ML20035C216
| ML20035C216 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 04/01/1993 |
| From: | BALTIMORE GAS & ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20035C213 | List: |
| References | |
| NUDOCS 9304060342 | |
| Download: ML20035C216 (14) | |
Text
9 4
ATTACHMENT (1)
UNIT 1 TECIINICAL SPECIFICATION REVISED PAGES B3/47-1 B3/47-3 B3/47-5 B 3/411-2 B 3/411-4 j
9304060342 930401 PDR ADOCK 05000317 P
3/4.7 PLANT SYSTEMS BASES 3/4.7.1 TURBINE CYCLE 3/4.7.1.1 Safety Valves pk The OPERABILITY of the - in steam line code safety valves ensures that the Secondary System 3res re will be limited to within 110% of its design pressure of 1000(ftg3during the most severe anticipated system operational transient. The total relieving capacity for all valves on all of the steam lines is 12.18 x 10' lbs/hr at 100% RATED THERMAL POWER. The maximum relieving capacity is associated with a turbine trip from 100% RATED THERMAL POWER coincident with an assumed loss of condenser heat sink (i.e., no steam bypass to the condenser). The main steam line code safety valves are tested and maintained in accordance with the r f
Sect on
-seth(ing,s m
_AjMEBoi
_e no Teifs L g p: y gig Jesy i
ar. w ps., to-ensumt.._ _ the lift settaints will mmM n wi t h i n - s necifications-dur4 fig-the-cych In MODE 3, two main steam safety valves are required OPERABLE per steam generator. These valves will provide adequate relieving capacity for removal of both decay heat and reactor coolant pump heat from the Reactor Coolant System via either of the two steam generators. This requirement is l
provided to facilitate the post-overhaul setting and OPERABILITY testing of the safety valves which can only be conducted when the RCS is at or above 500 F.
It allows entry into MODE 3 with a minimum number of main steam
- vi safety valves OPERABLE so that the set pressure for the remaining valves can be adjusted in the plant. This is the most accurate means for adjusting safety valve set pressures since the valves will be in thermal equilibrium with the operating environment.
STARTUP and/or POWER OPERATION is allowable with safety valves inoperable within the limitations of the ACTION requirements on the basis of the reduction in Secondary System steam flow and THERMAL POWER required by the reduced reactor trip settings of the Power Level-High channels.
The reactor trip setpoint reductions are derived on the following bases:-
For two loop operation SP = (X) - 0) @) x 106.5 X
For single loop operation (two reactor coolant pumps operating in the same loop) 3p _ (X) - (Y)(U) x 46.8 X
CALVERT CLIFFS - UNIT 1 3 3/4 7-1 Amendment No.,(ES/
9 3/4.7 PLANT SYSTEMS BASES w~
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?#P a ce. '
In t c 3pcctrum si events analyzed in which detomatic initiatien of l
aniliery feecwater umis, thc fcilcwing fico condi ti ciis arm all owed-wi-th Jfd-2n operator ection timc cf 10 minsteh s
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Loss of Foo & ater 0 gu AuAiliary Feedwater Flow 42)
F d14ne Paek 0 spa Auxilidij Feedweter Flew i
(1) u $ Stegr t$n 37;;g
- g;g ;3;;;37j 7;g3 e3 Flcw (This being the m uimum f ew through the Ar,! sucticn Linc, with one unit r;quirMg t
Bow, prior to pgsp csvitat4cn
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At 10 minutes after an Auxiliary Feedwater Actuation Signal the operator is assumed to be available to increase or decrease auxiliary feedwater flow to that required by the existing plant conditions.
3/4.7.1.3 Condensate Storace Tank The OPERABILITY of the condensate storage ' tank'with the minimum water volume ensures that sufficient water is available to maintain the RCS at HOT STANDBY conditions for 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> with steam discharge to atmosphere with 59 concurrentsand total loss of offsite power. The contained water volume limit includes an allowance for water not usable because of tank discharge line location or other physical characteristics.
3/4.7.1.4 Activity The limitations on Secondary System specific activity ensure that the resultant off-site radiation dose will be limited to a small fraction of 10 CFR Part 100 limits in the event of a steam line rupture. This dose also includes the effects of a coincident 1.0 GPM primary to secondary tube leak in the steam generator of the affected steam line and concurrent loss of offsite electrical pcwer. These values are consistent with the assumptions used in the accident analyses.
