ML20133E735
| ML20133E735 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 10/02/1985 |
| From: | Stolz J Office of Nuclear Reactor Regulation |
| To: | Philadelphia Electric Co, Public Service Electric & Gas Co, Delmarva Power & Light Co, Atlantic City Electric Co |
| Shared Package | |
| ML20133E739 | List: |
| References | |
| DPR-44-A-111, DPR-56-A-115 NUDOCS 8510090528 | |
| Download: ML20133E735 (34) | |
Text
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+4 jo, UNITED STATES
[
3 NUCLEAR REGULATORY COMMISSION t
WASHINGTON, D. C. 20555 g,
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PHILADELPHIA ELECTRIC COMPANY I
PUBLIC SERVICE ELECTRIC AND GA5 COMPANY DELMARVA POWER AND LIGHT COMPANY ATLANTIC CITY ELECTRIC COMPANY DOCKET NO. 50-277 PEACH BOTTOM ATOMIC POWER STATION. UNIT NO. 2 AMENDPENT TO FACILITY OPERATING LICENSE Amendment No. 111 License No. DPR-44 i
1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by Philadelphia Electric Company, et 4
al. (the licensee) dated April 19, 1984, as supplemented October 2, 1984, 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; 1
4 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 (1) 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; D.
The issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the 4
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 4
been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment.
l and paragraph 2.C.(2) of Facility Operating License No. OPR-44 is hereby amended to read as follows:
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Technical Specifications I
i The Technical Specifications contained in Appendices i
i A and B, as revised through Amendment No. Ill, are l
hereby incorporated in the license, pECO shall operate r
the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of its date of issuance.
1 1
FOR THE NUCLEAR REGUL TORY C0m !SSION 1
J F. Stolz, Chief i
0 ting Reactors Branch #4 Division of Licensing
Attachment:
i Changes to the Technical i
Specifications I
]
oate of Issuance:
October 2,1985 i
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ATTACHMENT TO LICENSE AMENDMENT NO.111 FACILITY OPERATING LICENSE NO. DPR-44
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DOCKET NO. 50-277 i
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Replace the following pages of the Appendix "A" Technical Specifications l
with the enclosed pages. The revised pages are identified by Amendment i
number and contain a vertical line indicating the area of change.
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Remove Insert 11a lla 12 12 15 15 l
21 21 61 61 63 63 72 72 79 79 80 80 29 89 90 90*
182 182 199 199 252 252 l
- Overleaf page included for document completeness.
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PBAPS SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING B.
Core Thermal Power Limit B. APRM Rod Block Trip Setting Reactor Pressure < 800 psia) i SRB < (0.66 W + 424 - 0.66 AW)
( FRP )
RFLeo,
where:
F RP = fraction of rated thermal power (3293 MWt).
MFLPD = maximum f raction of limiting power density where the limiting Power density is 13.4 KW/ft for all 8 x 8 fuel.
= hall be set equal to 1.0 unless the actual operating value is less than the design value of 1.0, in which case the actual operating value will be used.
C.
Whenever the reactor is in the C. Scram and isolation--> 539 in. above shutdown condition with reactor low water vessel zero irradiated fuel in the reactor level (0" on level vessel, the water level shall instruments) not bs less than minus 160 inches indicated level (378 inches above vessel zero).
Amendment No..W, 111
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PBAPS SAFETY LIMIT LIMITING SAFETY _ SYSTEM SETTING 2.1 (Cont'd)
D.
' Scram-- turbine stop 410 percent valve closure E.
Scram-- turbine control fast closure on loss of control oil pressure.
5004P(850 psig.
F.
Scram--low
> 23 inches J
condenser vacuum Hg vacuum G.
Scram--main steam (10%
line isolation valve closure H.
>850 psig
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i isolation valve closure--nuclear system low pressure I.
> minus 160 in.
actuation--reactor Tndicated level low-low-low
(> 378 inches water level above vessel sero)
J.
> minus 48 in.
actuation--reactor Tndicated level low-low water
(> 490 inches level a5ove vessel i
zero)
K.
> minus 160 in.
isolation valve Tndicated level l
closure--reactor
(> 378 inches low-low-low a5ove vessel water level zero) l l
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Amendment No. FI, 111..
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PBAPS 1.1.C B__AS_E_S_ ( Con t ' d. )
However, for this specification a safety Limit violation will be assumed when a scram is only accomplished by means of a backup feature of the plant design.
The concept of not approaching a Safety Limit, provided scram signals are operable, is supported by the extensive plant safety analysis.
The computer provided with Peach Bottom Unit 2 has a sequence annuciation program which will indicate the sequence in which events such as scram, APRM trip initiation, pressure scram initiation, etc. occur.
This program also indicates when the scram setpoint is cleared.
This will provide information on how long a scram condition exists and thus provide some measure of a the energy added during a transient.
Thus, computer information normally will be available for analyzing scrams; however, if the computer information should not be available for any scram analysis, Specification 1.1.C will be relied upon to determine if a Safety Limit has been violated.
1 D.
Reactor Water Level (Shutdown Condition)
During periods when the reactor is shutdown, 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 eleva.ted cladding temperatures and clad perforation.
The core can be cooled sufficiently should the water level be reduced to two-thirds the core height.
Establishment of the safety limit at qinus 160 inches indicated level (378 inches above vessel zero) pro
- sides adeauate margin to assure sufficient cooling during shutdown conditions.
This level will be continuously monitored.
E.
References 1.
General Electric BWR Thermal Analysis Basis (GETAB): Data, Correlation and Design Application, January 1977 (NEDO-10958-A).
2.
Process Computer Performance Evaluation Accuracy, General Electric Company BWR Systems Department, June 1974 (NEDO-20340).
3.
" General Electric Boiling Water Reactor Generic Reload Fuel Application", NEDE-240ll-P-A.
