be demonstrated that loss i
of power will not cause rod movement.
If the trip insertion time above is not met, the rod shall be declared inoperable.
4.7.1.2 If a control rod is misaligned with its group average by more j
than an indicated nine (9) inches, the rod shall be declared inoperable and the limits of Specification 3.5.2.2 shall apply.
The rod with the greatest udsalignment shall be evaluated first. The position of a rod declared inoperable due to adsalignment shall not be included in computing the average position of the group for detennining the operability of rods with lesser misalignments.
4.7.1.3 If a control rod cannot be exercised, or if it cannot be located with absolute or relative position indications or in or out limit lights, the rod shall be declared to be inoperable.
Bases The control rod trip insertion time is the total elapsed time from power interruption at the control rod drive breakers until the control rod has completed 104 inches of travel from the fully withdrawn position. The l
specified trip time is based upon the safety analysis in FSAR, Section 14, j
whose calculations are based on a rod drop from fully withdrawn to 2/3 inserted. Since the most accurate position indication is obtained from the zone reference switch at the 3/4 inserted position, this position is used j
instead of the 2/3 inserted position for data gathering.
j Each control rod drive mechanism shall be exercised by a movement approximately two (2) inches of travel every two (2) weeks. This requirement shall apply to either a partial or fully withdrawn control rod at reactor operating conditions.
Exercising the drive mechanisms in this manner provides assurance of reliability of the mechanisms.
A rod is considered inoperable if it cannot be exercised, if the trip insertion time is greater than the specified allowable time, or if the rod 1
i Amendment No. M,M6 102
)
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MARKFD.UP ANO-1 TECHNICAL SPECIFICATION BASES PAGES t
f r
f f
1
i RASES:
The plant is designed to operate with both reactor coolant loops and at least, one reactor coolant pump per loop in operation, and maintain DNBR 1
above 1.30 (for the RAN-2 correlation) and 1.18 ffor the BWC correlation)
.l
'during all normal operations and anticipated transients. (1)
I whenever the reactor coolant average temperature is above 280*F, single failure considerations require that two loops be operable.
The decay heat removal system suction piping is designed for 300*F thus, the system can remove decay heat when the reactor coolant system is below this temperature.
(2,3)
One pressurizer code safety valve is capable of preventing overpressurization when the reactor is not critical since its relieving capacity is greater than that required by the sum of the available heat sources which are pump energy, pressurizer heaters, and reactor decay heat.
(4)
Both pressurizer code safety valves are required to be in service prior to criticality to confons to the system design relief capabilities.
The code safety valves prevent overpressure for a rod withdrawal accident.
(5)
The pressurizer code safety valve lift setpoint shall be 2,500 psig il percent allowance for error and each valve shall be capable of relieving 300,000 lb/h of saturated steam at a pressure not greater than 3 percent above the set pressure. When testing the pressurizer code safety valves, the "as found" lif t setpoint may be 2500 psig +1,
-3 percent. However, if found outside the il percent tolerance band, they shall be reset to 2500 psig il percent.
i The internals vent valves are provided to relieve the pressure generated by steaming in the core following a LOCA so that the core remains sufficiently covered.
Inspection and manual actuation of the internal vent valves (1) ensure operability, (2) ensure that the valves are not open during normal operation, and (3) demonstrate that the valves begin to open and are fully open at the forces equivalent to the differential pressures assumed in the safety analysis.
The reactor coolant vents are provided to exhaust noncondensible gases and/or steam from the primary system that could inhibit natural circulation core cooling. The operability of at least one reactor coolant system vent path from the reactor vessel head, the reactor coolant system highpoints, and the pressurizer steam space ensures the capability exists to perform i
this function. The valve redundancy of the vent paths serves to minimize the probability of inadvertent actuation and breach of reactor coolant pressure boundary while ensuring that a single failure of a vent valve, power supply, or control system does not prevent isolation of the vent path. Testing requirements are covered in Section 4.0 for the class 2 valves and Table 4.1-2 for the vent paths. These are consistent with ASME Section XI and Item II.B.1 of NUREG-0737, " Clarification of TMI Action Plan Requirements," 11/80.
