ML20216D632

Forwards Response to 980330 RAI Re LAR Change to RCS Flow Requirements to Allow Increased SG Plugging.Response Does Not Change No Significant Hazards Determination Previously Provided

ML20216D632
Person / Time
Site:	Calvert Cliffs
Issue date:	04/08/1998
From:	Cruse C BALTIMORE GAS & ELECTRIC CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
TAC-M97855, TAC-M97856, NUDOCS 9804150362
Download: ML20216D632 (16)

Text

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e Cu Antes 11. CuesE Baltimore Gas and Electric Company Vice President Calvert Cliffs Nuclear Power Plant Nuclear Energy 1650 Calvert Cliffs Parkway Lusby. Maryland 20657 410 495-4455 April 8,1998 U. S. Nuclear Regulatory Commission Washington, DC 20555 ATTENTION: Document Control Desk

SUBJECT:

Calvert Cliffs Nuclear Power Plant Unit Nos.1 & 2; Docket Nos. 50-317 & 50-318 Request for Additional Information - License Amendment Request Change to Reactor Coolant System Flow Requirements to Allow Increased Steam Generator Tube Plugging (TAC Nos. M97855 and M97856)

REFERENCES:

(a) Letter from Mr. A. W. Dromerick (NRC), to Mr. C. H. Cruse (BGE),

dated March 30,1998, Request for Additional Information - License Amendment Request Change to Reactor Coolant System Flow Requirements to Allow Increased Steam Generator Tube Plugging (TAC Nos. M97855 and M97856)

(b) Phone call between NRC Staff and BGE Staff, March 25,1998 (c) Meeting between NRC Staff and BGE Staff, April 7,1998 in response to Reference (a), Attachment (1) contains our response to your questions. Additional information is provided in Attachment (2) to respond to some concerns raised in References (b) and (c).

These responses do not change the No Signincant Hazards Determination previously provided.

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Documrnt Control Desk April 8,1998 Page 2 Should you have questions regarding this matter, we will be pleased to discuss them with you.

Very truly yours,

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STATE OF MARYLAND  : //gg /MF IA

TO WIT:

COUNTY OF CALVERT  :

1, Charles 11. Cruse, being duly sworn, state that I am Vice President, Nuclear Energy Division, Baltimore Gas and Electric Company (BGE), and that I am duly authorized to execute and file this License Amendment Request on behalf of BGE. To the best of my knowledge and belief, the statements contained in this document are true and correct. To the extent that these statements are not based on my personal knowledge, they are based upon information provided by other BGE employees and/or consultants. Such information has been reviewed in accordance with company prac.tief and I believe it to be reliable. /

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/ n b4V/ An1V

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SMbscribed and sworn before me, a Notary Public in and for the State of Maryland and County of CA b>1rd , this 5% day of flhA ' L, .,1998.

'I WITNESS my Hand and Notarial Seal: [6 #/ i' Notary Public 9 My Com nission Expires: dA>> a A4t. /, ACo,1 F bate CIIC/ PSF /bjd Attachments: (1) Response to Request for Additional Information (2) Response to Concerns Raised by the Nuclear Regulatory Commission Staff ac: R. S. Fleishman, Esquire 11. J. Miller, NRC J. E. Silberg, Esquire Resident inspector, NRC Director, Project Directorate I-1, NRC R. I. McLean, DNR A. W. Dromerick, NRC J. II. Walter, PSC 1

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. . l ATTACHNIE.ET (n RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION Baltimore Gas and Electric Company Calvert Cliffs Nuclear Power Plant April 8,1998

, ATTACIIMENT (1)

RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION MAIN STEA'M LINE BREAK ACCIDENT The scenario postulated in the question is correct.

NRCInformation Request The duration ofprimary to secondary leakage to thefaulted steam generator and the timefollowing ths AiSLB at which primary to secondary leakage no longer occurs in thefaulted steam generator.

