ML20216H752

Responds to RAIs & 970609 Re 970131 Request for Amend Re SG Tube Plugging

ML20216H752
Person / Time
Site:	Calvert Cliffs
Issue date:	03/17/1998
From:	Cruse C BALTIMORE GAS & ELECTRIC CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9803230164
Download: ML20216H752 (57)

Text

CurnLEs II. CnosE Baltimore Gas and Electric Company Vice President Calvert Cliffs Nuclear Power Plant Nuclear Energy 1650 Calvert Cliffs Parkway Lusby. Maryland 20657 410 495-4455 March 17,1998 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 Response to Request for Additional Information: Control Room Habitability Analyses and Main Steam Line Break Analyses

REFERENCES:

(a) Letter from Mr. C. H. Cruse (BGE) to NRC Document Control Desk, dated January 31, 1997, License Amendment Request; Change to Reactor Ceolant System Flow Requirements to Allow increased Steam Generator Tube Plugging (b) Letter from Mr. A. W. Dromerick (NRC) to Mr. C. H. Cruse (BGE),

dated March 5,1997, Request for Additional Information - Proposed Technical Specification Changes to Reactor Coolant System Flow Limit (TAC Nos. M97855 and M97856)

(c) Letter from Mr. A. W. Dromerick (NRC) to Mr. C. H. Cruse (DGE),

dated August 28, 1997, Extension of Control Room Habitability Analysis Submittal Date (TAC Nos. M99013 and M99014)

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

dated June 9,1997, Request for Additional Information - Proposed Technical Changes to Reactor Coolant System Flow Limit (TAC Nos. M97855 and M97856)

\l By Reference (a), Baltimore Gas and Electric Company submitted a license amendment request to the /

NRC 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 last two open questions related to this amendment request. Reference (b) requested revised accident analyses for the effect of the increased steam generator g c tube plugging limit on Control Room dose. Reference (c) also requested these analyses be performed. i Four of the six required analyses are discussed in Attachment (1). The last two required analyses will be provided before March 31,1998. The last open question (Reference d) requires a basis for assumptions 9803230164 980317 PDR P ADOCK 05000317 l Illy ll

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Docum:nt Control Desk March 17,1998 Page 2 used in the main steam line break analysis. This letter provides a reanalysis of the main steam line break that resolves all of the concerns stated in the question (Attachment 2).

Should you have questions regarding this matter, we will be pleased to discuss them with you.

Very truly yours, ,

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CllC/ PSF /dlm

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Attachment (1): Response to the Request for Additional Information dated March 5,10)7

Enclosures:

(A) Main Steam Line Break (B) Steam Generator Tube Rupture (C) Seized Rotor Event Attachment (2): Response to the Request for Additional Information dated June 9,1997 t

cc: R. S. Fleishman, Esquire II. J. Miller, NRC J. E. Silberg, Esquire Resident inspector, NRC A. W. Dromerick, NRC R. I. McLean, DNR Director, Project Directorate I-1, NRC J.11. Walter, PSC I

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l ATTACHMENT (1) l RESPONSE TO TIIE REQUEST FOR ADDITIONAL INFORMATION DATED MARCII 5,1997 I

l Baltimore Gas and Electric Company Calvert Clifts Nuclear Power Plant March 17,1998

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ATTACHMENT (1)

RESPONSE TO TIIE REQUEST FOR ADDITIONAL INFORhtATION DATED MARCII 5,1997 Question l

All revised accident analyses must assess the consequences of the accident with respect to GDC 19.

Please transmit the assumptions and dose associated with your analyses to demonstrate that GDC 19 is met.

In addition to this question, the NRC made a similar request in Reference (1), as noted below.

1. It is imperative that Baltimore Gas and Electric Company (BGE) have on their docket an analysis that indicates the status of Calvert Chffs to meet General Design Criterion (GDC) 19, not onlyfor the whole body, butfor other organs.

2. It is a requirement to meet GDC 19for allpostulated accidents. While the timing issue of NUREG-1465 may impact a loss-of-coolant accident (LOCA), it will have no impact on a Steam Generator Tube Rupture (SGTR) or a Main Steam Line Break (MSLB) Event, nor is it likely to impact a Control RodEjection or a Locked Rotor Event.

BESPONSE This response addresses both requests for information.

Backcround There were six Control Room habitability analyses requested by the NRC Staff. Two of them are the LOCA analysis and the fuel handling analysis that support the design basis for the Control Room ventilation system. These analyses are not complete and will be provided later. The other analyses requested are-MSLB, Seized Rotor Event (SRE), SGTR Event, and the Control Rod Ejection Event.

These analyses were requested to support a license amendment that changes the flow rate through the reactor core. The analyses for three of these events are provided as enclosures to this attachment. The fourth event did not have to be revised.

Main Steam Line Break Chapter 14.14 of the Updated Final Safety Analysis Report 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. A loss of offsite power, with the turbine trip, results in the maximum doses, since the loss of offsite power causes the reactor coolant pumps to coast down, minimizing core flow, lowering the transient departure form nucleate boiling ratio, and maximizing the number of failed fuel pins. i i

For the affected steam generator, the maximum secondary system Technical Specification activity, and i that fraction of the primary system Technical Specification and failed fuel activity which leaks to the secondary side of the steam generator, are discharged into the main steam piping room and out the main steam piping room vent on the roof of the Auxiliary Building. For the intact steam generator, the maximum secondary system 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 i activity that leaks to the secondary side of the intact steam generator is assumed to be discharged from the atmospheric dump valves (ADVs) or the main steam safety valves. Note that with a loss of offsite power, the condensers are not available for cooldown.

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ATTACllMENT (1)

RESPONSE TO TIIE REQUEST FOR ADDITIONAL INFORMATION DATED MARCII 5,1997 l

The MSLB was analyzed in conjunction with the core analyses supporting the low How amendment, which analyzed the effect of increasing the number of plugged tubes per steam generator from 800 to i

2500. The results were previously presented to the NRC. The NRC questioned some of the assumptions in that analysis. A re-analysis to resolve the NRC concerns is presented in Attachment (2). The results of this analysis indicate that fuel failure will be less than or equal to 1.35%.

The NRC also requested additional information regarding the Control Room doses that would result from the MSLB. Detailed information about this analysis is contained in Enclosure (A). This analysis evaluates Control Room habitability for the MSLB based on International Commission for Radiation Protection (ICRP)-30 dose conversion factors, ARCON96 generated atmospheric dispersion coef6cients to the west road inlet plenum,1.35% failed fuel, and a 3000 cfm control room inleakage. The Control Room doses from an MSLB with the flow going out the main steam piping room vents and the safety valves and the ADV are the following:

Control Room Thyroid Dose 29.30 Rem Whole Body Dose 0.90 Rem Beta Skin Dose 0.05 Rem Note that for this case, the Control Room doses ere less than the 10 CFR Part 50, Appendix A, General Design Criteria 19 thyroid, whole body, and beta skin dose limits of 30,5, and 30 rem, respectively.

Steam Generator Tube Ruoture Chapter 14.15 of the Updated Final Safety Analysis Report presents the licensing basis evaluation of the SGTR Event. An SGTR Event is defined as the penetration of the barrier between the Reactor Coolant System (RCS) and the Main Steam System. The integrity of this barrier is of radiological significance, in that a leaking steam generator tube allows the transfer of reactor coolant into the Main Steam System.

Radioactivity contained in the reactor coolant would then mix with the Guid in the secondary side of the affected steam generator. This radioactivity would then be transported by steam to the turbine / condenser / vent stack / atmosphere or directly to the atmosphere via the ADVs.

The design basis SGTR Event comprises a double-ended tube rupture. The initial secondary activity, together with the initial primary activity and the iodine spike activity released to the primary system that then leaks into the secondary system will escape out of the steam generators via the ADVs and the condenser.

The SGTR Event was analyzed in conjunction with the core analyses supporting the low How amendment, which analyzed the effect of increasing the number of plugged tubes per steam generator from 800 to 2500. The results were previously presented to the NRC (Reference 2). The NRC also requested additional information regarding the Control Room doses that would result from the SGTR.

Detailed information about this analysis is contained in Enclosure (B).

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ATTACIIMENT (1)

RESPONSE TO Tile REQUEST FOR ADDITIONAL INFORMATION DATED MARCII 5,1997 This analysis evaluates Control Room habitability for the SGTR based on ICRP-30 dose conversion factors, ARCON96 generated atmospheric dispersion coefficients to the west road inlet plenum, and a 3000 cfm Control Room inleakage. Activity releases are consistent with Reference (2). The Control Room doses from an SGTR with all of the Dow going out the ADVs are the following:

Control Room Thyroid Dose 3.89E-4 Rem Whole Body Dose 1.17E-5 Rem Beta Skin Dose Negligible The Control Room doses from an SGTR with all of the flow going to the condenser over the entire accident are the following:

Control Room Thyroid Dose 1.34E-1 Rem Whole Body Dose 7.00E-3 Rem Beta Skin Dose Negligible The Control Room doses are less than the 10 CFR Part 50, Appendix A, GDC 19 thyroid, whole body, and beta skin dose limits of 30,5, and 30 rem, respectively. Note that the consequences of an SGTR with the Dow going out the ADVs is bounded by the consequences of an SGTR with the Dow going to the condenser for the ContrM Room analyses. Thus, the condenser results will be used as the design basis limits for the Control Room results.

Seized Rotor Event Chapter 14.16 of the Updated Final Safety Analysis Report presents the licensing basis evaluation of the SRE. An SRE is defined as a complete seizure of a single reactor coolant pump shaft. An SRE is )

initiated at hot full power by an instantaneous complete seizure of a single reactor coolant pump shaft.

With the reduction of core now due to the loss of a reactor coolant pump, the coolant temperatures in the core will increase. Assuming a positive moderator temperature coefficient, the core power will increase.

The insertion of the control element assemblies due to a low RCS How trip will terminate the power rise; however, a limited number of fuel pins will experience departure from nucleate boiling for a short period of time and thus fail. The initial secondary system activity, together with initial primary system activity and failed fuel activity released to the primary system that then leaks into the secondary system, will escape out of the steam generators via the ADVs and condenser.

The SRE was analyzed in conjunction with the core analyses supporting the low flow amendment, which analyzed the effect of increasing the number of plugged tubes per steam generator from 800 to 2500.

