|
|
| (One intermediate revision by the same user not shown) |
| Line 17: |
Line 17: |
|
| |
|
| =Text= | | =Text= |
| {{#Wiki_filter:SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING c.The neutron flux shall not exceed its scram setting for longer than 1.5 seconds as indicated by the process computer.When the process computer is out of service, a safety limit violation shall be assumed if the neutron flux exceeds the scram setting and control rod scram does not occur.To ensure that the Safety Limit established in Specifications 2.1.la and 2.l.lb is not exceeded, each required scram shall be initiated by its expected scram signal.The Safety Limit shall be assumed to be exceeded when scram is accomplished by a means other than the expected scram signal.d.e.The reactor water low level scram trip setting shall be no lower than-12 inches (53 inches indicator scale)>elative to the minimum normal water level (302'9").The reactor water low-low level setting for core spray initiation shall be no less than-5 feet (5 inches indicator scale)relative to the minimum normal water level (Elevation 302'9").f.~The flow biased APRM rod block trip settings shall be less than or equal to that shown in Figure 2.l.l.d.Whenever the reactor is in the shutdown condition with irradiated fuel in the reactor vessel, the water level shall not be more than 6 feet, 3 inches (-10 inches indicator scale)below minimum normal water level (Elevation 302'9")except as specifed in"e" below.e.For the purpose of performing major maintenance (not to exceed 12 weeks in duration)on the reactor vessel;the reactor water level may be lowered 9'elow the minimum normal water leve)(Elevation 302'9").Whenever the reactor.water level is to be lowered below the low-low-low level setpoint redundant instrumentation will be provided to monitor the reactor water level.840406032i 840402 PDR ADOCK 05000220 P PDR | | {{#Wiki_filter:SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING |
| \1~
| | : c. The neutron flux shall not exceed its d. The reactor water low level scram trip scram setting for longer than 1.5 seconds setting shall be no lower than -12 inches as indicated by the process computer. (53 inches indicator scale) elative to When the process computer is out of the minimum normal water level (302'9"). |
| BASES FOR 2.1.1 FUEL CLADDING-SAFETY LIMIT During periods when the reactor is shut down, consideration must also be given to water level requirements, due to the effect of decay heat.If reactor water level should drop below the top of the active fuel during this time, the ability to cool the core is reduced.This reduction in core cooling capability could lead to elevated cladding temperatures and clad perforation.
| | service, a safety limit violation shall be assumed if the neutron flux exceeds the e. The reactor water low-low level setting scram setting and control rod scram does for core spray initiation shall be no less not occur. than -5 feet (5 inches indicator scale) relative to the minimum normal water level To ensure that the Safety Limit (Elevation 302'9"). |
| The core will be cooled sufficiently to prevent clad melting should the water level be reduced to two-thirds of the core height.The lowest point at which the reactor water level can normally be monitored is approximately 7 feet 11 inches below minimum normal water level or 4 feet 8 inches above the top of the active fuel.This is the location of the reactor vessel tap for the low-low-low water level instrumentation.
| | established in Specifications 2.1.la and 2.l.lb is not exceeded, each required f. ~ |
| The actual low-low-low water level trip point is 6 feet 3 inches (-10 inches indicator scale)below minimum norma)water level (Elevation 302'-9").The 20 inch difference resulted from an evaluation of the recomnendations contained in General Electric Service Information Letter 299"High Drywell Temperature Effect on Reactor Vessel Water Level Instrumentation." The low-low-low water level trip point was raised 20 inches to conservatively account for possible differences in actual to indicated water level due to potentially high drywell temperatures.
| | The flow biased APRM rod block trip scram shall be initiated by its expected settings shall be less than or equal to scram signal. The Safety Limit shall be that shown in Figure 2.l.l. |
| The safety limit has been established here to provide a point which can be monitored and also can provide adequate margin.However, for performing major maintenance as specified in Specification 2.1.l.e, redundant instrumentation will be provided for monitoring reactor water level below the low-low-low water level set point.(For example, by installing temporary instrument lines and reference points to redundant level transmitters so that the reactor water level may he monitored over the required range.)In addition written procedures, which identify all the valves which have the potential of lowering the water level inadvertently, are established to prevent their operation during the major maintenance which requires the water level to he below the low-low level set point.The thermal power transient resulting when a scram is accomplished other than by the expected scram signal (e.g., scram from neutron flux following closure of the main turbine stop valves)does not necessarily cause fuel damage.However, for this specification a safety limit violation wi]l be assumed when a scram is only accomplished by means of a backup feature of the plant design.The concept of nqt approaching a safety limit provided scram signals are operable is supported by the extensive plant safety analysis.t 13
| | assumed to be exceeded when scram is accomplished by a means other than the expected scram signal. |
| | : d. Whenever the reactor is in the shutdown condition with irradiated fuel in the reactor vessel, the water level shall not be more than 6 feet, 3 inches (-10 inches indicator scale) below minimum normal water level (Elevation 302'9") except as specifed in "e" below. |
| | : e. For the purpose of performing major maintenance (not to exceed 12 weeks in duration) on the reactor vessel; the reactor water level may be lowered the minimum normal water leve) 9'elow (Elevation 302'9"). Whenever the reactor |
| | .water level is to be lowered below the low-low-low level setpoint redundant instrumentation will be provided to monitor the reactor water level. |
| | 840406032i 840402 PDR ADOCK 05000220 P PDR |
|
| |
|
| REFERENCES FOR BASES 2.1.1 AND 2.1.2 FUEL CLADDING (1)General Electric BHR Thermal Analysis Basis (GETAB)Data, Correlation and Design Application, NEDO-10958 and NE DE-10958.(2)Linford, R.B.,"Analytical Methods of Plant Transient Evaluations for the General Electric Boiling plater Reactor," NED0-10801, February 1973.(3)FSAR, Volume II, Appendix E.(4)FSAR, Second Supplement.
| | \ |
| (5)FSAR, Volume II, Appendix E.(6)FSAR, Second Supplement.
| | 1 ~ |
| (7)Letters, Peter A.Horr is, Director of Reactor Licensing, USAEC, to John E.Logan, Vice-President, Jersey Central Power and Light Company, dated November 22, 1967 and January 9, 1968.(8)Technical Supplement to Petition to Increase Power Level, dated April 1970.(9)Letter, T.J.Brosnan, Niagara Mohawk Power Corporation, to Peter A.Morris, Division of Reactor Licensing, USAEC, dated February 28, 1972.(10)Letter, Philip D.Raymond, Niagara Mohawk Power Corporation, to A.Giambusso, USAEC, dated October 15, 1973.-(ll)Nine Mile Point Nuclear Power Station Unit 1 Load Line Limit Analysis, NEDO 24012, May, 1977.(12)Licensing Topical Report General Electric Boiling Mater Reactor Generic Reload Fuel Application, NEOE-24011-P-A, August, 1978.(13)Nine Mile Point Nuclear Power Station UJ)it 1, Extended Load Line Limit Analysis, License A)))end)vent Submittal (Cycle 6), NED0-24185, April 1979.(14)General Electric SIL 299"High Oryuel1 Temperature Effect on Reactor Vessel Water Level Instrusientation." 20
| |
|
| |
|
| LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT c~If a redundant component in each of the core spray systems becomes inoperable, both systems shall be considered operable provided that the.component is returned to an operable condition within 7 days and the additional surveillance required is performed.