Amendments Nos.
R,.3fr,A&,JO,111
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1 PBAPS I
I 2.1 BASES (Cont'd.)
f g7;_c t_or Wa t e__r _L_o_w Lev el _S_c__ ram a_n_d__I so_l a_t i on ( Ex c_e_p t* Ma i n Rea C.
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j The setpoint for the low level scram is above the bottom of the separator skirt.
This level has been used in transient analyses dealing with coolant inventory decrease.
The results reported in l
FSAR subsection 14.5 show that scram and isolation of all process j
lines (except main steam) at this level adequately protects the j
f uel and the pressure barrier, because MCPR is greater than the i
fuel cladding integrity safety limit in all cases, and system l
pressure does not reach the safety valve settings.
The scram setting is approximately 23 inches below the normal operating range and is thus adequate to avoid spurious scrams.
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D.
Turbine Sto2 Valve Closure Scram f
I The turbine stop valve closure scram trip anticipates the i
pressure, neutron flux and heat flux increase that could result from rapid closure of the turbine stop valves.
With a scram trip setting of less than or equal to 10 percent of valve closure from j
full open, the resultant increase in surface heat flux is limited such that MCPR remains above the fuel cladding integrity safety limit even during the worst case transient that assumes the
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turbine bypass is closed.
This scram is bypassed when turbine j
steam flow is below 30% of rated, as measured by turbine first stage pressure.
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l E.
Turbine Control Valve Scram j
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The turbine control valve fast closure scram anticipates the i
pressure, neutron flux and heat flux increase that could result from fast closure of the turbine control valves due to a load t
rejection exceeding the capacity of the bypass valves or a i
failure in the hydraulic control system which results in a loss of oil pressure.
This scram is initiated from pressure switches in the hydraulic control system which sense loss of oil pressure l
due to the opening of the fast acting solenoid valves or a failure in the hydraulic control system piping.
Two turbine j
first stage pressure switches for cach trip systo.? initiate I
automatic bypass of the turbine control valve fast closure scram when the first stage pressure is below that required to produce i
30% of rated power.
Control valve closure time is approximately i
twice as long as that for stop valve closure.
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i Amendments Nos. 36; A&, 30' 111
-21.
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INSTRUMENTATION TilAT INITIATES PRIMARY CONTAINMENT ISOLATION 2
g o Man tmum No.
g of Operable Number of Instrument j
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Instrument Instrument Tra,p Level Setting Channels Provided Action 4
% Channels per By Design (2) j Trap System j
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(1) i 2 (6)
Reactor Low Kater 3 0" Indicated 4 Inst. Channels A
i Level Level (3) 75 psig 2 Inst. Channels D
1 Reactor High Pressure (Shutdown Cooling Isolation) s 2
Reactor Low-Low-Low at or above -160" 4 Inst. Channels A
Water Level indicated level (4) 2 (6)
High Drywell Pressure 2 psig 4 Inst. Channels A
l 2
High Radiation Main
~ Full Power Background
< 3 X Normal Rated (8) 4 Inst. Channels B
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Steam Line Tunnel 2
Low Pressure Main 1 850 psig (7) 4 Inst. Channels B
Steam Line l
j 2 (5) liigh Flow Main
< 140% of Rated 4 Inst. Channels B
j Steam Line Steam Flow I
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< 200 deg. F (9) 4 Inst. Channels B
i Tunnel Exhaust l
Duct Illgh Temperature 1
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PBAPS 1
1 NCTIS FOR TABLE 3. 2. A i
1.
i Whenever Primary rpntainment integrity is recuired by Section 3.7, there shall be two operable or tripped trip systems for i
each function.
4 2.
f If the first column cannot be met for one of the trip systems, that trip system shall be tripped or the appropriate i
action listed below shall be taken:
A.
Initiate an orderly shutdown and have the reactor in Cold Shutdown Condition in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
B.
Initiate an orderly load reduction and have Main Steam Lines isolated within eicht hours.
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C.
Isolate Reactor Water Cleanup System.
D.
Isclate Shutdown Cooling.
4 E.
Isolate Reacter Water Cleanup Filter Demineraliters unless the following provision is satisfied.
The RW2U j
Filter Demineralizer may be used (tne isolation overridden) to route the reactor water to the main i
j condenser or waste surge tank, with the high temperature trip incperable for up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, provided the water i
inlet j
temperature is monitored once per hour and confirmed to be below 180 degrees T.
3.
Instrument setcoint corresoond.s to 538 inches above vessel
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zero.
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4.
Instrument setpoint corresponds to 378 inches above vessel
- zero, i
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5.
Two recuired for each steam line.
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These signals also start SBGTS and initiate secondary centainment isolation.
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Only required in Run Mode (interlocked with Mode $sitch).
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8.
i At a radiation level of 1.5 times the normal rated power background, an alarm will be tripped in the control room to alert the control room operators to an increase in the main i
steam line tunnel radiation level.
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l In the event of a loss of ventilation in the main steam line tunnel area, the main steam line tunnel exhaust duct high temperature setpoint may be raised up to 250 degrees F for a period not to exceed 30 minutes to permit' restoration of the ventilation flow.
shall observe control room indications of the ductDuring the 30-min temperature so in the event of ranid increases (indicative of a steam line break) main steam.line isolation valves.the coerator shall promptly close the Amencment No. E t J44, lll,
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PBAPS Notes for Table 3.2.B r
1.
Whenever any CSCS subsystem is required by Section 3.5 to be operable, there shall be two operable trip systems.
If the first column cannot be met for one of the trip systems, that trip system shall be placed in the tripped condition or the reactor shall be placed in the Cold Shutdown Condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
2.
Close isolation valves in RCIC subsystem.
3.
Close isolation valves in HPCI subsystem.
