Text

Request for Additional Information Regarding Implementation for Alternative Source Term
	ML070870110
Person / Time
Site:	Calvert Cliffs
Issue date:	03/22/2007
From:	Spina J; Calvert Cliffs, Constellation Energy Group
To:	; Document Control Desk, NRC/NRR/ADRO
References
	TAC MC8845, TAC MC8846
	Download: ML070870110 (80)
	v • d • e

James A. Spina Calvert Cliffs Nuclear Power Plant, Inc.

Vice President 1650 Calvert Cliffs Parkway Lusby, Maryland 20657 410.495.5200 410.495.3500 Fax 0 Constellation Energy° Generation Group March 22, 2007 U. S. Nuclear Regulatory Commission Washington, DC 20555 ATTENTION: Document Control Desk

SUBJECT:

Calvert Cliffs Nuclear Power Plant Unit Nos. I & 2; Docket Nos. 50-317 & 50-318 Request for Additional Information Regarding Implementation for Alternative Source Term (TAC Nos. MC8845 and MC8846)

REFERENCES:

(a) Letter from Mr. B. S. Montgomery (CCNPP) to Document Control Desk, dated November 3, 2005, License Amendment Request: Revision to Accident Source Term and Associated Technical Specifications (b) Letter from Mr. P. D. Milano (NRC) to Mr. J. A. Spina (CCNPP), dated December 22, 2006, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2

-- Request for Additional Information Regarding Implementation of Alternative Source Term (TAC Nos. MC8845 and MC8846)

Reference (a) submitted a request to revise the accident source term in the design basis radiological consequence analyses. Reference (b) requested additional information needed to complete the Nuclear Regulatory Commission review of the proposed change. Our response to Reference (b) is contained in Attachment (1). When providing our response, we determined that two of the marked up Technical Specification pages provided in Reference (a) need to be revised. The revised Technical Specification pages are contained in Attachment (2) and replace the same pages in Reference (a).

Note that this additional information does change the No Significant Hazards Consideration by removing the waste gas incident and the waste processing incident from the list of incidents considered for this amendment request. However, the responses to the questions in the No Significant Hazards Consideration section of Reference (a) are not changed nor is the Environmental Consideration discussion affected by the responses to these questions.

Document Control Desk March 22, 2007 Page 2 Should you have questions regarding this matter, please contact Mr. Jay S. Gaines at (410) 495-5219.

Ve truly yours, STATE OF MARYLAND

TO WIT:

COUNTY OF CALVERT I, James A. Spina, being duly sworn, state that I am Vice President - Calvert Cliffs Nuclear Power Plant, Inc. (CCNPP), and that I am duly authorized to execute and file this License Amendment Request on behalf of CCNPP. To the best of my knowledge and belief, the statements contained in this document are true and correct. To the extent that these statements are not based on my personal knowledge, they are based upon information provided by other CCNPP employees and/or consultants. Such information has been reviewed in accordance with company practice and I believe it to be reliable.

Subscribed and sworn before me, a Notary Public in and for the State of Maryland and County of o ,this ha

day of "60a-I ,2007.

WITNESS my Hand and Notarial Seal: 7/ ey ý ýaý,

Notary Publiv/

My Commission Expires:

t Date JAS/PSF/bjd Attachments: (1) Request for Additional Information -- Implementation of Alternative Source Term Appendix 1: Sensitivity Study of Atmospheric Dispersion Coefficients for Releases from ADVs and MSSVs Appendix 2: Evaluation of Release from the RWT to Atmosphere Appendix 3: Sensitivity Study of Carbon Filter Dose Model (2) Marked up Technical Specification Pages

Document Control Desk March 22, 2007 Page 3 cc: D. V. Pickett, NRC Resident Inspector, NRC S. J. Collins, NRC R. I. McLean, DNR

ATTACHMENT (1)

REQUEST FOR ADDITIONAL INFORMATION -

IMPLEMENTATION OF ALTERNATIVE SOURCE TERM Calvert Cliffs Nuclear Power Plant, Inc.

March 22, 2007

ATTACHMENT (1)

REQUEST FOR ADDITIONAL INFORMATION -- IMPLEMENTATION OF ALTERNATIVE SOURCE TERM By letter dated November 3, 2005 (Reference 1) we requested an amendment to revise the accident source term in the design basis radiological consequence analyses at Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2. In order to complete its review, the Nuclear Regulatory Commission (NRC) staff requested the following additional information.

ACCIDENT AND TRANSIENT ANALYSES Steam Generator (SG) Tube Rupture (SGTR) Accident

1. Are the SG blowdown system and waste processing system safety-related systems? If not, what is the basisfor assumingcreditfor these systems in the mitigation of SGTR accident activity releases? Are these systems part of the TMI [Three Mile Island] Action Item III D. 1.1 Leakage Reduction Program, and was their leakage incorporated in the analyses of post-SGTR accident activity releases? If not, why?

Response: One of the cases presented in Reference 1, Attachment 1, Section 4.1.4, described an option for the operator to continue to cool down the Reactor Coolant System (RCS) via the atmospheric dump valve (ADV) of the unaffected SG but also allowed the use of the SG blowdown system (and the waste processing system) during the cooldown phase of the event. The SG blowdown system and the waste processing system are not safety-related and are not relied upon to provide cooldown support during the event as described in the Updated Final Safety Analysis Report (UFSAR) Section 14.15. Therefore, this case is removed from the supporting SGTR design basis calculation (Enclosure 4 to Reference 1) to better align the calculation with the cases presented in the UFSAR Section 14.15. We are revising Section 4.1.4 of Attachment 1 (Reference 1) as follows:

4.1.4 Steam Generator Tube Rupture (SGTR)

Section 14.15 of the CCNPP UFSAR describes the design basis evaluation of the SGTR Event. A SGTR Event is defined as the penetration of the barrier between the RCS and the main steam system. The integrity of this barrier is of radiological safety significance, in that a leaking SG tube allows the transfer of reactor coolant into the main steam system. Radioactivity contained in the reactor coolant would then mix with the fluid in the secondary side of the affected SG. This radioactivity would then be transported by steam to the turbine/condenser/vent stack/atmosphere or directly to the atmosphere via the atmospheric dump valves (ADVs) or main steam safety valves.

The limiting SGTR Event is considered to be a complete double-ended tube break and is postulated to occur due to a complete failure of a tube-to-sheet weld or the rapid propagation of a circumferential crack.

The SGTR Event allows primary coolant to leak into the secondary side via the SG. The current design basis accident assumes that cooldown of the RCS to shutdown cooling (SDC) conditions utilizes the ADVs of both SGs until SDC conditions are attained. In this analysis, the ADV of the affected SG must be isolated after two hours to limit activity emissions. The following two dhfee-cooldown operational modes are considered:

The operator continues the cooldown via the ADV of the unaffected SG until the SDC entry conditions are reached. It will take approximately 14 days for the decay heat generation to decline to a level that can be removed via a single SG and ADV. Note that a 30 day cooldown via the ADV of the unaffected SG is conservatively modeled in this analysis.

The operator- eontinues the cooldown v ia the ADV. of the unaffc.t. d SG but also establishs conditiens that allow the use of the SG blowdown system during the eeeldown phas.e~ te event. Nete that use of SG blewdown can be assumed to occur-at any timce dur-ing the cooldown

ATTACHMENT (1)

REQUEST FOR ADDITIONAL INFORMATION -- IMPLEMENTATION OF ALTERNATIVE SOURCE TERM phase of the event. However, the off-site and Control Room deses are bouinded by these of the firs*teptil.

The operators can re-open the ADV of the affected SG for up to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months after an initial cooldown of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months post-accident.

Enclosure (4) contains the detailed SGTR radiological consequences design basis calculation using the AST. The following supporting calculations are also provided: Atmospheric Dispersion Coefficient (X/Q) calculation (Enclosure 9), Source Terms calculation (Enclosure 10), and Primary and Secondary Isotopic calculations (Enclosure 12).

Major assumptions and required plant modifications considered in the SGTR re-analysis include:

A bounding Control Room in-leakage value of 3,500 cfm,

" Modification of the CREVS to a nominal 10,000 cfm flow with 90 percent filtration efficiency for elemental and organic iodine and 99 percent for particulate iodine was credited,

Installation of automatic isolation dampers and radiation monitors at Access Control HVAC Unit RTU-1 and Access Control Air Conditioning Unit 13 on the Auxiliary Building roof,

Revision to the TS 3.4.15 limit for RCS activity from 1.0 gCi/gm to 0.5 jtCi/gm.

Per Regulatory Guide 1.183 (Reference 3), if no or minimal fuel damage is postulated, the activity should be the maximum coolant activity allowed by the TSs, assuming two cases of iodine spiking. Thus, two cases are modeled to determine the maximum offsite and Control Room doses.

PIS Case: A reactor transient is assumed to occur prior to the postulated SGTR and has raised the primary coolant iodine concentration to the maximum value permitted by the TSs, which is 60 times the new TS 3.4.15 limit of 0.5 jCi/gm.

