ML20246L313
| ML20246L313 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 05/09/1989 |
| From: | Butler W Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20246L318 | List: |
| References | |
| NUDOCS 8905180293 | |
| Download: ML20246L313 (26) | |
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NUCLEAR REGULATORY COMMISSION 3 * \\l* / /
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s.~.. y PUBLIC SERVICE ELECTRIC & GAS COMPANY PHILADELPHIA ELECTRIC COMPANY DELMARVA POWER AND LIGHT COMPANY ATLANTIC CITY ELECTRIC COMPANY DOCKET NO. 50-272 SALEM GENERATING STATION, UNIT NO. 1 4
AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.96 License No. DPR 1 The Nuclear Regulatory Comission (the Comission or the NRC) has found that:
A.
The application for amendment filed by the Public Service Electric &
Gas Company, Philadelphia Electric Company,(Delmarva Power and Light Company and Atlantic City Electric Company thelicensees) dated December 30, 1988 and supplemented by letters dated April 19 and May 4, 1989, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the 4
provisions of the Act, and the rules and regulations of the Comission; C.
There is reasonable assurance: (i) +S t the activities authorized by this amendment can be conducted wit a t endangering the health and safety of the public, and (ii) that such activities will be conducted in coripliance with the Commissicn's regulations set forth in 10 CFR Chapter I; D.
The issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifica-tions as indicated in the attachment to this license amendment, and peragraph 2.C.(2) of Facility Operating License No. DPR-70 is hereby amended to read as follows:
8905180293 890509 PDR ADOCK 05000272 P
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% (2).TechnicalSpecificationsandEnvironmentalProtectionPlan The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 96, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of its date of issuance.
FOR THE. NUCLEAR REGULATORY COMMISSION
/s/
Walter R. Butler, Director Project Directorate I-2 Division of Reactor Projects I/II
Attachment:
Changes to the Technical Specifications Date of Issuance: May 9, 1989 al h
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2-(2) Technical Specifications and Environmental Protection Plan The Technicti Specifications contained in Appendices A and B,_as revised through Amendment No. 96, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
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3.
This license amendment is effective as of its date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION y 'Y U
Walter R. Butler, Director Project Directorate I-2 Division of Reactor Projects I/II
Attachment:
Changes to the Technical Specifications Date of Issuance: May 9, 1989 i
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ATTACHMENT TO LICENSE AMENDMENT NO. og 1
FACILITY OPERATING LICENSE NO. DPR-70 DOCKET NO. 50-272 Revise Appendix A as follows:
Remove Pages Insert Foges B 2-1 B 2-1 B 2-4 B 2-4 B 2-6 B 2-6 3/4 1-21 3/4 1-21 3/4 2-9 3/4 2-9 3/4 2-14 3/4 2-14 8 3/4 2-6 B 3/4 2-6
2.1 SATETY LIMITS BASES 2.1.1 REACTOR CORE The restrictions of this safety limit prevent overheating of the fuel anti possible cladding perforation which would result in the release of fission products to the reactor coolant. Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature.
Operation above the upper boundary of the nucleate boiling regime could result in excessive cladding temperatures because of the onset of departure from nucleate boiling (DNB) and the resultan't sharp reduction in heat transfer coefficient. DNB is not a directly measurable parameter during operation and therefore THERMAL, POWER and Re.2ctor Coolant Temperature and Pressure have been related to DNB through the W-3, R-Grid correlation for standard (LOPAR) fuel assemblies and WRB-1 correlation for Vantage SH fuel assemblies. The DNB correlation has been developed to predict the DNB flux and the location of DNB for axially uniform and non-uniform heat flux distributions. The local DNB heat flux ratio, DNBR, defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux, is indicative of the margin to DNB.