4.
Instrument set point corresponds to 378 inches above vessel Eero.
1 5.
HPCI has only one trip system for these sensors.
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I Amendment No.111 1 l
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TABLE 3.2.G INSTRUMENTATION TilAT INITIATES RECIRCULATION PUMP TRIP i
g-E, m
Minimum No.
Instrument Trip Level Setting Number of Instrument Action y
of Operable Channels Provided
.9 Instrument by Design Channels Per Trip System (1) 1 Reactor High Pres-
< 1120 ps19 4
(2) sure i
1 Reactor Low-Low Water
> -48 in. indicated 4
(2) j Level level I
Notes for Table 3.2.G 1.
Whenever the reactor is in the RUN Mode, there shall be one operable trip system for each parameter for each operating recirculation pump.
If this cannot be met, the indicated action shall be taken.
2.
Reduce power and place the mode selector-switch in a mode other than the RUN Mode.
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i TABLE 4.2.A r,. -
MINIMUM TEST AND CALIBRATION FREQUENCY FOR PCIS Instrument Channel (5)
Instrument Funct-lonal Test Calibration Freguency Instrument Check 1)
Reactor High Pressure (1)
Once/3 months None (Shutdown Cooling Permissive) 2)
Reactor Low-Low-Low (1) (3)
Once/ operating cycle Once/ day Water Level (7) 3)
Main Steam High Temp.
(1) (3)
Once/ operating cycle Once/ day 4)
Main Steam High Flow (7)
(1) (3)
Once/ operating cycle Once/ day 5)
Main Steam Low Pressure (1)
Once/3 months None 6)
Once/3 months once/ day High Flow 7)
Once/3 months None High Temp.
Logic System Functional Test (4) (6)
Freguency 1)
Main Steam Line Isolation Vvs.
Once/6 months
-Main Steam Line Drain Vvs.
Reactor Water Sample Vvs.
2)
RHR - Isolation Vv. Control Once/6 months Shutdown Cooling Vvs.
Head Spray 3)
Reactor Water Cleanup Isolation Once/6 months 4)
Drywell Isolation Vvs.
Once/6 months TIP Withdrawal
' Atmospheric Control Vvs.
Sump Drain Valves 5)
Standby Gas Treatment System Once/6 months Reactor Building Isolation b.N.
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PBAPS i
i 3.2 BASES In addition to reactor protection instrumentation which initiates a reactor scram, protective instrumentation has been provided which initiates. action to mitigate the consequences of accidents which are beyond the operator's ability to control, or terminates operator errors before they i
result in serious consequences.
This set of specifications provides the limiting conditions of operation for the primary system isolation function, initiation of the core cooling systems, control rod block and standby gas treatment systems.
The objectives of the Specifications are (i) to assure the effectiveness of the protective instrumentation when required i
even during periods when portions of such systems are out-of-service for maintenance, and (ii) to prescribe the trip l
settings required to assure adequate performance.
When necessary, one channel may be made inoperable for brief intervals to conduct required functional tests and calibrations.
i Some of the settings on the instrumentation that initiate or control core and containment cooling have tolerances explicitly stated where the high and low values are both critical and may have a substantial effect on safety.
The I
set points of other instrumentation, where only the high or low end of the setting has a direct bearing on safety, are l
chosen at a level away from the normal operating range to i -
prevent inadvertent actuation of the safety system involved and exposure to abnormal situations.
Actuation of primary containment valves is initiated by protective instrumentation shown in Table 3.2.A which senses the conditions for which isolation is required.
S uch 4
instrumentation must be available whenever primary containment integrity is required, i
The instrumentation which initiates primary system isolation is connected in a dual bus arrangement.
The low water level instrumentation set to trip at zero inches indicated level (538 inches above vessel zero) closes all isolation valves except those in Groups 1, 4 and 5.
Details of valve grouping and required closing times are given in Specification 3.7.
For valves which isolate at this level, this trip setting is adequate to prevent core uncovery in the case of a break in the largest line assuming a 60 second valve closing time.
Required closing times are less than this.
The low-low reactor water level instrumentation is set to trip when reactor water level is minus 48 inches indicated level (490 inches above vessel zero).
This trip initiates the UPCI and RCIC and trips the recirculation pumps.
The i
low-low-low reactor water level instrumentation is set to trip when the reactor water level is minus 160 inches indicated level (378 inches above vessel zero).
This trip j
closes Main Steam Line Isolation Valves, Main Steam Drain Valves and Recire Sample Valves (Group 1), activates the remainder of the CSCS subsystem, and starts l
l Amendment No. lll,
3.2 BASES (Cont ' d) the emergency diesel generators.
These trip level settings were chosen to be high enough to; prevent spurious actuation but low enough to initiate CSCS operation and primary system isolation so that post accident cooling can be accomplished 4
and the guidelines of 10 CFR 100 will not be exceeded.
For large breaks up to the complete circumferential break of a 28-inch recirculation line and with the trip setting given above, CSCS initiation and primary system' isolation are initiated in time to meet the above criteria.
Reference paragraph 6.5.3.1 FSAR.
The high drywell pressure instrumentation is a diverse signal i
for malfunctions to the water level instrumentation and in addition to initiating CSCS, it causes isolation of Group 2 and 3 isolation valves.
For the breaks discussed above, this instrumentation will generally initiate CSCS operation before the low - low - low water level instrumentation; thus the results given above are applicable here also.
See Spec.
3.7 for Isolation Valve Closure Group.
The water level instrumentation initiates protection for the full spectrum of loss-of-coolant accidents and causes isolation of all isolation valves except Groups 4 and 5.
Venturis are provided in the main steam lines as a means of measuring steam flow and also limiting the loss of mass inventory from the vessel during a steam line break accident.
