ML20044H051
| ML20044H051 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 06/01/1993 |
| From: | Butler W Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20044H052 | List: |
| References | |
| NUDOCS 9306070306 | |
| Download: ML20044H051 (17) | |
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UNITED STATES l
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NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 DUOVESNE LIGHT COMPANY OHIO EDISON COMPANY PENNSYLVANIA POWER COMPANY l
DOCKET NO. 50-334 i
i BEAVER VALLEY POWER STATION.' UNIT NO I 1'
AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.172' License No. DPR-66' 1.
The Nuclear Regulatory Comission. (the Comission) has found that:
l A.
The application for amendment by Duquesne Light Company, et al. (the licensee) dated February 19, 1993 as supplemented March 31, and i
April 19,.1993, complies with the standards and requirements of the l
Atomic Energy Act of 1954, as amended (the Act) and the Commission's i
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 public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and s,afety of the public; and E.
The issuance of this amendment is in accordance with :10 CFR Part 51 of the Comiss. ion's regulations and all applicable requirements have been s?tisfied.
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9306070306 930601 PDR ADOCK 05000334 P
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Accordingly, the license is amended by changes to the. Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility 0perating License No. DPR-66.is-hereby
-amended to read as follows:
(2) Technical Soecifications l
The Technical Specifications contained in' Appendix A,..as revised.
through Amendment No.172,-'are hereby incorporated in;..the license.
The licensee-shall operate the facility in aceprdance with the Technical Specifications.
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3.
This license amendment is effective as' of the date of its issuance,~ to.be implemented within 60 days of issuance.
FOR THE NUCi. EAR. REGULATORY' COMMISSION.
Rdu etL Walter R. Butler,. Director -.
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Project Directorate I-3 Division of Reactor Projects'- I/II' Office' of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance:
June 1, 1993 l
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l ATTACHMENT TO LICENSE AMENDMENT N0.172 FACIllTY OPERATING LICENSE NO. DPR-66 4
DOCKET NO 50-334 i
Replace the following pages of Appendix A Technical Specifications, with the enclosed pages as indicated. The revised pages are identified by amendment number and contain vertical lines indicating the areas of change.
Remove Insert l
XXV XXV 2-1 2-1 2-2 2-2 2-3 j
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2-4 2-6 2-6 B 2-4 B 2-4 3/4 2-12 3/4 2-12
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3/4 2-13 3/4 2-13 l
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DPR-66 Ficure Index FIGURE TITLE PAGE 2.1-1 Reactor Core Safety Limit - Three Loops in 2-2 Operation l
3.1-1 Rod Group Insertion Limits Versus Thermal 3/4 1-24 Power
.Three Loop Operation 3.1-2 Rod Group Insertion Limits Versus Thermal 3/4 1-25 Power - Two Loop Operation 3.2-1 Axial Flux Difference Limits as a Function 3/4 2-4 of Rated Thermal. Power 3.2-2 K(z) - Normalized Fo(z) as a function of 3/4 2-7 Core Height 3.4-1 Dose Equivalent I-131 Primary Coolant 3/4 4-21 Specific Activity Limit Versus Percent of Rated Thermal Power with the Primary Coolant Specific Activity > 1.0 pCi/ gram Dose Equivalent I-131 3.4-2 Beaver Valley Unit No. 1 Reactor Coolant 3/4 4-24 System Heatup Limitations Applicable for the First 9.5 EFPY 3.4-3 Beaver Valley Unit No. 1 Reactor Coolant 3/4 4-25 System Cooldown Limitations Applicable for the First 9.5 EFPY 3.6-1 Maximum Allowable Primary Containment Air 3/4 6-7 Pressure Versus River Water Temperature and RWST Water Temperature B 3/4.2-1 Typical Indicated Axial Flux Difference B 3/4 2-3 Versus Thermal Power at BOL.
B 3/4.4-1 Fast Neutron Fluence (E>l Mev) as a B 3/4 4-6a Function of Full Power Service Life B 3/4.4-2 Effect of Fluence, Copper Content, and B 3/4 4-6b Phosphorus Content on A RTmn for Reactor Vessel Steels Per Reg. Guide 1.99 BEAVER VALLEY - UNIT 1 XXV Amendment No.172
DPR-66 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS REACTOR CORE 2.1.1 The combination of THERMAL POWER, pressurizer pressure, and the highest operating loop coolant temperature (T,y) ion shall not exceed the limits shown in Figure 2.1-1 for 3 loop operat APPLICABILITl:
MODES 1 and 2.