The DNB design basis is as follows: there must be at least a 95 percent probability that the minimum DNBR of the limiting rod during condition I and II events is greater than or equal to the DNBR limit of the DNB correlation being used (the WRB-1 or W-3 R-Grid correlation). The correlation DNER limit is established based on the entire applicable experimental data set such that there is a 95 percent probability with 95 percent confidence that DNB will not occur when the minimum DNBR is at the DNER correlation limit (1.17 for the WRB-1 or 1.30 for the W-3 R-Grid).
The curves of Figures 2.1-1 and 2.1-2 show the loci of points of THERMAL POWER, Reactor Coolant System pressure and average temperature for which the minimum DNER is no less than the design DNBR value, or the average enthalpy at l the vessel exit is equal to the enthalpy of saturated liquid.
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l SALEM - UNIT 1 B 2-1 Amendment No.96 1
LIMITING S AFE'"Y SYSTEM SETTINGS EASES einimum DNBR is maintained above the design DNBR value for multi rod drop accidents. The analysis of a single control rod drop accident indicates a return to full power may be initiated by the automatic control system in response to a continued full power turbine load demand or by the l
negative moderator temperature feedback. This transient will not result in a DNBR of less than the design DNBR value, therefore single rod drop protection is not required.
j Intermediate and Source Range, Nuclear Flux The Intermediate and Source Range, NucIear Flux trips provide reactor core protection during reactor startup. These trips provide redundant protection to the low setpoint trip of the Power Range, Neutron Flyg channels.
7te Source Range Channels will initiate a reactor trip at about 10 counts per second unless manually blocked when P-6 becomes active. The Intermediate Range Channels will initiate a reactor trip at a current level proportional to approximately 25 percent of RATED THERMAL POWER unless manually blocked when P-10 becomes active. No credit was taken for operation of the trips associated with either the Intermediate or Source Range Channels in the accident analyses; however, their functional capability at the specified trip settings is required by this specification to enhance the overall reliability of the Reactor Protection System.
Overtemperature AT The Overtemperature AT trip provides core protection to prevent DNB for all combinations of pressure, power, coolant temperature, and axial power distribution, provided that the transient is slow with respect to piping transit delays from the core to the temperature detectors (about 4 seconds),
and pressure is within the range between the High and Low Pressure reactor trips. This setpoint includes corrections for changes in density and heat capacity of water with temperature and dynamic compensation for piping delays from the core to the loop temperature detectors. With normal axial power distribution, this reactor trip limit is always below the core safety limit as shown in Figure 2.1-1. If axial peaks are greater than design, as indicated by the difference between top and bottom power range nuclear detectors, the reactor trip is automatically reduced according to the notations in Table 2.2-1.
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S ALEM - UNIT 1 B 2-4 Amendment No. 96
O LIMITING SAFE ~Y SYSTEM SETTINGS BASES through the pressurizer safety valves. No credit was taken for operation of this trip in the accident analyses; however, its functional capability at the specified trip setting is required by this specification to enhance the overall reliability of the Reactor Protection System.
Loss of Flow The Loss of Flow trips provide core protection to prevent DNB in the event of a loss of one or more reactor coolant pumps.
Above 11 percent of RATED THERMAL POWER, an automatic reactor trip will occur if the flow in any two loops drop belob 90% of nominal fell loop flow.
Above 36% (P-8) of RATED THERMAL POWER, automatic reactor trip will occur if the flow in any single loop drops below 90% of nominal full loop flow. This latter trip will prevent the minimum value of the DNBR from going below t.he design DNBR value during normal operational transients and anticipated transients when ' loops are in operation and the Overtemperature AT tr19 set point is adjusted to the value specified for all loops in operation. With the Dvertemperature LT trip set point adjusted to the value specified for 3 loop '
operation, the P-8 trip at 76% RATED THERMAL POWER will prevent the minimum i
value of the DNBR from going below the design DNBR value during normal operational transients and anticipated transients with 3 loops in operation.
Steam Generator Water Level The Steam Generator Water Level Low-Low trip provides core protection by preventing operation with the steam generator water level below the minimum volume required for adequate heat removal capacity. The specified setpoint provides allowance that there will be sufficient water inventory in the steam generators at the time of trip to allow for starting delays of the auxiliary feedwater systam.
