ML20215L056
| ML20215L056 | |
| Person / Time | |
|---|---|
| Site: | Seabrook  |
| Issue date: | 10/15/1986 |
| From: | Novak T Office of Nuclear Reactor Regulation |
| To: | Harrison R PUBLIC SERVICE CO. OF NEW HAMPSHIRE |
| References | |
| NUDOCS 8610280445 | |
| Download: ML20215L056 (38) | |
Text
..
15 00T 1986 Docket Nos.:
50-443 and 50-444 Mr. Robert J. Harrison President & Chief Executive Officer Public Service Company of New Hampshire
Dear Mr. Harrison:
SUBJECT:
CHANGES TO THE SEABROOK FINAL. DRAFT TECHNICAL SPECIFICATIONS In a letter dated June 18, 1986, the Seabrook Unit 1 Final Draft Technical Specifications (TSI were transmitted to you for certification, under oath and affirmation that the Final Draft TS are consistent with the Final Safety Analysis Report (FSAR), the Safety Evaluation Report (SERI, and the as-built facility.
In letters dated June 26, 1986 and October 10, 1986 some changes to the Seabrook Final Draft TS were forwarded to you for incorporation into the Final Draft transmitted to you on June 18, 1986.
Enclosed are the remaining changes to be incorporated into the Final Draft transmitted to you on June 18, 1986.
You are requested to certify, under oath and affirmation, that all the final revised Final Draft TS are consistent with the FSAR, the SER and its supplements and the as-built facility. Your certification is requested by October 16, 1986.
1Thomas M. Novak, Acting Director Division of PWR I.icensing-A
Enclosure:
As stated cc: See next page Distribution '
V. Benaroya Docket File-NRC PDR l.,
Moon I.ocal PDR PWR PD#5 N. Thompson J. Partlow B. Grimes V. Nerses OGC-Bethesda M. Rushbrook T. Novak P
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8610280445 861015 PDR ADOCK 05000443 A
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NUCLEAR REGULATORY COMMISSION h
, j WASHINGTON, D. C. 20655 e
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15 OCT 1986 Docket Nos.:
50-443 and 50-444 Mr. Robert J. Harrison President & Chief Executive Officer Public Service Company of New Hampshire
Dear Mr. Harrison:
SUBJECT:
CHANGES TO THE SEABROOK FINAI. DRAFT TECHNICAI. SPECIFICATIONS In a letter dated June 18, 1986, the Seabrook Unit 1 Final Draft Technical Specifications (TS) were transmitted to you for certification, under oath and affirmation that the Final Draft TS are consistent with the Final Safety Analysis Report (FSAR), the Safety Evaluation Report (SER), and the as-built facility.
In letters dated June 26, 1986 and October 10, 1986 some changes to the Seabrook Final Draft TS were forwarded to you for incorporation into the Final Draft transmitted to you on June 18, 1986.
Enclosed are the remaining changes to be incorporated into the Final Draft transmitted to you on June 18, 1986.
You are requested to certify, under oath and affirmation, that all the final revised Final Draft TS are consistent with the FSAR, the SER and its supplements and the as-built facility.
Your certification is reauested by October 16, 1986.
I f
' %, }i 'q4 t\\i t
Thomas M. dovak, Acting Director Division of PWR I.icensing-A
Enclosure:
As stated l
cc: See next page A-
Mr. Robert J. Harrison Public Service Company of New Hampshire Seabrook Nuclear Power Statior cc:
Thomas Dignan, Esq.
E. Tupper Kinder, Esc.
John A. Ritscher, Esq.
G. Dana Bisbee, Esc.
Ropes and Gray Assistant Attorney General 225 Franklin Street Office of Attorney General Boston, Massachusetts 02110 208 State Hosue Annex Concord, New Hampshire 03301 Mr. Bruce 8. Beckley, Pro.iect Manager Public Service Company of New Hampshire Resident Inspector Post Office Box 330 Seabrook Nuclear Power Station Manchester, New Hampshire 03105 c/o US Nuclear Regulatory Commission Post Office Box 700 Dr. Mauray Tye, President Seabrook, New Hampshire 03874 Sun Valley Association 209 Summer Street Mr. John DeVincentis, Director Paverhill, Massachusetts 01839 Engineering and licensing Yankee Atomic Electric Company Robert A. Backus, Esc.
1671 Worchester Road O'Neil, Backus and Spielman Framingham, Massachusetts 01701 116 1.owell Street Manchester, New Pampshire 03105 Mr. A. M. Ebner, Project Manager United Engineers & Constructors William S. Jordan, III 30 South 17th Street Diane Curran Post Office Box 8225 Parmon, Weiss & Jordan Philadelphia, Pennsylvania 19101 20001 S Street, NW Suite 430 Washington, D.C.
20009 Mr. Philip Ahrens, Esq.
Assistant Attorney General State House, Station #6 Aucusta, Maine 04333 Carol S. Sneider Esq.
Office of the Assistant Attorney General Environmental Protection Division Mr. Warren Pall One Ashburton Place Public Service Company of Boston, Massachusetts 02108 New Hampshire Post Office Box 330 D. Pierre G. Cameron, Jr., Esq.
Seabrook, New Hampshire 03874 General Counsel Public Service Company of New Hampshire Seacoast Anti-Pollution f.eague Post Office Box 330 Ms. Jane Doughty Manchester, New Hampshire 03105 5 Market Street Portsmouth, New Fampshire 03801 Regional Administrator, Region I U.S. Nuclear Regulatory Commission Mr. Diana P. Randall 631 Park Avenue 70 Collins Street Kina of Prussia, Pennsylvania 19406 Seabrook, New Hampshire 03874 Richard Hampe, Esq.
New Hampshire Civil Defense Agency 107 Pleasant Street Concord, New Hampshire 03301
)*
=
Public Service Company of Seabrook Nuclear Power Station New Hampshire cc:
Mr. Calvin A. Canney, City Manager Mr. Alfred V. Sargent, City Hall Chairman 126 Daniel Street Board of Selectmen Portsmouth, New Hampshire 03801 Town of Salisbury, MA 01950 Ms. Letty Hett Senator Gordon J. Humphrey Town of Brentwood ATTN: Tom Burack RFD Dalton Road U.S. Senate Brentwood, New Hampshire 03833 Washington, D.C.
20510 Ms. Roberta C. Pevear Mr. Owen B. Durgin, Chairman Town of Hampton Falls, New Hampshire Durham Board of Selectmen Drinkwater Road Town of Durham Hampton Falls, New Hampshire 03844 Durham, New Hampshire 03824 Ms. Sandra Gavutis Charles Cross, Esq.