BGE Response The initial primary activity and failed fuel activity leaks to the affected steam generator at the maximum Technical Specification leakrate of 100 gpd. This activity is immediately released to the environment through the main steam vent for 0<t<9 hours at which time the Reactor Coolant System temperature decreases to less than 212*F and the accident ceases.

NRCInformation Request The duration ofprimary to secondary leakagefor the intact steam generator and the timefollowing the A(SLB at which RilR operation begins.

BGE Response The initial primary cetivity and failed fuel activity leaks to the intact steam generator at the maximum Technical Specification leakrate of 100 gpd. This activity is immediately released to the environment through the atmospheric dump valves (ADVs)/ main steam safety valves (MSSVs) for 0<t<9 hours at which time the Reactor Coolant System temperature decreases to less than 212 F and the accident ceases.

STEAM GENERATOR TUBE RUPTURE (SGTR) ACCIDENT ANALYSIS For the SGTR accident, thefollowing information is provided:

a. For thefaultedsteam generator

1. Afass ofbreakflow @ 0 - 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 304,313 lb

@ 2 - 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 684,575 lb

@ 9 - 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 0

11. Duration ofbreakflow 0 - 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 1

, ATTACHMENT (1)

RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION l Ill. ' Mass ofSteam Released @ 0 - 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> *209,696 lb l @ 2 - 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 339,400 lb

@ 9 - 21 hours2.430556e-4 days <br />0.00583 hours <br />3.472222e-5 weeks <br />7.9905e-6 months <br /> 0

• For the purpose of answering this NRC request, the steam released from the l

faulted steam generator through the ADV until isolation was assumed to be 60%

of the total ADV Dow based on engineering judgn.ent that How from the closer steam generator will encounter less resistance. This assumption is necessary due to the limitations of the Asea Brown Boveri, Inc/ Combustion Engineering Inc.

computer codes. CESEC and COOL do not calculate ADV Dow on a steam generator-specine basis.

IV. Flashing Fraction as afunction of time See Attached table

b. For the intact steam generator I. Mass ofSteam Released @ 0 - 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 131,179

@ 2 - 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 0

@,9 - 24 hoors 0

11. Beginning and ending times of steaming " 0 - 5889 see

" Although unisolation of the ADV for the intact steam generator and isolation of the faulted steam generator is completed by 5889 sec, in order to maximize releases from the faulted steam generator, the faulted steam generator-ADV continues to be used for steaming until 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> in an attempt to control level in the faulted steam generator,

c. Clarification of when releases occur through the condensers and through the ADVs Releases occur through the ADV for the off-site doses and through the condenser for the Control Room analysis; this results in the worst case consequences.

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l ATTACHMENT (1)

RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION Steam Generator Tube Rupture - i Flashing Fraction vs. Time l

Hours Sec - ff Ilours See ff 0 0 0.113 3 10,800 0.033 1689 0.113 11,700 0.036  !

1690 0.046 12,600 0.032 1700 0.044 4 14,400 0.024 0.5 1800 0.052 15,300 0.029 1 2000 0.051 16,200 0.027 2400 0.056 17,100 f

0.024 i 2800 0.057 5 18,000 0.026 1 3600 0.045 19,800 0.025 4400 0.024 6 21,600 0.018 5000 0.034 23,400 9.019 2 7200 0.040 7 25,200 0.025 7800 0.039 27,000 0.024 8400 0.032 28,800 0.022 9000 0.036 9600 0.037 10,000 0.033 in SEIZFD ROTOk ACCIDENT, NRCInformation Request Thefollowing information should be provided:

a. mass ofsteam releasedfrom 0-2 hr and 2-8 hr. ,

b. beginning and ending times ofsteaming.

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c. clarification of when releases occur through the condenser and releases occur through the ADVs. i BGE Response For the Seized Rotor Event, two scenarios were investigated, and the most conservative utilized.