Steam /Feedwater Flow Mismatch and Low Steam Generator Water Level The Steam /Feedwater Flow Mismatch in coincidence with a Steam Generator Low Water Level trip is not used in the transient and accident analyses but is included in Table 2.2-1 to ensure the functional capability of the specified trip settings and thereby enhance the overall i
SALEM - UNIT 1 B 2-6 Amendment No. 96 I
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REACTIVITY CONTROL SYSTEMS ROD DROP TIMI l
LIMITING CONDITION FOR OPERATION 3.1.3.3 The individual full length (shutdown and control) rod drop time from 228 steps withdrawn shall be 5 2.7 seconds from beginning of decay of stationary gripper coil voltage to dashpot entry with:
T,y 1 541*F, and a.
b.
All reactor coolant pumps operating.
APPLICABILITY: MODE 3.
t ACTION:
With the drop time of any full length rod determined to exceed the a.
above limit, restore the rod drop time to within the above limit prior to proceeding to MODE 1 or 2.
b.
With the rod drop times within limits but determined with 3 reactor coolant pumps' operating, operation may proceed provided THERMAL POWER is restricted to $71% of RATED THERMAL POWER.
SURVEILLANCE REQUIREMENTS 4.1.3.3 The rod drop time of full length rods shall be demonstrated through measurement prior to reactor criticality:
For til rods following each removal of the reactor vessel head, a.
b.
For specifically affected individual rods following any maintenance on or modification to the control rod drive system which could affect the drop time of those specific rods, and c.
At least once per 18 months.
SALEM - UNIT 1 3/4 1-21 Amendment No. 96
k POWF.R DISTRIBUTION LIMITS hTCLEAR ENTHALPY HOT CHANNEL FACTOR - F"H j
LIMITING CONDITION FOR OPERATION (H shall be limited by the following relationships 3.2.3
[h51.55[1.0+0.3(1-P)]
THERMAL POWER where: P
=
RATED THERMAL POWER APPLICABILITY: MODE 1 ACTION:
With N exceeding its limit:
Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within a.
2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and reduce the Power Range Neutron Flux-High Trip Setpoints to 5 557. of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, Demonstratethruin-coremappingthatF"kbuceTHERMALPOWER b.
is within its limit within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after exceeding the limit or r than 5% of RATED THERMAL POWER within the next 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, and Identify and correct the cause of the out of limit condition prior c.
to increasing THERMAL POWER above the reduced limit required by a; F'g b. above; subsequent POWER OPERATION may proceed provided that o
is demonstrated through in-core mapping to be within its limit akHa nominal 50% of RATED THERMAL POWER prior to exceeding this THERMAL POWER, at a nominal 75% of RATED THERMAL POWER prior to exceeding this THERMAL power and within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after attaining 95%
or greater RATED THERMAL POWER.
SALEM - WIT 1 3/4 2-9 Amendment No. 96
4 TABLE 3.2-1 DNE PARAMETERS j
l LIMITS 4 Loops In 3 Loops In PARAMITER Operation Operation Reactor Coolant System T,yg 5 582*F 5 572*F Pressurizer Pressure 1 2220 psia
- 2 2220 psia
- Reactor Coolant System 2 357,200 gpm#
1 284,500 gpm#
l
- Limit not applicable during either THERMAL POWER ramp increase in excess of
- 57. RATED THERMAL POWER per minute or a THERMAL POWER step increase in excess of 107. RATED THERMAL POWER.
- Includes a 2.2% flow measurement uncertainty plus a 0 1% measurement uncertainty due to feedwater venturi foul!ng.
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SALEM - UNIT 1 3/4 2-14 Amendment No.96
PO'-T.R DISTRIBUTION LIMITS EASES 3/4.2.5 DNB PAPJWETERS The limits on the DNB related parameters assure that each of the parameters are maintained within the normal steady state envelope of operation assumed in the transient and accident analyses. The limits are consistent with the initial FSAR assumptions and have been analytically demonstrated adequate to maintain a minimum DNBR of the design DNBR value throughout each analyzed transient.