The results were previously presented to the NRC (Reference 3). The results of the analysis indicate that fuel failure will be less than or equal to 5%.

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ATTACilMENT (1)

RESPONSE TO Tile REQUEST FOR ADDITIONA1, INFOlBIATION DATED MARCil 5,1997 The NRC subsequently requested additional information regarding the Control Room doses that would result from the SRE. Detailed information about this analysis is contained in Enclosure (C). This analysis evaluates Control Room habitability for the SRE based on high burnup fuel isotopics, ICRP-30 l

dose conversion factors, ARCON96 generated atmospheric dispersion coefficients to the west road inlet plenum,5% failed fuel, and a 3000 cfm Control Room inleakage. The Control Room doses from an SRE with the Dow for the first 30 minutes going out the ADVs and thereafter to the condenser are the following:

Control Room Thyroid Dose 1.39E-1 Rem Whole Body Dose 9.02E-3 Rem Beta Skia Dose 2.00E-1 Rem The Control Room doses from an SRE with the Dow going to the condenser over the entire accident are the following:

Control Room Thyroid Dose 1.66E-1 Rem Whole Body Dose 1.05E-2 Rem Beta Skin Dose 2.44E-1 Rem For all cases, the Control Room doses are less than the 10 CFR Part 50, Appendix A, GDC 19 thyroid, whole body, and beta skin dose limits of 30,5, and 30 rem, respectively. Note that the consequences of l an SRE with the now for the first 30 minutes going out the ADVs and thereafter to the condenser is bounded by the consequences of an SRE with the now going to the condenser over the entire accident for the Control Room analyses. Thus, the condenser-condenser results will be used as the design basis limits for the Control Room results.

Control Rod Election Event in our initial submittal for the low How amendment (Reference 3), we stated that the consequences of a Control Element Assembly Ejection Event did not change as a result of changing the Dow rate through the RCS. In a request for additional information (Reference 4), the NRC requested that we analyze five events. In our response (Reference 5), we stated that we would analyze four of the events. We also provided justi6 cation for not analyzing the Control Element Assembly Ejection Event. Our reasoning for not performing this analysis (i.e., that the quantity of fuel melting and gap release does not increase) was also discussed with NRC reviewers. Based on the fact that we do not have to revise the analysis for this event for the proposed change in the RCS How rate, we are not providing the results of a Control Room habitability analysis.

Lating To support the assumption made in the analyses, testing of the Control Room ventilation system was

! conducted in November and again in February. The testing was done to determine the amount of I unfiltered leakage into the Control Room envelope. The initial tests were conducted prior to any concentrated maintenance effort. The resulting inleakage was measured at 4300 cfm for IIcating, Ventilation, and Air Conditioning (liVAC) Train No.11 and 3000 cfm for llVAC Train No.12. This measured inleakage was greater than assumed in the analyses. Numerous maintenance activities were undertaken to reduce the unfiltered inleakage. The seats were replaced on some of the boundary dampers, and seams in the ductwork and fan housing were sealed. Additionally, the Hexible joints were 4

ATTACilMENT (1) l RESPONSE TO TIIE REQUEST FOR ADDITIONAL INFORMATION DATED MARCII 5,1997 l repaired and door seals were added to the boundary doors. It was discovered that the Control Room was l

operating at a negative pressure, thus making the inleakage worse. An attempt was made to balance the  ;

ventilation flow throughout all of the rooms that make up the Control Room ventilation envelope. The attempt was not successful. After the maintenance was completed, another test was run to determine if the inleakage had been reduced. This test simulated some additional sealing of the ventilation boundary (see below) and showed that the leakage had been reduced to approximately 2700 cfm for each train.

l The analyses had originally been done assuming 2000 cfm inleakage, but have now been changed to use l

3000 cfm inleakage for input. No additional tests are scheduled to be performed this year.

Cmumilments To achieve the dose analysis results given above, a modification must be performed. The analyses use atmospheric dispersion coefficients assuming that the intake point for the radioactive air is along the west face of the Auxiliary Building. The discharge point for the radioactive steam is on top of the Auxiliary Building roof. Currently there are openings in the roof that would allow this radioactive steam l into the Control Room envelope and the Control Room HVAC Equipment Room of the Auxiliary l Building. Therefore, we will seal up those openings in the roof, which would force the intake point to be i along the west face of the building, and the assumption used in the calculation will be validated. Due to the complexity of this work, it is not expected to be completed before December 1999. At the conclusion of this modification, habitability of the Control Room in accordance with GDC 19 will be assured for these accidents.

The Control Room habitability analyses for a LOCA and Fuel Handling Accident have not yet been l completed. Initial analyses were done based on an assumed leakage of 2000 cfm; however, the current test results do not support that assumption. New analyses are being prepared. Those analyses are expected to be provided to the NRC before March 31,1998. Plant modifi ations are also expected to be required to suppon the analytical assumptions. They will be detailed in the late March submittal.

In the interim period, while these analyses and modifications are being performed, the compensatory measures described in Reference (6) will remain in effect. This ensures continued operator safety during the period of these changes.

REFERENCES

1. Letter from Mr. A. W. Dromerick (NRC) to Mr. C.11. Cruse (BGE), dated August 28,1997, Extension of Control Room Habitability Analysis Submittal Date (TAC Nos. M99013 and M99014)

2. Letter from Mr. C. H. Cruse (BGE) to NRC Document Control Desk, dated January 22,1998, I

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 (TAC Nos. M97855 and M97856) l

3. Letter from Mr. C. H. Cruse (BGE) to NRC Document Conuol Desk, dated January 31,1997, License Amendment Request; Change to Reacior Coolant System Flow Requirements to Allow

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Increased Steam Generator Tube Plugging 5

A'ITACIIMENT (1)

RESPONSE TO TIIE REQUEST FOR ADDITIONAL INFORMATION DATED MARcil 5,1997

4. Letter from Mr. A. W. Dromerick (NRC) to Mr. C.11. Cruse (BGE), dated April 22,1997, Request for Additional Information - Proposed Technical Specification Changes to Reactor Coolant System Flow Limit Regarding Radiological Controls [sicJ (TAC Nos. M97855 and M97856)

5. Letter from Mr. C. II. Cruse (BGE) to NRC Document Control Desk, dated August 19,1997, Response to 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)

6. Letter from Mr. C.11. Cruse (BGE) to NRC Document Control Desk, dated May 6,1993, Control Room liabitability - Interim Engineering Analysis for Thyroid Dose l

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ENCLOSURE (A) TO ATTACHMENT (1) l l

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l MAIN STEAM LINE BREAK l

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Baltimore Gas and Electric Company Cdvert Cliffs Nuclear Power Plant March 17,1998 i

ENCLOSURE (A) TO ATTACilMENT (1)

MAIN STEAM LINE BREAK INTRODUCTION The Main Steam Line Break (MSLB) was analyzed in conjoen 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 NRC also requested additional information regarding the Control l Room doses that would result from the MSLB.

METIIODS OF ANALYSIS Chapter 14.14 of the Updated Final Safety Analysis Report 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 extraction by the steam generators and causes cooldown of the Reactor Coolant System (RCS). With a negative moderator temperature coefficient, the RCS cooldown will produce a positive reactivity addition that is terminated only when the control element assemblies insert af ter a reactor trip. The depressurization of the affected steam generator 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 generator to continue. The continued blowdown causes the RCS pressure to decrease l until the Safety Injection Actuation Signal is initiated, which automatically starts the high pressure safety l injection (IIPSI) pumps. Since the shutoff head to the IIPSI 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 HPSI pump flow will begin to add coolant mass, such that the l pressurizer level will be reestablished. A loss of offsite power, with the turbine trip, results in the l maximum site boundary doses. The loss of offsite power causes the reactor coolant pumps to coast

! down, minimizing core flow, lowering the transient departure from nucleate boiling ratio, and maximizing the number of failed fuel pins.

For the affected steam generator, the maximum secondary system Technical Specification activity, and that fraction of the primary system Technical Specification and failed fuel activity which leaks to the secondary side of the steam generator, are 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 l to withstand RCS operating pressure on the tube side with atmospheric pressure on the shell side, the l continued integrity of the RCS barrier is assured. Thus, only the maximum Technical Specification l primary-to-secondary leakage is assumed. A partition factor of one is assumed for all discharged radioactivity.

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l For the intact steam generator, the maximum secondary Technical Specification activity is also assumed l to be discharged out the main steam piping room vent. All of the primary-to-secondary Technical l Specification and failed fuel activity that leaks to the secondary is assumed to be discharged from the 1

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.

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ENCLOSURE (A) TO A'ITACllMENT (1)

MAIN STEAM LINE BREAK 1

This analysis was accomplished using 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 th; main steam safety valve. AX3 is equivalent to AX1 DENT, l

! TACT 5 and LOCADOSE.

( ASSUMPTIONS j The following assumptions were utilized in this analysis:

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

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l Buildup of daughter nuclides is not taken into account as source tenn nuclides decay per '

Reference (1).

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

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

e lleight oflower wind instrument: 10 meters e lleight of upper wind instrument: 60 meters e Wind speed units type: meters /second e Release type: vent I

• Release height: 17.15 meters Main steam piping room vent height is: 11 = 91.5" + 9.75' - 45.0' = 56.25' = 17.15 meters No thermal or momentum plume rise is assumed.