ACTION:
Whenever the point defined by the combination of the highest operating loop average temperature and THERMAL POWER has exceeded the appropriate pressurizer pressure line, be in HOT STANDBY within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
REACTOR COOLANT SYSTEM PRESSURE 2.1.2 The Reactor Coolant System pressure shall not exceed 2735 psig.
APPLICABILITY:
MODES 1, 2,
3, 4 and 5.
ACTIOl{:
MODES 1 and 2 Whenever the Reactor Coolant System pressure has exceeded 2735 psig, be in HOT STANDBY with the Reactor Coolant System pressure within its limit within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
MODES 3, 4 and 5 Whenever the Reactor Coolant System pressure has exceeded 2735 psig, reduce the Reactor Coolant System; pressure to within its limit within 5 minutes.
i BEAVER VALLEY - UNIT 1 2-1 Amendment No.172
^1 DPR-66 670
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1.1 1.2 FRACTION OF RATED THERMAL POWER FIGURE 2.1-1 REACTOR CORE SAFETY LIMIT THREE LOOP OPERATION BEAVER VALLEY - UNIT 1 2-2 Amendment No.172
DPR-66 TABLE 2.2-1 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES
- 1. Manual Reactor Trip Not Applicable Not Applicable
- 2. Power Range, Neutron Flux Low setpoint - s 25% of RATED-Low Setpoint - 5 27.3% of RATED THERMAL POWER THERMAL POWER High Setpoint - s 109% of High Setpoint
's 111:3% of RATED THERMAL RATED THERMAL POWER
' POWER 3.
Power Range, Neutron Flux, s 5% of-RATED -THERMAL POWER
's 6.3% of RATED THERMAL POWER with a time High Positive Rate with a time constant 2 2 constant 2 2 seconde seconds r
4.
Power Range,. Neutron Flux, s 5% of RATED THERMAL POWER s 6.3% of RATED THERMAL POWER with a time High Negative Rate with a time constant 2 2 constant a 2 seconds seconds t.
5.
Intermediate Range, s 25% of RATED THERHAL POWER s 31.1% of RATED THERMAL POWER Neutron Flux 5
5 16.
Source Range,. Neutron Flux s-10 counts per second-s 1.4 x 10 counts per-second 7.-
Overtemperature AT.
See Note 1 See Note 3'
^
8.
Overpower AT See Note 2_
.See Note 4 9.~' Pressurizer Pressure--Low ~
2 1945 peig
.2 1934 psig 4
- 10. Pressurizer Pressure--High 5 2385 psig s 2394 psig
- 11. Pressuriger Water s 92% of instrument span s 93.9% of instrument span Level--High 4
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- 12. Loss'of Flow 2 90% of. design flow
- per loop 2.89.0% of. design flow
- per loop-
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- Design flow is 87,200'gpm per loop.
l i BEAVER VALLEY - UNIT 1 2-6 Amendment No.172
l DPR-66 LIMITING SAFETY SYSTEM SETTINGS BASES i
i The Power Range Negative Rate trip provides protection to ensure that the minimum DNBR is maintained above the design DNBR limit for control rod drop accidents.- At high power a single or multiple rod drop accident could cause flux peaking which, when in conjunction with nuclear power being maintained equivalent to turbine power by action of the automatic rod control
- system, could cause an unconservative local DNBR to exist.
The Power Range Negative Rate trip will prevent this from occurring by tripping the reactor.
For those transients on which reactor trip on power range negative rate trip is not postulated, it is shown that the minimum DNBR is greater 4
than the design DNBR limit.
Intermediate and Source Rance. Nuclear Flux The Intermediate and Source Range, Nuclear Flux trips provide reactor core protection during reactor start-up.
These trips provide j
redundant protection to the low setpoint trip of the Power Range, i
Neutron Flux channels.
The Source Range Channels will initiate a reactor trip at about 10+5 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
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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 l limit as shown on 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.
