ML20217D292

Responds to Request for Addl Info Re LAR Change to Reactor Coolant Sys Flow Requirements to Allow Increased SG Tube Plugging

ML20217D292
Person / Time
Site:	Calvert Cliffs
Issue date:	04/21/1998
From:	Cruse C BALTIMORE GAS & ELECTRIC CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
TAC-M97855, TAC-M97856, NUDOCS 9804240262
Download: ML20217D292 (40)

Text

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CHARLES II. CROSE Baltimore Gas and Electric Company Vice President Calvert Cliffs Nuclear Power Plant

' , Nuclear Encrgy 1650 Calven Cliffs Parkway Lusby, Maryland 20657 410 495-4455 April 21,1998 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 frcm Mr. C. H. Cruse (BGE) to NRC Document Control Desk, dated January 31, 1997, License Amendment Request; Change to Reactor Coolant System Flow Requirements to Allow Increased Steam Generator Tube Plugging (b) Meeting between NRC Staff and BGE Staff, April 7,1998 (c) Phone call between NRC Staff and BGE Staff, April 17,1998 By Reference (a), Baltimore Gas and Electric Company submitted a license amendment request to the Nuclear Regulatory Commission to support operation of Calvert Cliffs Units 1 and 2 with up to 2500 steam generator tubes plugged in each steam generator. This letter responds to the additional requests made in References (b) and (c) to:

1. Provide a Control Room dose assessmentfor the current plant condition to demonstrate compliance with respect to GDC 19 using compensatory measures, and

2. Clarify the use of breathing rates in conjunction with occupancyfactors.

Attachment (1) contains our responses to your questions. These responses do not change the No Significant Hazards Determination previously provided.

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4 Document Control Desk April 21,1998 Page 2 Should you have questions regarding this matter, we will be pleased to discuss them with you.

Very truly yours, j! /

STATE OF MARYLAND  :

TO WIT:

COUNTY OF CALVERT  :

1, Charles H. 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 practice and I believe it to be reliable. / j, / f

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Su 'scri d and sworn before me, a Notary Public in -and for the State of Maryland and County of LXAfJ , this &/4Lday of /LbAAll ,1998.

I WITNESS my Hand and Notarial Seal: f A4Mr ) b 2 Notary Public My Commission Expires: S!l!hO~O1 bate CHC/GT/bjd

Attachment:

(1) Response to the Request for Additional Information From Meeting with NRC on l April 7,1998 and Phone Call on April 17,1998 l Enclosure (A)-Maximum Hypothetical Accident / Loss-of-Coolant Accident l Analysis Enclosure (B)- Fuel Handling Accident Analysis Enclosure (C)- Main Steam Line Break cc: R. S. Fleishman, Esquire H. J. Miller, NRC J. E. Silberg, Esquire Resident inspector, NRC Director, Project Directorate 1-1, NRC R. I. McLean, DNR A. W Dromerick, NRC J. H. Walter, PSC

o ATTACHMENT (1)

RESPONSE TO THE REQUEST FOR ADDITIONAL INFORMATION FROM MEETING WITH NRC ON APRIL 7,1998 AND PHONE CALL ON APRIL 17,1998 Baltimore Gas and Electric Company Calvert Cliffs Nuclear Power Plant April 21,1998

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ATTACHMENT (1)

RESPONSE TO TIIE REQUEST FOR ADDITIONAL INFORMATION FROM MEETING WITH NRC ON APRIL 7,1998 AND PHONE CALL ON APRIL 17,1998 Question / Request Provide a Control Room dose assessmentfor the current plant condition to demonstrate compliance with respect to GDC 19 using the compensatory measures ofReference 1.

RESPONSE

Baltimore Gas and Electric Company has completed Control Room dose assessments for the Maximum Hypothetical Accident (MHA), the Fuel llandling Accident (FHA), and the Main Steam Line Break (MSLB) Accident using inputs and assumptions that bound the current plant condition. In these

- analyses, no credit is taken for the plant modifications planned for the future that were discussed in Reference 2. These analyses assume a Control Room inleakage of 5500 cfm which far exceeds the i maximum inleakage recently measured during testing in February 1998 (3400 cfm). These analyses also conservatively assume the location of the inleakage is at the Control Room inlet or exhaust damper with respect to determination of the atmospheric dispersion coefficients. The Reactor Coolant System mass used is consistent with that described in Reference 3. The FHA in Containment assumes a release rate of l

one Containment volume in two hours. All other assumptions are identical to those used in Reference 4 for MHA and FHA, and in Reference 5 for MSLB.

l l The MHA, FRA, and MSLB are the limiting accidents with respect to Control Room habitability for l Calvert Cliffs. The Control Room operator doses for these accidents are significantly higher than for any

! other event as demonstrated in References 3,4, and 5. Additionally, the manner in which Control Room l operators identify the need and implement the compensatory measures of Reference 1 is similar for all of the events described in References 3,4, and 5. Radiation Monitoring System alarms in each of these events would cause operators to enter appropriate Emergency Response Plan Implementing Procedures l

that require consideration of the compensatory measures of Reference 1. If needed, these measures i would be implemented within 45 minutes in all cases.

! A description of the Control Room dose assessments for the current plant condition are provided in Enclosures A, B, and C for the MHA, FHA, and MSLB respectively. The results of these analyses demonstrate that operators have a minimum of 45 minutes to implement the compensatory measures of Reference 1. As previously reviewed and approved for Calvert Cliffs on an interim basis in Reference 6, these analyses demonstrate compliance with respect to General Design Criteria 19.

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ATTACHMENT (1)

RESPONSE TO TIIE REQUEST FOR ADDITIONAL INFORMATION FROM MEETING WITII NRC ON APRIL 7,1998 AND PIIONE CALL ON APRIL 17,1998 Question / Request Clarify the use ofbreathing rates in conjunction with occupancyfactors.

RESPONSE

The Control Room occupancy factors are:

0 - 24 hrs 1.0 24 - 96 hrs 0.6 96 - 720 hrs 0.4 3

Note that the Control Room breathing rate is 3.47E-04 m /sec multiplied by the occupancy factor.

The offsite breathing rates are:

3 0 - 8 hrs 3.47E-04 m /sec 8 - 24 hrs 1.75E 04 m'/sec 24 - 720 hrs 2.32E-04 m'/see The occupancy factors do not apply here.

REFERENCES

1. Letter from Mr. C. II. Cruse (BGE) to NRC Document Control Desk, dated May 6,1993, Control Room liabitability - Interim Engineering Analysis for Thyroid Dose

2. Meeting between NRC Staff and BGE Staff, April 7,1998

3. Letter from Mr. C.11. Cruse (BGE) to NRC Document Control Desk, dated April 8,1998, Request for Additional Information - License Amendment Request Change to Reactor Coolant System Flow Requirements to Allow Increased Steam Generator Tube Plugging

4. Letter from Mr. C.11. Cruse (BGE) to NRC Document Control Desk, dated April 9,1998, Response to Request for Additional Information: Accident Dose Analysis and Control Room Habitability Analyses for the Maximum flypothetical Accident, Fuel lianding Accident, and Control Element Assembly Ejection Event

5. Letter from Mr. C. II. Cruse (BGE) to NRC Document Control Desk, dated March 17,1998, Response to Request for Additional Information: Control Room Habitability Analyses and Main Steam Line Break Analyses

6. Letter from Mr. D. G. Mcdonald (NRC) to Mr. R. E. Denton (BGE), dated June 22,1995, Control Room liabitability Interim Analysis for Thyroid Dose, Calvert Cliffs Nuclear Power Plant Unit Nos. I and 2 1

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9 ENCLOSURE (A) TO ATTACHMENT (1)

MAXIMUM HYPOTHETICAL ACCIDENT /

LOSS-OF-COOLANT ACCIDENT ANALYSIS (FOR CURRENT PLANT CONDITION)

Baltimore Gas and Electric Company Calvert Cliffs Nuclear Power Plant April 21,1998

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'. I ENCLOSURE (A)

MAXIMUM HYPOTHETICAL ACCIDENT /

LOSS-OF-COOLANT ACCIDENT ANALYSIS INTRODUCTION Section 14.24 of the Updated Final Safety Analysis Report (UFSAR) describes a Maximum Hypothetical Accident (MHA) which involves a gross release of fission products from the fuel to the Containment.

During an accidental release, air containing radionuclides may enter the Control Room through inleakage into the Control Room ventilation system. The Control Room thyroid, whole body, and beta skin doses must meet 10 CFR Part 50, Appendix A, General Design Criteria (GDC) 19 limits. Similarly, during an accidental release, air containing radionuclides may travel offsite. The offsite thyroid and whole body doses must meet 10 CFR Part 100 guidelines.

! Reference 1 requires the inclusion of the results of dose calculations from the various fission product release paths to the atmosphere. These release paths include the containment leakage path, including l contributions through the four-inch hydrogen purge line and contributions from post-loss-of-coolant accident leakage from engineered safety feature systems outside Containment [the refueling water tank (RWT) release path and the Emergency Core Cooling System (ECCS) pump room release path). The likelihood that the hydrogen purge valves are open when the accident occurs is very small. In addition, the four-inch hydrogen purge line is assumed to close before activity release from fuel damage could occur. This will effectively eliminate the hydrogen purge line as a pathway. Due to the implementation of TMI Action Item III.D.I.1, no leakage is assumed from the ECCS into the Auxiliary Building. A leakage reduction program is implemented (Technical Specification 6.5.3), which allows us to neglect this pathway in the analysis. The release paths that are used in the calculation are described in this Attachment.

During an accidental release, airborne and sump radionuclides may also affect the Control Room through radiation penetrating (shine) through the Containment walls. The Control Room doses from airborne radioactivity inside and outside the Control Room as well as direct shine from all radiation sources must be included in the Control Room habitability analyses. This portion of the analysis is discussed later in l this attachment. j DOSES DUE TO CONTAINMENT LEAKAGE AND REFUELING WATER TANK LEAKAGE l

J METHOD OF ANALYSIS l l I A design-basis MHA occurs at time t = 0, instantaneously releasing 100% of the noble gases and 25% of the iodines to the Containment atmosphere, and 50% of the iodines to the Containment sump. After a minimum of 36 minutes, recirculation actuation signal occurs. During the recirculation phase, sump water is recirculated through the ECCS pumps and could leak through various valves and reach the RWT. The two pathways include the two valves in series in the minimum flow recirculation line header and the valve from the containment spray pumps. Thus, there is a potential for an unmonitored release l pathway resulting from the post-loss-of-coolant accident leakage of isolation valves in the safety (

l injection or containment spray system recirculation lines to the RWT, which is vented directly to the l l atmosphere. Refueling water tank leakage to the environment coupled with ARCON96 atmospheric l dispersion coefficients produce activity concentrations at the site boundary, low population zone, and l Control Room inlet. Time-dependent Control Room inleakage and damper flow allow activity to penetrate into the Control Room, which is subsequently cleaned-up by time-dependent Control Room filtration. The Control Room, site boundary, and low population zone activities result in thyroid, whole body, and beta skin doses based on the latest International Commission for Radiation Protection-30 dose 1 conversion factors, time-dependent breathing rates and occupancy factors, and a 5500 cfm Control Room l inleakage.