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• Building area: 1155 m l 2 The calculation of containment cross-sectional area yields 12435.63 ft above the rooftop level of elevation 91'6". The Auxiliary Building cross-sectional area is calculated to be 1938.93 ft2 ,

For a west-to-east wind direction, the total cross-sectional area of the Auxiliary Building and the 2

i two containments is 26810 ft . 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 total cross-sectional area of the containment and the turbine building is 21016 ft 2 The cross-2 2 sectional area of a single containment of 12435.63 ft or 1155 m will conservatively be used.

f j e Effluent vertical velocity: 0 meters /second

• 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

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ENCLOSURE (A) TO ATTACIIMENT (1)

MAIN STEAM LINE BREAK e Direction to source: Unit 1 - 25 degrees; Unit 2 - 63 degrees e Source window: 90 degrees e Distance from source to receptor: 59.04 meters e Intake height: 9.14 meters The Auxiliary Building roof above the Control Room and above the Control Room 11 eating, Ventilation, and Air Conditioning (11VAC) Equipment 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. The inlet plenum is 54'x10' with a bottom elevation of 70'. Thus, the intake height is 75'-45' = 30' = 9.14 m e

Grade elevation difference: 0 meters (Reference 2) e Surface roughness length: 0.1 meters (Reference 2) e Minimum wind speed: 0.5 meters /second (Reference 2) e Sector averaging constant: 4 (Reference 2) e Hours in average: 12 4 812 24 96168 360 720(Reference 2) e Minimum number of hours: 12 4 81122 87152 324 648(Reference 2) s

. Horizontal difTusion coefficient: 0.64 meters a y= r/2.15 = 1.38/2.15 = 0.64 meters (Reference 2)

. Vertical diffusion coefficient: 2.97 meters o z= 9.75' = 2.97 meters (Reference 2)

Example 5 of Reference 2 presumes a vertical diffusion coefficient of 1 meter for a capped vent, 1 meter above the roof. This case is similar, 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:

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

. Release height: 53.56 meters

> The weighted average wind speed is calculated using the 1991-1993 Joint Frequency Table (JFT) values for stability class F and the wind speed power law distribution with height relationship for stability class F. The JFT displays the nutnber of hours (n3) that the wind is at velocity vii n a specific velocity interval at the 10 meters or 60 meters primary meteorological tower (PMT) position. The average wind velocity for yeir 'x' at PMT position 'y' is thus:

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

• v,(x,y))/I,(n,(x,y)).

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ENCLOSURE (A) TO ATTACIIMENT (1)

MAIN STEAM LINE BREAK The average velocity over three years at PMT position 'y' is thus:

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

For the 10 meter and 60 meter PMT positions, the average velocities are 1.87 and 4.48 l

meters /second. The wind velocity at any height 'h' can then be determined from the wind speed i power law distribution with height relationship for stability class F v(h) = v(10m)*(h/10) 5 Use l of this relationship with the release height, All, calculated below, yields the wind velocity 2.46 meters /second.

> The ADV exit flow rate is:

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

> Thejet entrainment coefficient is:

beta = 1/3 + u/Vs = 0.37

> The minimum separation distance 'x' between the Units I and 2 ADVs and the west road inlet plenum is 61.70 meters.

> 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:

Fm = Vs2 *d 2

• (T,/4ffs) = 58.0090 m*/s 2

> The buoyancy flux is:

2 Fb = g*V 3*D *(AT/4ffs) = 2.2975 m'/s'

> The thermal plume rises is:

Allt = 1.6

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• x 2" / u = 13.4021 meters

> The momentum plume rise is:

2 o Alim = (3

• F. * [ sin (x* sos /u)/ beta /u/s .5))ia = 22.9420 meters where s = 0.02

• g / T, = 0.000669

> ADV height is:

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

> The combined release height is:

A11 tot = Allt + Alim + 11 = 53.5653 meters 2

e Building area:1155 m

. Effluent vertical velocity: 67.53 meters /second 221.5 ft/sec = 67.53 meters /second 4

f' ENCLOSURE (A) TO ATTACIIMENT (1)

MAIN STEAM LINE BREAK e Stack or vent flow: 3.44 m'/s 121.32 cf/sec = 3.44 m3/sec; Stack or vent radius: 0.13 meters d = 10.02" = 0.25 meters Direction to source: Unit 1 - 24 degrees; Unit 2 - 67 degrees e Source window: 90 degrees

• Distance from source to receptor: 61.7 meters e intake height: 9.14 meters e

Grade elevation difference: 0 meters (Reference 2) l .

Surface roughness length: 0.1 meters (Reference 2)

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

Hours in average: 12 4 812 24 96168 360 720 (Reference 2) e Minimum number of hours: 12 4 81122 87152 324 648 (Reference 2)

Horizontal diffusion coefficient: 0.0 meters e Vertical diffusion coefficient: 0.0 meters l

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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 e Expansion Factor: 0.704

• Limiting Diameter: 3.760 inches e Exit Diemeter: 10.020 inches l AX3 INPUTS The inputs into the AX3 code are as follows:

e Initial thermal power is 2754 MWt. (UFSAR Section 3.2. land Reference 4).

• The power peaking factor is 1.70. (Reference 5).

• The failed fuel fraction is 1.35%.

The specific activity of the secondary coolant system shall be less then 1.0E-7 Ci/ gram Dose Equivalent 1-131 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 i

I-131 per Technical Specification 3.4.8.

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ENCLOSURE (A) TO ATTACIIMENT (1)

MAIN STEAM LINE BREAK e

The speciGe 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.

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

The minimum RCS fluid mass is 457,437 lbm. This is based on an RCS volume without the pressurizer of 9601 cubic feet (UFSAR Table 4.1), a hot full power average temperature of 572.5 F (UFSAR Figure 4.9), and an RCS pressure of 2250 psia (UFSAR Table 4.1). The specific volume at those conditions is 0.02206 cf/lbm (45.33 lbm/cf- Reference 7). This is also based on a minimum pressurizer volume of 600 cubic feet (UFSAR Table 4-7), pressurizer pressure of 2250 psia (UFSAR Table 4-7), and a saturation temperature of 653 F (Reference 7). The specific volume at those conditians is 0.02700 cf/lbm (37.04 lbm/cf- Reference 7). Thus, the RCS mass is:

Macs = 9601*45.33 + 600*37.04 = 457437 lbm

. The me.ximum secondary fluid mass is 250,500 lbm.

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

. The tirr,e 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 seconds.

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

Unit 1 Unit 2 0-2 hrs 1.08E-03 sec/m3 1.05E-03 sec/m3 l 2-8 hrs 8.84E-04 sec/m3 7.90E-04 sec/m3 8-24 hrs 4.25E-04 sec/m3 3.49E-04 sec/m3 24-96 hrs 3.21E-04 sec/m3 2.39E-04 sec/m3 96-720 hrs 2.26E-04 sec/m3 1.69E-04 sec/m3 (Reference 2)

Atmospheric dispersion coefficients from the ADVs and/or the main steam safety valves to Control Room:

Unit 1 Unit 2 0-2 hrs 1.00E-09 sec/m7 1.00E-09 sec/m3 3 2-8 hrs 1.32E-09 sec/m3 1.00E-09 sec/m3 i 8-24 hrs 8.80E-10 sec/m3 1.00E-09 sec/m3 l 24-96 hrs 1.00E-09 sec/m3 I 1.00E-09 sec/m3 96-720 hrs 1.55E-08 sec/m3 5.68E-09 sec/m3 (Reference 2)

• The breasing rates are extracted from Reference 8:

0-8 hrs 3.47E-04 m3/sec 8-24 hrs 1.75E-04 m3/sec 24-720 hrs 2.32E-04 m3!sec e The Control Room occupancy factors are extracted from References 6 and 9:

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

l 6

l

r ENCLOSURE (A) TO ATTACIIMENT (1)

MAIN STEAM LINE BREAK e

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

e i Control Room inleakage: Control Room inteakage will be reduced to less than 3000 cfm. The l Auxiliary Building roof above the Control Room and above the Control Room HVAC Equipment j

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.

Control Room recirculation and filtration flow:

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

> Initiation delay time: 30 seconds [(Activates automatically on a SIAS (UFSAR Table 8-7)] 3 sec (SIAS) +10 sec (EDG startup) = 13 sec (Conservatively 30 see is assumed.)

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

> Iodine filter removal coefficients:

LF = (1800.cfm)*(60. min /hr)*0.90/(166000.cf) = 0.5855/hr = 1.6265E-04/sec e Partition Factors:

> Primary / secondary release through the ADVs: 1.0 I

> Primary / secondary release through the main steam piping room vents: 1.0 e The isotopic source terms (Cl/MWT) were extracted from Reference 10 and are consistent with TID-14844 methodology (Reference 11). The isotopic decay constants (1/sec) were also extracted from Reference 10.

SOURCE DECAY Isotope Cl/MWT 1/SEC I-131 2.508E+04 9.976E-07 I-132 3.806E+04 8.425E-05 1-133 5.622E+04 9.211 E-06 1-134 6.575E+04 2.200E-04 I-135 5.103 E+04 2.912E-05 XE-131M 2.595E+02 6.81 SE-07 XE-133M 1.384 E+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 2.11SE-05 XE-137 5.103E+04 3.024E-03 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 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 I

7 l

ENCLOSURE (A) TO ATTACIIMENT (1)

MAIN STEAM LINE BREAK t

Per References 1,12, and 13, 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

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

p INilALATION IMMERSION IMMERSION REM /CI REM-M3/Cl-S REM-M3/Cl S isotope TilYRotD W1IOLE IlODY TllYROID WilOLE BODY llETA SKIN 1-131 1.1 E+06 3.3 E+04 1-132 6.3 E+03 3.4 E+02 1-133 1.8E+05 5.6E+03 1-134 1.lE+03 1.lE+02 1-135 3.l E+04 1.!E+03 XE-131M 1.3E-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.lE-01  !

XE-135 4.7E-02 1.2E-01 XE-138 2.0E-01 2.0E-01 4.lE-01 KR-83h1 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.

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

COMPUTER CODES ARCON96 CODE METHODOLOGY The ARCON96 computer code implements a computational model for calculating atmospheric dispersion coefficients (E /Qs) in the vicinity of buildings (Reference 2). An atmospheric dispersion coefficient is simply the ratio of the relative concentration at the receptor (gm/m 3) 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 hourly meteorological data. This model also treats diffusion more realistically under low wind speed conditions than previous NRC-issued models.

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

8

ENCLOSURE (A) TO ATTACIIMENT (1)

MAIN STEAM LINE BREAK l

1 AX3 CODE METHODOLOGY l This calculation employed the AX3 computer code. AX3 models the transport of halogen and noble gas isotopes from a primary containment to a secondary containment and then to the environment and f Control Room.

