ML19323C564
| ML19323C564 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 05/14/1980 |
| From: | DUQUESNE LIGHT CO. |
| To: | |
| Shared Package | |
| ML19323C563 | List: |
| References | |
| TAC-10007, TAC-11055, NUDOCS 8005160334 | |
| Download: ML19323C564 (180) | |
Text
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80051 SON [
l DUQUESNE LIGHT COMPANY :
i Beaver Valley. Power Station, Unit No. 1 l
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Docket No. 50-334 License No. DPR-66
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Attachment A l f
Radiological Effluent Technical Specifications j
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I Proposed Revi~sions to ETS Submitted Previously on April 11, 1979 1 i
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INDEX DEFINITIONS T
SECTION Page 1.0 DEFINITIONS 1-1 Defined Terms...................................'............. 1-1 Thermal Power................................................
1-1 Rated Thermal Power..........................................
1-1 Operational Mode...................'.......................... 1-1 Action....................................................... 1-1 Operable - Operability.......................................
1-2 Reportable Occurrence........................................
1-2 Containment Integrity........................................
1-2 Channel Cal i b ra ti o n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
1-2 Channel Check................................................
1-3 Channel Functional Test......................................
1-3 C o r e Al t e r a ti o n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
1-3 Shut'down Margin..............................................
1-3 Identified Leakage...........................................
1-3 Unidentified Leakage.........................................
1-4 Pressure Boundary Leakage....................................
1-4 Controlled Leakage...........................................
. 1-4 Quadrant Power Til t Ratio . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
1-4 Dose Equivalent I-131...............................s........ '
1-4 Staggered Test Basis.........................................
1-4 Frequency Notation...........................................
1-5 Reacto r Trip Res pons e Time . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
1-5 Engineered Safety Fea ture Re'sponse Time. . . . . . . . . . . . . . . . . . . . . .
1-5 Axial Flux Difference........................................
1-5 Physics Test.................................................
1-5 E-Average Disentegration Energy.............................. . .
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1-6 Source Check.................................................
1-6 Process Control Program.......................................
1-6 So l i d i f i c a ti o n . . . . . - . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
1-6 Off-Si te Oose Cal cul ation Manual (00CM) . . . . . . . . . . . . . . . . . . . . . . .
1-6 Gaseous Radwaste Treatment System.............................
System..........................
1-6 Ventilation Exhaust Treatment i-7
.. .... ..... ......... ...................
O Purse /evr9 4 as....
Venting.......................................................
1-7 1-8 Operatioral Modes (Table 1.1).................................
I-8 Frequency Notation............................................
SEAVER VALLEY - UNIT 1 I
INDEX L IMITING C0!:0! TIC::S FCR OPEPAT!O!! Al:0 SUR'.'EILLAf!CE RECUIREM Page_
SECTI0t.{
3/4.2 POWER DISTRIBUTION LIMITS 3/? 2-1 3/4.2.1 Axi al Fl ux Di f fer e n c e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
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. 3/4 2-5 3/4.2.2 Hea t Fl ux Ho t Channel Factor. . . . . . . . . . . . . . . . . . . . . . . . . . .
3/4 2.-8 4
Nucl ear Enthal py Ho t Channel Factor. . . . . . . . . . . . . . . . . . . .
! 3/4.2.3 ~
3/4 2-10 3/4.2.4 Quadra nt Pcwer Til t Ra tio . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
3/4 2-12 3/4.2.5 DNB Parameters.........................................
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3/4.3 INSTRUMENTATIOf!
! 3/4 3-1 3/4.3.1 PROTECTIV E INSTRUMEtiT' TI0!l . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
3/4 3-14
) 3/4.3.2 E!!GI!!EERED SAFETY FEATURE INSTRUMEt:TATIO:1. . . . . . . . . . . . . .
f 3/4.3.3 MONITORIftG INSTRUMENTATION Rad i a ti o n Mo ni tori ng . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
3/4 3-33 3/4 3-37 Mova bl e Inc o r e De t ecto r s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
Q S ei smi c Ins trumenta ti on . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
3/4 3-38 Instrumentation......................... 3/4 3-41
' Heteorologicai 3/4 3 44 Remo te Shutdown Ins trumentation. . . . . . . . . . . . . . . . . . . . . . . '
3/4 3-47 Fire .De tecti o n Instrumenta tion. . . . . . . . . . . . . . . . . . . . . . . . .
Radioactive Liquid Effluent Instrumentation................ 3/4 3-49 i
Radioactive liquid Effluent Instrumentation................ 3/4 3-50 3/4 3-55 Radioactive Gaseous Effluent Moni tors . . . . . . . . . . . . . . . . . . . . . .
3/4.4 REACTOR COOLANT SYSTEM
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3/4.4.1 REACTOR COOLANT LOOPS 3/4 4-1 Normal Operation........................................
3/4 4-3 Isolat2d Leop.......................................... ,
Isolated Lcop Startup.................................. 3/4 4-4 .
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BEAVER VALLEY - UNIT 1 IV
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INDEX
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LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE
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3/4.9.10 WATER LEVEL-REACTOR VESSEL............................. 3/4 9-10 3/4.9.11 STORAGE POOL WATER LEVEL............................... 3/4 9-11 3/4.9.12 FUEL BUILDING VENTILATION SYSTEM - FUEL MOVEMENT....... 3/4 9-12 3/4.9.13 FUEL BUILDING VENTILATION SYSTEM - FUEL STORAGE. . . . . . . . 3/4 9-13
- 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOWN MARGIN........................................ 3/4 10-1 3/4.10.2 GROUP HEIGHT AND INSERTION LIMITS...................... 3/4 10-2 3/4.10.3 PRESSURE / TEMPERATURE LIMITATIONS-REACTOR CRITICALITY... 3/4 10-4
- 3/4.10.4 PHYSICS TEST........................................... 3/4 10-6 3/4.10.5 NO FLOW TESTS.......................................... 3/4 10-7 1
3/4.11 RADIOACTIVE EFFLUENTS-3/4.11.1 LIQUID EFFLUENTS .
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CONCENTRATION.......................................... 3/4 11-1
(]) 00SE................................................... 3/4 11-5 LIQUID WASTE TREATMENT................................. 3/4 11-6
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LIQUID HOLDUP TANKS.................................... 3/4 11-7 3/4 11.2 GASEOUS EFFLUENTS -
DOSE RATE.............................................. 3/4 11-8 DOSE - NOBLE GASES..................................... 3/4 11-11 DOSE - RADI0 IODINES, PARTICULATE AND RADIONUCLIDES OTHER THAN NOBLE GASES................................. 3/4 11-12 GASEOUS WASTE TREATMENT................................ 3/4 11-13 GAS STORAGE TANKS...................................... 3/4 11-15 3/4.11.3 SOLID RADI0 ACTIVE WASTE................................ 3/4 11-16 3/4.11.4 TOTAL 00SE............................................. 3/4 11-17 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.1 MONITOR!NG PR0 GRAM..................................... 3/4 12-1 3/4.12.2 LAND USE CENSUS........................................ 3/4 12-9 3/4.12.3 INTERLABORATORY COMPARISON PR0 GRAM..................... 3/4 12-10 O BEAVER VALLEY ,- UNIT 1 IX
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> 1.0 DEFINITIONS (Continued)
PURGE-PURGIrlG 1.33 PURGE or PURGIttG is the controlled process of discharging air or gas fro- a confinement to maintain temperature, pressure, hu.nidity, concentraticn or other operating conditions, in such a manner that replacement air or gas is required to purify the confinement.
VENTING 1.34 VEflTING is the controlled process of discharging air or gas fecm a confinement to maintain temperature, pressure, humidity, concentration or other operating conditions, in such a manner that replacement air or gas is not provided or required during VENTIflG. Vent, used in system names,
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does not imply a VENTING process.
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4 OPERATIONAL MODES
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REACTIVITY % RATED AVERAGE COOLANT CONDITION, Koff THERMAL POWER
- TEMPERATURE ,
MODE
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> 0.99 > 5% > 350*F
- 1. POWER OPERATIM
> 0.99 < 5% > 350*F
- 2. STARTUP
< 0.99 0 > 350*F
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- 3. HOT STANOBY
< 0.99 0 350*F > Tavg
- 4. HOT SHUTDC'nN
> 200*F
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< 0.'99 0 < 200*F
- 5. COLD SHUTDOWN _
< 0.95 0 < 140*F
- 6. REFUELING ** _
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- Exclucing cecay heat.
- Reactor vessel head unbolted or removed and fuel in the vessel.
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TABLE 1.2
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FREQUENCY NOTATION i
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1 FREQUENCY .
i NOTATION
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At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
5 At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
D '
At least once per 7 days.
W At least once per 31 days M
At least once per 92 days.
Q
- At least once per 6 months.
SA l .
1 At least once per 18 months.
f R
! Prior to each reactor startup.
' S/U Completed prior to each release.
P i
N.A.
Not applicable.
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BEAVER VALLEY - UNIT 1 i
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O t"stausentatto" RADIOACTIVE LIOUID EFFLUENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.8 The radioactive liquid effluent monitoring instrumentation channels
- shown in Table '.3-11 shall be OPERABLE with their alarm / trip setpoints set to ensure that t.ae limits of Specification 3.11.1.1 are not exceeded. The alarm / trip setpoints of these channels shall be determined in accordance with the Offsite Oose Calculation Manual (00CM).
APPLICABILITY: At all times.
ACTION:
- a. With a radioactive liquid effluent monitoring instrumentation channel alarm / trip setpoint less conservative than required by the above specification, immediately suspend the release of radioactive liquid effluents monitored by the affected channel or declare the channel inoperable.
- b. Wit . one or more radioactive liquid effluent monitoring instrumentation channels inoperable, take the ACTION shown
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in Table 3.3-11.
- c. The provisione of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REOUtREMENTS 4.3.3.8 Each radioactive liquid effluent monitoring instrumentation channel shall be demongtrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations at the frequencies she,,n in Table 4.3-11.
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TABLE 3.3 71
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RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION MINIMUM E CilANNELS E INSTRUMENT OPERABLE ACTION 9
-: 1. Gross Activity Monitors Providing Automatic h Tennination of Release
- a. Liquid Waste Effluents Monitor (RM-LW-104) (1) 23 23 s b. Liquid Waste Contaminated Drain Monitor (1)
% (RM-LW-ll6)
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- 2. Gross Activity Monitors Not Providing Termination of Release
- a. Component Cooling-Recirculation Spray (1) 24
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m lleat Exchangers River Water Monitor
) (RM-RW-100)
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- 3. Flow Rate Measurement Devices
- a. Liquid Radwaste Effluent Line (1) 25 (1; FR-LW-103 (2) FR-LW-104
- b. Cooling Tower Blowdown Line (1) 25 (1) FT-CW-101 (2) FT-CW-101-1,
- 4. Tank Level Indicating Devices (For tanks outside plant building)
- a. Primary Water Storage Tank (BR-TK-6A) (1) 26
- b. Primary Water Storage Tank (BR-TK-68) (1) 26
- c. Steam Generator Drain Tank (LW-TK-7A) (1) 26
- d. Steam Generator Drain Tank (LW-TK-7B) (1) 26
- e. Refueling Water Storage Tank (QS-TK-1) (1) 26
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_ TABLE 3.3-11 (Continued) )
TABLE NOTATION ACTION 23 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases may ba resumed for up to 14 days provided that prior to initiating a release:
- 1. At least two independent samples are analyzed in accordance with Specification 4.11.1.1.3, and;
- 2. At least two technically qualified members of the Facility Staff independently verify the release rate calculations and discharge valving; Otherwise, suspend release of radioactive effluents via this pathway.
ACTION 24 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days provided that at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> grab samples are at a analyzed for gross radioactivity (beta or gamm9)uCi/ml.
Lower Limit of Detection (LLD) of at least 10-
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O ACT10N as - With the numeer of chenneis OeEaABtE iess then requ4 red by the Minimum Channels OPERABLE requirement, effluent
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releases via this pathway may continue for up to 30 days provided the flow rate is estimated at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> during actual releases. Pump curve may be used to estimate flow. -
ACTION 26 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, liquid additions to this tank may continue for up to 30 days provided the tank liquid level is estimated during all liquid additions to the tank.
[D BEAVER VALLEY - UNIT 1 3/4 3-52
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O O O TABLE 4.3-11 03 9 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS i
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- u CllANNEL si SOURCE CllANNEL FUNCTIONAL CilANNEL E~ CALIBRATION TEST CIIECK _ CllECK "J _ INSTRUMENT h 1. Gross Beta or Ganana Radioactivity. Monitors Pro-E viding Alarm and Automatic Tennination of Release R(3) Q(1)
$ a. Liquid Radwaste Effluent Line D P(5)
(RM-LW-104)
- b. Liquid Waste Contaminated Drain Line D P(5) R(3) Q(1)
(IU4-LW-ll 6)
- 2. Gross Beta or Ganana Radioactivity Monitors Providing Alann but not providing Automatic y Tennination of Release Q(2)
- a. Component Cooling-Recirculation Spray D M R(3) w di lleat Exchangers River Water Monitor
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(RM-RW-100)
- 3. Flow Rate Monitors 0(4) NA R Q
- d. Liquid RadWaste Effluent Lines (1) FR-LW-103 (2) FR-LW-104 NA R Q
- b. Cooling Towar Blowdown Line (FT-CW-101) O(4)
- 4. Tank Level Monitors for tlie following Tanks D* NA 11 Q
- a. Primary Water Storage Tank - (PG-TK-6A)
D* NA R Q
- b. Primary Water Storage Tank - (PG-TK-6B)
D* NA R R
- c. Steam Generator Drain Tank - (LW-TK-7A)
D* NA R Q
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- d. Steam Generator Drain Tank - (LW-TK-7B) Q D* NA R
- e. Refueling Water Storage Tank (QS-TK-1)
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TABLE A.3-11 (Continued)
TABt.E NOTATION O
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Duri.ng liquid additions to the. tan.k.
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(1) - The CHANNEL FUNCTIONAL TEST shall also demonstrate that automatic isolation of this pathway and Control Room Alarm Annunciation occurs if any of the following conditions exist:
- 1. Instrument indicates measured levels above the alarm / trip
.setpoint.
- 2. Circuit failure.
- 3. Instrument controls not set in operate mode.
The CHANNEL FUNCTIONAL TEST shall also demonstrate that control
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(2) -
room alarm annunciation occurs if any of the following conditions exist:
- 1. Instrument indicates measured levels above the alann/ trip setpoint.
- 2. Circuit failure.
- 3. Instrument controls are no.t set in cperate mode.
(3) - The initial CHANNEL CALIBPJ. TION for radioactivity measurement instrumentatior shall be performed using one or more of the reference standards certified by the National Bureau of Standards or using standards that have been obtained from suppliers that
]L participate in measurement assurance activities with NBS. These standards should permit calibrating the system over its intended range of energy and rate capabilities. For subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration should be used, at intervals of at least once per eighteen montns.
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This can normally be accomplished during refueling outages.
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( 4) - CHANNEL CHECK shall consist of verifying indication of flow during periods of release. CHANNEL CHECX shall be made at least once daily on any day on which continuous, peri adic, or batch releases are made.
(5) - A channel check may verify system operation. In these instances, a source check may be waived, since high backgrounds may mask check
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source strength.
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1 h) BEAVER VALLEY - UNIT I 3/4 3-54
} INSTRUMENTATION RADIOACTIVE GASE0US EFFLUENT MONITORING INSTRUMENTATION .
LIMITING CONDITION FOR OPERATION 3.3.3.9 The radioactive gaseous effluent monitoring instrumentation channels shown in Table 3.3-12 shall be OPERABLE with their alarm / trip setpoints set to ensure that the limits of Specification 3.11.2.1 are not exceeded. The alarm / trip setpoints of these channels shall be determined in accordance with i the 00CM.
APPLICABILITY: As shown in Table 3.3-12.
ACTION:
- a. With a radioactive gaseous effluent monitoring instrumentation channel alarm / trip setpoint less conservative than required by
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the above Specification, declare the cnannel inoperable.
- b. With one or more radioactive gaseous effluent monitoring instrumentation channels inoperable, take the ACTION shown
- in Table 3.3-12.
() c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
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SURVEILLANCE RE0VIREMENTS i
4.3.3.9 Each radioactive gaseous effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations at the frequencies shown in Table 4.3-12.
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BEAVER VALLEY - UNIT 1 3/4 3-55' l i
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O o O TABLE 3.3-12 y
RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION
,
y
$ MINIMUM l~
CllANNELS
'2
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OPERABLE APPLICABILITY ACTION
. INSTRUMENT b 1. Gaseous Waste Particulate and Gas Monitors (RM-GW-108A & B)
]
- 27
- a. Noble Gcs Activity Monitor (1)
- 32
- b. Iodine Samf er Cartridge (1)
- 32
- c. Particulate ' ampler Filter (1)
- 28 u d. Effluent Flow Rate Measurii.g (1) 2 Device (FR-GW-108)
Sampler Flow Itate Measuring Device Il}
- e.
- 2. Auxiliary Building Ventilation System (ItM-VS-101A & B) 29
- a. Noble Gas Activity Monitor (1) ,
- 32
- b. Iodine Sampler Cartridge (1)
- 32
- c. Particulate Sampler Filter (1)
- 28
- d. Flow Rate Monitor (FR-VS-101) (1)
- 28
- e. Samp,er flow Rate Monitor (1)
- During Releases via this pathway
-_ _ _ - - - _ _ _ _ _ _ - _ _ - .
_ - _-
- __ - .- _ . _ . _ . - - - . . -- -
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m TABIE 3.3-12 (Continuedl m
?! RADI0 ACTIVE GASE0US EFFLUENI' MONITORING INSTRUMENTATIO!!
$
-:
MINIMUM
??
l; CilANNELS APPLICABILITY ACTION OPERABLE
{ INST _RUMENT
$ .3. Elevated Release (Located dn top of G Reactor Containment) (RM-VS-107A & B)
- 30
~
- a. Noble Gas Activity Monitor (1)
- 32
- b. lodine Sampler Cartridge (1)
- 32
- c. Particulate Sampler Filter (1)
- 28
- d. Flow Rate Monitor (FR-VS-112) (1)
R
- 28
- e. Sampler Flow Rate Monitor (1) t,. '
" Waste Gas Decay Tanks Monitor 4.
- a. Oxygen Monitor (0 2-AS-GW-110-1,2) ( j)' , ** 31
- During Releases via this pathway.
- Dur ing waste gas decay tank filling operation.
_____________-__ _ _______ - - - - _ _ _ _ _ _ - _ _ _ _ _ _ _ - - _ _ _
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TABLE 3.3-12 (Continued)
O TABLE tiOTATI0t1 ACTION 27 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, the contents of the tank may be released to the environment for up to 14 days provided that prior to initiating the release:
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- 1. At least two independent samples of the tank's content are analyzed, and
- 2. At least two technically qualified members of the Facility Staff independently verify the release rate calculations and discharge valve lineup.
Otherwise, suspend release of radioactive effluents via this pathway.
ACTION 28 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 28 days provided the flow rate is estimated at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
ACTION 29 - With the number of channels OPERABLE less than required by p the Minimum Channels OPERABLE requirement, effluent releases V via this pathway may continue for up to 28 days provided grab samples are taken at least once per 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />s-and these samples are analyzed for gross activity within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
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ACTION 30 - With the number of channels OPERABLE less than required by Minimum Channels OPERABLE requirement, immediately suspend PURGING of Reactor Containment via this pathway.
ACTION 31 - With the number of channels OPERABLE one less than required by the MINIMUM Channels OPERABLE requirement, operation of this system may continue provided grab samples are obtained every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
ACTION 32 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days, provided samples are continuously collected with auxiliary sampling equipment as required in Table 4.11-2.
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1 BEAVER VALLEY - UNIT 1 3/4 3-58 i
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O O O TABLE 4.3-12 h
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! ItADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CilANNEL h CllANNEL SOURCE CilANNEL FUNCTIONAL
,q CALIBRATION TEST
< INSTRUMENT CllECK CllECK e
E 1. Gaseous Waste Particulate and Gas Monitors 5 (RM-GW-108A & B)
Q(1)
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- a. . Noble Gas Activity Monitor P* P(5) R(3)
N/A N/A N/A
- b. lodine Sampler Cartridge W N/A N/A
- c. Particulate Sampler Filter W N/A P* N/A R Q
- d. System Effluent Flow Rate Meastaring Device D* N/A R Q
- e. Sampler Flow Rate Measuring Device
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m 2. Auxiliary Bldg. Vent. System +
- (RM-VS-101A & B)
Noble Gas Activity Monitor D M R(3) Q(2) y a.
N/A N/A N/A
$ o. lodine Sampler W Particulate Sampler W N/A N/A N/A c.
N/A R N/A
- d. System Effluent Flow Rate D Measurement Device D N/A R. Q
- e. Sampler Flow Rate Meastarement Device
_ _ _ _ _ - _ _ _ - _ _ _ _ _ _ _ _ _ - _ _
O O O ca TABLE 4.3-12 (Continued) 9 RADI0 ACTIVE GASE0US EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REmilREME g
w ClANNEL h CllANNEL SOURCE CilANNEL FUNCTIONAL l~; CALIBRATION TEST
- INSTRUMENT CllECK CllECK
..
C 35 3. Elevated Release (Top of Reactor
" Containment) Monitoring System
~
(RM-VS-107A & B)
Noble Gas Activity Monitor D* M R(3) Q(1) a.
D* N/A N/A N/A
- b. Iodine Sampler D* N/A N/A N/A
- c. Particulate Sampler D* N/A R N/A
- d. System Effluent Flow Rate Measuring Device
{ R Q
- e. Sampler Flow Rate Measuring D* N/A
'f g Device
- 4. Waste Gas Decay Tanks Monitor D N/A Q(4) M
- a. Oxygen Monitor
__ _ _ _ _ _ . _ _ _ __ -
_ _ _ _ _
TABLE 4.3-12 (Continued)
O
's> TABLE NOTATION
- Ouring releases via this pathway.
- During waste gas holdup system operation (treatment for primary system offgases).
(1) The CHANNEL FUNCTIONAL TEST shall also demonstrate that automatic isolation of this pathway and control room alarm annunciation occurs if any of the following conditions exist:
- 1. Instrument indicates measured levels above the alarm / trip setpoint.
- 2. Circuit failure.
- 3. Instrument controls not set in operate mode.
(2) The CHANNEL FUNCTIONAL TEST shall also demonstrate that control rcom alarm annunciation occurs if any of the following conditions exist:
t"stru=e"t 4"dicates me*s" red ieveis eeove the eier=/tria O i-setpoint.
- 2. Circuit failure.
- 3. Instrument controls not set in operate mode.
(3) The initial CHANNEL CALIBRATION for radioactivity measurement instrumentation shall be performed using one or more of the referenr.e standards certified by the National Bureau of Standards or using standards that have been obtained from suppliers that participate in measurement assurance activities with NBS. These standards should permit calibrating the system over its intended range of energy and rate capabilities. For subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration should be used, at intervals of at least once per eighteen months. This can normally be accomplished during refueling outages. (Existing plants may substitute previously established calibration procedures for this requirement.)
%
BEAVER VALLEY - UNIT 1 3/4 3- 61
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TABLE 4.3-12
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4 n
V (Continued) 4 i
(4) The CHANNEL CALIBRATION shall include the use of standard gas samples
containing a nominal:
- 1. One volume percent oxygen, balance nitrogen; and i 2. Four volume percent oxygen, balance nitrogen.
A channel check may verify system operation. In these instances, a (5) source check may be waived,.since background may mask check source
,
strength.
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BEAVER VALLEY - UNIT 1 3/4 3-62 l
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3/4.11 RADI0 ACTIVE EFFLUENTS 3/4.11.1 LIOUID EFFLUENTS CONCENTRATION LIMITING CONDITION FOR OPERATION 3.11.1.1 The concentration of radioactive material released from the site (see Figure 5.1-2) shall be limited to the concentrations scecified in 10 CFR Part 20, Appendix 3, Table II, Column 2 for radionuclides otner than dissolved or entrained noble gases. For diss the concentration shall be limited to 2 x 10 aci/ml gived total or entrained activity.noble gases, APPLICABILITY: At all times.
ACTION:
With the concentration of radioactive material released from the site exceeding the above limits, immediately restore the concentration to within the above limits.
SURVEILLANCE REQUIREMENTS O 4.11.1.1.1 The radioactivity content of each batch of radioactive liquid waste shall be determined prior to release by sampling and analysis in accordance with Table 4.11-1. The results of pre-release analyses shall be used with the calculational methods in. the ODCM to assure that the concentration at the point of release is maintained within the limits of Specification 3.11.1.1.
4.11.1.1.2 Post-release analyses of samples composited frcm batch releases shall be performed in accordance with Table 4.11-1. The results of the previous post-release analyses shall be used with the calculational methods in the 00CM to assure that the concentrations at the point of release were i maintained within the limits of Specification 3.11.1.1.
t 4.11.1.1.3 The radioactivity concentration of liquids discharged from continuous release points shall be determined by collection and analysis of samples in accordance with Table 4.11-1. The results of the analyses shall be used with the calculational methods in the 00CM to assure that the concentrations at the point of release are maintained within the limits of Specification 3.11.1.1.
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f BEAVER VALLEY - UNIT I 3/4 11-1 !
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TABLE 4.11-1 O RADIOACTIVE LICUID WASTE SAMPLING AND ANALYSIS PROGRAM Minimum Lower Limit
Sampling Analysis Type of Activity of Detection Liquid Release Frequency Frequency Analysis (LLD)
Type (uCi/ml)a P P A. Batch Waste d Each Batch Each Batch Princical Gama 5 x 10-7
! Release Tanks Emitters
-
I-131 1 x 10-p
-5 4
- One Batch /M M Dissolved and 1 x 10 Entrained Gases (Gama Emiters)
.
P Each Batch M H-3 1 x 10-5 3
1 Composite -7 Gross Alpha 1 x.10 P-32 1 x 10N O
'
P Each Batch Q Sr-89, Sr-90 5 x 10 -B b
-
Composite Fe-55 1 x 10-6 B. Grab Sample 9 W Principal Gama 5 x 10-7 Continuagg c Releases Composite Emittersf I-131 1 x 10 -6 Grab Sample 9 M Dissolved and 1 x 10-5 Entrained 3ases (Gama Emitters)
-5 Grab Sample 9 M c
H 1 x 10 Composite Gross Alpha 1 x 10-7
-6 P-32 1 x 10
-0 Grab Sample 9 Q Sr-89, Sr-90 5 x 10 c
Composite .
,
Fe-55 1 x 10~
d i
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BEAVER VALLEY - UNIT 1 3/4 11-2
. -_-
.-.
TABLE 4.11-1 (Continued)
TABLE NOTATION C:
- a. The Lower Limit of Detection (LLD) is defined in terms of the back-ground counting rate as follows:
LLD = 2/2 S b
2.22 x E x V Where S is the standard deviation of the background counting rate b
2.22 is the number of pCi per disintegration / min.
E is the counting efficiency in counts per disintegration V is the sample size analyzed in appropriate units
,
It should be recognized that the LLD is defined as an a, priori (before the fact) limit representing the capability of a measurement system and not as a posteriori (after the fact) limit for a particular measurement.
Analyses shall be performed in such a manner that the stated LLDs LLDs will be achieved under routine conditions. Occasionally back-ground fluctuations, unavoidably small sample sizes, the presence of interferring nuclides, or other uncontrollable circumstances may render these LLDs unachievable. In such cases, the contributing
' (~)
k- factors will be identified and described in the semi-annual Radio-active Effluent Release Report. ,
- b. A composite sample is one in which the quantity of liquid sampled is proportional to the quantity of liquid waste discharged and in which the method of sampling employed results in a speciment which is representative of the liquids released.
- c. To be representative of the quantities and concentrations of radio-active materials in liquid effluents, samples shall be collected continuously in proportion to the rate of flow of the effluent stream.
Prior to analyses, all samples taken for the composite shall be thoroughly mixed in order for the composite sample to be representa-tive of the effluent release.
- d. A batch release is the discharge of liquid wastes of a discrete
,
volume. Prior to sampling for analyses, each batch shall be isolated, and tnen thoroughly mixed by a method described in the CDCM, to assure representative sampling,
- e. A continuous release is the discharge of liquid wastes of a non-discrete volume; e.g., from a volume of system that has an input flow during the continuous release. This is applicable to the Turbine Building drains only.
(J9 BEAVER VALLEY - UNIT 1 3/4 11-3
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O TABLE 4.11-1 (Contineed>-
TABLE NOTATION
- f. The principal gamma emitters for which the LLD specification will apply are exclusively the following radionuclides: Mn-54, Fe-59, Co-58, Co-60, In-65, Mo-99, Cs-134, Cs-137, Ce-141, and Ce-144 This list does not mean that only these nuclides are to be detected and reported. Other peaks which are measurable and identifiable, together with the above nuclides, shall also be identified and reported. Nuclides which are below the LLD for the analyses should be reported as "less than" the nuclide's LLO, and should not be reported as being present at the LLO level for that nuclide. The "less than" values should not be used in the required dose calculations.
