ML20035H012

Discusses Proposed Changes to Plant TS Bases to Accurately Reflect Info in UFSAR & Previously Approved Amends to Ts. Rev Instructions & Revised TS Bases Pages Encl

ML20035H012
Person / Time
Site:	Calvert Cliffs
Issue date:	04/26/1993
From:	Mcdonald D Office of Nuclear Reactor Regulation
To:	Denton R BALTIMORE GAS & ELECTRIC CO.
References
TAC-M86136, NUDOCS 9304300352
Download: ML20035H012 (15)

Text

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[kAT UNITED STATES

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NUCLEAR REGULATORY COMMISSION

' f WASHINGTON. D.C. 20555-0001

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s April 26, 1993 Docket Nos. 50-317 and 50-318 Mr. Robert E. Denton Vice President - Nuclear Energy Baltimore Gas & Electric Company Calvert Cliffs Nuclear Power Plant 1650 Calvert Cliffs Parkway Lusby, Maryland 20657-4702

SUBJECT:

CHANGES TO THE CALVERT CLIFFS NUCLEAR POWER PLANT, UNITS 1 AND 2 TECHNICAL SPECIFICATION BASES, TAC NOS. M86136 (UNIT 1) AND M86137 (UNIT 2)

By letter dated April 1, 1993, Baltimore Gas and Electric Company (BG&E) identified revisions to the Bases Sections of the Calvert Cliffs Nuclear Power Plant, Units 1 and 2, Technical Specifications (TS).

Four specific changes to the TS Bases are being proposed to accurately reflect the information in the Updated Final Safety Analysis Report (UFSAR) and previously approved amendments to the TS.

The specific changes being proposed and the justification for the changes are:

Chance No. 1 This change proposes to resolve a discrepancy between the UFSAR, Chapter 4, Section 4.1.3.2, " Steam Generators," and TS Bases Section 3/4.7.1.1, " Safety Val ve s. " The design pressure of the main steam line code safety valves, as stated in the UFSAR, is 1000 psia and the TS Bases indicates that the secondary system (steam generators and main steam line piping) design pressure is 1000 psig.

BG&E performed a design evaluation and determined that the correct design pressure for the secondary system, which includes the steam generators,_is 1000 psia.

BG&E also proposes that the statement in the TS Bases, which indicates that the as-left lift settings of no less than 985 psig to assure that the lift setpoints will remain within specification during an operating cycle, be deleted.

BG&E indicated that the value is a nominal value which has tolerances and has led to confusion.

BG&E also noted that, in the Unit 2 TS Bases only, the word Vessel was inadvertently omitted in the citation relating to the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code.

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l Mr. Robert E. Denton s The design evaluation supports the proposed revisions to change the design pressure to 1000 psia and that the 985 psig value is a nominal value.

The current TS Bases state that the Code safety valves are to be tested and maintained in accordance with the requirements of Section XI of the ASME Boiler and Vessel Pressure Code, thus deleting the statement relat'ing to the as-left left settings will not detract from the required testing and maintenance.

Therefore, the NRC staff has no objections to the proposed changes including the administrative correction in the Unit 2 TS Bases.

Chanae No. 2

~

This change is proposed to clarify TS Bases Sections 3/4.11.l.3, " Liquid Radwaste Treatment System," and 3/4.11.2.4, " Gaseous Radwaste Treatment System." The request is to the add a sentence to the TS Bases section which will indicate that the dose limits apply to the combined effluent of the plant (e.g., two units).

BG&E noted that the two systems are shared between Units 1 and 2.

BG&E further indicates that the dose limits were approved as the combined effluent of the plant in Amendment No. 105 to facility Operating License No. DPR-53 and Amendment No. 86 to facility Operating License No. DPR-69, which were issued on July 4, 1985.

The amendments referenced by BG&E provided the final NRC action relating to Radiological Effluent Technical Specifications, Offsite Dose Calculation Manual, and the Process Control Program for the Calvert Cliffs site.

The issuance on February 22, 1985, of Amendment Nos. 100 and 82, Units 1 and 2, respectively, was the other NRC action. The supporting Safety Evaluation for these initial amendments, which was also referenced in the final amendments, indicates that the projected dose limits are for both reactors for each of the shared systems discussed above. Therefore, the NRC staff has no objections to the proposed changes.

Chance No. 3 This change proposes to update TS Bases 3/4.7.1.2 " Auxiliary Feedwater System," to more accurately reflect the analysis of the auxiliary feedwater i

system (AFW7 which was performed in support of Amendment No. 149 to facility Operating License No. DPR-53 and Amendment No. 130 to Facility Operating License No. DPR-69 which were issued on December 4, 1990. The current TS' Bases address the impact of AFW flow based on the initiation of the AFW System for the Loss of Feedwater, Feedline Break, and Main Steam Line Break Design l

Basis Events (DBEs)=

BG&E proposes'to replace the existing discussion to indicate that the AFW flow and' response times are conservatively accounted for in all the applicable DBE analyses.