Town of Kensington, New Hampshire Shaines, Mardrigan and RDF 1 McEaschern East Kingston, New Hampshire 03827 25 Maplewood Avenue Post Office Eox 366 Portsmouth, New Hampshire 03801 Chairman, Board of Selectmen RFD 2 South Hampton, New Hampshire 03827 Mr. Guy Chichester, Chaiman Rye Nuclear Intervention Mr. Angie Machiros, Chairman
' Committee Board of Selectmen c/o Rye Town Hall for the Town of Newbury 10 Central Road Newbury, Massachusetts 01950 Rye, New Hampshire 03870 Ms. Cashman, Chairman Jane Spector Board of Selectmen Federal Energy Regulatory Town of Amesbury Commission Town Hall 825 North Capital Street, NE Amesbury, Massachusetts 01913 Room 8105 Washington, D. C.
20426 Honorable Peter J. Matthews Mayor, City of Newburyport Mr. R. Sweeney Office of the Mayor New Hampshire Yankee Division City Hall Public Service of New Hampshire Newburyport, Massachusetts 01950 Company 7910 Woodmont Avenue Mr. Donald E. Chick, Town Manager Bethesda, Maryland 20814 Town of Exeter 10 Front Street Mr. William B. Derrickson Exeter, New Hampshire 03823 Senior Vice President Public Service Company of New Hampshire Post Office Box 700, Route 1 Seabrook, New Hampshire 03874
)
s see l'NACCEPTABLE OPERATION N0.LPsIa
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ACCEPTABLE
\\h OPERATION 3
ses 529 8.8 029 8.49 8.60 9.80 1.00 1.29 FRACTION OF RATED TERMAL POWER FIGURE 2.1-1 REACTOR CORE SAFETY LIMIT - FOUR LOOPS IN OPERATION SEABROOK - UNIT 1 2-2
POWER DISTRIBUTION LIMITS HEATFLUXHOTCHANNELFACTOR-Fg SURVEILLANCE REQUIREMENTS 4.2.2.2d. (Continued)
C RTP 2)
When the F is less than or equal to the F limit for the x
x appropriate measured core plane, additional power distribution P
maps shall be taken and F compared to F and F at least x
once per 31 EFPD.
e.
The F limits for RATED THERMAL POWER (F P) shall be provided for xy all core planes containing Bank "D" control rods and all unrodded core planes in a Radial Peaking Factor Limit Report per Specifica -
tion 6.8.1.6; f.
The F,y limits of Specification 4.2.2.2e., above, are not applicable in the following core planes regions as measured in percent of core height from the bottom of the fuel:
1)
Lower core region from 0 to 15%, inclusive, 2)
Upper core region from 85 to 100%, inclusive, 3)
Grid plane regions at 17.8 1 2%, 32.1 1 2%, 46.4 1 2%, 60.6 1 2%,
i and 74.9 1 2%, inclusive, and 4)
Core plane regions within i 2% of core height ('i 2.88 inches) about the bank demand position of the Bank "D control rods.
g.
With F exceeding Fx, the effects of F on F (Z) shall be evaluated x
xy 9
to determine if F (Z) is within its limits.
q 4.2.2.3 When F (Z) is measured for other than F determinations, an overall q
xy measured F (Z) shall be obtained from a power distribution map and increased 9
by 3% to account for manufacturing tolerances and further increased by 5% to account for measurement uncertainty.
SEABROOK - UNIT 1 3/4 2-7
8 TABLE 3.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION MINIMUM
^
TOTAL NO.
CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION E
y 1.
1 2
1, 2 1
2 1
2 3*, 4*, 5*
10 y
2.
Power Range, Neutron Flux a.
High Setpoint 4
2 3
1, 2 2#
b.
Low Setpoint 4
2 3
1###, 2 2#
3.
Power Range, Neutron Flux 4
2 3
1, 2 2#
High Positive Rate 4.
Power Range Neutron Flux, 4
2 3
1, 2 2#
s R
High NegatC,e Rate l
5.
Intermediate Range, Neutron Flux 2
1 2
1###, 2 3
6.
Source Range, Neutron Flux a.
Startup 2
1 2
2##
4 b.
Shutdown 2
0 1
3,4,5 5
c.
Shutdown 2
1 2
3*, 4*, 5*
10 7.
Overtemperature AT 4
2 3
1, 2 6#
8.
Overpower AT 4
2 3
1, 2 6#
9.
Pressurizer Pressure--Low 4
2 3
1**
6# (1) 10.
Pressurizer Pressure--High 4
2 3
1, 2 6# (1) 4 11.
Pressurizer Water Level--High 3
2 2
1**
6#
TABLE 3.3-1 (Continued) h REACTOR TRIP SYSTEM INSTRUMENTATION E8 MINIMUM
^
TOTAL NO.
CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION Ey 12.
Reactor Coolant Flow--Low 7
a.
Single Loop (Above P-8) 3/ loop 2/ loop in 2/ loop in 1
4#
g any oper-each oper-ating loop ating loop 1
b.
Two Loops (Above P-7 and 3/ loop 2/ loop in 2/ loop 1
4#
below P-8) two oper-each oper-l ating loops ating loop 13.
Steam Generator Water 4/stm. gen.
2/stm. gen.
3/sta. gen.
1, 2 6# (1)
Level--Low-Low in any oper-each oper-R ating stm.
ating stm.
gen.
gen.
j 14.
Undervoltage--Reactor Coolant 4-2/ bus 2-1/ bus 2 on one bus 1**
6#
Pumps
- 15. Underfrequency--Reactor Coolant 4-2/ bus 2-1/ bus 2 on one bus 1**
6#
Pumps 4
- i 16.
a.
Low Fluid Oil Pressure 3
2 2
1***
4#
b.
Turbine Stop Valve Closure 4
4 4
1***
11#
17.
Safety Injection Input from ESF 2
1 2
1, 2 9
18.
Reactor Trip System Interlocks l
a.
Intermediate Range Neutron Flux, P-6 2
1 2
2##
8 4
J
=
TABLE 3.3-1 (Continued)
ACTION STATEMENTS (Continued)
ACTION 3 - With the number of channels OPERABLE one less than the Minimum Channels OPERABLE requirement and with the THERMAL POWER level:
a.
Below the P-6 (Intermediate Range Neutron Flux Interlock)
Setpoint, restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER above the P-6 Setpoint, and b.
Above the P-6 (Intermediate Range Neutron Flux Interlock)
Setpoint but below 10% of RATED THERMAL POWER, restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER above 10% of RATED THERMAL POWER.
ACTION 4 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, suspend all operations involving' positive reactivity changes.
ACTION 5 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or open the Reactor Trip System breakers, suspend all operations involving positive reactivity changes and verify that valve RMW-V31 is closed and secured in position within the next hour.
ACTION 6 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:
The inoperable channel is placed in the tripped condition a.
within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and b.
The Minimum Channels OPERABLE requirement is met; however, the inoperable channel may be bypassed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for surveillance testing of other channels per Specification 4.3.1.1.