3/4.7.1.S Main Steam Line Isolation Valves The OPERABILITY of the main steam line isolation valves ensures that no more than one steam generator will blowdown in the event of a steam line rupture. This restriction is required to 1) minimize the positive reactivity effects of the Reactor Coolant System cooldown associated with the blowdown, and 2) limit the pressure rise within containment in the i
event the steam line rupture occurs within containment. The OPERABILITY of i
CALVERT CLIFFS - UNIT 1 3 3/4 7-3 Amendment No. FA j
=.
4 l
INSERT A Auxiliary Feedwater flow and response times are conservatively accounted for in the analyses of Design Basis Events. In Main Steam Line Break and Excess Imad analyses where Auxiliary Feedwater flow would increase the consequences of the accidents, the delay time for Auxiliary Feedwater actuation is minimized and Auxiliary Feedwater flow is maximized. In Feedline Break, Iess of Feedwater, and loss of Non-Emergency AC Power analyses, in which Auxiliary Feedwater flow would decrease the consequences of the accidents, the delay time for Auxiliary Feedwater actuation is maximized and Auxiliary Feedwater flow is minimized.
t i
i
1 3/4.7 PLANT SYSTEMS BASES r
accident conditions. The OPERABILITY of this system in conjunction with Control Room design provisions is based on limiting the radiation exposure to personnel occupying the Control Room to 5 rem or less whole body, or its equivalent. This limitation is consistent with the uirements of 10 CFR Part 50, Appendix A, General Design Criteri dr.
11 3/4.7.7 ECCS pVMP ROOM EXHAUST AIR FILTRATION SYSTEM The OPERABILITY of the ECCS Pump Room Exhaust Air Filtration System ensures that radioactive materials leaking from the ECCS equipment within the pump room following a LOCA are filtered prior to reaching the environment. The operation of this system and the resultant effects on offsite dosage calculations was assumed in the accident analyses.
3/4.7.8 SNUBBERS All safety related snubbers are required OPERABLE to ensure that the structural integrity of the Reactor Coolant System and all other safety related systems is maintained during and following a seismic or other event initiating dynamic loads. Snubbers excluded from this inspection program are those installed on non-safety related systems and then only if their t
failure or failure of the system on which they are installed would have no adverse effect on any safety-related system.
t The visual-inspection frequency is based upon maintaining a constant level of snubber protection to systems. Therefore, the required inspection interval. varies inversely with the observed snubber failures and is determined by the number of inoperable snubbers found during the previous inspection, the total population or category size, and the previous inspection interval.
Snubbers may be categorized, based upon their accessibility during power operation, as accessible or inaccessible. These categories may be examined separately or jointly. However, that decision must be made and documented before any inspection and that decision shall be used as the basis upon which to determine the next inspection interval for that category.
Inspections performed before that interval has elapsed may be used as a new reference point to determine the next inspection. However, the results of such early inspections performed before the original required time interval has elapsed (nominal time less 25%) may not be used to lengthen the required inspection interval. Any inspection whose results require a shorter inspection interval will override the previous schedule.
When the cause of the rejection of a snubber is clearly established and remedied for that snubber and for any other snubbers that may be generically susceptible, and verified by inservice functional testing, that i
l CALVERT CLIFFS - UNIT 1 3 3/4 7-5 Amendment No. 169'
s 3/4.11 RADI0 ACTIVE EFFLUENTS BASES 3/4.11.1.3 Licuid Radwaste Treatment System The requirement that the appropriate portions of this system be used, when 1
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 50.36a, of 10 CFR Part 50, Appendix A. General Design Criterion 60 and the design objective given in 10 CFR Part 50, Appendix I, Section II.D. 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 s o
10 CFR Part 50 A ndix I. Section
.A f r iqui eff1uents.
e dose./:i.e zap mhierop c[p/&paMp(g, (e.g %dk 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 and beyond the SITE BOUNDARY from gaseous effluents from all units on the site will be within the annual dose limits of 10 CFR Part 20 to 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 a MEMBER OF THE PUBLIC sWW in an UNRESTRICTED AREA, exceeding the limits specified in 10 CFR Part 20, Appendix B, Table II (10 CFR 20.106(b)). For MEMBERS OF THE PUBLIC who may at times be within the SITE BOUNDARY, the occupancy of that MEMBER OF THE PUBLIC will be sufficiently low to compensate for any increase in the atmospheric diffusion factor above that for the SITE BOUNDARY.
The required detection capabilities for radioactive materials in gaseous waste samples are tabulated in terms of the lower limits of detection (LLDs). Detailed discussion of the LLD, and other detection limits can be found in HASL Procedures Manual, HASL-300, Currie, L.A., " Limits for Qualitative Detection and Quantitative Determination - Application to Radiochemistry," Anal. Chem. a0, 586-93 (1968), and Hartwell, J.K.,
" Detection Limits for Radioanalytical Counting Techniques," Atlantic Richfield Hanford Company Report ARH-SA-215 (June 1975).