The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> periodic surveillance of these parameters thru instrument readout is sufficient to ensure that the parameters are restored within their limits following load changes and other expe'eted transient operation. The 18 month periodic measurement of the RCS total flow rate is adequate to detect flow degradation and ensure correlation of the flow indication channels with measured flow such that the indicated percunt flow will provide sufficient verification of flow rate on a 12 hcur baats.
Amendrnent No. 96 SALEM - UNIT 1 B 3/4 2-6
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UNITED STATES j
Ij NUCLEAR REGULATORY COMMISSION 1
WASHINGTON, D. C. 20555 PUBLIC SERVICElLECTRIC & GAS COMPANY PHILADELPHIA ELECTRIC COMPANY DELMARVA POWER AND LIGHT COMPANY ATLANTIC CITY ELECTRIC COMPANY-DOCKET NO. 50-311 SALEti GENERATING STATION, UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.72 License No. DPR-75 1.
The Nuclear Regulatory Commission (the Comission or the NRC) has found that:
A.
The application for amendment filed by the Public Service Electric &
Gas Company, Philadelphia Electric Company,(Delmarva Power and Light Company and Atlantic City Electric Company the licensees) dated December 30, 1988 and supplemented by. letters dated April 19 and May 4, 1989, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Comission; C.
There is reasonable assurance: (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the pub'fic, and (ii) that such activities will be conducted in compliance with the Commission's regulations set forth in 10 CFR Chapter I; D.
The issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifica-tions as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-75 is hereby amended to read as follows:
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(2) Technical Specifications and Environmental Protection Plan.
The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 72,-are hereby incorporated in the license. The licensee shall operate the facility in accordance with-
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the Technical Specifications.
3..
This'. license amendment is effective as of fuel load t.uring the fifth refueling cutage currently scheduled to begin in March 1990.
FOR'THE NUCLEAR' REGULATORY COMMISSION
/s/
Walter.R. Butler, Director Project Directorate I-2 Division of Reactor Projects I/II
Attachment:
Changes to the Technical.
L Specifications Date of' Issuance: May 9, 1989 b)
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2-l (2) Technical Specifications and Environmental Protection Plan The Technicel Specifications contained in Appendices A and 8, as revised through Amendment No. 72. ;re hereby incorporated in the l
license. The licensee shall operne the facility in accordance with the Technical Specifications.
3.
.Thic license amendment is effective as of fuel load during the fifth refueling outage currently scheduled to begin in March 1990.
FOR THE NUCLEAR REGULATORY COMMISSION l-Walter R. Butler, Director Project Directorate I-2 Division of Reactor Projects I/II
Attachment:
Changes to the Technical Specifications Date of Issuance: May 9, 1989
ATTACHMENT TC LICENSE AtiENDMENT NO. 72 FACILITY OPERATING LICENSE NO. DPR-75 DOCKET NO. 50-311 Revisa Appendix A as follows:
Remove Pages-Insert Pages B 2-1 B 2-1 B 2-4
.B 2-4 8 2-6 B 2-6 3/4 1-18 3/4 1-18 3/4 2-9 3/4 2-9 3/4 2-10 3/4 2-10 3/4.2-16 3/4 2-16 3/4 2-17 3/4 2-17 8 3/4 2-4 B 3/4 2-4 E 3/4 2-5 B 3/4 2-5 B 3/4 2-6 B 3/4 2-6 u.
2.1 SATETY LIMITS I
BASES 2.1.1 REACTOR CORE The restrictions of this safety limit prevent overheating of the fuel and possible cladding perforation which would result in the release of fission prodects to the reactor coolant. Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the i
I heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature.