CIS Case: The primary system transient associated with the SGTR is assumed to cause an iodine spike in the primary system. The increase in primary coolant iodine concentration is estimated using a spiking model that assumes the iodine release rate from the fuel rods to the primary coolant increases to a value 335 times greater than the release rate corresponding to the iodine concentration at the equilibrium value. Per Regulatory Guide 1.183, the assumed iodine spike duration should be 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

Table 4 summarizes the results of the SGTR re-analysis for the EAB, LPZ, and Control Room doses for the design-basis CIS and PIS cases for the two three-operational cooldown modes described above

[cooldown via ADV of unaffected SG from 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to 30 days, and additional cooldown via the ADV of the affected SG from 0-2 and 24-32 hours,...and ld.wn via the unaffected SG with blowdown to the Waste Preessing System (.,PS) . The corresponding regulatory dose limits are also shown. The calculated doses in all cases are less than the regulatory limit.

2

ATTACHMENT (1)

REQUEST FOR ADDITIONAL INFORMATION -- IMPLEMENTATION OF ALTERNATIVE SOURCE TERM Table 4 Control Room EAB (Rem) LPZ (Rem) (Rem)

Cases CIS Unaffected ADV 0-2 hr-30 days 0.1964 0.0484 1.7081 CIS Affected ADV 0-2/24-32 hr 0.1964 0.0476 1.6929 CIS Unaffccted ADV 0 2 hr/WP-S 04964 004484 4--7484-CIS Regulatory Limits 2.5 2.5 5 PIS Unaffected ADV 0-2 hr-30 days 0.4910 0.1164 4.1590 PIS Affected ADV 0-2/24-32 hr 0.4910 0.1162 4.1655 PIS Unafflctcd ADV 0 2 h25WPS 0491-0 05464 PIS Regulatory Limits 25 25 5 The above section replaces Section 4.1.4 in Attachment I of Reference 1.

2. How are tank ventilation and other possible activity leakage pathwaysfrom the SG blowdown and waste processingsystems accountedfor in taking creditfor their mitigativefunction?

Response: Since the SG blowdown system and the waste processing system are not necessary for mitigating the design basis SGTR Event, all references to the SG blowdown system and waste processing system are deleted from the supporting design basis calculation (Enclosure 4 to Reference 1).

3. What is the pathway "Iodine via ADV of Unaffected SG - Isolated after 30 Days" used to model, in relation to the design-basisaccident (DBA).

Response: In accordance with Reference 2, the primary-to-secondary ruptured tube leakage and Technical Specification (TS) leakage of 200 gpd (TS 3.4.13) is assumed to continue until SDC conditions are attained and releases from the SGs have been terminated. In addition Reference 2 notes that, "the TS leakage should be apportioned between affected and unaffected SGs in such a manner that the calculated dose is maximized." Therefore, since the primary-to-secondary flow from the RCS to the affected SG was maximized for the worst-case thermohydraulic conditions, all of the TS primary-to-secondary leakage is assumed to flow to the unaffected SG. This leakage path maximizes dose, since the affected SG is isolated after two hours, and after 10 minutes the flashing fraction for the unaffected SG exceeds that for the affected SG.

After the ADV of the unaffected SG is reopened at 5031 seconds post-accident, 200 gpd or 1.1586 Ibm/min of primary-to-secondary leakage occurs until SDC conditions are achieved. All of the noble gases associated with the primary-to-secondary leakage are assumed to be released to the atmosphere via either the ADVs or MSSVs. A flashing fraction of 7% was calculated which results in 7% of the iodines immediately released to the atmosphere via the ADVs or MSSVs. The remaining 93%

of the primary-to-secondary leakage is assumed to mix with the bulk water of the unaffected SG. In accordance with Reference 2, the radioactivity in the bulk water is assumed to become vapor at a rate that is a function of the steaming rate and the partition coefficient. A partition coefficient for iodine of 100 may be assumed per Reference 2. The unaffected SG steaming rate is 936.072 lbm/min.

3

ATTACHMENT (1)

REQUEST FOR ADDITIONAL INFORMATION -- IMPLEMENTATION OF ALTERNATIVE SOURCE TERM

4. Clearly identify the coolant mass flow rate for releasesfrom the tube break and ADV steamingfor both the faulted and intact SGs.

Response: The behavior of the primary and secondary systems during and after a double-ended tube break SGTR Event was modeled based on operator responses as defined by Emergency Operating Procedures. The results of the analysis were plots and tables from the initiation of a SGTR Event until the plant has been cooled to SDC entry conditions.

The masses used in Enclosure 4 of Reference 1 are as follows:

Reactor Coolant System mass is 391900 Ibm which is the minimum RCS inventory at 1318.8 sec at the start of the COOL execution.

Unaffected SG mass is 56420 Ibm which is the minimum unaffected SG inventory at 1318.8 sec at the end of the CESEC execution.

" Affected SG mass is 124644 Ibm which is the minimum affected SG inventory at 750 sec in the CESEC execution.

Minimum mass inventories of the primary and secondary systems were used to maximize isotopic transport and hence maximize doses. Employing a time-dependent or time-averaged mass inventory could reduce RCS radioisotope loss by 19%, affected SG radioisotope loss due to steaming by an additional 13%, and unaffected SG radioisotope loss due to steaming by an additional 71%. Therefore, the results listed in Reference 1 are conservative.

The leakage from the RCS to the unaffected SG is defined by the TS 3.4.13 primary-to-secondary limit of 100 gpd per SG. All of the TS primary-to-secondary leakage is applied to the unaffected SG, because this case maximizes the calculated dose. A 7% flashing fraction is applied. Thus, the leakage from RCS to unaffected SG can be calculated as:

Leak Flow from RCS to Unaffected SG Leakage Flashing Time Total Leakrate Nonflashing Flashing gpd Fraction Sec Ibm/min Ibm/min Ibm/min 200 0.07 0 0 0 0 200 0.07 115200 1.1586 1.0775 0.0811 200 0.07 2592000 1.1586 1.0775 0.0811 The leakage from RCS to affected SG is:

Leak Flow from RCS to Affected SG Leakage Flashing Time Total Leakrate Nonflashing Flashing Ibm Fraction sec Ibm/min Ibm/min Ibm/min 0 0 37303 0.135 415 5393.2 4665.1 728.1 48199 0.135 600 3533.8 3056.8 477.1 52142 0.135 685 2783.3 2407.5 375.7 85747 0.045 1318.8 3181.3 3038.1 143.2 312900 0.045 6491 2635.1 2516.5 118.6 312900 0.055 86400 0.0 0.0 0.0 424116 0.055 115200 231.7 219.0 12.7 424116 0.055 2592000 0.0 0.0 0.0 4

ATTACHMENT (1)

REQUEST FOR ADDITIONAL INFORMATION -- IMPLEMENTATION OF ALTERNATIVE SOURCE TERM Note that Enclosure 4 of Reference 1 stated that the flashing fractions were 0.135 from 0-600 sec, 0.045 from 600-1318.6 sec, and 0.055 from 1318.6 sec to end of accident. The flashing fractions used are those listed in the table above. However, the graphs of the CESEC and COOL flashing fractions are attached (Figures 4A and 4B) and yield flashing fractions of 0.135 from 0-600 sec, 0.045 from 600-1318.6 sec, and much less than 0.045 on average for greater than 1318.6 sec. The values used in the above table and in the transport calculations are conservative compared to the CESEC and COOL values.

The steaming from the unaffected SG to the environment is shown below. The partition factor (PF) adjusted steaming rate is the steaming rate adjusted for the PF of 100 per Reference 2.

Steam Release from Unaffected SG Steaming Rate Time Steaming PF Adjusted sec Ibm/min Ibm/min 0.000 0 0.0 4921.2 0 0.0 115200 936.072 9.361 2592000 936.072 9.361 The steaming from the affected SG to the environment is shown below. The PF adjusted steaming rate is the steaming rate adjusted for the PF of 100 per Reference 2.

Steam Release from Affected SG Affected Affected SG Unaffected Steaming Rate Time SG ADV MSSV SG MSSV Total Steaming Steaming Rate PF Adjusted sec Ibm Ibm Ibm Ibm Ibm/min Ibm/min 0 0 0 0 0 0.00 0.0 415 0 0 0 0 0.00 0.0 600 16127 35960 35464 87551 28394.92 283.9 685 23718 37373 38814 99905 8720.47 87.2 1318.8 76556 37373 38814 152743 5002.02 50.0 4921 163866 37373 38814 240053 1454.28 14.5 86400 163896 37373 38814 240083 0.02 0.0 115200 508156 37373 38814 522553 588.48 5.9 5

ATTACHMENT (1)

REQUEST FOR ADDITIONAL INFORMATION -- IMPLEMENTATION OF ALTERNATIVE SOURCE TERM

.I.t

-.. 1 ...