- g. Whenever there is primary to secondary leakage, sampling is done for
,
turbine building drain effluents by means of grab sampling taken every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> during the period of discharge and analyped for gross radioactivity (beta and gama) at a sensitivity of 10- uCi/mi and recorded in the plant records, along with the flow rate. Primary to secondary leakage is considered to be occurring whenever measurements indicate that secondary coolant gross activity (beta and gamma) is greater than 10-3 aci/ml. In addition, two plant personnel snall check release calculations to verify that the limits of 3.11.1.1 and 3.11.1.2 are not exceeded.
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BEAVER VALLEY - UNIT 1 3/4 11-4
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% RADI0 ACTIVE EFFLUENTS (G
COSE S
LIMITING CONDITION FOR OPERATION 3.11.1.2 The dose or dose commitment to an individual from radioactive materials in liquid effluents released from the site (see Figure 5.1-4) shall be limited:
- a. During any calendar quarter to less than or equal to 1.5 mrem to the total body and to less than or equal to 5 mrem to any organ, and
- b. During any calendar year to less than or equal to 3 mrem to the total body and to less than or equal to 10 mrem to any organ.
APPLICABILITY: At all times.
ACTION:
- a. With the calculated dose from the release of radioactive materials in liquid effluents exceeding any of the above limits, in lieu of any other report required by Specification 6.9.1, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report which identifies the cause(s)
(~)'
'
for exceeding the limit (s) and defines the corrective actions to be taken to reduce the releases of radioactive materials in liquid effluents during the remainder of the current calendar quarter and during the subsequent three calendar quarters, so that the cumulative dose or dose commitment to an individual from such releases during these four calendar quarters is within 3 mrem to the total body and 10 mrem to any organ. (This Special Report shall also include (1) the results of radiological analyses of the drinking water source and (2) the radiological impact on finished drinking water supplies with regard to the requirements of 40 CFR 141,
. Safe Drinking Water Act).*
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.11.1.2 Dose Calculations. Cumulative dose contributions frem liquid effluents snall be cetermined in accordance with the 00CM at least once per 31 days.
- Applicaole only if drinking water supply is taken from the receiving water body.
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() RADI0 ACTIVE EFFLUENTS LIQUID WASTE TREATMENT
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LIMITING CONDITION FOR OPERATION 3.11.1.3 The liquid radwaste treatment system shall be OPERABLE. The appropriate portions of the system shall be used to reduce the radioactive materials in liquid wastes prior to their discharge when the projected doses due to the liquid effluent from the site (see Figure 5.1-4) when averaged over 31 days, would exceed 0.06 mrem to the total body or 0.2 mrem to any organ.
APPLICABILITY: At all times.
ACTION:
- a. With the liquid radwaste treatment system inoperable for more than 31 days or with radioactive liquid waste being discharged without treatment.and in excess of the above limits, in lieu of any other report required by Specification 6.9.1, prepare and submit to the Commission within 30 days pursuant to Specification 6.9.2 a Special Report which includes the following information:
- 1. Identification of the inoperable equipment or subsystems and
({} the reason for inoperability,
- 2. Action (s) taken to restore the inoperable equipment to OPERABLE status, and
- 3. Summary description of action (s) taken to prevent a recurrence.
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE RECUIREMENTS 4.11.1.3.1 Doses due to liquid releases shall be projected at least once per 31 days, in accordance with ODCM.
4.11.1.3.2 The liquid radwaste treatment system shall be demonstrated OPERABLE by operating the liquid radwaste treatment system equipment for at least 60 minutes at least once per 92 days unless the liquid radwaste system has been utilized to process radioactive liquid effluents during the previous 92 days.
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RADI0 ACTIVE EFFLUENTS
{ LIOUID HOLDUP TANKS LIMITING CCNDITION FOR OPERATION 3.11.1.4 The quantity of radioactive material contained in each of the following tanks shall be limited to 1 10 curies, excluding tritium and dissolved or entrained noble' gases.
- a. BR-TK-6 A (Primary Water Storage Tank)
- b. BR-TK-6B (Primary Water Storage Tank)
- c. LW-TK-7A (Steam Generator Drain Tank)
- d. LW-TK-7B (Steam.. Generator Drain Tank)
- e. QS-TK-1 (Refueling Water Storage Tank)
APPLICABILITY: At all times. ACTION:
- a. With the quantity of radioactive material in any of the above
! listed tanks exceeding the above limit, immediately suspend all additions of radioactive material to the tank and within
.,
48 hours . reduce the tank contents. to within t.he limit.
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. -
([) . SURVEILLANCE REOUIREMENTS
. '
4.11.1.4 The quantity of radioactive material contained in each of the above listed tanks shall be determined to be within the above limit by analyzing a representative sample of the tank's contents at least once per 7 days wnen radioactive materials are being added to the tank. _. .
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(~% (-) BEAVER VALLEY - UNIT 1 3/4 11-7
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O aA0IO4CT m EFnUENTS 3/4.11.2 GASEOUS EFFLUENTS DOSE RATE LIMITING CONDITION FOR OPERATION 3.11.2.1 The dose rate due to radioactive materials released in gaseous effluents from the site (see Figure 5.1-3) shall be limited to the following:
- a. For noble gases: Less than or equal to 500 mrem /yr. to the total
body and less than or equal to 3000 mrem /yr. to the skin, and
- b. For all radiciodines and for all radioactive materials in I. particulate form and radionuclides (other than noble gases) l with half lives greater than 8 days: Less than or equal to
- 1500 mrem /yr. to any organ.
APPLICABILITY: At all times. ACTION: With the dose rate (s) exceeding the above limits, immediately decrease the release rate to within the above limit (s). ' SURVEILLANCE RECUIREMENTS 4.11.2.1.1 The dose rate due to noble gases in gaseous effluents shall , be determined to be within the above limits in accordance with the methods " and procedures of the 00CM. 4.11.2.1.2 The dose rate due to radioactive materials, other than noble gases, in gaseous effluents shall be determined to be within the above limits in accordance with the methods and procedures of the 00CM by obtaining representative samples and performing analyses in accordance with the sampling and analysis program, specified in Table 4.11-2. i ! .
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TAGLE 4.11-2 . w RADIDACTIVE GASE0lls ilASTE SAMPLillG Afl0 AtlALYSIS PROGRAH 9
< Lower Limit of Hininunn
!j Sampling Analysis Type of Detection (ll0) a Gaseous Release Type Frequency Fre<1!!ency Activity Analysis (oci/ml)a
# P _4 li" P Each Tank Principal Ganuita Emitters 9 1 x 10 A. llaste Gas Storage Eacli Tank
' Grair -6 Tank 1 x 10 Sample 11- 3 E -4 E P 9 G Principal Gansna Emitters 1 x 10
- II . Containment Purge Eacli Purge Each Purge Gral' -6 11- 3 1 x 10 Sample C. Ventilation Vent
*
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H Principal Ganuna Emitters ! l x 10-4 I' -6
- 1. Process Vent ,(, il e 11- 3 1 x 10
- 2. Containment Vent
%
u 3. Aux. Bidu. Vent
"
a I-l'31 -12 Continuous 5 1 x 10 Cha.rcoal 11' Samp,le g ,j 33
-
1 x 10"10
~
d 9 Continuous I 11 Principal Ganuna Emitters -II Particulate (1-131.Otliers) 1 x 10 Sample - Continunic I H Gross alpha 1 x 10 " CompasiLe Particulate _ Sample
-
Continuous f 9 Sr-fl9, Sr-96 1 x 10 " Composite Particulate _ Ed"'N'
.
4 TABLE 4.11-2 (Centinued)
,
/]) TABLE NOTATION )
!
- a. The Lower Limit of Detection (LLD) is defined in Table Notation (a) of Table 4.11-1 of Specification 4.11.1.1.
l
- b. Analyses shall also be performed following shutdcwn, startup, or similar operational occurrence which could alter the mixture of radionuclides if warranted by reactor coolant activity changes.
- c. Tritium grab samples shall be taken at least once per 24 hours when the refueling canal is flooded.
- d. Samples shall be changed at least once per 10 days and analyses shall be completed within 48 hours.after charging (or after removal from sampler). Sampling, and analyses shall also be performed at least once per 24 hours for 7 days following each shutdown, startup or similar operational occurrence which lead to increases or decreases 20% in radd9 iodine releases. When samples collected for
,
24 hours are analyzed, the corresponding LLD's may be increased by a factor of 10.
- e. Tritium grab samples shall be taken at least once per 7 days from the ventilation exhaust from the spent fuel pool area.
e f. The average ratio of the sample flow rate to the sampled stream
' l flow rate shall be known for the time period covered by each dose ;
or dose rate calculation made in accordance with Specification 3.11.2.1, 3.11.2. 2 and 3.11.2.3.
- g. The principal gamma emitters for which the LLD specification will apply are exclusively the following radionuclides: Kr-87, Kr-88, Xe-133, Xe-133m, Xe-135, and Xe-138 for gaseous emissions and Mn-54, Fe-59, Co-58, Co-60, In-65, Mo-99, Cs-134, Cs-137, Ce-141, and Ce-144 for particulate emissions. This list does not mean that only these nuclides are to be detected and reported. Other peaks which are
' measurable and identifiable, together with the above nuclides, shall also be identified and reported. Nuclides which are below the LLO for the analyses should not be reported as being present at the LLD level for that nuclide. ,
r~) l kJ BEAVER VALLEY - UNIT 1 3/4 11-10 f ! i
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RADI0 ACTIVE EFFLUENTS DOSE, NOBLE GASES ' LIMITING CONDITION FOR OPERATION 3.11.2.2 The air dose in unrestricted areas (See Figure 5.1-1) due to noble gases released in gaseous effluents shall be limited to the following: 5 mrad for gamma radiation
- a. During any calendar quarter, to 1 and 1 10 mrad for beta radiation.
- b. During any calendar year, to 1 10 mrad for gamma radiation
! and 1 20 mrad for beta radiation; APPLICABILITY: At all times. ACTION _:
- a. With the calculated air dose from radioactive noble gases in gaseous effluents exceeding any of the above limits, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report which identifies the cause(s) for exceeding the limit (s) and defines the corrsctive actions to be taken to reduce the releases of radioactive noble gases in gaseous effluent during the remainder of the current calendar quarter and during the subsequent three calendar quarters so that the average dose during these four ceieader suerters is withia (10) meed for samma raaietica O- and (20) mrad for beta radia1! ion.
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.11.2.2.1 Dose Calculations Cumulative dose contributions for the current calendar quarter and current calendar year shall be determined
in accordance with the Off-Site Dose Calculation Manual (00CM) at least once every 31 days.
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RADI0 ACTIVE EFFLUENTS DOSE - RADI0 IODINES, RADI0 ACTIVE MATERIALS IN PARTICULATE FORM, AND RADIONUCLIDES OTHER THAN NOBLE GASES LIMITING CONDITION FOR OPERATION 3.11.2.3 The dose to an individual from radiciodines and radioactive materials in particulate form, and radionuclides (other than noble gases) with half-lives greater than 8 days in gaseous effluents released from tne site (see Figure 5.1-3) shall be limited to the following:
- a. During any calendar quarter: Less than or equal to 7.5 mrem to any organ and,
- b. During any calendar year: Less than or equal to 15 mrem to any organ.
(The dose design objectives shall be reduced based on predicted carbon-14 releases and turbine building releases if effluent sampling is not provided). APPLICABILITY: At all times. ACTION:
- a. With the calculated dose frem the release of radiciodines, radioactive materials in particulate form, or radionuclides (other than noble gases) with half-lives greater than 8 days, in gaseous effluents exceeding any of the above limits, in lieu of any other report required by Specificition S.9.', prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report which identifies the cause(s) for exceeding the limit and defines the corrective actions to be taken to reduce the releases of radiciodines and radioactive materials in particulate form, and radionuclides (other than noble gases) with half-lives greater than 8 days in gasecu:: effluents during the remainder of the current calendar quarter and during the subsequent three calendar quarters, so that the cumulative dose or dose commitment to an individual from such releases during these four calendar quarters is within (15) mrem to any organ. '
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.11.2.3 Dose Calculations Cumulative dose contributions for the current calendar quarter and current. calendar year shall be determined in accordance 4 with the ODCM at least once every 31 days. ! 3/4 11-12 l BEAVER VALLEY - UNIT 1 i
RADIOACTIVE EFFLUENTS n U GASEOUS RA0 WASTE TREATMENT LIMITING CONDITION FOR OPERATION 3.11.2.4 The gaseous radwaste treatment system and the ventilation exhaust treatment system shall be OPERABLE. The appropriate portions of the gaseous radwaste treatment system shall be used to reduce radioactive materials in gaseous waste prior to their discharge when the projected gaseous effluent air doses due to gaseous effluent releases from the site (see Figure 5.1-3), when averaged over 31 days, would exceed 0.2 mrad for gamma radiation and 0.4 mrad for beta radiation. The appropriate portions of the ventilation exhaust treatment system shall be used to reduce radioactive materials in gaseous waste prior to their discharge when the projected doses due to gaseous effluent releases from the site (see Figure 5.1-3) when averaged over 31 days would exceed 0.3 mrem to any organ. APPLICABILITY: At all times.
'
ACTION:
- a. With the gaseous radwaste treatment system and/or the ventilation exhaust treatment system inoperable for more than 31 days or with gaseous waste being discharged without treatment and in excess of
/~ the above limits, in lieu of any other report required by U) Specification 6.9.1, prepare and submit to the Commission within 30 days, pursuant to Specifigation 6.9.2, a Special Report which includes the following information:
- 1. Identification of the inoperable equipment or subsystems and the reason for inoperability,
- 2. Action (s) taken to restore the inoperable equipment to OPERABLE status, and
- 3. Summary description of action's) taken to prevent a recurrence.
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SILRVEILLANCE RECUIREMENTS 4.11.2.4.1 Doses due to gaseous releases from the site shall be projected at least once per 31 days, in accordance with the 00CM. 4.11.2.4.2 The gaseous radwaste treatment system and ventilation exhaust system shall be demonstrated OPERABLE by operating the gaseous radwaste treatment system equipment and ventilation exhaust treatment system equipment for at least 60 minutes, at least once per 92 days unless the appropriate system has been utilized to process radioactive gaseous effluents during the ] orevious 92 days. SL VER VALLEY - UNIT 1_. 3/4 11-13
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RADIOACTIVE EFFLUE?lTS
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GAS STORAGE TAtlKS ' LIMITING C0tIDITI0ft FOR OPERATI0tl
.
3.11.2.6 The quantity of radioactivity contained in each gas storage tank shall be limited to < 52000 curies noble gases (considered as Xe-133). APPLICABILITY: At all times.
-
ACTION:
- a. With the quantity of rndicactive material in any gas storage tank exceeding the above limit, immediately suspend all additions of radioactive material to the tank and within 48 hours reduce the tank contents to within the limit,
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLAtlCE 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 hours when radioactive materials are being added.to the tank. 1
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l BEAVER VALLEY - UtlIT 1 3/4 11-15 O l
__ RADI0 ACTIVE EFFLUENTS 3/4.11.3 SOLID RADI0 ACTIVE WASTE O LIMITING CONDITION FOR OFERATION 3.11.3 The solid radwaste system shall be used, as applicable, to solidify and package radioactive wastes, and to meet the requirements of 10 CFR Part 20 and of 10 CFR Part 71.
~ -
' APPLICABILITY: At all times. ACTION:
- a. With the applicable requirements of 10 CFR Part 20 and 10 CFR Part 71 not satisfied, suspend affected shipments of solid radioactive wastes from the site.
<
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not
- applicable.
i SURVEILLANCE REQUIREMENTS
.
4.11.3.1 Prior to shipment, solidification shall be verified in accordance with Station Operating Procedures. ()
.
~ ' 2 .
Reoorts - The semi-annual Radioactive Effluent Release Report
,ec ification 6.9.1.12 shall include the following information for each
.jpe of solid waste shipped offsite during the report period: ,
- a. container volume,
- b. total curie quantity (aetermined by measurement or estimate), ,
l
- c. principal radionuclides (determined by measurement or estimate), j
;
- d. type of waste (e.g., spent resin, ccmpacted dry waste l i evaporatorbottoms),
- e. type of container (e.g., LSA, Type A, Type B, large Quantity),and
,
,
- f. solidification agent (e.g. , cement, urea formaldehyde).
() BEAVER VALLEY - UNIT 1 3/4 11-16
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n RADI0 ACTIVE EFFLUENTS U 3/4.11.4 TOTAL DOSE LIMITING CONDITION FOR OFERATION 3.11.4 The dose or dose comitment to any real individual from uranium fuel cycle sources shall be limited to less than or equal to 25 mrem to the total body or any organ (except the thyroid, which shall be limited to less than or equal to 75 mrem) over 12 consecutive months. APPLICABILITY: At all times.
,
ACTION:
- a. With the calculated dose from the release of radioactive materials in liquid or gaseous effluents exceeding twice the limits of Specifications 3.11.1.2.a, 3.ll.l.2.b, 3.ll.2.2.a, 3.11.2.2.b, 3.ll.2.3.a, or 3.11.2.3.b, in lieu of any other report required by Specification 6.9.1, prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 and limit the subsequent releases such that the dose or dose commitment to any real individual from uranium fuel cycle sources is limited to less than or equal to 25 mrem to the total body or any organ (except thyroid, which is q limited to less than or equal to 75 mrem) over 12 consecutive months.
V This Special Report shall include an l'aiy.;; which demonstrated that radiation exposures to any real individual from wanium fuel cycle sources (including all effluent paths cys and direct radiation) are less than the 40 CFR Part 190 Standard. Otherwise, obtain a variance from the Commissfcq to permit releasts which exceed the 40 CFR Part 190 Standard.
-
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS , 4.11.4 Dose Calculations Cumulative dose caritributions from liquid and gaseous effluents shall be determined in accordance w-ith Specifications 4.11.1.2, 4.11.2.2, and 4.11.2.3, and in accordance wit.1 the ODCM. l BEAVER VALLEY - UNIT 1 3/4 11-17
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3 (d 3/4.11 RADIOACTIVE EFFLUENTS BASES 3/4.11.1 LIOUID EFFLUENTS 3/4.11.1.1 CONCENTRATION This specification is provided to ensure that the concentration of radioactive materials released in liquid waste effluents from the site will be less than the concentration levels specified in 10 CFR Part 20, Appendix S, Table II, Column 2. This limitation provides additional assurance that the levels of radioactive materials in bodies of water outside the site will result in exposures within (1) the Section II.A design objectives of Appendix I, 10 CFR 50, to an individual and (2) the limits of 10 CFR 20.106(e) to the population. The concentration limit for dissolved or entrained noble gases is based upon the assumption that Xe-135 is the controlling radioisotope and its MPC *n air (submersion) was converted to an equivalent concentration in water using the methods described in International Commission on Radiological Protection (ICRP) Publication 2. 3/4.11.1.2 DOSE This specification is provided to implement the requirements of
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Sections II.A, III.A and IV.A of Appendix I,10 CFR Part 50. The Limiting Q' Condition for Operation implements the guides set forth in Section II.A of Appendix I. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive material in liquid effluents will be kept "as low as is reasonably achievable." Also, for fresh water sites with drinking water supplies which can be potentially affected by plant operations, there is reasonable r.ssurance that the operation of the facility will not result in radionuclide concentrations in the finished drinking water that are in excess of the requirements of 40 CFR 141. The dose calculations in the ODCM implement the requirements in Section III.A of , Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data such that the actual exposure of an individual through appropriate pathways is unlikely to be substantially underestimated. The equations specified in the 00CM for calculating the doses due to the actual release rates of radioactive materials in liquid effluents are consistent with the methodology provided in Regulatory Guide 1.109,
" Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I,"
Revision 1, October 1977 and Regulatory Guide 1.113, " Estimating Aquatic Dispersion of Effluents from Accidental and Routine Reactor Releases for the Purpose of Implementing Appendix I," April 1977. This specification applies to the release of liquid effluents from each reactor at the site. For units with shared radwaste treatment systems, the j liquid effluents form the shared system are proportioned among the units j sharing that system. ' b^ BEAVER VALLEY - UNIT 1 33/411-1 l
RADIOACTIVE EFFUJENTS BASES 3/4.11.1.3 LIOUID WASTE TREATMENT The OPERABILITY of the liquid radwaste treatment system ensures that this system will be available for use whenever liquid effluents require treatment prior to release to the environment. The requirement that the appropriate portions of this system be used when specified provides assurance that the releases of radioactive materials in liquid effluents will be kept "as low as is reasonably achievable." This specification implements the requirements of 10 CFR Part 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50 and the design objective given in Section II.0 of Appendix I to 10 CFR Part 50. The specified limits governing the use of appropriate portions of the liquid radwaste treatment system were specified as a suitable fraction of the dose design objectives set forth in Section II.A of Appendix I, 10 CFR Part 50, for liquid effluents. 3/4.11.1.4 LIOUID HOLDUP TANKS Restricting the quantity of radioactive material contained in the specified tanks provides assurance that in the event of an uncontrolled release of the tanks' contents, the resulting concentrations would be less than the limits of 10 CFR Part 20, Appendix B, Table II, Column 2, at the nearest potable water supply and the nearest surface water supply O in an unrestricted area. 3/4.11.2 GASEOUS EFFLUENTS 3/4.11.2.1 00SE RATE This specification is provided to ensure that the dose at any time at the site boundary from gaseous effluents from all units on the site will be within the annual dose limits of 10 CFR Part 20 for unrestricted areas. The annual dose limits are the dcses associated with the concentrations of 10 CR Part 20, Appendix B, Table II, Coltzn 1. These limits provide reasonable assurance that radioactive material discharged in gaseous effluents will not result in the exposure of an individual in an unrestricted area, either within or outside the site boundary, to annual average concentrations exceeding the limits specified in Appendixwho individuals B, Table may atII times of 10 CFR Part the be within 20 (10 site CFR Part 20.106(b)(1)). boundary, the occupancyFor, of the individual will be sufficiently low to compensate for any increase in the
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atmospheric diffusion factor above that for the site boundary. The specified release rate limits restrict, at all times, the corresponding gamma and beta I dose rates above background to an individual at or beyond the site boundary to l less than or equal to 500 mrem / year to the total body or to less than or equal to 3000 mrem / year to the skin. These release rate limits also restrict at all times, the corresponding thyroid dose rate above background to an infant via the cow-milk-infant pathway to less than or equal to 1500 mrem / year for the nearest cow to the plant. O SEAVER VALLEY - UNIT 1 B 3/4 11-2
RADI0 ACTIVE EFFLUENTS BASES This specification applies to the release of gaseous effluents from all reactors at the site. For units with shared radwaste treatment systems, the gaseous effluents.from the shared system are proportioned
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among the units sharing that system. 3/4.11.2.2 OOSE -NOBLE GASES This specification is provided to implement the requirements of Sections II.S. III.A, and IV. A of Appendix I,10 CFR Part .50. The Limiting Condition for Operation implements the guides set forth in Section II.B of Appendix I. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radio-active material in gaseous effluents will be kept "as low as is reasonably achievable". The Surveillance Requirements implement the requirements in Section III.A of Appendix ! that conformance with the guides of Appendix I be shown by calculational procedures based on models and data such that the actual exposure of an individual through the appropraite pathways is unlikely to be substantially underestimated. The dose calculations established in the 00CM for calculating the doses due to the actual release rates of radioactive noble gases in gaseous effluents will be consistent with the methodology provided in Regulatory Guide 1.109, " Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the
- O- Purpose of Evaluating Ccmpliance with 10 CFR Part 50, Appendix I, " Revision 1, October,1977 and Regulatory Guide 1.111 " Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors," Revision 1, July,1977. The
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00CM equations provided for determining the air doses at the site. boundary . .
~are' based'upof tih'e 6Et'o'rical average atmospheric conditions.
3/4.11.2.3 COSE - RADI0 IODINES, RADI0 ACTIVE MATERIALS IN PARTICULATE FORM AND RADIONUCLIGES OTHER THAN NOBLE GASES j This specification is provided to implement the requirements of Sections
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II.C, III.A and IV.A of Appendix I,10 CFR Part 50. The Limiting Conditions for Operation are the guides set forth in Section II.C. of Appendix I. , l
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The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix
.
I to assure that the releases of radioactive materials in gaseous effluents
! will be kept "as low as is reasonaoly achievable". The 00CM calculational methods specified .in the surveillance requirements implement the require-
', ments in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data such that the actual exposure of an individual through appropriate path-ways is unlikely to be substantially underestimated. The 00CM calcula-tional methods approved by NRC for calculating the doses due to the actual release rates of the subject materials are required to be consistent with the methodology provided in Regulatory Guide 1.109, O
SEAVER VALLEY - UNIT 1 - B 3/4 11-3
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.O RAoI0AC m m ertu nTS l
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BASES
" Calculating of Annual Doses to Man frem Routing Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," Revision 1, October, 1977, and Regulatory Guide 1.111,
" Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors,"
Revi3 ion 1, July, 1977. These equations also provide for determining l the actual doses based upon the historical average atmospheric conditions.
The release rate specifications for radiciodines, radioactive material in particulate form, and radionuclides other than noble gases are dependent on the existing radionuclide pathways to man, in the unrestricted area. The pathways which are examined in the development of these calculations are: 1) individual inhalation of airborne radionuclides, 2) deposition of ' radionuclides onto green leafy vegetation with subsequent consumption by man, 3) deposition onto grassy areas where milk animals and meat producing ' animals graze with consumption of the milk and meat by man, and 4) deposition on the ground with subsequent exposure of man. 3/4.11.2.4 GASEOUS RADWASTE TREATMEtiT
' The OPERABILITY of the gaseous radwaste treatment system and the
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O veati'etioa exaeust tre>tmeat srste= easures thet the siste== ~4u ee available for use whenever gaseous effluents require treatment prior to release to the environment. The requirement that the appropriate portions e of these systems be used, when specified, provides reasonable assurance
- that the releases of radioactive material in gaseous effluents will be kept "as low as is reasonably achievable." This specification implements the requirements of 10 CFR Part 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50, and the design objectives given in Section~II'.D
- of Appendix I to 10 CFR Part 50. The specified limits governing the use of appropriate protions of the sytems were specified as a suitable fraction of the dose design objectives set forth is Sections II.B and II.C of Appendix I,10 CFR Part 50, for gaseous effluents.
3/4.11.2.5 EXPLOSIVE GAS MIXTURE This specification is provided to ensure that the concentration of potentially explosive gas mixtures contained in the waste gas holdup system is maintained 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 Appeadix A to 10 CFR Part 50. SEAVER VALLEY - UNIT 1 B 3/4 11-4 O
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, y s RADIOACTIVE EFFLUENTS
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BASES 3/4.11.2.6 GAS STORAGE TANKS Restricting the quantity of radioactivity contained in each gas storage tank provides assurance that in the event cf an uncontrolled release of the tank's contents, the resulting total body exposure to an individual at the nearest exclusion area boundary will not exceed 0.5 rem. This is consistent with Standard Review Plan 15.7.1, " Waste Gas System Failure." 3/4.11.3 SOLIO RADI0 ACTIVE WASTE The OPERABILITY of the solid radwaste system ensures that the system will be available for use whenever solid radwastes require processing and packaging prior to being shipped offsite. This specification implements the requirements ' of 10 CFR Part 50.36a and General Design Criterion 60 of Appendix A to 10 LFR 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, mixing and curing times. 3/4.11.4 TOTAL DOSE O This specification is provided to meet the dose limitations of 40 CFR 190. The specification requires the preparation and submittal of a Special Report whenever the calculated doses from plant radioactive effluents exceed twice the design objective doses of Appendix I. For sites containing up to 4 reactors, it is highly unlikely that the resultant dose to a real individual will exceed 40 CFR 190 if the individual reactors remain within the reporting requirement level. The Special Report will describe a course of action which should result in the limitation of dose to a real individual for 12 consecutive months to within the 40 CFR 190 limits. For the purposes of the Special Report, it may be assumed that the dose commitment to the real individual from other uranium fuel cycle sources is negligible, with the exception that dose contributions from l other nuclear fuel cycle facilities at the same site or within a radius ! of 5 miles must be considered. l O 1 BEAVER VALLEY - UNIT 1 3 3/4 11-5 i
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_ -_ ___ _ _ .. _ i i l O ' 5.0 oEStGN FEATUaES
5.1 SITE EXCLUSION AREA 5.1.1 The exclusion area shall be as shown in Figure 5.1-1. LOW POPULATION ZONE 5.1.2 The low population zone shall be as shown in Figure 5.1-2. SITE BOUNDARY FOR GASEOUS EFFLUENTS 5.1.3 The site boundary for gaseous effluents shall be as shown in Figure 5.1-3. SITE 800NDARY FOR LIOUID EFFLUENTS l 5.1.4 The site boundary for liquid effluents shall be as shown in
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Figure 5.1-4. 5.2 CONTAINMENT . CONFIGURATION 5.2.1 The reactor containment building is a steel linec, reinforced concrete building of cylindrical shape, with a dome roof and having the following i design features:
- a. Nominal inside diameter = 125 feet.