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ENCLOSURE (A)

MAXIMUM IiYPOTIIETICAL ACCIDENT /

LOSS-0F-COOLANT ACCIDENT ANALYSIS This was accomplished via three distinct methodologies:

. An EXCEL spreadsheet was developed to calculate the iodine partition coefficient via NUREG/CR-5732.

e The AX2 code (which is a variant of AXIDENT and TACT-5) was executed to determine the 30-day site boundary, low population zone, and Control Room doses.

• Atmospheric dispersion coefficients from the release point to the Control Room were calculated via the ARCON96 computer code.

ASSUMPTIONS The following assumptions were utilized in this work:

• No credit is taken for deposition of the plume on the ground or decay ofisotopes in transit to the site boundary.

• Build-up of daughter nuclides is not taken into account as source term nuclides decay per Reference 2.

• This methodology assumes uniform containment leakage over the external Containment surface and neglects leakage internal to the Auxiliary Building, which would be subsequently filtered.

. This work also utilizes the more conservative Unit 1 atmospheric dispersion coefficients.

• A design-basis MHA occurs at time t = 0 instantaneously releasing 50% of the iodines and 100% of the noble gases uniformly into Containment. Fifty-percent of these iodines immediately plateout on Containment surfaces.

The input data to determine the Control Room dose from a MHA are the following:

ARCON96 INPUTS The X Q / was determined for the Containment to Control Room flow path using the following inputs:

. Height oflower wind instrument: 10 meters e Height of upper wind instrument: 60 meters e Wind speed units type: meters /sec e Release type: vent

• Release height: 14.17 meters The release point is assumed to be at the 91.5' roof elevation. H = 91.5' - 45.0' = 46.5' =

14.17 meters 2

e Building area: 1155 m 2

The calculation of Containment cross-sectional area yields 12435.63 ft above the rooftop level 2

of elevation 91'6". The Auxiliary Building cross-sectional area is calculated to be 1938.93 f1, For a west-to-east wind direction, the total cross-sectional area of the Auxiliary Building and the two Containments is 26810 ft 2. For an east-to-west wind direction, the total cross-sectional area 2

of the Turbine Building is 27167 ft . For a north-to-south and south-to-north wind direction, the 2

total cross-sectional area of the Containment and the Turbine Building is 21016 ft . The cross-2 2 sectional area of a single Containment of 12435.63 ft or 1155 m will conservatively be used.

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ENCLOSURE (A)

MAXIMUM IIYPOTHETICAL ACCIDENT /

LOSS-OF-COOLANT ACCIDENT ANALYSIS

• Effluent vertical velocity: 0 meters /sec

. Stack or vent flow: 0 m'/s e Stack or vent radius: 20.96 meters r = 137.5'/2 = 68.75' = 20.96 meters

• Direction to source: Unit 1 - 265 degrees; Unit 2 - 170 degrees

• Source window: 90 degrees e Distance from source to receptor: 44.62 meters (Unit 1),54.84 meters (Unit 2) l

• Intake height: 15.62 meters l 91.5' + 4.75' - 45' = 51.25' = 15.62 m where 91.5' is the height of the Auxiliary Building roof,4.75' is the Control Room exhaust height, and 45' is ground level e Grade elevation difference: 0 meters (Reference 3) e Surface roughness length: 0.1 meters (Reference 3) l

• Minimum wind speed: 0.5 meters /sec (Reference 3) l l

• Sector averaging constant: 4 (Reference 3)

• Hours in average: 12 4 812 24 96168 360 720 (Reference 3) e Minimum number of hours: 12 4 81122 87152 324 648 (Reference 4)

+ Horizontal diffusion coefficient: 9.75 meters The horizontal diffusion coefficient is defined as the Containment radius divided by 2.15 o y= r/2.15 = 20.96/2.15 = 9.75 meters (Reference 3) e Vertical diffusion coefficient: 22.62 meters o,= (193.4'-45')/2 = 74.2' = 22.62 meters (Reference 3) e Note that this methodology assumes uniform containment leakage over the external Containment surface and neglects leakage to the Auxiliary Building, which would be subsequently filtered. This analysis also utilizes the more conservative Unit I atmospheric dispersion coefficients.

These inputs are substituted when calculating the contribution due to leakage from the RWT.

• Release height: 14.99 meters Z = 94'2" - 45' = 49'2" = 14.99 meters

• Direction to source: 260 degrees (Unit 1); I 83 degrees (Unit 2)

• Distance from source to receptor: 71.5 meters (Unit 1),79.00 meters (Unit 2) e Horizontal diffusion coefficient: 0 meters e Vertical diffusion coefficient: 0 meters AX2 INPUTS The inputs into the AX2 code are as follows:

• Initial thermal power is 2754 MWt. (UFSAR Section 3.2.1 and Reference 4) e The Control Room volume of 166000 ft' is extracted from Reference 5.

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4 ENCLOSURE (A)

MAXIMUM IIYPOTHETICAL ACCIDENT /

LOSS-0F-COOLANT ACCIDENT ANALYSIS

• Contaimnent air recirculation and cooling system:

> Time delay: 3 sec [ safety injection actuation signal (SIAS)] + 35 sec (response) = 38 sec (Reference 6, UFSAR Table 7-4)

> 2 containment cooling units (2*55000 = 110000 cfm) (UFSAR Section 6.5.3)

• Containment volume:

> Net free volume: 1.989E+06 ft'(UFSAR Table 14.20-3)

> Containment sprayed volume: 1.447E+06 ft' Volume fraction: 0.7273

> Containment unsprayed volume: 0.542E+06 ft' Volume fraction: 0.2727

• Fission product inventories released to containment:

> Noble gases 100 %

> lodines 25% (References 7 and 8)

Note that the iodine release assumes 50% release and 50% subsequent plateout.

• The Control Room occupancy factors are extracted from References 5 and 9:

0- 24 hrs 1.0 24- 96 hrs 0.6 96-720 hrs 0.4 e Containment spray:

> Time delay: 3 sec (containment spray actuation signal) + 67 see (response) = 70 sec (Reference 6)

> Refueling water tank contains dissolved boric acid of 5 pli at 80 F (UFSAR Section 6.4.2)

> Sump contains dissolved boric acid nd trisodium phosphate of pl{ 7 (UFSAR Section 6.4.2)

> Elemental iodine spray removal coefficient: (UFSAR Section 6.4.2 and Reference 10)

LS = 6*K*T*F*(0.63)*(0.839)*(60 min /hr)/(VSR*D) = 13.7422 / hr K = Gas phase mass transfer coefficient = 13.2 ft/ min T = Fall time = 100 ft/735 ft/ min = 0.136 min F = Spray pump volume flow rate = 167 cfm (UFSAR Table 14.20-6)

VSR = Sprayed region volume = 1.447E+06 ft' D = Spray drop mass linear diameter = 2.87E-03 ft

> Elemental iodine spray decontamination factor:

DF = 1 + VS*ll/VC = 23.73 VS = Containment sump liquid volume = 68329 ft' VC = Cor.tainment net free volume = 1.989E+06 - 68329 = 1.9207E+06 ft' 11 = Effective iodine partition coefficient (References 11 and 12)

= Ratio ofiodine concentration in aqueous phase to gas phase

= exp(6.46+3.25E-06*t) = 639 for t = 0 see LS' = LS*VSR/VC = Containment average spray removal coefficient = 10.3531/hr

! T = Effective spray time = Ln(DF)/LS' = 0.3059 hr = 1101 sec 4

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ENCLOSURE (A) l MAXIMUM HYPOTIIETICAL ACCIDENT /

LOSS-0F-COOLANT ACCIDENT ANALYSIS

> Particulate iodine spray removal coefficient / decontamination factor:

LP = 1.5*(60 min /hr)*h*F*GJD)/VSR = 3.17/hr DF<50

= 0.317/hr DF>50 (Reference 10) h = fall height = 100 feet (UFSAR Section 6.4.2)

F = Spray pump volume Gow rate = 167 cfm (UFSAR Table 14.20-6)

VSR = Containment volume sprayed region = 1.447E+06 ft' E/D = Ratio of a dimensionless collection efficiency E to the average spray drop diameter D (Reference 10)

= 10/m = 3.048/ft for DF<50.

= 1/m = 0.3048/ft for DF>50.

LP' = LP*VSR/VC = Containment average spray removal coefGeient = 2.3852 /hr T = Effective spray time = Ln(DF)/LP' = 1.6401 hr = 5904 sec e Containment plateout:

No credit is taken for plateout other than the initial 50% (References 7 and 8) e Containment leak rate:

0-24 hrs: 0.2%/ day = 2.3148E-08/sec = 2.7625 cfm (UFSAR Section 14.24.3)24-720 hrs: 0.1%/ day = 1.1574E-08/sec = 1.3812 cfm (Reference 8) e Control Room inleakage: Control Room inleakage is assumed to be 5500 cfm. This bounds the current plant condition.

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• The site boundary X /Q of 1.30E-4 sec/m was extracted from UFSAR Section 2.3.6.

• The low population zone (2 miles) atmospheric dispersion coefficients were extracted from UFSAR Figure 2.3-3 and UFSAR Section 14.24.3:

0- 2 hrs 3.30E-05 sec/m' 3

2- 24 hrs 2.20E-06 sec/m 24-720 hrs 5.40E-07 sec/m'

• Atmospheric dispersion coefficients from Containment to the Control Room: (Reference 3)

Unit i Unit 2 3

0- 2 hrs 3.72E-04 sec/m' 3.06E-04 sec/m 3

2- 8 hrs 2.33E-04 sec/m 2.03E-04 sec/m' 3 3 8- 24 hrs 9.03E-05 sec/m 8.69E-05 sec/m 24- 96 hrs 7.75E-05 sec/m' 6.35E-05 sec/m'96-720 hrs 6.36E-05 sec/m' 5.24E-05 sec/m' l e Containment iodine removal system:

> Time delay : 3 (SIAS) + 10 (emergency diesel generator startup) + 20 (sequencer) = 33 sec (Reference 6 and UFSAR Table 8-7)

20 minutes for the second filter (UFSAR Section 14.24.2)

> Flow capacity: 18000 cfm each (Technical SpeciGcation 3.6.3)

> Filter efficiencies: 90% for elemental iodine 90% for particulate iodine 30% for organic iodine (Reference 2)

> Elemental and particulate iodine filter removal coefficients:

LF = (18000 cfm)*(60 min /hr)*0.9/(1.989E+06 ft') = 0.4887/hr = 1.3575E-4/see per filter 5

ENCLOSURE (A)

MAXIMUM IIYPOTIIETICAL ACCIDENT /

l LOSS-OF-COOLANT ACCIDENT ANALYSIS

> Organic iodine filter removal coefficient:

LF = (18000 cfm)*(60 min /hr)*0.3/(1.989E+06 ft') = 0.1629/hr = 4.5249E-5/sec per filter i

> Iodine decontamination factor: None (Reference 8) 1

• Control Room recirculation and filtration flow:

> Flow rate: One filter train at 1800 cfm (Technical Specification 3.7.6) J l > Initiation delay time: 30 sec [(Activates automatically on a SIAS (UFSAR Table 8-7)] 3 sec l (SIAS) +10 sec (emergency diesel generator startup) = 13 see (Conservatively 30 see will be l assumed.)