(i) The initial secondary activity from both steam generators is ejected from the ADVs at time t = 0. '

The initial primary activity and failed fuel activity leaks to the two steam generators at the maximum Technical Specification leakrate of 100 gpd for each steam generator. This activity is immediately released to the environment through the ADVs for 0<t<30 minutes and then through the condenser for 30 minutes <t<6 hours at which time shutdown cooling is initiated and the accident ceases.

(ii) The initial secondary activity from both steam generators is ejected from the condensers at time t = 0. The initial primary activity and failed fuel activity leaks to the two steam generators at the maximum Technical Specification leakrate of 100 gpd for each steam generator. This activity is immediately released to the environment through the condensers for 0<t<6 hours at which time shutdown cooling is initiated and the accident ceases.

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, ATTACHMENT (1)

RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION Note that the consequences of a Seized Rotor Event with the flow for the first 30 minutes going out the ADVs and thereafter to the condenser bounds the consequences of an Seized Rotor Event with the flow l

going to the condenser over the entire accident for the site boundary and low population zone analyses.

Also note that the consequences of a Seized Rotor Event with the flow for the first 30 minutes going out l the ADVs and thereafler to the condenser is bounded by the consequences of a Seized Rotor Event with the flow going to the condenser over the entire accident for the Control Room analyses. Thus the ADV- 1 l

I condenser results will be used as the design basis limits for the site boundary and low population zone l while the condenser-condenser results will be used as the design basis limits for the control room results.

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l ATTACHMENT (2) i l

RESPONSE TO CONCERNS RAISED BY THE NUCLEAR REGULATORY COMMISSION STAFF l

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Haltimore Gas and Electric Company Calvert Cliffs Nuclear Power Plant April 8,1998

. ATTACHMENT (3) 1 RESPONSE TO CONCERNS RAISED BY THE NUCLEAR REGULATORY COMMISSION STAFF NRC Concern l The NRC has questioned the consistency of the BGE offsite and control room dose submittals. They are l concerned that the control room operator doses were calculated with one set ofparameters and the ofsite doses with another.

BGE Response Some of the previously submitted work was performed by Baltimore Gas and Electric Company (BGb) and some by Asea Brown Boveri, Inc/Ccmbustion Engineering Inc. (ABB/CE) which could have led to some inconsistency in the assumptions. However, all of the doses (offsite and Control Room) are now calculated by BGE using consistent parameters and methodology except as noted below.

(1) The latest site boundary, low population zone (LPZ), and Control Room doses for the Main Steam l Line Break (MSLB), Seized Rotor, and Steam Generator Tube Rupture (SGTR) events are calculated by BGE with consistent parameters and methodology except as noted below. The results are as follows:

Results The Seized Rotor Event (SRE) results are as follows:

Site Boundary Thyroid Dose 12.0 Rem Whole Body Dose 0.2 Rem LPZ Thyroid Dose 1.0 Rem Whole Body Dose 0.04 Reia Control Room Thyroid Dose 0.166 Rem Whole Body Dose 1.05E-2 Rem Beta Skin Dose 0.244 Rem Note that the site boundary whole body dose exceeds the 0.06 Rem result presented in Reference (1).

The LPZ results are being presented for the first time. The other results are consistent with those presented in References (1) and (4).

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l ATTACIIMENT (2) 1

, RESPONSE TO CONCERNS RAISED BY i TIIE NUCLEAR REGULATORY COMMISSION STAFF l l The SGTR results assuming a Generated Iodine Spike of 500 are as follows:

Site Boundary

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I Thyroid Dose 13.0 Rem i Whole Body Dose 0.55 Rem LPZ j l Thyroid Dose 5.0 Rem j Whole Body Dose 0.15 Rem Control Room Thyroid Dose 0.134 Rem )

Whole Body Dose 7.00E-3 Rem '