BEAVER VALLEY - UNIT 1 B 2-4 Amendment No.172
l DPR-66 POWER DISTRIBUTION LIMITS 3/4.2.5 DNB PARAMETERS l
LIMITING CONDITION FOR OPERATION 3.2.5 The following DNB related parameters shall be maintained within the limits shown on Table 3.2-1"):
l-a.
Reactor Coolant System T,y l
I b.
Pressurizer Pressure c.
Reactor Coolant System Total Flow Rate APPLICABILITY:
MODE 1G) l ACTION:
j With any of the above. parameters' exceeding its limit,. restore the i
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 />.
J SURVEILLANCE REQUIREMENTS 4.2.5.1 Each of the parameters of Table 3.2-1 shall be verified to be indicating 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 System total flow rate -shall be determined to be within its limit by measurement at least once per 18 months.
(1) The values presented in Table 3.2-1 correspond to analytical limits used in the safety analyses.
(2) The provisions of Specification 4.0.4 are not applicable for Reactor Coolant System total flow rate to allow a calorimetric flow measurement and the calibration of the Reactor Coolant System total flow rate indicators.
BEAVER VALLEY - UNIT 1 3/4 2-12 Amendment No.172
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DPR-66 t
TABLE 3.2-1 DNB PARAMETERE 3 Loops In PARAMETER Operation Reactor Coolant System T.y, s 580.7'F j
m Pressurizer Pressure 2 2220 psia Reactor Ccolant System 2 261,600 gpm Total Flow Rate i
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(1)
Limit not applicable during either a THERMAL POWER'. ramp l increase in excess of 5% RATED THERMAL POWER per minute or j
a THERMAL POWER step increase in excess of 10% RATED THERMAL l
POWER.
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BEAVER VALLEY - UNIT 1 3/4 2-13 Amendment No.172
o "c e-E, UNITED STATES S
i NUCLEAR REGULATORY COMMISSION
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WASHINGTON, D.C. 20555-0001
....+
DUQUESNE LIGHT COMPANY i
OHIO EDIS0N COMPANY THE CLEVELAND-ELECTRIC ILLUMINATING COMPANY THE TOLEDO EDISON COMPANY-DOCKET NO. 50-412 BEAVER VALLEY POWER STATION. UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSI Amendment No. 51 1
License No. NPF-73 1.
The Nuclear Regulatory Commission (the Commission) has found that:
]
A.
The application for amendment by Duquesne Light Company, et al. (the
]
licensee) dated February 19, 1993, as supplemented March 31, and April 19, 1993, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, 1
the provisions of the Act, and the rules and regulations of the Commission; i
C.
There is reasonable assurance (i) that the activities. authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that'such activities will be conducted in compliance with the Commission';s regulations; 1
l D.
The issuance of this amendment will not be inimic'al to the common defense and security or to.the health and safety of the public; and j
i E.
1he issuance of this amendment is in_accordance with_10 CFR Part 51 of the Commission's regulations and all applicable requirements l
have been satisfied.
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Accordingly, the license is amended by changes to the Technical-Specifications as indicated in the attachment to.this license amendment, I
and paragraph 2.C.(2) of Facility _ Operating License No. NPF-73 is~hereby.
amended to' read as follows:
1 (2) Technical Specifications The Technical Specifications contained in~ Appendix.A as revised'
-l through Amendment No. 51, and the Environmental ~ Protection' Plan-1 contained in Appendix B, both of which are attached hereto are i
hereby incorporated in the license. DLCO shal'1 operate the facility in accordance with the Technical-Specifications and.the -
1 Environmental Protection Plan..
3.
This license amendment is effective as of the date of its issuance, to be implemented within 60 days of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
/ol W O L Walter R. ' Butler, Director.-
' Project Directorate I-3 Division of. Reactor Projects - I/II:
)
Office' of Nuclear Reactor Regulation Att:chment:
Changes to the Technical Specifications -
Date of Issuance: June 1, 1993 t
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ATTACHMENT T0 LICENSE AMENDMENT N0,51 i
' FACILITY OPERATING LICENSE NO. NPF-73 j
DOCKET NO. 50-412-Replace the following pages of Appendix A, Technical Specifications,L with the
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enclosed pages as indicated. The revised pages are identified by amendment number and contain vertical lines indicating the areas of change.