- b. Nominal inside height = 185 feet.
- c. Minimum thickness of concrete walls = 4.5 feet.
- d. Minimum thickness of concrete roof = 2.5 feet,
- e. Minimum thickness of foundation mat = 10 feet.
- f. Nominal thickness of vertical portion of steel liner = 3/8 inch.
4 g. Nominal thickness of steel liner, dome portion = 1/2 inch, 0
- h. Net free volume = 1.8 x 10 cubic feet.
DESIGN PRESSURE AND TEMPERATURE 5.2.2 The reactor containment building is designed and shall be maintained for a maximum internal pressure of 45 psig and a temperature of 280*F. O 5-1
, BEAVER VALLEY - UNIT 1-S
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Operatior.s C Divislon
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- Vice
;'. Prestdent E i i l
- General Supt. of 3 Power Stations
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= Tech. Asst. ___ _.__ . ____3 ' Iluclear ' g e i e I
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1 Director of aluclear Oper. I i 1'
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Superintendent Power Stations Power Stations Superintendent Superintendent Superintendent BVP5 - 2 Superintendent AC5H Co. Lic. & Compt. Superintendent Superintendent Pers. & Records Iech. Services 'suclear Services Startup Coord. of Operations Superintendent
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Superintendent Shigiptagpurt of m intenance Cheswick flrama Phillips Office Nnager alcalth invironmental Core Analysis 8)uelity Control Physicist Coordinator Engineer Supervisor Senior Coal 2' Inspector 0 l l m
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Core Operations Quality Control Sr. Compliance Sr. Licensing ** fageneer fagineer lageneer Engineer r.*
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Grr5 lit oic RAllon (rt.itilit)
- Fire Protection responsibility Flcunt 6.r.1
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b 5tation Supt. Nuclear
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Radtstion Ng lear Chief Nuclear ingr. Office m aager [ Results Nuclear Coordtator Control Supv. Engineer & Refuellag Supv I I I I Reactor Control Technical Nuclear Station Iechnical Senior
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5enior fagr. Chemist Supv. - Nuclear Oper. Supv. Advisory Engr. (ngineer Design Control I
""'I ih0II Ih- ""'I - I'" I*' I'"3'-
Radlochemist Shift Supv. AJvisors training Supv. Refueling
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fluclear Station MdI"L'"4"'8 -
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Chem'6sts (e.gle ers. Oper. f oremen Supervisor personnel Records Asst. Radiation _ tiuclear Malatenance Control ingr. Control Oper. fagencers l Cleths Guards I Ncclear Nintenance Operators foremen 44Jiatics Control Nint enance
- Emergency Planning and Fos .wn lethnicians Fjre frotection Responsibilities
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Radiation Tech 4' clans FACIllIV ONGANilATION iIGURI 6.2.2 - _ _ _ _ . _ _ _ __ - __-
ACMINISTRATIVE CCNTROLS , O e.S UNIT STArg OUAtIr1 CATIONS 6.3.1 Each member of the unit staff shall meet or exceed the minimum qualifications of ANSI N18.1-1971 for ccmparable positions, except for the Radiation Control Supervisor who shall meet or exceed the qualificatior.s of Regulatory Guide 1.8, September,1975. 6.4 TRAINING 6.4.1 A retraining and replacement training program for the unit sta.ff shall be maintained under the directiJn of the Training Supervisor and shall meet or exceed the requirements and recomendations of Section 5.5 of ANSI N18.1-1971 and Appendix "A" of 10 CFR Part 55. 6.4.2 A training program for the Emergency Squad shall be maintained under the direction of the Training Supervisor and shall meet or exceed the
- requirements of Section 27 of the NFPA Code - 1976.
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_, 6.5 REVIEW AND AUDIT
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6.5.1 On-Site Safety Cemittee (OSC) FUNCTION
) 6.5.1.1 The On-Site Safety Committee (OSC) shall function to advise the Statiors Superintendent on all matters related to nuclear safety.
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COMPOSITION 6.5.1.2 The OSC shall be composed' of the: Chairman: Chief Engineer Member: Operations Supervisor Member: Radiation Control Supervisor Member: Maintenance Supervisor Member: Nuclear Engineering & Refueling Supervisor Member: Results Coordinator Member: Training Supervisor Member: Office fianager Nuclear (Security Officer)
!! ember: Senior Engineer - Emergency Pla'nning and gire Protection Member: Technical Advisory Engineer i
O BEAVER VALLEY - UNIT 1 6-5 O 1 l
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.. ADMINISTRATIVE CONTROLS O c. Rev4ew of ail proposed chenses to appendix ^ Techn4cai-Specifications.
- d. Review of all proposed changes or modifications to unit systems or equipment that affect nuclear safety.
- e. Investigation of all violations of the Technical Specifications including the preparation aid forwarding of reports covering evaluation and recommendations to prevent recurrence to the General Superintendent of Power Stations and to the Chairman of the Off-Site Review Committee.
- f. Review of events requiring.24-hour written notification to the Commission.
g, Review of unit operations to detect potential nuclear safety hazards.
- h. Performance of special raviews, investigations, or analyses and reports thereon at requested by the Station Superintendent or the Chairman of the Off-Site Review Committee.
- i. Review.of the Security Plan and implementing procedures and shall submit reccmmended changes to the Chairman of the Off-Site Review Comittee.
Revie'.. or every uapiaaaea re'eese or radioactive meter 4e1 to O J. the environs; evaluate the event; specify remedial action to prevent recurrence; and document the event description, evaluation, and corrective action, and the disposition of the . corrective action in the plant records. _ _ _
- k. Review of changes to the Process Control Program, Offsite Oose Manual, and radwaste treatment systems.
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AUTHORITY _. 6.5.1.7 The On-Site Safety Committee (OSC) shall:
- a. Recommend to the Station Superintendent written approval or disapproval of items considered under 6.5.1.6(a) through (d) above.
- b. Render deteminations in writing with regard to whether or not each item considered under 6.5.1.6(a) through (e) above constitutes an unreviewed safety question.
- c. Pruide written notification within 24 hours to the General Superintendent of Power Stations and Off-Site Review Comittee of disagreement between the OSC and the S;ation Superintendent; however, tht: Station Superintendent shall have responsibility for resolution of such disagreements pursuant to 6.1.1 above.
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BEAVER VALLEY - UNIT 1 6-7
ADMINISTRATIVE CONTROLS i G , U COMPOSITI0ft , 6.5.2.2 The ORC shall be composed of the: Chairman: Director of Nuclear Operations, Power Stations Department Vice Chairman: Beaver Valley Power Station Project Manager Member: Nuclear Engineer, Mechanical Engineering Department Member: Beaver Valley Power Station Superintendent Member: Superintendent of Nuclear Services, Power ' Stations Department Member: Technical Assistant to the Vice President-Operations Member: Mechanical Engineer - Mechanical Engineering Department Member: Technical Assistant Nuclear, Power Stations Department Member: Quality Assurance Manager - Quality Assurance Departmut 11 ember: Superintendent of Licensing and Compliance, Power Stations Department , ALTERNATES , 6.5.2.3 All alternate members shall be appointed in writing by the ORC chairman to serve on a temporary basis; however, no more than two alternates shall participate as voting members in ORC activities O at ear oae time. . CONSULTANTS 6.5.2.4 Consultants shall be utilized as determined by the ORC Chairman to provide expert advice to the ORC. MEETING FREQUENCY 6.5.2.5 The ORC shall meet at least once per six months. OUORUM 6.5.2.6 A quorum of ORC shall consist of the Chairman or his designated alternate and at least 4 members including alternates. No more than a minority of the quorum shall have line responsibility for operation of 1 the facility.
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BEAVER VALLEY - UNIT 1 6-9
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ACMINISTRATIVE CONTROLS
'O AUDITS
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6.5.2.8 Audits of unit activities shall be performed under the cognizance of the ORC. These audits shall enccmpass:
- a. The conformance of unit operation to provisions contained
' within the Technical Specifications and applicable license conditions at least once per 12 months.
- b. The ;:erformance, training, and qualifications of the Facility staff at least once per 12 months.
- c. The results of actions taken to correct deficiencies occurring in unit equipment, structures, systems, or methods of oceration that affect nuclear safety at least once per 6 months.
- d. The performance of activities required by the Operations Quality Assurance Program to meet the criteria of Appendix "B",10 CFR 50, at least once per 24 months.
- e. The Facility Emergency Plan and implementing procedures at least once per 24 months.
- f. The Facility Security Plan and implementing procedures at least once per 24 conths.
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O g. Any other area of unit operation considered appropriate by the ORC or the General Superintendent cf Power Stations Department.
- h. The Facility -Fire Protection Program and implementing procedures at least once per 24 months.
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- 1. An independent fire protection and loss prevention program inspection and audit shall be perfomed annually utilizing either qualified off-site licensee personnel or an outside fire protection firm.
- j. An inspection and audit of the fire protection and loss prevention program shall he performed by an outside qualified fire consultant at wtervals no greater than 3 years.
- k. The radiological er.vironmental monitoring program anc the results thereof at least once per 12 months.
- 1. The OFFSITE 00SE CALCULATION MANUAL and implementine procedures at least once per 24 months.
- m. The PROCESS CONTROL PROGRAM and implementing procedures for solidification of radioactive wastes at least once per 24 months.
- n. The performance of activities required by the Quality Assurance Program to meet the criteria of Regulatory Guide 4.15, Cecember 1977 at least C
. l ' once per 12 months. SEAVER VALLEY - UNIT 1 6-11 __
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6.7 SAFETY LDtIT VIOLATION O 6.7.i The foiiewins ect4 ens sheii ee texen in the event e Sefety tim 4t is violated:
- a. The unit shall be placed in at least NOT STANDBY within one hour.
- b. The Safety Limit violation shall be reported to the Comission, the General Superintendent of Power Stations, and to the Off-Site Review Comittee Chairman or his alternate within 24 hours.
- c. A Safety Limit Violation Report shall be preparec. The report shall be reviewed by the On-Site Safety Cemittee. This recort shall describe (1) applicable circumstances preceding the violation, (2) effects of the violation upon facility components, systems or structures, and (3) corrective action taken to prevent recurrence.
- d. The Safety Limit Violation Report shall be submitted to the Comission, the Off-Site Review Comittee, and the General Superintendent of Power Stations within 14 days of the violation.
6.8 PROCEDURES 6.8.1 Written procedures shall be established, implemented, and maintained covering the activities referenced below:
- a. The applicable procedures recomended in Appendix "A" $f Regulatory Guide 1.33, Revision 2, February, 1978,
- b. Refueling operations.
- c. Surveillance and test activities of safety related equipment.
- d. Security Plan implementation.
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- e. Emergency Plan implementation.
- f. Fire Protection Program implementation.
- g. PROCESS CONTROL PROGRAM implementation. I
- h. OFF-SITE DOSE CALCULATION MANUAL . implementation. ,
- i. Quality Assurance Program for effluent and environmental monitoring using the guidance in Regulatory Guide 4.15, December 1977.
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BEAVER VALLEY - UNIT 1 6-13 l _
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ADMINISTRATIVE CONTROLS (')) %
- d. Reactivity anomalies involving disagreement with the predicted value of reactivity balance under steady state conditions during power operation greater than or equal to 1% ak/k; a calculated reactivity balance indicating a SHUTDOWN MARGIN less conservative than specified in the technical specifications; short-tenn reactivity increases that correspond to a reactor period of less than 5 seconds or, if subcritical, an unplanned reactivity insertion of more than 0.5% ak/k; or occurrence of any unplanned criticality.
- e. Failure or malfunctier. of. one or more ccmoonents which prevents or could prevent, by itself, the fulfillment of the functional requirements of system (s) us d to cope with accidents analyzed in the SAR.
- f. Personnel error or procedural inadequacy which prevents or could prevent, by itself, the fulfillment of the fuictional requirements of systems required to cope with accidents analyzed in the SAR.
- g. Conditions arising frem natural or man-made events that, as a direct result of the event require unit shutdown, operation of safety systems, or other protective measures s required by technical specifications.
- h. Errors discovered in the transient or accident analyses or in the methods used for such analyses as described in the safety analysis report or in the bases for the technical specifications that have or could have permitted reactor operation in a manner less conservative than assumed in the analyses.
- i. Performance of structures, systems, or components that requires remedial action or corrective measures to prevent operation in a manner less conservative than assumed in the accident analyses in the safety analysis report or technical specification bases; or discovery during unit life of conditions not specifically considered in the safety analysis report or technical :pecifica-tions that require remedial action. or corrective measures to prevent the existence or development of an unsafe condition.
- j. Occurrence of an unusual or important event that causes a significant radiological impact, or that has high public or potential public interest concerning environmental impact frcm l unit operation.
- k. Offsite releases of radioactive materials in liquid and gaseous effluents which exceed the limits of Specification 3.11.1.1 or 3.11.2.1.
(] 1. Exceeding the limits in Specification 3.11.1.4 or 3.11.2.5 for storage of radioactive materials in the listed tanks. The written l l follow up activities planned and/or taken to reduce the centents
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to within specified limits.
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BEAVER VALLEY - UNIT 1 6-17
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ADMINISTRATIVE CONTROLS > 0 THIRTY OAY WRITTEN REPORTS 6.9.1. 9 The types of events listed below shall be the subject of written reports to the Director of the Regional Office within thirty days of occurrences of the event. The written report shall include, as a minimum, a completed copy of a licensee event report form. Informa-tion provided on the license'e event report form shall be supplemented, as needed, by additional narrative material to provide complete explana-tion of the circumstances surrounding the event.
- a. Reactor. protection system or engineered safety feature instrument settings which are found to be less conservative
! than those established by the technical specifications but l which do not prevent the fulfillrhent of the functional requirements of affected systems.
- b. Conditions leading to operation in a degraded mode permitted by a limiting condition for operation or plant shutdown required by a limiting condition for operation.
NOTE: Routine surveillance testing, instrument calibration, or preventive maintenance which require system configurations as described in items a and b need , not be reported except where test results themselves i reveal a degraded m a as described above.
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O~ c.
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Observed inadequacies in the implementation of administrative or procedural controls which threaten to cause reduction of degree of ridundancy provided in reactor protection systems or engineered safety feature systems. ,
- d. Abnormai degradation of systems other,than those specified
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in 6.9.1.8 above designed to contain radioactive material , resulting from the fission process. 1
- e. An unplanned offsite release of 1) more than 1 curie of radioactive material in liquid effluents, 2) more than 150 curies of noble gas in gaseous effluents, or 3) more than 0.05 curies of radiciodine in gaseous effluents.
Th'e report of an unplanned offsite release of radioactive
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material shall include the following infomation:
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- l. A description of the event and equipsent involved.
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- 2. Cause(s) for the unplanned release.
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- 3. Actions taken to prevent recurrence.
- 4. Consequences of the unplanned release. l
. f. Measured levels of radioactivity in an environmental sampling medium determined to exceed the reporting level values of Q Table 3.12-2 when averaged over any calendar quartij!r sampling period.
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6-18 ' BEAVER VALLEY - UNIT-1
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ADMINISTRATIVE CONTROLS O ' The reports shall also include the following: a sumary description of the radiological environmental monitoring program and a map of all sampling locations keyed to a table giving distances and directions from one reactor. The result of land use censuses required by the Specification 3.12.2 and the results of licensee participation in the Quality Assurance Program required by Specification 3.12.3. SEMI-ANNUAL RADI0 ACTIVE EFFLUENT RELEASE REPORT 1/ t . 6.9.1.12 Routine radioactive effluent release reports covering the operating of the unit during the previous 6 months of operation shall be submitted within 60 days after January 1 and July 1 of each year. 6.9.1.13 The radioactive effluent rel' ease reports shall include a sumary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit as outlined in Regulatory Guide 1.2, " Measuring, Evaluating, and Reporting Radioactivity in Solid ' Wastes and Releases of Radioactive Materials in Liquid and Gaseous Effluents from Light-Water-Cooled Nuclear Power Plants", with data-sumarized on a quarterly basis following the format of Appendix B thereof. ,
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The radioactive effluent release report to be submi d d 60 days after January 1 of each year shall include an annual summary of hourly meterological data collected over the previous year. This ' pV annual summary may be either in the form on an hour-by-hour listing of wind speed, wind direction, atmospheric stability, and precipitation (if measured) on magnetic tape, or in the form of joint frequency distributions of wind speed, wind direction, and atmospheric stability. A suggested magnetic tape format is available from the Hydrology-Meteorology Branch, Division of Site Safety and Environmental Analysis, NRR, NRC, Washington, D.C. 20555. This same report shall include an assessment of the radiation doses due to the radioactive liquid and 7 gaseous effluents released from the unit or station during the previous calendar year. This same report shall also include an assessment of the radiation doses from liquid and gaseous radioactive effluents to individuals due to their activities inside the site boundary (Figure 5.1-1) during the report period. All assumptions used in making these assessments (e.g., specific activity, exposure time and location) shall be included in these reports. The meteorological conditions concurrent with the time of release of radioactive materials in gaseous effluents (as determined by i sampling frequency and measurement) shall be used for determining the gaseous pathway doses. The assessment of radiation doses shall be performed in accordance with the Off-Site Case Calculation Manual (00CM). ,
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3/ A single submittal may be made for a multiple unit station. The submittal should combine those sections that are comen to all units at the station; however, for units with separate radwaste systems, the submittal shall specify the releases of radioactive material from each unit.
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BEAVER VALLEY - UNIT 1 6-21
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O ^ost"tsta^ttve contao's The radioactive effluent release reports shall include the following information for each type of solid waste shipped offsite during the report period:
- a. Container volume.
- b. Total curies quantitiy (specify whether determined by measurement or estimate.
- c. Principal radionuclides (specify whether determined by measurement or estimate.
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- d. Type of waste (e.g., spend resin, compacted dry waste, evapcrator
, bottoms).
- e. Type of container (e.g., LSA, Type A, Type B, Large quantity), and
! 4 f. Solidification agent (e.g., cement, urea formaldehyde). 1' The radioactive effluent release reports shall include unplanned releases from the site to unrestricted areas of radioactive material in gaseous and liquid effluents on a quantity basis. ' The radioactive effluent release reports shall include any changes to the (/
) PROCESS CONTROL PROGRAM (PCPM) made during the reporting period, as
(_ provided in Speci,fication 6.14. SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the Director of the Office of Inspection and Enforcement Regional Office within the time period specified for each report. These reports shall be submitted covering the activities identified below pursuant to the requirements of the applicable reference specification:
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J ! , 4 O BEAVER VALLEY - UNIT 1 6-22
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O AostNtsta^Ttve coNTaols radiation area and entrance thereto shall be controlled by requiring issuance of a Radiation Work Permit *. Any individual or group of individuals permitted to enter such areas shall be provided with or acccmpanied by one or more of the following:
- a. A radiation monitoring device which continuously indicates the radiation dose rate in the area.
- b. A radiation monitoring device which continuously integrates the radiation dose rate in the area and alarms when a preset inte-grated dose is received. Entry into such areas with this monitoring device may be made after the dose rate level in the area has been established and personnel have been made kncw-ledgeable of them.
- c. An individual qualified in radiation protection procedures who is equipped with a radiation dose rate monitoring device. This individual shall be responsible for providing positive control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified by a facility Health Physics Supervisor in the Radiation Work Permit.
6.12.2 The requirements of 6.12.1, above, also apply to each high radia-p tion area in which the intensity of radiation is greater than 1000 mrem /hr.
\ In addition, locked doors shall b'e provided to prevent unauthorized entry into such areas and the keys shall be maintained under the administrative control of the Shif t Supervisor on duty and/or the Plant Health Physicist.
6.13 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM , t The Director of Huclear Operations delegates the responsibility for the Radiological Environmental Monitoring Program to the environmental administrator (Superintendent of Licensing ~ Compliance, [ Figure 6.13.1]) or his designated alternate. The environmental administrator is responsible for administering the off-site Radiological Environmental Monitoring Program. He shall determine that the sampling program is being implemented as described to verify that the environment is adequately protected under existing procedures. He shall also have the responsibility for establishing, implementing, maintaking, and approving off-site environmental program sampling, analyses, and calibration procedures.
- Health Physics personnel, or personnel escorted by Health Physics personnel in accordance with approved emergency procedures, shall be exempt from the RWP issuance requirement during the performance of their radiation protection duties, provided they comply with approved radiation protection procedures for entry into high radiation areas.
O t V BEAVER VALLEY - UNIT 1 6-26 _ _
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O O O m C < General Supt. $ Power Stations 5 (.------.------------ I
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I Director of E I Nuclear Operation-. G I - I
' Off-site Review Conmittee Station Supt. Supt. of Supt. of
- Beaver Valley Licen. & Compl. Nuclear Services g,
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nv roninental Station Staff Coordinator Review (Radiolojical ' Review (Ecologi]al
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{0ff-SiteProgram) Of f Site PrograA,i) On-Site lleal tii Of f-Site Environ. Core Analysis Env or am Physicist Program Engineer FIGURE 6.13.1 - Duquesne Light Company Power Stations Departu:ent Partial Organization Chart Relative to Environmental Matters
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O AcMrNrSTaATtvE CONTa0LS 6.14 PROCESS CONTROL PROGRAM (pCP) FUNCTION 6.14.1 The pCP shall be a manual containing the processing steps and a set of established process parameters detailing the program of sampling, analysis, and evaluation witnin which solidification of radioactive wastes from liquid systems is assured, consistent with Specification 3.11.3.1 , and the surveillance requirements of these Technical Specifications. 6.14.2 Licensee initiated changes:
- 1. Shall be submitted to the Ccemission by inclusion in the semi-annual Radioactive Effluent Release Report for the period in which the change (s) was made.and sna11 contain:
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- a. Sufficiently detailed information to totally support the rationale for the change without benefit of additional or supplemental informa-tion;
- b. a detemination that the change did not reduce the overall conforma'ce of the solidified waste O- product to existing criteria for solid wastes; -
and -
- c. documentation of the m that the changes has been reviewed and four acceptable by tne OSC.
- 2. Shall become effective upon review and acceptance by the ORC. ,
6.15 0FFSITE DOSE CALCULATION MANUAL (00CM) FUNCTION 6.15.1 The 00CM shall describe the methodology and parameters to be used in the calculation of offsite doses due to radioactive gaseous and liquid effluents and in the calculation of gaseous and liquid effluent monitoring instrunentation alarm / trip setpoints consistent with the applicable LCO's ccatained in these Technical Specifications. Methodologies ar.d caf cula-tional procedures acceptable to the Commission are contained in NUREG-0133.
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t 5.0 ADMINISTRATIVE CONTROLS ' , O Objective To define the organization, assign responsibilities, describe the environmental surveillance procedures, provide for a review and audit 1 function, and prescribe the reporting requirements in order to insure continuing protection of the environment and implement the Environmental Technical Specifications. . 5.1 ORGANIZATION Figure 5.1-1 is a partial organization chart of the Power Stations f' Oepartment which includes groups and individuals responsible for environmental protection, environmental monitoring and the implementation of the Environmental Technical Specifications following the issuance of i ' ' an operating license for Beaver Valley Unit No.1. 5.2 RESPONSIBILITY All activities within the Power Stations Department are the responsibility of the General Superintendent. He delegates the responsibility for ecological environmental monitoring programs to the Superintendent of
- Licensing and Compliance. He is advised by the Off-Site Review Committee in matters relating to the environmental impact of station operation which involve proposed changes to the Environmental Technical Specifications, proposed changes to written procedures, proposed changes or modifications n to station systems or equipment, evaluations of investigations conducted as U a result of environmental technical specification violations, and when
appropriate, recommendations to prevent recurrence of such violations. The Station Superintendent is the individual cnarged with the responsibility for assuring that all releases to the environment shall be limited in accordance with the Limiting Conditions for Operations
as described in the Environmental Technical Specifications. Controlled releases are made utilizing written procedures under the direction of ' the Shift Supervisor at the request of the Reactor Control Chemist (for non-radiological discharges) or the Radiation Control Supervisor (for discharges requiring a radiological discharge permit) who specifies the release rate, the station operating conditions, and the monitoring , requirements which must be maintained'during the release to comply with ! the technical specifications. These activities are reviewed by the , Radiochemist and the Station Operating Supervisor and audited periodically l under the direction of the Superintendent of Licensing and Ccmpliance, ' Power Stations Department. The Station Superinter#,nt is also responsible for the issuance of the 24-hour reporting requirement to the Director of the Regional Regulatory Operations Office, whn required. l O- 5-1 l ! ! __. . . _ . _ m
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O . O - O - General Supt. Powgr Stations
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(___---------------- l l Director of I Nuclear Operation' . I 1
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Off-site Review Coninittee
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Station Supt. SUFt. of Supt. of
<s. Beaver Valley Licen. & Compl. Nuclear Services so . i l
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Station Staff Coordinator
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Review (Ecologi]al
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Of f Site Prograh) I , On-Site . Off-Site Environ. Core Analysis Environ. Program Prograni Engineer
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(Ef fluents)
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FIGURE 5.1-1 - Duquesne Liglit Company
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Power Stations Department Partial Organization Chart Relative to Environmental Matters
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5.3 REVIEW AND AUDIT
- O The Seper4nteneent e ucensing ane Compiience is resxns4bie to the Off-Site Review Comittee for reviewing the results of the Environmental Surveillance Program to determine that the sampling program is being implemented as described to verify that the environment is adequately
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2 protected under the existing station operating procedures. He is also
- responsible for requesting that analyses be repeated or that new samples be obtained from particular locations in the case of anomalous results.
The Superintendent of Licensing and Compliance is responsible to the Off-Site Review Comittee for scheduling audits utilizing various company personnel to verify that station operations including releases to the
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environment are conducted in such a manner that continuing protection to i the environment is provided. The Superintendent of Licensing and Compliance shall also have the responsibility for reviewing off-site environmental program sampling, analyses and calibration procedures as sumarized in Table 3.1-3 and described in Sections 3.0 and 5.5.1 of these specifications. The Beaver Valley Power Station Superintendent is responsible for reviewing on-site procedures relative to the environment as described in Section 2.0 and Section 5.5.1 of these specifications. The Off-Site Review Comittee functions as they relate to the Environmental Technical Specifications are as follows:
- a. Reviews the reports of the Superintendent of Licensing and Compliance en the results of the environmental monitoring program prior to their submittal to the NRC in each Annual O Non-Radiological Environmental Report. See Section 5.6.1.
- b. Review and make recomendations on proposed changes to the Environmental Technical Specifications and the evaluated impact of the changes.
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- c. Review proposed changes or modifications to plant systems or equipment and the evaluated impact which would require a change in the procedures described in (d) below.
- d. Review reported instances of violations of Environmental Technical Specifications and abnormal environmental occurrences. Where investigation indicates, evaluate and formulate recommendations to prevent recurrence.
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- e. Schedule audits of Station Operation with regard to the control of releases to the environment.
- f. Review and compare the Safety Technical Specifications and the Environmental Technical Specifications to a'roid conflicts and maintain consistency.
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- ! ACTION TO BE TAKEN IF A LIMITING CONDITION FOR OPERATION IS EXCEEDED
' 5.4.1 Remedial Action i Innediate remedial actions as permitted by these technical
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specifications will be implemented until such time as the
; limiting condition for operation is met.
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! 5.4.2 Investigation The Station Superintendent shall direct that an investigation be conducted to determine the cause of the occurrence and recommend i
both temporary and permanent corrective action to prevent a recurrence.
; The results of this investigation shall be reviewed and approved by the Off-Site Review Committee and the General Superintendent, Power Staticns Department.