Operation above the upper boundary of the nucleate boiling regime could result in excessive cladding temperatures because of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coefficient., DNB is not a directly measurable parameter during operation and therefore THERMAL POWER and Reactor Coolant Temperature and Pressure have been related to DNB through the W-3, R-Grid correlation for standard (LOPAR) fuel assemblies and WRB-1 correlations for Vantage SH fuel assemblies. The DNB correlation has been developed to predict the DNB flux and the location of DNB for axially uniform and non-uniform heat flux distributions. The local DNB heat flux ratio, DNBR, defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux, is indicative of the margin to DNB.
The DNB design basis is as follows: there must be at least a 95 percent probability that the minimum DNBR of the limiting rod during Condition I and II events is greater than or equal to the DNBR limit of the DNB correlation being used (the WRB-1 or W-3, R-Grid correlation). The correlation DNBR limit is established based on the entire applicable experimental data set such that there is a 95 percent probability with 95 percent confidence that DNB will not occur when the minimum DNBR is at the DNBR correlation limit (1.17 for the WRB-1 or 1.30 for the W-3 R-Grid).
The curves of Figures 2.1-1 and 2.1-2 show the loci of points of THERMAL POWER, Reactor Coolant System pressure and average temperature for which the minimum DNBR is no less than the design DNBR value, or the average enthalpy at l the vessel exit is equal to the enthalpy of saturated liquid.
The curves are based on an enthalpy hot channel factor, F of 1.55 and a reference cosine with a peak gf 1.55 for axial power shape. Akallowanceis included for an increase in T" at reduced power based on the expression:
g rlg s 1.55 h.M o.30 6 l
where P is the fraction of RATED THERMAL POWER These limiting heat flux conditions are higher than those calculated for the range of all control rods FULLY WITHDRAWN to the maximum allowable control l rod insertion assuming the axial power imbalance is within the limits of the fy (delta I) function of the Overtemperature trip. When the axial power Amendnent No. 72 EM-W2 BM
l l
l LIMITING SAFETY SYSTEM SETTINGS EASES The Power Range Negative Rate trip provides protection to ensure that the l
i minimum DNER is maintained above tne design DNBR value for control rod drop accidents.
At high power a single or multiple rod drop accident could cause local flux peaking which, when in conjunction with nuclear power being maintained equivalent to turbine power by action of the automatic rod control l
{
system, could cause an unconservative local DNBR to exist. The Power Range j
Negative Rate trip will prevent this from occurring by tripping the reactor for all single or multiple dropped rods.
Intermediate and Source Range, Nuclear Flux The Intermediate and Source Range, Nuclear Flux trips provide reacter core protection during reactor startup. These trips provide redundant protectiontothelowsetpointtripofthePowerRange,NeutronFlygchannels.
The Source Range Channels will initiate a reactor trip at about 10 counts per second unless manuclly blocked when P-6 becomes active. The Intermediate Range Channels will initiate a reactor trip at a current level proportional to approximately 25 percent of RATED THERMAL POWER unless manually blocked when F-10 becomes active. No credit was taken for operation of the trips associated with either the Intermediate or Source Range Channels in the accident analyses; however, their functional capability at the specified trip settings is required by this specification to enhance the overall reliability of the Reactor Protection System.
Overtemperature Delta T The Overtemperature delta T trip provides core protection to prevent DNB for all combinations of pressure, power, coolant temperature, and axial power distribution, provided that the transient is slow with respect to piping transit delays from the core to the temperature detectors (about 4 seconds),
and pressure is within the range between the High and Low Pressure reactor trips. This setpoint includes corrections for changes in density and heat capacity of water with temperature and dynamic compensation for piping delays from the core to the loop temperature detectors. With normal axial power distribution, this reactor trip limit is always below the core safety limit as shown in Figure 2.1-1. If axial peaks are greater than design, as indicated by the difference between top and bottom power range nuclear detectors, the reactor trip is automatically reduced according to the notations in Table 2.2-1.
SAI.EM - UNIT 2 B 2-4 Amendment No. 72
LIMITINO SAFETY SYSTDI SEMINGS BASES Loss of Flow The Loss of Flow trips provide core protection to prevent DNB in the event of a loss of one er mers reactor coolant pumps.