ACTION 7
' t. oud)
MT ith less than the Minimum Number of Channels OPERABLE, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> determine by observation of the associated permissive annunciator window (s) that the interlock is in its required state fnr the existing plant condition, or apply Specification 3.0.3.
ACTION 9 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; however, one channel may be bypassed for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for surveillance testing per Specification 4.3.1.1, provided the other channel is OPERABLE.
the number of OPERABLt channels one less tIiaii the Total
')
~
Number of Channels, STARTUP and/or POWER OPERATION may proceed until performance of the next required ANALOG CHANNEL OPERATIONAL '
'/
SEABROOK - UN: TEST provided the inoperable channel is placed in the tripped condition within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
.x'
,/
TABLE 3.3-1 (Continued)
ACTION STATEMENTS (Continued)
ACTION 10 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or open the Reactor Trip System breakers within the next hour.
ACTION 11 - With the number of OPERABLE channels less than the Total Number of Channels, operation may continue provided the inoperable channels are placed in the tripped condition within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
ACTION 12 - With one of the diverse trip features (undervoltage or shunt trip attachment) inoperable, restore it to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or declare the breaker inoperable and apply ACTION 9.
The breaker shall not be bypassed while one of the diverse trip features is inoperable except for the time requirbd for performing maintenance to restore the breaker to OPERABLE status.
SEABROOK - UNIT 1 3/4 3-7
TABLE 3.3-9 i
g; REMOTE SHUTDOWN SYSTEM E
TOTAL NO.
MINIMUM READOUT OF CHANNELS 7c INSTRUMENT LOCATION CHANNELS OPERABLE i
gg
- 1. Intermediate Range Neutron Flux CP-108 A and B 2
1
- q
- 2. Source Range Neutron Flux CP-108 A and B 2
1
- 3. Reactor Coolant Temperature -
y ide Range for Loops 1 and 4 a.
T CP-108 A 2
2 c
b.
T CP-108 8 2
2 H
- 4. Pressurizer Pressure CP-108 A and B 2
2
- 5. Pressurizer Level CP-108 A and B 2
2
- 6. Steam Generator Pressure CP-108 A and B 1/sta gen.
1/sta. gen.
i
- 7. Steam Generator Water Level CP-108 A and B 1/sta. gen.
1/stm. gen.
- 8. Steam Generator-Emergency Feedwater m
i 3
Flow Rate CP-108 A and B 1/sta. gen.
1/sta. gen.
- 9. Boric Acid Tank Level CP-108 A and B 1/ tank 1/ tank f;
TRANSFER SWITCHES / CONTROL CIRCUITS LOCATION l
_me u
e4 - -.. _.
...--_+A
-.4 A
,_._a
._A_.__g.
a 4..
TABLE 3.3-9 (Continued)
REMOTE SHUTDOWN SYSTEM y,
i 9
~o
-REAuvus i
8 TRANSFER SWITCHES / CONTROL CIRCUITS LOCATION l
i
E.
- 19. MS Isolation Valves MS-V-86/88/90/92 CP-108 A
{
Z
- 20. MS Isolation Valves MS-V-86/88/90/92 CP-108 8 i
- 21. Pressurizer Heaters, Group A CP-108 A g
- 22. Pressurizer Heaters, Group B CP-108 8
- 23. Charging Pump CS-P-2A BUS 5 SWGR
- 24. Charging Pump CS-P-28 BUS 6 SWGR
5
t'
- 31. High Pressure Injection SI-V-138 CP-108 A
- 32. High Pressure Injection SI-V-139 CP-108 8 Y
l t
4 l
j e
4 4
i
TABLE 4.3-6 (Continued)
RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS 8
R CHANNEL.
MODES FOR WHICH CHANNEL SOURCE CHANNEL OPERATIONAL SURVEILLANCE INSTRUMENT CHECK CHECK CALIBRATION TEST
- 15 REQUIRED w
3.
GASEOUS WASTE PROCESSING SYSTEM (Providing Alarm and Automatic Termination of Release) a.
Ga Activity Monitor
]<-@]
1]
]< h)
] <-O 4.
TURBINE GLAND SEAL CONDENSER EXHAUST
- a. Iodine Sampler W
N.A.
N.A.
N.A
- b. Particulate Sampler W
N.A.
N.A.
N.A.
- c. Sampler Flow Rate Indicator D
N.A.
N.A.
N.A.
)
4 I
~
i 1
REACTOR COOLANT SYSTEM gr 3/4.4.7 RELIEF VALVES LIMITING CONDITION FOR OPERATION 3.4.4 All power-operated relief valves (PORVs) and their associated block valves shall be OPERABLE.
APPLICABILITY: MODES 1, 2, and 3.
ACTION:
With one or more PORV(s) inoperable, because of excessive seat a.
leakage, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> either restore the PORV(s) to OPERABLE status or close the associated block valve (s); otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
b.
With one PORV inoperable due to causes other than excessive seat leakage, within I hour either restore the PORV to OPERABLE status or close the associated block valve and remove power from the block valve; restore the PORV to OPERABLE status within the following 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
4 With both PORV(s) inoperable due to causes other than excessive seat c.
leakage, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> either restore each of the PORV(s) to OPERABLE status or close their associated block valve (s) and remove power from the block valve (s) and be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
d.
With one or more block valve (s) inoperable, within 1 hour:
(1) restore the block valve (s) to OPERABLE status, or close the block valve (s) and remove power from the block valve (s), or close the PORV and remove power from its associated solenoid valve; and (2) apply the ACTION b. or c. above, as appropriate, for the isolated PORV(s).
The provisions of Specification 3.0.4 are not applicabit.
e.
t 1
SEABROOK - UNIT 1 3/4 4-11
cJ REACTOR COOLANT SYSTEM PRESSURE / TEMPERATURE LIMITS OVERPRESSURE PROTECTION SYSTEMS 3 77,
LIMITING CONDITION FOR OPERATION 3.4.9.3 At least one of the followit.g overpressure Protection Systems shall be OPERABLE:
a.
Two residual heat removal (RHR) suction re ief valves each with a setpoint of le.e ther, er.;.;l t: 450 psig or b.
Two power-operated relief valves (PORVs) with lift setpoints that vary with RCS temperature which do not exceed the limit established in Figure 3.4-4, or c.
The Reactor Coolant System (RCS) denressurized with an RCS vent of greater than or equal to 1.58 square inches.
APPLICABILITY: MODE 4 when the temperature of any RCS cold leg is less than or equal to 329'F; MODE 5 and MODE 6 w'.th the reactor vessel head on.
ACTION:
a.
With one PORV and one RHR suction relief valve inoperable, either restore two PORVs or two RHR suction relief valves to OPERABLE status within 7 days or depressurize and vent the RCS through at least a 1.58-square-inch vent within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, i
i b.
With both PORVs and both RHR suction relief valves inoperable, depressurize and vent the RCS through at least a 1.58-square-inch vent within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
c.