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-J C:

(0 0

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.6E-O'1 J.r L .I , 1"

ýt .Iýt ..i ... ..: I ... 1. 1. i :

-4(-Oi' I

II Ii i 0" 2:ý00 .400 Time.(s) 1200:7 .I-Figure 4A - Flashing fractions versus time via CESEC-II1 6

ATTACHMENT (1)

REQUEST FOR ADDITIONAL INFORMATION -- IMPLEMENTATION OF ALTERNATIVE SOURCE TERM

.6E-O1 -

SE-al -

.4E-Ol C,)

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,.Affected' I~~~~A I.l~AhIAL(fA:.I.-

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.0. 50000 100000 150000 400000 .250000 Figure 4B - Flashing fractions versus time via COOL-I1 7

ATTACHMENT (1)

REQUEST FOR ADDITIONAL INFORMATION -- IMPLEMENTATION OF ALTERNATIVE SOURCE TERM

5. On page 20 of Enclosure 4 of the November 3, 2005, application,it is indicatedthat steaming isfrom the main steam safety valves (MSSVs). However, elsewhere in the analysis the steaming release is describedasflowingfrom the ADVs. Clarify the reasonfor this apparentdifference/discrepancy. If the release points are in fact being correctly described, verify and explain how the correct control room atmosphericdispersionfactors are being used Response: While the ADVs are the primary steam release pathways for the SGs, if the SG pressures exceed the MSSV lift setpoints as delineated in TS 3.7.1 "MSSVs," the MSSVs will lift, and steam will be released through the MSSVs into the atmosphere. While the ADV is the preferred steam release pathway, this can not always be guaranteed. However, it can be shown that the ADV-to-West Road and ADV-to-Turbine Building atmospheric dispersion coefficients generated in Enclosure 9 to Reference 1 are bounding for all ADVs and MSSVs. A sensitivity study was performed and is contained in Appendix 1.

Seized Rotor Event (SRE)

6. In the SRE analysis, it appears that the fraction of moisture carryoverwas applied only to alkaline metals releasedfrom the failed fuel. If this is indeed the case, provide the justificationfor not applying thisfraction to other particulatenuclides and iodine, in particular.

Response: This analysis is based on UFSAR Section 14.16 ("Seized Rotor Event"). The analysis of record reflected in UFSAR Section 14.16 calculates that 1.02% of fuel pins experience departure from nucleate boiling (DNB) for less than 3.5 seconds based on probabilistic convolution methodology; however, 5% of fuel pins were assumed to experience DNB for conservatism. Note that all fuel pins experiencing DNB are assumed to fail due to cladding overheating, thus releasing all of the radionuclides in the fuel pin gas gaps. However, a SRE does not cause an approach to the centerline temperature melt specified acceptable fuel design limits. Therefore, no fuel pins are assumed to reach fuel melt initiation temperature as a result of a SRE. Only the gas gap radionuclide released due to pin failure are modeled.

Per Reference 2, Regulatory Position 3.2, the non-loss-of-coolant accident (LOCA) gas gap contains iodines, noble gases, and alkali metals. These are modeled in Enclosure 5 of Reference 1.

" The alkali metals are modeled as particulates with a moisture carryover of 0.001.

All noble gas radionuclides released from the primary system are assumed to be released to the environment without reduction or mitigation.

" The TS primary iodines and TS secondary iodines are all modeled as 97% elemental and 3% organic.

" The gas gap iodines are all also modeled as 97% elemental and 3% organic. This is more conservative than modeling the gas gap iodine as 95% particulate, 4.85% elemental, and 0.15% organic. Modeling the iodine as a mixture of elemental and organic chemical forms maximizes iodine release from the SGs, since all of the iodine entrained due to flashing and steaming are assumed lost from the SGs, rather than the 0.001 fraction included in the moisture carryover. In addition, since the control room filter efficiency for particulate iodine exceeds that for elemental and organic iodine, assuming all iodine is either elemental or organic is an additional conservatism.

Thus, application of moisture carryover to other particulate nuclides or iodines is unnecessary, since all releases are conservatively accounted for.

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ATTACHMENT (1)

REQUEST FOR ADDITIONAL INFORMATION -- IMPLEMENTATION OF ALTERNATIVE SOURCE TERM

7. As stated in the submitted SRE Design Analysis CA06451, Revision 0, as a design basis, it is assumed that an 8-hour period is required to establish shutdown cooling (SDC) from a hot full power (HFP)condition, which is consistent with "actualplant operation." After the 8-hour period wherein SDC is established, the postulatedRCS activity release to the environment ceases. Because of this dynamic activity release, it cannot be assumed that the activity concentration in the environment, resultingfrom this release, is at its highest from the onset of the accident-initiated leakage,; moreover, and more specifically, the determination of a worst 2-hour Exclusion Area Boundary (EAB) dose must account for the actual activity release profile. It appears that the licensee attempted to compensatefor this effect by postulating an activity release case that assumes SDC is restored in 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , as opposed to 8. However, this case misses the peak environmental activity concentration associatedwith the design basis 8-hour release period, which may in fact be even higher than that associatedwith the hypothetical 2-hour releaseperiod case. By assigning the EAB dispersioncoefficient from time equal to 0 through the end of the accident, in the 8-hourperiod case, the code would be allowed to identify the peak 2-hour dose period associatedwith the DBA releasescenario.

Verify that the methodology used to determine the worst-case 2-hour EAB dose for the design-basis SRE scenario, as described in the application,was both conservative and accurate.

Response: The methodology used to determine the worst-case 2-hour EAB dose for the design basis SRE scenario, as described in the application, was not the most conservative. The RADTRAD SRE 8-hour cases were rerun to calculate the maximum 2-hour EAB dose by assigning the EAB dispersion coefficient from beginning of accident to end of accident. The results are as follows:

SRE EAB Results 0-2 hr 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months SRE 0-2 hr 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months SRE 6-8 hr 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months SRE EAB Rem EAB Rem EAB Rem Failed Fuel Iodine 0.025081 0.020530 0.035600 Failed Fuel Noble Gas 0.012240 0.012240 0.002579 Failed Fuel Alkali Metals 0.000627 0.000332 0.002101 Primary TS Activity Iodine 0.000002 0.000002 0.000003 Primary TS Activity NG 0.000037 0.000037 0.000025 Secondary TS Activity I 0.000580 0.000367 0.000260 Total 0.038567 0.033507 0.040569 The worst-case 2-hour EAB dose occurred during the last two hours of the accident and equaled 0.041 Rem TEDE. Note that this is still substantially less than the regulatory limit of 2.5 Rem TEDE per Reference 2. We are revising Section 4.1.5 of Attachment 1 (Reference 1) as follows:

4.1.5 Seized Rotor Event (SRE)

Section 14.16 of the CCNPP UFSAR describes the design basis evaluation of the SRE. A SRE is defined as a complete seizure of a single RCP shaft. The seizure is postulated to occur due to a mechanical failure or a loss of component cooling to the pump shaft seals. The most limiting SRE is an instantaneous RCP shaft seizure at HFP. The reactor coolant flow through the core would be asymmetrically reduced to three pump flow as the result of a shaft seizure on one pump. With the reduction of core flow due to the loss of an RCP, the core coolant temperatures 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 (CEAs) due to a low RCS 9

ATTACHMENT (1)

REQUEST FOR ADDITIONAL INFORMATION -- IMPLEMENTATION OF ALTERNATIVE SOURCE TERM flow 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 are thus predicted to fail. The initial secondary activity together with initial primary activity and failed fuel activity released to the primary system that then leaks into the secondary system will escape out of the SGs via the atmospheric dump valves. Note that per the requirements of Regulatory Guide 1.183 (Reference 3), the release of fission products from the secondary system should be evaluated with the assumption of a coincident LOOP. Thus, the use of condensers can not be credited in this analysis.

Enclosure (5) contains the detailed SRE radiological consequences design basis calculation using the AST. The following supporting calculations are also provided: Atmospheric Dispersion Coefficient (X/Q) calculation (Enclosure 9), Source Terms calculation (Enclosure 10), Gas Gap Isotopic Fraction calculation (Enclosure 11), and Primary and Secondary Isotopic calculations (Enclosure 12).

Major assumptions and required plant modifications considered in the SRE re-analysis include:

A bounding Control Room in-leakage value of 3,500 cfm,

Modification of the CREVS to a nominal 10,000 cfm flow with 90 percent filtration efficiency for elemental and organic iodine and 99 percent for particulate iodine was credited,

Installation of automatic isolation dampers and radiation monitors at Access Control HVAC Unit RTU-1 and Access Control Air Conditioning Unit 13 on the Auxiliary Building roof,

Revision to the TS 3.4.15 limit for RCS activity from 1.0 gCi/gm to 0.5 gCi/gm.