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) primary dose equivalent 1-131 and Eisenbud noble gas activity with Kr-85 half-life; and (3) secondary system dose equivalent I 131 activity. 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 Room 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 AX1 DENT: (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 DCF 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 dispersion 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 with the flow going out the main steam piping room vents and the ADVs and/or main steam safety valves are the following:

Control Room Thyroid Dose 29.30 Rem Whole Body Dose 0.90 Rem Beta Skin Dose 0.05 Rem CONCLUSIONS The Control Room doses are less than the 10 CFR Part 50, Appendix A, General Design Criterion 19 thyroid, whole body, and beta skin dose limits of 30,5, and 30 rem, respectively.

l 9

I.

l ENCLOSURE (A) TO ATTACIIMENT (1)

MAIN STEAM LINE BREAK REFERENCES

1. NRC Safety Guide 25, " Assumptions Used for Evaluating the Potential Radiological Consequences of a Fuel IIandling Accident in the Fuel llandling and Storage Facility for

! Boiling and Pressurized Water Reactors," March 23,1972 l

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

l May 1997

3. BGE Drawing 60-330-E, "lieating 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

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

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

" Control Room Dose"

7. Electrical Research Association, St. Martin's Press,"1967 Steam Tables," 1967

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 IIabitability System," July 1981

10. LOCADOSE NE319, Revision 3

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

March 23,1992

12. Letter from Mr. D. G. Mcdonald (NRC) to Mr. G. C. Creel, dated July 6,1992, " Approval for Calvert Cliffs Units 1 and 2 Fuel Pin Burnup Limit of 60 MWD /KG"

13. 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 Accegability of a 1-Pin Burnup Limit of 60 MWD /kg for CE 16x16 PWR Fuel"

14. IIaliburton NUS Report NUS-1954, Revision 3, "AX1 DENT: A Digital Computer Dose Calculation Model," February 1984 1

l 10

i ENCLOSURE (B) TO ATTACIIMENT (1)

{

l l

I l

STEAM GENERATOR TUBE RUPTURE I

i i

l l

l llattimore Gas and Electric Company Calvert Cliffs Nuclear Power Plant March 17,1998

ENCLOSURE (B) TO ATTACIIMENT (1) 4 STEAM GENERATOR TUBE RUPTURE INTRODUCTION The Steam Generator Tube Rupture (SGTR) Event was analyzed in conjunction with the low How amendment, which analyzed the effect of increasing the number of plugged tubes per steam generator from 800 to 2500. The results were presented to the NRC. The NRC also requested additional information regarding the Control Room doses that would result from the SGTR.

METIIODS OF ANALYSIS Chapter 14.15 of the Updated Final Safety Analysis Report (UFSAR) presents the licensing basis evaluation of the SGTR Event. A SGTR Event is defined as the penetration of the barrier between the Reactor Coolant System (RCS) and the Main Steam System. The integrity of this barrier is of radiological signi0cance, in that a leaking steam generator tube allows the transfer of reactor coolant into the main steam system. Radioactivity contained in the reactor coolant would then mix with the Guid in the secondary side of the affected steam generator. This radioactivity would then be transported by steam to the turbine and condenser, then to the vent stack and from there to the atmosphere or directly to the atmosphere via the atmospheric dump valves (ADVs).

The design basis SGTR Event comprises a double-ended tube rupture. The primary-to-secondary coolant transfer far exceeds the charging pump capacities and, consequently, steam generator level increases while pressurizer level decreases. The decrease in the pressurizer level and the inability of the heaters to maintain pressurizer pressure causes RCS depressurization, which initiates a Thermal Margin / Low Pressure and a Safety injection Actuation Signal reactor / turbine trip ensuring that the speciDed acceptable fuel design limits are not exceeded; however, a generated iodine spike with a 500 spiking factor does result. As the pressure drops below the high pressure safety injection pump shut-off head, safety injection How is delivered to the core and stabilizes the RCS pressure at the high pressure safety injection pump head. The operator then identifies and manually isolates the steam generator with the ruptured tube and initiates a cooldown per the Emergency Operating Procedures. The initial secondary activity, together with the initial primary activity and the iodine spike activity released to the primary that then leaks into the secondary, will escape out of the steam generators via the ADVs and the condenser.

This calculation was accomplished using the following methodologies:

Atmospheric dispersion coefficients from the release point to the intet plenum were calculated via the ARCON96 computer code, e

The AX2 and AX3 codes were executed to determine the 30-day Control Room doses from an SGTR with the now exiting the ADV or the condenser. The AX2 and AX3 codes are radiation transport codes similar to AXIDENT, LOCADOSE, and TACTS.

ASSUMPTIONS The following assumptions were utilized in this analysis:

e No credit 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 and offsite doses from an SGTR are the following:

1

ENCLOSURE (B) TO ATTACHMENT (1)

STEAM GENERATOR TUBE RUPTURE ARCON96 INPUTS The X /Q was determined for the vent stack 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 /second

. Release type: vent

. Release height: 48.29 meters Ventilation Stack height is extracted from Reference 2. No thermal or momentum plume rise is assumed.

H = 203' 5"- 45.0' = 158.42' = 48.29 m

. Building area: 1155 m 2 2

The calculation of containment cross-sectional area yields 12435.63 ft above the rooftop level of 91'6". The Auxiliary Building cross-sectional area can be calculated to be 1938.93 ft2 . For a west-to-east wind direction, the total cross-sectional area of the Auxiliary Building and the two 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-tc,-south and south-to-north wind direction, the total cross-sectional area of the containment and the turbine building is 21016 ft2 . The cross-2 2 sectional area of a single containment of 12435.63 ft or 1155 m will conservatively be used.

• Effluent vertical velocity: 0 meters /second e Stack or vent flow: 0 m'/s

• Stack or vent radius: 0.91 meters The exit diameter is extracted from Reference 3.

d = 6' = 1.8288 meters

• Direction to source: Unit 1 - 15 degrees; Unit 2 - 75 degrees e Source window: 90 degrees e Distance from source to receptor: 66.49 meters e intake height: 9.14 meters The Auxiliary Building roof above the Control Room and above the Control Room IIVAC Equipment 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. Per Reference 10, the inlet plenum is 54'x10' with a bottom elevation of 70'. Thus, the intake height is 75'-45' = 30'

= 9.14 meters.

Grade elevation difference: 0 meters (Reference 4) e Surface roughness length: 0.1 meters (Reference 4)

• Minimum wind speed: 0.5 meters /second (Reference 4)

• Sector averaging constant: 4 (Reference 4)

+ Hours in average: 12 4 812 24 96168 360 720 (Reference 4) l 2

I ENCLOSURE (B) TO ATTACIIMENT (1)

STEAM GENERATOR TUBE RUPTURE e

Minimum number of hours: 12 4 81122 87152 324 648 (Reference 4) e llorizontal diffusion coefficient: 0.0 meters ey= r/2.15 = 0.9144/2.15 = 0.43 meters (Reference 4) l e Vertical diffusion coefficient: 0.0 meters The Z / Q was determined for the atmospheric dump valve to Control Room flow path using the .

following inputs:

[

• lleight oflower wind instrument: 10 meters q

e lleight of upper wind instrument: 60 meters

. Wind speed units type: meters /second e Release type: vent j

e Release height: 53.56 meters

> The weighted average wind speed is calculated using the 1991-1993 Joint Frequency Table (JFT) values for stability class F and the wind speed power law distribution with height relationship for i stability class F. The JFT displays the number of hours (n,) that the wind is at velocity vi in a '

specific velocity interval at the 10 meter or 60 meter primary meteorological tower (PMT) position. The average wind velocity for year 'x' at PMT position 'y' is thus:

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

• v,(x,y))/I,(n,(x,y))

1 The average velocity over three years at PMT position 'y' is thus: 1 v(y) = E x(v(x,y))/3 For the 10 meter and 60 meter PMT positions, the average velocities are thus 1.87 and 4.48 meters /second. The wind velocity at any height 'h' can then be determined from the wind speed power law distribution w-ith height relationship for stability class F.

v(h) = v(10m)*(h/10)"

Use of this relationship with the release height, All, calculated below yields the wind velocity 2.46 meters /second.

> The ADV exit flow rate is:

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

> Thejet entrainment coefficient is:

beta = 1/3 + u/Vs = 0.37

> The minimum separation distance 'x' between the Units 1 and 2 ADV and the west road inlet plenum is 61.70 meters.

> The exit diameter is extracted from Reference 3.

d = 10.02" = 0.2545 m

> 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.

3

ENCLOSURE (H) TO ATTACIIMENT (1)

STEAM GENERATOR TUBE RUPTURE

> The momentum flux is:

F = Vs 2*d2*(T,/4/r )s= 58.0090 m*/s2

> The buoyancy flux is:

Fb = g*Vs*D 2*(AT/4/r )s = 2.2975 m'/s 3

> The thermal plume rises is:

AHt = 1.6

• F3 "

• x" / u = 13.4021 m t

> The momentum plume rise is: '

2 Allm = (3

• F *m[ sin (x*s .5/u)/ beta /u/s 5j )u3 = 22.9420 meters where s = 0.02

• g / T, = 0.000669 l

> ADV height is:

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

> The combined release height is:

Alitot = Allt + Allm + 11 = 53.5653 raeters e Building area: 1155 m 2

. Effluent vertical velocity: 67.53 meters /second 221.5 ft/sec = 67.53 meters /second e Stack or vent flow: 3.44 m'/s 3

121.32 cf/sec = 3.44 m /sec e Stack or vent radius: 0.13 meters The exit diameter is:

d = 10.02" = 0.25 meters

• Direction to source: Unit 1 - 24 degrees; Unit 2 - 67 degrees

• Source window: 90 degrees

• Distance from source to receptor: 61.7 meters e intake height: 9.14 meters

• Grade elevation difference: 0 meters (Reference 4)

• Surface roughness length: 0.1 meters (Reference 4) e Minimum wind speed: 0.5 meters /second (Reference 4)

• Sector averaging constant: 4 (Reference 4) e flours in average: 12 4 812 24 96168 360 720(Reference 4) l Minimum number of hours: 12 4 81122 87152 324 648(Reference 4)

!

• Ilorizontal diffusion coefficient: 0.0 meters I.

e Vertical diffusion coefficient: 0.0 meters f ,

I1 4

ENCLOSURE (B) TO ATTACHMENT (1)

STEAM GENERATOR TUBE RUPTURE The steam flow rates through the ADVs as a function of steam generator temperature are calculated with the following inputs:

• Atmospheric pressure: 14.700 psia e Resistance coefDcient: 9.820 K e Maximum DP/Pl: 0.771

. Expansion Factor: 0.704

. Limiting Diameter: 3.760 inches e Exit Diameter: 10.020 inches I l

1 AX3 INPUTS i The AX3 inputs are as follows:

j e

Initial thermal power is 2754 MWt. (UFSAR Section 3.2.1 and Reference 5) e The power peaking factor is 1.70. (Reference 6)

The failed fuel fraction is 0%. (Reference 7)

The specific activity of the secondary coolant system shall be less than 1.0E-7 Ci/ gram Dose Equivalent 1-131 per Technical Specification 3.7.1.