In the event the PORVs, or the RHR suction relief valves, or the RCS vent (s) are used to mitigate an RCS pressure transient, a Special c
Report shall be prepared and submitted to the Commission pursuant to Specification 6.8.2 within 30 days. The report shall describe the circumstances initiating the transient, the effect of the PORVs, or the RHR suction relief valves, or RCS vent (s) on the transient, and any corrective action necessary to prevent recurrence.
d.
The provisions of Specification 3.0.4 are not applicable.
SEABROOK - UNIT 1 3/4 4-34
mg 2500 A
a
/
VAUD FOR THE F1RST 16 EFPY. SETPOINT
__ CONTAINS MARGIN Or 50 F FOR C
TRANSl[NT [rFECTS.
3 F
-t A
w 2000 l
G
-f Y
i E
/
uirs..rs or l
[
- P=412.7 + s.597 e 8*I; r>tso r E
)
i i
/
O 1500 j
/
w 2
i
/
t S
/
5 1000
}
3
/
y I
/
l
/
I
/
i 7
500 I
f a;.
I 200 i
so too iso 200 sao soo aso RCS TEMPER ATURE ('F)
FIGURE 3.4-4 RCS COLD OVERPRESSURE PROTECTION SETPOINTS l
=
EMERGENCY C0RE COOLING SYSTEMS 3/4.5.2 ECCS SUBSYSTEMS - T,y GREATER THAN OR EQUAL TO 350*F LIMITING CONDITION FOR OPERATION 3.5.2 Two independent Emergency Core Cooling System (ECCS) subsystems shall be OPERABLE with each subsystem comprised of:
a.
One OPERABLE centrifugal charging pump, b.
One OPERABLE Safety Injection pump, c.
One OPERABLE RHR heat exchanger, d.
One OPERABLE RHR pump, esnd e.
An OPERABLE flow path
- capable of taking suction from the refueling water storage tank on a Safety Injection signal and automatically transferring suction to the containment sump during the recirculation phase of operation.
APPLICABILITY:
MODES 1, 2, and 3**.
ACTION:
a.
With one ECCS subsystem inoperable, restore the inoperable subsystem to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
b.
In the event the ECCS is actuated and injects water into the Reactor Coolant System, a Special Report shall be prepared and sub:nitted to the Commission pursuant to Specification 6.8.2 within 90 days describ-ing the circumstances of the actuation and the total accumulated actuation cycles to date. The current value of the usage factor foreachaffectedSafetyInjectionnozzleshallbeprovidedinthis l
Special Report whenever its value exceeds 0.70.
- During MODE 3, the discharge paths of both Safety Injection pumps may be isolated by closing for a period of up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to perform surveillance testing as required by Specification 4.4.6.2.2.
- The provisions of Specifications 3.0.4 and 4.0.4 are not applicable for entry into MODE 3 for the centrifugal charging pump and the Safety Injection pumps declaredinoperablepursuanttoSpecification4.5.3.2providedthecentrifugal 7
charging pump and the Safety Injection pumps are restored to OPERABLE status within at least 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or prior to the temperature of one or more of the RCS cold legs exceeding 375'F, whichever comes first.
SEABROOK - UNIT 1 3/4 5-4
CONTAINMENT SYSTEMS PRIMARY CONTAINMENT CONTAINMENT LEAKAGE SURVEILLANCE REQUIREMENTS 4.6.1.2 (Continued) d.
Type B and C tests shall be conducted with gas at a pressure not less than P,, 49.6 psig, at intervals no greater than 24 months except for tests involving:
1)
Air locks, and 2)
Purge supply and exhaust isolation valves with resilient material seals.
The combined bypass leakage rate shall be determined to be less than e.
or equal to 0.60 L, by applicable Type B and C tests at least once per 24 months.
f.
Purge supply and exhaust isolation valves with resilent material seals shall be tested and demonstrated OPERABLE by the requirements of Speci-fication 4.6.1.7.3 or 4.6.1.7.#, as applicable; 2.-
3 g.
Air locks shall be tested and demonstrated OPERABLE by the requirements of Specification 4.6.1.3; and h.
The provisions of Specifications 4.0.2 are not applicable.
SEABROOK - UNIT 1 3/4 6-4
~
^
PLANT SYSTEMS TURBINE CYCLE ATMO5PHERIC RELIEF VALVES LIMITING CONDITION FOR OPERATION 3.7.1.6 At least four atmospheric relief valves and associated manual controls including the safety-related gas supply systems shall be OPERABLE.
APPLICABILITY:
MODES 1, 2, 3, and 4.*
ACTION:
7 (t,With$ssthantherequiredatmosphericreliefvalvesOPERABLE, estore the required atmospheric relief valves to OPERABLE status within 72,,a,;; or be in at least HOT STANDBY within the next,8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
SURVEILLANCE REQUIREMENTS 4.7.1.6 Each atmospheric relief valve and associated manual controls including the safety-related gas s;pply systems shall be demonstrated OP.ERABLE:
a.
At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by verifying that the nitrogen accumulator tank is at a pressure greater than or equal to 500 psig.
b.
Prior to startup following any refueling shutdown or cold shutdown of 30 days or longer, verify that all valves will open and close fully by operation of manual controls.
N-Y bo
/ss s Y W.ey.& d~ap M
& be b e p/)N r e r /n e.
d Ad fA e L al-agkny A A c iwa s s u= s<
ll Mr lc tv. $oh 72 w
S YMOBy t u. 'Yk n d m{ f
- When steam generators are being used for decay heat removal.
SEABROOK - UNIT 1 3/4 7-10
9 PLANT SYSTEMS 3/4.7.5 ULTIMATE HEAT SINK LIMITING CONDITION FOR OPERATION 3.7.5 The ultimate heat sink (UHS) shall be OPERABLE with:
a.
A service water pumphouse water level at or above 5'-0", minus 37'-0" Mean Sea Levci, USGS datum, and gg b.
A mechanical draft cooling tower comprised of one cooling tow cell with one OPERABLE fan and a second cell with two OPERABLE fa s, and a contained basin water level of equal to or greater than feet at a bulk average water temperature of less than or equal to 67.3*F, and c.
A portable tower makeup pump system stored to be OPERABLE for 30 days following a Safe Shutdown Earthquake.
APPLICABILITY: MODES 1, 2, 3, and 4.
ACTION:
a.
With the service water pumphouse inoperable, restori the service water pumphouse to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
b.
With the mechanical draft cooling tower inoperable, restore the cooling tower to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
c.
With the portable tcwer makeup pump system inoperable, continue operation and notify the NRC within I hour in accordance with the procedure of 10 CFR 50.72 of actions or contingencies to ensure an adequate supply of makeup water to the mechanical draft cooling tower for a minimum of 30 days.
- With the cooling tower in operation with valves aligned for tunnel heat treat-ment, the tower basin level shall be maintained at greater than or equal to feet.