Above 11 percent of RATED THERMAL POWER, an automatic reactor trip will occur if the flow in any two loops drop below 90% of nominal full loop flow.
Above 36% (P-8) of RATED THERMAL POWER, automatic reactor trip will occur if the flow in any single loop drops below 90% of nominal full loop flow. This latter trip will prevent the minimum value of the DNBR from going below the design DNBR value during normal operational.* transients and anticipated transients when 3 loops are in, operation and the Overtemperature delta T trip set point is adjusted to the value specified for all loops in operation. With the Overtemperature delte T trip set point adjusted to the value specified for 3 loop operation, the P-8 trip at 76% RATED THERMAL POWER will prevent the minimum value of the DNBR from going below the design DNBR value during normal operational transients and anticipated transients with 3 loops in operation.
Steam Generator Water Level The Steam Generator Water Level Low-Low trip provides core protection by preventing operation with the steam generator water level below the minimum volume required for adequate heat ramoval capacity. The specified setpoint provides allowance that there will be sufficient water inventory in the steam generators at the time of trip to allow for starting delays of the auxiliary feedwater systam.
Steam /Feedwater Flow Mismatch and Low Steam Generator Water Level The Steam /Feedwater Flow Mismatch in coincidence with a Steam Generator Low Water Level trip is not used in the transient and accident analyses but is included in Table 2.2-1 to ensure the functional capability of the specified trip settings and thereby enhance the overall reliability of the Reactor Protection System. This trip is redundant to the Steam Generator Water Level Low-Low trip. The Steam /Feedwater Flow Mismatch portion of this trip is activated when the steam fl w exceeds the feedwater flow by greater than or 6
equal to 1.42 x 10 lbs/ hour. The Steam Generator Low Water level portion of the trip is activated when the water level drops below 24 percent, as indicated by the narrow range instrument. These trip values include sufficient allowance in excess of normal operating values to preclude spurious trips but will initiate a reactor trip before the steam generators are dry. Therefore, the required capacity and starting time requirements of the auxiliary feedwater pumps are reduced and the resulting thermal transient.a the Reactor Coolant System and steam generators is minimized.
SALEM - UNIT 2 B 2-6 Amendment No. 72
1 l
l REACTIVITY CONTROL SYSTEMS RCD DROP TIME L!MITING CONDITION FOR OPERATION
+
3.1.3.3 The individual full length (shutdown and control) ro0 drop time from 228 steps withdrawn shall be less than or equal to 2.7 seconds from beginning I
of decay of stationary gripper coil voltage to dashpot entry with:
T,yg greater than or equal to 541*F, and a.
.b.
All reactor coolant pumps operating.
APPLICABILITY: MODES 1 & 2.
l ACTION:
With the drop time of any full length rod determined to exceed the a.
above limit, restore the rod drop time to within the above limit prior to proceeding to MODE 1 or 2.
b.
With the rod drop times within limits but determined with 3 reactor I
coolant pumps operating, operation may proceed provided THERMAL POWER is restricted to less than or equal to 76% of RATED THERMAL POWER.
SURVEILLA.NCE REQUIREMENTS 4.1.3.3 The rod drop time of full length rods shall be demonstrated through measurement prior to reactor criticality:
For all rods following each removal of the reactor vessel head, a.
b.
For specifically affected individual rods following any maintenance on or modification to the control rod drive system which could affect the drop time of those specific rods, and At least once per 18 months.
c.
SALEM - UNIT 2 3/4 1-18
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4 I
1 POWER DISTRIBUTION LIMITS 3/4.2.3 NUCLEAR ENTHALPY HOT CHANNEL FACTOR F H LIMITING CONDITION FOR OPERATION 3.2.3 7
shall be limited by the following relationship:
g 1 1.55 [1.0 + 0.3 (1.0-P)]
H where:
T THERMAL POWER e
RATED THERMAL POWER APPLICABILITY: MODE 1 ACTIOM:
With F exceeding its limit:
H Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 2 a.
hours and reduce the Power Range Neutron Flux-High Trip Setpoints to 1 55% of RATED THERMAL POWER within sne next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, i
b.