F l
CALVERT CLIFFS - UNIT 1 B 3/4 11-2 AmendmentNo.jui(
j
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k 3/4.11 RADI0 ACTIVE EFFLUENTS BASES parameters for calculating the doses due to the actual release rates of the subject materials 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, 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, and NUREG-0133, " Preparation of Radiological Effluent Technical-Specifications for Nuclear Power Plants". These equations also provide for deternining the actual doses based upon the historical annual average atmospheric conditions. The release rate specifications for iodine-131 and radionuclides in particulate fann with half lives greater than 8 days are dependent upon the existing radionuclide pathways to man, in the areas at and beyond the SITE BOUNDARY. The pathways that were examined in the development of these calculations were:
- 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 TREATMENT SYSTEM The requirement that the appropriate portions of these systems be used.
when specified, provides reasonable assurance that the releases of radioactive materials in gaseous effluents will be kept "as low as is reasonably achievable".
This specification implements the requirements of 10 CFR 50.36a, 10 CFR Part 50. Appendix A, General Design Criterion 60 and the design objectives given in 10 CFR Part 50, Appendix I, Section II.D.
The specified limits governing the use of appropriate portions of the systems were specified as a suitable fraction of the dose design objectives set forth in 10 50, Appendix IQcti_ons II.B _and LI.C for a +
effJ uents.
dose.ma
~ [p4 h N cosEMcWM+%/od(.cq m Mhs).
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3/4.11.2.5 Exolosive 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 flamability limit of oxygen. Maintaining the concentration of oxygen below its flamability limit provides assurance that the releases of radioactive materials will be controlled in conformance with the requirements of 10 CFR Part 50, Appendix A, General Design Criterion 60.
CALVERT CLIFFS - UNIT 1 B 3/4 11-4 Amendment No. J591
ATTACHMENT (2)
UNIT 2 TECIINICAL SPECIFICATION REVISED PAGES B3/47-1 B3/47-3 B3/47-5 B 3/411-2 1-B 3/411-4 t
j i
3/4.7 PLANT SYSTEMS BASES ~
i 3/4.7.1 TURBINE CYCLE 3/4.7.1.1 Safety Valves ps; The OPERABILITY of the in steam line code safety valves ensures that the Secondary System re will be limited to within 110% of its design pressure of 1000 during the most severe anticipated system operational transient. The relieving capacity for all valves on all of the steam lines is 12.18 x 10' lbs/hr at 100% RATED THERMAL POWER. The maximum relieving capacity is associated with a turbine trip from 100% RATED Ve55c/
THERMAL POWER coincident with an assumed loss of condenser heatD (i.e., no steam bypass to the condenser). The Eline code safety valves are tested and maintained in accord gMTs,ujtements of ith the req Section XI of theJSM
- . W ~e W i Ws h an'E_ Roller and Pressur eft li.t settings h
985'pT7thnsura lelif-taetpoints will ca6
=
.within :pecificat4cn-during the cycle-In MODE 3, two main steam safety valves are required OPERABLE per steam generator. These valves will provide adequate relieving capacity for removal of both decay heat and reactor coolant pump heat from the Reactor Coolant System via either of the two steam generators. This requirement is provided to facilitate the post-overhaul setting and OPERABILITY testing of the safety valves which can only be conducted when the RCS is at or above 500 F.
It allows entry into MODE 3 with a minimum number of main steam e*4 safety valves OPERABLE so that the set pressure for the remaining valves can be adjusted in the plant. This is the most accurate means for adjusting safety valve set pressures since the valves will be in thermal equilibrium with the operating environment.
STARTUP and/or POWER OPERATION is allowable with safety valves inoperable within the limitations of the ACTION requirements on the basis of the reduction in Secondary System steam flow and THERMAL POWER required by the reduced reactor trip settings of the Power Level-High channels. The reactor trip setpoint reductions are derived on the following bases:
For two loop operation SP =
~
x 106.5 X
For single loop operation (two reactor coolant pumps operating in the same loop) 3p, (X) - (Y)(V) x 46.8 X
CALVERT CLIFFS - UNIT 2 B 3/4 7-1 Amendment No. A47
3/4.7 PLANT SYSTEMS BASES f
7 the sput= of+ei.a analsed ta which :utmetic a itiauon or
~
? Reela c e an Q
f -aux 411ary feMwater neem s the Sile*; #!cw conditimis are 4iicweu wi th an_g erM or actica tdre of 10 *inutes,
Irl5EKT A (1)
Lcz of Fccdwater 0 gpm Auxilicry Fccd &ter Flow (2)
Fcedline Break-0 CNi Auxilidry Fudwater Flow 0)
"cin Stam Lin: Srcak 1550 gpm Auxiliary F cdwater-
-Flow (this being the max 4 mum ficw thram che AFW >Jctiso-Linc, Mh onc unit-requiring f4c., grive La pwup car i L4Liuw h
de Lu iow iiPs~it}.