4s.55 SEABROOK - UNIT 1 3/4 7-14
PLANT SYSTEMS ULTIMATE HEAT SINK SURVEILLANCE REQUIREMENTS 4.7.5 The ultimate heat sink shall be determined OPERABLE:
At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by:
a.
1)
Verifying the water level in the service water pumphouse to be at or above 5'-0", minus 37'-0" Mean Sea Level, and 2)
Verifying the water in the mechanical draft cooling tower basin to be greater than or equal to a level of 35-t feet.
42.15 b.
At least once per week by verifying that the water in the mechanical draft cooling tower basin to be at a bulk average temperature of b tb ego,) to 67.3*F.
e At least once per 31 days by:
c.
1)
Starting from the control room each UHS cooling tower fan that is required to be OPERABLE and operating each of those fans for at least 15 minutes, and 2)
Verifying that the portable tower makeup pump system is stored in its design operational readiness state.
d.
At least once per 18 months by verifying automatic actuation of each cooling tower fan on a Tower Actuation test signal.
e SEABROOK - UNIT 1 3/4 7-15
.-n...
o PLANT SYSTEMS 3/4.7.6 CONTROL ROOM AREA VENTILATION SYSTEM LIMITING CONDITION FOR OPERATION 3.7.6 Two Control Room Area Ventilation Systems shall be OPERABLE.'
APPLICABILITY:
All MODES.
ACTION:
MODES 1, 2, 3, and 4:
With one Control Room Area Ventilation System inoperable, restore the inoperable system to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
MODES 5 and 6:
a.
With one Control Room Area Ventilation System inoperable, restore the inoperable system to OPERABLE status within 7 days or initiate and maintain operation of the remaining OPERABLE Control Room Area Ventilation System in the recirculation mode.
b.
With both Control Room Area Ventilation Systems inoperable, suspend all operations involving CORE ALTERATIONS or positive reactivity changes.
SURVEILLANCE REQUIREMENTS 4.7.6 Each Control Room Area Ventilation System shall be demonstrated OPERABLE:
a.
At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by verifying that the Control Room Area Ventilation System is maintaining the temperature of equipment and instrumentation in the control room area below its limiting equipment qualification temperature.
b.
At least once per 18 months or after any significant modification to the Control Room Area Ventilation Systems by verifying a system flow rate of 25,700 cfm i 10% through the air conditioner unit (3A and 3B) and a flow rate of at least 1200 cfm i 10% makeup from each intake to the emergency filtration unit with a discharge of 2300 cfm i 10% from the filtration unit.
sJe o l
l SEABROOK - UNIT 1 3/4 7-16
PLANT SYSTEMS CONTROL ROOM AREA VENTILATION SYSTEM SURVEILLANCE REQUIREMENTS 4.7.6 (Continued) c.
At least once per 18 months by:
1)
Verifying that on a high radiation signal from the control room makeup air intake, the subsystem automatically switches to the emergency recirculation mode of operation and the isolation dampers close within 5 seconds.
2)
Verifying that on an S signal the emergency filtration fans start.
3)
Verifying that the system maintains the control room area at a positive pressure of greater than or equal to a pressurization 1/8-inch Water Gauge relative to adjacent areas during system operation at '---
t' :- -
7::' t a pressurization flow of 1200 cfm i 10%.
SEABROOK - UNIT 1 3/4 7-17
ELECTRICAL POWER SYSTEMS A.C. SOURCES OPERATING SURVEILLANCE REQUIREMENTS 4.8.1.1.2 (Continued) a)
Verifying deenergization of the emergency busses and load shedding from the emergency busses, and Verifying the diesel starts on the 91+^ff'r't signaT,NW* G"
/as.s e b) t energizes the emergency busses with permanently connected loads within 12 seconds, energizes the auto-connected shutdown loads through the emergency power sequencer and operates for greater than or equal to 5 minutes while its:
generator is loaded with the shutdown loads.
After ener-gization, the steady-state voltage and frequency of the emergency busses shall be maintained at 4160 420 volts and 60 1 1.2 Hz during this test.
5)
Verifying that on an SI actuation test signal, without loss-of-offsite power, the diesel generator starts on the auto-start signal and operates on standby for greater than or equal to 5 minutes. The generator voltage and frequency shall be 4160 1 420 volts and 60 1 1.2 Hz within 10 seconds after the auto-start signal; the steady-state generator voltage and frequency shall be maintained within these limits during this test; 6)
Simulating a loss-of-offsite power in conjunction with an SI ac uation test signal; '--4 t;.;;r actuati. test eiw.al (T^.)1.__.c j
and a)
Verifying deenergization of the emergency busses and load shedding from the emergency busses;
/0 b)
Verifying the diese arts on the auto-start signal, energizes the em ency busses with permanently connected loads within seconds, energizes the auto-connected emergency (accident) loads through the emergency power sequencer and operates for greater than or equal to 5 minutes while its generator is loaded with the emergency loads.
After energization, the steady-state voltage and frequency of the emergency busses shall be maintained at 4160 420 volts and 60 1.2 Hz during this test; and c)
Verifying that all automatic diesel generator trips, except engine overspeed, low lube oil pressure, 4160-volt bus fault, and generator differential, are automatically bypassed upon loss of voltage on the emergency bus concur-rent with a Safety Injection actuation signal.
SEAb?)0K - UNIT 1 3/4 8-6
TABLE 4.8-1 DIESEL GENERATOR TEST SCHEDULE NUMBER OF FAILURES IN NUMBER OF FAILURES IN LAST 20 VALID TESTS
- LAST 100 VALID TESTS
- TEST FREQUENCY
<1
<5 At least once per 31 days
[2
[6 At least once per 7 days **
- Criteria for determining the number of failures and number of valid tests shall be in accordance with Regulatory Positior. C.2.e of Regulatory Guide 1.108, E~/.vov p Aqvs fj777; but determined on a per diesel generator basis.
For the purpose of determining the required test frequency, the previous test failure count may be reduced to zero if a complete diesel overhaul to like-new condition is completed, provided that the overhaul, including appropriate post-maintenance operation and testing, is specifically approved l
by the manufacturer and if acceptable reliability has been demonstrated. The reliability criterion shall be the successful completion of 14 consecutive tests in a single series. Ten of these tests shall be in accordance with the routine Surveillance Requirements 4.8.1.1.2.a.5 and 4.8.1.1.2.a.6 and four tests in accordance with the 184-day testing requirement of Surveillance Requirements 4.8.1.1.2.a.5 and 4.8.1.1.2.a.6.
If this criterion is not satisfied during the first series of tests, any alterr. ate criterion to be used to transvalue the failure count to zero requires NRC approval.
- This test frequency shall be maintained until seven consecutive failure-free demands have been performed and the number of failures in the last 20 valid demands has been reduced to one or less.
SEABROOK - UNIT 1 3/4 8-10 l
EM[W& Wbd !