5.4.3 Reoort The occurrence shall ce reported to the NRC as specified in Section 5.6.2. 5.5 PROCEDURES 5.5.1 Written Procedures Detailed written procedures, including check lists and special instructions when required, will be prepared and implemented for
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all activities involving releases to the environment, described l in Sections 2.0 and 3.0, the sampling and monitoring of these ( releases and the conduct of the Environmental Surveillance Program. These procedures will include sampling, instrument calibration, analysis, and actions to be taken when limits are approached or exceeded. Testing and calibration frequencies of effluent alarms and monitoring instruments will be included. These frequencies will be determined from experience with similar instruments in similar environments and from manufacturers' recommendations. l 5.5.2 Standard Goeratino Procedures Plant standard operating procedures will include, when required, sufficient precaution, limitations, and setpoints to ensure that
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l I all piant systems and components are operated in compliance with j the limiting conditions for operations established in Section 2.0
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of these techtkal specifications. 5.5.3 Reviews l All procedures described above, and all changes thereto, will be
' reviewed prior to implementation and periodically thereafter as described in Section 5.3. Temporary changes to procedures which r
do not change the intent of the original procedure may be made l without the need for review by NRC. Changes to procedures for ' programs in Section 3 will be reviewed by the NRC. Such changes will be documented. l 5-4 l w
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r 1 ! DUQUESNE LIGHT COMPANY ; ! Beaver Valley Power Station, Unit No. 1 ! 1 O oecx c " so-33' License No. DPR-66
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; Attachment B I
j Radiological Effluent Technical Specifications i t
, Table of Exceptions or Changes ;
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O i The item numbers correspond to exceptions or l
- changes noted on the attached Technical Specifications t
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O Item Comment Number Pm Section h 3/4 3-51 Table 3.3-11 Item 3.b Device number added to clarify specific device h 3/4 3-53 Table 4.3-11 Items 1.a & 1.b Note (5) added in source check column as in previous submittal h 3/4 3-54 Table 4.3-11 Table Notation (5) Note (5) added as in previous submittal 3/4 3-55 3.3.3.9 For certain ventilarion Action A systems, the shutdown of the system =ay result in a less safe condition h 3/4 3-57 Table 3.3-12 Item 4.a Device number added to clarify specific device. ACTION state-ment should be as in previous submittal. The requirement to go to cold shutdown may result in more gas created, and there-g ( fore, a less safe condition h 3/4 3-57 Table 3.3-12 Table 4.3-12 To clarify elevated release point Item 3
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3/4 3-60 h 3/4 3-56 3/4 3-57 Table 3.3-12 Items 2.b & 2.c To correct numbering of action statements Items 3.b & 3.c h 3/4 3-58 ACTION 30 Supplementary leak collection and release system does not service containment areas - this would be in conflict with other technical specifications h 3/4 3-60 Table 4.3-12 Item 4.a To correct numbering of Uote (4) in channel l calibration column h 3/4 3-61 Table 4.3-12 Table Notation (3) To clarify frequency of channel calibration h 3/4 11-3 Table 4.11-1 Table Notation Definition of LLD inappro-priate for batch releases of l
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effluent. Use previous sumittal _
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1 !O ~ Item ,
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Number Me, Section Comment l h 3/4 11-6 4.11.1.3.2 Provide time for system operation h 3/4 11-8 4.11.2.1.2 Correct numbering of section h 3/4 11-9 Table 4.11-2 Item C Corrset ventilation vent description 3/4 11-13 4.11.2.4.2 Provide time for system operation 3/4 11-16 3.11.3 Use previous submittal h B3/4 11-4 3/4.11.2.5 f.utomatic control features not applicable at Beaver Valley
h 5-1 5.2.2 To provide missing pressure and temperature data
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O h 6-22 6.9.1.13 Delete paragraph regarding
<=e1 c7c1 eerce - sec
! required unless twice limits exceeded
' h 6-27 6.14.1 and 6.15.1 The PCP and ODCM will be approved by the ORC prior i
to implementation i
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INDET DEFINITIONS
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) SECTION
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Page 1.0 DEFINITIONS _ 1 -1 Defined Terms................................................ 1-1 Thermal Power................................................ 1-1 Rated Thermal Power.......................................... 1-1 Operational Mode............................................. 1-1 Action....................................................... 1-1 Operable - Operability....................................... 1-2 Reportable Occurrence........................................ 1-2 Containment Integrity........................................ 1-2 Channel Cal i b ra ti o n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-2 Channel Check................................................ 1-3 Channel Func ti onal Tes t. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-3 Co r e Al te ra ti o n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-3 Shut'down Margin.............................................. 1-3 Iden ti fi ed Le akag e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-3 Unidentified Leakage......................................... 1-4 Pressure Boundary Leakage.................................... 1-4~
^] Co ntro11 ed le akag e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
.. 1-4 Quadrant Power Til t Ratio . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
1-4 Dose Equivalent I-131........................................ 1-4 Staggered Test Basis......................................... 1-4 Frequency Notation........................................... 1-5 Reacto r Trip Res pons e Time . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-5 Engineered Safety Feature Re'spense Time...................... 1-5 Axial Fl u x Di f f e re nc e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-5 Physics Test................................................. Energy.............................. 1-5 E-Average Disentegration .
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1-6 Source Check................................................. 1-6 P roce s s Co n t rol P rog ram . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . , 1-6 Solidification................................................ 1-6 Off-Site Dose Calculation Manual (00CM)....................... 1-6 Gaseous Radwaste Treatment System............................. System.......................... 1-6 l Ventilation Exhaust Treatment 1-7 Purge / Purging................................................. 1-7 Venting....................................................... 1-8
" Opera tional Mo de s (Tabl e 1.1 ) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
I-6 Frequency Notation-........................................... BEAVER VALLEY - UNIT 1- I
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INDEX LIMITING CO::0ITIC::S FCTt OPERATIC:t At:0 SL'R'!EILLA : E RECUIRE!!E?tTS . Pace SECTION 3/4.2 PO'!ER DISTRIBUTIO!! LIltITS 3/4 2-1 3/4.2.1 Axi al Fl ux Di f f e r en c e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
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3/4 2-5 3/4.2.2 Hea t Fl ux Ho t Cha nnel Fa ctor. . . . . . . . . . . . . . . . . . . . . . . . . . . Nucl ear Enthal py Ho t Channel Factor. . . . . . . . . . . . . . . . . . . . 3/4 2.-8 3/4.2.3 ' 3/4 2-10 3/4.2.4 Quadrant Fewer Ti l t Ra ti o . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 2-12 3/4.2.5 DNS Parameters......................................... 3/4.3 I?tSTRUMENTATIOri 3/4 3 *. 3/4.3.1 P ROT ECT IV E I NSTRUM EliTAT!0 f t . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 3-14 3/4.3.2 ENGINEERED SAFETY FEATURE INSTR" MENTATION.............. 3/4.3.3 MONITORING INSTRUMENTATION
~ Rad i a ti o n Mo ni to ri ng . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
3/4 3-33 3/4 3-37 Nova bl e Incor e De tecto r s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 3-33 O ~s e t s=i c t a s trc=ea:> :4 c a . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Me teorological Ins trumentation. . . . . . . . . . . . . . . . . . . . . . . . . 3/4 3-41 3/4 3-44 Remote Shutdown Ins trumentation . . . . . . .. . . . . . . . . . . . . . . . . ~ 3/4 3-47 Fire De tecti on Ins trument ation. . . . . . . . . . . . . . . . . . . . . . . . . Radioactive Liquid Effluent Instrumentation................ 3/4 3 49 Radioactive Liquid Effluent Instrumentaticn................ 3/4 3-50 ' Radioactive Gaseous Effluent !!cni tors . . . . . . . . . . . . . . . . . . . . . . 3/4 3-55
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3 / 4 . 4_ REACTOR COOLANT SYSTER
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3/4.4.1 REACTOR CCOLANT LOOPS 3/4 4-1 ' Normal Operation....................................... 3/4 4-3 Isolatad loop.......................................... , Isolated Loop Startup.................................. 3/4 4-4 . O IV BEAVER YALLEY - UNIT 1 \
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_._ INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS C) SECTt0N PAGE 3/4.9.10 WATER LEVEL-REACTOR VESSEL............................. 3/4 9-10 3/4.9.11 STORAGE POOL WATER LEVEL............................... 3/4 9-11 3/4.9.12 FUEL BUILDING VENTILATION SYSTEM - FUEL MOVEMENT. . . . . .. 3/4 9-12 3/4.9.13 FUEL BUILDING VENTILATION SYSTEM - FUEL STORAGE. . . . . . . . 3/4 9-13
- 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOWN MARGIN........................................ 3/4 10-1 3/4.10.2 GROUP HEIGHT AND INSERTION LIMITS...................... 3/4 10-2 3/4.10.3 PRESSURE / TEMPERATURE LIMITATIONS-REACTOR CRITICALITY... 3/4 10-4 3/4.10.4 PHYSICS TEST........................................... 3/4 10-6 3/4.10.5 NO FLOW TESTS.......................................... 3/4 10-7 3/4.11 RADIOACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS CONCENTRATION.......................................... 3/4 11-1 r3 00SE................................................... 3/4 11-5 t '
LIQUID WASTE TREATMENT................................. 3/4 11-6 L I QU I D HO L DU P TAN KS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 11-7 3/4 11.2 GASEOUS EFFLUENTS DOSE RATE.............................................. 3/4 11-8 DOSE - NOBLE GASES..................................... 3/4 11-11 DOSE - RADI0 IODINES, PARTICULATE AND RADIONUCLIDES OTHER THAN NOBLE GASES................................. 3/4 11-12 GASEOUS WASTE TREATMENT................................ 3/4 11-13 GAS STO RAG E TAN KS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 11-15 3/4.11.3 SOLID RADI0 ACTIVE WASTE................................ 3/4 11-16 3/4.11.4 TOTAL 00SE............................................. 3/4 11-17 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.1 MONITORING PR0 GRAM..................................... 3/4 12-1 3/4.12.2 LAND USE CENSUS........................................ 3/4 12-9 3/4.12.3 INTERLABORATORY COMPARISON PROGRAM..............., .... 3/4 12-10 x_-
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O 1.0 CEFINITI0 tis (Continued) . PURGE-PURGING 1.33 PURGE or PURGING is the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating conditions, in such a manner that replacement air or gas is
required to purify the confinement.
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VENTING 1.34 VENTING is the controlled process of discharging air or gas frem a confinement to maintain temperature, pressure, humidity, concentration or other operating conditions, in such a manner that replacement air or gas is not provided or required during VENTING. Vent, used in system names, does not imply a VENTING process.
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OPERATIONAL MODES
% RATED AVERAGE COOLANT REACTIVITY CONDITION. X,ff THERMAL POWER
- TEMPERATURE MODE
> 0.99 > 5% > 350*F
- 1. POWER OPERATION _
< 5% >
- 2. STARTUP > 0.99 _
_350*F
>
- 3. HOT STANDBY < 0.99 0 _350*F
< 0.99 0 350*F > Tavg
- 4. HOT SHtJiDOWN
> 200*F
< 0.99 0 < 200'F
- 5. COLD SHUTDOWN _
< 0.95 0 < 140'F I
- 6. REFUELING **
- Exclucitig decay heat.
** Reactor vessel head unbolted or removed and fuel in the vessel.
l l BEAVER VALLEY - UNIT 1 1-8
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., -,w-- =y- -- , ,-,,.r-.-.,,------------.-,we. , -.m,,-- y,_ ,,w--y+,--.m-.
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5 TABLE 1.2 ( ])- , FRECUENCY NOTATION l FREQUENCY I NOTATION At least once per 12 hours. S 1 i At least once per 24 hours. D j At least cnce per 7 days. W
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At least once per 31 days M At least once per 92 days. I - Q At least once per 6 months. SA At least once per 18 mcnths. R Prior to each reactor startup. S/U ~ Completed prior to each release. I P i N.A. Not applicable.
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_ _ . - _ . . . _ . _ _ _ . _ . - . _ . . _ - - , . . _ . _ , _ . ..__.._-_ ... , _
' () INSTRUMENTATION , RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.8 The radioactive liquid effluent monitoring instrumentation channels shown in Table 3.3-11 shall be OPERABLE with their alarm / trip setpoints set to ensure that the limits of Specification 3.11.1.1 are not exceeded. The alarm / trip setpoints of these channels shall be determined in accordance with the Offsite Dose Calculation Manual (00CM). APPLICABILITY: At all times. ACTION:
- a. With a radioactive liquid effluent monitoring instrumentation
' channel alarm / trip setpoint less conservative than required by the above specification, immediately suspend the release of radioactive liquid effluents monitored by the affected channel or declare the channel inoperable.
- b. With one or more radioactive liquid effluent monitoring instrumentation channels inoperable, take the ACTION shown in Table 3.3-11.
O s/ c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. SURVEILLANCE REQUIREMENTS
;
4.3.3.8 Each radioactive liquid effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECX, SOURCE CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operatiens at the frequencies shown in Table 4.3-11. i 1
- BEAVER VALLEY - UNIT 1 3/4 3-50
[) l l l
, __ . _ _ - _ _ _ _
O O O TABLE 3.3-11 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION MINIMUM N CilAtaELS
- U INSTRUMENT OPERABLE ACTION
!3 ' < l. Gross Activity Monitors Providing Automatic E Tennination of Release m
- a. Liquid Waste Effluents Monitor (RM-LW-104) (1) 23 7
23 s b. Liquid Waste Contaminated Drain Monitor (1) $ (RM-LW-116) ~
- 2. Gross Activity Monitors Not Providing Tennination of Release
- a. Component Cooling-Recirculation Spray (1) 24 m
lleat Exchangers River Water Monitor 2 (RM-RW-100)
- 3. Flow Rate Measurement Devices
- a. Liquid Radyaste Effluent Line (1) 25 (1) FR-LW-103 (2) FR-LW-104
- b. Cooling Tower Blowdown Line (1) 25 (1) FT-CW-101
((2) FT-CW-101-1
- 4. Tank Level Indicating Devices (For tanks outside plant building)
- a. Primary Water Storage Tank (BR-TK-6A) (1) 26
- b. Primary Water Storage Tank (BR-TK-6B) (1) 26
- c. Steam Generator Drain Tank (LW-TK-7A) (1) 26
- d. Steam Generator Drain Tank (LW-TK-7B) (1) 26
- e. Refueling Water Storage Tank (QS-TK-1) (1) 26
TABLE 3.3-11 (Continued) , TABLE NOTATION ACTION 23 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases may be resumed for up to 14 days provided that prior to initiating a release:
- 1. At least two independent samples are analyzed in accordance with Specification A.11.1.1.3, and;
- 2. At least two technically qualified memcers of the Facility Staff independently verify the release rate calculations and discharge valving; Otherwise, suspend release of radioactive effluents via this pathway.
ACTION 24 - With the number of channels OPERABLE less than required by tne Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days provided that at least once per 8 hours grab samples are at a analyzed lower Limitforofgross radioactivity Detection (LLD) of(beta or gamm9)uCi/ml. at least 10-ACTION 25 - With the number of channels OPERABLE less than required pd by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days provided the flow rate is. estimated at least once per 4 hours during actual releases. Pump curve may be used to estimate ficw. ACTION 25 - With the number of channels OPERABLE less than required
; by the Minimum Channels OPERABLE requirement, liquid additions to this tank may continue for up to 30 days provided the tank liquid level is estimated during all liquid additions to the tank.
. f BEAVER VALLEY - UNIT 1 3/4 3-52 g) L
;
l __-_ ____ _ __ . _ _ _ _ __ _
. - - - . - - ..
_ O O O
,
TABLE 4.3-11 un 9 RADI0 ACTIVE LIQUID EFFLUENT MONITORING 1
,
w INSTRUMENTATION SURVEILLANCE REQUIREMENTS CllANNEL si CllANNEL SOURCE CilANNEL FUNCTIONAL i~ CllECK CllECK CAllt1 RATION TEST
!
.
INSTRUMENT h 1. Gross Beta or Ganana Radioactivity Monitors Pro-
= viding Alanu and Automatic Tennination of Release R(3) Q(1) 2 a. Liquid Radwaste Effluent Line D P (RM-LW-104)
R(3) Q(1)
- b. Liquid Waste Contaminated Drain Line D P S)
(RM-LW-ll6)
- 2. Gross Deta or Ganina Radioactivity Monitors Providing Alann but not providing Automatic t' Tennination of Release R(3) Q(2)
[, a. Component Cooling-Recirculation Spray 0 M di lleat Exchangers River Water Monitor
""
(RM-RW-100)
- 3. Flow Rate Monitors NA R Q
- d. Liquid RadWaste Effluent Lines D(4)
(1) FR-LW-103 (2) FR-LW-104 NA It Q
- b. Cooling Tower Blowdown Line (FT-CW-101) U(4)
'
- 4. Tank Level Monitors for the Following Tanks D* HA R Q
- a. Primary Water Storage Tank - (PG-TK-6A)
D* NA R Q
- b. Primary Water Storage Tank - (PG-TK-6B)
D' NA R Q
- c. Stearii Generator Drain Tank - (LW-TK-7A)
D* NA R Q
'
- d. Steam Generator Drain Tank - (LW-TK-711) R Q ItefuelingWaterStorageTank(QS-TK-1) D' HA e.
.
*
.
TABLE 3.3-11 (Continued) TABLE NOTATION
~
* - Ouring liquid additions to the. tank.
.
--
(1) - The CHANNEL Fi1NCTIONAL TEST shall also demonstrate that automatic isolation of this pathway and Control Room Alarm Annunciation occurs if any of the following conditions exist:
- 1. Instrument indicates measured levels above the alarm / trip setpoint.
- 2. Circuit failure.
- 3. Instrurent controls not set in operate mode.
' (2) - The CHANNEL FUNCTIONAL TEST shall also demonstrate that control room alarm annunciation occurs if any of the following conditions exist:
- 1. Instrument indicates measured levels above the alann/ trip setpoint.
- 2. Circuit failure.
- 3. Instrument controls are rio.t set in operate mode.
(3) - The initial CHANNEL CALIBRATION for radioactivity measurement instrumentation shat 1 be performed using one or more of the reference standards certified by the National Bureau of Standards
-
or using standards that have been obtained from suppliers that
) participate in measurement assurance activities with NBS. These standards should permit calibrating the system over its intended range of energy and rate capabilities. For subsequent CHANNEL '
CALIBRATION, sources that have been related to the initial calibration should be used, at intervals of at least once per eighteen months.
'
This can normally be accomplished during refueling outages.
- - - . . -- -- _ .. . _
.. .
__ __
~
( 4) - CHANNEL CHECK shall consist of verifying indication of flow during periods of release. CHANNEL CHECK shall be made at least once daily on any day on which car,tinuous, periodic, or batch releases are made. (5) - A channel check may verify system operation. In these instances,
-
3 ~ a source check may be waived, since high backgrounds may mask check source strength. BEAVER VALLEY - UNIT 1 3/4 3-54
_ _ _
.
0
% IftSTRUMENTATI0ff
-
RADI0 ACTIVE GASEOUS EFFLUEtiT M0:11TORING ITISTRUMENTATION LIMITING CONDITION FOR OPERATI0tl 3.3.3.9 The radioactive gaseous effluent monitoring instrumentation channels shown in Table 3.3-12 shall be OPERABLE with their alarm / trip setpoints set to ensure that the limits of Specification 3.11.2.1 are not exceeded. The alarm / trip setpoints of-these channels shall be determined in accordance with tne 00CM. APPLICABILITY: As shown in Table 3.3-12. ACTION:
- a. With a radioactive gaseous effluent monitoring instrumentation IL' channel alam/ trip setpoint less conservative than required by
\ ,> the above Specification declare the channel inoperable,
- b. With one or more radioactive gaseous effluent monitorir.g instrumentation channels inoperable, take the ACTION shown in Table 3.3-12.
- c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
b SURVEILLAftCE REOUIREMENTS 4.3.3.9 Each radioactive gaseous effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations at the frequencies shown in Table 4.3-12. , i e SEAVER VALLEY - UNIT 1 3/4 3-55
,
.
. . - . .
_ _ _ - __ _ - - _ . _ _ - - _. _ _ O O O
.
- TABLE 3.3-12 9 RADIDACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION p]
5 MINIMUM l- CllANNELS Q OPERABLE APPLICABILITY ACTION
, INSTRUMENT b 1. ' Gaseous Waste Particulate and Gas Monitors (RM-GW-108A & B)
]
- 27
- a. Noble Gas Activity Monitor (1) 32
'
- b. lodine Sampler Cartridge (li
- l
* ' 32
- c. Particulate Sampler Filter (1)
- 28 m d. Ef fluent Flow Rate licasuring (1)
,1 2 Device (FR-GW-108) i
- e. Sampler Flow Rate Measuring Device
$.
o
- 2. Auxiliary Building Ventilation System
.
(RM-VS-101A & B) 29
*
'
- a. Nuble Gas Activity Monitor (1) 1
- 8 Iodine Sampler Cartridge- (1)
- c. Particulate Sampler Filter (1)
- Q
- 28
- d. Flow Rate Monitor (FR-VS-101) (1)
* ~28
- e. Sampler Flow Rate Monitor (1)
t
- During Releases via this pathway i
.
- _ . . - _ - .- - _ . . _. .. _ _ - . - . ..
~ O o o
;
m . TAllLE 3.3-12 (Continued) , h RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION
$
<
MINIMUM 72 CitANNELS l~ ACTION
- INSTRUMENT OPERABLE APPLICAlllLITY
- 3. Elevated Release ((Located on top D
] G (Reactor Containsent)f RM-VS-107A & u)
- 30
~
- a. Notele Gas Activity Monitor (1)
- b. lodine Sampler Cartridge (1)
- I k
4
- c. particalate Sampler Filter (1)
- h
- 28
- d. Flow Rate Monitor (FR-VS-ll2) (1)
' M
- 28
- e. Sampler Flow Rate Monitor (1)
]
ln
" Waste Gas Decay Tanks Monitor 4.
**
- a. Oxygen Monitor 2-AS-GW-110-1,2 (1)
, i
- During Releases via this path.say.
** During waste gas decay tank filling operation.
, k i
' .
l
. _ _ _ _ - - _ _ - - - --___ _ .
._-. - - . . . - _ _ _ _ _ , TABLE 3.3-12 (Continued) -
TABLE NOTATION ACTION 27 - With the number of channels OPEP.ABLE less than required by the Minimum Channels OPERABLE requirement, the contents of the tar.k may be released to the environment for up to 14 days provided that, prior to initiating the release:
- 1. At least two independent sar;les of the tank's ,
content are analyzed, and
.
- 2. At leas two technically qualified members of the Facili t. Staff independently verify the release rate cr.iculations and discharge valve lineup.
Otherwise, suspend release of radioactive effluents via
,
this pathway. ACTION 28 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 28 days provided the flow rate is estimated at least once per 4 , hours. ACTION 29 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent relea'ses s) .
- via this pathway may continue for up to 28 days provided i grab samples are taken at least once per 8 hours and these samples are analyzed for gross activity within 24 hours.
ACTION 30 - With the number of channels OPERABLE less than required o by Minimum Channels OPcDABLE recuirement, immediately O suspendPURGINGof@eactorcontainment}/iathispathway. ACTION 31 - With the number of channels OPERABLE one less than required by the MINIMUM Channels OPERABLE requirement, cperation of this system may continue provided grab samples are obtained every 4 hours. ACTION 32 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days, provided samples are continuously collected with auxiliary sampling
. equipment as required in Table 4.11-2.
! O se^vea v4LLev - uNtT i 3/4 3-s8 .
.y , -- -
- . , . . .n. y y, -.a.. -
---en
- . _ _ . ._ - _ _ . . __ . . _ _ _ - __ _
O o O TABLE 4.3-12 h RADIDACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS h CilANilEL h CllANNEL SOURCE CilANNEL FUNCTIONAL p, CllECK CALIBRATION. TEST ' < INSTRUMENT _ CllECK
,
E 1. Gaseous Haste Particulate and Gas Monitors 3 (RM-GW-108A & B) R(3) Q(1)
~
- a. Noble Gas Activity Monitor P* P(S)
W N/A N/A N/A
- b. todine Sampler Cartridge N/A N/A N/A
- c. Particulate Sampler Filter W P* N/A R 4
- d. System Effluent Flow Rate Measuring Device D* N/A R Q
- e. Sampler Flow Rate Measuring Device
- 2. Auxiliary Bldg. Vent. System m
3 (RM-VS-101A & B) tioble Gas Activity Mont tor D M R(3) Q(2)
<r a.
W N/A N/A N/A E b. Iodine Sampler N/A N/A N/A
- c. Particulate Sampler W D ,N/A R N/A l d. System Effluent Flow Rate
- Measurement Device D N/A R Q
- e. Sampler Flow Rate Me.isurement Device
! .
-- .-- - - - -
__ _ . __ . _. -- . - __. - . O o O TABLE 4.3-12 (Continued) g
,j RADIDACTIVE GASEOUS rFFLtlENT MONITORING INSTRUMENTATION SURVEILLANCE REQtilREMEN
, w CilANNEL
$
,
CilANNEL SOURCE CilANNEL FUNCT10NAL G
- INSTRUMENT CilECK CllECK CAllBRATION TEST
.
S 3. ElevatedRelease$opofReactop
~i (Containmentg)>ni toring 5ys teiii
"
(RM-v5-lu/A & B) 2 H R(3) Q(1)
- a. Noble Gas Activity Monitor D*
D* N/A N/A N/A
- b. Iodine Sampler D* 'N/A N/A N/A
- c. Particulate Sampler D* N/A R N/A
- d. System Effluent Flow Rate Measuring Device g
D* N/A R Q
?> e. Sampler Flow Rate Measuring g Device Waste Gas Decay Tanks Monitor
' 4. D N/A f M
- a. Oxygen Monitor
. 1
.
k e _
TABLE 4.3-12 p d (Continued)
~
TABLE NOTATION
- During releases via this pathway.
** During waste gas holdup system operation (treatment for primary system offgases).
(1) The CHANNEL FUNCTIONAL TEST shall also demonstrate that automatic isolation of this pathway and control room alarm annunciation occurs if any of the following conditions exist:
- 1. Instrument indicates measured levels above the alarm / trip setpoint.
- 2. Circuit failure.
- 3. Instrument controls not set in operate mode.
(2) The CHANNEL FUNCTIONAL TEST shall also demonstrate that control room alarm annunciation occurs if any of the following conditions exist:
- 1. Instrument indicates measured levels above the alarm / trip setpoint.
- 2. Circuit failure.
- 3. Instrument controls not set in operate mode.
(3) The initial CHANNEL CALIBRATION for radioactivity measurement instrumentation shall be performed using one or more of the reference standards certified by the National Bureau of Standards or using standards that have been obtained from suppliers that participate in measurement assurance activities with NBS. These standards should permit calibrating the system over its intended range of energy and rate capabilities. For subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration should be used,(at intervals go 5f at least once per eighteen montns) This can normally be accomplisneo curing reruei ng outages. Wx1 sting plants may substitute previously established calibration procedures for this requirement.) O BEAVER VALLEY - UNIT 1 3/4 3-61
. -
.__ _ . _ . __ _ _ _. .. _ _ . _ . _ _._- - _ ____.
N
.
;
^
t i
,
.
'
'
TABLE 4.3-12 ' j l (Continued)
i (4) The CHANNEL CALIBRATION shall include the use of standard gas samples containing a nominal:
- 1. One, volume percent oxygen, balance nitrogen; and
'
- 2. Four volume percent oxygen, balance nitrogen.
(5) A channel check may verify system operation. In these instances, a
source check may be waived, since background may mask check source strength. l
.
t I i O
*
1 ' ' l b I i, i , i l l ' O BEAVER VALLEY - UNIT 1 3/4 3-62
.
. - _ - _ . _ - - _-__
O 3/4.11 RADIOACTIVE EFFLUENTS
-
3/4.11.1 LICUID EFFLUENTS CONCENTRATION LIMITING CONDITION FOR OPERATION 3.11.1.1 The concentration of radioactive material released from the site (see Figure 5.1-2) shall be limited to the concentrations scecified in 10 CFR Part 20, Appendix B, Table II, Column 2 for radionuclides other than dissolved or entrained noble gases. For diss theconcentrationshallbelimitedto2x10givedorentrainednoblegases,
':.Ci/ml total activity.
,
APPLICABILITY: At all times. ACTION: , With the concentration of radioactive material released from the site exceeding the above limits, immediately restore the concentration to within the above limits. . SURVEILLANCE REOUIREMENTS O 4.11.1.1.1 The radioactivity content of each batch of radioactive liquid waste shall be determined prior to release by sampling and analysis in accordance with Table 4.11-1. The results of pre-release analyses shall be used with the calculational methods in the CDCM to assure that the concentration at the point of release is maintained within the limits of Specification 3.11.1.1.
- 4.11.1.1.2 Post-release analyses of samples composited from batch releases shall be performed in accordance with Table 4.11-1. The results of the previous post-release analyses shall be used with the calculational methods in the ODCM to assure that the concentrations at the point of release were maintained within the limits of Specification 3.11.1.1.
4.11.1.1.3 The radioactivity concentration of liquids discharged from continuous release points shall be determined by collection and analysis of samples in accordance with Table 4.11-1. The results of the analyses shall be used with the calculational methods in the 00CM to assure that the concentrations at the point of release are maintained within the limits of Specification 3.11.1.1. b" BEAVER VALLEY - UNIT 1 3/4 11-1
- -. . - -.