Demonstrate thru in-core mapping that F is within its limit within 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />safterexceedingthelimitorredub!THERMALPOWERtolessthan5%
of RATED THEPMAL POWER within the next 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, and Identify and correct the cause of the out of limit condition prior to-c.
increasing THEPMAL POWER above the reduced limit required by g. or b.
above; subsequent POWER OPERATION may proceed provided that F 18 demonstratedthroughin-coremappingtobewithinitslimitakHa nominal 50% of RATED THERMAL POWER prior to exceeding this THERMAL POWER, at a nominal 75% of RATED THERMAL POWER prior to exceeding this THERMAL POWER and within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after attaining 95% or greater RATED THERMAL POWER.
l SALEM - UNIT 2 3/4 2-9 Amendment No. 72
7 l
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o POLTJ DISTRIBUTION LIMITS St%'EILLANCE REQUIREMD.TS 4.2.3.1 [k!ctors to obtain a power distribution maps shall be determined to be within its limit by using the movable incere de Prior to operation above 75% of RATED THERMAL POWER after each fuel a.
loading, and b.
At least once per 31 Effective Full Power Days.
The provisions of Specification 4.0.4 are not applicable.
c.
4.2.3.2 The measured N f
a ve, s increased by 4% for measurement uncertainty H t
SALEM - UNIT 2 3/4 2-10 Amendment No. 72
l POLTR DISTRIBUTION LIMITS 3/1. 2.5 DNB PARA.'ETERS LIMITING CONDITION FOR OPERATION 3.2.5 The following'DNB related parameters shall be maintained within the limits shown on Table 3.2-1:
Reactor Coolant System T, a.
b.
Pressurizer Pressure.
c.
Reactor Coolant Systm Tc*.a1 flow Rate.
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~
t APPLICABILITY: MODE I ACTION:
With any of the above parameters exceeding its limit, restore the parameter to within its limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
SURVEILLANCE REQUIREMENTS 4.2.5.1 Each of the parameters of Table 3.2-1 shall be verified to be within their limits at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
4.2.5.2 The Reactor Coolant Syr. tem Total Flow Rate shall be determined to be within its limit by measurement at least once per 18 months.
i Amendment No. 72 SALEM - UNIT 2 3/4 2-16
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TABLE 3.2-1 u
DNB PARAMETERS
[1 PARAMETER I,IMITS 4 Loops in Operation Reactor Coolant System T,yg 5 582*F Pressurizer Pressure 5 2220 psia
> 357200 gpm#
l l
1
- Limit not applicable during either a THERMAL POWER ramp in excess of 5% RATED THERMAL POWER per minute or a THERMAL POWER step in excess of 10% RATED THERMAL POWER.
- Incivdes a 2.2% flow uncertainty plus a 0.1% measurement uncertainty due to feedwater venturi fouling.
SALEM - UNIT 2 3/4 2-17 Amendment No. 72
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PO'a*EE DiSTEIEl.' TION LIMITS BASES 3/4.2.2 and 3/l. 2.3 HEAT FLUX AND NUCLEAR ENTHALPY HOT CHANNEL j
AND RADIAI PEAT.ING FACTORS - F (Z) AND 9
g l
The limits on heat flux and nuclear enthalpy hot channel factors and RCS I
flow rate ensure that 1) the design limits on peak local powcr density and minimum DNBR are not exceeded and 2) in the event of a LOCA the peak fuel clad temperature will not exceed the 2200*F ECCS acceptance criteria limit.
Each of these hot channel factors are measurable but will normally only l
be determined periodically as specified in Specifications 4.2.2 and 4.2.3.
This periodic surveillance is sufficie:t to insure that the limits are maintained provided:
Control rod in a single group move together with no individual rod a.
insertion differing by more than i 12 steps from the group demand position.
b.