The primary function of the instrumentation is to detect a i
break in the main steam line.
For the worst case accident, j
main steam line break outside the drywell, a trip setting of 140% of rated steam flow in conjunction with the flow limiters and main steam line valve closure, limits the mass inventory loss such that fuel is not uncovered, fuel temperatures peak at approximately 1000 F and release of radioactivity to the environs is below CFR 100 guidelines.
Reference Section 1
14.6.5 FSAR.
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Temperature monitoring instrumentation is provided in the main steam line tunnel exhaust duct and along the steam line in the turbine building to detect leaks in these areas. Trips are provided on this instrumentation and when exceeded, cause 1
closure of isolation valves.
See Spec. 3.7 for valve Group.
The setting is 2000F for the main steam line tunnel detector.
Fo; large breaks, the high steam flow instrumentation is a O
backup to the temperature instrumentation.
no High radiation monitors in the main steam line tunnel have 5
been provided to detect gross fuel failure as in the control 6
rod drop accident.
With the established setting of 3 times normal background, and main steam line isolation valve closure, 19 i
F fission product release is limited so that 10 CFR 100 guide-lines are not exceeded for this accident.
Reference Section i
14.6.2 FSAR.
i 90-l APRIL 1974
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PBAPS NOTES FOR TABLE NO. 3.7.1 C = Closed SC = Stays Closed GC = Goes Closed Note:
Isolation grouoings are as follows:
GROU? 1:
The valves in Group 1 are actuated by any one of the following conditions:
1.
Reactor vessel low-low-low water level.
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- 2. Main steam line high radiation.
- 3. Main steam line high flow.
- 4. Main steam line space high temperature.
- 5. Main steam lino low pressure (RUN mode only).
GROUP 2A:
The valves in Group 2A are actuated by any one of the following conditions:
1.
Reactor vessel low water level.
l 2.
Reactor water cleanup system heat exchanger discharge high temperature.
3.
Reactor water cleanup system suction line break.
4.
Standby liquid control system actuation.
GROUP 28:
The valves in Group 2B are actuated by any one of the following conditions:
1.
Reactor vessel low water level.
- 2. High drywell pressure.
3.
Reactor high pressure of shutdown mode.
GROUP 2C:
The valves in Group 2C are actuated by any one of the following conditions:
- 1. Reactor low water level.
- 2. High reactor vessel pressure, (600 PSIG)
- 3. High drywell pressure.
i GROUP 2D:
The valves in Group 2D are actuated by the following conditions:
I
- 1. High drywell pressure.
- 2. Reactor low water level.
GROUP 3:
The valves in Group 3 are actuated by any one of the following conditions:
Amenditent No. 3& 111
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l PBAPS 3.7.D & 4.7.D BASES Primary Containment Isolation Valves Double isolation valves are provided on lines penetrating the primary containment and open to the free space of the containment.
j Closure of one of the valves in each line would be sufficient to
)
maintain the integrity of the pressure suppression system.
l Automatic initiation is required to minimize the potential leakage paths from the containment in the event of a loss-of-coolant accident.
Group 1:
Actuation for valves associated with the isolation of the main steam system.
The main steam lines are isolated by reactor vessel low-low-low water level in order to allow for l
removal of decay heat subsequent to a scram, yet isolate in time for proper operation of the core standby cooling systems.
The valves in group 1 are also closed when process instrumentation detects excessive main steam line flow, high radiation, low pressure, or main steam space high temperature.
Grouo 2:
Actuation for valves associated with the isolation of the reactor auxiliary systems.
Some of the reactor auxiliary systems such as the RWCU and RHR shutdown cooling systems connect into the reactor coolant boundary while others such as the drywell equipment and floor drain discharge valves do not penetrate the reactor coolant boundary.
Group 2 actuation is subdivided as follows:
Group 2A - process lines are normally in use and it is therefore not desirable to cause spurious isolation due to high drywell pressure resulting from non-safety related J
causes.
To protect the reactor from a possible pipe break in the system, isolation is provided by high temperature at the cleanup system heat exchanger / outlet or high flow through the inlet to the cleanup system.
Also, since the vessel could potentially be drained through the cleanup system, a low level isolation is provided.
An alarm of high temperature in the cleanup system area will provide an indication of suction line break resulting in manual isolation of the system.
During actuation of the standby liquid control system, the cleanup system is isolated.
Group 2B - isolation valves are not normally in use and are closed by reactor vessel low water level, high drywell pressure or high reactor pressure of the shutdown mode.
Groug,2C - isolation valves can only be opened when the reactor is at low pressure and the core standby cooling systems are not required.
Also, since the reactor vessel could potentially be drained through these process lines, these valves are closed by low water level.
Amendment No. E, 111
_199_
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PBAPS G.5.2.S Centinued e.
The Facility Emergency Plan and implementing procedures at least once per year.
l f.
The Facility Security Plan and i=plementing procedures at least once per two years.
g.
The offsite Dese Calculation Manual and i=plementing procedures at least once per two years.
h.
The perfermance of activities recuired by the Cuality Assurance Program regarding the radiological monitoring program to meet the provisiens of Regulatory Guide 4.1, Revisien 1, April 1975, at least enee per calendar year.
1.
Any other area cf facility operatien eensidered appropriate by the CSR Cc==itt ee er the Vice President, Electric Preduc.icn.
Autheritv 6.5.2.9 The CSR Cc==ittee shall report to and advise the Vice President, Electric Production on these areas of responsibility specified to Sectien 6.5.2.7 and 6.5.2.2.
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Reecrds 6.5.2.10 Records of CSR Committee activities shall be prepared, l
approved, and distributed as indicated below:
(
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Minutes of each OSR Committee meeting shall be a.
prepared, approved and forwarded to the Vice President, Electric Production within 14 days following each meeting.
b.