I TAELE a.11-1 ]
-
RADI0 ACTIVE LICUID WASTE SAMPLIrlG Afl0 AtlALYSIS PROGRAM Minimum Lower Limit Sampling Analysis Type of Activity of Detection Liquid Release Frequency Frequency Analysis (LLD) Type (uCf/mi)a P P ,
- 4. Batch Waste d Each Batch Each Batch Principal _ Gama 5 x 10-#
Release Tanks Emi ttersi
-6 I-131 1 x 10 P
One Batch /M M Dissolved and 1 x 10' Entrained Gases (Gama Emmiters) P Each Batch M H-3 1 x 10'a' b Composite -7 ' Gross Alpha 1 x 10 P-32 1 x 10:6 Q '
- P ~g Each Batch Q Sr-89, Sr-90 5 x 10 b
Composite Fe c o- 1 x 10-6
.
B. Continuog9 Grab Sample 9 W c Principal Gama 5 x 10-7' Releases Composite Emittersf l I-131 1 x 10-6
-
Grab Sample 9 M Dissolved and 1 x 10 ' Entrained Gases i (Gama Emitters) Grab Sample 9 M e H 1 x 10-5 Composite Gross Alpha 1 x 10 ~7
-6
P 1 x 10
-0 Grab Sample 9 1 Sr-89, Sr-90 5 x 10 c
Composite Fe-55 1 x 10-6 ! O BEAVER VALLEY - Ut41T 1 3/4 11-2
._ _ , , , . _ _ . - . - . _ .
TABLE 4.11-1 (Continued) TABLE NOTATION
.
. The Lower Limit of Detection (LLD) is defined in terms of the back-ground counting rate as follows:
2 ,' 2 S
\\ LLD = b 2.22 x E x V Where S is the standard deviation of the background counting rate b
2.22 is the numcer of pCi per disintegration / min. E is the counting efficiency in counts per disintegration V is the sample size analyzed in appropriate units It should be recognized that the LLD is defined as an a orfori (before the fact) limit representing the capability of a measurement system and not as a posteriori (after the fact) limit for a particular measurement. Analyses shall be performed in suca a manner that the stated LLDs LLDs will be achieved under routine conditions. Occasionally back-ground fluctuations, unavoidtMy small sample sizes, the presence of interferring nuclides, or 7tner uncontrollable circumstances may render these LLDs unachievabic. In such cases, the contricuting [~}
- factors will be identified and described in the semi-annual Radio-active Effluent Release Report.
- b. A composite sample is one in which the quantity of liquid sampled is proportional to the quantity of 1 auid waste discharged and in which the method of sampling employed results in a speciment which is representative of the liquids released.
- c. To be representative of the quantities and concentrations of radio-active materials in liquid effluents, samples shall be collected continuously in proportion to the rate of flow of the effluent stream.
Prior to analyses, all samples taken for the composite shall be thoroughly mixed in order for the composite sample to be representa-tive of the effluent release.
- d. A batch release is the discharge of liquid wastes of a discrete volume. Prior to sampling for analyses, each batch shall be isolated, and then thoroughly mixed by a method described in the ODCM, to assure represegtativesampling.
- e. A continuous release is the discharge of liquid wastes of a non-j discrete volume; e.g., from a volume of system that has an input flow during the continuous release. This is applicable to the Turbine Building drains only.
l A ' L] SEAVER VALLEY - UNIT 1 3/4 11-3 l
- --
- .
_ -_
*
,
-
8
'
O~ TABLE d.11-1 (Ccntinued). TABLE NOTATION
- f. The principal gamma emitters for which the LLD specification will apply are exclusively the following radionuclides: Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141, and Ce-144.
This list does not mean that only these nuclides are to be detected and reported. Other peaks which are measurable and identifiable, together with the above nuclides, shall also be identified and , reported. Nuclides which are below the LLD for the analyses should be reported as "less than" the nuclide's LLD, and should not be reported as being present at the LLD level for that nuclide. The "less than" values should not be used in the required dose calculations.
- g. Whenever there is primary to secondary leakage, sampling is done for turbine building drain effluents by means of grab sampling taken every 4 hours during the period of discharge and analyted for gross radioactivity (beta and gamma) at a sencitivity of 10-' uCi/ml and recorded in the plant records, along with the flow rate. Primary to secondary leakage is considered to be occurring whenever measurements indicate that secondary coolant gross activity (beta and gamma) is greater than 10-3 uCi/ml. In addition, two plant personnel snali check release calculations to verify that the limits of 3.11.1.1 and 4 3.11.1.2 are not exceeded.
.
l . . BEAVER VALLEY - UNIT 1 3/4 11-4 (
;
!
I
# I l
e.
' --- - . . . .- _ . . _
_ . O RADI0 ACTIVE EFFLUENTS . DOSE LIMITING CONDITION FOR OPERATION 3.11.1.2 The dose or dose commitment to an individual from radioactive materials in liquid effluents released from the site (see Figure 5.1-4) shall be limited:
- a. During any calendar quarter to less than or equal to 1.5 mrem to the total body anc to less than or equal to 5 mrem to any organ, and
- b. During any calendar year to less than or equal to 3 mre:n to the total body'and to less than or equal to 10 mrem to any organ.
APPLICABILITY: At all times. ACTION:
- a. With the calculated dcse from the release of radioactive materials in liquid. effluents exceeding any of the above limits, in lieu of any other report required by Specification 6.9.1, prepare e$ and submit to the Commission within 30 days, pursuant to (s) Specification 6.9.2, a Special Report which identifies the cause(s)
, for exceeding the limit (s) and defines the corrective actions to be taken to reduce the releases of radioactive materials in liquid
effluents during the remainder of the current calendar quarter and l during the subsequent three calendar quarters, so that the cumulative dose or dose comnitment to an individual from such releases during these four calendar quarters is within 3 mrem to the total body and 10 mrem to any organ. (This Special Report shall also include (1) the results of radiological analyses of the drinking water source and (2) the radiological impact on finished drinking water supplies with regard to the requirements of 40 CFR 141, Safe Drinking Water Act).*
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE RECUIREMENTS 4.11.1.2 Dose Calculations. Cumuistive dose contributions from liquid effluents snall be determined in accordance with the 00CM at least once per 31 days.
- Applicable only if drinking water supply is taken from the receiving water body.
.
~
BEAVER VALLEY - UNIT 1 3/4 11-5
.
4
, -r
O) V - RADIOACTIVE EFFLUENTS LICUID WASTE TREATMENT LIMITING CONDITION FOR OPERATION 3.11.1.3 The liquid radwaste treatment system shall be OPERABLE. The appropriate portions of the system shall be used to reduce the radioactive materials in liquid wastes prior to their discharge when the projected doses due to the liquid effluent frem ;he site (see Figure 5.1-4) when i averaged over 31 days, would exceed 0.06 mrem to the total body or 0.2 mrem to any organ. APPLICABILITY: At all times. l ACTION:
- a. With the liquid radwaste treatment system inoperable for more than 31 days or with radioactive liquid waste being discharged without treatment and in excess of the above limits, in lieu of any other report required by Specification 6.9.1, prepare and submit to the Commission within 30 days pursuant to Specification 6.9.2 a Special Report which includes the following information:
O 1. tdeatification of the inoPereele eauiPment or sees > stems ene the reason for inoperability,
- 2. Action (s) taken to restore the inoperable equipment to OPERABLE status, and
- 3. Summary description of action (s) taken to prevent a recurrence.
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REGUIREMENTS~ 4.11.1.3.1 Doses due to liquid releases shall be projected at least once per 31 days, in accordance with CDCM. . 4.11.1.3.2 The liquid radwaste treatment system shall be demonstrated l OPERABLE v opt 'ating the liquid radwaste treatment system equipment for i d at least 60 ..inutes at least once per 92 days unless the liquid radwaste system has oeen utili::ed to process radioactive liquid effluents during the previous 92 days. , ! ) O SEAVER VALLEY - UNIT 1 3/4 11-6
._
. .
RADIOACTIVE EFFLUENTS
.
LIQUID HOLDUP TANXS
>
LIMITING CONDITION FOR OPERATION 3.11.1.4 The cuantity of radioactive material contained in each of the following tanks shall be limited to < 10 curies, excluding tritium and _, dissolved or entrained noble gases.
- a. BR-TK-6 A (Primary Water Storage Tank)
- b. BR-TK-6B (Primary Water Storage Tank)
- c. LW-TK-7A (Steam Generator Orsin Tank)
- d. LW-TX-7B (Steam,Generatcr Orain Tank)
- e. QS-TK-1 (Refueling Water Storage Tank)
APPLICABILITY: At all times. , ACTION:
- a. With the quantity of, radioactive material in any of the above listed tanks exceeding the above limit, immediately suspend all additions of radioactive material to the tank and within 48 hours reduce the tank contents, to within the limit. _.
.
~
- b. The provisions of Specifications 3'.0.3 and 3.0.4 are not applicable. '
SURVEILLANCE REQUIREMENTS 4.11.1.4 The quantity of radioactive material contained in each of the above listed tanks shall be determined to be within the above limit by analyzing a representative sample of the tank's contents at least once per 7 days when radioactive materials are being added to the tank.
-~
_ _ . BEAVER VALLEY - UNIT 1 3/4 11-7 ) l
'
l
O RADIOACTIVE EFFLUENTS
-
3/4.11.2 GASEOUS EFFLUEt4TS DOSE RATE LIMITING CONDITION FOR OPERATION 3.11.2.1 The dose rate due to radioactive materials released in gaseous effluents from the site (see Figure 5.1-3) shall be limited to the following:
- a. For noble gases: Less than or equal to 500 mrem /yr. to the total
, ,
body and less than or equal to 3000 mrem /yr. to the skin, and
- b. For all radioiodines and for all radioactive materials in particulate form and radionuclides (other than noble gases) with half lives greater than 8 days: Less than or equal to 1500 mrem /yr. to any organ.
APPLICABILITY: At all times. ACTION: With the dose rate (s) exceeding the above limits, immediately decrease the release rate to within the above limit (s).
.
SURVEILLANCE RECUIREMENTS 4
'
4.11.2.1.1 The dose rate due to noble gases in gasecus effluents shall be determined to be within the above limits in accordance with the methods and procedures of the ODCM. 4.ll.2.1h The dose rate due to radioactive materials, other than noble i O)3 gases, in gaseous effluents shall be determined to be within the above limits in accordance with the methods and procedures of the 00CM by obtaining representative samples and performing analyses in accordance with the sampling and analysis program, specified in Table 4.11-2.
.
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TABLE 4.11-2 (Continued)
/] ~
TABLE'N0TATION
- a. The Lower Limit of Detection (LLD) is defined in Table NotationJa) of Table 4.11-1 of Specification 4.11.1.1.
- b. Analyses shall also be performed following shutdown, startup, or similar operational occurrence which could alter the mixture of radionuclides if warranted by reactor coolant activity changes.
- c. Tritium grab samples shall be taken at least once per 24 hours when the refueling canal is flooded.
- d. Samples shall be changed at least once per 10 days and analyses.
shall be completed within 48 hours after charging (or after removal from sampler). Sampling, and analyses shall also be perfonned at least once per 24 hours for 7 days following each shutdown, startup or similar operational occurrence which lead to increases or decreases 20% in radiciodine releases. When samples collected for 24 hours are analy:ed, the corresponding LLD's may be increased by a factor of 10.
- e. Tritium grab samp'es shall be taken at least once per 7 days from the ventilation exhaust from the spent fuel pool area.
- f. The average ratio of the sample flow rate to the sampled stream
, flow rate shall be known for the time period covered by each dose Os or dose rate calculation made in accordance with Specification 3.11.2.1, 3.11.2.2 and 3 11.2.3.
- g. The principal gama emitters for which the LLO specification will apply are exclusively the following radionuclides: Kr-87, Kr-88, Xe-133, Xe-133m, Xe-135, and Xe-138 for gaseous emissions and Mn-54, Fe-59,-Co-58, Co-60, In-65, Mo-99, Cs-134, Cs-137, Ce-141, and Ce-ld4 for particulate emissions. This list does not mean that only these nuclides are to be detected and reported. Other peaks which are measurable and identifiable, together with the above nuclides, shall also be identified and reported. Nuclides which are below the LLO for the analyses should not be reported as being present at the LLD level for that nuclide.
BEAVER VALLEY - UNIT 1 3/4 11 7 . i 1
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RADI0 ACTIVE EFFLUENTS
) .
DOSE, NOBLE GASES 4 LIMITING CONDITICN FOR OPERATION 3.11.2.2 The air dose in unrestricted areas (See Figure 5.1-1) due to noble gases released in gaseous effluents shall be limited to the following: 5 mead for gamma radiation
- a. During any calendar quarter, to 1 and 1 10 mrad for beta radiation.
10 mrad for gamma radiation
- b. During any calendar year, to 1 and 1 20 mrad for beta radiation; APPLICABILITY: At all times. .
ACTION _:
- a. With the calculated air dose from radioactive noble gases in gaseous effluents exceeding any of the above limits, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report which identifies the cause(s) for exceeding the liait(s) and defines the corrective actions to be taken to reduce the releases of radioactive noble gases in gaseous effluents during the remainder of the
, current calendar quarter and during the subsequent three calendar quarters so that the average dose during these four
- calendar quarters is within (10) mrad for gamma radiation and (20) mrad for beta radiation.
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
.
SURVEILLANCE REOUIREMENTS 4.11.2.2.1 Dose Calculations Cumulative dose contributions for the current calendar quarter and current calendar year shall be determined in accordance with the Off-Site Oose Calculation Manual (00CM) at least once every 31 days. BEAVER VALLEY - UNIT 1 3/4 11-11
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V RADI0 ACTIVE EFFUJENTS , DOSE - RADI0 IODINES, RADI0 ACTIVE MATERIALS IN PARTICULATE FORM, AND RADI0huCLIOES OTHER IHAN NOBLE GASES LIMITING CONDITION FOR OPERATION 3.11.2.3 The dose to an individual from radiciodines and radioactive materials in particulate form, ar.d radionuclides (other than noble gases) with half-lives greater than 8 days in gaseous effluents released from the site (see Figure 5.1-3) shall be limited to the following:
- a. During any calendar quartar: Less than or equal to 7.5 mrem to any organ and,
- b. During any calendar year: Less than or equal to 15 mrem to any organ.
(The dose design objectives shall be reduced based on predicted carbon-14 releases and turbine building releases if effluent sampling is not provided). APPLICABILITY: At all times. ACTION: 4 (' With the calculated dose from the release of radiciodines,
-
a. radioactive materials in particulate form, or radionuclides (other than noole gases) with half-lives greater than 8 days, in gaseous effluents exceeding any of the above limits, in lieu of any other report required by Specification 6.9.1, prepare and submit to the Ccmission within 30 days, pursuant to Specification 6.9.2, a Special Report which identifies the cause(s) for exceeding the limit and defines tne corrective actions to be taken to reduce the releases of radiciodines and radioactive materials in particulate form, and radionuclides (other than noble gases) with half-lives greater than 8 days in gaseous effluents during the remainder of the current calendar quarter and during the subsequent three calendar quarters, so
'
that the cumulative dose or dose commit: ent to an individual from such releases during these four calendar quarters is within (15) crem to any organ.
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE RECUIREMENTS 4.11.2.3 Dose Calculations Cumulative dose contributions for the current I calendar quarter and current calendar year shall be determined in accordance with the 00CM at least once every 31 days. G
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SEAVEP. VALLEY - UNIT 1 l l __ ,
RADIOACTIVE EFFLUENTS
.
GASECUS RADWASTE TREATMENT LIMITING CONDITION FOR OPERATION 3.11.2.4 The gaseous radwaste treatment system and the ventilation exhaust treatment system shall be OPERABLE. The appropriate portions of the gaseous radwaste treatment system shall be used to reduce radioactive materials in gaseous waste prior to their dischatge when the projected gaseous effluent air doses due to gaseous effluent releases from tne site (see Figure 5.1-3), when averaged over 31 days, would exceed 0.2 mrad for gama radiation and 0.4 mrad for beta radiation. The appropriate portions of the ventilation exhaust treatment system shall be used to reduce radioactive materials in gaseous waste prior to their discharge when the projected doses due to gaseous effluent releases from the site (see Figure 5.1-3) when averaged over 31 days would exceed 0.3 mrem to any organ. APPLICABILITY: At all times. ACTION:
- a. With the gaseous radwaste treatment system and/or the ventilation exhaust treatment system inoperable for more than 31 days or with gaseous waste being discharged without treatment and in excess of the above limits, in lieu of any other report required by
/') (- Specification 6.9.1, prepare and submit to the Comission within 30 days, pursuant to Specificaticn 6.9.2, a Special Report which includes the following information:
- 1. Identification of the inoperable equipment or subsystems and the reason for inoperability,
- 2. Action (s) taken to restore the inoperable equipment to OPERABLE status, and
- 3. Summary description of action (s) taken to prevent a recurrence.
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SjlRVEILLANCE REQUIREMENTS 4.11.2.4.1 Ooses due to gaseous releases from the site shall be projected at least once per 31 days, in accordance with the CDCM. 4.11.2.4.2 The gaseous radwaste treatment system and ventilation exhaust system shall be demonstrated OPERABLE by operating the gaseous radwaste treatment sy equipment and ventilation exhaust treatment system equipment for at least minutes, at least once per 92 days unless the appropriate Ot5 system has been utilized to process radioactive gaseous effluents during the previous 92 days. BEAVER VALLEY - UNIT 1 3/4 11-13 _ _ _ _
THIS PAGE It4TENTI0flALLY LEFT BLAtlK O
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RADIOACTIVE EFFLUENTS GAS STORAGE TANKS f-') V LIMITING CONDITION FOR OPERATION 3.11.2.6 The quantity of radioactivity contained in each gas storage tank shall be limited to < 52000 curies noble gases (considered as Xe-133). APPLICABILITY: At all times.
'
ACTION:
- a. With the quantity of radioactive material in any gas storage tank exceeding the above limit, immediately suspend all additions of radioactive material to the tank and within 48 hours reduce the tank contents to within the limit.
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
.
SURVEILLANCE REOUIREMENTS 4.11.2.6 The quantity of radioactive material contained in each gas storage tank shall be determined to be within the above lim'it at least once per 24 hours when radioactive materials are being added to the tank.
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' BEAVER VALLEY - UNIT 1 3/4 11-15 0 . 9
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RADIOACTIVE EFFLUENTS 3/4.11.3 SOLID RADI0 ACTIVE WASTE
- l
, LIMITING CONDITION FOR OPERATION 3.11.3 The solid radwaste system shall be used, as applicable, to
.
solidify and. package radioactive wastes, and to meet the requirements of 10 CFR Part 20 and of 10 CFR Part 71. '
~
APPLICABILITY: At all times. ACTION: .
- a. With the applicable requirements of 10 CFR Part 20 and 10 CFR Part 71 not satisfied, suspend affected shipments of solid radioactive wastes from the site.
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.'
SURVEILLANCE REOUIREMENTS 4.11.3.1 Prior to shipment, solidification shall be verified in accordance with Station Operating Precedures. (~} 4.11.3.2 Reports _ - The semi-annual Radioactive Effluent Release Report in Specification 6.9.1.12 shall include the following information for each type of solid waste shipped offsite during the report period:
.a. container volume,
- b. total curie quantity (determined by measurement or estimate),
- c. principal radionuclides (determined by measurement or estimate),
- d. ' fpe of waste (e.g. , spen't resin, compacted dry waste evaporatorbottoms),
-
- e. type of container (e.g., LSA, Type A, Type B, large
' Quantity),and
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- f. solidification agent (e.g., cement, urea formaldehyde).
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g BEAVER VALLEY - UNIT 1 3/4 11-16
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' h RADI0 ACTIVE EFFLUENTS 3/4.11.4 TOTAL DOSE LIMITING CONDITION FOR OPERATION , 3.11.4 The dose or dose commitment to any real individual from uranium fuel
cycle sources shall be limited to less than or equal to 25 mrem to the total
'
body or any organ (except the thyroid, which shall be limited to less than. or equal to 75 mrem) over 12 consecutive months. 4 APPLICABILITY: At all times. l ACTION:
- a. With the calculated dose from the rele'ase of radioactive materials in liquid or gaseous effluents exceeding twice the limits of Specifications 3.11.1.2.a, 3.11.1.2.b, 3.ll .2.2.a , 3.11.2.2.b, 3.11.2.3.a, or 3.11.2.3.b, in lieu of any other report required by Specification 6.9.1, prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 and limit the subsequent releases such that the dose or dose commitment to any real individual from uranium fuel cycle sources is limited to less than or equal to
,' 25 mrem to the total body or any organ (except thyroid, which is limited to less than or equal to 75 mrem) over 12 consecutive months. l (]) This Special Report shall include an analysis which demonstr'ated that radiation exposures to any real individual from uranium fuel _ cycle sources (including all effluent pathways and direct radiation) are less than the 40 CFR Part 190 Standard. Otherwise, obtain a variance ; from the Cc= mission to permit releases which exceed the '40 CFR ~ Part 190 Standard- .
- b. The provisions of Specificaticns 3.0.3 and 3.0.4 are not applicable. I SURVEILLA'4CE REQUIREMENTS
, 4.11.4 Dose Calculations 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 00CM.
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O BEAVER VALLEY - UNIT 1 3/4 11-17
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, O 3/4.11 RADI0 ACTIVE EFFLUENTS , i BASES ! 3/4.11.1 LIQUID EFFLUENTS 3/4.11.1.1 CONCENTRATION I This specification is provided to ensure that the concentration of radioactive materials released in liquid waste effluents from the site will be less than the concentration levels specified in 10 CFR Part 20, Appendix 3,
,
Table II, Column 2. This limitation provides additional assurance tnat the levels of radioactive materials in bodies of water outside the site will result in exposures within (1) the Section II.A design objectives of Appendix I, 10 CFR 50, to an individual and (2) the limits of 10 CFR 20.106(e) to the population. The concentration limit for dissolved or entrained noble gases is based upon the assumpticr.' t..-t Xe-135 is the controlling radioisotope and its MPC in air (submersion) was converted to an equivalent concentration in water using the methods described in International Commission on Radiological
,
l Protection (ICRP) Publication 2. 3/4.11.1.2 00SE This specification is provided to implement the requirements of r Sections II.A, III.A and IV.A of Appendix I, 10 CFR Part 50. The Limiting i ( Condition for Operation implements the guides set forth in Section II.A of
! Appendix I. The ACTTON statements provide the required operating flexibility j
and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive material in liquid effluents will be kept "as low as is reasonably achievable." Also, for fresh water sites with drinking water supplies which can be potentially affected by plant operations, there is reasonable assurance that the operation of the facility will not result in radionuclide concentrations in the finished drinking water that are in excess of the requirements of 40 CFR 141. The dose calculations in the 00CM implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calcu t ational procedures based on models and data such that the actual exposure of an individual through appropriate pathways is unlikely to be substantially underestimated. The equations specified in the 00CM for calculating the doses due to the actual release rates of radioactive materials in liquid effluents are consistent with the methodology provided in Regulatory Guide 1.109,
" Calculation of Annual Doses to Man frcm Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I,"
Revision 1, October 1977 and Regulatory Guide 1.113, " Estimating Aquatic 4
'
Dispersion of Effluents from Accidental and Rcutine Reactor Releases for the Purpose of Implementing Appendix I," April 10/7. This specification applies to the release of liquid effluents from each reactor at the site. For units with shared radwaste treatment systems, the liquid effluents form the shared system are proportioned among the units sharing that system. 1 b V
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, SEAVER VALLEY ' UNIT 1 8 3/4 11-1
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'd RADI0 ACTIVE EFFLUENTS
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BASES 3/4.11.1.3 LIQUID WASTE TREATMENT The OPERABILITY of the liquid radwaste treatment system ensures that this system will be available for use whenever liquid effluents require treatment prior to release to the environment. The requirement that the appropriate portions of this system be used when specified provides assurance that the releases of radioactive materials in liquid effluents will be kept "as low as is reasonably achievable." This specification implements the requirements of 10 CFR Part 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50 and the design objective given in Section II.0 of Appendix I to 10 CFR Part 50. The specified limits governing the use of appropriate portions of the liquid radwaste treatment system were specified as a suitable fraction of the dose design objectives set forth in Section II.A of Appendix 1, 10 CFR Part 50, for liquid effluents. .
.
3/4.11.1.4 LICUID HOLDUP TANKS Restricting the quantity of radicactive material contained in the specified tanks provides assurance that in the event of an uncontrolled release of the tanks' contents, the resulting concentrations would be s less than the limits of 10 CFR Part 20, Appendix 3, Tabi II, Column 2, at the nearest potable water supply and the nearest surface water supply ( }) in an unrestricted area. 3/4.11.2 GASEOUS EFFLUENTS 3/4.11.2.1 DOSE RATE This specification is provided to ensure that the dose at any time at the site boundary from gaseous effluents from all units on the site will be within the annual dose limits of 10 CFR Part 20 for unrestricted areas. The annual dose limits are the doses associated with the concentrations of 10 CFR Part 20, Appendix B, Table II, Column 1. These limits provide reasonable assurance that radioactive material discharged in gaseous effluents will not result in the exposure of an individual in an unrestricted area, either within or outside the site boundary, to annual average concentrations exceeding the limits specified in Appendix B, Table II of 10 CFR Part 20 (10 CFR Part 20.106(b)(1)). For individuals who may at times be within the site boundary, the occupancy of the individual will be sufficiently low to compensate for any increase in the atmospheric diffusion factor above that for the site boundary. The specified release rate limits restrict, at all times, the corresponding gamma and beta dose rates above background to an individual at or beyond the site boundary to less than or equal to 500 mrem / year to the total body or to less than or equal to 3000 mrem / year to the skin. These release rate limits also restrict at all times, the corresponding thyroid dose rate above background to an infant via the cow-milk-infant pathway to less than or equal to 1500 mrem / year for the nearest cow to the plant. () SEAVER VALLEI - UNIT 1 B 3/4 11-2
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O C/ RADICACTIVE EFFLUENTS
.
BASES
..
This specification applies to the release of gaseous effluents from all reactors at the site. For units with shared radwasta treatment systems, the gaseous effluents from the shared system are proportioned among the units sharing that system. FA,ll.2.2 00SE -NOBLE GASES This specification is provided to implement the requirements of Sections II.3, III.A, and IV.A of Appendix I,10 CFR part 50. The Limiting Condition for Operation implements tne guides set forth in ~ Section II.B of Appendix I. The ACTION statements provice the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radio-active material in gaseous effluents will be kept "as low as is reasonably achievable". The Surveillance Requirements implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data such that the actual exposure of an individual through the appropraite pathways is unlikely to be substantially underestimated. The dose calculations established in the 00CM for calculating the doses due to the actual release rates of radioactive noble gases in gaseous effluents will be consistent with the methodology provided in Regulatory Guide 1.109, " Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the h- Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I, " Revision 1, October,1977 and Regulatory Guide 1.111, " Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors," Revision 1, July, 1977. The 00CM equations provided for determining the air doses at the site. boundary _ .
- ~
j are ba'seFupisth'e Mito'rical average atmospheric conditions. 3/4.11.2.3 COSE - RADI0ICUINES, RADI0 ACTIVE MATERIALS IN PARTICIILATE FORM AND RADIONUCLIDES OTHER THAN NOBLE GASE5 This specification is provided to implement the requirements of Sections II.C, III.A &nd IV.A of Appendix I,10 CFR Part 50. The Limiting Conditions
;
for Operation are the guides set forth in Section II.C. of Appendix I. . The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Aopendix I to assure that the releases of radioactive materials in gaseous effluents
,!
will be kept "as low as is reasonably achievable". The 00CM calculational methods specified in the surveillance requirements implement the require-ments in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data ; such that the actual exposure of an individual through appropriate path- l ways is unlikely to be substantially underestimated. The 00CM calcula-tional methods approved by NRC for calculating the doses due to the actual release rates of the subject materials are required to be _ consistent with the methodology provided in Regulatory Guide 1.109, n* BEAVER VALLEY - UNIT 1 3 3/4 11-3
!
.