| | BASES FOR 2.1.1 FUEL CLADDING SAFETY LIMIT During periods when the reactor is shut down, consideration must also be given to water level requirements, due to the effect of decay heat. If reactor water level should drop below the top of the active fuel during this time, the ability to cool the core is reduced. This reduction in core cooling capability could lead to elevated cladding temperatures and clad perforation. The core will be cooled sufficiently to prevent clad melting should the water level be reduced to two-thirds of the core height. |
| d.If a core spray system becomes inoperable and all the components are operable in the other system, the reactor may remain in operation for a period not to exceed 7 days.check calibrate test Once/day Once/3 months Once/3 months d.Core spray header<P instrumentation e.If Specifications a, b, c and d are not met, a normal orderly shutdown shall be initiated within one hour and the reactor shall be in the cold shutdown condition within ten hours.If both core spray systems become inoperable the reactor shall be in the cold shutdown condition within ten hours and no work (except as specified in"f" and"h" below)shall be performed on the reactor or its connected systems which could result in lowering the reactor water level to more than six feet, three inches below minimum normal water level (-10 inches indicator scale).e.Surveillance with Inoperable Components llhen a component or system becomes inoperable its redundant component or system shall be demonstrated to be operable immediately and daily thereafter.
| | The lowest point at which the reactor water level can normally be monitored is approximately 7 feet 11 inches below minimum normal water level or 4 feet 8 inches above the top of the active fuel. This is the location of the reactor vessel tap for the low-low-low water level instrumentation. The actual low-low-low water level trip point is 6 feet 3 inches (-10 inches indicator scale) below minimum norma) water level (Elevation 302'-9"). The 20 inch difference resulted from an evaluation of the recomnendations contained in General Electric Service Information Letter 299 "High Drywell Temperature Effect on Reactor Vessel Water Level Instrumentation." The low-low-low water level trip point was raised 20 inches to conservatively account for possible differences in actual to indicated water level due to potentially high drywell temperatures. The safety limit has been established here to provide a point which can be monitored and also can provide adequate margin. However, for performing major maintenance as specified in Specification 2.1. l.e, redundant instrumentation will be provided for monitoring reactor water level below the low-low-low water level set point. (For example, by installing temporary instrument lines and reference points to redundant level transmitters so that the reactor water level may he monitored over the required range.) In addition written procedures, which identify all the valves which have the potential of lowering the water level inadvertently, are established to prevent their operation during the major maintenance which requires the water level to he below the low-low level set point. |
| f.Surveillance during control rod'drive maintenance which is simultaneous with the suppression chamber unwatered shall include at least hourly checks that the conditions listed in 3.1.4f are met.
| | The thermal power transient resulting when a scram is accomplished other than by the expected scram signal (e.g., |
| | scram from neutron flux following closure of the main turbine stop valves) does not necessarily cause fuel damage. |
| | However, for this specification a safety limit violation wi]l be assumed when a scram is only accomplished by means of a backup feature of the plant design. The concept of nqt approaching a safety limit provided scram signals are operable is supported by the extensive plant safety analysis. |
| | t 13 |
|
| |
|
| LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREf1ENT h.For the purpose of performing major maintenance (not to exceed 12 weeks in duration)on the reactor vessel, the reactor water level may be lowered to 9'elow the minimum normal water level (elevation 302'9").Whenever the reactor water level is to be lowered below the low-.low-low level set point redundant isntrumentation will be provided to monitor the reactor water level and written procedures will be developed and followed whenever the reactor water level is lowered below the low-low level set point.The procedures will define the valves that will be used to lower the vessel water level.All other vaves that have the potential of lowering the vessel water level will be identified by valve number in the procedures and these valves will be red tagged to preclude their operation during the major maintenance with the water level below the low-low level set point.During the period of major maintenance requiring lowering the water level to more than 6 feet, 3 inches below minimum normal water level (-10 inches indicator scale), either both Core Spray Systems must be operable or, if one Core Spray System is inoperable because of the maintenance, all of the redundant components of the other Core Spray System must be operable.53a
| | REFERENCES FOR BASES 2.1.1 AND 2.1.2 FUEL CLADDING (1) General Electric BHR Thermal Analysis Basis (GETAB) Data, Correlation and Design Application, NEDO-10958 and NE DE-10958. |
| | (2) Linford, R. B., "Analytical Methods of Plant Transient Evaluations for the General Electric Boiling plater Reactor," NED0-10801, February 1973. |
| | (3) FSAR, Volume II, Appendix E. |
| | (4) FSAR, Second Supplement. |
| | (5) FSAR, Volume II, Appendix E. |
| | (6) FSAR, Second Supplement. |
| | (7) Letters, Peter A. Horr is, Director of Reactor Licensing, USAEC, to John E. Logan, Vice-President, Jersey Central Power and Light Company, dated November 22, 1967 and January 9, 1968. |
| | (8) Technical Supplement to Petition to Increase Power Level, dated April 1970. |
| | (9) Letter, T. J. Brosnan, Niagara Mohawk Power Corporation, to Peter A. Morris, Division of Reactor Licensing, USAEC, dated February 28, 1972. |
| | (10) Letter, Philip D. Raymond, Niagara Mohawk Power Corporation, to A. Giambusso, USAEC, dated October 15, 1973.- |
| | (ll) Nine Mile Point Nuclear Power Station Unit 1 Load Line Limit Analysis, NEDO 24012, May, 1977. |
| | (12) Licensing Topical Report General Electric Boiling Mater Reactor Generic Reload Fuel Application, NEOE-24011-P-A, August, 1978. |
| | (13) Nine Mile Point Nuclear Power Station UJ)it 1, Extended Load Line Limit Analysis, License A)))end)vent Submittal (Cycle 6), NED0-24185, April 1979. |
| | (14) General Electric SIL 299 "High Oryuel1 Temperature Effect on Reactor Vessel Water Level Instrusientation." |
| | 20 |
|
| |
|
| BASES FOR 3.1.5 AND 4.1.5 SOLENOID-ACTUATED PRESSURE RELIEF VALVES Pressure Bl owdown In the event of a small line break, substantial coolant loss could occur from the reactor vessel while it was stil I at relatively high pressures.
| | LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT c~ If a redundant component in each of the core spray systems becomes inoperable, both systems shall be considered operable provided that the .component is returned to an operable condition within 7 days and the additional surveillance required is performed. |
| A pressure blowdown system is provided which in conjunction with the core spray system will prevent significant fuel damage for all sized line breaks (Appendix E-11.2.0*).
| | : d. If a core spray system becomes inoperable d. Core spray header < P instrumentation and all the components are operable in the other system, the reactor may remain in check Once/day operation for a period not to exceed 7 calibrate Once/3 months days. test Once/3 months |
| Operation of three solenoid-actuated pressure relief valves is sufficient to depressurize the primary system to 110 psig which wi 11 permit full flow of the core spray system within required time limits (Appendix E-11.2~).Requiring-'all six of the relief valves to be operable, therefore, provides twice the minimum number required.Prior to or following refueling at low reactor pressure, each v'alve will be<aanually opened to verify valve operability.
| | : e. If Specifications a, b, c and d are not met, e. Surveillance with Inoperable Components a normal orderly shutdown shall be initiated within one hour and the reactor shall be in llhen a component or system becomes the cold shutdown condition within ten inoperable its redundant component or system hours. shall be demonstrated to be operable immediately and daily thereafter. |
| The malfunction analysis (Section II.XV,"Technica'I Supplement to Petition to Increase Power Level,"<late<i April 1970).demonstrates that no serious consequences result if one valve fails to close since the resulting blowdown is well.within design limits.In the event of small line break, considerable time is available for the operator to permit core spray operation by manually depressurizing the vessel using the solenoid-actuated valves.However, to ensure that the depressurization will be accomplished, automatic features are provided.The relief valves shall be capable of automatic initiation from simultaneous low-low-low water level (6 feet, 3 inches below minimum normal water level at Elevation 302'",-10 inches indicator scale)and high containment pressure (3.5 psig).The system response to small breaks requiring depressurization is discussed in Section VII-.A.3.3*
| | If both core spray systems become inoperable the reactor shall be in the cold shutdown f. Surveillance during control rod 'drive condition within ten hours and no work maintenance which is simultaneous with the (except as specified in "f" and "h" below) suppression chamber unwatered shall include shall be performed on the reactor or its at least hourly checks that the conditions connected systems which could result in listed in 3.1.4f are met. |
| and the time available to take operator action is summarized in Table VII-1*.Additional information is included in the answers to guestions III-1 and III-5 of the First Supplement.
| | lowering the reactor water level to more than six feet, three inches below minimum normal water level (-10 inches indicator scale). |
| ~Steam from the reactor vessel is discharged to the suppression chamber during valve testing.Conducting the tests with the reactor at low pressure such as just prior to or just after refueling minimizes the stress on the reactor coo 1 an t sys tern.a The test interval of once per operating cycle results in a system failure probability of 7.0 x 10-7 (Fifth Supplement, p.115)and is consistent with practical consideration.
| |
| *FSAR 59 E
| |
| '
| |
| LIMITING CONDITION FOR OPERATION SURVEILLANCE RE(UIREtiENT c.If a redundant component in each of the containment spray systems or their associated raw water systems become inoperable, both systems shall be considered operable provided that the component is returned to an operable condition within 7 days and that the additional surveillance required is performed. | |
| C.Raw Water Cool ing Pumps At least once per quarter manual startup and operability of the raw water cooling pumps shall be demonstrated.
| |
| d.If a containment spray system or its associated raw water system becomes inoperable and all the components are operable in the other systems, the reactor may remain in operation for a period not to exceed 7 days.d.Surveillance with Inoperable Components | |
| ,/hen a component or system becomes inoperable its redundant component or system shall be demonstrated to be operable immediately and daily thereafter.
| |
| e.If Specifirations"a" or"b" are not met, shutdown shall begin within one hour and the reactor coolant shall be below 215F within ten hours.If both containment spray systems become inoperable the reactor shall be in the cold shutdown condition within ten hours and no work (except as specified in"f" below)shall be performed on the reactor which could result in lowering the reactor water level to more than six feet, three inches (-10 inches indicator scale)below minimum normal water level: (Elevation 302'").e.Surveillance during control rod drive maintenance which is simultaneous with the suppression chamber unwatered shall include at least hourly checks that the conditions listed in 3.3.7.f are met.159 QJ E C (D E
| |
|
| |
|
| Tab 1 e 3.6.2f INSTRUMENTATION THAT INITIATES AUTO DEPRESSUR IZATION Liiig ii O>>i Parameter Minimum No.of Tripped or Operable~Tri Systems 4 Minimum No.of Operable Instrument Channels per Operable T~ri System Set-Point Reactor Mode Switch Position in Which Function Must Be Operable INITIATION (1)a.b.Low-Low-Low Reactor Water Level High Orywell Pressure 2 (a)2 (a)2 (a)~-10 inches*(Indicator scale)<3.5 psig (b)(b)(b)x (b)x*greater than (>)means less negative 213}} | | LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREf1ENT |
| | : h. For the purpose of performing major maintenance (not to exceed 12 weeks in duration) on the reactor vessel, the reactor water level may be lowered to 9'elow the minimum normal water level (elevation 302'9"). Whenever the reactor water level is to be lowered below the low-.low-low level set point redundant isntrumentation will be provided to monitor the reactor water level and written procedures will be developed and followed whenever the reactor water level is lowered below the low-low level set point. |
| | The procedures will define the valves that will be used to lower the vessel water level. All other vaves that have the potential of lowering the vessel water level will be identified by valve number in the procedures and these valves will be red tagged to preclude their operation during the major maintenance with the water level below the low-low level set point. |