Reports of review encompassed by Section
- 6. 5.2. 7. e, f, g, and h a bove, shall be prepared,
approved and forwarded to the Vice President, Electric Producticn within 14 days following completion of the review.
Amencment No. 72, 37, y 111
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i p regu g'o, UNITED STATES l
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NUCLEAR REGULATORY COMMISSION li.
'j WASHINGTON, D. C. 20555 j
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PHILADELPHIA ELECTRIC COMPANY PUBLIC SERVICE ELECTRIC AND GAS COMPANY DELMARVA POWER AND LIGHT COMPANY ATLANTIC CITY ELECTRIC COMPANY DOCKET NO. 50-278 PEACH BOTTOM ATOMIC POWER STATION, UNIT NO. 3 AMENDMENT TO FACILITY OPERATING LICENSE 4
Amendment No, 115 License No. DPR-56 1.
The Nuclear Regulatory Commission (the Commission) has found that:
4 s
A.
The application for amendment by Philadelphia Electric Company, et al. (the licensee) dated April 19, 1984, as supplemented October 2, 1984, 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, the provisions of the Act, and the. rules and regulations of the Commission; 3
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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; D.
The issuance of this amendment will not be inimical to the common 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 Commission's regulations and all applicable requirements have been satisfied.
2.
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 Operating License No. DPR-56 is hereby amended to read as follows:
l l
t Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.115, are hereby incorporated in the license. PEC0 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 Q
F. Stolz, Chief rating Reactors Branch #4 l
Division of Licensing
Attachment:
Changes to the Technical Specifications Date of Issuance: October 2,1985
ATTACHMENT TO LICENSE AMENDMENT NO.115 FACILITY OPERATING LICENSE NO. DPR-56 DOCKET NO. 50-278 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages. The revised pages are identified by Amendment number and contain a vertical line indicating the area of change.
Remove Insert 11a 11a 12 12 15 15 21 21 61 61 63 63 72 72 79 79 80 80 89 89 90 90*
182 182 199 199 252 252
- Overleaf page included for document completeness.
4
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PBAPS SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING B.
Core Thermal Power Limit B. APRM Rod Block Trip Setting Reactor Pressure < S00 psia)
SRB < (0.66 W + 42% - 0.66 AW)
(FRP)
~FLPD M
where FRP = fraction of rated thermal power (3293 MWt).
MFLPD = maximum f raction of limiting power density where the limiting Power density is 13.4 KW/ft for all 8 x 8 fuel.
The ratio of FRP to MFLPD shall be set equal to 1.0 unless the actual operating value is less than the design value of 1.0, in which case the actual operating value will be used.
C.
Whenever the reactor is in the C.
Scram and isolation--> 538 in, above shutdown condition with reactor low water vessel zero irradiated fuel in the reactor level (0" on level vessel, the water level shall instruments) not be less than minus 160 inches indicated level (378 inches above vessel zero).
Amendments Nos. JP, J&,115
-lla-9 e
+
_____ SAFETY LIMIT
' LIMITING SAFETY SYSTEM SET?ING 2.1 (Cont'd)
D.
Scram-- turbine stop (10 percent valve closure E.
Scram-- turbine control fast closure on loss of control oil pressure.
500cPc850 psig.
F.
Scram--low
> 23 inches condenser vacuum Hg vacuum G.
Scram--main steam (10%
line isolation valve closure H.
>850 psig isolation valve closure--nuclear system low pressure I.
> minus 160 in.
actuation--reactor Tndicated level low-low-low
(> 378 inches water level above vessel zero)
J.
> minus 48 in, actuation--reactor Indicated level low-low water
(> 490 inches level above vessel zero)
K.
> minus 160 in.
isolation valve indicated level
' closure--reactor
(> 378 inches low-low-low above vessel water level zero) 1 i
t 6
Amendment No. H,115 c
l PBAPS 1.1.C BASES (Cont'd.)
However, for this specification a Safety Limit violatioh will be assumed when a scram is only accomplished by means of a -backup feature of the plant design.
The concept of not approaching a safety Limit, provided scram signals are operable, is supported by the extensive plant safety analysis.
The computer provided with Peach Bottom Unit 3 has a sequence annuciation program which will indicate the sequence in which events such as scram, APRM trip initiation, pressure scram initiation, etc. occur.
This program also indicates when the scram setpoint is cleared.
This will provide information on how long a scram condition exists and thus provide some measure of the energy added during a transient.
Thus, computer information normally will be available for analyzing scramst however, if the computer information should not be available for any scram analysis, Specification 1.1.C will be relied upon to determine if a Safety Limit has been violated.
D.
Reactor Water Level (Shutdown Condition)
During periods when the reactor is shutdown, 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 can be cooled sufficiently should the water level be reduced to two-thirds the core height.
Establishment of the safety limit at minus 160 inches indicated level (378 inches above vessel zero) provides adequate margin to assure sufficient cooling during shutdown conditions.
This level will be continuously monitored. -
E.
References 1.
General Electric BWR Thermal Analysis Basis (CETAB): Data, Correlation and Design Application, January 1977 (NEDO-10958-A).
2.
Process Computer Performance Evaluation Accuracy, General Electric Company BWR Systems Department, June 1974 (NEDO-20340).
3.
" General Electric Boiling Water Reactor Generic Reload Fuel Application", NEDE-24011-P-A.
Amendments Nos. J3', R', f>F 115,
l L
-m-~~-----
l PBAPS 2.1 BASES (Cont'd.)
C.
Reactor Water Low Level Scram and Isolation (Except' Main st e a Ef_ in e s T-----= =
l
~~~==
The setpoint for the low level scram is above the bottom of the separator skirt.
This level has been used in transient analyses dealing with coolant inventory decrease.