One Twe-SRE models was wefe-constructed: a tw. heur SG rcel*as mdodl t. m..aximize the EAB .

and-an eight-hour SG release model. to maximize the Control Room dosec. The Eaeh-SRE model is composed of six release components: gas gap iodine releases, gas gap noble gas releases, gas gap alkali metal releases, TS primary iodine activity releases, TS primary noble gas activity releases, and TS secondary iodine activity releases. The SRE is assumed to occur at time t=0 releasing the failed fuel gas gap iodine, noble gas, and alkali metal activities immediately and homogeneously into the primary system.

Table 5 summarizes the results of the SRE re-analysis for the EAB, LPZ, and Control Room doses. The corresponding regulatory dose limits are also shown in the last row of the table. The calculated doses for bet-the eight-hour and ve.heu....cases are less than the regulatory limit.

Table 5 SRE Results Cases EAB (Rem) LPZ (Rem) Control Room (Rem) 8 Hour Secondary Pathway 0.04 10,0336 0.0095 0.7885 2 Hour S..ndar.y P .Ahway 040386 0.094 02294 Regulatory Limits 2.5 (RG 1.183) 2.5 (RG 1.183) 5 (10 CFR 50.67)

The above section replaces Section 4.1.5 in Attachment I of Reference 1.

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ATTACHMENT (1)

REQUEST FOR ADDITIONAL INFORMATION -- IMPLEMENTATION OF ALTERNATIVE SOURCE TERM

8. Describe (preferablyby showing equations) the derivation of the individual releasefractions input to the RFT files that are shown in Attachments H through M of Design Analysis No. CA06451, Revision 0.

Response: Six release fraction files are utilized in a typical SRE calculation: failed fuel iodine, noble gas, and alkali metal radionuclides released from the gas gaps; primary TS iodine and noble gas radionuclides; and secondary TS iodine radionuclides. The equations for all six release fraction files are shown below.

(A) Failed fuel iodine and noble gas radionuclides released from the gas gaps:

The initial failed fuel isotopic activity Aj0 in Ci/MWt contained in the gas gaps of all of the assemblies in the core for isotope 'i' is calculated in Attachment A Column E of Enclosure 5 of Reference 1 and is based on the following algorithm:

" Ai0 = ASTi

RFj

ASTi = Isotopic activity per unit power (Ci/MWT) (Attachment A Column C)

" RFi = Isotopic gas gap release fraction (Attachment A Column D)

SREFFI = Failed fuel fraction = 0.05

" SREFFN = Failed fuel fraction = 0.05 The iodine and noble gas isotopic activities Ai0 were inserted into the nuclear inventory file GAP14.NIF for use by RADTRAD. The file is listed in Attachment D of Enclosure 5 of Reference 1 and consists of the 14 gas-gap noble gas and iodine isotopes. The corresponding iodine and noble gas release fractions in SREFFI.RFT (Attachment H) and SREFFN.RFT (Attachment I) are 0.05, which corresponds to the 5%

failed fuel. The release is instantaneous: 0.0001 sec.

(B) Failed fuel alkali metal radionuclides released from the gas gaps:

The alkali metal isotopic activities A10 are contained in the base deck CRCB63.NIF, which is listed in Attachment G of Enclosure 5 of Reference 1. The corresponding alkali metal release fraction in SREFFA.RFT (Attachment M) is 0.05*0.24 = 0.012, where 0.05 is the failed fuel fraction and 0.24 is twice the non-LOCA alkali metal release fraction of Table 3 of Reference 2. The factor of two is from 1 of Reference 1, where all gas gap release fractions were conservatively doubled. All releases are instantaneous: 0.0001 sec.

" Ai0 = ASTi

" ASTi = Isotopic activity per unit power (Ci/MWT) (Attachment A Column C)

" SREFFA = Failed fuel fraction

RFi = 0.05

0.24 = 0.0 12

" RFj = Isotopic gas gap release fraction (Attachment A Column D)

(C) Primary TS iodine and noble gas radionuclides:

The initial primary specific activities in p[Ci/gm consistent with the TS 3.4.15 1.0 gCi/gm limit were extracted from Enclosure 12 of Reference I and are listed in Column F of Attachment A of Enclosure 5 of Reference 1. These were converted to total primary isotopic source terms in Ci in Attachment A column G via the following algorithm:

" Ai0 = AGMi

MRcs

0.000001

" AGMi = Isotopic activity per unit mass (gCi/gm) (Attachment A Column F)

MRCS = Water mass in RCS (gm) 11

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REQUEST FOR ADDITIONAL INFORMATION -- IMPLEMENTATION OF ALTERNATIVE SOURCE TERM

" SREPRI = 1

SREPRN =1 The primary iodine isotopic source terms were then halved in Attachment A Column H to reflect that the TS 3.4.15 limit for primary activity will be reduced from 1.0 gtCi/gm to 0.5 tCi/gm. These isotopic activities were inserted into the nuclear inventory file PRI14.NIF for use by RADTRAD. The file is listed in Attachment E of Enclosure 5 of Reference I and consists of the 14 primary noble gas and iodine isotopes. The activities are the total primary activities and are not per unit power. Thus a power of one should be designated when employing these files. The corresponding iodine and noble gas release fractions in SREPRI.RFT (Attachment J) and SREPRN.RFT (Attachment K) are unity, since the radionuclides are homogeneously and instantaneously distributed throughout the primary system at the beginning of the event.

(D) Secondary TS iodine radionuclides:

The initial secondary specific activities in jICi/gm consistent with the TS 3.7.14 0.1 [tCi/gm limit were extracted from Enclosure 12 of Reference I and are listed in column I of Attachment A of Enclosure 5 of Reference 1. These were converted to total primary isotopic source terms in Ci in Attachment A Column J via the following algorithm:

Ai0 =ASECi

MSG

0.000001

" ASECi = Isotopic activity per unit mass (ptCi/gm)

MSG = Water mass in SG (gm)

SRESEC = 1 These isotopic activities were inserted into the nuclear inventory file SEC05.NIF for use by RADTRAD.

The file is listed in Attachment F of Enclosure 5 of Reference 1 and consists of the 5 iodine isotopes. The activities are the total secondary activities and are not per unit power. Thus a power of one should be designated when employing these files. The corresponding iodine release fraction in SRESEC.RFT is unity, since the radionuclides are homogeneously and instantaneously distributed throughout the secondary at the beginning of the event.

9. Regulatory Guide (RG) 1.183, Appendix H, can be interpretedas assumingthat allfuel reachingfuel melt initiation temperaturecompletely melts and is releasedin accordance with the specified release fractions. Verify that the licensee's assumption of 5% fuel failure for the SRE conservatively addressesthis interpretation.

Response: The failed fuel factor of 5% is the same as that assumed in UFSAR Section 14.16 ("Seized Rotor Event"). Note that the analysis of record reflected in UFSAR Section 14.16 calculates that 1.02%

of fuel pins experience DNB for less than 3.5 seconds based on probabilistic convolution methodology; however, 5% of fuel pins were assumed to experience DNB for conservatism. Note that all fuel pins experiencing DNB are assumed to fail due to cladding overheating, thus releasing all of the radionuclides in the fuel pin gas gaps. However, a SRE does not cause an approach to the centerline temperature melt specified acceptable fuel design limits. Therefore, no fuel pins are assumed to reach fuel melt initiation temperature as a result of a SRE.

Main Steamline Break (MSLB) Accident

10. Provide additionaljustificationfor the postulation that the main steam piping room is the worst-case, credible, MSLB accident activity releasepoint. In this justification, include a discussion about 12

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REQUEST FOR ADDITIONAL INFORMATION -- IMPLEMENTATION OF ALTERNATIVE SOURCE TERM the validity of assuming a vent release, as opposed to leakage (causedby differentialpressure),from this main steam piping room.

Response: The main steam lines originate in containment, pass through the containment wall into the main steam piping penetration room (Room A315 for Unit 1 and Room A309 for Unit 2), and then pass into the Turbine Building. The main steam isolation valves (MSIVs) are in the main steam piping penetration room. Since the safety function of these valves is to shut, a MSLB in the Containment will be contained in the Containment Building and will have minimal offsite and control room dose effects. A MSLB in the Turbine Building will be rapidly isolated by the MSIV and again have minimal offsite and control room dose effects. The MSIV closes within a maximum of six seconds after a trip signal is initiated, preventing loss of SG inventory. A MSLB in the main steam piping penetration room upstream of the MS1Vs will discharge into the penetration room. Therefore, that room is the release point for a MSLB.

After the MSLB, the affected SG discharges into the main steam piping penetration room. This room has a 7.5' diameter hole in the ceiling, which discharges to the Auxiliary Building roof through an eight foot square J-neck galvanized steel duct (BGE Drawing 63770SH0001 Rev. 6). The initial SG blowdown for a guillotine break occurs very rapidly, results in an initial large pressure rise in the main steam piping penetration room, but accounts for less than 5% of the control room dose. Additionally, any breaching of the main steam piping penetration room walls would result in a chimney effect, enhancing flow out of the main steam gooseneck. Based on computational models of the main steam piping penetration room pressure and temperature profiles, the pressure spike dissipates rapidly through the 7.5' diameter main steam gooseneck duct. The remainder of the control room dose is released over the length of the accident via primary to secondary leakage into the affected SG. The main steam piping penetration room is not characterized by a positive pressure differential due to the large 7.5' diameter main steam gooseneck duct.