| | During the period of major maintenance requiring lowering the water level to more than 6 feet, 3 inches below minimum normal water level (-10 inches indicator scale), |
| | either both Core Spray Systems must be operable or, if one Core Spray System is inoperable because of the maintenance, all of the redundant components of the other Core Spray System must be operable. |
| | 53a |
| | |
| | BASES FOR 3.1.5 AND 4.1.5 SOLENOID-ACTUATED PRESSURE RELIEF VALVES Pressure Bl owdown In the event of a small line break, substantial coolant loss could occur from the reactor vessel while it was stil I at relatively high pressures. A pressure blowdown system is provided which in conjunction with the core spray system will prevent significant fuel damage for all sized line breaks (Appendix E-11.2.0*). |
| | Operation of three solenoid-actuated pressure relief valves is sufficient to depressurize the primary system to 110 psig which wi 11 permit full flow of the core spray system within required time limits (Appendix E-11.2~). Requiring |
| | -'all six of the relief valves to be operable, therefore, provides twice the minimum number required. Prior to or following refueling at low reactor pressure, each v'alve will be <aanually opened to verify valve operability. The malfunction analysis (Section II.XV, "Technica'I Supplement to Petition to Increase Power Level," <late<i April 1970) |
| | .demonstrates that no serious consequences result if one valve fails to close since the resulting blowdown is well. |
| | within design limits. |
| | In the event of small line break, considerable time is available for the operator to permit core spray operation by manually depressurizing the vessel using the solenoid-actuated valves. However, to ensure that the depressurization will be accomplished, automatic features are provided. The relief valves shall be capable of automatic initiation from simultaneous low-low-low water level (6 feet, 3 inches below minimum normal water level at Elevation 302'", -10 inches indicator scale) and high containment pressure (3.5 psig). The system response to small breaks requiring depressurization is discussed in Section VII-.A.3.3* and the time available to take operator action is summarized in Table VII-1*. Additional information is included in the answers to guestions III-1 and III-5 of the First Supplement. |
| | ~ |
| | Steam from the reactor vessel is discharged to the suppression chamber during valve testing. Conducting the tests with the reactor at low pressure such as just prior to or just after refueling minimizes the stress on the reactor coo 1 an t sys tern. a The test interval of once per operating cycle results in a system failure probability of 7.0 x 10-7 (Fifth Supplement, p. 115) and is consistent with practical consideration. |
| | * FSAR 59 E |
| | |
| | LIMITING CONDITION FOR OPERATION SURVEILLANCE RE(UIREtiENT |
| | : c. If a redundant component in each of the C. Raw Water Cool ing Pumps containment spray systems or their associated raw water systems become At least once per quarter manual startup and inoperable, both systems shall be considered operability of the raw water cooling pumps operable provided that the component is shall be demonstrated. |
| | returned to an operable condition within 7 days and that the additional surveillance required is performed. |
| | : d. If a containment spray system or its d. Surveillance with Inoperable Components associated raw water system becomes inoperable and all the components are ,/hen a component or system becomes operable in the other systems, the reactor inoperable its redundant component or system may remain in operation for a period not to shall be demonstrated to be operable exceed 7 days. immediately and daily thereafter. |
| | : e. If Specifirations "a" or "b" are not met, e. Surveillance during control rod drive shutdown shall begin within one hour and the maintenance which is simultaneous with the reactor coolant shall be below 215F within suppression chamber unwatered shall include ten hours. at least hourly checks that the conditions listed in 3.3.7.f are met. |
| | If both containment spray systems become inoperable the reactor shall be in the cold shutdown condition within ten hours and no work (except as specified in "f" below) shall be performed on the reactor which could result in lowering the reactor water level to more than six feet, three inches (-10 inches indicator scale) below minimum normal water level: |
| | (Elevation 302'"). |
| | 159 QJ E |
| | C (D |
| | E |
| | |
| | Tab 1 e 3.6. 2f INSTRUMENTATION THAT INITIATES AUTO DEPRESSUR IZATION Liiig ii O>>i 4 |
| | Minimum No. of Minimum No. Operable Instrument Reactor Mode Switch of Tripped or Channels per Position in Which Operable Operable Function Must Be Parameter ~Tri Systems T~ri System Set-Point Operable INITIATION (1) a. Low-Low-Low Reactor Water (a) (a) ~ -10 inches* |
| | 2 2 (b) (b) x Level ( Indicator scale) |
| | : b. High Orywell 2 (a) < 3.5 psig (b) (b) x Pressure |
| | * greater than ( >) means less negative 213}} |
Revised Pages to Tech Specs Re Triple Low Reactor Water Level Setpoint.Changes Involve Replacing 147.1 Inch Indicator Scale w/-10 Inch Indicator Scale| ML18038A666 |
| Person / Time |
|---|
| Site: |
Nine Mile Point  |
|---|
| Issue date: |
04/02/1984 |
|---|
| From: |
NIAGARA MOHAWK POWER CORP. |
|---|
| To: |
|
|---|
| Shared Package |
|---|
| ML17054A598 |
List: |
|---|
| References |
|---|
| NUDOCS 8404060321 |
| Download: ML18038A666 (16) |
|
|
|---|
Category:TECHNICAL SPECIFICATIONS
MONTHYEARML17059C7911999-09-30030 September 1999
Proposed Tech Specs Re Rev B to Conversion of Unit 2 Current TS to ITS
ML20211P5931999-09-10010 September 1999
Marked-up Tech Specs Pages Reflecting Conforming Administrative License Amend Associated with Proposed Transfer of Facilities to Amergen Energy Co,Llc
ML20211H2231999-08-26026 August 1999
Proposed Tech Specs,Supporting Implementation of Noble Metal Chemical by Raising Reactor Water Conductivity Limit in TSs 3.2.3.a,3.2.3.c.1 & 3.2.3.b
ML20206P2661998-12-30030 December 1998
Proposed Tech Specs Making Application of Emergency Condenser Vent Ng Activity Monitor Channel Operability Requirement & Daily Sensor Check SR Consistent with Conditions Stated in LCO 3.1.3.a Re Emergency Cooling Sys
ML20197K1041998-12-0404 December 1998
Revised Proposed TS Bases Reflecting Previous Removal of Condenser Low Vacuum Scram Function from TS as Well as Plant Design & Changes to EDG Ratings,Design Basis Load Limits & Loading Profiles