> Filter efficiencies: 90% for all iodine species (Technical Specification 3.7.6) i > Iodine filter removal coefficients:

3 l LF=(1800 cfm)*(60 min /hr)*0.90/(166000 ft ) = 0.5855/hr = 1.6265E-04/sec l l

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• Iodine chemistiy:

Ele' mental 91 %

Particulate 5%

Organic 4% (Reference 8) l e The isotopic source terms (Ci/MWt) were extracted from Reference 13 and are consistent with l TID-14844 methodology (Reference 7). The isotopic decay constants (1/sec) were also extracted from Reference 13.

SOURCE DECAY lsotope Ci/MWt 1/sec I-131 2.508E+04 9.976E-07 1-132 3.806E+04 8.425 E-05 l l-133 5.622E+04 9.211E-06 1-134 6.575E+04 2.200E-04 1-135 5.103 E+04 2.912E-05 XE-131M 2.595E+02 6.815E-07 XE-133M 1.384E+03 3.663E-06 XE-133 5.622E+04 1.528E-06 XE-135M 1.557E+04 7.380E-04 XE-135 5.363E+04 2.115E-05 XE-137 5.103E+04 3.024E-03 XE-138 4.775E+04 8.151E-04 KR-83M 4.152E+03 1.052E-04 KR-85M 1.297E+04 4.297E-05 KR-85 4.102E+02 2.054E-09 KR-87 2.335E+04 1.514E-04 KR-88 3.200E+04 6.731E-05 l

KR-89 3.979E+04 3.632E-03 l

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O-t ENCLOSURE (A)

MAXIMUM HYPOTHETICAL ACCIDENT /

LOSS-0F-COOLANT ACCIDENT ANALYSIS e

International Commission for Radiation Protection-30 dose conversion factors are listed below.

INilALATION SUBMERSION SUBMERSION Rem /Ci Rem-m'/Ci-sec Rem-m'/Ci-sec Isotope TilYROID WilOLE BODY TilYROID WilOLE BODY BETA SKIN I-131 1.lE+06 3.3 E+04 I-132 6.3 E+03 3.4E+02 I-133 1.8E+05 5.6E+03 1-134 1.lE+03 1.lE+02 1-135 3.l E+04 1.lE+03 XE-131M 1.3 E-03 1.8E-02 XE-133M 5.4E-03 3.8E-02 XE-133 7.3 E-03 6.3E-03 2.0E-02 XE-135M 7.7E-02 1.1E-01 XE-135 4.7E-02 1.2E-01

{

XE-138 2.0E-01 2.0E-01 4.lE-01 KR-83M 3.7E-06 1.8E-04 KR-85M 3.lE-02 3.0E-02 8.5E-02 )

KR-85 4.7E-04 4.8E-02 KR-87 1.4E-01 1.5E-01 5.2E-01 KR-88 3.8E-01 3.7E-01 5.4E-01

• The isotopic gamma energies and fractions are detailed in Reference 14.

• The energy-dependent total and energy absorption coefficients are detailed in Reference 14.

These inputs are substituted when calculating the contribution due to leakage from the RWT.

• Fission product inventories released to containment:

Noble Gas 0%

lodines 50 %

• Containment spray: Not credited e Containment plateout: Not credited e Primary containment leak rate to the RWT:

> Based on actual measurements which show that the leakage through MOV659/660 and S1459 are less than 200 cc/hr, a very conservative leakage rate of 12500 cc/hr or 0.054 gpm is assumed for the combined MOV659/660 and SI459 pathways.

> LR1 = (0.054 gpm) * (min /60 sec) / (7.4805 gal /ft') / (68329 ft')= 1.7587E-09/sec Note that recirculation actuation signal will not start until 36 minutes. (UFSAR Sections 6.3.3 and 6.4.2).

e in order to determine the amount of RWT release to the atmosphere, the breathing of the RWT l

through the vent due to diurnal temperature variations was calculated. The breathing was determined by performing a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> transient thermal analysis of the RWT using the computer code TSAP and l

) assuming RWT atmospheric temperature characteristics based on the maximum solar load during I summer solstice and a minimum ambient night time temperature.

LR2 = (4.18 cfm) * (min /60 sec) / (52109.75 ft') = 1.336922E-6/sec 1

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ENCLOSURE (A)

MAXIMUM HYPOTHETICAL ACCIDENT /

l LOSS-0F-COOLANT ACCIDENT ANALYSIS l

This is decreased by the ratio of volatile iodine in the RWT atmosphere to RWT total to yield:

LR2 = 4.00E-07/sec l where the partition coefficient is calculated via NUREG/CR-5732.

• Atmospheric dispersion coefficients from the RWT to the Control Room: (Reference 3)

Unit 1 Unit 2 l 0- 2 hrs 1.26E-03 sec/m' l.06E-03 sec/m' l 2- 8 hrs 9.96E-04 sec/m' 9.23E-04 sec/m' 3

I 8- 24 hrs 3.84E-04 sec/m' 4.14E-04 sec/m l 24- 96 hrs 3.18E-04 sec/m' 2.87E-04 sec/m' l 96-720 hrs 2.43E-04 sec/m' 2.32E-04 sec/m'

e Containment iodine removal system

Not Credited a lodine chemistry in Containment: Per NUrd.0/CR-5732 for a pli of 7 COMPUTER CODES AX2 CODE METHODOLOGY The AX2 computer code calculates individual gamma and beta whole body and thyroid doses to personnel in the Control Room per 10 CFR Part 50, Appendix A, GDC 19 resulting from any postulated i accident which releases radioactivity within the Containment or within any primary system. Computer l Code AX2 models the transport of radioactivity (elemental, particulate, and organic iodine isotopes and l krypton and xenon isotopes) from the sprayed and unsprayed regions of a primary Containment, through l the secondary Containment if any, and then to the environment and to the Control Room. The Code j includes the capability to model time-dependent activity release; containment spray, filtration, and leakage; Control Room filtration and inleakage; primary and secondary containment purge filters; Control Room intake filters; atmospheric dispersion; and natural decay. Doses are calculated for individuals residing in the Control Room.

The MHA in Containment model is constructed assuming that a design-basis MHA occurs at time t = 0 instantaneously releasing 50% of the iodines and 100% of the noble gases uniformly into Containment.

Fifty percent of these iodines immediately plateout on Containment surfaces. The Containment j clemental, particulate, and organic iodine activities are reduced by time-dependent sprays, filtration, and  ;

isotopic decay, while the noble gases only experience isotopic decay. No secondary containment is l modeled. Time-dependent leakage to the environment coupled with appropriate atmospheric dispersion l coefficients produce activity concentrations in the Control Room inlet. Time-dependent Control Room l inleakage and damper flow allow activity to penetrate into the Control Room, which is subsequently l cleaned-up by time-dependent Control Room filtration. The Control Room activities result in thyroid, whole body, and beta skin doses based on the latest International Commission for Radiation Protection-30 dose conversion factors and time-dependent breathing rates and occupancy factors.

ARCON96 CODE METHODOLOGY The ARCON96 computer code implements a computational model for calculating atmospheric dispersion coefficients (X /Q's) in the vicinity of buildings (Reference 3). An atmospheric dispersion coefficient is simply the ratio of the relative concentration at the receptor (gm/m') to the release rate at the release point (gm/sec). Thus atmospheric dispersion coefficients are in units of sec/m'. The model estimates impacts from ground-level, vent, and elevated releases using a single-year or multi-years of 8

. 1 ENCLOSURE (A)

MAXIMUM HYPOTIIETICAL ACCIDENT /

LOSS-OF-COOLANT ACCIDENT ANALYSIS hourly meteorological data. This model also treats diffusion more realistically under low wind speed conditions than previous Nuclear Regulatory Commission-issud models.

This work calculates the atmospheric dispersion coefficients from the Containment to the Control Room air inlet assuming no thermal plume or momentum plume rise.

EXCEL SPREADSHEET METIIODOLOGY Some inputs for the AX2 and ARCON96 computer programs were generated via an EXCEL spreadsheet.

The iodine partition coefficient (IPC) employed in the spray calculation is computed using the methodologies of References 11 and 12. From Reference 1I for a molecular iodine density less than 1.E-5 gm-12/ liter and a buffered pli of 7, the following IPC can be calculated:

IPC = exp (6.46+3.25E-6*t)

IPC = 639.0611 @ t=0 see The molecular iodine density can be calculated as follows. The initial isotopic activity in Curies released to the Containment for isotope 'i' is based on the following algorithm assuming TID-14844 ,

(Reference 7) iodine generation and 100% release to the sump:

A cio = ATID,

• P in Ci where l ATID, = Isotopic activity per unit power (Ci/MWt)

P = Core power (MWt)

The isotopic molal density ofiodine released to the sump can thus be calculated to be New = Acio

• Ci / Ai / No / Vs in moles / liter where C i= 3.7E+10 disintegrations /sec/Ci Ai = lsotopic decay constant in 1/sec N = Avogadro's number = 6.022E+23 atoms / mole Vs = Sump volume in liters The diatomic molecular iodine density in the sump can thus be calculated as peio = Ncio

• Ai

• F in gm-12/ liter where A i= lsotopic molecular weight in gm/ mole F = Molecular iodine equilibrium factor = {l2 1/(Il2 }+[f])

[x] = Isotopic concentration in gm-atom / liter 9

l 4

ENCLOSURE (A)

MAXIMUM IIYPOTHETICAL ACCIDENT /

LOSS-0F-COOLANT ACCIDENT ANALYSIS Per Reference 12 in the presence of radiation, the equilibrium formation of 12 from F is strongly dependent on pli and weakly dependent on temperature and concentration. Ignoring the latter, the  ;

molecular iodine equilibrium factor can be calculated assuming a pil of 7 and the following algorithm l In(1/F-1) = 1.72*pII - 6.08 F = 2.57E-3 Thus the total diatomic molecular iodine concentration in the sump can be calculated as pc = E pcio = 9.709E-7 gm-12/ liter which is less than 1.E-5 gm-12/ liter assumed above, l

l i DOSES DUE TO DIRECT SHINE METIiOD OF ANALYSIS A design-basis MIIA occurs at time t = 0, instantaneously releasing 50% of the halogens and 100% of the noble gases uniformly into the containment atmosphere and releasing 50% of the halogens,1% of the l solids, and 100% of H3, Cn, Ni6 into the containment sump. During an accidental release, airborne and sump radionuclides may affect the Control Room through shine through the containment walls.