O V RADIOACTIVE EFFLUENTS BASES
" Calculating of Annual Doses to Man from Routing Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," Revision 1, October,1977, and Regulatory Guide 1.111,
" Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors,"
Revision 1, July, 1977. These equations also provide for determining the actual doses based upon the historical average atmospheric conditions. The release rate specifications for radiciodines, radioactive material in particulate form, and radionuclides other than noble gases are dependent on the existing radionuclide pathways to man, in the unrestricted area. The pathways which are examined in the development of these calculations are: 1) individual inhalation of airborne radionuclides, 2) deposition of radionuclides onto green leafy vegetation with subsequent consumption by man, 3) deposition onto grassy areas wnere milk animals and meat producing animals graze with consumption of the milk and meat by man, and 4) deposition on the ground with subsequent exposure of man. 3/4.11.2.4 GASECUS RADWASTE TREATMENT The OPERABILITY of the gaseous radwaste treatment system and the ventilation exhaust treatment system ensures that the systems will be ['] available for use whenever gaseous effluents require treatment prior to release to the environment. The requirement that the appropriate portions of these systems be used, when specified, provides reasonable assurance that the releases of radioactive material in gaseous effluents will be
' eat "as low as is reasonably achievable." This specification implements requirements of 10 CFR Part 50.36a, General Design Criterion 60 of '
Apgendix A to 10 CFR Part 50, and the design objectives given in Section~II.O of Appendix I to 10 CFR Part 50. The specified limits governing the use of appropriate protions of the sytems were specified as a suitable fraction of the dose design objectives set forth is Sections II.B and II.C of Appendix I,10 CFR Part 50, for gaseous effluents. 3/4.11.2.5 EXPLOSIVE GAS MIXTURE , This specification is provided to ensure .that the concentration of potentially explosive gas mixtures contained in the waste gas holdup svstet is maintained below the flammability limits of hydrogen and oxygen. C 3 p Maintaining the concentration of hydrogen and oxygen below their flamability 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. SEAVER VALLEY - UNIT 1 8 3/4 11-4 O
.
O v . RADIOACTIVE EFFLUENTS ' BASES 3/4.11.2.6 GAS STORAGE TANKS _ Restricting the quantity of radioactivity contained in each gas storage tank provides assurance that in the event of an uncontrolled release of the tank's contents, the resulting total body exposure to an individual at the , i nearest exclusion area boundary will not exceed 0.5 rem. This is consistent with Standard Review Plan 15.7.1, " Waste Gas System Failure." 3/4.11.3 SOLIO RADI0 ACTIVE WASTE The OPERABILITY of the solid radwaste system ensures that the system will be available for use whenever solid radwastes require processing and packaging prior to being shipped offsite. This specification implements the requirements of 10 CFR Part 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, mixing and curing times. 3/4.11.4 TOTAL DOSE i O This specification is provided to meet the dose limitations of 40 CFR 190. The specification requires the preparation and submittal of a Special Report whenever the calculated doses from plant radioactive effluents exceed twice the design objective doses of Appendix I. For sites containing up to 4 reactors, it is highly unlikely that the resultant dose to a real individual will exceed 40 CFR 190 if the individual reactors remain within the reporting requirement level. The Special Report will describe a course of action which should result in the limitation of dose to a real individual for 12 consecutive months to within the 20 CFR 190 ! limits. For the purposes of the Special Report, it may be assumed that the dose commitment to the real individual frcm other uranium fuel cycle sources is negligible, with the exception that dose contributions from other nuclear fuel cycle facilities at the same site or within a radius of 5 miles must be considered. . . O BEAVER VALLEY - UNIT 1 B 3/4 11-5 i
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1 O 5.0 CESIGN FEATURES 5.1 SITE EXCLUSION AREA 5.1.1 The exclusion area shall be as shown in Figure 5.1-1. LOW POPULATION ZONE 5.1.2 The low population zone shall be as shown in Figure 5.1-2. SITE BOUNDARY FOR GASEOUS EFFLUEtiTS
5.1.3 The site boundary for gaseous effluents shall be as shown in [ Figure 5.1-3. SITE BOUNDARY FOR LIQUID EFFLUENTS 5.1.4 The site boundary for liquid effluents shall be as shown in Figure 5.1-4. 5.2 CONTAINMENT CONFIGURATION 5.2.1 The reactor containment building is a steel lineo, reinforced concrete building of cylindrical shape, with a dome roof and having the following design features:
- a. Nominal inside diameter = 125 feet.
- b. Nominal inside height = 185 feet.
- c. Minimum thickness of concrete walls = 4.5 feet. !
\ ,
- d. Minimum thickness of concrete roof = 2.5 feet.
- e. Minimum thickness of foundation mat = 10 feet.
- f. Nominal thickness of vertical portion of steel liner = 3/8 inch.
- g. Nominal thickness of steel liner, dome portion = 1/2-inch.
6
- h. Net free volume = 1.8 x 10 cubic feet.
DESIGN PRESSURE AND TE?iPERATURE 5.2.2 The reactor cantainment cuil g is cesigned and shall be ma tained i for a maximum internal pressure of psig and a temperature of . p' ' BEAVER VALLEY - UNIT 1 5-1 l . l
, . . . - .,, m , e 7-- - -
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s s v i
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! i l i ' l 1 o
. Operations O Division
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=. vice .
% President P i O l
- General Supt. of k Power Stations.
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Tech. Asst. Nuc lear _._____________-_________3, e i e ,
. I I
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I Cirector of I Nuclear Oper. l 1
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Power Stations Superintendent Superintendent Superintendent BVP5 - 2 Superintendent Power Stations Superintendent AC518 Co. L1C. & Compt. Superintendent of Operations Superinterwtent Superintendent Pers. & Records Tech. Services riuclear Services Startup Coord.
"" #" O ' I' U ' F Superintendent nf Maintenance Cheswick Shippingtort
[lrania ' Phillips
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Office Mar.ager Health f avironmental Core Analysis Quality Control Physicist Coorstaator Engineer Supervisor l Senior Coal 2 Inspectar p 5 Core Operat6cas Quality Control' s'>r. Compliance Sr. Licensing p inginces in9tneer Engineer Ingineer N l -
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6Ff 511E OkG"a;I7 Allo.1 (P,'.NII AL )
- Fire protection responsibility
-IIGURE 6.2.1
_ _ - _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ - _ -
_ . _ , . ~.-. - __ f
'd s k
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M - R 5tation Supt. O Nuclear w b
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, Results Radittion Nuclear Chief Nucle $r Engr. Office Hanager Coordinator Control Supv. Engineer & Refueltny aupv. Nuclear
. I I I l Reactor Control Technical fluclear Station Technical Senior
- _ 5enior Engr.
Che*Ist Supv, - Nuclear Oper. Supv. Advisory feigr. Ingineer Design Control
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- l .
fluclear Shift Icch. Nuclear 5entur Ingr. Radiochemist training Supv. _ Refueling Shift Supv. Advisors
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u Huclear Station Maintenance l Chemists Ingineers Oper. foremen Supervisor
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Personnel , Records Asst. . . Radiation _ fluclear Maintenance Control ingr. Control Oper. Ingineers
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Clerks Cuaeds Nmlear Maintenance Ogwrators Ios enen
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' aJiation Control Maintenance
- Emergency Planning and Fort men iethnicians Fire Protection Responsibilities Radia tion Ie(hn* clans FACitlif ORGANilAll0N FIGURE 6.2.2
_ ______ ___-_______________ _ _ - _ _ _ _ - __
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5 ACMINISTRATIVE C0fiTROLS O 6.3 UNIT STAFF QUALIFICATIONS
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, 6.3.1 Each member of the unit staff shall meet or exceed the minimum qualifications of ANSI N18.1-1971 for comparable positions, except for the Radiation Control Supervisor who snall meet or exceed the qualificatior.s of Regulatory Guide 1.8, September, 1975. 6.4 TRAINING 6.4.1 A retraining and replacement training program for the unit staff shall be maintained under the direction of the Training Supervisor and shall meet or exceed the requirements and recommendations of Section 5.5 of ANSI N18.1-1971 and Appendix "A" of 10 CFR Part 55. 6.4.2 A training program for the Emergency Squad shall be maintained under
the direction of the Training Supervisor and shall meet or exceed the requirements of Section 27 of,the NFPA Code - 1976. . . _ , 6.5 REVIEW AND AUDIT 6.5.1 On-Site Safety Comittee (OSC) . FUNCTION l 6.5.1.1 The On-Site Safety Comittee (OSC) shall function to advise the Station Superintendent on all matters related. to nuclear safety. COMPOSITION . 6.5.1.2 The OSC shall be composed' of the: Chairman: Chief Engineer Member: Operations Supervisor Member: Radiation Control Supervisor Member: Maintenance Supervisor ! Member: Nuclear Engineering & Refueling Supervisor Member: Results Coordinator Member: Training Supervisor Member: Office flanager Nuclear (Security Officer)
!1 ember: Senior Engineer - Emergency Pla'nning and Fire Protection-Member: Technical Advisory Engineer BEAVER VALLEY - UNIT 1 6-5 ,
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-, _. _ _ _ , _ - - , - _ . _ _ . _. ,
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ADMINISTRATIVE CONTROLS ('] c. Review of all proposed changes to Appendix "A" Tecnnical - Specifications.
- d. Review of all proposed changes or modifications to unit systems or equipment that affect nuclear safety.
- e. Investigation of all violations of the Technical Specifications including the preparation and forwarding of reports covering evaluation and recommendations to prevent recurrence to the General Superintendent of Power Stations and to the Chairman of the Off-Site Review Committee.
#. Review of events requiring 24-hour written notification to the Comission.
- g. Review of unit operations to detect potential nuclear safety hazards.
- h. Performance of special reviews, investigations, or analyses and reports thereon as requested by the Station Superintendent or the Chairman of the Off-Site Review Comittee.
- i. Review of the Security Plan and implementing procedures and shall submit recorarended changes to the Chaiman of the Off-Site Review Comittee.
O j. Review of every unplanned release of radioactive material to " V the environs; evaluate the event; specify remedial action to prevent recurrence; and document the event description, evaluation, and corrective action, and the disposition of the
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corrective action in the plant records. __
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- k. Review of changes to the Process Control Program, Offsite Dose Manual, and radwaste treatment systems.
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AUTHORITY 6.5.1.7 The On-Site Safety Comittee (OSC) shall:
- a. Recommend to the Station Superintendent written apcroval or disapproval of items considered under 6.5.1.6(a) through (d) above.
- b. Render dettminations in wr%ing with regard to whether or not each item considered under 6.5.1.6(a) through (e) above constitutes an unreviewed safety question.
- c. Provide written notification within 24 hours to the General Superintendent of Power Stations and Off-Site Review Comittee of disagreement between the OSC and the Station Superintendent; however, the Station Superintendent shall have responsibility for resolution of such disagreements pursuant to 6.1.1 above.
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n v 6-7 BEAVER VALLEY - 1) NIT I i
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ADMINISTRATIVE CONTROLS
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COMPOSITION 6.5.2.2 The ORC snail be composed of the: Chairman: Director of Nuclear Operations, Power Stations Department Vice Chairman: Beaver Valley Power Station Project Manager Member: Nuclear Engineer, Mechanical Engineering Department Member: Beaver Valley Power Station Superintendent Member: Superintendent of Nuclear Services, Power Stations Department Memoer: Technical Assistant to the Vice President-Operations Member: Mechanical Engineer - Mechanical Engineering Department Member: Technical Assistant Nuclear, Pcwer Stations Department Member: Quality Assurance Manager - Quality Assurance Department
!1 ember: Superintendent of Licensing and Compliance, Power Stations Department ALTERNATES 6.5.2.3 All alternate members shall be appointed in writing by the ORC chairman to serve on a temporary basis; however, no more than two alternates shall participate as voting members in ORC activities
(~T at any one time. N- / .
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CONSULTANTS 6.5.2.4 Consultants shall be utilized as determined by the ORC Chairman to provide expert advice to the ORC. MEETING FREGUENCY 6.5.2.5 The ORC shall meet at least once per six months. QUORUM 6.5.2.6 A quorum of ORC shall consist of the Chairm 1 3r his designated alternate and at least 4 members including alternates. No more than a minority of the quorum shall have line responsibility for operation of the facility.
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() BEAVER VALLEY - UNIT 1 6-9
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__ ACf4INISTRATIVE C0f1TROLS V AUDITS 6.5.2.8 Audits of unit activities shall be performed under the cognizance of the ORC. These audits shal1 encompass:
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- a. The conformance of unit operation to provisions conta!ned within the Technical Specifications and applicable license conditions at least once per 12 months.
- b. The performance, training, and qualifications of the Facility staff at least once per 12 months.
- c. The results of actions taken to correct deficiencies occurring in unit equipment, structures, systems, or methods of operation that affect nuclear safety at least once per 6 months.
- d. The performance of activities required by the Operations Quality Assurance Program to meet the criteria of Appendix "B",10 CFR 50, at least once per 24 months.
- e. The Facility Emergency Plan and implementing procedures at least once per 24 months.
- f. The Facility Security Plan and implementing tracedures at least once per 24 months.
O -
- g. Any other area of unit operation considereo appropriate by the ORC or the General Superintendent of Power Stations Department.
1
- h. The Facility Fire Protection Program and implementing procedures at least once per 24 months.
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! 1. An independent fire protection and loss prevention program inspection and audit shall be performed annually utilizing . , ' either qualified off-site licensee personnel or an outside i fire protection firm. l
- j. An inspection and audit of the fire protection and loss prevention program shall be performed by an outside qualified fire consultant at intervals no greater than 3 years.
- k. The radiological environmental monitoring program and the l results thereof at least once per 12 months.
!
- 1. The OFFSITE 00SE CALCULATION MANUAL and implementing procedures at least once per 24 months.
- m. The PROCESS CONTROL PROGRAM and implementing procedures for solidification of radioactive wastes at least once per 24 months.
- n. The performance of activities required by the Quality Assurance Program to meet the criteria of Regulatory Guide 4.15, December 1977 at least once per 12 months.
] 6-11 BEAVER VALLEY - UNIT 1
_ _ - , n 6.7 SAFETY LD1IT VIOLATION V 6.7.1 The following actions shall be taken in the event a Safety Limit is . violated:
- a. The unit shall be placed in at least HOT STAND 8Y within one hour.
- b. The Safety Limit violation shall be reported to the Comission, the General Superintendent of Power Stations, and to the Off-Site Review Comittee Chairman or his alternate within 24 hours.
- c. A Safety Limit Violation Report shall be prepared. The report shall be reviewed by the On-Site Safety Comittee. This report shall describe (1) applicable circumstances preceding the violation, (2) effects of the violation upon facility components, systems or structures, and (3) corrective action taken to prevent recurrence.
- d. The Safety Limit Violation Report shall be submitted to the Comission, the Off-Site Review Ccmittee, and the General Superir.tendent of Power Stations within 14 days of the viciation.
6.8 PROCEDURES 6.8.1 Written procedures shall be established, implemented, and maintained covering the activities referenced below: Q V
- a. The applicable procedures recommended in Appendix "A" of Regulatory Guide 1.33, Revision 2, February, 1978.
- b. Refueling operations.
- c. Surve:11ance and test activities of safety related equipment.
- d. Security Plan implementation.
Emergency Plan implementation.
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e.
- f. Fire Protection Program implementation.
- g. PROCESS CONTROL PROGRAM implementation.
- h. OFF-SITE DOSE CALCULATION MANUAL implementation.
- i. Quality Assurance Program for effluent and environmental monitoring using the guidance in Regulatory Guide 4.15, December 1977.
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n ) t 8EAVER VALLEY - UNIT 1 6-13
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O ADMINISTRATIVE CONTROLS
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- d. Reactivity anomalies involving disagreement with the predicted value of reactivity balance under steady state conditions during power operation greater than or equal to 1". ak/k; a calculated reactivity balance indicating a ShVTDOWN MARGIN less conservative than specified in the technical specifications; short-term reactivity increases that correspond to a reactor period of less than 5 seconds or, if subcritical, an unplanned reactivity insertion of more than 0.5". ak/k; or occurrence of any unplanned criticality,
- e. Failure or malfunction of one or more components which prevents or could prevent, by itself, the fulfillment of the functional requirements of system (s) used to cope with accidents analyzed in the SAR. ,
- f. Personnel error or procedural inadequacy which prevents or could prevent, by itself, the fulfillment of the functional requirements of systems required to cope with accidents analyzed in the SAR.
- g. Conditions arising frca natural or man-made events that, as a direct result of the event require unit shutdown, operation of safety systems, or other protective measures O
U required by technical specifications.
- h. Errors discovered in the transient or accident analyses or in the methods used for such analyses as described in the safety analysis report or in the bases for the technical specifications that have or could have permitted reactor operation in a manner less conservative than assumed in the analyses.
- i. Performance of structures, systems, or components that requires remedi-; action or corrective measures to prevent operation in a manner less conservative than assumed in the accident analyses in the safety analysis report or technical specification bases; or discovery during unit life of conditions not specifically considered in the safety analysis report or technical specifica-tions that require remedial action or corrective measures to -
prevent the existence or development of an unsafe condition.
- j. Occurrence of an unusual or important event that causes a significant radiological impact, or that has high public or potential public interest concerning environmental impact from unit operation.
- k. Offsite releases of radioactive materials in liquid and gaseous effluents which exceed the limits of Specification 3.11.1.1 or 3.11.2.1.
A 1. Exceeding the limits in Specification 3.11.1.4 or 3.11.2.6 for V storage of radioactive materials in the listed tanks. The written follow up activities planned and/or taken to reduce the contents
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to within specified limits. 3EAVER VALLEY - UGIT 1 6-17
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O AosIntsTRATtve ConTRcts
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THIRTY DAY WRITTEN rep 0RTs 6.9.1. 9 The types of events listed below shall be the subject of written reports to the Director of the Regional Office within thirty days of occurrences of the event. The written report shall include, as a minimum, a completed copy of a licensee event report form. Informa-tion provided on the license'e event report forn shall be supplemented, as needed, by additional narrative material to provide complete explana-tion of the circumstances surrounding the event.
- a. Reactor protection system or engineered safety feature instrument settings which are found to be less conservative than those established by the technical specifications but which do not prevent the fulfillinent of the functional requirements of affected. systems.
- b. Conditions leading to operation in a degraded made permitted by a limiting condition for operation or plant shutdcwn required by a ifmiting condition for operation.
NOTE: Routine surveillance testing, instrument calibration, or preventive maintenance which require system l configurations as described in items a and b need not be reported except where test results themselves - O reve>> e desr>eed mode es described eeove-Observed inadequacies in the implementation of administrativo c. or procedural controls which threaten to cause reduction of degree of re'dundancy provided in reactor protection systems or engineered safety feature systems. ,
- d. Abnomal degradation of systems other.than those specified
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in 6.9.1.8 above designed to contain radioactive material resulting from the fission process.
- e. An unplanned offsite release of 1) more than 1 curie of radioactive material in liquid effluents, 2) more than 150 curies of noble gas in gaseous effluents, or 3) more than 0.05 curies of radiciodine in gaseous effluents.
The report of an unplanned offsite release of radioactive material shall include the following information: ! A description of the event and equipment involved. l
- 1. J i 2. Cause(s) for the unplanned release.
- 3. Actions taker. to prevent recurrence. 1
- 4. Consequences of the unplanned release.
- f. Measured levels of radioactivity in an environmental sampling Q medium determined to exceed the reporting level . values of
, (V Table 3.12-2 when averaged over any calendar quarter sampling pericd. S' AVER VALLEY - UNIT 1 6-18 l
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f% V ACMINISTRATIVE CCNTROLS , The reports shall also include the following: a sumary description of the radiological environmen?al monitoring program and a map of all sampling locations keyed to a table giving distances and directions from one reactor. The result of land use censuses required by the Specification 3.12.2 and the results of licensee participation in the Quality Assurance Program required by Specification 3.12.3. SEMI-ANNUAL RADI0 ACTIVE EFFLUENT RELEASE rep 0RT 3/ 6.9.1.12 Routine radioactive effluent release reports covering the operating of the unit during the previous 6 months of operation shall be submitted within 60 days after January 1 and July 1 of each year. 6.9.1.13 The radioactive effluent reTease reports shall include a sumary of the quantities of radioactive liquid and gaseous effluents and solid waste released frcm the unit as' outlined in Regulatory Guide 1.2, " Measuring, dvaluating, and Reporting Radioactivity in Solid Wastes and Releases of Rad 1oactive Materials in Liquid and Gaseous Effluents frcm Light-Water-Cooled Nuclear power Plants", with data sumarized on a quarterly basis following the format of Appendix B thereof. , e The radioactive effluent release report to be submitted 60 days after January 1 of each year shall include an annual sumary of hourly meterological data collected over the previous year. This annual summary may be either in the form on an hour-by-hour listing of wind speed, wind direction, atmospheric stability, and precipitation (if measured) on magnetic tape, or in the form of joint frequency distributions of wind speed, wind direction, and atmospheric stability. A suggested magnetic tape format is availule from the Hydrology-Meteorology Branch, Division of Site Safety and Environmental Analysis, NRR, NRC, Washington, D.C. 20555. This same report shall include an assessment of the radiation doses due to the radioactive liquid and gaseous effluents released from the unit or station during the previous calendar year. This same report shall also include an assessment of the radiation doses from liquid and gaseous radioactive effluents to individuals due to their activities inside the site boundary (Figure 5.1-1) during the report period. All assumptions used in making these assessments (e.g., specific activity, exposure time and location) shall be included in these reports. The meteorological conditions concurrent with the time of release of radioactive materials in gaseous effluents (as determined by sampling frequency and measurement) shall be used for determining the gaseous pathway doses. The assessment of radiation doses shall be performed in accordance with, the Off-Site Case Calculation Manual (00CM). .
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3/ A single submittal may be made for a multiple unit station. The submittal should ccmbine those sections that are comen to all units at the station; however, for units with separate radwaste ] systems, the submittal shall specify the releases of radioactive i l p material from each unit. d BEAVER VALLEY - UNIT 1 6-21 _
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O ADMINISTRATIVE C0t1TROLS N C a The racicactive effluent release recorts snait include the following information for each type of solid waste shipped offsite during the report period:
- a. Container volume.
- b. Total curies quantitiy (specify whether determined by measurement or estimate.
- c. Principal radionuclides (specify whether determined by measurement or estimate.
- d. Type of waste (e.g., spend resin, compacted dry waste, evaporator bottoms),
- e. Type of container (e.g., LEA, Type A, Type B, large quantity), and
- f. Solidification agent (e.g., cement, urea formaldehyde).
The radioactive effluent release reports shall include unplanned releases from the site to unrestricted areas of radioactive materia' 4 gaseous and liquid effluents on a quantity basis. The radioactive effluent release reports shall include any changes to the O'V PRf' CESS C0t4 TROL PROGRAM (PCPM) made during the reporting period, as pre ided in Specification.6.14. SPECIAL REPORTS . 6.9.2 Special reports shall be submitted to the Director of the Office of Inspection and Enforcement Regional Office within the time period specified for each report. These reports shall be submitted covering the activities identified below pursuant to the requirements of the applicable reference specification: i BEAVER VALLEY - UtlIT 1 6-22
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O V . ADMINISTRATIVE CONTROLS radiation area and entrance thereto shall be controlled by requiring issuance of a Radiation Work Pemit*. Any individual or group of individuals permitted to enter such areas shall be provided with or acccmpanied by one or more of the following: ,
- a. A radiation monitoring device which continuously indicates the radiation dose rate in the area.
- b. A radiation monitoring device which centinuously integrates the radiation dose rate in the area and alarms when a preset inte-grated dose is received. Entry into such areas with this monitoring device may be made after the dose rate level in the area has been established and personnel have been made know-ledgeable of them.
- c. An individual qualified in radiation protection procedures who is equipped with a radiation dose rate monitoring device. This individual shall be responsible for providing positive control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified by a facility Health Physics Supervisor in the Radiation Work Permit.
6.12.2 The requirements of 6.12.1, above, also apply to each high radia-p
- tion area in which the intensity of radiation is greater than 1000 mrem /hr.
LJ In addition, locked doors shall be provided to prevent unauthorized entry into such areas and the keys shall be maintained under the administrative control of the Shift Supervisor on duty and/or the Plant Health Physicist. 6.13 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM The Director of Nuclear Operations delegates the responsibility for the Radiological Environmental :lonitoring Program to the environmental , administrator (Superintendent of Licansing & Ccmpliance, (Figure 6.13.1]) or his designated alternate. The environmental administrator is responsible for administering the off-site Radiological Environmental Monitoring Program. He shall cetermine that the sampling program is being implemented as described to verify that the environment is adequately protected under existing procedures. He shall also have the responsibility for establishing, implementing, mdintaining, and approving off-site environmental program sampling, analyses, and calibration procedures.
- Health hysics personnel, or personnel escorted by Health Physics personnel is accordance with approved emergency procedures, shall be exwpt from the RWP issuance requirement during the performance of their radiation protection duties, provided they comply with approved radiation protection procedures for entry into high radiation areas.
{} BEAVER VALL U - UNIT 1 6-26
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_ _ _ - _ - _ _ . - . _ _ _ _ . .- - ex b L Ni U Genera') Supt. d
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Power Stations 5 p ___.-.--___- U {. __._________
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l Director of E I Nuclear Operation-. G l _. I ___
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Off-site Review Cosmii ttee Station Supt. Supt. of Supt. of Beaver Valley Licen. & Compl. Nuclear Services [ m
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L 1 I Environmental i Station Staff
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Coordinator i Review (Radiolojical ' Review (Ecologi al
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{0ff-SiteProgram) Off-Site Progron,i) On-Site lleal th Off-Site Environ. Core Analysis Environ. Program Physicist Program Engineer (E f fluents) FIGURE 6.13.1 - Duquesne Light Company Power Statioris Departisient Partial Organization Chart Relative to Environmental Matters t _ _ _ _ _ _ - . _ _ . _ _ . _ _ _ _ _ _ - - - _ _ - -
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O -
- ADMINISTRATIVE CONTROLS 6.14 PROCESS CONTROL PROGRAM (PCP)
FUNCTION 6.14.1 The PCP shall be a manual containing the processing steps and a set of established process carameters detailing the program of sampling,
, analysis, and evaluation within which solidification of radioactive wastes from liquid systems is assured, consistent with Specification 3.11.3.1 and the surveillance requirements of these Technical Specifications.
6.14.2 Licensee initiated changes:
- 1. Shall be submitted to the Ccemission by inclusion in the semi-annual Radioactive Effluent Release Report for the period in which the change (s) was made.and g shall contain:
- a. Sufficiently detailed information to totally support the rationale for the change without benefit of additional or supplemental informa-tion; o b. a determination that the change did not reduce U the overall conformance of the solidified waste product to existing criteria for solid wastes; and 1
- c. d'ocumentation of the fact that the changes has been reviewed and found acceptable by the OSC.
- 2. Shall become effective upon review and acceptance by the ORC. .
6.15 0FFSITE DOSE CALCULATION MANUAL (00CM) FUNCTION [6.15.1 The 00CM shall describe the methodology and parameters to be used in the calculatier. of offsite doses due to radioactive gaseous and liquid effluents and in the calculation of gaseous and liquid effluent confioring instrumentation alarm / trip setpoints consistent with the applicable LCO's contained in these Technical Specifications. Methodologies and calcula-(tional procedures acceptable to the Commission are contained in NUREG-0133.
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(V SEAVER VALLEY - UNIT 1 6-27
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, .,e.
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O , 5.0 ACMINISTRATIVE CONTROLS Objective To define the organization, assign responsibilities, describe the environmental surveillance procedures, provide for a review and audit function, and prescribe the reporting requirements in order to insure '
, continuing protection of the environment and implement the Environmental Technical Specifications.
5.1 ORGANIZATION Figure 5.1-1 is a partial organization chart of the Power Stations Department which includes groups and individuals responsible of the Environmental Te ,hnical Specifications following the issuance of l an operating license for Seaver Valley Unit No. 1. ! 5.2 RESPONSIBILITY All activities within the Power Stations Department are the responsibility of the General Superintendent. He delegates the responsibility for ecological environmental monitoring programs to the Superintendent of Licensing and Ccmpliance. He is advised by the Off-Site Review Ccmmittee in matters relating to the environmental impact of station operation
*aica 4"vo've aroaosed cae"ses to tae e"v'romme"te' teca"'ce' saec4'icetio">
O proposed changes to written procedures, proposed changes or modifications to station systems or equipment, evaluations of investigations conducted as a result of environmental technical specification violations, and when
- appropriate, recommendations to prevent recurrence of such violations.
The Station Superintendent is the individual charged with the responsibility for assuring that all releases to the environment shall be limited in accordance with the Limiting Conditions for Operations as described in the Environmental Technical Specifications. Controlled releases are made utilizing written procedures under the direction of the Shift Supervisor at the request of the Reactor Control Chemist (for non-radiological discharges) or the Radiation Control Supervisor (for discharges requiring a radiological Escharge permit) who specifies the
-release rate, the station operating condtiens, and the monitoring requirements which must be maintained during the release to ccmply with the technical specifications. These activities are reviewed by the Radiochemist and the Station Operating Supervisor and audited periodically under the direction of the Superintendent of Licensing and Ccmpliance, Power Stations Department.
The Station Suoerintendent is also responsible for the issuance of the 24-hour reporting requirement to the Director of the fagional Regulatory Operations Office, wnen required. O 5-1
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___ _ _ _ _ _ _ _ . _ __ . _ _ _ _ O O O General Supt. Power Stations (------------------ 1 I Director of i Nuclear Operation s . I i
' Off-site Review Comnittee
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.
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Station Supt. Supt. of Supt. of Beaver Valley - Licen. & Compl . Nuclear Services
<p N .