ML20196H0671998-11-30030 November 1998
Proposed Tech Specs,Correcting LCO & Associated Bases for TS Section 3.1.2, Liquid Poison Sys, That Had Been Incorporated Into TS as Part of NMP Unit 1 TS Amend 101
ML20196D2081998-11-24024 November 1998
Proposed Tech Specs B 2.1.1, Fuel Cladding - Safety Limit, B 3.2.5 & 4.2.5, Reactor Coolant Sys Leakage Rate
ML20154R1101998-10-16016 October 1998
Proposed Ts,Revising LCO 3.7.1.1 & Associated Actions & SRs to Provide Assurance That Four SW Pumps Are Operable & Are Operating within Acceptable Sys Parameters,With Divisional Cross Connect Valves Open
ML20217N7911998-04-24024 April 1998
Revised Pages 2-1,3/4 4-1 & B2-1 to Replace Previously Submitted Pages Contained within Attachment a of 971215 Proposed Change to TS to License NPF-49
ML20202G3751998-02-0505 February 1998
Proposed Tech Specs Pages,Reflecting References to 10CFR50.55a(f) & (G) as Well as Terminology Used in Second Ten Year Isi/Ist,Beginning on 980405
ML20197H3371997-12-15015 December 1997
Proposed Tech Specs Revising SLMCPR from 1.07 to 1.09 for Two Recirculation Loop Operation & from 1.08 to 1.10 for Single Loop Operation
ML20211M5231997-10-0707 October 1997
Proposed Tech Specs Section 4.9.6,reflecting New Setpoints Due to Difference in Weights of Two Existing Triangular Refueling Platform Masts
ML20217G9011997-07-31031 July 1997
Proposed Tech Specs Changing Wording in Action 36 of TS Table 3.3.3-1, Emergency Core Cooling Sys Actuation Instrumentation
ML20148A7751997-04-30030 April 1997
Proposed Change to Tech Specs,Removing Section 3.3.7.3 & Associated Surveillance Section 4.3.7.3,associated Tables 3.3.7.3-1 & 4.3.7.3-1,Bases Section 3.4.3.7.3 & Revising Section 0.0 (Index) Pages VII & Xvii
ML20116D8121996-07-26026 July 1996
Proposed Tech Specs Re Option B of App J to 10CFR50
ML20115H1521996-07-12012 July 1996
Proposed Tech Specs Section 6.2.2.i Re App a
ML20117F9501996-05-15015 May 1996
Proposed Tech Specs Section 3/4.3.2 Re Isolation Actuation Instrumentation
ML20101F4281996-03-20020 March 1996
Proposed TS 3/4.3.1 Re RPS Instrumentation,Deleting Operability Requirements for APRM Neutron flux-upscale, Setdown & Inoperative Functions in Operational Conditions (Oc) 3 & 4 & Modifying Operability Requirements in Oc 5
ML20101D7551996-03-15015 March 1996
Proposed Tech Specs Section 4.6.2 Re Depressurization Sys - Suppression Pool
ML20097D2391996-02-0707 February 1996
Proposed Tech Specs Re Administrative Controls
ML20100C1021996-01-25025 January 1996
Proposed Tech Specs Table 3.3.3-1 Re Emergency Core Cooling Sys Actuation Instrumentation
ML20096G8401996-01-17017 January 1996
Proposed Tech Specs,Representing Revs to Specs 3/4.3.1, 3/4.3.2,3/4.3.3,3/4.3.4.2 & Associated Bases to Relocate Response Time Limit Tables from TSs to Plant USAR
ML20094B5421995-10-25025 October 1995
Proposed Tech Specs Re Position Title Changes & Reassignments of Responsibilities at Upper Mgt Level
ML20091Q9571995-08-28028 August 1995
Proposed TS 3.6.1.7,increasing Time 12-inch & 14-inch Containment Purge Sys Supply & Exhaust Valves May Be Open in Operational Condition 1,2 & 3 from 90 H Per 365 Days to 135 H Per 365 Days & Deleting Expired Footnotes for Clarity
ML20081E2491995-03-0909 March 1995
Proposed Tech Specs,Revising SR 4.6.1.2.a,allowing Second Primary Containment ILRT (Type a) to Be Performed at Refueling Outage 5 or 72 Months After First Type a Test
ML20077N1591995-01-0606 January 1995
Proposed Tech Specs Re Min Gallons of Fuel Oil Required in Day Tanks & Storage Tanks
ML20077N2071995-01-0606 January 1995
Proposed Tech Specs Re Deletion of Certain Instruments Not Classified as Category 1
ML20078Q9241994-12-13013 December 1994
Proposed Tech Specs Re Change to Table 3.6.1.2-1 That Will Allow Maximum Leakage of 24.0 Scfh for Each of Eight MSIVs
ML20077E1561994-12-0202 December 1994
Proposed TS Page 3/4 1-20,reflecting Rev of SLCS Relief Valve Setpoint to Show Influence of Back Pressure & Rev of Bases Page B3/4 5-2 to Show That Hpsc Sys Designed to Supply 517 Gpm Flow Rate at 1,200 (Instead of 1,175) Psid
ML20078J3781994-11-14014 November 1994
Proposed Tech Specs Reducing Leak Rate Test Pressure for Safety Related ADS Nitrogen Receiving Tanks from 385 Psig to 365 Psig
ML20078E0331994-10-28028 October 1994
Proposed Tech Specs 1.0, Definitions, 3/4.3.2, Isolation Actuation Instrumentation & 3/4.9.3, Control Rod Position
ML20078B9841994-10-21021 October 1994
Proposed Tech Specs SR 4.8.1.1.2.e.8,reflecting Addition of Footnote Re 24 H Functional Test of DGs
ML20073L8781994-10-0505 October 1994
Proposed TS Table 3.3.7.1-1 Re Radiation Monitoring Instrumentation
ML20072U3491994-09-0202 September 1994
Proposed Tech Specs Re 18 Month Operability Test of Svc Water Pumps Sys & Resistance Test of Intake Deicing Heater Sys
ML17059A4471994-09-0101 September 1994
Proposed Tech Specs Section 3.2.2, Minimum Reactor Vessel Temperature for Pressurization, & Associated Bases
ML20072P6211994-08-26026 August 1994
Proposed TS 3/4.6.1.3, Primary Containment Air Locks, Allowing Continued Plant Operation If Interlock Becomes Inoperable as Long as Operable Door Locked Shut & Periodically Checked as Being Locked Shut
ML20071G2771994-07-0101 July 1994
Proposed Tech Specs Surveillance Requirements 4.6.5.1.c.1 & 4.6.5.1.c.2
ML20029D4001994-04-27027 April 1994
Proposed Tech Specs Re Drawdown Time Testing & Inleakage Testing for Secondary Containment Integrity
ML20058N3491993-12-14014 December 1993
Proposed Tech Specs 3/4.8.2, DC Sources, 3/4.8.4, Electrical Equipment Protective Devices & Bases for 3/4.6.3, Primary Containment Isolation Valves
ML20056G9391993-09-0202 September 1993
Proposed Tech Specs Section 3/4.1.3.5, CR Scram Accumulators
ML20056G1951993-08-27027 August 1993
Proposed Tech Specs,Revising TS 4.8.1.1.2.e, AC Sources - Operating & Adding TS 4.8.1.1.2.f
ML20044G6261993-05-26026 May 1993
TS Section 6.8 Re Review & Approval Process for Procedure Changes
ML20044F7421993-05-21021 May 1993
Proposed Tech Specs Pages 2-3,3/4 2-2,3/4 3-60,3/4 3-62, 3/4 3-63,3/4 3-64,3/4 3-65 & 6-22 & Bases Page B3/4 2-1
ML20044F3041993-05-19019 May 1993
Proposed Tech Specs Sections 3.4.3.1 & 3.4.3.2 Re Generic Ltr 88-01,drywell Leak Detection Requirements
ML17058B7751993-05-14014 May 1993
Proposed Tech Specs,Reflecting Editorial Changes, Administrative Corrections & Retyping of TS
ML17056C3281993-03-29029 March 1993
Proposed TS Pages 139 & 140 Re LCO SR for Type a & Local Leak Rate Type B & C Tests
ML20012G5441993-02-27027 February 1993
Proposed TS Sections 1.0, Definitions & 3/4.3.6, Control Rod Block Instrumentation.