These calculations were accomplished via three distinct steps:

( e EXCEL spreadsheets were developed to determine airborne and sump isotopic activities, t

i e The MICROSlilELD code was executed to determine time-dependent dose rates in the Control .

Room from Containment shine.

• The MICROSlilELD dose rates were integrated via an EXCEL spreadsheet to yield 30-day Control Room doses.

t ASSUMPTIONS l The following assumptions were utilized in this work:

• A design-basis MilA occurs at time t = 0 instantaneously releasing 5,0% of the halogens and 100% of the noble gases unifonnly into the containment atmosphere and 50% of the halogens,1% of the solids, and 100% of11,3 Cu, and Ni6 uniformly to the containment sump.

• No credit is assumed for containment spray, plateout, or filtration.

l e No containment leakage is conservatively assumed, e The containment sump surface elevation is ~17', while ground level and the Control Room are at 45' elevation. No credit is taken for the elevation difference or for additional shielding afforded by the ground or the Auxiliary Building.

The input data to determine the Control Room doses from the containment shine pathway after a MilA

[

are the following:

• Initial thennal power is 2754 MWt (UFSAR Section 3.2.1 and Reference 4).

l 10 l

f ENCLOSURE (A)

MAXIMUM HYPOTHETICAL ACCIDENT /

LOSS-0F-COOLANT ACCIDENT ANALYSIS e Containment dimensions

> Net free volume: 1.989E+06 ft'(UFSAR Table 14.20-3) l > Inside diameter: 130 ft (UFSAR Section 5.1.2.1) .

> Vertical wall thickness: 3.75 ft (UFSAR Section 5.1.2.1)

> Sump volume: 68329 ft' e Control room wall thickness 2' e The isotopic decay constants and half-lives were extracted from References 13 and 15.

• The isotopic source terms (Ci/MWt) were extracted from Reference 13, are consistent with l TID-14844 methodology (Reference 7).

. Fission product inventories released to containment:

Airborne l Noble gases 100 % (References 8 and 16)

IIalogens 50% (References 8 and 16)

Sump l Noble gases 0% (Reference 16)

IIalogens 50 % (Reference 16)

Solids 1% (Reference 16)

H3, Cn, Ni6 100 % (Conservative) l l

l COMPUTER CODES EXCEL SPREADSHEET METHODOLOGY TO CALCULATE ACTIVITIES l The initial isotopic activity in Curies released to the containment atmosphere and sump for isotope 'i' is l based on the following algorithm based on TID-14844 (Reference 7):

Acio = ATID,

• P

• RFi

• exp(-loi

• to

• 3600) where ATID i= Isotopic activity per unit power (Ci/MWt) l P = Core power (MWt) l RFi = lsotopic release fraction  :

l Aoi = Isotopic decay constant (1/sec) to = Initial decay time (hr) l The corresponding isotopic activity density in Ci/m' released to the containment for isotope 'i' is l l pcio = Acio / Ve

! where f

V, = Containment airborne or sump volume (m')

11

ENCLOSURE (A)

MAXIMUM IIYPOTHETICAL ACCIDENT /

LOSS-0F-COOLANT ACCIDENT ANALYSIS MICROSHIb D CODE METHODOLOGY MICROSHIELD is a menu-driven point-kernel shielding code, which calculates gamma dose rates for various activity sources and shielding geometries. The activities calculated above, together with the i

relevant shielding geometries and materials determine the Control Room dose rates.

l

! Control room calculations:

Airborne Containment: Air cylinder of 65' radius and 149.8505' height Sump Containment Water cylinder of 65' radius and 5.15' height Wall Concrete cylinder of 3.75' width Gap Air gap of l' between Containment and Control Room Cor. trol Room wall Concrete slab of 2' thickness EXCEL SPREADSHEET METHODOLOGY TO INTEGRATE DOSE RATES Dose rate integration is accomplished via the trapezoidal rule (Reference 17).

" 2, F= f(x)

• dx = 0.5 * (x2 - x1) * [f(x2) + f(x1)]

s'n 1

where the integration is performed over each time step and summed. j RESULTS l The Control Room doses from an MHA for the current plant condition assuming 5500 cfm inleakage:

Containment RWT Shine Tntal Thyroid Dose 70.1 Rem 2.35 Rem 72.5 Rem (U Whole Body Dose 2.27 Rem 0.070 Rem 0.141 Rem 2.48 Rem Beta Skin Dose 10.5 Rem Negligible 10.5 Rem (U

The thyroid dose is 30 Rem in 45 minutes.

CONCLUSIONS The Control Room operators will accumulate a thyroid dose of 30 Rem after 45 minutes. Therefore, ,

J adequate time is available to implement the compensatory measures of Reference 18 such that compliance with GDC 19 using compensatory measures for thyroid dose is demonstrated.

The whole body and beta skin doses meet the GDC 19 limits of 5 and 30 Rem, respectively.

REFERENCES

1. Standard Review Plan 15.6.5, Appendix A, Revision 1, " Radiological Consequences of a Design Basis Loss-of-Coolant Accident including Containment Leakage Contribution,"

July 1981 i

! 12

ENCLOSURE (A)

MAXIMUM HYPOTHETICAL ACCIDENT /

LOSS-0F-COOLANT ACCIDENT ANALYSIS

2. Regulatory Guide 1.52, Revision 2, " Design, Testing, and Maintenance Criteria for Post Accident ESF Atmospheric Cleanup System Air Filtration and Absorption Units of LWRs,"

March 1978

3. NUREG/CR-6331, Revision 1, " Atmospheric Relative Concentrations in Building Wakes,"

May 1997

4. Regulatory Guide i.49, Revision 1, " Power Levels of Nuclear Power Plants," December 1973

5. Letter from Mr. J. A. Tiernan (BGE) to Mr. A. C. Thadani (NRC), dated March 5,1986,

" Control Room Dose"

6. Letter from Mr. A. E. Lundvall (BGE) to Mr. E. J. Butcher (NRC), dated August 9,1985,

" Containment Vent System"

7. TID-14844, " Calculation of Distance Factors for Power and Test Reactor Sites,"

March 23,1962

8. Regulatory Guide 1.4, Revision 2, " Assumptions for Evaluating the Potential Radiological Consequences of a LOCA for PWRs," June 1974

9. Standard Review Plan 6.4, Revision 2, " Control Room Habitability System," July 1981

10. Standard Review Plan 6.5.2, Revision 2, " Containment Spray as a Fission Product Cleanup System," December 1988

11. NUREG/CR-4697," Chemistry and Transport of lodine in Containment," October 1986

12. NUREG/CR-5732," lodine Chemical Forms in LWR Severe Accidents," April 1992

13. LOCADOSE NE319. Revision 3

14. Haliburton NUS Report NUS-1954, Revision 3, "AXIDENT: A Digital Computer Dose Calculation Model," February 1984

15. 1989 General Electric Company, Fourteenth Edition, "Nuclides and Isotopes - Chart of the Nuclides"

16. Regulatory Guide 1.89, Revis. ion 1, " Environmental Qualification of Certain Electric Equipment important to Safety for Nuclear Power Plants," June 1984

17. M. Abramowitz and I. A. Stegun, Eds., Dover Publications, Inc., NY, " Handbook of l Mathematical Functions"

18. Letter from Mr. C. H. Cruse (BGE) to NRC Document Control Desk, dated May 6,1993, Control Room Habitability - Interim Engineering Analysis for Thyroid Dose 13

4 ENCLOSURE (B) TO ATTACHMENT (1) l l

I

! w l

FUEL HANDLING ACCIDENT ANALYSIS I

(FOR CURRENT PLANT CONDITION) l

)

I l

i l l

! l l

l i

f-i Baltim9re Gas and Electric Company Calvert Cliffs Nuclear Power Plant April 21,1998

. 1 l .

ENCLOSURE (B)

FUEL HANDLING ACCIDENT l

INTRODUCTION i

l The Fuel llandling Accident (FHA) was initially analyzed to support fuel movement in Containment and in the Spent Fuel Pool (SFP). Analyses have been done assuming the Containment was closed and also assuming that both doors of the personnel air lock are open (Technical Specification 3.9.4). Similarly, an l analysis of the FIIA to support fuel movement in the SFP with the SFP filtration system in operation i (Technical Specification 3.9.12) was performed. These analyses are described in our Updated Final l Safety Analysis Report (UFSAR). Based on the results of these previous analyses, the most limiting l FHA would be the Containment analysis, assuming the personnel air lock doors are open.

l METHODS OF ANALYSIS Section 14.18 of the UFSAR presents the licensing basis evaluation of the FHA, which is assumed to occur when a fuel assembly is dropped during fuel movement operations in the SFP or in Containment.

All 176 rods from the highest power fuel assembly are damaged in the FHA.

The FHA in Containment model is constructed assuming that an FHA occurs at time t = 0 and assuming that the isotopes calculated in the EXCEL spreadsheet are released at time t = 0 to Containment. No cleanup mechanisms (spray, filtration, plateout) are assumed in Containment. The activity leaks from Containment assuming a single Containment turnaround in two hours. This activity is transported to the l site boundary and to the Control Room via appropriate atmospheric dispersion coefficients. Time-dependent Control Room inleakage is modeled. The Control Room and site boundary thyroid, whole body, and beta skin doses are calculated based on appropriate breathing rates and occupancy factors and on International Commission for Radiation Protection-30 dose conversion factors. I The FHA in the SFP model is constructed assuming that an FHA occurs at time t = 0 and assuming that l

% isotopes calculated in the EXCEL spreadsheet are released immediately and uniformly into the SFP. I no spray or plateout cleanup mechanisms are assumed in the SFP. The SFP ventilation system processes  !

i 35,200 cfm of the SFP volume into the environment with no credit for the high efficiency particulate air or charcoal filters for the duration of the accident. This activity is transported to the site boundary and to l

the Control Room via appropriate atmospheric dispersion coefficients. Time-dependent Control Room i inleakage is modeled. The Control Room and site boundary thyroid, whole body, and beta skin doses are calculated based on appropriate breathing rates and occupancy factors and on International Commission for Radiation Protection-30 dose conversion factors.

This was accomplished via three distinct methodologies:

• An EXCEL spreadsheet was developed to calculate the activity released to the containment or SFP atmosphere.