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1.1 1.2 FRACTION OF RATED THERMAL POWER j
FIGURE 2.1-1 REACTOR CORE SAFETY LIMIT i
THREE LOOP OPERATION l
BEAVER VALLEY - UNIT 2 2-2 Amendment No. 51
NPF-73 TABLE 2.2-1 l
j REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS FUNCTIONAL UNIT ALLOWANCE (TA)
Z S
TRIP SETPOINT ALLOWADLE VALUE 1.
Manual Reactor Trip N/A N.A.
N.A.
N.A.
N.A.
2.
Power Range, Neutron Flux, a.
High Setpoint 7.5 4.56 0
5 109% of RTP*
s 111.1% of RTP*
b.
Low Setpoint 8.3 4.56 0
s 25% RTP*
s 27.1% of RTP*
3.
Power Range, Neutron Flux, 1.6 0.50 0
5 5% of RTP* with 5*6.3% of RTP* with High Positive Rate a time constant a time constant 2 2 seconds a 2 seconds 4.
Power Range, Neutron Flux 1.6 0.50 0
5 5% of RTP* with 5 6.3% of RTP* with High Negative Rate a time constant a time constant 2 2 seconds 2 2 seconds 5.
Intermediate Range, 17.0 8.41 0
s 25% RTP*
s 30.9% of RTP*
Neutron Flux 8
6.
Source Range, Neutron Flux 17.0 10.01 0
s 10 cps s 1.4 x 105 cps 7.
Overtemperature AT 7.0 5.10 See Note 5 See Note 1 See Note 2 8.
Overpower AT 4.9 1.71 1.49 See Note 3 See Note 4 9.
Pressurizer Pressure-Low 3.1 0.71 1.67 2 1945 psig***
2 1935 psig***
- 10. Pressurizer Pressure-High 6.2 4.96 0.67 s 2375 peig s 2383 psig
- 11. Pressurizer Water Level-High 8.0 2.18 1.67 s 92% of s 93.8% of instrument span instrument span
- 12. Loss of Flow 2.5 1.39 0.60 2 90% of loop 2 88.9% of loop l
design flow **-
design flow **
- = RATED THERMAL POWER
- Loop design flow = 87,200 gpm l
- Time constants utilized in the lead-lag controller for Pressurizer Pressure-Low are 2 seconds for lead and i second for lag.
Channel calibration shall ensure that these time constants are adjusted to those values, l
BEAVER VALLEY - UNIT 2 2-4 Amendment No.51
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NPF-73 POWER DISTRIBUTION LIMITS I
e DNB PARAMETERS LIMITING CONDITION FOR OPERATION 3.2.5 The following DNB related parameters shall be maintained.
within the limits shown on Table 3.2-1m l
l a.
Reactor Coolant System T,y J
b.
Pressurizer Pressure c.
Reactor Coolant System Total Flow Rate APPLICABILITY:
MODE 1W l
ACTION:
With any of the above parameters exceeding its limit, restore the l
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 percent of RATED THERMAL POWER within the next 4 l
hours.
l SURVEILLANCE REQUIREMENTS 4.2.5.1 Each of the parameters of Table 3.2-1 shall be verified to l be indicating 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 />.
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I 4.2.5.2 The Reactor Coolant System total flow rate shall-be determined to be within its limit by measurement at least once per 18 j
months.
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(1) The values presented in Table 3.2-1 correspond to analytical limits used in the safety analyses.
(2) The provisions of Specification 4.0.4 are not applicable for Reactor Coolant System total flow rate to allow a calorimetric flow measurement and the calibration of the Reactor Coolant l
System total flow rate indicators.
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BEAVER VALLEY - UNIT 2 3/4 2-11 Amendment No.51 i
.i NPF-73 i
TABLE ~3.2-1
)
l DNB PARAMETERS-I
-3 Loops In-
'I PARAMETER
' Operation Reactor Coolant System T,y,
's 580.2*F'
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Pressurizer Pressure 2 2220 psia")
l Reactor Coolant System:
2 261',600 gpm l
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Total Flow Rate i
(1)'
Limit not applicable during either a THERMAL POWER. ramp' ] ;
increase in excess of 5 ' percent. RATED _. THERMAL POWER.per minute or a THERMAL' POWER step increase.in-excess of-10%
RATED THERMAL POWER.
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' BEAVER. VALLEY - UNIT.2 3/4~2-12'
' Amendment No.51
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