ML17056C2741993-02-18018 February 1993
Proposed TS Table 3.2.7 Re RCS Isolation Valves,Table 3.2.7.1 Re Primary Coolant Sys Pressure Isolation Valves & Table 3.3.4 Re Primary Containment Isolation Valves,Lines Entering Free Space of Containment
ML17056C2571993-02-12012 February 1993
Proposed,Revised TS Pages 204,204a,207,225a & 230 to Clarify Operator Actions in Event of Loss of Two Instrument Channels
ML20126H9941992-12-30030 December 1992
Proposed Tech Specs 3.4.3.1 & 3.4.3.2 Re RCS Leakage Detection Sys & Operational Leakage to Conform W/ Recommendations of Generic Ltr 88-01
1999-09-30
[Table view]
Category:TECHNICAL SPECIFICATIONS & TEST REPORTS
MONTHYEARML17059C7911999-09-30030 September 1999
Proposed Tech Specs Re Rev B to Conversion of Unit 2 Current TS to ITS
ML20211P5931999-09-10010 September 1999
Marked-up Tech Specs Pages Reflecting Conforming Administrative License Amend Associated with Proposed Transfer of Facilities to Amergen Energy Co,Llc
ML20211H2231999-08-26026 August 1999
Proposed Tech Specs,Supporting Implementation of Noble Metal Chemical by Raising Reactor Water Conductivity Limit in TSs 3.2.3.a,3.2.3.c.1 & 3.2.3.b
ML18041A0731999-07-27027 July 1999
Rev 1 to NMP2-ISI-006, Second Ten Year Interval Inservice Insp Program Plan for Nine Mile Point Nuclear Power Station Unit 2.
ML20206P2661998-12-30030 December 1998
Proposed Tech Specs Making Application of Emergency Condenser Vent Ng Activity Monitor Channel Operability Requirement & Daily Sensor Check SR Consistent with Conditions Stated in LCO 3.1.3.a Re Emergency Cooling Sys
ML20197K1041998-12-0404 December 1998
Revised Proposed TS Bases Reflecting Previous Removal of Condenser Low Vacuum Scram Function from TS as Well as Plant Design & Changes to EDG Ratings,Design Basis Load Limits & Loading Profiles
ML20196H0671998-11-30030 November 1998
Proposed Tech Specs,Correcting LCO & Associated Bases for TS Section 3.1.2, Liquid Poison Sys, That Had Been Incorporated Into TS as Part of NMP Unit 1 TS Amend 101
ML20196D2081998-11-24024 November 1998
Proposed Tech Specs B 2.1.1, Fuel Cladding - Safety Limit, B 3.2.5 & 4.2.5, Reactor Coolant Sys Leakage Rate
ML20154R1101998-10-16016 October 1998
Proposed Ts,Revising LCO 3.7.1.1 & Associated Actions & SRs to Provide Assurance That Four SW Pumps Are Operable & Are Operating within Acceptable Sys Parameters,With Divisional Cross Connect Valves Open
ML20217N7911998-04-24024 April 1998
Revised Pages 2-1,3/4 4-1 & B2-1 to Replace Previously Submitted Pages Contained within Attachment a of 971215 Proposed Change to TS to License NPF-49
ML20202G3751998-02-0505 February 1998
Proposed Tech Specs Pages,Reflecting References to 10CFR50.55a(f) & (G) as Well as Terminology Used in Second Ten Year Isi/Ist,Beginning on 980405
ML20197H3371997-12-15015 December 1997
Proposed Tech Specs Revising SLMCPR from 1.07 to 1.09 for Two Recirculation Loop Operation & from 1.08 to 1.10 for Single Loop Operation
ML20211M5231997-10-0707 October 1997
Proposed Tech Specs Section 4.9.6,reflecting New Setpoints Due to Difference in Weights of Two Existing Triangular Refueling Platform Masts
ML20217G9011997-07-31031 July 1997
Proposed Tech Specs Changing Wording in Action 36 of TS Table 3.3.3-1, Emergency Core Cooling Sys Actuation Instrumentation
ML20148A7751997-04-30030 April 1997
Proposed Change to Tech Specs,Removing Section 3.3.7.3 & Associated Surveillance Section 4.3.7.3,associated Tables 3.3.7.3-1 & 4.3.7.3-1,Bases Section 3.4.3.7.3 & Revising Section 0.0 (Index) Pages VII & Xvii
ML17059B2911996-09-0404 September 1996
NMP Nine Mile Point Nuclear Station Unit 2,Pump & Valve First Ten-Yr Inservice Testing Program Plan.
ML20116D8121996-07-26026 July 1996
Proposed Tech Specs Re Option B of App J to 10CFR50
ML20115H1521996-07-12012 July 1996
Proposed Tech Specs Section 6.2.2.i Re App a
ML20117F9501996-05-15015 May 1996
Proposed Tech Specs Section 3/4.3.2 Re Isolation Actuation Instrumentation
ML20101F4281996-03-20020 March 1996
Proposed TS 3/4.3.1 Re RPS Instrumentation,Deleting Operability Requirements for APRM Neutron flux-upscale, Setdown & Inoperative Functions in Operational Conditions (Oc) 3 & 4 & Modifying Operability Requirements in Oc 5
ML20101D7551996-03-15015 March 1996
Proposed Tech Specs Section 4.6.2 Re Depressurization Sys - Suppression Pool
ML20097D2391996-02-0707 February 1996
Proposed Tech Specs Re Administrative Controls
ML20100C1021996-01-25025 January 1996
Proposed Tech Specs Table 3.3.3-1 Re Emergency Core Cooling Sys Actuation Instrumentation
ML20096G8401996-01-17017 January 1996
Proposed Tech Specs,Representing Revs to Specs 3/4.3.1, 3/4.3.2,3/4.3.3,3/4.3.4.2 & Associated Bases to Relocate Response Time Limit Tables from TSs to Plant USAR
ML20094B5421995-10-25025 October 1995
Proposed Tech Specs Re Position Title Changes & Reassignments of Responsibilities at Upper Mgt Level
ML20091Q9571995-08-28028 August 1995
Proposed TS 3.6.1.7,increasing Time 12-inch & 14-inch Containment Purge Sys Supply & Exhaust Valves May Be Open in Operational Condition 1,2 & 3 from 90 H Per 365 Days to 135 H Per 365 Days & Deleting Expired Footnotes for Clarity
ML20081E2491995-03-0909 March 1995
Proposed Tech Specs,Revising SR 4.6.1.2.a,allowing Second Primary Containment ILRT (Type a) to Be Performed at Refueling Outage 5 or 72 Months After First Type a Test
ML17059B0221995-02-15015 February 1995
Rev 0 to UT-NMP-311V0, Procedure for Manual Ultrasonic Exam of Nozzles Inner Radius & Bore.