REFERENCES (1)
FSAR, Tables 9-10 and 4-3 through 4-7 (2)
FSAR, Section 4.2.5.1 and 9.5.2.3 (3)
FSAR, Section 4.2.5.4 (4)
FSAR, Section 4.3.10.4 and 4.2.4 (5)
FSAR, Section 4.3.7 I
Amendment No. M,M,94 17 REVISED BY NRC LETTER DATED t - DECSP.S"S 15, 1491 l
l t
e
Minimum volumes (including a 10% safety factor) as specified by Figure 3.2-1 for the boric acid addition tank or 44,5t0 g:11::: Of 2270 ;; 5:::n
- 5::10 ::id ::12 tion in thean operable borated water storage tank (3) will each satisfy this requirement. The specification assures that adequate supplies are available whenever the reactor is heated above 200*F so that a single failure will not prevent boration to a cold condition. The minimum volumes of boric acid solution given include the boron necessary to account for xenon decay.
The principal method of adding boron to the primary system is to pump the concentrated boric acid solution (8700 ppm boron, minimum) into the umkeup tank using the 25 gpm boric acid pumps.
The alternate method of addition is to inject boric acid from the borated water storage tank using the makeup pumps.
Concentration of boron in the boric acid addition tank may be higher than the concentration which would crystallize at ambient conditions.
For this reason and to assure a flow of boric acid is available when needed this tank and its associated piping will be kept 10'F above the crystallization temperature for the concentration present. Once in the makeup system, the concentrate is sufficiently well mixed and diluted so that normal systen temperatures assure boric acid solubility.
REFERENCES 1.
FSAR, Section 9.1; 9.2 2.
FSAR, Figure 6-2 3.
FS.' ", S : tion 3.3SAR. Section 3.1 l
l j
i Amendment No. 44,43,64 35
370,100 gallens of beretcd w: tor cro cupplicd for emerg:ncy caro cooling end coactor building rpray in tha cv:nt of a locc-of-coolent accid:nt. This amount fulfills requirements for emergency core cooling. Approximately 16,000 gallons of borated water are required to reach cold shutdown. The original nominal borated water storage tank capacity of 380,000 gallons is based on refueling volume requirements. Heaters maintain the borated water supply at a temperature to prevent crystallization and local freezing of the boric acid. The miniggm reauired BWST boron concentaation i; ::t :t : v:lu: th:t will : int inof 2270 Dom assures that the core will be maintained at least 1 percent Ak/k suberitical at 70*r without any control rods in the core. The ;;;;;ntr:ti:n f : it ik/h emberiticality i: 1500 ;;1 h:::n i; th: ::::, =hil: th mini-v:1;; :p::ified 4: th: 5:::ted :t; :t:::;; tank i: 2270 pe; h::::.
Specification 3.3.2 assures that above 350*r two high pressure injection pumps are also available to provide injection water as the energy of the reactor coolant system is increased.
Specification 3.3.3 assures that above 800 psig both core flooding tanks are operational.
Since their design pressure is 600 1 25 psig, they are not brought into the operational state until 800 psig to prevent spurious injection of borated water. Both core flooding tanks are specified as a single core flood tank has insufficient inventory to reflood the core. (1)
Specification 3.3.4 assures that prior to going critical the redundant train of reactor building emergency cooling and spray train are operable.
The spray system utilizes common suction lines with the low pressure injection system.
If a single train of equipment is removed from either system, the other train must be assured to be operable in each system.
Reference 6 provides an assessment of the impact of level indicator instrument error on the allowed NaOH tank level variation. Note that the indicated level variation of 33.2 + 1.8 feet includes an allowance for instrument loop error.
When the reactor is critical, maintenance is allowed per Specification 3.3.5.
Operability of the specified components shall be based on the results of testing as required by Technical Specification 4.5.
The maintenance period of up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is acceptable if the operability of equipment redundant to that removed from service is demonstrated within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to removal.
Exceptions to Specification 3.3.6 permit continued operation for seven days if one of two DWST level instrument channels is operable or if either the pressure or level instrument channel in the CFT instrument channel is operable.
In the event that the need for emergency core cooling should occur, functioning of one train (one high pressure injection pump, one low pressure injection pump, and both core flooding tanks) will protect the core and in the event of a main coolant loop severance, limit the peak clad temperature to less than 2200*r and the metal-water reaction to that representing less than 1 percent of the clad.
The service water system consists of two independent but interconnected, full capacity, 100% redundant systema, to ensure continuous heat removal.(4)
One serv!.ce water pump is required for normal operation. The normal operating requirements are greater than the emergency requirements following a loss-of-coolant accident.
Amendment No, MO,M&,M4 39 REVISED BY NRC LETTER DATED:
0/21/91; '/15/92
)
3.