• Atmospheric dispersion coefficients from the release point to the Control Room were calculated via the ARCON96 computer code.

! e The AX2 (similar to AXIDENT and TACT-5) code was executed to determine the two hour site l boundary doses and the 30 day Control Room and low population zone doses for a FIIA in l

Containment and in the SFP.

This current work re-analyzes Control Room habitability for Containment and SFP FilAs based on high burnup fuel isotopics, International Commission for Radiation Protection-30 dose conversion factors, ARCON96 generated atmospheric dispersion coefficients to the Control Room dampers, and a 5500 cfm I

l

ENCLOSURE (B)

FUEL IIANDLING ACCIDENT Control Room inleakage. Note that 5500 cfm far exceeds the inleakage values measured for the current plant condition and is used for conservatism.

ASSUMPTIONS The following assumptions were utilized in this work:

• All 176 rods from only one fuel assembly will be damaged in the FHA.

e No credit is taken for atmospheric cleanup systems in Containment (spray, filter, plateout).

l l e No credit is taken for deposition of the plume on the ground or decay ofisotopes in transit to the site i boundary.

• Buildup of daughter nuclides is not taken into account as source tenn nuclides decay per Reference 1.

• This methodology assumes uniform Containment leakage over the external Containment surface and I neglects leakage internal to the Auxiliary Building, which would be subsequently filtered.

• This work also utilizes the more conservative Unit I atmospheric dispersion coefficients.

The input data to determine the Control Room dose from a Fila are the following:

AX2 INPUTS The inputs into the AX2 code are as follows:

. Initial thermal power is 2754 MWt (UFSAR Section 3.2.1 and Reference 2).

l e The power peaking factor is 1.65 (Reference 1).

l

• Fuel movement does not occur until 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> after reactor shutdown per Technical Specification 3.9.3.

• The Containment volume is assumed to be 1.989E+06 ft' per UFSAR Table 14.20-3. 72.73% of this volume comprises the sprayed region.

e The isotopic source terms (Ci/MWt) were extracted from Reference 3 and are consistent with I

TID-14844 methodology (Reference 4). The isotopic decay constants (1/sec) were also extracted from Reference 3.

SOURCE DECAY Isotope Ci/MWt 1/sec I-131 2.508E+04 9 976E-07 l l-132 3.806E+04 8.425E-05 l

I-133 5.622E+04 9.211E-06 l l-134 6.575E+04 2.200E-04 1-135 5.103 E+04 2.912E-05 XE-131M 2.595E+02 6.815E-07 XE-133M 1.384E+03 3.663 E-06 XE-133 5.622E+04 1.528E-06 XE-135M 1.557E+04 7.380E-04 XE-135 5.363E+04 2.11 SE-05 XE-137 5.103E+04 3.024E-03 e

2 1

( . . . _ _ _ _ _ _ _

ENCLOSURE (B)

FUEL IIANDLING ACCIDENT SOURCE DECAY Isotope Ci/MWt 1/sec XE-138 4.775E+04 8.151 E-04 KR-83M 4.152E+03 1.052E-04 KR-85M 1.297E+04 4.297E-05 KR-85 4.102E+02 s 2.054E-09 KR-87 2.335 E+04 1.514E-04 KR-88 3.200E+04 6.731 E-05 KR-89 3.979E+ 04 3.632E-03 e Per References 5 and 6, damaged fuel rods are assumed to release their gas gap activities consisting of the following isotopes:

12 % I-131 10% other iodines 30% Kr-85 10% other noble gases e Per Reference 1, the iodine gap activity is composed of 99.75% inorganic species ar.d 0.2 i% organic species of iodine. The pool decontamination factors are 133 for the inorganic iodine ard I for the ,

organic iodine, yielding an overall effective decontamination factor of 100. This diFerence in decontamination factor for inorganic and organic iodine species results in the iodine above the fuel pool being composed of 75% inorganic and 25% organic species.

. The decontamination factor of noble gases in the pool is unity per Reference 1.

• The Control Room volume of 166000 ft' is extracted from Reference 7.

• The breathing rates are extracted from Reference 8: l 3

0- 8 hrs 3.47E-04 m /sec 8- 24 hrs 1.75E-04 m'/sec 3

24-720 hrs 2.32E-04 m /sec l

e The Control Room occupancy factors are extracted from References 7 and 9:

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

• The site boundary X /Q of 1.30E-4 sec/m' was extracted from UFSAR Section 2.3.6.

• Atmospheric dispersion coefficients from Containment to the low population zone:

0- 2 hrs 3.30E-5 sec/m' 2-24 hrs 2.20E-6 sec/m'24-720 hrs 5.40E-7 sec/m' 3

T U ENCLOSURE (B)

FUEL HANDLING ACCIDENT l e

International Commission for Radiation Protection-30 dose conversion factors are as follows:

INilALATION SUBMERSION SUBMERSION Rem /Ci Rem-m'/Ci-sec Rem-m'/Ci-sec j isotope iIlYROID WilOLE BODY Tl!YROID WilOLE BODY BETA SKIN I-131 1.lE+06 3.3 E+04 1-132 6.3 E+03 3.4E+02 1-133 1.8E+05 5.6E+03 1-134 1.lE+03 1.lE+02 1-135 3.l E+04 1.lE+03 XE-131M 1.3E-03 1.8E-02 XE-133N1 5.4 E-03 3.8E-02 XE-133 7.3E-03 6.3 E-03 2.0E-02 XE-135M 7.7E-02 1.lE-01 XE-135 4.7E-02 1.2E-01 XE-138 2.0E-01 2.0E-01 4.lE-01 l KR-83M 3.7E-06 1.8E-04 l KR-85M 3.lE-02 3.0E-02 8.5E-02 KR-85 4.7E-04 4.8E-02 KR-87 1.4E-01 1.5E-01 5.2E-01 KR-88 3.8E-01 3.7E-01 5.4E-01 e Atmospheric dispersion coefficients from Containment to Control Room: (Reference 10)

Unit 1 Unit 2 3

0- 2 hrs 3.72E-04 sec/m 3.06E-04 sec/m' 2- 8 hrs 2.33E-04 sec/m' 2.03E-04 sec/m' 8- 24 hrs 9.03E-05 sec/m' 8.69E-05 sec/m' 3 3 24- 96 hrs 7.75E-05 sec/m 6.35E-05 sec/m 3

96-720 hrs 6.36E-05 sec/m 5.24E-05 sec/m'

• Atmospheric dispersion coefficients from Vent Stack to Control Room: (Reference 10)

Unit i Unit 2 3

0- 2 hrs 1.83E-04 sec/m' l.21E-04 sec/m 3

2- 8 hrs 1.10E-04 sec/m' 6.81E-05 sec/m 8- 24 hrs 5.11E-05 sec/m' 2.50E-05 sec/m' 3 3 l 24- 96 hrs 3.14E-05 sec/m 1.77E-05 sec/m 3 3

!96-720 hrs 2.33E-05 sec/m 1.19E-05 sec/m e The isotopic gamma energies and fractions are detailed in Reference 11.

. The energy-dependent total and energy absorption coefficients are detailed in Reference 11.

• Control Room inleakage is assumed to be 5500 cfm. This bounds the current plant condition.

e The leak rates are calculated as follows:

> Containment:

0-2 hrs: A = 1/7200 sec = 1.3889E-4/sec (One Containment turnover in two hours) 2-720 hrs: A = 0%/sec l

4

r ENCLOSURE (B) j FUEL HANDLING ACCIDENT l

1

> SFP:-

0-720 hrs A = Fsrp/Vsrp/60 = 1.1251E-3/sec l where i l Fsrp = 1.l *32000 cfm = Maximum SFP No. I1 fan Dow rate (Technical Specincation 3.9.12) )

Vsrp = 5.2144E+05 ft' l )

e Activity release rate from the refueling pool to the Containment or the SFP to the Auxiliary Building:

> Containment: 100% of activity conservatively released at time t = 0

> SFP: 100% of activity released at time t = 0 e Control Room recirculation and filtration Dow:

> Flow rate

One 61ter train at 1800 cfm (Technical Specification 3.7.6)

> Initiation delay time:

30 see (Activates automatically on a CRS, UFSAR Table 8-7) l 3 sec (containment radiation signal) = 3 sec (No loss of offsite power is assumed.

j Conservatively 30 see will be assumed.)

> Filter efficiencies: 90% for all iodine species (Technical Specifications 3.7.6)

> lodine filter removal coefficients:

LF = (1800 cfm)*(60 min /hr)*0.90/(l66000 ft') = 0.5855/hr = 1.6265E-04/sec

. The spent fuel pool filters are not credited, although the SFP ventilation system is.

I ARCON961NPUTS The X /Q was determined for the Vent Stack to Control Room Dow path using the following inputs:

. Height oflower wind instrument: 10 meters e Height of upper wind instrument: 60 meters

. Wind sp2ed units type: meters /sec l

• Release type: vent l e Release height: 48.31 meters l > Ventilation stack height is extracted from Reference 12.

l > H = 203.5' - 45.0' = 158.5' = 48.31 meters 2

e Building area: 1155 m 2

! The calculation of Containment cross-sectional area yields 12435.63 ft above the rooftop level of 2

91'6". The Auxiliary. Building cross-sectional area can be calculated to be 1938.93 ft . For a west-to-east wind direction, the total cross-sectional area of the Auxiliary Building and the two l 2

Containments is 26810 ft For an east-to-west wind direction, the total cross-sectional area of 2

the Turbine Building is 27167 ft . For a north-to-south and south-to-north wind direction, the 2

total cross-sectional area of the Containment and the Turbine Building is 21016 ft . The cross-2 2 sectional area of a single containment of 12435.63 ft or 1155 m will conservatively be used.

• Effluent vertical velocity: 21.26 (meters /sec)

The vent stack exit Dow rate assumes that No. I1 or No.12 Main Exhaust Fan is in operation prior to SFP area ventilation startup and fuel movement. As a conservatism, this How rate is reduced by 10%. The exit velocity is thus the flow rate divided by the vent cross-sectional area.