ML17059B0211995-02-10010 February 1995
Rev 0 to UT-NMP-309V0, Procedure for Manual Ultrasonic Exam of Planar Flaw Sizing for Nozzle Inner Radius & Bore Regions.
ML17059B0201995-02-10010 February 1995
Rev 0 to UT-NMP-703V0, Procedure for Geris 2000 Ultrasonic Exam of RPV Nozzle Inner Radius & Bore Regions.
ML17059A8881995-01-18018 January 1995
Rev 0 to Field Disposition Instruction 0245-90800, Shroud.
ML20077N1591995-01-0606 January 1995
Proposed Tech Specs Re Min Gallons of Fuel Oil Required in Day Tanks & Storage Tanks
ML20077N2071995-01-0606 January 1995
Proposed Tech Specs Re Deletion of Certain Instruments Not Classified as Category 1
ML20078Q9241994-12-13013 December 1994
Proposed Tech Specs Re Change to Table 3.6.1.2-1 That Will Allow Maximum Leakage of 24.0 Scfh for Each of Eight MSIVs
ML20077E1561994-12-0202 December 1994
Proposed TS Page 3/4 1-20,reflecting Rev of SLCS Relief Valve Setpoint to Show Influence of Back Pressure & Rev of Bases Page B3/4 5-2 to Show That Hpsc Sys Designed to Supply 517 Gpm Flow Rate at 1,200 (Instead of 1,175) Psid
ML20078J3781994-11-14014 November 1994
Proposed Tech Specs Reducing Leak Rate Test Pressure for Safety Related ADS Nitrogen Receiving Tanks from 385 Psig to 365 Psig
ML17059A5661994-11-0909 November 1994
Rev 2 to Emergency Plan Maint Procedure EPMP-EPP-08, Maint, Testing & Operation of Oswego County Prompt Notification Sys.
ML20078E0331994-10-28028 October 1994
Proposed Tech Specs 1.0, Definitions, 3/4.3.2, Isolation Actuation Instrumentation & 3/4.9.3, Control Rod Position
ML20078B9841994-10-21021 October 1994
Proposed Tech Specs SR 4.8.1.1.2.e.8,reflecting Addition of Footnote Re 24 H Functional Test of DGs
ML20073L8781994-10-0505 October 1994
Proposed TS Table 3.3.7.1-1 Re Radiation Monitoring Instrumentation
ML20072U3491994-09-0202 September 1994
Proposed Tech Specs Re 18 Month Operability Test of Svc Water Pumps Sys & Resistance Test of Intake Deicing Heater Sys
ML17059A4471994-09-0101 September 1994
Proposed Tech Specs Section 3.2.2, Minimum Reactor Vessel Temperature for Pressurization, & Associated Bases
ML20072P6211994-08-26026 August 1994
Proposed TS 3/4.6.1.3, Primary Containment Air Locks, Allowing Continued Plant Operation If Interlock Becomes Inoperable as Long as Operable Door Locked Shut & Periodically Checked as Being Locked Shut
ML17059A4241994-07-0606 July 1994
Rev 14 to Nine Mile Point Nuclear Station Unit 1 Odcm.
ML20071G2771994-07-0101 July 1994
Proposed Tech Specs Surveillance Requirements 4.6.5.1.c.1 & 4.6.5.1.c.2
ML20029D4001994-04-27027 April 1994
Proposed Tech Specs Re Drawdown Time Testing & Inleakage Testing for Secondary Containment Integrity
ML18040A2941993-12-17017 December 1993
Rev 13 to Nine Mile Point Nuclear Station Unit 1 Odcm.
ML20058N3491993-12-14014 December 1993
Proposed Tech Specs 3/4.8.2, DC Sources, 3/4.8.4, Electrical Equipment Protective Devices & Bases for 3/4.6.3, Primary Containment Isolation Valves
ML17059A0971993-10-25025 October 1993
Rev 4 to NMP2-IST-001,NMP,Nine Mile Point Nuclear Station, Unit 2,Pump & Valve First Ten-Yr Inservice Testing Program Plan.
ML20056G9391993-09-0202 September 1993
Proposed Tech Specs Section 3/4.1.3.5, CR Scram Accumulators
1999-09-30
[Table view] |
Text
SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING
- c. The neutron flux shall not exceed its d. The reactor water low level scram trip scram setting for longer than 1.5 seconds setting shall be no lower than -12 inches as indicated by the process computer. (53 inches indicator scale) elative to When the process computer is out of the minimum normal water level (302'9").
service, a safety limit violation shall be assumed if the neutron flux exceeds the e. The reactor water low-low level setting scram setting and control rod scram does for core spray initiation shall be no less not occur. than -5 feet (5 inches indicator scale) relative to the minimum normal water level To ensure that the Safety Limit (Elevation 302'9").
established in Specifications 2.1.la and 2.l.lb is not exceeded, each required f. ~
The flow biased APRM rod block trip scram shall be initiated by its expected settings shall be less than or equal to scram signal. The Safety Limit shall be that shown in Figure 2.l.l.
assumed to be exceeded when scram is accomplished by a means other than the expected scram signal.
- d. Whenever the reactor is in the shutdown condition with irradiated fuel in the reactor vessel, the water level shall not be more than 6 feet, 3 inches (-10 inches indicator scale) below minimum normal water level (Elevation 302'9") except as specifed in "e" below.
- e. For the purpose of performing major maintenance (not to exceed 12 weeks in duration) on the reactor vessel; the reactor water level may be lowered the minimum normal water leve) 9'elow (Elevation 302'9"). Whenever the reactor
.water level is to be lowered below the low-low-low level setpoint redundant instrumentation will be provided to monitor the reactor water level.
840406032i 840402 PDR ADOCK 05000220 P PDR
\
1 ~
BASES FOR 2.1.1 FUEL CLADDING SAFETY LIMIT During periods when the reactor is shut down, consideration must also be given to water level requirements, due to the effect of decay heat. If reactor water level should drop below the top of the active fuel during this time, the ability to cool the core is reduced. This reduction in core cooling capability could lead to elevated cladding temperatures and clad perforation. The core will be cooled sufficiently to prevent clad melting should the water level be reduced to two-thirds of the core height.