The specific activity of the primary coolant is included in the AX2 executions as part of the generated iodine spike.

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/ dis.

The Control Room volume of 166000 ft' is extracted from Reference 8.

• The minimum RCS fluid mass is 457,437 lbm. This is based on an RCS volume without the pressurizer of 9601 cubic feet (UFSAR Table 4.1), a hot full power average temperature of 572.5 F (UFSAR Figure 4.9), and an RCS pressure of 2250 psia (UFSAR Table 4.1). The specific volume at those conditions is 0.02206 cf/lbm (45.33 lbm/cf- Reference 9). This is also based on a minimum pressurizer volume of 600 cubic feet (UFSAR Table 4-7), pressurizer pressure of 2250 psia (UFSAR Table 4-7), and a saturation temperature of 653 F (Reference 9). The specific volume at those conditions is 0.02700 cf/lbm (37.04 lbm/cf- Reference 9). Thus, the RCS mass is:

Macs = 9601*45.33 + 600*37.04 = 457437 lbm

. 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 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> or 21,600 seconds. Technical Specification 3.4.9 sets a maximum allowable cooldown rate of 100 F/hr, while UFSAR Figure 4-9 defines the hot full power RCS coolant temperature as 572.5 F. Thus, assuming a cooldown rate at only half of the Technical Specification limit will result in a shutdown cooling temperature of

~270 F in 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

5

i l

ENCLOSURE (B) TO ATTACIIMENT (1)

STEAM GENERATOR TUBE RUPTURE l Atmospheric dispersion coefficients from the vent stack to the inlet plenum:

Unit 1 Unit 2 0-2 hrs 1.08E-03 sec/m3 9.81E.04 sec/m3 2-8 hrs 7.99E-04 sec/m3 6.62E-04 sec/m3 8-24 hrs 3.83E-04 sec/m3 2.71E-04 sec/m3 24-96 hrs 2.86E-04 sec/m3 1.84E-04 sec/m3 96-720 hrs 2.04E-04 sec/m3 1.34E-04 sec/m3 (Reference 4) l e Atmospheric dispersion coefficients from ADV to Control Room: j Unit 1 Unit 2 0-2 hrs 1.00E-09 sec/m3 1.00E-09 sec/m3 2-8 hrs 1.32E-09 sec/m3 1.00E-09 sec/m3 8-24 hrs 8.80E-10 sec/m3 1.00E-09 sec/m3 24-96 hrs 1.00E-09 sec/m3 1.00E-09 sec/m3 l 96-720 hrs 1.55E-08 sec/m3 5.68E-09 sec/m3

{

(Reference 4) e The breathing rates are extracted from Reference 10:

{

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 8 and 11:

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

The primary-to-secondary leak rate is 200 gal / day (100 gal / day per steam generator) or 0.1389 gpm 1 per Technical Specification 3.4.6. I e Control Room inleakage: Control Room inleakage will be reduced to less than 3000 cfm. The Auxiliary Building roof above the Control Room and above the Control Room IIVAC Equipment 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.

• Control Room recirculation and filtration flow:

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

> Initiation delay time: 30 seconds (Activates automatically on a SIAS; UFSAR Table 8-7) 3 sec (SIAS) +10 sec (EDG startup) = 13 sec (Conservatively 30 see will be assumed.)

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

> Iodine filter removal coefficients:

LF = (1800 cfm)*(60 min /hr)*0.90/(166000 cf) = 0.5855/hr = 1.6265E-04/sec i

l I

6 i 1

i

ENCLOSURE (B) TO ATTACIIMENT (1)

STEAM GENERATOR TUBE RUPTURE e Partition Factors:

> Primary release through ADVs: 1.0 (Reference 7)

> Secondary release through ADVs: 1.0 (Reference 7)

> Release through condenser: 0.0005 (Reference 7)

The isotopic source terms (Cl/MWT) were extracted from Reference 12 and are consistent with TID-14844 methodology (Reference 13). The isotopic decay constants (1/sec) were also extracted from Reference 12.

SOURCE DECAY Isotope CI/MWT 1/SEC 1-131 2.508E+04 9.976E-07 l-132 3.806E+04 8.425E-05 I-l33 5.622E+04 9.211 E-06 1-134 6.575 E+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 5.363 E+04 2.I 15 E-05 XE-137 5.103 E+04 3.024E-03 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 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 e

Per References 1,14, and 15, 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 7

ENCLOSURE (B) TO ATTACIIMENT (1)

STEAM GENERATOR TUBE RUPTURE e

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

INIIALATION IMMERSION IMMERSION REM /Cl REM-M3/CI-S REM-M3/CI-S isotope IllYROID WilOLE 130DY lilYROID WilOLE 130DY llETA SKIN I-131 1.lE+06 3.3E+04 I-132 6.3E+03 3.4E+02 1-133 1.8E+05 5.6E+03 1-134 1.lE+03 1.lE+02 I-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 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-83hi 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 AX2 INPUTS The AX2 inputs not included in the above are as follows:

The generated iodine spike 1-131 activity released to the environment from the ADV in the SGTR Event is:

7200 sec 200.9 Ci 28800 sec 1036.0 Ci These values are conservatively increased by 20%.

7200 sec 250 Ci 28800 see 1250 Ci Thus,1250 Ci ofI-131 are released as a result of an SGTR to the environment. Twenty percent of this activity is conservatively released at t = 0 sec, while the remaining 80% is released at t = 7200 seconds. The ADV partition factor is one.

The generated iodine spike I-131 activity released to the environment from the condenser in the SGTR Event is identical to that given above. The condenser partition factor is 0.0005.

COMPUTER CODES ARCON96 CODE METIlODOLOGY The ARCON96 computer code implements a computational model for calculating atmospheric dispersion coefficients (E /Qs) in the vicinity of buildings (Reference 4). 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/m3 . The model f estimates impacts from ground-level, vent, and elevated releases using a single year or multi-years of 8

ENCLOSURE W TO ATTACIIMENT (1)

STEAM GN"CRATOR TUBE RUPTURE hourly meteorological data. This model also treats diffusion more realistically under low wind speed conditions than previous NRC-issued models.

This work calculates the atmospheric dispersion coefficients from the vent stack to the inlet plenum assuming no thermal or momentum plume rise and from the ADVs to the inlet plenum assuming momentum and thermal plume rises.

AX2 CODE METHODOLOGY AX2 models the transport of halogen and noble gas isotopes from a primary containment to a secondary containment and then to the environment and Control Room.

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 that releases radioactivity within the containment or within any primary system. 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; Contro! Room 6!tration and inleakage; primary and secondary containment purge filters; Control Room intake Elters; atmospheric dispersion; and natural decay. Doses are calculated for individuals residing in the Control Room.

This code was executed to determine the 30-day Control Room doses from an SGTR generated iodine spike with the flow going out the ADVs, and from an SGTR generated iodine spike with the flow going to the condenser.

AX3 CODE METHODOLOGY AX3 models the transport of halogen and noble gas isotopes from a primary containment to a secondary containment and then to the environment and Control Room.

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) primary dose equivalent 1-131 and dose equivalent Kr-85 activity; and (3) secondary system dose equivalent 1-131 activity. AX3 models the transport of radioactivity (iodine, krypton, and xenon isotopes) from the primary to the secondary and then to the environment and to the Control Room. The code assumes instantaneous failed fuel activity release 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-l dependent Control Room filtration and inleakage, time-dependent atmospheric dispersion; and natural l decay. Doses are calculated for individuals residing in the Control Room.

l This code was executed to determine the 30-day Control Room doses from primary noble gas and l secondary iodine activity released after an SGTR with the flow going out the ADVs or to the condenser, f

[

9

l I

ENCLOSURE (B) TO ATTACHMENT (1)

STEAM GENERATOR TUBE RUPTURE RESULTS The Control Room doses from an SGTR with all of the flow going out the ADVs are the following:

Control Room Thyroid Dose 3.89E-4 Rem Whole Body Dose 1.17E-5 Rem Beta Skin Dose Negligible The Control Room doses from an SGTR with all of the flow going to the condenser over the entire ,

accident are the following: l Control Room Thyroid Dose 1.34E-1 Rem

{

Whole Body Dose 7.00E-3 Rem l Beta Skin Dose Negligible CONCLUSIONS For all cases, the Control Room doses are less than the 10 CFR Part 50, Appendix A, General Design Criterion 19 thyroid and whole body dose limits of 30 and 5 rem, respectively. Note 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 condenser results will be used as the design basis limits for the Control Room results.

REFERENCES

1. NRC Safety Guide 25, " Assumptions Used for Evaluating the Potential Radiological I

Consequences of a Fuel Handling Accident in the Fuel Handling and Storage Faciity for Boiling and Pressurized Water Reactors," March 23,1972

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

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

( Sections and Details," Revision 14 l

l 4. NUREG/CR-6331 Revision 1, " Atmospheric Relative Concentrations in Building Wakes,"

May 1997

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

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

7. Letter from Mr. C. H. Cruse (BGE) to NRC Document Control Desk, dated January 31,1997,

" Supplement to LAR; Change to Reactor Coolant System Flow Requirements to Allow Increased Steam Generator Tube Plugging"

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

" Control Room Dose"

9. Electrical Research Association, St. Martin's Press,"1967 Steam Tables," 1967

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

11. Standard Review Plan Section 6.4, Revision 2, " Control Room liabitability System," July 1981 10

ENCLOSURE (B) TO ATTACIIMENT (1)

STEAM GENERATOR TUBE RUPTURE

12. LOCADOSE NE319 Revision 3

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

March 23,1962

14. Letter from Mr. D. G. Mcdonald (NRC) to Mr. G. C. Creel, dated July 6,1992, " Approval for Calvert Cliffs Units 1 and 2 Fuel Pin Burnup Limit of 60 MWD /KG"

15. 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" i

i i

lI

t ENCLOSURE (C) TO ATTACHMENT (1)  !

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SEIZED ROTOR EVENT l

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Baltimore Gas and Electric Company Calvert Cliffs Nuclear Power Plant March 17,1998

ENCLOSURE (C) TO ATTACIIMENT (1)

SEIZED ROTOR EVENT l INTRODUCTION I The Seized Rotor Event (SRE) was analyzed in conjunction with the core analyses supporting the low flow amendment, which analyzed the effect of increasing the number of plugged tubes per steam I generator from 800 to 2500. The results were previously presented to the NRC. The NRC subsequently l requested additional information regarding the Control Room doses that would result from the SRE.