- l 8
- I I -
"Y "'"*"" '
Station Staff Coordinator i
- Review (Ecologi al Of !f Site Prograsi) 1 On-Site .
Off-Site Environ. Core Analysis Environ. Program Program Engineer.
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(E f fluents)
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FIGURE 5 1-1 - Duquesne Light Company
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- Power Stations Department Partial Organization Chart Itelative to Environmental
' Matters
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- . _ . _ . _ . _ . . _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _______ ____ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ___ _
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O s.3 REvrEw tr<o euotr . The Superintendent of Licensing and Complie ce is responsible to the Off-Site Review Ccmmittee for reviewing the results of the Environmental Surveillance Program to determine that the sampling program is being implemented as described to verify that the environment is adequately protected under the existing station operating procedures. He is also responsible for requesting that analyses be repeated or that new sampic s be obtained from particular locations in the case of anomalous result!. I The Superintendent of Licensing and Compliance is responsible to the , Off-Site Review Committee for scheduling audits utilizing various company i personnel to verify that station operations including releases to the environment are conducted in such a manner that continuing protection to the environment is provided. The Superintendent of Licensing and Compliance
shall also have the responsibility for reviewing off-site environmental program sampling, analyses and calibration procedures as summarized in Table 3.1-3 and desJibed in Sections 3.0 and 5.5.1 of these specifications. The Beaver Valley Power Station Superintendent is responsible for reviewing on-site procedures relative to the environment as described in Section 2.0 and Section 5.5.1 of these specifications. The Off-Site Review Comittee functions as they relate to the Environmental Technical Specifications are as follows:
- a. Reviews the reports of the Superintendent of Licensing and Compliance on the results of the environmental monitoring O arcere= Pr4or to t8eir seemittei to the riac 4a eeca ^#auei Non-Radiological Environmental Report. See Section 5.6.1.
I
- b. Review and make recommendations on proposed changes to the Environmental Technical Specifications and the evaluated impact of the changes.
;
- c. Review proposed changes or modifications to plant systems or equipment and the evaluated impact which would require a change in the procedures described in (d) below.
- d. Review reported instances of violations of Environmental
. Technical Specifications and abnormal environmental occurrences. Where investigation indicates, evaluate and formulate recommendations to prevent recurrence.
- e. Schedule audits of Station Operation with regard to the control of releases to the environment.
- f. Review and compare the Safety Technical Soecifications and the Environmental Technical Specifications to avoid conflicts and maintain consistency.
hN 5-3 l 1 1
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ACTICH TO BE TAKE.l IF A LIMITING CONDITICN FOR OPERATION IS EXCEEDED O 5.4.1 Remedial Action
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Imediate remedial actions as permitted by these technical specifications will be implemented until such time as the limiting condition for operation is met. 5.4.2 , Investigation The Station Superintendent shall direct that an investigation be conducted to determine the cause of the occurrence and recommend both temporary and permanent corrective action to prevent a recurrence. The results of this investigation shall be reviewed and approved by the Off-Site Review Committee and the General Superintendent, Power Stations Department. 3.4.3 Report The occurrence shall be reported to the NRC as specified in Section 5.5.2. 5.5 PROCEDURES 5.5.1 Written Procedures Detailed written procedures, including check lists and special instructions when required, will be prepared and implemented for all activities involving releases to the environment, described
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O 4# sect 4oos 2 o e#4 3 o. the semai4#9 e#d mo#4 tor 4#9 or taese releases and the conduct of the Environmental Surveillance Program.
'
These procedures will include sampling, instrument calibration, analysis, and actions to be taken when limits are approached or exceeded. Testing and calibration frequencies of effluent alarms and monitoring instruments will be included. These frequencies will be determined from experience with similar instruments in similar environments and from manufacturers' recomendations. 5.5.2 Standard Ooeratina Precedures Plant standard operating procedures will include, when required, sufficient precaution limitations, and setpoints to ensure that all plant systems and components are operated in ccepliance with the limiting conditions for coerations established in Section 2.0 of these technical specifications. 5.5.3 Reviews
-All procedures described above, and all changes thereto, will be reviewed prior to implementation and periodically thereafter as described in Section 5.3. Temporary changes to procedures which do not change.the intent of the original procedure may be made without the need for review by NRC. Changes to procedures for programs in Section 3 will be reviewed by the NRC. Such changes will be documented.
i
/"i i 5-4 !
l L
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- DUQUESNE LIGHT COMPANY Beaver Valley Power Station, Unit No.1 O Occket No. 50-334 License No. OPR-66
!
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l i Attachment C Addition of Safety-Related Snubbers O O
.
--- .- - - , .
- - - -
- - , - - - , - - -
O O O
.
TABLE 3.7-4 (Continued)
- t. ,
SAFETY RELATED llYDRAULIC SNUBBERS *
. 'S ESPECI ALLY Dif flCllt.T SYSTEM S!!UBDER INSTALLED ACCESSIBt.E OR li1Gil RADIATION ,~.1 Silitl'0ER ON, LOCAT10ll AND ELEVATI0ll lilACCESSIBLE 7ONE TO RElflVC ] __ fl0 .
739' I Yes Yes RC-IISS-101 RC RCP Cub. " A " " "
"
'O RC-IISS-102
"
739' " " "
. SI-ilSS-102A SI " "
745' " "
"
c: 51-1155-1020 "
" "
745' " " "
"
33 51-1155-414
" "
741' " "
" -'
0C-1855-103 RC RCP Cub. B' 739' " " "
"
RC-IISS-104
" " "
739' " " "
"
S1-1155-410 51
"
741' llo 'No ttC-IISS-23 RC Reac. Cnt. Bldg. 749' A I Yes Yes liC-IISS-105 RC RCP Cub. C 739' " " " RC-IISS-106 RC
"
.
"
739' _
" "
739' I " RC-IISS-130 "RC RCP Cub.
"
C I
"
ItC-IISS-131
" 739
" "
"
SI-ilSS-114A 51
" "
745' " " " S1-1155-1140 "
" "
745' " "
"
M S1-1155-422
" " " 741' " "
* -"
S1-1155-423
" " "
739' 784' A Yes Yes Y RC-IISS-22 RC Pressurtzer Cub. " " M RC-IISS-41 A
" "
784' A
" "
ItC-IISS-44 A
" "
784' A' " ItS-IISS-235 RS Reac. Cnt. Bldg. 832' I No
" "
"
115 1155-210 "
" " 817' " "
"
ItS.IISS.211
" " "
817' " "
"
115 1155-212
" " " 017' " "
"
I!S.IISS-228
" " " 817' " " "
I:S.1155-213
" " 800' " " "
ItS .itSS-214
" " " !!00' "
,
"
"
ItS-IISS-222
" " " 814' " " "
" " " 014 '
Its.llSS-223 " " " ' 115 1155-224
" " " 813' " " "
ItS.ilSs-225 " " " 013' No !!o lif t'0-IISS-201FW
" " -
780' A
.
. - - _
O o O TABLE 3.7-. (Continued) . SAFETY RELATED IIYDRAULIC SNUDDERS* lilGli RADIAT10tl ESPECI ALLY OlfflCill.T SYSTEM SNUBBER INSTALLED ACCESSIBLE OR ' c' T. Nill: DER ZONE TO REMOVE Otl, LOCATION AND ELEVATION INACCESSIBLE _ _ 7j Il0. tet Yes fio ItC-IISS-Il9 RC Reac. Cnt. Bldg. 7 34 A
-
N No No l 728' A y S1-115S-337 SI Reac. Cnt. Bldg.
" " 729'
" No
"
p, SI-IISS-409 SI " " "
" " "
'-
$1-185S-410 7 31 ' " "
"
,
'
'
SI-ilSS-411
" "
'.' 731' "
" " 731'
" Yes "
& RS-IISS-201 RS " " "
" "
g itS-IISS-202 7 31 ' . "
" " " 731'
"- No "
Its -1155-237 " " IIS-IISS-238
" -
" " 731' " "
"
115-1155-22 9
" " " 731' " " ~
" " " 7 31 ' "
RS IISS-236 " " ltS.-IISS-234
" " " 726' Yes
" " 711' I Yes "
CC-IISS-406A CC " " CC-1155-406B "
" " 711' " " "
" " 707_' "
w CC-IISS-40SA " " " CC-IISS-40Sil "
" " 707' "
i " " 711'
" "
"
y CC-IISS-407A " " " CC-IISS-4078 "
" " 711' llo L3 No
" " 702' A ll R5-ilSS-205 RS " " " 702'
" " "
"
RS-IISS-206 .
" "
115-1155-219
" " " 702' " " "
ItS-IISS-220
" " " 7 02' " "
"
ItS-IISS-207
" " " 702' " "
"
115-1155 -2011
" " " 702' " " "
RS-IISS-209
" " " 710' " " "
115 -1155 -2 1 5
" " " 715' " " "
itS-IISS-216
" " " 715' I Yes .Yes
, > 1111 -11 5 S - 1 0 S Ril
" " 704' " "
' "
!t11-1155-107
" " " 704' " " "
[UI-ilSS-108
" " " 704' " "
Illi-ilS S-I l l
" " " 704'
.
.
- . _ _ - .
O O O . TABLE 3.7-4_ (Continued)
.
SAFETY RELATED llYDRAULIC SHUBBERS*
'S '
ESPECI ALLY DIFFICUI.T ACCESSIBLE OR ilIGli RADIATI0fl SYSTEli StIUBBER INSTALLED TO Rett 0VE [b SillfuSER INACCESSIBLE ZONE __ Ott,,LOCAT10ll A110 ELEVAT10tl
,j 11 0 .
No ria 728' A "
' l, t-lit-ilSS-3048 "Rtl Cabic Vault "
"
'#
111t-1155-316
" " 733' "
"
"
', Illt-ilSS-307
" " " 733' " "
"
" " 733' "
"
H 111(-11S 5.- 3 0 6 . "
" " " 731'
" "
*i
- 111t-1155-308
" " " 7 31 *
" "
"
"
-, 111t-1155-30') "
tilt-ilSS-300
" " " 724' "
"
"
S1-1155-522 SI
" " 731' " "
"
"
" " 7 31 ' "
SI-ilSS-523A " " S1-115S-5233 ""
" " 731' " "
"
" " 7 31 ' "
"
SI-ilSS-521 "
" " 731' "
"
SI-ilSS-516A " "
" " 731 ' "
"
w SI-ilSS-516B "" " "
"
"
7 31
- 1 S1-1155-520
" " " 731'
"
"
"
"
- s. 51-1155-515 "
- 2. S 1-1155-51 9
" " " 731' " "
"
" " " 7 31 ' "
E S1-1155-514 " "
, 51-1855-512 "
* " 7 3fl' " "
"
SI-ilSS-512A "" .
" " 734' " "
"
" " 730' "
"
S I -ItSS-511 "
" " " 733' "
"
S t -ilSS'-518 "
" " " 73fl' "
"
SI-liSS-517 741'
"
"
05-1155-504 .QS Safeguards Area " "
" " " 741' " "
"
05-1155-503 " -"
"
7 37 ' "
" "
45-115S-502 " " 737' "
, QS-!!SS-50)
SI-IISS-211
"
51.
" "
7 34'. ". " No
-
No 743' A SI-ilSS-002 SI Safeguards' Area A No No. SI-ilSS-003 SI Safeguards Area 743' No A No 51-1155-009 SI Safeguards Area 743' No A No SI-ilSS-010 SI Safequards Area 743' "
" "
" "
735' a e QS-ilSS-205A QS " " -" 73S'
"
*
"
0S-1155-20 % "" " " 737'
'
QS-IISS-; G
'
_ _ _ _ __ _ _ _ _ _ . . _ _ _ _ _ _ _ _ _ , I ! DUQUEstlE LIGHT COMPAtlY Beaver Valley Power Station, Unit flo.1 O Docket flo. 50-334 License No. DPR-66 i I
\
1 i P.tachment D Revised Surveillance Requirements For AFW System I l O
. .
, l l O
. . . .. . . . .. --. - . - _ .
l h) v PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (continued)
- 3. Verifying that each pump operates for at least 15 minutes.
- 4. Cycling cach testable power operated or automatic valve in the flow path through at least one complete cycle of full travel.
- 5. Verifying that each valve (manual, power operated or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.
- 6. Reverifying the requirements of Tech. Spec. surveillance 4.7.1.2.a.5 by a second and independent means.
- 7. Establish and maintain constant communications between the control room and the auxiliary feed pump room while any tormal discharge valve is closed during surveillance testing.
- 8. Verifying operability of each River Water auxiliary supply valve
/ by cycling each manual River Water to Auxiliary Feedwater J k System valve through one complete cycle.
- 9. Following an extended plant outage verify Auxiliary Feedwater Flow from WT-TK-10 to the Steam Generators.
- b. At least once per 18 months during shutdown by:
j
- 1. lycling each power operated (excluding automatic) valve in the flow path that is not testable during plant operation, through at least one complete cycle of full travel.
^
- 2. Verifying that each automatic valve in the flow path actuates to its correct position on a test signal.
- 3. Verifying that each pump starts automatically upon receipt of a test signal.
, O stivtx v^tter - uutt 1 3/4 7-6
.
. ___ __ __ __ _ _ __ _ _ _ _ _
!
,
I ! i DUQUESNE LIGHT COMPANY Beaver Valley Power Station, Unit fio.1 O Docket No. 50-334 License flo. OPR-66
,
Attachment E AP4 Flow Rate Indication at Remote Shutdown Panel O O
;
y,_ .. - .. . .__ _, _ , - - _ _ _ _ . . . - _ _ . _ . . . , _ - - . . - . _., e- - _ , _ -. - _ ...,__;,_ .
-- -. -- . .-- - - .-
O O O TABLE 4.3-6 , REMOTE SHUTDOWN MONITORING INSTRUMENTATION { SURVEILLANCE REQUIREMENTS 5 CHANNEL CHANNEL l~ INSTRUMENT CllECK CALIBRATION , E
. 1. Intermediate Range Nuclear Flux M N.A.
- 2. Intermediate Range Startup Rate M N.A.
"
- 3. Source 'ange Nuclear Flux M N.A.
- 4. Source E v' Startup Rate M N.A.
,
- 5. Reactor Coolant Temperature - Hot Leg M R s
[ 6. Reactor Coolant Temperature - Cold Leg M R b 7. Pressurizer Pressure M R
- 8. Pressurizer Level M R
- 9. Steam Generator Pressure h R i 10. Steam Generator Level M R
- 11. RilR Temperature - IlX Outlet M R
- 12. Auxiliary Feedwater Flow Rate S/U (2) R l
Notation (1) If not performed in previous 7 days. (2) Channel check to be performed in conjunction with Surveillance Requirement 4.7.1.2.a.9 following an extended plant outage. I
_ _ l
!
t i
! DUQUESNE LIGHT COMPANY i Beaver Valley Power Station, Unit No.1 '
Docket No. 50-334 License No. DPR-66 I I < i , Attachment F ! Installation of Control Room Chlorine Detection
O
I l i 1 l O '
>
l
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INDEX LIMITIflG CONDITI0fiS FOR OPERATION AND SURVEILLAtlCE REQUIREMENTS SECTION Page 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 Axi al Fl ux Di f fe rence. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 2-1 3/4.2.2 Hea t Flux Hot Channel Facto r. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 2-5 3/4.2.3 Nuclear Enthalpy Hot Channel Factor. . . . . . . . . . . . . . . . . . . . . . . 3/4 2-8 3/4.2.4 Quadrant Power Til t Ratio. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 2-10 3/4.2.5 D:18 Parameters............................................ 3/4 2-12 3/4.3 INSTRUMENTATION 3/4.3.1 PROTECTIVE INSTRUMENTATION................................ 3/4 3-1 3/4.3.2 ENGINEERED SAFETY FEATURE INSTRUMENTATION................. 3/4 3-14 3/4.3.3 MONITORIflG INSTRUMENTATION O aedie ti on Mon 4 to ri n2. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . u4 3-n Movable Incore Detectors.................................. 3/4 3-37 Seismic Instrumentation................................... 3/4 3-38 Meteorological Ins trumenta tion. . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 3.-41 Remote Shutdown Instrumentation........................... 3/4 3-44 Fire Detection Instrumentation............................ 3/4 3-47 Chl ori ne Detecti on Sys tem. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 3-49 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS Normal 0peration.......................................... 3/4 4-1 Isolated Loop............................................. 3/4 4-3 Isolated Loop Startup..................................... 3/4 4-4 O V BEAVER VALLEY - UNIT 1 iv 1
)
,
__ _ O tusrau"Eur4 Tron CHLORINE DETECTION SYSTEMS LIMITING CONDITION FOR OPERATION 3.3.3.7 Three independent chlorine detection systems, with their alarm / trip setpoints adjusted to actuate at a chlorine concentration of less than or equal to 5 ppm, shall be OPERABLE. APPLICABILITY: MODES 1, 2, 3 and 4. ACTION:
- a. With one chlorine detection system inoperable, operation may contirue provided the inoperable detector is placed in the tripped condition within 1 hour. .
- b. With two chlorine detection systems inoperable, restore one of the inoperable detection systems to OPERABLE status within 7 days, or within the next 6 hours, initiate and maintain operation of the con-trol room emergency ventilation system in the recirculation mode of operation.
- c. With no chlorine detection system OPERABLE, within I hour initiate and maintain operation of the control room emergency ventilation system in the recirculation mede of operation.
- d. The provisions of Specification 3.0.4 are not applicable.
.
SURVEILLANCE REQUIREMENTS 4.3.3.7 Each chlorine detection system shall be demonstrated OPERABLE by performance of a CHANNEL FUNCTION TEST at least once per 31 days and a CHANNEL CALIBRATION at least once per 18 months. O BE^vEa v>LLEv - UNIT i 3/4 3-49 i
.. _ _ _ _ _ _ .
. .
G ,: 3w ,p. non i i
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[v'ijb Mu ud!
; l coaeoaArioN .
/=/p yo. 27x- 7 5' 7/ -+ l A QFK A)o'.' il8 92 Dc.?- /Cfa
.
,
'
INSTALLATION, OPERATION 8 MAINTENANCE > INSTRU.CTIONS CHLORINE DETECTOR SERIES 50-125 PANEL MOUNTED (SHOCK-RESISTANT)
OO
-
.
f
.
..
*
.
h i Alo' S1 I EJ L
-
DIVISION Oo ' 25 MAIN STREET.BELLEVILLE NEW JERSEY 07109 000x no.was so.ias i,_77
-
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- 1 Stressing dependability, this unit monitors for chlorine gas h
*
~. with its sensor in continuous contact with ambient air. De.
<**
'. CHLOR:ng CETscion sign simplicity and positive air sampling malte it capable
-
' * ' of detecting in seconds at 1 ppm. This level corresponds S
- to OSHA regulations and AWWA guidelines concernmg esposure to chlortne.
Q .
,
FEATURES
'D. G FIRST NON. INSTRUMENT. TYPE DETECTOR
[ N
-
(g Design simplicity removes this unit from the class of so-
.
a* phisticated instrumentation. It is the nrst truly uncom-
!n '
O plicated chlorine detector... easy to understand, operate. 4 and maintain,
%
d, ' , .~ # 7 DEPENDABLE y.
- p"' .. g.:@pj, t
e
..
A high capacity, integral fan provides positive air sam. pling. The measuring electrode is continuously cleaned by gravity Sow of the electrolyte.
.,
' LOW IN COST, LOW COST IN OPERATION 9,,E " $f[. >Nt
*
. k.-p[e Design simplicity means low initial cost' there is no light.
[,, ' g "g sensing system: few moving parts. Takes only 4 or of
. m
','-
'E P
.
I- p}1[
.-
electrolyte (glycerin.hs ad pota:sium iodide solution)
, - ' L3 'N every 3 4 .veeks and V2 oz of activated 61:er carbon after f,5
-
a chlorine leak. M. .
.
,
CHolCE OF MODELS, EASY TO INSTALL O[D 6 ,., - Installation requires only mounting on a wall with bracket ~ l'
& supplied and connecting !!5. volt power to a termtnal strip.
Two optional models for remote sam'pling up to 80 feet: one is in a WkT Chlorination module (free-standing cab-inet); ths other is mounted on a panel. Both have a high capacity blower junction box and hose connections for
,
sample. air inlet and vent. The panel-mounted .nodel can i
;
have an optional, audible alarm. i I EASY TO OPERATE r Requires only two periodic visual checks: electrolyte sup-
;
ply as shown by a red level indicator and electrolyte Sow i
; as shown by a wet electrode. Reservoir is nited with the t
detector in place. No samrte-air adjustments. After a
5
.
y' M
'
power failure, the alarm resets if no chlorine is present. P After an alarm. it can be reset only when chlortne is no
!. .-
L Ionger present. [
! LOW MAINTENANCE Materials are chemical-resistant plastics and alloys. The
{' electronics compartment and connections are air-tight.
'
W&T Chion.ne Detecto' Solid state components are on a quality printed. circuit
,
mounted m a fiber glass board. Operation of the alarm and circuitry is easily
; chlorinotor module for checked by placing a drop of chlorine bleach on the
=
remote ,nstanotion. ,3,eggag,,
'
! JCS 63999 m .- . .- . c; . .
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/ +_ TECHNICAL DATA
- *
- temperature limits 35 to 125 F.
.
. . .
-
-
.- A _d _
,
"
O DESIGN AND OPERATION [[e"i indication by red alarm light: relay contacts pro-Nd'd I* ' "'** I U
.
"
The WhT Chlorine Detector consists of an electrolyte tank also a powWn lig d** and *an* alam-reset "d # *' 'N Mon*".' ** with a level indicator and an air Alter. The activated. carbon Siter keeps chlorine gas away from the electrolyte. relay contacts Frorr tne bottom of this tank a sensor projects down mro *wo pairs rated 3 ampares (resistive load),250 volts ac:
, a sensing chamber where it contacts sample air driven one pair for equipment normally operated with an open by the fan or blower, circuit, the other pair for closed circuit equipm nt.
The sensor is a plastic holder containing two platinum electric I reWrements Wall mounted detector: 115-volt. 50/60 Hr. s,mgle.phare, electrodes. Electrolyte drains slowly down the holder 0.5 smpere; module mounted detector: 115-volt, 60 Hr. keeping it constantly wet and continuously washing ort , single-phase 0.75 ampere. dirt and contaminants. Excess solution drops into a tray 1 some of it evaporates and the remainder drains through electrical connections plastic tubing. Airtight ,'s-inch condwt connections for customer-furnished 8 P""*" " * " * *** * * * ' When chlortne-laden air enters the sensor chamber / chlo. rine reacts with the electrolyte at the electrodes to produce installation Wall mounted with brackets supplied. 51ust be approxi. an electriest current. The current is amplined in the solid state electronic unit to light a built in red alarm and de. mately 12" above door for proper sampling of the arnient air. But for convenient servicing. may be mounted higher energize a double pole, double-throw relay. The relay with optional package consisting of 4 ft of 1" PVC pipe contacts are wired to a terminal strip to permit pick-up of and locknut 6ttings (maximum length for 1" sample pipe a contact opening or closure for operating fans. chlorine shut off valves, or external alarms. Contact rating is S l8, M,), ,,,,,;,g amperes at 250 volts ac. Optional model has the detector in a WLT Chlorination As well as the alarm light.,an alarm-reset button and an Stodule (free. standing Ster glass cabinet). Another option amber power-on light are meluded. The latter indicates has the detector and blower on a panel for convenient when the unit is operating. Upon power interruptien, the wall mounting. Blower units in both models have con-amber light goes out but the relay is de-energized to the nections for customer-furnish *d 1-inch pipe for sample alarm state. When power is restored, the amber light intet and vent, comes on, the alarm relay resets automatically and will alarm if a leak occurred during the power interruption.. blower capacity
- 13 'f" *ith '" 5"**i " * "d
- P P'**i"*
- e SHORT DESCRIPTION S0 feet of suction and 20 ft of discharge pipe (1.,y I cfm with or hose).
The Wallace & Tiernan Series 50-125 Chlorine Detector standard accessories operates amperometrically. Its sensor is continuously wet-pd ted with electrolyte solution and is in continuous contact with f an-driven sample air. It is specide for chlorine gas, 4 ft. drain tubing; about 2 years'si.oply of electrolyte; felt
'
wicking for the electrode holder: two 4-os packages of The detector consists of ar' electrolyte tank. a sensor. sensor activated carbon for recharging the air 51ter; bracket for chamber electronic unit. and mounting hardware.The tang wall mounting. has an activated carbon air 61ter and a level indicator. optionci accessories The sensor has two platinum electrodes which detect chlor. 1" vinyl hose or 1" PVC pipe for sample inlet and vent: ine gas in seconds at 1 ppm (by volume) m sample air. 4 or bottles (3-4 weeks' supply) and 1-gallon containers The electronics compartment and connections are air-tight. (approx. 2 years' supply) of electrolyte; remote WhT There is a red alarm light. an amter power-on !tsht and Central Alarm (red alarm light and buz:er); remote WkT an alarm reset beton en the front of the detector. The solid Individus2 Alarm (red alarm light and green light to state electronic unit has a printed circuit board. a current indicate the detector and remote alarm are in use); auto-ampliner, a double-pole double throw relay, and a terminal matic chlorine-line shut. ort valves. And for the pane!- mounted detector, audible alarm with alarm acknewledge-strrp containing two pairs of relay contacts. These permit ment button, pick:up of a contact opening or c:ntact closure for opera-tien of external alarms or other equipment. The contacts equipment furnished are rated at S amperes. 250 volts ac. The detector can be Items su.:h as external tubing, piping, and wiring, conduit. furnished: by itself for wall mounting; in a free-standing and elective features are included only as specideally modular cabinet for remote installation: on a panel for listed in a quotation. remote installation. The wall-mounted model has an integ* overali dimensions ra! fan; other models have separate blower units. Detector on wall with bracket. 9%"x1S%"x9F "; detector in cabinet. 20%"x5'S%"x15", detector on panel, 24"x 2 4"x 10 %". TECHNICAL DATA weight and shipping weight .
**"* 'Y Detector only.12 lb and 32 lb: detector in cabinet. 70 lb Detects m seconds at 1 ppm chlorine by volume (3 mg/
.
and 120 lb: detector on panet 35 lb and 55 lb. m4 ) in air. SERVICE & REFERENCES elec,roly,e Dilute glycerin based potassium iodide solution. 4 oz of Prom;t service on Wallace k Tiernan equipment is avail-this concentrate mixed with 4 quarts of disti!!ed water d!!s able from branch of5ces in principal cities. Pubitcatiens en the reservoir (3 4 weeks' supply). One gallon plastic con. chlortnators and other related equipment are availat:e en tainer (about 2 years's:.pply) supphed. f'4u'5L O r,s
- Progressive changes iridesign may be made without prior announcement.
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_ -. . __.. $ DUQUESNE LIGHT COMPAfiY Beaver Valley Power Station, Unit No.1 O' Docket No. 50-334 License No. DPR-66 l Attachment G Containment Liner Weld Channels and Plugs Integrity 1 l O I
,
i l 1 1 ' O l l l ! l l
_ _ _ _ _ - _ _ - _ . CONTAINMENT SYSTEMS CONTAINMENT STRUCTURAL INTEGRITY LIMITING CONDITIONS FOR OPERATION 3.6.1.6 The structural integrity of the containment shall be maintained at a level consistent with the scceptance criteria in Specification 4.6.1.6.1. APPLICABLITY: MODES 1, 2, 3 and 4. ACTION: With the structural integrity of the containment not conforming to the above reguirements, restore the structural integrity to within the limits prior to increasing the Reactor Coolant System temperature above 200*F. SURVEILLANCE REQUIREMENTS 4.6.1.6.1 Liner Plate and Concre_t_e, e The structural integrity of the (~T, containment liner plate and concrete shall be determined during the shutdown v for each Type A containment leakage rate test (reference Specification 4.6.i.2) by:
- a. a visual inspegtion of the accessible surfaces and verifying no apparent changes in appearance or other abnormal degradation.
- b. a visual inspection of accessible containment liner test channels prior to each Type A containment leakage rate test. Any contain-ment liner test channel which is found to be damaged to the extent that channel integrity is -impaired or which is discovered with a vent plug removed, shall be removed and a protective coating shall be applied to the liner in that area.
- c. a visual inspection of the dome area prior to each Type A contain-ment leakage rate test to insure tne integrity of the protective
' coating. If a loss of integrity of the protective coating is observed, any vent plug to a test channel which may be in the area ; where the protective coating has failed shall be seal welded and j then the protective coating shall be repaired. , ex
'
(f BEAVER VALLEY - UNIT-1 3/4-6-10 i 1
- -
i i () SURVEILLANCE REQUIREMENTS (continued) 4.6.1.6.2 Reoorts An initial report of any abnormal degradation of the containment structure detected during the above required tests and inspections
! shall be made within 10 days after completion of the surveillance requirements
, of this s;..cification, and the detailed report shall be submitted pursuant to ; i Specification 6.9.1 within 90 days after completion. This report shall include a description of the condition of the liner plate and concrete, the inspection procedure, the tolerances on cracking and the corrective actions taken.