The lowest point at which the reactor water level can normally be monitored is approximately 7 feet 11 inches below minimum normal water level or 4 feet 8 inches above the top of the active fuel. This is the location of the reactor vessel tap for the low-low-low water level instrumentation. The actual low-low-low water level trip point is 6 feet 3 inches (-10 inches indicator scale) below minimum norma) water level (Elevation 302'-9"). The 20 inch difference resulted from an evaluation of the recomnendations contained in General Electric Service Information Letter 299 "High Drywell Temperature Effect on Reactor Vessel Water Level Instrumentation." The low-low-low water level trip point was raised 20 inches to conservatively account for possible differences in actual to indicated water level due to potentially high drywell temperatures. The safety limit has been established here to provide a point which can be monitored and also can provide adequate margin. However, for performing major maintenance as specified in Specification 2.1. l.e, redundant instrumentation will be provided for monitoring reactor water level below the low-low-low water level set point. (For example, by installing temporary instrument lines and reference points to redundant level transmitters so that the reactor water level may he monitored over the required range.) In addition written procedures, which identify all the valves which have the potential of lowering the water level inadvertently, are established to prevent their operation during the major maintenance which requires the water level to he below the low-low level set point.
The thermal power transient resulting when a scram is accomplished other than by the expected scram signal (e.g.,
scram from neutron flux following closure of the main turbine stop valves) does not necessarily cause fuel damage.
However, for this specification a safety limit violation wi]l be assumed when a scram is only accomplished by means of a backup feature of the plant design. The concept of nqt approaching a safety limit provided scram signals are operable is supported by the extensive plant safety analysis.
t 13
REFERENCES FOR BASES 2.1.1 AND 2.1.2 FUEL CLADDING (1) General Electric BHR Thermal Analysis Basis (GETAB) Data, Correlation and Design Application, NEDO-10958 and NE DE-10958.
(2) Linford, R. B., "Analytical Methods of Plant Transient Evaluations for the General Electric Boiling plater Reactor," NED0-10801, February 1973.
(3) FSAR, Volume II, Appendix E.
(4) FSAR, Second Supplement.
(5) FSAR, Volume II, Appendix E.
(6) FSAR, Second Supplement.
(7) Letters, Peter A. Horr is, Director of Reactor Licensing, USAEC, to John E. Logan, Vice-President, Jersey Central Power and Light Company, dated November 22, 1967 and January 9, 1968.
(8) Technical Supplement to Petition to Increase Power Level, dated April 1970.
(9) Letter, T. J. Brosnan, Niagara Mohawk Power Corporation, to Peter A. Morris, Division of Reactor Licensing, USAEC, dated February 28, 1972.
(10) Letter, Philip D. Raymond, Niagara Mohawk Power Corporation, to A. Giambusso, USAEC, dated October 15, 1973.-
(ll) Nine Mile Point Nuclear Power Station Unit 1 Load Line Limit Analysis, NEDO 24012, May, 1977.
(12) Licensing Topical Report General Electric Boiling Mater Reactor Generic Reload Fuel Application, NEOE-24011-P-A, August, 1978.
(13) Nine Mile Point Nuclear Power Station UJ)it 1, Extended Load Line Limit Analysis, License A)))end)vent Submittal (Cycle 6), NED0-24185, April 1979.
(14) General Electric SIL 299 "High Oryuel1 Temperature Effect on Reactor Vessel Water Level Instrusientation."
20
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT c~ If a redundant component in each of the core spray systems becomes inoperable, both systems shall be considered operable provided that the .component is returned to an operable condition within 7 days and the additional surveillance required is performed.
- d. If a core spray system becomes inoperable d. Core spray header < P instrumentation and all the components are operable in the other system, the reactor may remain in check Once/day operation for a period not to exceed 7 calibrate Once/3 months days. test Once/3 months
- e. If Specifications a, b, c and d are not met, e. Surveillance with Inoperable Components a normal orderly shutdown shall be initiated within one hour and the reactor shall be in llhen a component or system becomes the cold shutdown condition within ten inoperable its redundant component or system hours. shall be demonstrated to be operable immediately and daily thereafter.
If both core spray systems become inoperable the reactor shall be in the cold shutdown f. Surveillance during control rod 'drive condition within ten hours and no work maintenance which is simultaneous with the (except as specified in "f" and "h" below) suppression chamber unwatered shall include shall be performed on the reactor or its at least hourly checks that the conditions connected systems which could result in listed in 3.1.4f are met.
lowering the reactor water level to more than six feet, three inches below minimum normal water level (-10 inches indicator scale).
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREf1ENT
- h. For the purpose of performing major maintenance (not to exceed 12 weeks in duration) on the reactor vessel, the reactor water level may be lowered to 9'elow the minimum normal water level (elevation 302'9"). Whenever the reactor water level is to be lowered below the low-.low-low level set point redundant isntrumentation will be provided to monitor the reactor water level and written procedures will be developed and followed whenever the reactor water level is lowered below the low-low level set point.
The procedures will define the valves that will be used to lower the vessel water level. All other vaves that have the potential of lowering the vessel water level will be identified by valve number in the procedures and these valves will be red tagged to preclude their operation during the major maintenance with the water level below the low-low level set point.
During the period of major maintenance requiring lowering the water level to more than 6 feet, 3 inches below minimum normal water level (-10 inches indicator scale),
either both Core Spray Systems must be operable or, if one Core Spray System is inoperable because of the maintenance, all of the redundant components of the other Core Spray System must be operable.
53a
BASES FOR 3.1.5 AND 4.1.5 SOLENOID-ACTUATED PRESSURE RELIEF VALVES Pressure Bl owdown In the event of a small line break, substantial coolant loss could occur from the reactor vessel while it was stil I at relatively high pressures. A pressure blowdown system is provided which in conjunction with the core spray system will prevent significant fuel damage for all sized line breaks (Appendix E-11.2.0*).
Operation of three solenoid-actuated pressure relief valves is sufficient to depressurize the primary system to 110 psig which wi 11 permit full flow of the core spray system within required time limits (Appendix E-11.2~). Requiring
-'all six of the relief valves to be operable, therefore, provides twice the minimum number required. Prior to or following refueling at low reactor pressure, each v'alve will be <aanually opened to verify valve operability. The malfunction analysis (Section II.XV, "Technica'I Supplement to Petition to Increase Power Level," <late>i 4
Minimum No. of Minimum No. Operable Instrument Reactor Mode Switch of Tripped or Channels per Position in Which Operable Operable Function Must Be Parameter ~Tri Systems T~ri System Set-Point Operable INITIATION (1) a. Low-Low-Low Reactor Water (a) (a) ~ -10 inches*
2 2 (b) (b) x Level ( Indicator scale)
- b. High Orywell 2 (a) < 3.5 psig (b) (b) x Pressure
- greater than ( >) means less negative 213