W At 10 minutes after an Auxiliary Feedwater Actuation Signal, the operator is assumed to be available to increase or decrease auxiliary feedwater flow to that required by the existing plant conditions.
3/4.7.1.3 Condensate Storace Tank The OPERABILITY of the condensate storage tank'with the minimum water volume ensures that sufficient water is available to maintain the RCS at HOT STANDBY conditions for 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> with steam discharge to atmosphere with W
concurrent-and total loss of offsite power. The contained water volume limit includes an allowance for water not usable because of tank discharge line location or other physical characteristics.
3/4.7.1.4 Activity The limitations on Secondary System specific activity ensure that the resultant off-site radiation dose will be limited to a small fraction of 10 CFR Part 100 limits in the event of a steam line rupture.
This dose also includes the effects of a coincident 1.0 GPM primary to secondary tuce leak in the steam generator of the affected steam line and a concurrent loss of offsite electrical power. These values are consistent with the assumptions used in the accident analyses.
3/4.7.1.5 Main Steam Line Isolation Valves The OPERABILITY of the main steam line isolation valves ensures that no more than one steam generator will blowdcwn in the event of a steam line rupture. This restriction is required to 1) minimize the positive reactivity effects of the Reactor Coolant System cooldewn associated with the blowccwn, and 2) limit the pressure rise within containment in the event the steam line rupture occurs within containment. The OPERABILITY of CALVERT CLIFFS - UNIT 2 3 3/4 7-3 AmencmentNo.{49
i 1
l INSERT A
)
Auxiliary Feedwater flow and response times are conservatively accounted for in the analyses of j
Design Basis Events. In Main Steam Line Break and Excess load analyses where Auxiliary Feedwater flow would increase the consequences of the accidents, the delay time for Auxiliary Feedwater actuation is minimized and Auxiliary Feedwater flow is maximized. In Feedline Break, Imss of Feedwater, and Loss of Non-Emergency AC Power analyses, in which Auxiliary Fecawater flow would decrease the consequences of the accidents, the delay time for Auxiliary Feedwater i
actuation is maximized and Auxiliary Feedwater flow is minimized.
l v
5 k
+
l t
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e t
f a
f
P 3/4.7 PLANT SYSTEMS BASES accident conditions. The OPERABILITY of this system in conjunction with Control Room design provisions is based on limiting the radiation exposure to personnel occupying the Control Room to 5 rem or less whole body, or its equivalent. This limitation is consistent with the recuirements of 10 CFR Part 50, Appendix A,. General Design Criteria 3/4.7.7 ECCS PUMP ROCM EXHAUST AIR FILTRATION SYSTEM The OPERABILITY of the ECCS Pump Room Exhaust Air Filtration System ensures that radioactive materials leaking from the ECCS equipment within the pump room following a LOCA are filtered prior to reaching the environment. The operation of this system and the resultant effects on offsite dosage calculations was assumed in the accident analyses, e
3/4.7.8 SNUBBERS All safety related snubbers are required OPERABLE to ensure that the structural integrity of the Reactor Coolant System and all other safety related systems is maintained during and following a seismic or other event initiating dynamic loads. Snubbers excluded from this inspection program are those installed on non-safety related systems and then only if their failure or failure of the system on which they are installed would have no adverse effect on any safety-related system.
The visual ~ inspection frequency is based upon maintaining a constant level of snubber protection to systems. Therefore, the required inspection interval varies inversely with the observed snubber failures and is determined by the number of inoperable snubbers found during the previous inspection, the total population or category size, and the previous inspection interval.
Snubbers may be categorized, based upon their accessibility.during power operation, as accessible or inaccessible. These categories may be examined separately or jointly. However, that decision must be made and documented before any inspection and that decision shall be used as the basis upon which to deternine the next inspection interval for that category.
Inspections perfomed before that interval has elapsed may be used as a new reference point to determine the next inspection. However, the results of such early inspections performed before the original required time interval has elapsed (nominal time less 25%) may not be used to lengthen the t
required inspection interval. Any inspection whose results require a shorter inspection interval will override the previous schedule.