5 l

)

J

ENCLOSURE (B)

FUEL HANDLING ACCIDENT 3

e Flow rate = 0.9

• 131500 cfm = 118350 cfm = 55.8550 m /sec 2

Vs(VS) = Flow rate /(n*d /4)/60 = 69.76 ft/sec = 21.26 m/sec e Stack or vent Dow: 55.86 (m'/s) e Stack or vent radius: 0.91 meters The exit diameters is extracted from Reference 13.

d = 6' = 1.8288 meters

• Direction to source: 290 degrees (Unit 1); I 50 degrees (Unit 2) e Source window: 90 degrees e Distance from source to receptor: 24.5 meters (Unit 1); 38.0 meters (Unit 2) )

e Intake height: 15.93 meters 91.5' + 5.75' - 45' = 52.25' = 15.93 m where 91.5' is the height of the Auxiliary Building roof,5.75' is the Control Room intake height, and 45' is ground level e Grade elevation difference: 0 meters (Reference 10)

• Surface roughness length: 0.1 meters (Reference 10) e Minimum wind speed: 0.5 meters /sec (Reference 10)

• Sector averaging constant: 4 (Reference 10) e liours in average: 12 4 812 24 96168 360 720 (Reference 10)

Minimum number of hours: 12 4 81122 87152 324 648 (Reference 10) e llorizontal diffusion coefficient: 0.0 meters a y= r/2.15 = 0.9144/2.15 = 0.43 meters (Reference 10) Conservatively use 0.

• Vertical diffusion coefficient: 0.0 meters The X /Q was determined for the Containment to Control Room flow path using the following inputs:

. Height oflower wind instrument: 10 meters e lleight of upper wind instrument: 60 meters e Wind speed units type: meters /sec e Release type: vent i

e Release height: 14.17 meters l The release height is assumed to be at the 91.5' roof elevation.

l Z = Average containment release height = 91.5'-45' = 46.5' = 14.17 meters 2

e Building area: 1155 (m )

e Effluent vertical velocity: 0.0 meters /sec

. Stack or vent flow: 0.0 (m'/s) e Stack or vent radius: 20.96 meters (UFSAR Section 5.1.2.1) r = 137.5'/2 = 68.75' = 20.96 meters 6

[

l .

l ENCLOSURE (B)

FUEL IIANDLING ACCIDENT

• Direction to source: 265 degrees (Unit 1); 170 degrees (Unit 2)

• Source window: 90 degrees

• Distance from source to receptor: 44.62 meters (Unit 1); 54.84 meters (Unit 2) e Intake height: 15.62 meters l 91.5' + 4.75' - 45' = 51.25' = 15.62 m l where 91.5' is the height of the Auxiliary Building roof,4.75' is the Control Room exhaust height, and 45' is ground level e Grade elevation difference: 0 meters (Reference 10)

!

• Surface roughness length: 0.1 meters (Reference 10) e Minimum wind speed: 0.5 meters /sec (Reference 10)

• Sector averaging constant: 4 (Reference 10)

• Hours in average: 12 4 812 24 96168 360 720 (Reference 10)

• Minimum number of hours: 12 4 81122 87152 324 648 (Reference 10) e Horizontal diffusion coefficient: 9.75 meters The horizontal diffusion coefficient is defined as the containment radius divided by 2.15. This is j consistent with the methodology of Reference 10.

a y= r/2.15 = 20.96/2.15 = 9.75 meters (Reference 10)

• Vertical diffusion coefTicient: 22.62 meters The vertical diffusion coefficient is defined as 1/2 of the height of the leakage area. This is consistent with the methodology of Reference 10.

L2 = Upper containment elevation 193.4' Reference 14 oz = (l93.4'-45')/2 = 74.2' = 22.62 meters (Reference 10)

COMPUTER CODES ARCON96 CODE METHODOLOGY The ARCON96 computer code implements a computational model for calculating atmospheric dispersion coefficients ( X /Q's) in the vicinity of buildings (Reference 10). An atmospheric dispersion 3

coefficient is simply the ratio of the relative concentration at the receptor (gm/m ) to the release rate at 3

the release point (gm/sec). Thus atmospheric dispersion coefficients are in units of sec/m . The model l estimates impacts from ground-level, vent, and elevated releases using a single-year or multi-years of l hourly meteorological data. This model also treats diffusion more realistically under low wind speed

! conditions than previous Nuclear Regulatory Commission-issued models.

This calculation determines the atmospheric dispersion coefficients from the vent stack to the Control Room air inlet assuming no thermal plume rise but assuming a momentum plume rise consistent with No. I1 or No.12 ventilation stack fans being in service.

u AX2 CODE METHODOLOGY The AX2 computer code calculates individual gamma and beta whole body and thyroid doses to personnel in the Control Room resulting from any postulated accident which releases radioactivity within l

7

O l l

ENCLOSURE (B) l FUEL HANDLING ACCIDENT the Containment or within any primary system. The AX2 models the transport of radioactivity (elemental, particulate, and organic iodine isotopes and krypton and xenon isotopes) from the sprayed and unsprayed regions of a primary Containment, through the secondary Containment if any, and then to the environment and to the Control Room. The code includes the capability to model time-dependent activity release; containment spray, filtration, and leakage; Control Room filtration and inleakage; primary and secondary containment purge filters; Control Room intake filters; atmospheric dispersion; and natural decay. Doses are calculated for individuals residing in the Control Room.

EXCEL SPREADSHEET METHODOLOGY This spreadsheet was used to calculate the activity released to the containment or SFP atmosphere. The initial isotopic activity in curies released to the Containment for isotope 'i' is based on the following algorithm based on TID-14844 (Reference 4):

Acio = ATIDi

• P

• PPF

• RFi / NASSM / DFi

• exp(-loi

• to

• 3600.)

where:

ATID i= lsotopic activity per unit power (Ci/MWt)

P = Core power (MWt)

PPF = Power peaking factor RFi = Isotopic gas gap release fraction DFi = Isotopic decontamination factor Aoi = Isotopic decay constant (1/sec) to = Time from power shutdown to FHA (hr)

NASSM = Number of assemblies in core = 217 3

The corresponding isotopic activity density in Ci/m released to the containment for isotope 'i' is peio = Acio / V c where 3

V e= Containment volume (m )

RESULTS The doses from a FHA in Containment with the personnel air lock doors open for the current plant condition assuming 5500 cfm inleakage are the following:

Control Room Thyroid Dose 31.0 Rem (U Whole Body Dose 0.937 Rem Beta Skin Dose 0.356 Rem (O

The thyroid dose is less than 30 Rem in two hours.

The Control Room doses from a FHA in the SFP with only the Control Room filtration system in operation, were also performed and are bounded by the above results.

8

ENCLOSURE (B) ]

FUEL IIANDLING ACCIDENT l CONCLUSIONS The Control Room thyroid dose for the case of the personnel air lock open is bounded by the thyroid dose calculated by the Nuclear Regulatory Commission Staff (33 Rem). The referenced safety evaluatien considered 33 Rem an acceptable response when meeting the requirements of 10 CFR Part 50, Appendix A, General Design Criteria 19. The whole body and beta skin doses are well within the General Design Criteria limits of 5 and 30 Rem, respectively. Therefore, based on the Nuclear Regulatory Commission Staff's prior acceptance, we consider the FIIA to continue to meet General Design Criteria 19 acceptance criteria. Note that if compensatory measures of Reference 15 are implemented within two hours the thyroid dose will remain below 30 Rem.

REFERENCES

1. Nuclear Regulatory Commission Safety Guide 25, " Assumptions Used for Evaluating the Potential Radiological Consequences of a FIIA in the Fuel llandling and Storage Facility for BWRs and PWRs," March 23,1972

2. Regulatory Guide 1,49 Revision 1, " Power Levels of Nuclear Power Plants," December 1973

3. LOCADOSE NE319, Revision 3

4. TID-14844, " Calculation of Distance Factors for Power and Test Reactor Sites,"

March 23,1962

5. Letter from Mr. D. G. Mcdonald (NRC) to Mr. G. C. Creel (BGE), dated July 6,1992,

" Approval for Calvert Cliffs Units 1 and 2 Fuel Pin Burnup Limit of 60 MWD /KG"

6. Letter from Mr. A. C. Thadani (NRC) to Mr. A. E. Scherer (CE), dated June 22,1992, " Generic Approval of CE Topical Report CEN-386-P, Verification of the Acceptability of a 1-Pin Burnup Limit of 60 MWD /kg for CE 16x16 PWR Fuel"

7. Letter from Mr. J. A. Tiernan (BGE) to Mr. A. C. Thadani (NRC), dated March 5,1986,

" Control Room Dose"

8. Regulatory Guide 1.4 Revision 2, " Assumptions for Evaluating the Potential Radiological Consequences of a LOCA for PWRs," June 1974

9. Standard Review Plan, Section 6.4, Revision 2, " Control Room liabitability System,"

July 1981

10. NUREG/CR-6331, Revision 1, " Atmospheric Relative Concentrations in Building Wakes,"

May 1997

11. IIaliburton NUS Report NUS-1954, Revision 3, "AXIDENT: A Digital Computer Dose Calculation Model," February 1984

12. BGE Drawing 62-006-E," General East and South Elevations," Revision 4

13. BGE Drawing 60-330-E, "licating and Ventilation System, Auxiliary Building, El. e, 0,"

Sections and Details," Revision 14

14. BGE Drawing 61-740-E, " Containment Liner Plan, Elevation & Penetrations," Revision 19

15. Letter from Mr. C. II. Cruse (BGE) to Document Control Desk (NRC), dated May 6,1993,  ;

" Control Room Habitability - Interim Engineering Analysis for Thyroid Dose" I

j 9

ENCLOSURE (C) TO ATTACHMENT (1) l MAIN STEAM LINE BREAK (FOR CURRENT PLANT CONDITION)

Baltimore Gas and Electric Company Calvert Cliffs Nuclear Power Plant April 21,1998

. l ENCLOSURE (C)

MAIN STEAM LINE BREAK ACCIDENT INTRODUCTION The Main Steam Line Break (MSLB) was analyzed in conjunction with the core analyses supporting the low flow amendment, which analyzed the effect ofincreasing the number of plugged tubes per steam generator from 800 to 2500. The Nuclear Regulatory Commission also requested additional information regarding the Control Room doses that would result from the MSLB.

METHODS OF ANALYSIS Chapter 14.14 of the Updated Final Safety Analysis Report (UFSAR) presents the licensing basis evaluation of the MSLB. An MSLB is defined as the pre-trip guillotine-type rupture of a main steam line outside Containment in the Main Steam Piping Room. This rupture increases the rate of heat l extraction by the steam generators and causes cooldown of the Reactor Coolant System (RCS). With a i

negative moderator temperature coefficient, the RCS cooldown will produce a positive reactivity addition that is terminated only when the control element assemblies insert after a reactor trip. The depressurization of the affected steam generator causes a low steam generator isolation signal which causes the main steam isolation valves to close. It is assumed that the steam line break occurs between the steam generator and the main steam isolation valve, allowing blowdown of the affected steam i generator to continue. The continued blowdown causes the RCS pressure to decrease until the Safety l Injection Actuation Signal is initiated, which automatically starts the high pressure safety injection i l pumps. Since the shutoff head to the high pressure safety injection pumps is equal to 1280 psia, no safety injection flow is delivered immediately. Therefore, the pressure continues to decrease and the l pressurizer empties. At 1280 psia, the high pressure safety injection pump flow will begin to add coolant l mass, such that the pressurizer level will be reestablished. A loss of offsite power, with the turbine trip, l results in the maximum site boundary doses. The loss of offsite power causes the reactor coolant pumps l to coast down, mirimizing core flow, lowering the transient departure from nucleate boiling ratio, and l maximizing the number of failed fuel pins.