Exc:pt for phycies tosto or ex:rcicing control reda, tha
)
control rod position setpoints are specified in the CORE OPERATING LIMITS REPORT for 4, 3, AND 2 pump operation.
If the applicable control rod position setpoints are l
exceeded, corrective measures shall be taken immediately to achieve an acceptable control rod position. Acceptable control rod positions shall be attained within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
4.
Except for physics tests or exercising axial power shaping j
rods (APSRs), the limits for APSR position are specified in the CORE OPERATING LIMITS REPORT.
With the APSRs outside the specified Itmit provided in the CORE OPERATING LIMITS REPORT, corrective measures shall be taken inmediately to achieve the correct position.
Acceptable APSR positinns shall be attained within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
3.5.2.6 Reactor Power Imbalance:
1.
Reactor power imbalance shall be monitored on a frequency not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> during power operation above 40% rated power.
2.
Except for physics tests, reactor power imbalance shall be maintained within the envelope defined by the CORE OPERATING LIMITS REPORT.
3.
If the reactor power imbalance is not within the envelope defined by the CORE OPERATING LIMITS REPORT, corrective measures shall be taken to achieve an acceptable reactor power imbalance.
4.
If an acceptable reactor power imbalance is not achieved within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, reactor power shall be reduced until reactor power imbalance setpoints are met.
3.5.2.7 The control rod drive patch panels shall be locked at all times with limited access to be authorized by the Superintendent.
Bases The reactor power-imbalance envelope defined in the CORE OPERATING LIMITS REPORT is based on either LOCA analyser Jwhich have defined the maximum linear l
heat rate (see CORE OPERATING LIMITS REPORT), such that the maximum cladding temperature will not exceed the Final Acceptance Criteria) or loss of forced reactor coolant flow analysis (such that the hot fuel rod does not experience a departure from nucleate boilino condition). Corrective measures will be taken immediately should the indicated quadrant power tilt, control rod position, or reactor power imbalance be outside their specified boundaries.
Operation in a situation that would cause the Final Acceptance Criteria to be approached should a LOCA or loss of forced reactor coolant flow occur is highly improbable because l
all of the power distribution parameters (quadrant power tilt, rod position, and reactor power imbalance) must be at their limits while Amendment No. 6,M,M,44,93,4M, 48 444,4M, M9
9 4.7 REACTOR CONTROL ROD SYSTEM TESTS a
4.7.1 Control Rod Drive System Functional Tests Applicability Applies to the surveillance of the control rod system.
Objective To assure operability of the control rod system.
Specification 4.7.1.1 The control rod trip insertion time shall be measured for each control rod at either full flow or no flow conditions following each refueling outage prior to return to power. The maximum control rod trip insertion time for an operable control rod drive mechanism, except for the Axial Power Shaping Rods (APSRs), from the fully withdrawn position to 3/4 insertion (104 inches travel) shall not exceed 1,66 seconds at reactor coolant full flow conditions or 1.20 seconds for no flow conditions.
For the APSRs it shall be demonstrated that loss of power will not cause rod movement. If the trip insertion time above is not met, the rod shall be declared inoperable.
4.7.1.2 If a control rod is misaligned with its group average by more than an indicated nine (9) inches, the rod shall be declared inoperable and the limits of Specification 3.5.2.2 shall apply.
The rod with the greatest misalignment shall be evaluated first. The position of a rod declared inoperable due to misalignment shall not be included in computing the average position of the group for determining the operability of rods with lesser misalignments.
4.7.1.3 If a control rod cannot be exercised, or if it cannot be located with absolute or relative position indications or in or out limit lights, the rod shall be declared to be inoperable.
Bases The control rod trip insertion time is the total elapsed time from power interruption at the control rod drive breakers until the control rod has completed 104 inches of travel from the fully withdrawn position. The specified trip time is based upon the safety analysis in FSAR, Section 14 6
whose calculations are based on a rod drop from fully withdrawn to 2/3 inserted.
Since the most accurate position indication is obtained from the zone reference switch at the 3/4 inserted position. this position is used instead of the 2/3 inserted oosition for data aatherina.
Each control rod drive mechanism shall be exercised by a movement approximately two (2) inches of travel every two (2) weeks. This requirement shall apply to either a partial or fully withdrawn control rod at reactor operating conditions. Exercising the drive mechanisms in this manner provides assurance of reliability of the mechanisms.
A rod is considered inoperable if it cannot be exercised, if the trip insertion time is greater than the specified allowable time, or if the rod i
Amendment No. M,M6 102 i
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