I i

METIIODS OF ANALYSIS l Chapter 14.16 of the Updated Final Safety Analysis Report presents the licensing basis evaluation of the SRE. An SRE is defined as a complete seizure of a single reactor coolant pump shaft. The seizure is postulated to occur due to a mechanical failure or a loss of component cooling to the pump shaft seats.

The most limiting SRE is an instantaneous reactor coolant pump shaft seizure at hot full power. The reactor coolant flow through the core would be asymmetrically reduced to three pump Dow as the result of a shaft seizure on one pump.

An SRE is initiated at hot full power by an instantaneous complete seizure of a sir.gle reactor coolant pump shaft. With the reduction of core How due to the loss of an reactor coolant pump, the coolant ,

temperatures in the core will increase. Assuming a positive moderator temperature coefficient, the core power will increase. The core average heat flux will decrease slightly due to the increasing core temperatures. The insertion of the control element assemblies due to a low RCS flow trip will terminate i the power rise; however, a limited number of fuel pins will experience departure from nucleate boiling l for a short period of time and, thus, fail. The initial secondary system activity, together with initial primary system activity and failed fuel activity released to the primary system that then leaks into the secondary system, will escape out of the steam generators via the atmospheric dump valves (ADVs) and condenser.

This analysis was accomplished using the following methodologies:

• Atmospher;c dispersion coefficients from the release point to the inlet plenum were calculated via the ARCON96 computer code.

e The AX3 code was executed to determine the 30-day Control Room doses from an SRE with the flow for the first 30 minutes going out the ADVs and thereafter to the condenser and from an SRE with the flow going to the condenser over the entire accident.

ASSUMPTIONS The following assumptions were utilized in this work:

e No credit is taken for deposition of the plume on the ground or decay ofisotopes 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 Contrcl Room doses from a Seized Rotor Event are the following:

ARCON96 INPUTS The 7 /Q was determined for the vent stack to Control Room flow path using the following inputs:

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

l 1 l

ENCLOSURE (C) TO ATTACHMENT (1)

SEIZED ROTOR EVENT

• Wind speed units type: meters /second

• Release type: vent

• Release height: 48.29 meters Ventilation Stack height is extracted from Reference 2. No thermal or momentum plume rise is assumed.

H = 203' 5"- 45.0' = 158.42' = 48.29 meters

• Building area: 1155 m 2 2

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

containments is 26810 ft . For an east-to-west wind direction, the total cross-sectional area 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.

• Effluent vertical velocity: 0 meters /second

• Stack or vent flow: 0 m'/s

• Stack or vent radius: 0.91 seters The exit diameter is e:.tracted from Reference 3.

d = 6' = 1.8288 meters

• Direction to source: Unit 1 - 15 degrees; Unit 2 - 75 degrees

• Source window: 90 degrees

• Distance from source to receptor: 66.49 meters

• Intake height: 9.14 meters The Auxiliary Building roof above the Control Room and above the Control Room IIVAC Equipment 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. Per Reference 3, the inlet plenum is 54'x10' with a bottom elevation of 70'. Thus, the intake height is 75'-45' = 30' '

9.14 meters.  !

1 Grade elevation difference: 0 meters (Reference 4)

• Surface roughness length: 0.1 meters (Reference 4) l

• Minimum wind speed: 0.5 meters /second (Reference 4) f

• Sector averaging constant: 4 (Reference 4)

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

Minimum number of hours: 12 4 81122 87152 324 648(Reference 4)

• Horizontal diffusion coefficient: 0.0 meters c y= r/2.15 = 0.9144/2.15 = 0.43 meters (Reference 4)

• Vertical diffusion coefficient: 0.0 meters 1

2

ENCLOSURE (C) TO ATTACIIMENT (1)

SEIZED ROTOR EVENT The X /Q was determined for the atmospheric dump valves to Control Room Dow path using the following inputs:

e Height oflower wind instrument: 10 meters e Height of upper wind instrument: 60 meters e Wind speed units type: meters /second

. Release type: vent e Release height: 53.56 meters

> The weighted average wind speed is calculated using the 1991-1993 Joint Frequency Table (JFT) values for stability class F and the wind speed power law distribution with height relationship for stability class F. The JFT displays the number of hours (ni) that the wind is at velocity vi ni 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) = I,(a,(x,y)

• v,(x,y))/I(n,(x,y)).

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

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

For the 10 meter and 60 meter PMT positions, the average velocities are 1.87 and 4.48 l meters /second. 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) = v(10m)*(h/10)". Use l of this relationship with the release height, AH, calculated below, yields the wind velocity 2.46 l meters /second.

> The ADV exit flow rate is:

Flow rate = 7279 efm = 3.44 m'/sec 2

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

> Thejet entrainment coefficient is:

beta = 1/3 + u/Vs = 0.37

> The minimum separation distance 'x' between the Units 1 and 2 ADVs and the west road inlet plenum is 61.70 meters.

> 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:

Fm = Vs2 *d2

• (T,/4fTs) = 58.0090 m*/s2

> The buoyancy flux is:

F3 = g*Vs*D'*(AT/4ffs) = 2.2975 md /s' 3

ENCLOSURE (C) TO ATTACIIMENT (1)

SEIZED ROTOR FX?.NT

> The thermal plume rises is:

2 AHt = 1.6

• Fb'"

• x / u = 13.4021 meters

> The momentum plume rise is:

2 l Allm = {3

• F,n * [ sin (x*s'5/u)/ beta fyj,a5))it3 = 22.9420 meters where s = 0.02

• g / T = 0.000669

> ADV height is:

11 = 91.5' + 10' - 45.0' = 56.5' = 17.22 meters s l

l > The combined release height is- '

Alltot = AHt + Allm + 11 = 53.5653 meters 2

e Building area:1155 m

. Effluent vertical velocity: 67.53 meters /second 221.5 ft/sec = 67.53 meters /second 3

• Stack or vent flow: 3.44 m /s 121.32 cf/sec = 3.44 m'/sec; e Stack or vent radius: 0.13 meters d = 10.02" = 0.25 meters

• Direction to source: Unit 1 - 24 degrees; Unit 2 - 67 degrees e Source window: 90 degrees

• Distance from source to receptor: 61.7 meters

• Intake height: 9.14 meters e

Grade elevation difference: 0 meters (Reference 4) e Surface roughness length: 0.1 meters (Reference 4)

e Minimum wind speed

0.5 meters /second (Reference 4) l

• Sector averaging constant: 4 (Reference 4) e liours in average: 12 4 812 24 96168 360 720 (Reference 4) e Minimum number of hours: 12 4 81122 87152 324 648 (Reference 4) e llorizontal diffusion coefficient: 0.0 meters e Vertical diffusion coefficient: 0.0 meters The steam flow rates through the ADVs as a function of steam generator temperature are calculated with the following inputs:

• Atmospheric pressure: 14.700 psia e Resistance coefficient: 9.820 K e Maximum DP/Pl: 0.771 4

ENCLOSURE (C) TO ATTACHMENT (1) j SEIZED ROTOR EVENT e Expansion Factor: 0.704 e Limiting Diameter: 3.760 inches e 1 Exit Diameter: 10.020 inches AX3 INPUTS The AX3 inputs are as follows:

Initial thermal power is 2754 MWt. (UFSAR Section 3.2.1 and Reference 5)

• The power peaking factor is 1.70. (Reference 6) e The failed fuel fraction is 5%. (Reference 7) j e

The specific activity of the secondary coolant system shall be less than 1.0E-7 Ci/ gram Dose l Equivalent 1-131 per Technical Specification 3.7.1.

The specific iodine activity of the primary coolant shall be less than 1.0E-6 Ci/ gram Dose Equivalent I-131 per Technical Specification 3.4.8.

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/ dis.

The Control Room volume of 166000 ft' is extracted from Reference 8.

. The minimum RCS fluid mass is 457,437 lbm. This is based on an RCS volume without the l pressurizer of 9601 cubic feet (UFSAR Table 4.1), a hot full power average temperature of 572.5*F l (UFSAR Figure 4.9), and an RCS pressure of 2250 psia (UFSAR Table 4.1). The specific volume at those conditions is 0.02206 cf/lbm (45.33 lbm/cf- Reference 9). This is also based on a minimum

pressurizer volume of 600 cubic feet (UFSAR Table 4-7), pressurizer pressure of 2250 psia (UFSAR i

Table 4-7), and a saturation temperature of 653 F (Reference 9). The specific volume at those conditions is 0.02700 cf/lbm (37.04 lbm/cf- Reference 9). Thus, the RCS mass is:

hincs = 9601*45.33 + 600*37.04 = 457437 lbm

. 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 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> or 21,600 seconds. Technical Specification 3.4.9 sets a maximum allowable cooldown rate of 100 F/hr, while UFSAR Figure 4-9 defines the hot full powe- RCS coolant temperature as 572.5 F. Thus, assuming a cooldown rate at only half of the Technical Specification limit will result in a shutdown cooling temperature of

~270*F in 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

5

ENCLOSURE (C) TO A'ITACIIMENT (1)

SEIZED ROTOR EVENT e

Atmospheric dispersion coefficients from the vent stack to the inlet plenum:

Unit 1 Unit 2 0-2 hrs 1.08E-03 sec/m3 9.81E-04 sec/m3 2-8 hrs 7.99E-04 sec/m3 6.62E-04 sec/m3 8-24 hrs 3.83E-04 sec/m3 2.71E-04 sec/m3 24-96 hrs 2.86E-04 sec/m3 1.84E-04 sec/m3 96-720 hrs 2.04E-04 sec/m3 1.34E-04 sec/m3 (Reference 4)

. Atmospheric dispersion coef ficients from ADV to Control Room:

Unit i Unit 2 0-2 hrs 1.00E-09 sec/m3 1.00E-09 sec/m3 2-8 hrs 1.32E-09 sec/m3 1.00E-09 sec/m3 8-24 hrs 8.80E-10 sec/m3 1.00E-09 sec/m3 24-96 hrs 1.00E-09 sec/m3 1.00E-09 sec/m3 96-720 hrs 1.55E-08 sec/m3 5.68E-09 sec/m3 (Reference 4)

• The breathing rates are extracted from Reference 10:

0-8 hrs 3.47E-04 m3/sec 8-24 hrs 1.75E-04 m3/sec 24-720 hrs 2.32E-04 m3/sec e The Control Room occupancy factors are extracted from References 8 and 11:

0-24 hrs 1.0 24-96 hrs 0.6 96-720 hrs 0.4 The primary-to-secondary leak rate is 200 gal / day (100 gal / day per steam generator) or 0.1389 gpm per Technical Specification 3.4.6.

l e Control Room inleakage: Control Room inleakage will be reduced to less than 3000 cfm. The Auxiliary Building roof above the Control Room and above Control Room liVAC Equipment 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.