The results reported in FSAR subsection 14.5 show that scram and isolation of all process lines (except main steam) at this level adequately protects the fuel and the pressure barrier, because MCPR is greater than the fuel cladding integrity safety limit in all cases, and system pressure does not reach the safety valve settings.
The scram
'
- setting is approximately 23 inches below the normal operating range and is thus adequate to avoid spurious scrams.
D.
Turbine Stop Valve Closure Scram The turbine stop valve closure scram trip anticipates the pressure, neutron flux and heat flux increase that could result from rapid closure of the turbine stop valves.
With a scram trip setting of less than or equal to 10 percent of valve closure from full open, the resultant increase in surface heat flux is limited such that MCPR remains above the fuel cladding integrity safety limit even during the worst case transient that assumes the turbine bypass is closed.
This scram is bypassed when turbine steam flow is below 30% of rated, as measured by turbine first
, stage pressure.
3 E.
Turbine Control Valve Scram l
The turbine control valve fast closure scram anticipates the pressure, neutron flux and heat flux increase that could result from fast closure of the turbine control valves due to a load l
rejection exceeding the capacity of the bypass valves or a failure in the hydraulic control system which results in a loss of oil pressure.
This scram is initiated from pressure switches in the hydraulic control system which sense loss of oil pressure due to the opening of the fast acting solenoid valves or a failure in the hydraulic control system piping.
Two turbine first stage pressure switches for each trip system initiate automatic bypass of the turbine control ~ valve' fast closure scram when the first stage pressure is below that required to produce 30% of rated power.
Control valve closure time is approximately twice as long as that for stop valve closure.
Amendments Nos. X, AP, K 115 "
-.n.
v--..--
t TAlli E_3. 2. A
.F g
INSTRUMENTATION TilAT INITIATES PRIMARY CONTAINMENT ISOLATION 2
m
_=
of Operable Number of Instrument p
Instrument Instrument Trip Level Setting Channels Provided Action
.T Channels per By Design (2)
Trip System (1) hk 2 (6)
Reactor Low Water
> 0" Indicated 4 Inst. Channels A
Level Level (3) 1 Reactor High Pressure j 75 psig 2 Inst. Channels D
i (Shutdown Cooling Isolation) e 2
Reactor Low-Low-Low at or above -160" 4 Inst. Channels A
Water Level indicated level (4) 2 (6)
High Drywell Pressure 3 2 psig 4 Inst. Channels A
2 High Radiation Main j 3 X Normal Rated (8) (10) 4 Inst. Channels B
Steam Line Tunnel Full Power Background 2
Low Pressure Main
> 850 psig (7) 4 Inst. Channels B
i Steam Line 2 (5)
High Flow Main 3 140% of Rated 4 Inst. Channels B
Steam Line Steam Flow 2
Main Steam Line f 200 deg. F (9) 4 Inst. Channels B
Tunnel Exhaust Duct liigh Temperature s
PBAPS NCTIS FOR TA3*I 3.2.A 1.
Wheneve: ?:imary containment integriiy is required by Se:: ion 3.7, the:e snall ce two operable or ::ipped ::ip systems fe:
each funesion.
- f :he fi:s: =clumn carnet be met for one of the ::ip systems, tha: ::ip sys:em shall be ::ipped or the app: priate a:tica listed be.'ow sna11 de taken:
A.
Initiate an orderiv shutdown and have the reae:o in Cold She:down Condition in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
3.
Initiate an orderly load reduction and have Main Steam L4es isciated within eight hours.
C.
Isolate Rea or Water Cleanup System.
D.
Isolate Shutdown Cooking.
I.
Iselate Reactor Water Cleanup Filter Demineralizers unless the fellowing p cvision is satisfied.
The RW U F;1:e: Demineralizer may be used (the iscla:icn everridden) to route the resetor water to the main condense
- waste surge sank, with the hign temperatu:e
- p incperable for up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, provided the wa:e:
inlet tem erature is monitored once per heu: and c nfirmed to ce celow 190 degrees F.
3.
Instrument setpoint corresponds to 539 inchas above vessel zero.
-4.
Instrument setpoint corresponds to 378 inches above vessel zero.
5.
Two recuired'for each steam line.'
6.
These signals also start 330:5 and is.itiate secondary containment is=la: ion.
7 Only required in Run Mode (interlocked with Mode Switch).
B.
At a radiation level of 1.5 times the normal rated power background, an ala:m will be tripped in the cont:ol room to alert the control room operators to an increase in the main steam line tunnel radiation level.
9.
In the event of a loss of ventilaItion in the main steam line tunnel area, the main steam line tunnel exhaust duct high temperature se: point may be raised up to 250 degrees F for a period not to exceed 30 minutes to permit' restoration of the ventilation flow.
During the 30-minute period, an cperator l
(
shall observe control room indications of the duct i
temperature so in the event of rapid increases (indicative of a steam line b:eak) the operator shall promptly close the main steam line isolation valves.
Amendment No.
O, Je( 115 I
PBAPS Notes for Table 3.2.B 1.
Whenever any CSCS subsystem is required by Section 3.5 to be operablo, there shall be two operable trip systems.
If the first column cannot be met for one of the trip systems, that trip system shall be placed in the tripped condition or the reactor shall be placed in the Cold Shutdown Condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
2.
Close isolation valves in RCIC subsystem.
3.
Close isolation valves in HPCI subsystem.
4.
Instrument set point corresponds to 378 inches above vessel zero.
5.
HPCI has only one trip system for these sensors.
Amendment No. )( 115 #
TABLE 3.2.G INSTRUMENTATION TilAT INITIATES RECIRCULATION PUMP TRIP a
Minimum No.
Instrument Trip Level Setting Number of Instrument Action g>
of Operable Channels Provided y
Instrument by Design 9
Channels Per Trip System (1) 1 Reactor High Pres-
< 1120 psi 9 4
sure (2) i 1
Reactor Low-Low Water
> -48 in. indicated 4
l Level level (2)
Notes for Table 3.2.G 1.