Beta Skin Dose Negligible The LPZ results are being presented for the first time. The other results are consistent with those presented in References (3) and (4). All of the inputs for these results are consistent except as noted ,

below, i 1

1 The SGTR results assuming a Pre-Existing lodine Spike of 60 pCi/gm are as follows- l 1

Site Boundary Thyroid Dose 22.0 Rem Whole Body Dose 0.66 Rem LPZ Thyroid Dose 6.0 Rem  ;

Whole Body Dose 0.18 Rem l Control Room Thyroid Dose 8.80E-2 Rem Whole Body Dose 6.20E-3 Rem Beta Skin Dose Negligible These results are being presented for the first time. All of the inputs for these results are consistent except as noted below.

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, ATTACIIMENT (2) l RESPONSE TO CONCERNS RAISED BY l TIIE NUCLEAR REGULATORY COMMISSION STAFF The MSLB results for 1.35% failed fuel are as follows:

J Site Boundary Thyroid Dose 5.0 Rem Whole Body Dose 0.2 Rem LPZ Thyroid Dose 1.2 Rem Whole Body Dose 0.04 Rem l Control Room Thyroid Dose 29.30 Rem Whole Body Dose 0.90 Rem Beta Skin Dose 0.05 Rem The LPZ results are being presented for the first time. The other results are consistent with those presented in Reference (4). All of the inputs for these results are consistent except as noted below.

l Parameters I

The minimum Reactor Coolant System (RCS) fluid mass is 457,437 lbm for the 0-30 day LPZ and

, Control Room calculations and 384,260 lbm for the 0-2 hour site boundary calculations.

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• The mass for the 0-2 hour site boundary calculation is defined as the initial undiluted water mass in the RCS without pressurizer. This can be calculated as:

l Macs = (9601 - 2

• 338.1)

• 45.33 = 404561.2 lbm where the volume of the RCS with no plugs is 9601 R' [ Updated Final Safety Analysis Report (UFSAR) Table 4.1], the volume of the 2500 plugged tubes is 338.1 ft' per steam generator, and the l water density 45.33 lbm/ft' based on 2250 psia (UFSAR Table 4.1) and 572.5'F (UFSAR l Figure 4.9). This value will be conservatively decreased 5% to 384,260 lbm to conform with the value used in the ABB/CE calculation.

• The mass for the 0-30 day LPZ and Control Room calculations is defined as the average coolant i mass in the RCS without pressurizer over the accident. The initial coolant mass is defined above.

l The final coolant mass is the RCS water mass at start of shutdown cooling, which is defined as l 300'F saturation per UFSAR Section 1.2.9.2. Thus the mass can be calculated as: l l

1 Macs = (9601 - 2

• 338.1) * (45.33 + 57.3066) / 2 = 458005.6 lbm l This value will be conservatively decreased to 457,437 lbm to conform with initial RCS mass with

! pressurizer, where the minimum pressurizer volume is 600 R' (UFSAR Table 4-7), the pressurizer l pressure is 2250 psia (UFSAR Table 4-7), and the saturation temperature is 653 F. The specific l

volume at those conditions is 0.02700 ft'/lbm (37.04 lbm/ff). thus the mass is Macs = 9601 *45.33 + 600*37.04 = 457437 lbm 3

ATTACHMENT (2)

RESPONSE TO CONCERNS RAISED BY THE NUCLEAR REGULATORY COMMISSION STAFF The atmospheric dispersion coemeients are as follows:

The site boundary atmospheric dispersion coemeient remains unchanged at 0- 2 hrs 1.3E-04 sec/m'

• The LPZ atmospheric dispersion coemeients are 3

0- 2 hrs 3.30E-05 sec/m 2- 24 hrs 2.20E-06 sec/m' i 24-720 hrs 5.40E-07 sec/m' Atmospheric dispersion coefficients from the condenser / vent stack to the Control Room:

0- 2 hrs 1.08E-03 sec/m' 3

2- 8 hrs 7.99E-04 sec/m I 8- 24 hrs 3.83E-04 sec/m' 24- 96 hrs 2.86E-04 sec/m'96-720 hrs 2.04E-04 sec/m' e Atmospheric dispersion coemeients from the atmospheric dump valve (ADV) to the Control Room:

l' 0- 2 hrs 1.00E-09 sec/m' 2- 8 hrs 1.32E 09 sec/m' 8- 24 hrs 8.80E-10 sec/m' 24 96 hrs 1.00E-09 sec/m' 3

96-720 hrs 1.55E-08 sec/m e Atmospheric dispersion coemeients from the main steam vent to the Control Room:

0- 2 hrs 1.08E-03 sec/m' 3

2- 8 hrs 8.84E-04 sec/m 8- 24 hrs 4.25E-04 sec/m' 24- 96 hrs 3.21E-04 sec/m'96-720 hrs 2.26E-04 sec/m' NRC Concern The NRC has questioned the continuity of the BGE offsite and control room dose submittals. They are concerned that the ca/culations may have been updated without informing the NRC.

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i l , ATTACIIMENT (2) l RESPONSE TO CONCERNS RAISED BY TIIE NUCLEAR REGULATORY COMMISSION STAFF BGE Response Some of the previously submitted work was recalculated using consistent and updated inputs and

, methodologies. The results, listed above, remain unchanged except for an increase in the SRE site l boundary whole body dose. The inputs and assumptions used to derive the above are consistent with the Reference (4) responses. The main parameters are as follows with changes as noted: i e The failed fuel fraction is 5% for Seized Rotor,0% for SGTR, and 1.35% for MSLB. The MSLB l assumption differs from that presented in earlier submittals and was necessitated by General Design Criteria 19 requirements.

• The pretrip power level is 102% (2754 MWt) and is unchanged.

• The power peaking factor is 1.7 and is unchanged.

The dose conversion factors are based on International Commission for Radiation Protection-30 and differ from the earlier International Commission for Radiation Protection-2 values.

• The source terms are based on TID-14844 and are unchanged.

• Fractional gas gap releases are 12% I-131,10% other iodines,30% Kr-85, and 10% other noble i gases and are unchanged.

• The secondary 1-131 activity is 1.E-7 Ci/gm and is unchanged.

• The primary 1-131 activity is 1.E-6 Ci/gm and is unchanged.

• The primary noble gas activity is 1.E-4/Ebar Ci/gm and is unchanged.

• The Control Room volume is 166000 ft' and is unchanged.

• The Auxiliary Building roof above the Control Room will be sealed. Most Control Room inleakage can then be assumed to originate at the Auxiliary Building inlet plenum on the West Road side of the Auxiliary Building. The atmospheric dispersion coefficients are then calculated via ARCON96 (NUREG/CR-6331 Revision 1). The previous submitted values assumed Control Room inleakage through the inlet dampers on the roof oft he Auxiliary Building, e The steam generator mass is 125,250 lbm per steam generator and is unchanged.

• The primary to secondary leakrate is 100 gpd at 572.5 F and 2250 psia. This differs from that reported previously (100 gpd at 70*F) in that the Technical Specification leakage is now calculated at the more restrictive hot conditions.

• The Control Room inflow rate is 3000 cfm at all times. This differs from the earlier submittal (21100 cfm for 0<t<2.1 sec,11005 cfm for 2.1 sec<t<25.4 sec, 910 cfm for 25.4 sec<t), in that updated measured data is utilized. This also assumes that the roof of the Auxiliary Building will be scaled.

• Control room recirculation and filtratian flow:

=> Flowrate: One filter train at ) 800 cfm (formerly 2000 cfm)

=> Initiation delay time: 30 sec (formerly 2.1 sec)

=> Filter efficiencies: 90% for all iodine species (same as before)

The values were changed to conform to Technical Specifications and to be more conservative.