• Control Room recirculation and filtration flow:

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

> Initiation delay time: 30 seconds [(Activates automatically on a SIAS (UFSAR Table 8-7)] 3 sec (SIAS) +10 sec (EDG startup) = 13 sec (Conservatively 30 see will be assumed.)

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

> Iodine filter removal coefficients:

LF = (1800.cfm)*(60. min /hr)*0.90/(166000.cf) = 0.5855/hr = 1.6265E-04/sec 6

I

( e .

ENCLOSURE (C) TO ATTACHMENT (1)

SElZED ROTOR EVENT e Partition Factors:

i

> Primary / secondary release through the ADVs: 1.0

> Primary / secondary re: ease through the main steam piping room vents: 1.0

> Primary / secondary release through the condensers: 0.0005 e The isotopic source terms (CI/MWT) were extracted from Reference 12 and are consistent with TID-14844 methodology (Reference 13). The isotopic decay constants (1/sec) were also extracted j from Reference 12. '

SOURCE DECAY Isotope C1/MWT 1/SEC l l-131 2.508E+04 9.976E-07 i 1-132 3.806E+04 8.425E-05 I-133 5.622E+04 9.211 E-06 i 1-134 6.575E+04 2.200E-04  !

I-135 5.103 E+04 2.912E-05 l 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 l XE-135M 1.557E+04 7.380E-04 i 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 1.297E404 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

l Per References 1,14, and 15, 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 l

10% other noble gases 7

ENCLOSURE (C) TO ATTACilMENT (1)

SEIZED ROTOR EVENT e

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

INilALATION IMMERSION IMMERSION REhi/CI REM M3/CI-S REM-M3/CI-S isotope TilYROID WilOLE BODY TIIYROID WilOLE BODY llETA SKIN I-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-13 t hi 1.3 E-03 1.8E-02 XE-133M 5.4 E-03 3.8E-02 XE-133 7.3 E-03 6.3 E-03 2.0E-02 XE-135M 7.7E-02 1.1E-01 XE-135 4.7E-02 1.2E-01 I XE-138 2.0E-01 2.0E-01 4.lE-01 KR-83M 3.7E-06 1.8E-04 KR-85M 3.1 E-02 3.0E-02 8.5E-02 I KR-85 4.7E-04 4.8E-02 l KR-87 1.4E-01 1.5E-01 5.2E-01 KR-88 3.8E-01 3.7E-01 5.4E-01 e The isotopic gamma energies and fractions are detailed in Reference 16.

. The energy-dependent total and energy absorption coefficients are detailed in Reference 16.

COMPUTER CODES ARCON96 CODE METHODOLOGY The ARCON96 computer code implements a computational model for calculating atmospheric dispersion coefficients (X /Qs) in the vicinity of buildings (Reference 4). 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/m2 . The model j 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 l

conditions than previous NRC-issued models.

This work calculates the atmospheric dispersion coefficients from the vent stack to the inlet plenum assuming no thermal or momentum plume rise and from the ADVs to the inlet plenum assuming momentum and thermal plume rises.

AX3 CODE METHODOLOGY AX3 models the transport of halogen and noble gas isotopes from a primary containment to a secondary containment, and then to the environment and Control Room. 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) primary system dose equivalent 1-131 arid Eisenbud noble gas activity with Kr-85 half-life; and (3) secondary system dose equivalent 1-131 activity. 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 8

l ENCLOSURE (C) TO ATTACIIMENT (1)

SEIZED ROTOR EVENT Control Room. The code assumes instantaneous failed fuel activity release 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 l

activities, time-dependent Control Room filtration and inleakage, time-dependent atmospheric dispersion; and natural decay. Doses are calculated for individuals residing in the Control Room.

l To efficiently model accidents such as the Seized Rotor, AX3 incorporates the following enhancements l over AXIDENT. (1) The format and contents of the data file were altered to incorporate the isotopic gas j gap fractions and to incorporate either ICRP-2 or ICRP-30 dose conversion factor data. (2) Primary l system and secondary system initial activity were incorporated into the model. (3) An instantaneous j 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 system and failed fuel iodine activities. (6) The ability to utilize time-dependent Control Room atmospheric dispersion 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.

This work was done to determine the 30-day Control Room doses from an SRE with the flow for the first 30 minutes going out the ADVs and thereafter to the condenser and from an SRE with the flow going to the condenser over the entire accident.

RESULTS The Control Room doses from an SRE with the flow for the first 30 minutes going out the ADVs and thereafter to the condenser are the following:

l Control Room Thyroid Dose 1.39E-1 Rem Whole Body Dose 9.02E-3 Rem Beta Skin Dose 2.00E-1 Rem The Control Room doses from an SRE with the flow going to the condenser over the entire accident are the following.

( Control Room Thyroid Dose 1.66E-1 Rem Whole Body Dose 1.05E-2 Rem Beta Skin Dose 2.44E-1 Rem CONCLUSIONS For all cases, the Control Room doses are less than the 10 CFR Part 50 Appendix A General Design j Criterion 19 thyroid and whole body dose limits of 30 and 5 rem, respectively. Note that the i consequences of an SRE with the flow for the first 30 minutes going out the ADVs and thereafter to the l 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 condenser-condenser results will be used as the design basis limits for the Control Room results..

9

ENCLOSURE (C) TO ATTACHMENT (1)

SElZED ROTOR EVENT REFERENCES 1.

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

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

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

Sections and Details," Revision 14 4.

{

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

May 1997

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

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

7. Letter from Mr. C. H. Cruse (BGE) to NRC Document Control Desk, dated January 31,1997,

" Supplement to LAR; Change to Reactor Coolant System Flow Requirements to Allow Increased Steam Generator Tube Plugging"

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

" Control Room Dose"

9. Electrical Research Association, St. Martin's Press, "1967 Steam Tables," 1967

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

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

12. LOCADOSE NE319 Revision 3 l 13. TID-14844, " Calculation of Distance Factors for Power and Test Reactor Sites,"

March 23,1962

14. Letter from Mr. D. G. Mcdonald (NRC) to Mr. G. C. Creel, dated July 6,1992, " Approval for Calveit Cliffs Units I and 2 Fuel Pin Burnup Limit of 60 MWD /KG"

15. 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 ~0 MWD /kg for CE 16x16 PWR Fuel"

16. Haliburton NUS Report NUS-1954, Revision 3, "AX1 DENT: A Digital Computer Dose Calculation Model," February 1984 10

ATTACHMENT (2) I i

RESPONSE TO TIIE REQUEST FOR ADDITIONAL INFORMATION ,

DATED JUNE 9,1997 Baltimore Gas and Electric Company Calvert Cliffs Nuclear Power Plant March 17,1998

ATTACIIMENT (2)

RESPONSE TO Tile REQUEST FOR ADDITIONAL INFORMATION DATED JUNE 9,1997 l

l Question Please provide a basis for the 3 second time delay ofloss-of-offsite power LOOP) assumed in the main steam line break (MSLB) analyses described in the amendment request.

Main Steam Line Break Event This analysis responds to the NRC request by removing the 3-second time delay from the analysis instead of providing a basis for it.

t Event DescripliDD l

The MSLB analyzes a break in a main steam line. An MSLB may occur as a result of thermal stress or I cracking. The break increases the rate of steam extraction by the steam generators (SGs) and causes a Reactor Coolant System temperature reduction. With a negative moderator temperature coefficient (MTC), the Reactor Coc: ant System (RCS) temperature reduction will produce a positive reactivity insertion. Two analyses are performed to evaluate the reactivity insertion: the post-trip analysis and the pre-trip analysis.

The post-trip analysis evaluates the system response and core reactivity balance long after insertion of the control element assemblies (CEAs). The cooldown associated with a large break, combined with the most negative MTC, may result in enough positive reactivity insertion to overcome the negative

) reactivity inserted by the CEAs, and present the possibility of a return-to-power. The main stearn line flow restrictors limit the steam flow and resultant cooldown for a break outside of the containment.

l Therefore, the post-trip MSLB analysis is performed for an MSLB inside the containment. A post-trip MSLB analysis was performed to analyze the effect of reduced primary coolant mass flow associated with an increase in the number of plugged SG tubes. An increase of 2 F in the core inlet temperature was considered. The results of the post-trip MSLB analysis demonstrate that the site boundary dose is less than 10 CFR Part 100 guidelines, and the core remains coolable; therefore, the post-trip MSLB results do not constitute an unreviewed safety question.

l The pre-trip MSLB analysis evaluates departure from nucleate boiling ratio (DNBR) at or near the time of trip. The pre-trip MSLB analysis involves parametric analysis on a range of break sizes and MTC l

values to determine the combination that leads to the most limiting DNBR and highest fuel failure rate.

Breaks inside containment and outside containment are considered. Inside containment breaks are l considered due to the instrument uncenainty degradation associated with a harsh environment.

Degradation of instrument uncertainties may result in an additional delay between event initiation and l reactor trip.

The MSLB is classified as a Postulated Accident. The action of the Limiting Safety System Settings, in conjunction with the Limiting Condition for Operations, will limit the number of fuel pins that experience departure from nucleate boiling (DNB). Site boundary dose must not exceed the 10 CFR Part 100 limits. Since the event involves an RCS temperature reductica, primary system pressure decreases, and the RCS pressure upset limit is not approached.

I

ATTACIIMENT (2)

RESPONSE TO Tile REQUEST FOR ADDITIONAL INFORMATION DATED JUNE 9,1997 Analnis i The objectives of the pre-trip MSLB are to demonstrate that a coolable core geometry is maintained and that the site boundary dose is within the 10 CFR Part 100 limits.