For the affected steam generator the source term consists of the maximum secondary system Technical Specification activity and that fraction of the primary system Technical Specification and failed fuel i activity which leaks to the secondary side of the steam generator. This combined activity is then discharged into the main steam piping room and out the main steam piping room vent on the roof of the Auxiliary Building Since the steam generators are designed to withstand RCS operat*mg pressure on the tube side with atmospheric pressure on the shell side, the continued integrity of the RCS barrier is ,

assured. Thus, only the maximum Technical Specification primary-to-secondary leakage is assumed. A l partition factor of one is assumed for all discharged radioactivity.

1 For the intact steam generator, the maximum secondary Technical Specification activity is also assumed to be discharged out the main steam piping room vent. All of the primary-to-secondary Technical Specification and failed fuel activity that leaks to the secondary is assumed to be discharged from the atmospheric dump valves (ADVs) or the main steam safety valves. A partition factor of one is assumed for all discharged radioactivity. Note that with a loss of offsite power, the condensers are not available for cooldown.

l I

1 1

ENCLOEURE (C)

MAIN STEAM LINE BREAK ACCIDENT This analysis was accomplishcd usinF the following methodologies:

!

• Atmospheric dispersion coefficients from the release point to the inlet plenum were calculated using the ARCON96 computer code.

• The AX3 code was executed to determine the 30-day Control Room doses from an MSLB with the

, affected steam generator flow exiting the main steam piping room vent and the intact steam l generator flow exiting the ADV or the main steam safety valve. The AX3 code is equivalent to AXIDENT, TACT 5 and LOCADOSE.

l l

ASSUMPTIONS i

l The following assumptions were utilized in this analysis:

l

• No cred9 is taken for deposition of the plume on the ground or decay of isotopes in transit to the site boundary.

• Buildup of daughter nuclides is not taken into account as source term nuclides decay per Reference 1.

The input data to determine the Control Room from an MSLB are the following:

ARCON96 INPUTS l The X /Q was determined for the main steam piping room vent to Control Room flow path using the l following inputs:

l l

e Height oflower wind instrument: 10 meters e Height of upper wind instrument: 60 meters

. Wind speed units type: meters /sec

. Release type: vent l

• Release height: 17.15 meters Main steam piping room vent height is: H = 91.5" + 9.75' - 45.0' = 56.25' = 17.15 meters l No thermal or momentum plume rise is assumed.

t 2

l e Building area: 1155 m l

2 l The calculation of Containment cross-sectional area yields 12435.63 ft above the rooftop level 2

of elevation 91'6". The Auxiliary Building cross-sectional area is calculated to be 1938.93 ft ,

For a west-to-east wind direction, the total cross-sectional area of the Auxiliary Building and the two Containments is 268'.0 ft 2. For an east-to-west wind direction, the total cross-sectional area 2

of the Turbine BuildinF is 27167 ft . For a north-to-south and south-to-north wind direction, the 2

total cross-sectional area of the Containment and the Turbine Building is 21016 ft . The cross-2 2 sectional area of a single Containment of 12435.63 ft or 1155 m will be used conservatively.

• Effluent vertical velocity: 0 meters /sec e Stack or vent flow: 0 m'/s e Stack or vent radius: 1.38 meters The exit radius is:

2 Area = 8'*8' = 64 ft , thus r = sqrt(64/n) = 1.38 m 2

ENCLOSURE (C)

MAIN STEAM LINE BREAK ACCIDENT

+ Direction to source: Unit 1 - 278 degrees; Unit 2 - 157 degrees e Source window: 90 degrees

• Distance from source to receptor: 21 meters (Unit 1),32 meters (Unit 2) e Intake height: 15.62 meters 91.5' + 4.75' - 45' = 51.25' = 15.62 m where 91.5' is the height of the Auxiliary Building roof,4.75' is the Control Room exhaust height, and 45' is ground level.

• Grade elevation difference: 0 meters (Reference 2) i l

• Surface roughness length: 0.1 meters (Reference 2) e Minimum wind speed: 0.5 meters /sec (Reference 2) I i

e Sector averaging constant: 4 (Reference 2) e llours in average: 12 4 812 24 96168 360 720(Reference 2) e Minimum number of hours: 12 4 81122 87152 324 648(Reference 2) e llorizontal diffusion coefGcient: 0.64 meters ,

1 c y= r/2.15 = 1.38/2.15 = 0.64 meters (Reference 2) e Vertical diffusion coefficient: 2.97 meters o, = 9.75' = 2.97 meters (Reference 2)

Example 5 of Reference 2 presumes e vertical diffusion coefficient of 1 meter for a capped vent, I meter above the roof. This care is s:milar, in that we have a downward pointing gooseneck 9.75 feet above the surface of the Auxiliary Building roof. The physics of the downward pointing capped vent and gooseneck should be identical.

The X /Q was determined for the ADV to Control Room flow path using the following inputs:

• Height oflower wind instrument: 10 meters e Height of upper wind instrument: 60 meters e Wind speed units type: meters /sec e Release type: vent

• Release height: 53.56 meters

> The weighted average wind speed is calculated using the 1991-1993 Joint Frequency Table values for stability Class F and the wind speed power law distribution with height relationship for stability Class F. The Joint Frequency Table displays the number of hours (ni) that the wind is at velocity vi in a specific velocity interval at the 10 meters or 60 meters primary meteorological tower (PMT) position. The average wind velocity for year 'x' at PMT position

'y' is thus:

v(x,y) = E(n,(x,y)

• v,(x,y))/E(n,(x,y)).

3

ENCLOSURE (C)

MAIN STEAM LINE BREAK ACCIDENT

. The average velocity over three years at PMT position 'y' is thus:

v(y) = L(v(x,y))/3.

For the 10 meter and 60 meter PMT positions, the average velocities are 1.87 and 4.48 meters /sec. The wind velocity at any height 'h' can then be determined from the wind speed power law distribution with height relationship for stability class F v(h) =

s(10m)*(h/10) 5. Use of this relationship with the release height, AH, calculated below, yields the wind velocity 2.46 meters /sec.

> The ADV exit flow rate is:

3 Flow rate = 7279 cfm = 3.44 m /sec 2

Vs(VS) = Flow rate /(n*d /4)/60 = 221.54 ft/sec = 67.53 m/sec

> The jet entrainment coefficient is:

beta = 1/3 + u/Vs = 0.37

> The exit diameter is extracted from Reference 3.

d = 10.02" = 0.2545 meters

> The ambient temperature T, is assumed to be 293.15 K = 20 C = 68 F.

> The exit temperature Ts is assumed to be 373.15 K = 100 C = 212 F.

> The momentum flux is:

2 2 d 2 Fm = Vs *d *(T,/4fl's) = 58.0090 m /s

> The buoyancy flux is:

2 d F3= g*Vs*D *(AT/4fI's) = 2.2975 m /s'

> The thermal plume rises is:

3 2 AHt = 1.6

• F b

• x ' / u = 13.4021 meters

> The momentum plume rise is:

5 2 AHm = {3

• Fm * [ sin (x*s /u)/ beta /u/s 5]}'" = 22.9420 meters where s = 0.02

• g / T, = 0.000669

> ADV height is:

H = 91.5' + 10' - 45.0' = 56.5' = 17.22 meters

> The combined release height is:

AHtot = AHt + AHm + H = 53.5653 meters 2

e Building area: 1155 m

. Effluent vertical velocity: 67.53 meters /sec 221.5 ft/sec = 67.53 meters /sec

• Stack or vent flow: 3.44 m'/s 121.32 cf/sec = 3.44 m'/sec; 4

ENCLOSURE (C)

! MAIN STEAM LINE BREAK ACCIDENT e Stack or vent radius: 0.13 meters d = 10.02" = 0.25 meters j e Direction to source: Unit 1 - 285 degrees; Unit 2 - 152 degrees e Source window: 90 degrees e Distance from source to receptor: 18 meters (Unit 1),31 meters (Unit 2) j

• Intake height: 15.62 meters e

Grade elevation difference: 0 meters (Reference 2) e Surface roughness length: 0.1 meters (Reference 2)

• Minimum wind speed: 0.5 meters /sec (Reference 2) e Sector averaging constant: 4 (Reference 2)

• 11ours in average: 12 4 812 24 96168 360 720 (Reference 2) e Minimum number of hours: 12 4 81122 87152 324 648 (Reference 2) l e llorizontal diffusion coefficient: 0.0 meters 1

• Vertical diffusion coefficient: 0.0 meters I

The steam flow rates through the ADVs as a function of steam generator temperature are calculated using the following inputs:

)

)

. Atmospheric pressure: 14.700 psia e Resistance coefficient: 9.820 K e Maximum DP/Pl: 0.771

. Expansion Factor: 0.704

• Limiting Diameter: 3.760 inches

e Exit Diameter

10.020 inches l

l AX3 INPUTS The inputs into the AX3 code are as follows:

• Initial thermal power is 2754 MWt. (UFSAR Section 3.2.1 and Reference 4) e The power peaking factor is 1.70. (Reference 5) l e The failed fuel fraction is 1.35%.

• The specific activity of the secondary coolant system shall be less than 1.0E-7 Ci/ gram Dose i Equivalent I-131 per Technical Specification 3.7.1. ,

• The specific iodine activity of the primary coolant shall be less than 1.E-6 Ci/ gram Dose Equivalent 1-131 per Technical Specification 3.4.8.

5 l 1

l i  !

ENCLOSURE (C)

MAIN STEAM LINE BREAK ACCIDENT e The specific activity of the primary coolant shall be less than 100.E-6/<E> Ci/ gram noble gas per Technical Specification 3.4.8.

. The average gamma and beta energies are 0.624 and 0.470 Mev/ disintegration.

• The Control Room volume of 166000 ft' is extracted from Reference 6.

• The minimum RCS fluid mass is 457437 lbm for the 0-30 day low population zone and Control Room calculations and 384260 lbm for the 0-2 hour site boundary calculations.

. The maximum secondary fluid mass is 250,500 lbm.

. The initial and final primary densities are conservatively assumed to be 45.33 lbm/cf.

. The time to shutdown cooling is conservatively assumed to be 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> or 32,400 sec.