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l ATTACHMENT (2) l RESPONSE TO CONCERNS RAISED BY f

' THE NUCLEAR REGULATORY COMMISSION STAFF l l

1 Time to release secondary activity: All secondary activity released at t = 0. This methodology is I conservative and consistently utilized.

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• The breathing rates are unchanged:

0- 8 hrs 3.47E-04 m3/sec 8- 24 hrs 1.75E-04 m3/sec 24-720 hrs 2.32E-04 m3/sec

. The control room occupancy factors are unchanged:

0 24 hrs 1.0 24-96 hrs 0.6 96-720 hrs 0.4

• The lodine Partition Factors are unchanged:

=> Primary release through ADVs/MSSVs: 1.0

=> Secondary release through ADVs/MSSVs: 1.0

=> Primary release through main steam vents: 1.0

=> Secondary release through main s :am vents: 1.0

=> Condenser partition factor: 0.0005 The release pathways are accident specific and are as follows:

• For the SRE, two scenarios were investigated, and the most conservative utilized.

(i) The initial secondary activity from both steam generators is ejected from the ADVs at time t = 0.

The initial primary activity and failed fuel activity leaks to the two steam generators at the maximum Technical Specification leakrate of 100 gpd for each steam generator. This activity is immediately released to the environment through the ADVs for 0<t<30 minutes and then through the condenser for 30 minutes <t<6 hours at which time shutdown cooling is initiated and the accident ceases.

l (ii) The initial secondary activity from both steam generators is ejected from the condensers at time i t = 0. The initial primary activity and failed fuel activity leaks to the two steam generators at the maximum Technical Specification leakrate of 100 gpd for each steam generator. This activity is immediately released to the environment through the condensers for 0<t<6 hours at which time i shutdown cooling is initiated and the accident ceases.

Note that the consequences of an SRE with the flow for the first 30 minutes going out the ADVs and thereafter to the condenser bounds the consequences of an SRE with the flow going to the condenser over the entire accident for the site boundary and LPZ analyses. Also note that the consequences of an SRB with the flow for the first 30 minutes going out the ADVs and thereafter to the condenser is bounded by the consequences of an SRE with the flow going to the condenser over the entire accident for the Control Room analyses. Thus the ADV-condenser results will be used as the design basis limits for the site boundary and LPZ, while the condenser-condenser results will be used as the design basis limits for the Control Room results.

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' A'ITACHMENT (3) I 1

RESPONSE TO CONCERNS RAISED BY THE NUCLEAR REGULATORY COMMISSION STAFF e For the MSLB, the following scenario was assumed:

l The initial secondary activity from both steam generators is conservatively ejected from the main I steam vent at time t = 0. The initial primary activity tnd failed fuel activity leaks to the affected steam generator at the maximum Technical Specification leakrate of 100 gpd. This activity is immediately released to the environment through the main steam vent for 0<t<9 hours at which time the RCS temperature decreases to less than 212'F and the accident ceases. The initial primary l activity and failed fuel activity leaks to the intact steam generator at the maximum Technical l Specification leakrate of 100 gpd. This activity is immediately released to the environment through l the ADVs/MSSVs for 0<t<9 hours at which time the RCS temperature decreases to less than 212'F l and the accident ceases.

• For the SGTR Event, two scenarios were investigated, and the most conservative utilized.

(i) The initial secondary activity from both steam generators is ejected from the ADVs at time t = 0.

l The initial primary activity including the Generated Iodine Spike or Pre-Existing lodine Spike l activity that escapes to the steam generators is calculated by the ABB/CE code CESEC. This l activity escapes via the ADVs to the environment. Code CESEC calculates that the accident and releases cease eight hours after initiation.

l l (ii) The initial secondary activity from both steam generators is ejected from the condenser at time t = 0 The initial primary activity including the Generated lodine Spike or Pre Existing lodine Spike activity that escapes to the steam generators is calculated by the ABB/CE code CESEC.

l This activity escapes via the condenser to the environment. Code CESEC calculates that the l accident and releases cease eight hours after initiation.