The pre-trip MSLB transient analyses include a parametric analysis to determine the limiting break size and MTC combination that lead to the highest power at the time of reactor trip. The trips that are credited are the Variable Iligh Power trip, Low SG Pressure trip, Low SG Level trip, and liigh Containment Pressure trip. Both the Delta-T Power and the excore Nuclear Instrument Power signals are credited as part of the Vs.%ble liigh Power trip. For the Variable High Power trip signal based upon l Delta-T Power, the worst case initial deviation between the Delta-T Power and Nuclear Instrument I l Power signals is considered along with the worst case time response of the Delta-T Power signal. For the l Variable liigh Power trip signal based upon the excore detectors, the decalibration of the detectors that occurs as the reactor coolant temperature decreases is considered.

l The core mass How used in the analysis reDects the reduction it. flow associated with an increase in the j number of plugged SG tubes. The core inlet temperature was increased by 2 F. A loss of AC power was

assumed concurrently with the reactor trip. No credit is taken for a time delay between the reactor trip l and the loss of AC power. The primary effect of a loss of AC power is that primary Dow decreases. If I the Dow reduction begins prior to CEA insertion, the power (heat Hux) to Dow ratio will increase, and a lower DNBR will result.

l The calculation of the percentage of fuel failures used convolution methodology. This methodology was previously approved for use for the MSLB in Reference (1). As the number of fuel failures is limited, DNB propagation is postulated not to occur, and a coolable core geometry is maintained.

l The calculation of the offsite dose uses the Technical Specification primary-to-secondary leakage rate limit of 100 gallons per day per SG. The site boundary atmospheric dispersion coef0cient used is 1 3

1.3 x 10" second/m , and the breathing rate is 3.47 x 10" m'/second. The dose conversion factors from  !

[ International Committee on Radiation Protection] ICRP-30 were used. These dose conversion factors j have previously been approved for the Calvert Cliffs Fuel liandling incident (Reference 2). It is j conservatively assumed for the offsite dose calculation that 1.35% of the fuel pins fail. The Control Room dose calculation for this event is addressed separately in Attachment (1).

Results The results of the pre-trip MSLB analysis show that the predicted number of fuel pin failures is limited l to s 1.35%. As the number of fuel failures is limited, DNB propagation is postulated not to occur, and a ]

coolable geometry is maintained. The resultant site boundary doses, based on 1.35% fuel failures, are-0 - 2 Hr Thyroid Dose = 5 REM 0 - 2 Hr Whole Body Dose = 0.2 REM j Table 1 provides the initiel conditions and input parameters. Table 2 provides the assumptions used for the MSLB dose calculation. Table 3 provides the sequence of events. Figures 1 through 7 plot the following parameters for the MSLB: Moderator Reactivity versus Moderator Density, Reactor Power and Heat Flux versus Time, Core Average Heat Flux and Core Average Flow versus Time, RCS Pressure versus Time, RCS Temperatures versus Time, Reactivities versus Time, and SG Pressures versus Time.

2

f ATTACilMENT (2) {

RESPONSE TO T!!E REQUEST FOR ADDITIONAL, INFORMATION DATED JUNE 9,1997 REFERENCES

1. Letter from Mr. D. G. Mcdonald, Jr. (NRC) to Mr. R. E. Denton (BGE), dated May 11,1995,

" Approval to Use Convolution Technique in Main Steam Line Break Analysis - Calvert Cliffs Nuclear Power Plant, Unit Nos. I and 2 (TAC Nos. M90897 and M90898)"

2. Letter from Mr. M. J. Case (NRC) to Mr. R. E. Denton (BGE), dated August 31, 1994,

" Issuance of Amendments for Calvert Cliffs Nuclear Power Plant, Unit No.1 (TAC i

No. M88193) and Unit No. 2 (TAC No. M88194)  !

l l l I

I I

I I

+

3

ATTACIIMENT (2) l RESPONSE TO Tile REQUEST FOR ADDITIONAL INFORMATION DATED JUNE 9,1997 i l I

TABLE 1 1

INITIAL CONDITIONS AND INPUT PARAMETERS ASSUMED f TIIE OUTSIDE-CONTAINMENT MAIN STEAM LINE BREAK EVENT l INITIATED FROM IIOT FULL POWER j PARAMETER i UNITS VALUE

{

initial Core Power MWt 2754 Initial Core Inlet Temperature F 550 Initial RCS Pressure psia 2300 Initial SG Pressure psia 865 Low SG Pressure Analysis psia 650 Trip Setpoint Minimum CEA Worth Available at Trip  % Ap -5.4 1

Doppler Multiplier ---

0.85 i Moderator Cooldown Curve  % Ap vs Figure 2 density Most Negative MTC x 10" Ap/ F -3.0 Beta Fraction (Including Uncertainty) ---

.0044 Number of Plugged Tubes per SG ---

0(')

RCS Flow Ibte gpm 340,000

)

(*)

Up to 2500 plugged tubes per SG was considered; zero plugged tubes is more limiting.

4

ATTACIIMENT (2)

RESPONSE TO TIIE REQUEST FOR ADDITIONAL INFORMATION DATED JUNE 9,1997 TABLE 2 ASSUMPTIONS FOR TIIE RADIOLOGICAL EVALUATION OF TIIE MAIN STEAM LINE BREAK EVENT PARAMETER UNITS VALUE RCS Maximum Allowable pCi/gm 1.0 Concentration (dose equivalent I-131)(*)

Secondary Maximum Allowable pCi/gm 0.1 Concentration (dose equivalent I-131)(')

Partition Factor Assumed for ---

1.0 All Doses Atmospheric Dispersion Coefficient *) sec/m' l.3 x 10" Breathing Rate 3 d m /sec 3.47 x 10 Dose Conversion Factors REM /Ci O

(*)

Technical Specification limits.

• ) 0-2 hour accident condition.

(')

Dose conversion factors obtained from ICRP-30.

5

ATTACIIMENT (3)

RESPONSE TO TIIE REQUEST FOR ADDITIONAL INFORMATION DATED JUNE 9,1997 TABLE 3 SEOUENCE OF EVENTS FOR OUTSIDE CONTAINMENT STEAM LINE BREAK ANALYSIS TIME (scs) EVENT SETPOINT OR VALUE 0.0 Steam Line Break Occurs 2 0/15 ft 18.0 Variable High Power Trip Generated on Delta-T Power I;t0%

18.4 Trip Breakers Open --

18.5 Maximum Core Power Occurs 130.0% of rated power 18.9 CEAs Begin to Drop into Core ---

19.5 Minimum DNBR Occurs ---

l 33.9 Main Steam Isolation Valves Close ..- l l

l l

6

l i O ATTACIIMENT (2)

RESPONSE TO THE REQUEST FOR ADDITIONAL INFORMATION DATED JUNE 9,1997 FIGURE 1 MSLB OUTSIDE CONTAINMENT MODERATOR REACTIVITY VS MODERATOR DENSITY i 8

i I  !

! I g 6 -

+

t E l @ -

l BR i -

D 4 -

s i:

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m m

$ 2 -

ta m

m -

t O ,

1 O l 2

0 -

i

-2 40 45 50 55 60 65 MODERATOR DENSITY, LB/FT3 .

7

ATTACHMENT (2)

RESPONSE TO THE REQUEST FOR ADDITIONAL INFORMATION DATED JUNE 9,1997 FIGURE 2 1 MSLB OUTSIDE CONTAINMENT l CORE POWER AND IIEAT FLUX VS TIME 1.4 CORE POWER l

1.2 Y 1

z O

P O

k x l 3 0.8 u.

l H

d CORE HEAT FLUX I

O z l 4 l a:

W 0.6

[

O o.

E i O l 0 l 0.4 l

[ i 0.2 0 ' '

O 10 20 30 40 50 60 70 80 TIME, SECONDS 8

ATTACHMENT (2)

RESPONSE TO THE REQUEST FOR ADDITIONAL INFORMATION DATED JUNE 9,1997 FIGURE 3 MSLB OUTSIDE CONTAINMENT CORE HEAT FLUX AND CORE AVERAGE FLOW VS TIME 1.4 CORE HEAT Fl.UX

\4 1.2 1

z o $ \

6 /

CORE FLOW l 5 0.8 l d ,

I H l

@ l I

i Z 4 0.6 _

it S

u e

o 0

0.4 l

0.2 0

0 20 40 60 80 100 120 TIME, SECONDS 9

8 O ATTACIIMENT (2)

RESPONSE TO Tile REQUEST FOR ADDITIONAL INFORMATION DATED JUNE 9,1997 FIGURE 4 MSLB OUTSIDE CONTAINMENT RCS PRESSURE VS TIME i

2400 l

l 2200 2000-5 E

S 1800 e

m Q.

E  !

N m

1600 i.

i g  !

. 5 i

O i O U

1400 -

l l so  !

6 x

1200-1000 800 ' ' '

0 20 40 60 80 100 120 140 160 180 200 TIME, SECONDS 1

10 i

e o ATTACHMENT (2)

RESPONSE TO TIIE REQUEST FOR ADDITIONAL INFORMATION DATED JUNE 9,1997 FIGURE 5 MSLB OUTSIDE CONTAINMENT RCS TEMPERATURES VS TIME 650 600 m

h E 550 5

u e

5 g 500 o

$ T outlet o a d l E

verag  !

450

• j l

I 400 '

0 20 40 60 80 100 120 140 160 180 200 TIME, SECONDS 11

o ATTACIIMENT (2)

RESPONSE TO TIIE REQUEST FOR ADDITIONAL INFORMATION DATED JUNE 9,1997 FIGURE 6 MSLB OUTSIDE CONTAINMENT REACTIVITIES VS TIME 0.02 0.01 MODERATOR I

DOPPLER 0 3 40 60 80 10 1;:0

-0.01 o

x 4

5 ai

-0.02 i

o b

ct:

-0.03

\

TOTAL

-0.04

-0.05 CEA'S

-0.06 TIME, SECONDS 12 l

' 9 ATTACHMENT (2)

RESPONSE TO THE REQUEST FOR ADDITIONAL INFORMATION DATED JUNE 9,1997 FIGURE 7 MSLB OUTSIDE CONTAINMENT SG PRESSURE VS TIME 900 800 700

$ 600 E

EI a

E 500 E

M J

& 400 I

w 0

E 300 200 100 0

0 20 40 60 80 100 120 TIME, SECONDS l 13

.. ..