• Atmospheric dispersion coefficients from the main steam piping room vent to the Control Room:

(Reference 2)

Unit 1 Unit 2 0- 2 hrs 5.12E-03 sec/m' 2.86E-03 sec/m' 3

2- 8 hrs 3.57E-03 sec/m 2.12E-03 sec/m' 3

i 8- 24 hrs 1.53E-03 sec/m' 7.61E-04 sec/m l 24- 96 hrs 1.06E-03 sec/m' 5.44E-04 sec/m'96-720 hrs 8.40E-04 sec/m' 4.67E-04 sec/m' l

• Atmospheric dispersion coefficients from the ADVs and/or the main steam safety valves to Control Room: (Reference 2)

Unit 1 Unit 2 l 3 0- 2 hrs 1.00E-10 sec/m' l.00E-19 sec/m 3

2- 8 hrs 1.00E-10 sec/m 1.00E-19 sec/m' 3 3 8- 24 hrs 1.00E-10 sec/m 4.78E-18 sec/m 3

24- 96 hrs 1.00E-10 sec/m 1.73E-18 sec/m'96-720 hrs 5.02E-10 sec/m' l.62E-18 sec/m'

• The breathing rates are extracted from Reference 7:

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 extracted from References 6 and 8: I 0-24 hrs 1.0 24-96 hrs 0.6 96-720 hrs 0.4

• The primary to secondary leak rate is 100 gal / day per steam generator or 6.9444E-2 gpm per Technical Specification 3.4.6.

• Control Room inteakage: Control Room inleakage is assumed to be 5500 cfm. This bounds the current plant condition.

l 6 l l

e ENCLOSURE (C)

MAIN STEAM LINE BREAK ACCIDENT

• Control Room recirculation and filtration flow:

> Flow rate: One filter train at 1800 cfm (Technical Specification 3.7.6)

> Initiation delay time: 30 sec [(Activates automatically on a safety injection actuation signal (UFSAR Table 8-7)] 3 sec (safety injection actuation signal) +10 sec (emergency diesel generator startup) = 13 sec (Conservatively 30 see is assumed.)

> Filter efficiencies: 90% for all iodine species (Technical Specification 3.7.6)

> lodine filter removal coefficients:

LF = (1800.cfm)*(60. min /hr)*0.90/(l66000.cf) = 0.5855/hr = 1.6265E-04/sec

• Partition Factors:

> Primary / secondary release through the ADVs: 1.0 l

> Primary / secondary release through the main steam piping room vents: 1.0 e The isotopic source terms (Ci/MWt) were extracted from Reference 9 and are consistent with TID-14844 methodology (Reference 10). The isotopic decay constants (1/sec) were also extracted from Reference 9.

SOURCE DECAY Isotope Ci/MWt 1/sec l-131 2.508E+04 9.976E-07 I-132 3.806E+04 8.425E-05 1-133 5.622E+04 9.211 E-06 I-134 6.575E+04 2.200E-04 j I-135 5.103 E+04 2.912E-05 i XE-131M 2.595E+02 6.81SE-07 XE-133M 1.384E+03 3.663E-06 XE-133 5.622E+04 1.528E-06 XE-135M 1.557E+04 7.380E-04 l XE-135 5.363E+04 2.115E-05

! XE-137 5.103 E+04 3.024E-03 XE-138 4.775E+04 8.151E-04 KR-83M 4.152E+03 1.052E-04 KR-85M l.297E+04 4.297E-05 KR-85 4.102E+02 2.054E-09 KR-87 2.335E+04 1.514E-04 KR-88 3.200E+04 6.731 E-05 ,

KR-89 3.979E+04 3.632E-03 1

• Per References 1,11, and 12, damaged fuel rods are assumed to release their gas gap activities !

consisting of the following isotopes:

12 % l-131 10% other iodines 30% Kr-85 i

10% other noble gases 1

7 l

e o ..

ENCLOSURE (C)

MAIN STEAM LINE BREAK ACCIDENT e

International Commission for Radiation Protection (ICRP)-30 dose conversion factors are listed below.

INIIALATION IMMERSION IMMERSION Rem'Ci Rem-m'/Ci-sec Rem-m'/Ci-sec l isotope TilYROID WilOLE BODY TilYROID WilOLE BODY BETA SKIN 1-131 1.lE+06 3.3E+04 I-132 6.3 E+03 3.4E+02 1-133 1.8E+05 5.6E+03 1-134 1.lE+03 1.lE+02 1-135 3.l E+04 1.lE+03

]

{

XE-131M 1.3 E-03 1.8E-02 XE-133M 5.4E-03 3.8E-02 XE-133 7.3 E-03 6.3 E-03 2.0E-02 3 XE-135M 7.7E-02 1.lE-01 )

XE-135 4.7E-02 1.2E-01 XE-138 2.0E-01 2.0E-01 4.lE-01 KR-83M 3.7E-06 1.8E-04 4 KR-85M 3.1E-02 3.0E-02 8.5E-02 KR-85 4.7E-04 4.8E-02 KR-87 1.4E-01 1.5E-01 5.2E-01

)

KR-88 3.8E-01 3.7E-01 5.4E-01

. The isotopic gamma energies and fractions are detailed in Reference 13.

e The energy-dependent total and energy absorption coefficients are detailed in Reference 13.

i

[ COMPUTER CODES l ARCON96 CODE METHODOLOGY l

l The ARCON96 computer code implements a computational model for calculating atmospheric l dispersion coefficients (X /Qs) in the vicinity of buildings (Reference 2). An atmospheric dispersion 3

! coefficient is simply the ratio of the relative concentration at the receptor (gm/m ) to the release rate at the release point (gm/sec). Thus, atmospheric dispersion coefficients are in units of sec/m'.

The model estimates impacts from ground-level, vent, and elevated releases using a single year or multi-l years of hourly meteorological data. This model also treats diffusion more realistically under low wind i speed conditions than previous Nuclear Regulatory Commission-issued models.

l This work calculates the atmospheric dispersion coefficients from the main steam piping room vent to the Control Room assuming no thermal or momentum plume rise and from the ADVs and/or main steam safety valves to the Control Room assuming momentum and thermal plume rises.

AX3 CODE METHODOLOGY This calculation employed the AX3 computer code. The AX3 models the transport of halogen and noHe gas isotopes from a primary Containment to a secondary Containment and then to the environment and Control Room.

8

0

=

r ENCLOSURE (C)

MAIN STEAM LINE BREAK ACCIDENT The AX3 computer code calculates individual gamma and beta whole body and thyroid doses resulting from any postulated accident that releases radioactivity via the following sources: (1) failed fuel gas gap activity; (2) pnmary dose equivalent I-131 and Eisenbud noble gas activity with Kr-85 half-life; and (3) secondary system dose equivalent 1-131 activity. The AX3 models the transport of radioactivity (iodine, krypton, and xenon isotopes) from the primary system to the secondary system and then to the environment and to the Control Room. The code assumes instantaneous failed fuel activity is released into the primary system, a time-dependent primary-to-secondary leak rate, instantaneous secondary system release of activity to the environment with a time-dependent partition factor for the primary system and failed fuel iodine activities, time-dependent Control Rbom filtration and inleakage, time-dependent atmospheric dispersion coefficients; and natural decay. Doses are calculated for individuals residing in the Control Room.

To efficiently model the main steam line break, AX3 incorporates the following enhancements over AXIDENT: (1) The format and contents of the data file were altered to incorporate the isotopic gas gap fractions and to incorporate either ICRP-2 or ICRP-30 dose conversion factors data. (2) Primary system and secondary system initial activity were incorporated into the model. (3) An instantaneous release of failed fuel activity into the primary system was modeled. (4) A time-dependent primary-to-secondary leak rate was incorporated into the model. (5) Activity initially in or leaking into the secondary system was assumed to be instantaneously released to the environment with a time-dependent partition factor for the primary and failed fuel iodine activities. (6) The ability to utilize time-dependent Control Room atmospheric dispe:sion coefficients was incorporated into the model. (7) The ability to use time-dependent Control Room occupancy factors and filtration rates was incorporated. (8) Note that since Containment cleanup is not an issue in these types of analyses, for each iodine isotope a single species (e.g., elemental) was modeled.

RESULTS The Control Room doses from an MSLB for the current plant condition assuming 5500 cfm inleakage:

Control Room Thyroid Dose 149 Rem,30 Rem after 60 minutes Whole Body Dose 4.55 Rem Deta Skin Dose 0.19 Rem CONCLUSIONS The Control Room operators will accumulate a thyroid dose of 30 Rem after 60 minutes. Therefore, adequate time is available to implement the compen;atory measures of Reference 14 such that compliance with General Design Criteria 19 using compensatory measures for thyroid dose is demonstrated.

The whole body and beta skin doses meet the General Design Criteria 19 limits of 5 and 30 Rem, respectively.

9

' 1

.. i e i ENCLOSURE (C)

MAIN STEAM LINE BREAK ACCIDENT REFERENCES

1. Nuclear Regulatory Commission Safety Guide 25, " Assumptions Used for Evaluating the Potential Radiological Consequences of a Fuel Handling Accident in the Fuel IIandling and Storage Facility for Boiling and Pressurized Water Reactors," March 23,1972

2. NUREG/CR-6331 Revision 1, " Atmospheric Relative Concentrations in Building Wakes,"

May 1997 l

3. BGE Drawing 60-330-E, " Heating and Ventilation System, Auxiliary Building, El. 69'0,"

Sections and Details," Revision 14

4. Regulatory Guide 1.49, Revision 1, " Power Levels of Nuclear Power Plants," December 1973 1 1

5. "CCNPP Core Operating Limits Report for Unit 2 Cycle 12," Revision 0 l

6. Letter from Mr. J. A. Tiernan (BGE) to Mr. A. C. Thadani (NRC), dated March 5,1986,

" Control Room Dose" l 7. Regulatory Guide 1.4 Revision 2, " Assumptions for Evaluating the Potential Radiological ,

l Consequences of a LOCA for PWRs," June 1974 l l

l 8. Standard Review Plan Section 6.4, Revision 2, " Control Room Habitability System," July 1981

9. LOCADOSE NE319, Revision 3

10. TID-14844, " Calculation of Distance Factors for Power and Test Reactor Sites,"

March 23,1992  ;

11. Letter from Mr. D. G. Mcdonald (NRC) to Mr. G. C. Creel, dated July 6.1992, " Approval for Calvert Cliffs Units I and 2 Fuel Pin Burnup Limit of 60 MWD /KG" l

12. Letter from Mr. A. C. Thadani (NRC) to Mr. A. E. Scherer (CE), dated June 22,1992, " Generic

! Approval of CE Topical Report CEN-386-P, Verification of the Acceptability of a 1-Pin Burnup Limit of 60 MWD /kg for CE 16x16 PWR Fuel"

13. Haliburton NUS Report NUS-1954, Revision 3, "AXIDENT: A Digital Computer Dose Calculation Model," February 1984

14. Letter from Mr. C. H. Cruse (BGE) to NRC Document Control Desk, dated May 6,1993, Control Room Habitability - Interim Engineering Analysis for Thyroid Dose 10