Note that the consequences of an SGTR with the flow going out the ADVs bounds the consequences l of an SGTR with the flow going to the condenser for the site boundary and LPZ analyses. Also note I

that the consequences of an SGTR with the flow going out the ADVs is bounded by the consequences of an SGTR with the flow going to the condenser for the Control Room analyses.

, Thus, the ADV results will be used as the design basis limits for the site boundary and LPZ, while l the condenser results will be used as the design basis limits for the Control Room results.

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NRC Concern l

The NRC requested clarification concerning the timing of a reactor trip, turbine trip and loss of offsite l power during the main steam line break accident.

BGE Response 1

l For the MSLB accident, the reactor trip, turbine trip, and loss of offsite power are assumed to occur concurrently.

NRC Concern I

The NRC requestedfurther discussion concerning the proposed schedulefor the modifications to the roof of the Ataillary Building needed to limit inleakagefrom the roofarea during secondary side accidents.

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, ATTACIIMENT (2)

RESPONSE TO CONCERNS RAISED BY TIIE NUCLEAR REGULATORY COMMISSION STAFF BGE Response A number of activities are paposed to eliminate air infiltration for the area of the Auxiliary Building roof into the Control Room IIVAC system and the Control Room }{VAC equipment Room. The modifications are necessary not only to limit the rate of infiltration into the Control Room fleating, Ventilation, and Air Conditioning (liVAC) system, but to climinate the source of infiltration from the roof area, and therefore allow using a more favorable atmospheric dispersion factor or meet the General Design Criteria 19 rec,airements. These activities include:

e Isolate the Control Room IIVAC system roof penetration inlet and exhaust ducts.

• Isolate the smoke cemoval system.

• Remove the Access Control Area IIVAC units and relocate them. I

• Seal all openings between the roof and the Control Room flVAC equipment room. I e Remove the Control Room IIVAC equipment room ventilating system.

• Remove the associated ducts penetrating the roofline and seal the openings.

The scope of these modifications has recently been defined. Engineering work on the modifications will be completed this fall. The engineering work is rather extensive because of the removal and relocation of a number of components. Once engineering is complete, construction will start. The llVAC equipment room is very crowded with equipment, ductwork, and tubing. Due to the proximity of the Control Room }{VAC equipment to the work location, we will proceed with the demolition and construction work at a deliberate and well planned pace. Additionally, work will stop during the spring 1999 refueling outage. Other outage activities may require that all work in the area of the Control Room IIVAC system be suspended for the duration of the outage to ensure that the Control Room IIVAC system remains operable. Additionally, outage activities are very labor intensive and workers from this project will be needed in other areas. Work on the roof sealing modification will continue following the outage and will be completed before the end of 1999.

REFERENCES

1. Letter from Mr. C.11. Cruse (BGE) to Document Control Desk (NRC), dated January 31,1997, License Amendment Request; Change to Reactor Coolant System Flow Requirements to Allow Increased Steam Generator Tube Plugging i

2. Letter from Mr. C.11. Cruse (BGE) to Document Control Desk (NRC), dated March 25,1997, Seccnd Request for Additional Information: License Amendment Request; Change to Reactor Coolant System Flow Requirements to Allow Increased Steam Generator Tube Plugging

3. Letter from Mr. C.11. Cruse (BGE) to Document Control Desk (NRC), dated January 22,1998, Supplementary Responses to the April 22 and July 25, 1997, Requests for Additional Information: License Amendment Request; Change to Reactor Coolant System Flow Requirements to Allow increased Steam Generator Tube Plugging j

4. Letter from Mr. C.11. Cruse (BGE) to Document Control Desk (NRC), dated March 17,1998,  !

Response to Request for AdditionalInformation: Control Room liabitability Analysis and Main  !

Steam Line Break Analysis j l

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