For those DBEs where AFW flow would increase the consequences, a conservative minimum time for initiation is assumed and where AFW flow would decrease the consequences, a conservative maximum time for initiation is assumed.

i

Mr. Robert E. Denton April 26, 1993 s

The NRC staff's Safety Evaluation in support of the referenced amendments indicated that the analysis performed by BG&E was conservative and bounded the DBEs.

Therefore, since the NRC staff's safety evaluation supports the proposed changes, the NRC staff has no objections.

Change No. 4 This request is to correct a typographical error in TS Bases 3/4.7.6, " Control Room Emergency Ventilation System," which currently references 10 CFR Part 50, Appendix A, General Design Criteria 10. The reference should be-General Design Criteria 19.

General Design Criteria 10 provides principal design criteria for reactor design considerations and not control room design considerations.

As noted the correct citation should be General Design Criteria 19, " Control Room."

Therefore, the NRC staff has no objections to the proposed TS Bases change.

Enclosed are the revision instructions and the Revised Technical Specification Bases pages for the Calvert Cliffs, Units 1 and 2, Technical Specifications.

This concludes our actions relating to the referenced TAC numbers.

Sincerely, Daniel G. Mcdonald, Senior Project Manager Project Directorate 1-1 l

Division of Reactor Projects - 1/11 Office of Nuclear Reactor Regulation

Enclosure:

Revision Instructions and the revised Technical Specification Bases pages cc w/ enclosure:

See next page_

Mr. Robert E. Denton Calvert Cliffs Nuclear Power Plant Baltimore Gas & Electric Company Unit Nos. I and 2 cC' Mr. Michael Moore, President Mr. Joseph H. Walter Calvert County Board of Engineering Division Commissioners Public Service Commission of 175 Main Street Maryland Prince Frederick, Maryland 20678 American Building 231 E. Baltimore Street D. A. Brune, Esquire Baltimore, Maryland 21202-3486 General Counsel Baltimore Gas and Electric Company Kristen A. Burger, Esquire P. O. Box 1475 Maryland People's Counsel Baltimore, Maryland 21203 American Building, 9th Floor 231 E. Baltimore Street Jay E. Silberg, Esquire Baltimore, Maryland 21202 Shaw, Pittman, Potts and Trowbridge 2300 N Street, NW Patricia T. Birnie, Esquire Washington, DC 20037 Co-Director Maryland Safe Energy Coalition Mr. G. L. Detter, Director, NRM P. O. Box 33111 Calvert Cliffs Nuclear Power Plant Baltimore,. Maryland 21218 1650 Calvert Cliffs Parkway Lusby, Maryland 20657-47027 Resident Inspector c/o U.S. Nuclear Regulatory Commission P. O. Box 287 St. Leonard, Maryland 206B5 Mr. Richard I. McLean Administrator - Radioecology Department of Natural Resources 580 Taylor Avenue Tawes State Office Building B3 Annapolis, Maryland 21401 Regional Administrator, Region I U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, Pennsylvania 19406

i Enclosure TECHNICAL SPECIFICATION l

BASES REVISIONS l

i UN,I_1_- FACILITY OPERATING LICENSE NO. DPR-53 UNIl 2 - FACILITY OPERATING LICENSE NO. DPR-69 DOCKET NOS. 50-317 AND 50-318 s

Revise Appendix A, Bases Section, as follows:

Remove Paaes Insert Paaes l

B 3/4 7-1 B 3/4 7-1 B 3/4 7-3 B 3/4 7-3 i

B 3/4 7-5 B 3/4 7-5 B 3/4 11-2 B 3/4 11-2 B 3/4 11-4 8 3/4 11-4 i

t i

l l

i 3/4.7 PLANT SYSTEMS l

f BASES 3/4.7.1 TURBINE CYCLE 3/4.7.1.1 Safety Valves The OPERABILITY of the main steam line code safety valves ensures that the Secondary System pressure will be limited to within 110% of its design r

pressure of 1000 psia during the most severe anticipated system operational l

transient. The total relieving capacity for all valves on all of the steam i

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 l

THERMAL POWER coincident with an assumed loss of condenser. heat sink t

(i.e., no steam bypass to the condenser). The main steam line code safety i

valves are tested and maintained in accordance with the requirements of Section XI of the ASME Boiler and Pressure Vessel Code.

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 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 thennal equilibrium with the operating environment.

STARTUP and/or POWER OPERATION is allowable with safety valves inoperable.

l 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 l

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)

}~

}("} x 46.8 SP =

X 4

Changed by NRC letter dated Anril 26. 1993

-l-1 CALVERT CLIFFS - UNIT 1 B 3/4 7-1 i

i 1

3/4.7 PLANT SYSTEMS BASES Auxiliary Feedwater flow and response times a're conservatively accounted for in the analyses of Design Basis Events.

In Main Steam Line Break ard 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, Loss 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.

At 10 minutes after an Auxiliary Feedwater Actuation Signa.1 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 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 tube leak in the steam generator of the affected steam line and 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 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 event the steam line rupture occurs within containment.