Whenever the reactor is in the RUN Mode, there shall be one operable trip system for each i
parameter for each operating recirculation pump.
shall be taken.
If this cannot be met, the indicated action z
2.
Reduce power and place the mode selector-switch in a mode other than the RUN Mode.
V e
TABLE 4.2.A MINIMUM TEST AND CALIBRATION FREQUENCY FOR PCIS Instrument Channel (5)
Instrument Functional Test Calibration Freguency Instrument Check 1)
Reactor High Pressure (1)
Once/3 months None (Shutdown Cooling Permissive) 2)
Reactor Low-Low-Low (1) (3)
Once/ operating cycle Once/ day Water Level (7) 3)
Main Steam High Temp.
(1) (3)
Once/ operating cycle Once/ day 4)
Main Steam High Flow (7)
(1) (3)
Once/ operating cycle Once/ day 5)
Main Steam Low Pressure (1)
Once/3 months None 1
6)
Once/3 months once/ day High Flow 7)
Once/3 months None High Temp.
Logic System Functional Test (4) (6)
Freguency i
1)
Main Steam Line Isolation Vvs.
Once/6 months Main Steam Line Drain Vvs.
Reactor Water Sample Vvs.
2)
RHR - Isolation Vv. Control Once/6 months Shutdown Cooling vvs.
Head Spray 3)
Reactor Water Cleanup Isolation Once/6 months 4)
Drywell Isolation Vvs.
Once/6 months i
TIP Withdrawal Atmospheric Control Vvs.
Sump Drain Valves 5)
Standby Gas Treatment System Once/6 months Reactor Building Isolation
)
t Amendment No. Jf, 115 -
i
_ _ ~. _. _ _ _. _
PBAPS 3.2 BASES In addition to reactor protection instrumentation which initiates a reactor scram, protective instrumentation has been provided which initiates action to mitigate the consequences of accidents which are beyond the operator's ability to control, or terminates operator errors before they result in serious consequences.
This set of specifications provides the limiting conditions of operation for the primary system isolation function, initiation of the core cooling systems, control rod block and standby gas treatment systems.
The objectivec of the Specifications are (i) to assure the effectiveness of the protective instrumentation when required even during periods when portions of such systems are out-of-service for maintenance, and (ii) to prescribe the trip settings required to assure adequate performance.
When necessary, one channel may be made inoperable for brief intervals to conduct required functional tests and calibrations.
a Some of the settings on the instrumentation that initiate or control core and containment cooling have tolerances explicitly stated where the high and low values are both critical and may have a substantial effect on safety.
The set points of other instrumentation, where only the high or low end of the setting has a direct bearing on saf ety, are s
chosen at a level away from the normal operating range to prevent inadvertent actuation of the safety system involved and exposure to abnormal situations.
I Actuation of primary containment valves is initiated by j
protective instrumentation shown in Table 3.2.A which senses the conditions for which isolation is required.
Such instrumentation must be available whenever primary containment integrity is required.~
=.
The instrumentation which initiates primary system isolation I
is connected in a dual bus arrangement.
The low water level instrumentation set to trip at zero 4
inches indicated level (538 inches above vessel zero) closes l
all isolation valves except those in Groups 1. 4 and 5.
Details of valve grouping and required closing times are f
given in Specification 3.7, For valves which isolate at this level, this trip setting is adequate to prevent core uncovery in the case of a break in the largest line assuming a 60 second valve closing time.
Required closing times are less 4
than this, i
The low-low reactor water level instrumentation is set to trip when reactor water level is minus 48 inches indicated level (490 inches above vessel zero).
This trip initiates i
the HPCI and RCIC and trips the recirculation pumps.
The l
low-low-low reactor water levn1 instrumentation is set to l
trip when the reactor water level is minus 160 inches indicated level (378 inches above vessel sero).
This trip i
closes Main Steam Line Isolation Valves, Main Steam Drain Valves and Recire Sample Valves (Group 1), activates the remainder of the CSCS subsystem, and starts Amendment No.
115
i 3.2 BASES (Cont'd) the smargency diesel generators.. These trip level settings l
were chosen to be high enough to' prevent spurious actuation but low enough to initiate CSCS operation and primary system isolation so that post accident cooling can be accomplished and the guidelines of 10 CFR 100 will not be exceeded.
For large breaks up to the complete circumferential break of a i
28-inch recirculation line and with the trip setting given J
above, CSCS initiation and primary system isolation are i
initiated in time to meet the above criteria.
Reference
)
paragraph 6.5.3.1 FSAR.
The high drywell pressure instrumentation is a diverse signal for malfunctions to the water level instrumentation and in addition to initiating CSCS, it causes isolation of Group 2 and 3 isolation valves.
For the breaks discussed above, this instrumentation will generally initiate CSCS. operation before the low - low - low water level instrumentation; thus the results given above are applicable here also.
See Spec.
3.7 for Isolation Valve Closure Group.
The water level instrumentation initiates protection for the full spectrum of loss-of-coolant accidents and causes isolation of all isolation valves except Groups 4 and 5.
Venturis are provided in the main steam lines as a means of j
measuring steam flow and also limiting the loss of mass inventory from the vessel during a steam line break accident.
The primary function of the instrumentation is to detect a break in the main steam line.
For the worst case accident, main steam line break outside the drywell, a trip setting of 140% of rated steam flow in conjunction with the flow limiters
~
and main steam line valve closure, limits the mass inventory
-l loss such that fuel is not uncovered, fuel temperatures peak at approximately 10000F and release of radioactivity to the l
environs is below CFR 100 guidelines.
Reference Section 14.6.5 FSAR.