Control rod groups are sequenced with overlapping groups as described in Specification 3.1.3.5.
The control rod insertion limits of Specifications 3.1.3.4 and c.
3.1.3.5 are maintained.
d.
The axial power distribution, expressed in terms of AXIAL FLUX DIFFERENCE, is maintained within the limits.
The relaxation in as a function of N N allows chgnges in H
the radial power shape for all permissible rod insertion limits.
F will be maintained ithin its limits provided conditions a thru d above, areg maintained.
When an F measurement is taken, both experimental error and g
manufacturing tolerance must be allowed for. Five percent is the appropriate allowance for a full core map taken with the incore detector flux mapping system and 3% is the appropriate allowance for manufacturing tolerance.
When F is measured, experimental error must be allowed for and 4% is theapproprNteallowanceforafuf1coremaptakenwiththeincoredrisction system. ThespecifiedlimitforFf ce fN uncertaintieswhichmeanthatnorma$alsocontainsan8% allow operation will result in The 8% allowance is based on the following considerations:
H $
a 1.08.
SALEM - UNIT 2 B 3/4 2-4 Amendment No. 72
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e POWER DISTRIBUTION LIMITS BASES 3/4.2.2 and 3/4.2.3 HEAT FLUX AND NUCLEAR ENTHALPY HOT CHANNEL 1
i AND RADIAL PEAKING YACTORS - F (Z) AND H (Continued) g abormal perturbations p the radial power shape, such as from rod a.
misalignment, effect r more directly than F.
g g
b.
although rod sovement has a direct influence upon limiting F to withinitslimit,suchcontrolisnotreadilyavailabletoikmit M' **b in prediction for control power shape detected during startup c.
errort physics test can be compensated for in F by restricting axial flux Nq distributions. This compensation for F available, g is less rapidly j
The radial peaking factor F (Z) is measured periodically to provide assurancs that the hot channel factJo FQ(Z), remains within its limit. The F limit for RATED THERMAL POWER ( r
), as provided in the Radial Pesking FaIEor Limit Reportperspecifteation6*E.1.10,wasdeterminedfromexpectedpowercontrol maneuvers over the full range cf burnup conditions in the core.
l 3/4.2.4 OUADRANT POWER TILT RATI_0 i
The quadrant power tilt ratio limit assures that the radial power distribution satisfies the design values used in the power capability analysis. Radial power distribution measurements are made during startup testing and periodically during power operation.
i emrea e eve o am e _ c4 m m ma
i.
Pom DISTRIBl' TION LIMITS BASES The limit of 1.02 at which corrective action is required provides DNB and linear heat generation rate protection with x-y plane power tilts. A limiting tilt of 1.025 can be tolerated before the margin for uncertainty in F is depleted. The limit of 1.02 was selected to provide an allowance for tne g
uncertainty associated with the indicated power tilt.
The 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> time allowance for operation with a tilt condition greater than 1.02 but less than 1.09 is providet. to allow identification and correction of a dropped or misaligned rod. In the event such action does not correct the tilt, the margin for uncertainty on F is reinstated by reducing g
the power by 3% from RATED THZRMAL POWER for each percent of tilt in excess of 1.0.
,3/4.2.5 DNB PARA.507'ERS The limits on the DNB related parameters assure that each of the parameters are maintained with the nornal steady state envelope of operation assced in the transient and accidant analyses. The limits are consistent with the initial FSAR assumptions and have been analytically demonstrated adequate to maintain a minimum DNBR of the design DNBR value thrcughout each analyzed l
The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> periodic surveillance of these parameters through instrument readout is sufficient to ensure that the paraneters are restored within their limita following load changes and other expected transient operation. The 18 nonth periodic measurement of the RCS total flow rate is adequate to detect ilow degradation and ensure correlation of the flow indication channels with measured flow such that the indicated percent flow will provide sufficient verification of flow rate on a 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> basis.
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i SALEM - UNIT 2 B 3/4 2-6 Amendment No. 72 1