When the cause of the rejection of a snubber is clearly established and remedied for that snubber and for any other snubbers that may be generically susceptible, and verified by inservice functional testing, that l
1 CALVERT CLIFFS - UNIT 2 3 3/4 7-5 Amendment No. [
l i
3/4.11 _RADI0 ACTIVE EFFLUENTS i!
BASES i
3/4.11.1.3. Licuid Radwaste Treatment System The requirement that the appropriate portions of this system be used, when specified, provides assurance that the releases of radicactive materials in liquid effluents will be kept "as low as is reasonably achievable". This specification implements the requirements of 10 CFR 50.36a,10 CFR Part 50, i
Appendix A. General Design Criterion 60 and the design objective given in 10 CFR Part 50, Appendix 1.Section II.D. The specified limits governing l
the use of appropriate portions of the Liquid Radwaste Treatment System were specified as a suitable fraction of the dose design objectives set l
R_Part_50 A cendix I, S tion II_._
ffluents.
D comb;ms e /acd 441 pad &g.k.uaHg) t SEOUS tFFLUENTT i
3/4.11.2.1 Dose Rate This specification is provided to ensure that the dose at any time at and beyond the SITE B0UNDARY from gaseous effluents from all units on the site will be within the annual dose limits of 10 CFR Part 20 to UNRESTRICTED i
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 a MEMBER OF THE PUBLIC in an UNRESTRICTED AREA, exceeding the limits specified in Appendix B, g
Table II of~ 10 CFR 20.106(b).
For MEM8ERS OF THE TUBLIC who may at times be within the SITE SOUNDARY, the occupancy of that MEMBER OF THE PUBLIC will be sufficiently low to compensate for any increase in the atmospheric diffusion factor above that for the SITE B0UNDARY.
The required detection capabilities for radioactive materials in gaseous waste samples are tabulated in terms of the lower limits of detection (LLDs). Detailed discussion of the LLD, and other detection limits can be found in HASL Procedures Manual, HASL-300, Currie, L.A., " Limits for Qualitative Detection and Quantitative Detennination - Application to Radiochemistry," Anal. Chem. 40, 586-93 (1968), and Hartwell, ~ J.K.,
" Detection Limits for Racioanalytical Counting Techniques," Atlantic Richfield Hanford Company Report ARH-SA-215 (June 1975).
3/4.11.2.2 Dose - Noble Gases This specification is provided to implement the requirements of 10 CFR Part 50, Appendix I, Sections II.B. III.A and IV.A.
The Limiting Condition l
for Operation implements the guides set forth in Appendix I, i
CALVERT CLIFFS - UNIT 2 B 3/4 11-2 AmendmentNo.Ipp J
3/4.11 RADI0 ACTIVE EFFLUENTS BASES of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors,"
Revision 1. July 1977, and NUREG-0133, " Preparation of Radiological Effluent Technical Specifications for Nuclear Power Plants". These equations also provide for determining the actual doses based upon the historical annual average atmospheric conditions. The release rate specifications for iodine-131 and radionuclides in particulate form with half lives greater than 8 days are dependent upon the existing radionuclide pathways to man, in the areas at and beyond the SITE BOUNDARY. The pathways that were examined in the development of these calculations were:
- 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 o1 the ground with subsequent exposure of man.
3/4.11.2.4 GASE0US RADWASTE TREATHENT SYSTEM The requirement that the appropriate portions of these systems be used, when specified, provides reasonable assurance that the releases of radioactive materials in gaseous effluents will be kept "as low as is reasonably achicianle".
This specification implements the requirements of 10 CFR 50.36a,10 CFR g
Part 50, Appendix A, General Design Criterion 60 and the design objectives given in 10 CFR Part 50, Appendix I, Section II.D'.
The specified limits governing the use of appropriate portions of the systems were specified as a suitable fraction of the dose design objectives set forth in 10 R
P n
eo e fluents.,he ne_
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 flamability limit of oxygen. Maintaining the concentration of oxygen below its flamability limit provides assurance that the releases of radioactive materials will be controlled in conformance with the requirements of 10 CFR Part 50, Appendix A, General Design Criterion 60.
3/4.11.2.5 Gas Storace Tanks The tanks included in this specification are those tanks for which the quantity of radioactivity contained is not limited directly or indirectly by another Technical Specification to a quantity that is less than the quantity that provides assurance that in the event of an l
l J
CALVERT CLIFFS - UNIT 2 B 3/4 11-4 AmendmentNo.Id7 l