The high temperature of the primary to secondary leakage, which causes the entire leakage to flash, would result in an upward buoyant force through the main steam gooseneck duct.

11. Provide the basisfor the reactor cooldown time assumedfor this accident analysis.

Response: The time to cool the RCS to 212'F during a MSLB is required as input to the 30-day control room dose calculation. The primary pathway for release of activity during a MSLB is via primary-to-secondary leakage to the affected SG. This leakage will flash when the RCS temperature is above 212'F and transport the RCS activity through the break into the atmosphere. Therefore, when the RCS temperature is reduced below 212'F, the release of activity is stopped. For conservatism, a maximum cooldown time was calculated.

To determine the time required to cool the RCS to 212'F during a MSLB, the assumptions with regard to what equipment is available to the Operators must first be established. Since the Calvert Cliffs licensing basis for MSLB includes consideration of a loss-of-offsite power (LOOP), equipment that is not diesel-backed is assumed unavailable. In addition, the ADVs are also assumed unavailable. Therefore, the plant must be cooled using once through core cooling (OTCC). Consistent with the worst single failure for the MSLB analysis, failure of a diesel generator, only two high pressure safety injection (HPSI) pumps are assumed available, and only after time is allotted for Operators to cross-tie the appropriate electrical buses. The cross-tie of electrical buses is needed to open both power-operated relief valves (PORVs).

Calvert Cliffs Emergency Operating Procedures provide adequate guidance for Operators to conduct these actions in a timely manner.

13

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REQUEST FOR ADDITIONAL INFORMATION -- IMPLEMENTATION OF ALTERNATIVE SOURCE TERM Given the above assumptions, and based on discussions with Operations personnel, no more than two hours is required after the MSLB in order to cross-tie the necessary electrical buses and begin OTCC using two HPSIs and two PORVs. During this initial two hour period, the plant response to the MSLB will be to rapidly cool down until the break is isolated or the affected SG blows dry. With the ADVs and the main condenser (due to the LOOP) unavailable, the Operators will be unable to prevent the plant from re-heating to approximately 550'F at which point the MSSVs for the unaffected SG will open to stabilize plant temperature (Note that all of the primary-to-secondary leakage is in the affected SG, so this has a negligible contribution to dose). Therefore, after two hours the plant cooldown using OTCC will begin at an RCS temperature of 550'F.

Evaluation of OTCC with two HPSI pumps and two PORVs available shows that RCS hot leg temperature decreases to 300'F in approximately five hours after the event begins (5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months , 8 min).

Therefore, for an MSLB, it is assumed that Operators would establish SDC, continue the cooldown, and cool the RCS to 300'F within 5.2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months after the OTCC begins.

While on SDC, the RCS cooldown rate would be 100lF/Hr from 300'F to 256°F and 40°F/Hr from 256°F to 212'F per TS 3.4.3 requirements. As a result, approximately 1.6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months will be required to reduce plant temperature from 300'F to below 212'F. Therefore, from the time of the MSLB, no more than a total of nine hours will be required to cool the RCS to 212'F. At this point, i.e., nine hours into the event, the steam and associated activity releases from the affected SG will stop.

Note that the above determination of cooldown post-MSLB assumes no credit for ADVs in the cooldown other than their use during OTCC. The ADV from the unaffected SG will be available for use, and thus the 9 hour1.041667e-4 days 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months cooldown assuming only OTCC and SDC is conservative.

12. For the MSLB accident-initiated,concurrent, iodine release rate spike, RG 1.183, Appendix E, recommends an assumed spike duration of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . Explain the reason why a 9-hour accident-initiated, concurrent, iodine release rate spike duration was assumed in the licensee's MSLB accident analysis.

Response: Since a cooldown time of 9 hours1.041667e-4 days 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months was assumed in the MSLB accident analysis, the spike duration for the concurrent iodine release was conservatively increased to 9 hours1.041667e-4 days 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months . Since the concurrent iodine spike Nuclear Inventory File iodine isotopics generated in Enclosure 12 to Reference 1 assumed an 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months spike duration, the iodine activities were increased by a factor of 9/8 to account for the longer accident as detailed in Enclosure 3 to Reference 1, Section 9.2.e. Thus, the assumption is conservative.

Also note that the concurrent iodine spike MSLB analysis is bounded by the failed fuel MSLB analysis and was included in the license submittal only for completeness. It will not be considered as part of the design basis MSLB accident.

Fuel-Handling Accident (FHA)

13. In paragraph 4 on page 8 of Design Analysis No. CA06450, Revision 0, it is indicated that a decontaminationfactor (DF)of 200 is usedfor FHA analysis cases that assume a drop over seated assemblies in the spent fuel pool (SFP), despite a Technical Specification (TS) requiring only 21.5 feet of water coverage over seated assemblies. The rationalegiven for using a DF of 200 relies on an assumption that an assembly drop over the SFP will strike the bottom of the pool, as opposed to the top of the racks, thereby maintainingmore than the 23 feet of water coverage that is typically requiredto assume a DF of 200. The result of this analysis ultimately shows that the FHA over the 14

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REQUEST FOR ADDITIONAL INFORMATION -- IMPLEMENTATION OF ALTERNATIVE SOURCE TERM SFP during reconstitution/inspection,and not over seated assemblies, is the limiting FHA case. As explained, this is because a DF of 120, associatedwith a conservatively assumed value of 20.5 feet of water coverage, is used for analyzing an FHA over the SFP during reconstitution/inspection.

However, there is no verification of or justificationfor, the assumption that a dropped assembly must strike the bottom of the SFP before resulting in an FHA for the case assuming only seated assemblies. Therefore if, in the design basis F-L4 analysis, the licensee intends to continue to include the assessments that assume a DF of 200 for an F-L4 resultingfrom a drop over seated assemblies in the SFP,provide verification that the dropped assembly will strike the bottom of the SFP and indeed be covered by at least 23 feet of water.

Response: The current Calvert Cliffs design basis FHA is presented in UFSAR Section 14.18. An FHA is assumed to occur in the SFP handling area by dropping a fuel assembly during fuel movement operations. The analyses for a F1-A in the SFP assumes that gas gap activity from all fuel rods (176) of the highest power assembly is released. As noted in the question above, a fuel assembly striking the SFP floor is no longer the worst case FHA. The worst case FHA involves a fuel assembly seated in the fuel reconstitution stand.

In the SFP the fuel assemblies are stored within the racks in the SFP. The top of the rack extends above the tops of the stored fuel assemblies. There are five possible scenarios for a dropped fuel assembly:

" A fuel assembly dropped on the end of an assembly seated in the storage racks -- A dropped fuel assembly could not strike more than one fuel assembly in the storage racks. Impact could occur only between the ends of the involved fuel assemblies, the bottom end fitting of the dropped fuel assembly impacting against the top end fitting of the stored fuel assembly. The results of an analysis of the end on energy absorption capability of a fuel assembly indicate that a fuel assembly is capable of absorbing the kinetic energy of the drop with no fuel rod failures.

" A fuel assembly dropped horizontally across the top of the storage racks -- Dropping an assembly horizontally on top of the SFP racks from the spent fuel handling machine is not possible at Calvert Cliffs due to the design of the spent fuel handling machine and due to the height of the SFP racks.

The bottom of the outer mast assembly is at elevation 49'5", while the top of the SFP racks is at elevation 45'0". Since the height of a fuel assembly is 13'1" per UFSAR Figure 3.3-5, it is not possible to drop an assembly horizontally onto the SFP racks from the spent fuel handling machine.

A fuel assembly dropped by the cask handling crane -- Dropping an assembly on top of the SFP racks from the cask handling crane is also not a credible accident. The cask handling crane is designed in accordance with the single-failure proof criteria of NUREG-0554 and NUREG-0612 and is used to move assemblies into the new fuel elevator.

" A fuel assembly dropped from the new fuel elevator -- Dropping an assembly on top of the SFP racks from the new fuel elevator is also not a credible accident. The new fuel elevator is utilized to lower new fuel from the operating floor to the bottom of the SFP, where it is then grappled by the spent fuel handling machine. The elevator is powered by a cable winch, and the assembly is contained in a simple support structure whose wheels are captured on two rails. Dropping an

.assembly from the new fuel elevator would require a catastrophic failure of the new fuel elevator, which is not considered a credible event.

" A fuel assembly dropped from the spent fuel handling machine to the SFP floor -- A FHA could occur in the SFP by dropping a fuel assembly to the SFP floor, where it falls and strikes the fuel pool floor vertically. The only credible location for such a drop is the SFP upender trench between the SFP racks and the SFP wall. A fuel assembly falling into this area would be covered by 23' of 15

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REQUEST FOR ADDITIONAL INFORMATION -- IMPLEMENTATION OF ALTERNATIVE SOURCE TERM water because the SFP floor in the upender trench is 2' lower than the bottom of the SFP rack.