{
4 P:*v ELECTRICAL POWER SYSTEMS ELECTRICAL EQUIPMENT PROTECTIVE DEVICES
(
CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES ANO N 7e c n e Geww3 Fa st & c As.s E E' Pa.~ e a h o ru cs Gmverreo 7"o ne rv. L tes s TC Liarv>Tr IIMITING CONDITION FOR OPERATION 3.8.4.2 Each containment penetration conductor overcurrent protective device3 shall be OPERABLE.
s (l
'O M g n /,ed.n. k APPLICABILITY: MODES 1, 2, 3, and 4 y
_][y{
[g- - '
ACTION:
y (v, With one or more of the,gn'[--
^ g ;t ;t'; ;; ;'rOrr eyr purr...,^.-protective device (s) inoperable:
I)
Restore the protective device (s) to OPERABLE status or deenergize p.
the circuit (s) by tripping the associated circuit breaker or racking out or removing the inoperable protective device within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, declare the affected system or component inoperable, and verify the circuit breaker to be tripped or the inoperable protective device to' he racked out or removed at least once per 7 days thereafter; the provisions of Specification 3.0.4 are not applicable to overcurrent devices in circuits which have their circuit breakers tripped, or their inoperable protective devices racked out, or removed; or p.
Be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
l
)n.sk k.
[mL.- "_
^
y
~.
SURVEILLANCE REQUIREMENTS 4.8.4.2 Each-2...;.-..= ;:n:--M an
^
- r -
- n-e protective device
^ '
rn""
shall be demonstrated OPERABLE:
a.
At least once per 18 months:
By verifying that the medium voltage 13.8-kV circuit breakers 1) are OPERABLE by selecting, on a rotating basis, at least one of the circuit breakers, and performing the following:
a)
A CHANNEL CALIBRATION of the associated protective relays (because of the large currents involved, it is impractical to inject primary side signals to current transformers; therefore, the channel calibration will be performed by injecting a signal on the secondary side of those trans-I formers at their test plug),
h-M 4
enjy&
L / E~ q+ S w g, l
SEABROOK - UNIT 1 3/4 8-21
n s c p.p I17 &.ve s' -v %
3
~
- 'With one or more of the r__-
._,.=nrniez ;;.2='--21:n _.J;t protectivi device (s) inoperablef "
A
- f p estore the protective device (s) to OPERABLE status or deenergize the circuit (s) by tripping the enamqutumm8 circuit breaker or racking out or removing the inoperable protective device within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, declare the affected spagguer component inoperable, and verify the circuit breaker to be tripped or the inoperable protective device to be racked out or removed at least once per 7 days thereafter; the provisions of $pecification 3.0.4 era not applicable to overcurrent /
devices in circuita which have their circuit breakecs tripped, or their inoperable protective devices racked out, or removed {se-
[
l l
l l
l
REFUELING OPERATIONS FUEL STORAGE BUILDING EMERGENCY AIR CLEANING SYSTEM SURVEILLANCE REQUIREMENTS 4.9.12b (Continued) 1)
Verifying that the cleanup system satisfies the in place penetration and bypass leakage testing acceptance criteria of less than 0.05% and uses the test procedure guidance in Regulatory Positiens C.S.a, C.S.c, and C.S.d of Regulatory Guide 1.52, Revision 2, March 1978,* and the system flow rate is lh000 cfm i 10%;
/k+EO 2)
Verifying, within 31 days after removal, that a laboratory analysis of a representative carbon sample obtained in accor,
dance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978,* meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978, by showing a methyl iodide penetration of less than 1.0% when tested at a temperature of 30*C and at a relative hu-midity of 95% in accordance with ASTM-D-3803; and
/l96o 3)
Verifying a system flow rate of #, 000 cfm i 12% during system operation when tested in accordance with ANSI N510-1980.
c.
After every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of charcoal adsorber operation by verifying, within 31 days after removal, that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978,*
meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978,* by showing a methyl iodide penetration of less than 1.0% when tested at a tem-perature of 30*C and at a relative humidity of 95% in accordance with ASTM-D-3803.
d.
At least once per 18 months by:
1)
Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than 6 inches Water Gauge while operating the system at a flow rate of
-17,00^rcfm i 10%,
/Q 4670 2)
Verifying that the system maintains the spent fuel storage pool area at a negative pressure of greater than or equal to 1/4 inch Water Gauge relative to the outside atmosphere during system operation,
- ANSI N510-1980 shall be used in place of ANSI N510-1975 as referenced in Regulatory Guide 1.52, Rev. 2, March 1978.
SEABROOK - UNIT 1 3/4 9-14
~
REFUELING OPERATIONS FUEL STORAGE BUILDING EMERGENCY AIR CLEANING SYSTEM SURVEILLANCE REQUIREMENTS 4.9.12d (Continued) 3)
Verifying that the filter cross connect valve can be manually opened, and 4)
Verifying tha, the heaters dissipate 95 i 11 kW when tested in accordance with ANSI N510-1980.
e.
After each complete or partial replacement of a HEPA filter bank, by verifying that the cleanup system satisfies the in place penetration leakage testing acceptance criteria of less than 0.05% in accordance with ANSI N510-1980 for a DOP test aerosol while operating the system 000 cfm i 10%.
at a flow rate of 17'efo M,.
f.
Af ter each complete or partial replacement of a charcoal adsorber bank, by verifying that the cleanup system satisfies the in place penetration leakage testing acceptance criteria of less than 0.05%
in accordance with ANSI N510-1980 for a halogenated hydrocarbon re-frigerant test gas while operating the system at a flow rate of 19 600 cfm i 10%.
7
/Q450 SEABROOK - UNIT 1 3/4 9-15
RADI0 ACTIVE EFFLUENTS h6 GASEOUS EFFLUENTS GAS STORAGE TANKS LIMITING CONDITION FOR OPERATION 3.11.2.6 The quantity of radioactivity contained in the hydrogen surge tank shall be limited to less than or equal to 198,000 Curies of noble gases (con-sidered as Xenon-133 equivalent).
APPLICABILITY:
At all times.
ACTION:
4 a.
With the quantity of radioactive material in the hydrogen surge tank exceeding the above limit, immediately suspend all additions of radioactive material to the tank, within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> reduce the tank
- contents to within the limit, and describe the events leading to this condition in the next Semiannual Radioactive Effluent Release Report pursuant to Specification 6.8.1.4.
b.
The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.11.2.6 The quantity of radioactive material contained in each gas storage tank shall be determined to be within the above limit at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when radioactive materials are being added to the tank.
I SEABROOK - UNIT 1 3/4 11-10
i RADI0 ACTIVE EFFLUENTS 3/4.11.3 SOLID RADIOACTIVE WASTES LIMITING CONDITION FOR OPERATION 3.11.3 Radioactive wastes shall be SOLIDIFIED or dewatered in accordance with the PROCESS CONTROL PROGRAM to meet shipping and transportation requirements during transit, and disposal site requirements when received at the disposal site.