The OPERABILITY of Changed by NRC letter dated April 26, 1993 l

CALVERT CLIFFS - UNIT 1 B 3/4 7-3 J

3/4.7 PLANT SYSTEMS BASES 1

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 requirements of 10 CFR Part 50, Appendix A General Design Criteria 19.

l 3/4.7.7 ECCS PUMP 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 insp ction 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 detemined 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 detemine the next inspection. However, the results of 1

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 Changed by NRC letter dated AprH 26.1993 l

CALVERT CLIFFS - UNIT 1 B 3/4 7-5

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 specified, provides assurance that the releases of radioactive materials in l

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 l

Treatment System were specified as a suitable fraction of the dose design objectives set forth in 10 CFR Part 50, Appendix I, Section II.A for liquid effluents. The dose limits apply to the combined effluent'of the plant (e.g., two units).

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 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 i

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. 40, 586-93 (1968), and Hartwell, J.K.,

" Detection Limits for Radioanalytical Counting Techniques," Atlantic Richfield Hanford Company Report ARH-SA-215 (June 1975).

j i

Changed by NRC letter dated April 26.1993 l

CALVERT CLIFFS - UNIT 1 B 3/4 11-2

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 detennining the actual doses based upon the historical annual average atmospheric conditions. The release rate specifications for iodine-131 and radionuclides in particulate fonn 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.

i 3/4.11.2.4 GASE0US 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 CFR Part 50, Appendix I, Sections II.B and II.C for gaseous effluents.

The dose limits apply to the combined effluent of the plant (e.g.,twounits).

3/4.11.2.5 Explosive Gas Mixture This sifecification 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.

Changed by NRC letter dated April 26. 1903 I

q CALVERT CLIFFS - UNIT 1 B 3/4 11-4

3/4.7 PLANT SYSTEMS BASES Auxiliary Feedwater flow and response times a're conservatively accounted for in the analyses of 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, Loss 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.

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 Storage 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 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 tube 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 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 event the steam line rupture occurs within containment. The OPERABILITY of i

i Changed by NRC letter dated Anril 26, 1993 l

CALVERT CLIFFS - UNIT 2 B 3/4 7-3

3/4.7 PLANT SYSTEMS BASES accident conditions. The OPEP/BILITY 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 requirements of 10 CFR Part 50, Appendix A General Design Criteria 19.

l 3/4.7.7 ECCS PUMP 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 an 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 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 detemine 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, j

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 t

Changed by NRC letter dated April 26, 1993 l

CALVERT CLIFFS - UNIT 2 B 3/4 7-5

3/4.11 RADI0 ACTIVE EFFLUENTS BASES 3/4.11.1.3 Liouid Radwaste Treatment System 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 50.36a, of 10 CFR Part 50, 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 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 10 CFR Part 50, Appendix I, Section II.A for liquid effluents. The dose limits apply to the combined effluent of the plant (e.g., two units).

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 in an UNRESTRICTED AREA, exceeding the limits specified in Appendix B.

Table II of 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 tenns 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. 40, 586-93 (1968), and Hartwell, J.K.,

" Detection Limits for Radioanalytical 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 for Operation implements the guides set forth in Appendix I, Changed by NRC letter dated April 26, 1993 l

CALVERT CLIFFS - UNIT 2 B 3/4 11-2

3/4.11 RADIOACTIVE 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 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, 1

when specified, provides reasonable assurance that the releases of radioactive materials in gaseous effluents will be kept "as' low as is reasonably achievable".

a

?

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 CFR Part 50, Appendix I, Sections II.B and II.C for gaseous effluents. The j

dose limits apply to the combined effluent of the plant (e.g. two units).

}

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 limit of oxygen. Maintaining the i

concentration of oxygen below its flammability 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

+

J Design Criterion 60.

3/4.11.2.6 Gas Storage 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 Changed by NRC letter dated April 26, 1903 l

CALVERT CLIFFS - UNIT 2 B 3/4 11-4 4

Mr. Robert E. Denton April 26,1993 The NRC staff's Safety Evaluation in support of the referenced amendments indicated that the analysis performed by BG&E was conservative and bounded the DBEs. Therefore, since the NRC staff's safety evaluation supports the proposed changes, the NRC staff has no objections.

Chance No. 4 This request is to correct a typographical error in TS Bases 3/4.7.6, " Control Room Emergency Ventilation System," which currently references 10 CFR Part 50, Appendix A, General Design Criteria 10. The reference should be General Design Criteria 19. General Design Criteria 10 provides principal design criteria for reactor design considerations and not control room design considerations.

As noted the correct citation should be General Design Criteria 19, " Control Room."

Therefore, the NRC staff has no objections to the proposed TS Base _, change.

i Enclosed are the revision instructions and the Revised Technical Specification Base; pages for the Calvert Cliffs, Units 1 and 2, Technical Specifications.

This concludes our actions relating to the referenced TAC numbers.

Sincerely, Original signed by-Daniel G. Mcdonald, Senior Project Manager Project Directorate 1-1 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation

Enclosure:

i Revision Instructions and the revised Technical Specification Bases pages cc w/ enclosure:

j See next page Di st ribution:

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OFFICIAL RECORD COPY FILENAME: A:\\CC86136.LTR u