Technical Specification 3.7.13 requires that the SFP water level be maintained greater than or equal to 21.5' over the top of fuel seated in the storage racks. Therefore, a DF of 200 is appropriate for an assembly dropped to the SFP floor, because there would be at least 23' of water over the top of the dropped assembly.

Control Element Assembly Ejection (CEAE) Accident

14. In the CEAE Design Analysis CA06454, Revision 0, as a design basis, it is assumed that an 8-hour period is required to establish shutdown cooling (SDC) from a hot full power (HFP)condition, which is consistent with "actual plant operation." After the 8-hour period where SDC is established,the postulated reactor coolant system (RCS) activity release to the environment ceases.

Because of this dynamic activity release, it cannot be assumed that the activity concentration in the environment, resulting from this release, is at its highest from the onset of the accident-initiated leakage. moreover, and more specifically, the determination of a worst 2-hour EAB dose must accountfor the actual activity release profile. It appears that the licensee attempted to compensate for this effect by postulating an activity release case that assumes SDC is restored in 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , as opposed to 8. However, this case misses the peak environmental activity concentrationassociated with the design basis 8-hour release period, which may in fact be even higher than that associated with the hypothetical 2-hour releaseperiod case. By assigningthe EAB dispersion coefficient from time equal to 0 through the end of the accident, in the 8-hourperiod case, the code would be allowed to identify the peak 2-hour dose periodassociatedwith the DBA release scenario.

Provide verification that the methodology used to determine the worst-case 2-hour EAB dose for the design basis CEAE accident scenario, as described in the submittal was both conservative and accurate.

Response: The methodology used to determine the worst-case 2-hour EAB dose for the design basis CEAE accident, as described in the application, was conservative; however, the methodology did not address NRC concerns as reflected in the question above. The RADTRAD CEAE accident 8-hour cases were rerun to calculate the maximum 2-hour EAB dose by assigning the EAB dispersion coefficient from beginning-of-accident to end-of-accident. In addition, the EAB dose due to particulates (Alkali Metals, Tellurium Metals, Ba, Sr, Noble Metals, Cesiums, and Lanthanides) were also included. The results are as follows:

CEAE EAB Results Reference 1 NRC Requested Analysis 0-2 hr 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months CEAE 0-2 hr 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months CEAE 6-8 hr 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months CEAE EAB Rem EAB Rem EAB Rem Failed Fuel Iodine 0.155210 0.125030 0.204530 Failed Fuel Noble Gas 0.200730 0.200730 0.042293 Failed Fuel Particulates 0.001816 0.001816 0.011395 Primary TS Activity Iodine 0.000002 0.000002 0.000003 Primary TS Activity NG 0.000037 0.000037 0.000025 Secondary TS Activity I 0.000600 0.000362 0.000255 Total 0.358395 0.327976 0.258501 While the 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months CEAE accident EAB dose was conservative, it is replaced by the worst-case 2-hour dose calculated above. The worst-case 2-hour EAB dose occurred during the first two hours of the accident and equaled 0.328 Rem TEDE. Also note that while the worst-case EAB dose for failed fuel iodine, 16

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REQUEST FOR ADDITIONAL INFORMATION -- IMPLEMENTATION OF ALTERNATIVE SOURCE TERM

failed fuel particulates, and primary TS iodine occurred between 6-8 hours post-CEAE, the worst case EAB dose for failed fuel noble gas, primary TS noble gas, and secondary TS iodine occurred between 0-2 hours post-CEAE.

Section 4.1.6, Attachment I of Reference 1 is modified as follows:

4.1.6 Control Element Assembly Ejection Accident (CEAEA)

Section 14.13 of the CCNPP UFSAR describes the design basis evaluation of the CEAEA. A CEAEA is defined as a rapid, uncontrolled, total withdrawal of a single or dual CEA, where a dual CEA is two CEAs connected to a single CEA extension shaft. The event is postulated to occur as a result of a complete, instantaneous, circumferential rupture of either the control element drive mechanism pressure housing or the control element drive mechanism nozzle from the reactor vessel closure head. The pressure of the RCS causes the ejection of the extension shaft through the rupture and the movement of the CEA to a fully-withdrawn position. The most limiting CEAEA is a rapid total withdrawal of the highest worth CEA within 0.05 seconds and the breaching of the RCS pressure boundary. The immediate reactor core response is an exponential increase in nuclear power.

Enclosure (6) contains the detailed CEAEA radiological consequences design basis calculation using the AST. The following supporting calculations are also provided: Atmospheric Dispersion Coefficient (X/Q) calculation (Enclosure 9), Source Terms calculation (Enclosure 10), and Primary and Secondary Isotopic calculations (Enclosure 12).

Major assumptions and required plant modifications considered in the CEAEA re-analysis to meet regulatory requirements are:

A bounding Control Room in-leakage value of 3,500 cfm,

Modification of the CREVS to a nominal 10,000 cfm flow with 90 percent filtration efficiency for elemental and organic iodine and 99 percent for particulate iodine was credited,

" Installation of automatic isolation dampers and radiation monitors at Access Control HVAC Unit RTU-1 and Access Control Air Conditioning Unit 13 on the Auxiliary Building roof,

Revision to the TS 3.4.15 limit for RCS activity from 1.0 gCi/gm to 0.5 pCi/gm,

Revision to the TS 5.5.16 maximum allowable containment leakage rate, La, from 0.20 percent of containment air weight per day at Pa to 0.16 percent of containment air volume per day at Pa.

The CEAEA was analyzed for a 30-day containment pathway, a 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to shutdwn . eeling

. .""ndary pathway, and an 8-hour to shutdown cooling secondary pathway. Table 6 summarizes the results of the CEAEA re-analysis for the EAB, LPZ, and Control Room doses. The corresponding regulatory dose limits are shown in the last row of the table. All calculated doses are less than the regulatory limit.

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REQUEST FOR ADDITIONAL INFORMATION -- IMPLEMENTATION OF ALTERNATIVE SOURCE TERM Table 6 CEAEA Results Control Room (Rem)

Results EAB (Rem) LPZ (Rem) 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Secondary Pathway 0.327980.32616 0.0881480.0873-5 4.76274z59464-2 h..ur Se..ndary Pathway 0-4356-58 0.08394 30 day Containment Pathway 0.85130.456-7 0.21900.1187 2.02810.96-79 Regulatory Limits 6.3 6.3 5 The above section replaces Section 4.1.6 in Attachment I of Reference 1.

15. It appears that the CEAE accident analysis only accounts for releases of activity associated with iodine and noble gas isotopes. However, in accident scenarioswhere fuel failure is postulated,other particulatenuclides are availablefor releasefrom the gap offailedfuel, as well as from the melted fuel itself RG 1.183, Position 3, gives guidance to indicate that at least fractions of the "Alkali Metal" isotopic inventory should be accounted for when fuel failure is postulatedfor non-loss-of-coolant accident (LOCA) events. Table 3 of RG 1.183 specifies these fractions. This guidance appears[not] to have been followed in the SRE analysis. Therefore, explain why not accountingfor the release of these isotopes, in the case of the CEAE accident, is conservative and accurate. Also, indicate how RG 1.183, Footnote 11, is addressedin the analysis of this accident.

Response: Our analysis is based on Reference 2 Appendix H.1 which states:

For the rod ejection accident, the release from the breached fuel is based on the estimate of the number of fuel rods breached and the assumption that 10% of the core inventory of the noble gases and iodines is in the fuel gap. The release attributed to fuel melting is based on the fraction of the fuel clad that reaches or exceeds the initiation temperature for fuel melting and the assumption that 100% of the noble gases and 25% of the iodines contained in that fraction are available for release from containment. For the secondary system release pathway, 100% of the noble gases and 50% of the iodines contained in that fraction are released to the reactor coolant.

The accident-specific Reference 2 Appendix H gas gap fractions of 10% of the iodine and noble gas inventories are twice the Regulatory Position 3 fractions, except for 1-131 which is increased by 25% and Kr-85 which is not increased. Since Appendix H specifically prescribes a gas gap fraction different from that detailed in Regulatory Position 3, we believe that the Appendix H values supersede those in Regulatory Position 3. Noting that the gas gap fractions peak at end-of-life burnup while the pin power peaking factor of 1.7 applies at beginning-of-life, additional conservatism as suggested by Footnote 11 of Reference 2 seems excessive and contrary to Appendix.H guidance.

In accordance with Reference 2 Appendix H guidance, only iodines and noble gases are addressed as possible releases. However, to ensure that the calculation is conservative enough, the contribution of alkali metals, tellurium metals, bariums, strontiums, noble metals, ceriums, and lanthanides to the offsite and control room doses are evaluated and included in the results.

Two fission product release paths to the environment are considered independently.