APPLICABILITY: At all times.
ACTION:
a.
With SOLIDIFICATION or dewatering not meeting disposal site and shipping and transportation requirements, suspend shipment of the inadequately processed wastes and correct the PROCESS CONTROL PROGRAM, the procedures, and/or the Solid Waste System as necessary to prevent recurrence.
b.
With SOLIDIFICATION or dewatering not performed in accordance with the PROCESS CONTROL PROGRAM, test the improperly processed waste in each container to ensure that it meets burial ground and shipping requirements and take appropriate administrative action to prevent recurrence.
c.
The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.11.3 For cement SOLIDIFICATION of at least one representative test speci-men from at least every tenth batch of each type of wet radioactive wastes (e.g., filter sludges, spent resins, evaporator bottoms, boric acid solutions, and sodium sulfate solutions) shall be verified in accordance with the PROCESS CONTROL PROGRAM:
a.
If any test specimen fails to verify SOLIDIFICATION, the SOLIDIFICATION of the batch under test shall be suspended until such time as additional test specimens can be obtained, alternative SOLIDIFICATION parameters can be determined in accordance with the PROCESS CONTROL PROGRAM, 4
and a subsequent test verifies SOLIDIFICATION. SOLIDIFICATION of the batch may then be resumed using the alternative SOLIDIFICATION parameters determined by the PROCESS CONTROL PROGRAM; b.
If the initial test specimen from a batch of waste fails to verify SOLIDIFICATION, the PROCESS CONTROL PROGRAM shall provide for the collection and testing of representative test specimens from each consecutive batch of the same type of wet waste until at least three consecutive initial test specimens demonstrate SOLIDIFICATION.
The PROCESS CONTROL PROGRAM shall be modified as required, as provided in Specification 6.12, to assure SOLIDIFICATION of subsequent batches of waste; and
/D SEABROOK - UNIT 1 3/4 11-1f
~v.
~
RADIOACTIVE EFFLUENTS SOLID RADIOACTIVE WASTES SURVEILLANCE REQUIREMENTS 4.11.3 (Continued) c.
With the installed equipment incapable of meeting Specification 3.11.3 or declared inoperable, restore the equipment to OPERABLE status or provide for contract capability to process wastes as necessary to satisfy all applicable transportation and disposal requirements.
Il SEABROOK - UNIT 1 3/4 11-tr
RADIOACTIVE EFFLUENTS 3/4.11.4 TOTAL DOSE 1
LIMITING CONDITION FOR OPERATION 3.11.4 The annual (calendar year) dose or dose commitment to any MEMBER OF THE PUBLIC due to releases of radioactivity and to radiation from uranium fuel cycle sources shall be limited to less than or equal to 25 mrems to the whole body or any organ, except the thyroid, which shall be limited to less than or equal to 75 mrems.
APPLICABILITY:
At all times.
ACTION:
a.
With the calculated doses from the release of radioactive materials 4
in liquid or gaseous effluents exceeding twice the limits of Specif.i-cation 3.11.1.2a., 3.11.1.2b., 3.11.2.2a., 3.11.2.2b., 3.11.2.3a., br 3.11.2.3b., calculations shall be made including direct radiation contributions from the units and from outside storage tanks to deter-mine whether the above limits of Specification 3.11.4 have been exceeded.
If such is the case, prepare and submit to the Commission within 30 days, pursuant to Specification 6.8.2, a Special Report that defines the corrective action to be taken to r. educe subsequent releases to prevent recurrence of exceeding the above limits and includes the schedule for achieving conformance with the above limits.
This Special Report, as defined in 10 CFR 20.405(c), shall include an analysis that estimates the radiation exposure (dose) to a MEMBER OF THE PUBLIC from uranium fuel cycle sources, including all effluent pathways and direct radiation, for the calendar year that includes the release (s) covered by this report. It shall also describe levels of radiation and concentrations of radioactive material involved, and the cause of the exposure levels or concentrations.
If the estimated dose (s) exceeds the above limits, and if the release condition result-ing in violation of 40 CFR Part 190 has not already been corrected, the Special Report shall include a request for a variance in accor-dance with the provisions of 40 CFR Part 190.
Submittal of the report is considered a timely request, and a variance is granted until staff action on the request is complete.
b.
The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.11.4.1 Cumulative dose contributions from liquid and gaseous effluents shall be determined in accordance with Specifications 4.11.1.2, 4.11.2.2, and 4.11.2.3, and in accordance with the methodology and parameters in the ODCM.
4.11.4.2 Cumulative dose contributions from direct radiation from the units and from radwaste storage tanks shall be determined in accordance with the methodology and parameters in the ODCH.
This requirement is applicable only under conditions set forth in ACTION a. of Specification 3.11.4.
l 17' SEABROOK - UNIT 1 3/4 11-33' il
.--r
PLANT SYSTEMS BASES 3/4.7.1 TURBINE CYCLE (Continued) 3/4.7.1.2 AUXILIARY FEEDWATER SYSTEM I
The OPERABILITY of the Auxiliary Fe water System ensures that the Reactor Coolant System can be cooled down to le s than 350*F from normal operating conditions in the event of a total los -of-offsite power.
f 2/2. I The electric motor-driven emerge cy feedwater pump is pable of deliver-ing a total feedwater flow of 650 gpm at a pressure of psig to the en-trance of the steam generators. The steam-driven emergency feedwater pump is capable of delivering a total feedw er flow of 650 gpm at a pressure of 1221 psig to the entrance of the st am generators. The startup feedwater pump serves as the third auxiliary feedw ter pump and can be manually aligned to be powered from an emergency bus (Bus 5).
The startup feedwater pump is capable' of taking suction on the dedicated emergency feedwater volume of water in the condensate storage tank and deliv ing a total feedwater flow of in excess of 650 gpm at a pressure of 1221 psi to the entrance of the steam generator via t
either the main feedwater header or with manual alignment to the emergency feedwater flow path. This capacity is sufficient to ensure that adeauate feed-water flow is available to remove decay heat and reduce the Reactor Coolant System temperature to less than 350*F when the Residual Heat Removal System may be placed into operation.
3/4.7.1.3 CONDENSATE STORAGE TANK The OPERABILITY of the condensate storage tank with the minimum water vol-une ensures that sufficient water is available to cool the RCS to a temperature i
of 350*F. The OPERABILITY of the concrete enclosure ensures this availability of water following rupture of the condensate storage tank by a tornado generated missile. The contained water volume limit includes an allowance for water not usable because of tank discharge line location or other physical characteristics, 3/4.7.1.4 SPECIFIC ACTIVITY The limitations on Secondary Coolant System specific activity. ensure that the resultant offsite radiation dose will be limited to a small fraction of 10 CFR Part 100 dose guideline values in the event of a steam line rupture.