! ,
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-(]) BEAVER VALLEY - UNIT 1 3/4 6-10a
- - - .- - - - -.
-- - _ _ _ _ _ _ . _ _ .
c DUQUESNE LIGHT CCMPANY Beaver Valley Power Station, Unit No.1 ' O Docket No. 50-334 License No. OPR-66 l Attachment H Installation of Station Service Bus Undervoltage Relays
'
O O _
. _-. . ..
O. O O TABLE 3.3-3 (Continued) ENGINEERED SAFET's FEATURE ACTUATION SYSTEM INSTRUMENTATION MINIMUM h TOTAL NO. CilANNELS CilANNELS APPLICABLE ,1 N FUNCTIONAL UNIT OF CilANNELS TO TRIP 0PERABLE_ _ MODES ACTION _ OR, COINCIDENT WITil .
##
Steam Line Pressure- 1, 2, 3 c- Low 5 1 pressure -i Three loops 1 pressure / 2 pressures 14 ~ Operating loop any loops any 2 loops Two loops 1 pressure / l### pressure 1 pressure 15 Operating operating in any oper- any operating loop ating loop loop R#*
- 5. TURBINE TRIP &
Y _., FEEDWATER ISOLATION
- a. Steam Generator 3/ loop 2/ loop in 2/ loop in 1, 2, 3 14 Water Level - any oper- each oper-liigh-liigh ating loop. ating loop
- 6. LOSS OF POWER
- a. 4 kv Hus 1/4kv Bus 1/4kv Bus 1/4kv Bus 1, 2, 3 33*
Loss of Voltage
- b. Grid Degraded Voltage 2/4kv Bus 2/ Bus 2/ Bus on 1, 2, 3 34*
(4kv Bus) both 4kv Bus and 480v Bus
- c. Grid Degraded Voltage 2/480v Bus 2/ Bus 2/ Bus 1, 2, 3 34*
(480v Bus) on botts 4kv - Bus and 480v Bus
_ _ _ - __ -- - 1 4 PApt3.3-3(Continued) (] TABLE NOTATION
# Trip function may be bypassed in this MODE below P-11.
## Trip function may be byp ssed in this MODE below P-12.
### The channel (s) associated with the protective functions derived from the out of service Reactor Coolant Loop shall be placed in the tripped mode.
- The provisions of Specification 3.0.4 are not applicable.
ACTION STATEMENTS ACTION 13 - With the number of OPERABLE Channels one less than the Total Number of Channels, be in HOT STANDBY within 6 hours and in COLD SHUTDOWN within the following 30 hours; however, one channel may be bypassed for up to I hour for surveillance testing per Specification 4.3.2.1.1. ACTION 14 - With the number of OPERABLE Channels one less than the Total Number of Channels: .
- a. Below P-ll or P-12, place the inoperable channel in the
.
tripped condition within I hour; restore the inoperable channel to OPERABLE status within 24 hours after exceed-
. ing P-ll or P-12; otherwise be in at least HOT STANDBY within the following 6 hours.
- b. Above P-ll and P-12, place the inoperable channel in the tripped condition within 1 hour; operation may continue until performance of the next required CHANNEL FUNCTIONAL TEST.
ACTION 15 - With a channel associated with an operating loop inoperable, restore the inoperable channel to OPERABLE status within 2 hours or be in HOT SHUTDOWN within the following 12 hours; however, one channel associated with an operating loop may be bypassed for up to 2 hours for surveillance testing per Specification 4.3.2.1.1. ACTION 16 - With the number of OPERABLE Channels one less than the Total Number of Channels:
- a. Below P-ll or P-12, place the inoperable channel in the bypass condition; restore the inoperable channel to OPERABLE status within 24 hours after exceeding P-ll or P-12; otherwise be in at least HOT SHUTDOWN within the following 12 hours.
" BEAVER VALLEY - UNIT 1 3/4 3-20
TABLE 3.3-3 (Continued) , /^N U b. Above P-ll or P-12, demonstrate that the Minimum Channels CPGABLE requirement is met within 1 hour; operation may cc:.;.nue with the inope.?able cnannel bypassed and one aoditional channel may be bypassed for up to 2 hours for surveillance testing per Specification 4.3.2.1. ACTION 17 - With less than the Minimum Channels OPERABLE, operation may continue provided the containment purge and exhaust valves are maintained closed. ACTION 18 - With the number of OPERABLE Channels one less than the Total Number of Channels, restore the inoperable channel to OPERABLE status within 48 hours or be in at least HOT STANDBY within the next 6 hours and COLD SHUTDOWN within the following 30 hours. ACTION 33 - With tha number of OPERABLE Channels one less than the Total Number of Channels, the Emergency Diesel Generator associated with the 4kv Bus shall be declared inoperable and the ACTION Statements for Specifications 3.8.1.1 or 3.8.1.2, as appro-priate shall apply. ACTION 34 - With the number of OPERABLE Channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the inoperable channel is placed in the tripped con-dition within 1 hour. O ENGINEERED SAFETY FEATURES INTERLOCKS DESIGNATION CONDITION AND SETPOINT FUNCTION P-ll With 2 of 3 pressurizer P-il prevents or defeats pressure channels > 2010 the manual block of safety psig. injection actuation on low pressurizcr pressure. P-12 With 2 of 3 T avg channels P-12 prevents or defeats
> 545'F. the manual block of safety injection actuation on high steem line flow and low steam line pressure.
With 2 of 3 T**9 channels Allows manual block of
> 541*F. safety injection actuation on high steam line flow and low steam line pressure.
Causes steam line isolation on high steam flow. Affects steam dump blocks. BEAVER VALLEY - UNIT 1 3/4 3-21 O LJ
_ _ _ _ _ _
~ __ - ~ - _. . _ - .
O O TABLE 3.3-4 (Lontinued) o ENGillEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP SETP0lN15 FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES en 9 4. STEAM LINE ISOLATION M
- Not Applicable Not Applicable
- a. Manual Automatic Actuation Logic Not Applicable Not Applicable h b.
'h c. Containment Pressure--
,
Intennediate-liigh-liigh 1 5.0 psig 1 5.5 psig E5 d. Steam Flow in Two Steam Lines-- < A function defined as < A function defined as y liigh Coincident with Tayg-- follows: A ap correspond- follows: A Ap correspond-
- Low-Low or Steam Line Pressure-- ing to 40% of full steam ing to 44% of full steam Low flow between 0% and 20% flow between 0% and 20%
load and then a Ap load and then a ap increasing linearly to a increasiag linearly to a ap corresponding to 110% Ap corresponding to 111.5% w of full steam flow at of full steam flow at 2 full loa < full load Y T. avg > h43 F Tavy > 541"F Z 2 500 ps49 2 480 psig steam line pressure steam line pressure
- 5. TURBINE TRIP AND FEEDWATER ISOLATION
- a. Steam Generator Water Lcvel-- 1 75% of narrow range 1 76% of narrow range liigh-liigh instrument span each steam instrument span each steam
, generator generator
- 6. LOSS OF POWER
- a. 4.16 kv Emergency Bus > 75% of nominal bus > 74% of nominal bus
_ Undervoltage (Loss of Voltage) voltage with a 1 1 0.1 voltage with a 1 1 0.1 second time delay second time delay
- b. 4.16 kv Emergency Bus > 90% of nominal bus > 89% of nominal bus Undervoltage (Degraded Voltage) voltage with a 90 1 5 voltage with a 90 1 5 second time delay second time delay
- c. 480v Emergency Bus Undervoltage > 90% of nominal bus > 89% of nominal bus (Degraded Voltage) voltage with a 90 1 5 voltage with a 90 1 5 second time delay second time delay
_ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ - _ _ .- -__---__ _ _ -
.. .
.
%
4 O TA8'e 3 3-s-(co# tim"ed) ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS
- 6. Steam Flow in Two Steam Lines-High Coincident With Steam Line Pressure-Low t
- a. Safety Injection (ECCS) 3,13.0#/23.0##
.
- b. Reactor Trip (from SI) f, 3.0
- c. Feedwater Isolation < 75.0(1)
- d. Containment Isolation-Phase "A" < 22.0#/33.0##
,
- e. Auxiliary Feedwater Pumps Not Applicable i f. Rx Plant River Water System 5,77.0#110.0##
l g. Steam Line Isolation ;[ 8.0
- 7. Containment Pressure - Hich-Hich t'$ a. Containment Quench Spray 5,77.0 L) Not Applicable
- b. Containment Isolation-Phase "B"
,
.
- c. Control Roc.n Ventilation Isolation < 17.0#/30.0##
r
- 8. Steam Generator Water Level - Hich-Hich
- a. Turbine Trip-Reactor Trip ;i 2.5
- b. Feedwater Isolation 3.78.0(1)
- 9. Containment Pressure - Intermediate High-Hiah
-a. . Steam Line Isolation ' < 8.0
- 10. Loss _of Power
- a. 4.16kv Emergency Bus Undervoltage j 1.3
- . (Loss of Voltage)'
- b. 4.16kv and 480v Emergency Bus Under- ;i 95
' voltage (Degraded Voltage) l BEAVER VALLEY - UNIT 1 (}) 3/43'i. L l
... .. - . , -
_ _ _. __ _ _ _ _ _ _ _. . ._ . _ _ O O O TABLE 4.3-2 (Continued) ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS co 9 CllANNEL MODES IN WillCll
- 5 - CilANNEL CilANNEL FUNCTIONAL SURVEILLANCE
" FUNCTIONAL UNIT CllECK CAllBRATION TEST REQUIRED $ F 4. STEAM LINE ISOLATI0tl '2 -
- a. Manual N.A. N.A. M(1) 1, 2, 3, 4
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- b. Automatic Actuation Logic N.A. N.A. M(2) 1, 2, 3, 4
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- c. Containment Pressure-- S R H 1, 2, 3 Intermediate-liigh-iligh
- d. Steam Flow in Two Steam S R M 1, 2, 3 Lines--Iligh Coincident with R T ay L
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Pre 2s-reow-LoworSteamLine u -Low Y S 5. TURBINE TRIP AND FEEDWATER ISOLATION
- a. Steam Generator Water S R M 1, 2, 3 Level--Iligh-liigh
- 6. LOSS OF POWER
- a. 4.16kv Emergency Bus N.A R M 1, 2, 3 Undervoltage (Loss of Voltage)
- b. 4.16kv and 480v Emergency Bus N.A. R M 1, 2, 3 Undervoltage (Degraded Vol tage)
__ _ - _ _ _ _ _ _ - _ _ _ _____ _ _ _ _ _ _ _ _ _ _ - _ DUQUESNE LIGHT COMPANY Beaver Valley Power Station, Unit No.1 O Docket No. 50-334 License No. OPR-66 i Attachment I Modification of RCP Breaker Position Trip Logic
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O 't"ttt"o s^re svstt" seTTt"os BASES Safety Injection Inout From ESF , If a reactor trip has not already been generated by the reactor protective l instrumentation, the ESF automatic actuation logic channels will initiate a reactor trip upon any signal which initiates a safety injection. This trip is provided to protect the core in the event of a LOCA. The ESF instrumen-tation channels which initiate a safety injection signal are shown in Table 3.3-3. i . Reactor Coolant Puma Breaker Position Trio The Reactor Coolant Pump Breaker Position Trip is an anticipatory trip which provides reactor core protection against DNB resulting free the opening of two or more pump breakers above P-7. This trip is blocked below P-7. The open/close position trip assures a reactor trip signal
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is generated before the low flow trip set point is reached. No credit was taken in the accident analyses for operation of this trip. The functional capability at the open/close position settings is required to enhance the overall reliability of the Reactor Protection System.
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j i O a BEAVER VALLEY - UNIT 1 B 2-8
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O O O TABLE 3.3-1 (C6ntinued)
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REACTOR TRIP SYSTEM INSTRUMENTATION m
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$ TOTAL NO. CilANNELS CilANNELS APPLICABLE FUNCTIONAL UNIT OF CilANNELS TO TRIP OPERABLE MODES ACTION
- 18. Turbine Trip c a. Auto Stop Oil Pressure 3 2 2 1 7 y b. Turbine Stop Valve Closure 4 4 4 1 8
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- 19. Safety Injection Input 2 1 2 1, 2 1 from ESF
- 20. Reactor Coolant Pump Breaker Position Trip g Above P-7 1/ breaker 2 1/ breaker 1 11 a per oper-y ating loop
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- 21. Reacter Trip Breakers 2 1 2 1, 2* 1
- 22. Automa tic Trip Logic 2 1 2 1, 2* 1
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() TABLE 3.3-1 (Continued) ACTION 9 - With a channel associated with an operating loop inoperable, restore the inoperable channel to OPERABLE status within 2 hours or be in HOT STANDBY within the next 6 hours; however, one channel associated with an operating loop may be bypassed for up to 2 hours for surveillance testing per Specification 4.3.1.1. ACTION 10 - Not Applicable ACTION 11 - With less than the Minimum Number of Channels OPERABLE, operation may continue provided the inoperable channel is placed in the tripped condition within 1 hour. ACTION 12 - With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours or be in HOT STANDBY within the next 6 hours and/or open the reactor trip breakers.
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REACTOR TRIP SYSTEM INTERLOCKS DESIGNATION CONDITION AND SETPOINT FUNCTION P-6 With 2 of 2 Intermediate Range _)) P-6 prevents or defeats the Neutron Flux Channels < 6 x 10 manual block of source range amps. reactor trip. P-7 With 2 of 4 Power Range Neutron P-7 prevents or defeats the Flux Channels > 11% of RATED automatic block of reactor ' THERMAL POWER -r 1 of 2 Turbine trip on: Low flow in more impulse chamber pressure channels than one primary coolant
> 80 psia.
_ loop, reactor coolant pump under-voltage and under-frequency, turbine trip, pressurizer low pressure, and pressurizer high level. P-8 With 2 of 4 Power Range Neutron P-8 prevents or defeats the Flux channels > 31% of RATED automatic block of reactor THERMAL POWER.- trip on low coolant flow in a single loop. i n BEAVER VALLY - UNIT 1 3/4 3-7 V>
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l 1 DUQUESflE LIGHT COMPAflY Beaver Valley Power Station, Unit No. 1 , O Docket flo. 50-334 License flo. DPR-66 l Attachment J Redefining Operability
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O i.0 OeFINITIONS
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DEFINED TERMS l.1 The DEFINED TERMS of this section appear in capitalized type and are applicable throughout these Technical Specifications. . THERMAL POWER 1.2 THEPFAL POWER shall be the total reactor core heat transfer rate to the reactor coolant. RATED THERMAL POWER 1.3 RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 2652 MWt. OPERATIONAL MODE 1.4 An OPERATIONAL MODE shall correspond to any one inclusive combina-tion of core reactivity condition, power level and average reactor
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coolant temperature specified in Table 1.1. O v ACTION 1.5 ACTION shall be those additional requirements specified as corollary statements to each principle specification and shall be part of the specifications. OPERABLE - OPERABILITY l 1.6 A system, subsystem, train, component or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified function (s). Implicit in this definition shall be the assumption that all necessary attendant instrumentation, controls, normal and emergency electrical power sources, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component or device to perform its function (s) are also capable of performing their related support function (s). BEAVER VALLEY - UNIT.1 1-1
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(. 3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.0 APPLICABILITY LIMITING CONDITION FOR OPERATION 3.0.1 Limiting Conditions for Operation and ACTION requirements shall be applicable during the OPERATIONAL MODES or other conditions specified for each specification. 3.0.2 Adherence to the requirements of the Limiting Condition for Operation and/or associated ACTION within the specified time interval shall constitute compliance with the specification. In the event the Limiting Condition for Operation is restored prior to expiration of the specified time interval, completion of the ACTION statement is not required. 3.0.3 In the event a Limiting Condition for Operation and/or associated ACTION requirements cannot be satisfied because of circumstances in excess of those addressed in the specification, the unit shall be placed in at least HOT STANDBY within 1 hour, in at least H0T SHUTDOWN within the next 6 hours, and in at least COLD SHUTDOWN within the following 30 hours unless corrective measures are completed that permit operation under the permissi-ble ACTION statements for the specified time interval as measured from pd initial discovery or until the reactor is placed in a MODE in which the specification is not applicable. Exceptions to these requirements shall be stated in the individual specifications. 3.0.4 EntryintoanOPERATIONdLMODEorotherspecifiedapplicability condition shall not be made unless the conditions of the Limiting Con-dition for Operation are met without reliance on provisions contained in the ACTION statements unless otherwise excepted. This provision j shall not prevent passage through OPERATIONAL MODES as required to com-
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ply with ACTION statements. 3.0.5 When a system, subsystem, train, component or device is determined to be inoperable soley because its emergency power source is inoperable,
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4 or solely because its normal power source is inoperable, it may be con-sidered OPERABLE for the purpose of satisfying the requirements of its applicable Limiting Condition for Operation, provided: (1) its corre-sponding normal or emergency power source is OPERABLE; and (2) all of its redundant system (s), subsystem (s), trains (s), component (s) and device (s) are OPERABLE, or likewise satisfy the requirements of this specification. Unless both conditions (1) and (2) are satisfied, the unit shall be placed in at least HOT STANDBY within I hour, in at least HOT SHUTDOWN within the next 6 hours, and in at least COLD SHUTDOWN within the following 30 hours. This specification is not applicable in MODES 5 or 6.
.O staveR v4LLEY - UNIT i 3,4 0-i
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_ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ - O s.J SURVEILLANCE REQUIREMENTS 4.0.1 Surveillance Requirements shall be applicable during the OPERATIONAL MODES or other conditions specified for individual Limiting Conditions for Operation unless otherwise stated in an individual Surveillance Requirement. 4.0.2 Each Surveillance Requirement shall be performed within the specified time interval with:
- a. A maximum allowable extension not to exceed 25% of the surveil-lance interval, and
- b. A total maximum combined interval time for any 3 consecutive surveillance intervals not to exceed 3.25 times the specified surveillance interval.
4.0.3 Performance of a Surveillance Requirement within the specified time interval shall constitute compliance with OPERABILITY requirements for a Limiting Condition for Operation and associated ACTION statements unless otherwise required by the specification. 4.0.4 Entry into an OPERATIONAL MODE or other specified applicability condition shall not be made unless the Surveillance Requirement (s) associated with the Limiting Condition for 0?eration have been perform-ed within the stated surveillance interval or as otherwise specified. The provisions of Specification 4.0.4 are not applicable to the perform-
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ance of surveillance activities associated with fire protection technical specifications 4.7.14 and 4.7.15 until the completion of the initial .sur-veillance interval associated with each specification. 4.0.5 Surveillance Requirements for inservice inspection and testing of ASME Code Class 1, 2 and 3 components shall be applicable as follows:
- a. Inservice inspection of ASME Code Class 1, 2 and 3 components and inservice testing of ASME Code Class i, 2 and 3 pumps and valves shall be performed in accordance with Section XI of tne ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50, Section 50.55a(g), except wnere specific written relief has been granted by the Commission pursuant to 10 CFR 50, Section 50.55a(g)(6)(i).
- b. Surveillance intervals specified in Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda for the inservice inspection and testing activities required by the ASME Boiler and Pressure Vessel Code and applicable Addenda shall be applicable as follows 'in these Technical Specifications:
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() APPLICABILITY SURVEILLANCE REQUIREMENTS (continued) ASME Boiler and Pressure Vessel Code and applicable Required frequencies Addenda terminology for for performing inservice inservice inspection and inspecticn and testing testino activities activities Weekly At least once per 7 days Monthly At least once per 31 days Quarterly or every 3 months At least once per 92 days Semiannually or every 6 months At least once per 184 days Yearly or annually At least once per 366 days
- c. The provisions of Specification 4.0.2 are applicable to the l
above required frequencies for performing inservice inspec-tion and testing activities.
- d. Performance of the above inservice inspection and testing activities shall be in addition to other specified Surveil-lance Requirements.
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- e. Nothing in the ASME Boiler and Pressure Vessel Code shall be construed to supersede the requirements of any Technica,1 Specification.
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p C 3/4.0 APPLICABILITY BASES The specifications of this section provide the general requirements applicable to each of the Limiting Conditions for Operation and Surveil-lance Requirements within Section 3/4. 3.0.1 This specification defines the applicability of each specifi-cation-in terms of defined OPERATIONAL MODES or other specified conditions and is provided to delineate specifically when each specif1 cation is applicable. 3.0.2 This specification defines those conditions necessary to con-stitute compliance with the terms of an individual Limiting Condition for Operation and associated ACTION requirements. 3.0.3 This specification delineates the ACTION to be taken for circumstances not directly provided for in the ACTION statements and whose occurrence would violate the intent of the specification. For example, Specification 3.5.1 requires each Reactor Coolant System accumulator to be OPERABLE and provides explicit ACTION requirements if one accumulator is inoperable. Under the terms of Specification 3.0.3, if more than one accumulator is inoperable, the unit is required to be in at least H0T STANDBY (m) within 1 hour and in at least HOT SHUTDOWN within the follcuing 6 hours. As a further example, Specification 3.6.2.1 requires two Containment Spray Systems to be OPERABLE and provides explicit ACTION requirements if one spray system is inoperable. Under the terms of Specification 3.0.3, if both of the i required Contair. ment Spray Systems are inoperable, the unit is required to be in at least HOT STANDBY within 1 hour, in at least HOT SHUTDOWN within the
' following 6 hours and in at least COLD S?UTDOWN in the next 30 hours. It is assumed that the unit is brought to tha required MODE within the required times by promptly initiating and c2rrying out the appropriate ACTION state-ment.
3.0.4 This specification provides taat entry into an OPERABLE MODE-or other specified applicability condition me;t be made with (a) the full complement of required systems, equipment or components OPERABLE and (b) all other parameters as specified in the Limiting Conditions for Operation being met without regard for allowaole -deviations and out of service pro- . visions contained in the ACTION statements. The intent of this provision is to insure that facility operation is not initiated with either required equipment or systems inoperable or other specified limits being exceeded. Exceptions to this provision have been provided for a limited number of specifications when startup with inoperable equipment would not affect plant safety. These exceptions are stated in the ACTION statements of
' the appropriate specifications.
L') t BEAVER VALLEY - UNIT ~1- B 3/4 0-1
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- n APPLICABILITY
BASES i I 3.0.5 This specification delineates what additional conditions must i be satisfied to permit operation to continue, consistent with the ACTION statements for power sources when a normal or emergency power source is It specifically prohibits operation when one division is . ' not OPERABLE.
- inoperable because its normal or emergency power source is inoperable and
! a system, subsystem, train, component or device in another division is inoperable for another reason. i The provisions of this specification permit the ACTION. statements
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associated with individual systems, subsystems, trains, components or ? devices to be consistent with the ACTION statements of the associated electrical power source. It allows operation to be governed by the time
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limits of the ACTION statement associated with the Limiting Condition i for Operation for the normal or emergency power source, not the individual ACTION statements for each system, subsystem, train, component or device that is determined to be inoperable solely because of the inoperability of its normal or emergency power source. For example, Specification 3.8.1.1 requires in part that two emergency diesel generators be OPERABLE. The ACTION statement provides for a 72 hour q out-of-service time when one crergency diesel generator is not OPERABLE. V If the definition of OPERABLE were applied without consideration of Specification 3.0.5, all' systems, subsystems, trains, components and devices supplied by the inoperable emergency power source would also be inoperable. This would dictate invoking the applicable ACTION statements for each of the applicable Limiting Conditions for Operation. However, the provisions of Specification 3.0.5 permit the time limits for continued operation to_ be consistent with the ACTION statement for the inoperable emergency diesel generator instead, provided the other specified conditions are satisfied. In this case, this would mean that the corresponding normal power-source must be OPERABLE and all redundant systems, subsystems, trains, components and devices must be OPERABLE, or otherwise satisfy Specification 3.0.5 (i.e., be capable of performing their design function and have at least one normal or one emergency power source 0PERABLE). If they are not satisfied, shutdown is required in accordance with this specification. As a further example, Specification'3.8.1.1 requires in part that two physically independent' circuits between the offsite transmission network and the onsite Class;IE distribution system be OPERABLE. The ACTION statement
.provides a 24 hour out-of-service time when both required offsite ' circuits' are not OPERABLE. If the definition of OPERABLE were applied without con-sideration of Spfeification 3.0.5, all systems, subsystems, trains, compon-ents and devices supplied by the inoperable nonnal. power sources, both of the offsite circuits, would'also be inoperable. This,would dictate invoking the appil. cable ACTION statements'for each of the applicable LCOs. However, O BE M R VAL UY - UNIT 1 , B 38 0-2
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em APPLICABILITY U BASES __ the provisions of Specification 3.0.5 permit the time limits for continued operation to be consistent with the ACTION statements for the inoperable normal power sources instead, provided the other specified conditions are satisfied. In this case, this would mean that for one division, the emer-gene.y power source must be OPERABLE (as must be the components supplied by the emergency power source) and all redundant systems, subsystems, trains, components and devices in the other division must be OPERABLE, or likewise satisfy Specification 3.0.5 (i.e, be capable of performing their design functions and have an emergency power source OPERABLE). In other words, both emergency power sources must be OPERABLE and all redundant systems, subsystems, trains, ccmponents and devices in both divisions must also be OPERABLE. If these conditions are not satisfied, shutdown is required in accordance with this specification. In MODES 5 or 6 Specification 3.0.5 is not applicable, and thus the individual ACTION statements for each applicable Limiting Condition for Operation in these MODES must be adhered to. 4.0.1 This specification provides that surveillance activities necessary to insure the Limiting Conditions for Operation are met and will p be performed during the OPERATIONAL MODES or other conditions for which d the Limiting Conditions for Operation are applicable. Provisions for additional surveillance activities to be performed without regard to the applicable OPERATIONAL MODES or other conditions are provided in the individual Surveillance Requirements. 4.0.2 The provisions of this specification provide allowable tolerances. for performing surveillance activities beyond those specified in the nominal surveillance interval. These tolerances are necessary to provide operational flexibility because of scheduling and performance considerations. The tolerance values, taken either individually or consecutively over three test intervals, are sufficiently restrictive to ensure that the reliability associated with the surveillance activity is not significantly degraded beyond that obtained from the nominal specified interval. 4.0.3 The provisions of this specification set forth the criteria for determination of compliance with the OPERABILITY requirements of the Limiting Conditions for Operation. Under this criteria, equipment, systems or components are assumed to be OPERABLE if the associated surveillance activities have been satisfactorily performed within the specified time interval. Nothing in this provision is to be construed as defining equip-ment, . systems or components OPERABLE, when such items are found or known to be inoperable although still meeting the Surveillance Requirements. p) n BEAVER VALLEY - UNIT 1 B 3/4 0-3 i L
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- 4.0.4 This specification ensures that the surveillance activities associated with a Limiting Condition for Operation have been performed i within the specified time interval-prior to entry into an OPERATIONAL MODE
' or other applicable condition. The intent of this provision is to ensure- ' that surveillance activities have been satisfactorily demonstrated on a j current basis as required to meet the OPERABILITY. requirements of the
Limiting Condition for Operation.
- Under the terms of this specification, for example, during initial i plant startup or following extended plant outages, the applicable
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surveillance activities must.be performed within the stated surveillance interval prior to placing or returning the system or equipment into OPERABLE status.
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4.0.5 This specification ensures that inservice inspection of ASME Code Class 1, 2 and 3 components and inservice testing of ASME Code
- Class 1, 2 and 3 pumps and valves will be performed in accordance with a
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periodically updated version of Section XI of the ASME Boiler and Pressure Vessel Code and Addenda as required by_10 CFR 50.55a. Relief from any of the above requirements has been provided in writing by the Commission and
- is not a part of these Technical Specifications.
J This specification includes a clarification of the frequencies for performing the inservice inspection and testing activities required by Section XI aof the ASME Boiler and Pressure Vessel Code and applicable Addenda. This clarification is provided to ensure consistency in sur-veillance intervals throughout these Technical Specifications and to remove any ambiguities relative to the frequencies for performing the required inservice inspection and testing activities. Under. the terms of this specification, the more restrictive require-ments of the Technical Specifications take precendence over the ASME-
- Boiler and Pressure Vessel Code and applicable Addenda. For example, the requirements of Specification' 4.0.4 to perform surveillance activities-prior to entry into an OPERATIONAL MODE or other specified applicability
, i conditon takes precedence over the ASME Boiler and Pressure Vessel. Code provision which allows' pumps to be tested up to one week after return to.
' normal operation and for example, the Technical Soarification definition of OPERABLE does not grant a' grace period before .evice that is not .
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capable of performing its 'specified functions is declared inoperable and takes precedence over the ASME Soiler and Pressure Vessel Code provision which allows a valve to be incapable of performing its specified function for up to 24 hours before being . declared: inoperable. >
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