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REQUEST FOR ADDITIONAL INFORMATION -- IMPLEMENTATION OF ALTERNATIVE SOURCE TERM

The failed/melted fuel activity, resulting from a postulated CEAE accident and consisting of 100%

of the noble gases, 25% of the iodines, 30% of the alkali metals, 5% of the tellurium metals, 2% of the bariums and strontiums, 0.25% of the noble metals, 0.05% of the cerium group, and 0.02% of the lanthanides contained in the fuel that is estimated to reach initiation of melting and 10% of the noble gases, 10% of the iodines, and 24% of the alkali metals that are contained in the gas gaps of the fuel that experience clad failure, is released into the primary system, which is released in its entirety into the containment via the ruptured control rod drive mechanism housing. The released activity is instantaneously and uniformly mixed in the free volume of the Containment and is then released at the containment TS leak rate into the environment.

" The failed/melted fuel activity, resulting from a postulated CEAE accident and consisting of 100%

of the noble gases, 50% of the iodines, 30% of the alkali metals, 5% of the tellurium metals, 2% of the bariums and strontiums, 0.25% of the noble metals, 0.05% of the cerium group, and 0.02% of the lanthanides contained in the fuel that is estimated to reach initiation of melting and 10% of the noble gases, 10% of the iodines, and 24% of the alkali metals that are contained in the gas gaps of the fuel that experience clad failure is released into the primary system, which is then transmitted into the secondary system via the TS SG tube leakage. The condenser is assumed to be unavailable due to LOOP. Environmental releases occur from both SGs via the ADVs and MSSVs.

The results for the CEAE accident secondary pathway are as follows:

CEAE Accident Secondary Pathway Results 0-2 hour 6-8 hour EAB Rem EAB Rem LPZ Rem CR Rem Failed Fuel Iodine 1.2503E-01 2.0453E-01 3.6871E-02 4.1672E+00 Failed Fuel Noble Gas 2.0073E-01 4.2293E-02 5.0368E-02 4.1885E-01 Failed Fuel Particulates 1.8156E-03 1.1395E-02 8.0176E-04 1.6811E-01 Primary TS Activity Iodine 1.5225E-06 2.7295E-06 4.5645E-07 5.5874E-05 Primary TS Activity NG 3.6795E-05 2.5381 E-05 9.9443E-06 1.5932E-04 Secondary TS Activity I 3.6208E-04 2.5464E-04 9.7236E-05 8.3421 E-03 Total 3.2798E-01 2.5850E-01 8.8148E-02 4.7627E+00 Regulatory Limits 6.3000 6.3000 6.3000 5.0000 The results for the CEAE accident containment pathway are as follows:

CEAE Accident Containment Pathway Results EAB Rem LPZ Rem CR Rem 30 day Containment Pathway 0.8513 0.2190 2.0281 Regulatory Limits 6.3000 6.3000 5.0000 The results in all cases are below the regulatory limits.

16. RG 1.183, Appendix H, can be interpretedas assuming that all fuel that reachesfuel melt initiation temperature, completely melts and is released in accordance with the specified release fractions.

Explain how the assumption of 8% melted fuel and 2% clad damage used for the CEAE accident addresses this interpretation.

Response: Failed fuel for the CEAE accident falls into three categories: clad failure, incipient centerline fuel melt, and full centerline fuel melt. For the Unit 1 Cycle 15 CEAE accident analysis, the fraction of 19

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REQUEST FOR ADDITIONAL INFORMATION -- IMPLEMENTATION OF ALTERNATIVE SOURCE TERM fuel experiencing incipient centerline melting was 8.0% for HFP and 0.824% for HZP. The Calvert Cliffs current design basis, UFSAR Section 14.13, shows that no fuel rod will experience clad damage and only a small fraction will reach incipient centerline melt condition (4.2% for a HFP CEAE and 1.8% for a HZP CEAE).

Based on the above, Calvert Cliffs assumes the worst-case historical fraction for incipient centerline melting of 8%. For this 8%, full melting of the fuel including clad failure is assumed releasing 100% of the noble gases and 25% of the iodines into the containment and 100% of the noble gases and 50% of the iodines into the primary coolant. In addition, based on Reference 3, Calvert Cliffs accepted a 10% fuel damage fraction for a CEAE and since 8% of the fuel is assumed to melt with clad failure, an additional 2% of the fuel is assumed to experience clad failure with no fuel melting, releasing the gas gap contents of 10% noble gases and 10% iodines. Note that since only the highest power fuel pins are assumed to be damaged, the releases are increased by the pin power peaking factor of 1.7. Compared to the current design basis as described in UFSAR Section 14.13, this assumption is conservative.

17. Clarify (preferablyby showing equations) the derivation of the individual releasefractions input to the RFTfiles, shown in Attachments H through M of Design Analysis CA 06454, Revision 0.

Response: Six release fraction files are utilized in the CEAE secondary pathway calculations: failed fuel iodine, noble gas, and particulate radionuclides released from the gas gaps and melted fuel; primary TS iodine and noble gas radionuclides; and secondary TS iodine radionuclides. A single release fraction file is utilized in the CEAE containment pathway calculation containing the failed fuel iodine, noble gas, and particulate radionuclides released from the gas gaps and melted fuel. The derivation of these release fractions are shown below for all six cases.

(A) Iodine released from the gas gaps and melted fuel for secondary pathway:

The isotopic activities A10 in CEA14.NIF contain the iodine, krypton, and xenon isotopics and are identical to those derived in Enclosure 10 to Reference 1 (CRCB63.NIF). The corresponding iodine release fraction in CEAFFI.RFT is comprised of 8% of the 50% melted iodine release and 2% of the 10%

gap iodine release. All releases are instantaneous: 0.0001 sec.

Ai 0 = ASTi o Ai0 =Isotopic activity per unit power (Ci/MWT) in CEAI4.NIF o ASTi = Isotopic activity per unit power (Ci/MWT) in CRCB63.NIF

" CEAFFI(Iodine) = Fm

Fmi + Fc

Fci = 0.042 o Fm = 0.08 = Melted fuel fraction o Fc = 0.02 = Clad damaged fuel fraction o Fmi = 0.50 = Melted fuel iodine release fraction for secondary pathway o Fci = 0.10 = Iodine release fraction from clad damaged fuel (B) Noble Gases released from the gas gaps and melted fuel for secondary pathway:

The isotopic activities A10 in CEA14.NIF contain the iodine, krypton, and xenon isotopics and are identical to those derived in Enclosure 10 to Reference 1 (CRCB63.NIF). The corresponding noble gas release fraction in CEAFFN.RFT is comprised of 8% of the 100% melted noble gas release and 2% of the 10% gap noble gas release. All releases are instantaneous: 0.0001 sec.

20

ATTACHMENT (1)

REQUEST FOR ADDITIONAL INFORMATION -- IMPLEMENTATION OF ALTERNATIVE SOURCE TERM

" Ai0 = ASTi o A10 =Isotopic activity per unit power (Ci/MWT) in CEA14.NIF o ASTi = Isotopic activity per unit power (Ci/MWT) in CRCB63.NIF

" CEAFFN(Noble Gas) = Fm

Fmn + Fc

Fcn = 0.082 o Fm = 0.08 = Melted fuel fraction o Fc = 0.02 = Clad damaged fuel fraction o Fmn = 1.00 = Melted fuel noble gas release fraction o Fcn = 0.10 = Noble gas release fraction from clad damaged fuel (C) Particulates released from the gas gaps and melted fuel for secondary pathway:

Per Reference 2, Regulatory Position 3 and Enclosure 11 to Reference 1, the particulate isotopic release fractions are defined as 30% of the alkali metals, 5% of the tellurium metals, 2% of the bariums and strontiums, 0.25% of the noble metals, 0.05% of the cerium group, and 0.02% of the lanthanides contained in the fuel that is estimated to reach initiation of melting and 24% of the alkali, metals that are contained in the gas gaps of the fuel that experience clad failure. The isotopics are identical to those derived in Enclosure 10 to Reference 1 (CRCB63.NIF). The corresponding particulate release fractions in CEAAMS.RFT are comprised of 8% of the melted fuel release fractions and 2% of the gap release fractions. All releases are instantaneous: 0.000 1 sec.