This dose also includes the effects of a coincident 1,gpm reactor-to-secondary-tube leak in the steam generator of the affected steam line. These values are consistent with the assumptions used in the safety analyses.
3/4.7.1.5 MAIN STEAM LINE ISOLATION VALVES The OPERASILITY of the main steam line isolation valves ensures that no more than one steam generator will blow down in the event of a steam line rupture. This restriction is required to:
(1) minimize the positive reac-tivity effects of the Reactor Coolant System cooldown associated with the blowdown, and (2) limit the pressure rise within containment in the event the steam line rupture occurs within containment.
The OPERABILITY of the main steam isolation valves within the closure times of the Surveillance Require-ments are consistent with the assumptions used in the safety analyses.
SEABROOK - UNIT 1 B 3/4 7-2
RADI0 ACTIVE EFFLUENTS BASES GASEOUS EFFLUENTS 45nv.s 84O*4fFF 3/4.11.2.5 EXPLOSIVE GAS MIXTURE FOR THE WAEfWiM:debMP SYSTEM This specification is provided to ensure that the concentration of poten-tially explosive gas mixtures contained in the GASEOUS RADWASTE SYSTEM is main-tained below the flammability limits of hydrogen and oxygen. Maintaining the concentration of hydrogen and oxygen below their flammability limits provides assurance that the releases of radioactive materials will be controlled in conformance with the requirements of General Design Criterion 60 of Appendix A to 10 CFR Part 50.
1 W.11.2.6 GAS STORAGE TANK The t cluded in this spec'fication is that or which the quan-Ick tity of radioact v tained is not limit actly or indirectly by anothe Technical Specification.
ctin uantity of radioactivity contained in the gas storage tank provid nce that in the event of an uncontrolle release of the tank's e s, the resu hole body exposure to a MEMBER OF THE PUBLIC at earest SITE BOUNDARY will n d 0.5 rem.
This is consistent Standard Review Plan Section 11.3, Branch 1 Position ETSB
" Postulated Radioactive Releases Due to a Waste Gas Syste r
ure," in NUREG-0800, July 1981.
3/4.11.3 SOLID RADI0 ACTIVE WASTES This specification implements the requirements of 10 CFR 50.36a and General Design Criterion 60 of Appendix A to 10 CFR Part 50.
The process parameters included in establishing the PROCESS CONTROL PROGRAM may include, but are not limited to, waste type, waste pH, waste / liquid / SOLIDIFICATION agent / catalyst ratios, waste oil content, waste principal chemical constituents, and mixing and curing times.
3/4.11.4 TOTAL DOSE This specification is provided to meet the dose limitations of 40 CFR Part 190 that have been incorporated into 10 CFR Part 20 by 46 FR 18525.
The specification requires the preparation and submittal of a Special Report when-ever the calculated doses due to releases of radioactivity and to radiation from uranium fuel cycle sources exceed 25 mrems to the whole body or any organ, except the thyroid, which shall be limited to less than or equal to 75 mrems.
For sites containing up to four reactors, it is highly unlikely that the resul-tant dose to a MEMBER OF THE PUBLIC will exceed the dose limits of 40 CFR Part 190 if the individual reactors remain within twice the dose design objec-tives of Appendix I, and if direct radiation doses from the units (including outside storage tanks, etc.) are kept small.
The Special Report will describe a course of action that should result in the limitation of the annual dose to a MEMBER OF THE PUBLIC to within the 40 CFR Part 190 limits.
For the purposes of the Special Report, it may be assumed that the dose commitment to the MEMBER OF SEABROOK - UNIT 1 B 3/4 11-5
cd TABLE 6.2-1 MINIMUM SHIFT CREW COMPOSITION (1)
POSITION
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NUMBER OF INDIVIDUALS REQUIRED TO FILL POSITION MODE 1, 2, 3, or 4 MODE 5 or 6.
SRO 1
None RO 2
1 A0 2(3) 1 STA 1
None Shift Superintendent with a Senior Operator license on Unit 1 SS Individual with a Senior Operator license on Unit 1 SRO RO Individual with an Operator license on Unit 1 Auxiliary Operator AO STA Shift Technical Advisor l
TABLE NOTATIONS (1) The shift crew composition may be one less than the minia'um requirements of Table 6.2-1 for a period of time not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore the shift crew composition to within the minimum requirements of Table b.2-1.
This provision does not permit any shift crew position to be unmanned upon shift change due to an oncoming shift crewman being late or absent.
(2) During any absence of the Shift Superintendent from the control room while the unit is in MODE 1, 2, 3, or 4, an individual with a valid Senior Operator license shall be designated to assume the control room command function. During any absence of the Shift Superintendent from the control room while the unit is in MODE 5 or 6, an individual with a valid Senior Operator license or Operator license shall be designated to assume the control room command function.
(3) The STA position shall be manned in MODES 1, 2, 3, and 4 unless the Shift Superintendent or the individual with a Senior Operator license meets the qualifications for the STA as required by the NRC.
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ADMINISTRATIVE CONTROLS RECORD RETENTION 6.9.3 (Continued) f.
Records of reactor tests and experiments; Records of training and qualification for current members of the g.
station staff; h.
Specifications; Records of inservice inspections performed pursuant to i.
Records of quality assurance activities required by the Operational Quality Assurance Manual; j.
Records of reviews performed for changes made to procedures or equipment or reviews of tests and experiments pursuant to 10 CFR 50 k.
Records of meetings of the SORC and the NSARC; 1.
Records of the service lives of all hydraulic and mechanical snubbers required by Specification 3.7.7 including the date at which the service life commences and associated installation and maintenance records; Records of secondary water sampling and water quality; and m.
Records of analyses required by the Radiological Environmental n.
Monitoring Program that would permit evaluation of the accuracy of the analysis at a later date. This should include procedures effective at specified times and QA records showing that these procedures were followed.
_6.10 RADIATION PROTECTION PROGRAM 6.10.1 with the requirements of 10 CFR Part 20 and shall be app i'
adhered to for all operations involving personnel radiation exposure.
_6.11 HIGH RADIATION AREA i
6.11.1 Pursuant to paragraph 20.203(c)(5) of 10 CFR Part 20, in lieu of the
" control device" or "alara, signal" required by paragraph 20.203(c), each high radiation area, as defined in 10 CFR Part 20, in which the intensity of radia-tion is equal to or less than 1000 mR/h at 45 cm source or from any surface that the radiation pene(18 in.) from the radiation i
trates shall be barricaded and conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of a Radiation Work Permit (RWP).
qualified in radiation protection procedures (e.g., Health Physics Technician)
Individuals or personnel continuously escortec by such individuals may be exempt from the RWP g
tion areas with exposure rates equal to or less than 1000 t
j otherwise following plant radiation protection procedures for entry into such high SEABROOK - UNIT 1 6-20
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