Temperature monitoring instrumentation is provided in the main steam line tunnel exhaust duct and along the steam line in the turbine building to detect leaks in these areas. Trips are provided on this instrumentation and when exceeded, cause
)
closure of isolation valves.
See Spec. 3.7 for Valve Group.
]
The setting is 2000F for the main steam line tunnel detector.
1 For large breaks, the high steam flow instrumentation is a O
backup to the temperature instrumentation.
no High radiation monitors in the main steam line tunnel have 5
been provided to detect gross fuel failure as in the control d
rod drop accident.
With the established setting of 3 times normal background, and main steam line isolation valve closure, h1*
fission product release is limited so that 10 CFR 100 guide-lines are not exceeded for this accident.
Reference Section 14.6.2 FSAR.
e.
l APRIL 1974
PBAPS NOTES FOR TABLE NO. 3.7.1
~~~~~---~~~~~~~~~-~~--
Key:
0
= Open C
= Closed SC = Stays Closed GC = Goes Closed Mote:
Isolation groupings are as follows:
GROU? 1:
The valves in Group 1 are actuated by any one of the following conditions:
1.
Reactor vessel low-low-low water level.
l
- 2. Main steam line high radiation.
1
- 3. Main steam line high flow.
4.
Main steam line space high temperature.
- 5. Main steam line low pressure (RUN mode only).
GROUP 2A:
The valves in Group 2A are actuated by any one of the following conditions:
1.
Reactor vessel low water level.
2.
Reactor water cleanup system heat exchanger discharge high temperature.
3.
Reactor water cleanup system suction line break.
4.
Standby liquid control system actuation.
GROUP 20:
Tne valves in Group 23 are actuated by any one of the following conditions:
- 1. Reactor vessel low water level.
- 2. High drywell pressure.
- 3. Reactor high pressure of shut,down mode.
GROUP 2C:
The valves in Group 2C are actuated by any one of the following conditions:
- 1. Reactor low water level.
- 2. High reactor vessel pressure, (600 PSIG)
- 3. High drywell pressure.
GROUP 2D:
The valves in Group 2D are actuated by the following conditions:
- 1. High drywell pressure.
- 2. Reactor low water level.
GROUP 3:
The valves in Group 3 are actuated by any one of the following conditions:
Amendment No. )( 115
-182-i l
. _, _ _ _ ~ _ _ _., _.
i PBAPS i
-BASES 3.7.D & 4.7.D Primary Containment Isolation Valves Double isolation valves are provided on lines penetrating the primary containment and open to the free space of the containment.
1 Closure of one of the valves in each line would be sufficient to maintain the integrity of the pressurs suppression system.
Automatic initiation is required to minimize the potential leakage paths from the containment in the event of a loss-of-coolant accident.
Groug_1:
Actuation for valves associated with the isolation of i
the main steam system.
The main steam lines are isolated by reactor vessel low-low-low water level in order to allow for l
removal of decay heat subsequent to a scram, yet isolate in time i
for proper operation of the core standby cooling systems.
The valves in group 1 are also closed when process instrumentation detects excessive main steam line flow, high radiation, low i
pressure, or main steam space high temperature.
1 Group 2:
Actuation for valves associated with the isolation of the reactor auxiliary systems.
Some of the reactor auxiliary systems such as the RWCU and RHR shutdown cooling systems connect into the reactor coolant boundary while others such as the drywell
~
equipment and floor drain discharga valves do not penetrate the reactor coolant boundary.
Group 2 actuation is subdivided as follows:
Group 2A - process lines are normally in use and it is therefore not desirable to cause spurious isolation due to high drywell pressure resulting from non-safety related causes.
To protect the reactor from a possible pipe break in j
the system, isolation is provided by high temperature at the j
cleanup system heat exchanger / outlet or high flow through the inlet to the cleanup system.
Also, since the vessel could i
potentially be drained through the cleanup system, a low level isolation is provided.
An alarm of high temperature in l
the cleanup system area will provide an indication of suction line break resulting in manual isolation of the system.
During actuation of the standby liquid control system, the l
cleanup system is isolated.
l Group 2B - isolation valves are not normally in use and are j
closed by reactor vessel low water level, high drywell pressure or high reactor pressure of the shutdown mode.
Group 2C - isolation valves can only be opened when the reactor is at low pressure and the core standby cooling j
systems are not required.
Also, since the reactor vessel could potentially be drained through these process lines, l
these valves are closed by low water level.
1 i
i l
Amendment No. R, 115
_199_
PBAPS 6.5.2.8 Centinued e.
The Facility Emergency Plan and implementing procedures at least once per year.
l f.
The Facility Security Plan and implementing procedures at least enee per two years, g.
The Offsite Des e Calculaticn Manual and implementing procedures at least once per two years.
h.
The perfer=ance of activities recuired by the Ouality Assurance Program regarding the radiological acnitoring program to meet the provisions of Regulatory Guide 4.1, Revisien 1, April 1975, at least once per calendar year, i.
Any other area of f acility operation considered appropriate by the OSR Committee or the Vice Presi d ent, Electric Producticn.
Authority 6.5.2.9 The OSR Cc==ittee shall report to and advise the Vice President, Electric Production en those areas of responsibility specified to Section 6.5.2.7 and 6.5.2.8.
Reecrds 6.5.2.10 Records of OSR Committee activities shall be prepared, approved, and distributed as indicated belows a.
Minutes of each OSR Committee meeting shall be prepared, approved and forwarded to the Vice President, Electric Production within 14 days following each meeting.
b.
Reports of review encompassed by Section 6.5.2.7.e, f,g, and h above, shall be prepared,
approved and forwarded to the Vice President, Electric Productice within 14 days following completion of the review.
Amencment No. J9, 37, F,115
-2n-I
[
_ - -.