" Ai0 = ASTi o Ai0 = ASTI = Isotopic activity per unit power (Ci/MWT) in CRCB63.NIF

CEAAMS(Alkali Metal) = Fm

Fma + Fc

Fca = 0.0288

" CEAAMS(Tellurium) = Fm

Fmt = 4.OE-03

" CEAAMS(Strontium) = Fm

Fms = 1.6E-03

" CEAAMS(Barium) = Fm

Fmb = 1.6E-03

CEAAMS(Ruthenium) = Fm

Fmr = 2.OE-04

" CEAAMS(Cerium) = Fm

Fmc = 4.OE-05

" CEAAMS(Lanthanum) = Fm

Fml = 1.6E-05 o Fm 0.08 = Melted fuel fraction o Fc = 0.02 = Clad damaged fuel fraction o Fma = 0.30 = Melted fuel alkali metal release fraction o Fca 0.24 = Alkali Metal release fraction from clad damaged fuel o Fmt - 0.05 = Melted fuel tellurium release fraction o Fms = 0.02 = Melted fuel strontium release fraction o Fmb = 0.02 = Melted fuel barium release fraction o Fmr = 0.0025 = Melted fuel ruthenium release fraction o Fmc = 0.0005 = Melted fuel cerium release fraction

" Fml = 0.0002 = Melted fuel lanthanum release fraction 21

ATTACHMENT (1)

REQUEST FOR ADDITIONAL INFORMATION -- IMPLEMENTATION OF ALTERNATIVE SOURCE TERM (D) Primary TS iodine and noble gas releases via the secondary pathway:

The initial primary specific activities in gCi/gm consistent with the TS 3.4.15 1.0 RCi/gm limit were extracted from Enclosure 12 to Reference 1 and are listed in Column G of Attachment A of Enclosure 6 to Reference 1. These were converted to total primary isotopic source terms in Ci in Attachment A column H via the following algorithm:

Ai0 = AGMi

MRCS

0.000001 (Attachment A Column H)

AGMi = Isotopic activity per unit mass (gCi/gm) (Attachment A Column G)

MRCS = Water mass in RCS (gm)

" CEAPRI = 1

" CEAPRN = 1 The primary iodine isotopic source terms were then halved in Attachment A Column I to reflect that the TS 3.4.15 limit for primary activity will be reduced from 1.0 gCi/gm to 0.5 gCi/gm. These isotopic activities were inserted into the nuclear inventory file PRI14.NIF for use by RADTRAD. The file is listed in Attachment E of Enclosure 6 to Reference 1 and consists of the 14 primary noble gas and iodine isotopes. The activities are the total primary activities and are not per unit power. Thus a power of one should be designated when employing these files. The corresponding iodine and noble gas release fractions in CEAPRI.RFT (Attachment J) and CEAPRN.RFT (Attachment K) are unity, since the radionuclides are homogeneously and instantaneously distributed throughout the primary at the beginning of the event.

(E) Secondary TS iodine gas release via the secondary pathway:

The initial secondary specific activities in gCi/gm consistent with the TS 3.7.14 0.1 RCi/gm limit were extracted from Enclosure 12 to Reference I and are listed in Column J of Attachment A of Enclosure 6 to Reference 1. These were converted to total primary isotopic source terms in Ci in Attachment A column K via the following algorithm:

" Ai0 = ASECi

MsG

0.000001

" ASECi = Isotopic activity per unit mass (g1Ci/gm)

MSG = Water mass in SG (gin)

CEASEC = I These isotopic activities were inserted into the nuclear inventory file SECO5.NWF for use by RADTRAD.

The file is listed in Attachment F of Enclosure 6 to Reference 1 and consists of the 5 iodine isotopes. The activities are the total secondary activities and are not per unit power. Thus a power of one should be designated when employing these files. The corresponding iodine release fraction in CEASEC.RFT is unity, since the radionuclides are homogeneously and instantaneously distributed throughout the secondary at the beginning of the event.

(F) lodines, Noble Gases, and Particulates released from the gas gaps and melted fuel for the containment pathway:

Per Reference 2 Regulatory Position 3, Reference 2 Appendix H, and Enclosure 11 to Reference 1, the isotopic release fractions for the containment pathway are defined as 100% of the noble gases, 25% of the iodines, 30% of the alkali metals, 5% of the tellurium metals, 2% of the bariums and strontiums, 0.25% of the noble metals, 0.05% of the cerium group, and 0.02% of the lanthanides contained in the fuel that is estimated to reach initiation of melting and 10% of the noble gases, 10% of the iodines, and 24% of the alkali metals that are contained in the gas gaps of the fuel that experience clad failure. The isotopics are 22

ATTACHMENT (1)

REQUEST FOR ADDITIONAL INFORMATION -- IMPLEMENTATION OF ALTERNATIVE SOURCE TERM identical to those derived in Enclosure 10 to Reference 1 (CRCB63.NIF). The corresponding release fractions in CEACTMT2.RFT are comprised of 8% of the melted fuel release fractions and 2% of the gap release fractions. All releases are instantaneous: 0.0001 sec.

Ai0 = ASTi o Ai0 = ASTi = Isotopic activity per unit power (Ci/MWT) in CRCB63.NIF

" CEAFFI(Iodine) = Fm

Fmi + Fc

Fci = 0.022

" CEAFFN(Noble Gas) = Fm

Fmn + Fc *Fcn = 0.082

CEAAMS(Alkali Metal) = Fm

Fma + Fc

Fca = 0.0288

" CEAAMS(Tellurium) = Fm

Fmt = 4.OE-03

CEAAMS(Strontium) = Fm

Fms = 1.6E-03

" CEAAMS(Barium) = Fm

Fmb = 1.6E-03

" CEAAMS(Ruthenium) = Fm

Fmr = 2.OE-04

" CEAAMS(Cerium) = Fm

Fmc = 4.OE-05

" CEAAMS(Lanthanum) = Fm

Fml = 1.6E-05 o Fm = 0.08 = Melted fuel fraction o Fc = 0.02 = Clad damaged fuel fraction o Fmi = 0.25 = Melted fuel iodine release fraction for secondary pathway o Fci = 0.10 = Iodine release fraction from clad damaged fuel o Fmn = 1.00 = Melted fuel noble gas release fraction o Fcn = 0.10 = Noble gas release fraction from clad damaged fuel o Fma = 0.30 = Melted fuel alkali metal release fraction o Fca = 0.24 = Alkali Metal release fraction from clad damaged fuel o Fmt 0.05 Melted fuel tellurium release fraction o Fms = 0.02 = Melted fuel strontium release fraction o Fmb = 0.02 = Melted fuel barium release fraction o Fmr = 0.0025 Melted fuel ruthenium release fraction o Fmc = 0.0005 = Melted fuel cerium release fraction o FmIl = 0.0002 = Melted fuel lanthanum release fraction

18. RG 1.183, Appendix H, Section 1, indicates that, for airborne activity releasesfrom the primary system following a CEAE accident, an assumption of 25% of the melted fuel fraction of iodine is availablefor releasefrom containment. One would typically assume that the 25% value is meant to take credit for plateout associated with the release to containment resultingfrom this particular CEAE accident pathway. Where no containment leakage pathway is postulated,afraction of 50% is typically used, as can be seen for the boiling-waterreactor control room emergency airguidance in Attachment C of RG 1.183. Therefore, provide information to justify the use of Powers' Containment Natural Deposition Modelfor crediting iodine removal in the CEAE, reactorpressure vessel (RPV) breech case, while also assuming a 25% (as opposed to 50%) iodine activity release from containmentthrough this pathway.

Response: Per Reference 2 Appendix H.1, 25% of the iodines contained in the melted fuel fraction are available for release from containment. Per Reference 2 Appendix H.6.1, the amount of radioactive iodine available for leakage from containment may be reduced due to natural deposition according to the guidance contained in Reference 2 Appendix A. Per Reference 2, Appendix A.3.2, the amount of airborne radioactivity in containment may be reduced using acceptable models in SRP 6.5.2 and 23

ATTACHMENT (1)

REQUEST FOR ADDITIONAL INFORMATION -- IMPLEMENTATION OF ALTERNATIVE SOURCE TERM NUREG/CR-6189 (i.e., the Power's Containment Natural Deposition Model). We used the guidance in NUREG/CP-6189 to reduce the amount of airborne radioactivity in containment after assuming the required release in Reference 2, Appendix H. 1. In addition, the CEAE accident release was increased by the radial peaking factor of 1.70, yielding a total release of 42.50%. Therefore, we believe we remain in compliance with the provisions of Reference 2.

Also note that Reference 2 Appendix A.3.2 states that the practice of deterministically assuming that a 50% plateout of iodine from the fuel is no longer allowed since it is inconsistent with the characteristics of the revised source, term. Therefore, we believe that the 25% iodine release does not include a deterministically derived plateout. This position is consistent with Reference 2 Appendix H. 1.

Maximum Hypothetical Accident (MHA)

19. In the MHA Design Analysis CA06449, Revision 0, as a design basis, it is assumed that an 8-hour period is requiredto establish SDC from a HFP condition, which is consistent with "actualplant operation,"After the 8-hour period where SDC is estdblished, the postulatedRCS activity release to the environment ceases. Because of this dynamic activity release, it cannot be assumed that the activity concentration in the environment, resultingfrom this release, is at its highestfrom the onset of the accident-initiatedleakage; moreover, and more specifically, the determination of a worst 2-hour EAB dose must accountfor the actual activity release profile. It appears that the licensee attempted to compensatefor this effect by postulating an activity release case that assumes SDC is restored in 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , as opposed to 8. However, this case misses the peak environmental activity concentration associated with the design basis 8-hour release period, which may in fact be even higher than that associatedwith the hypothetical 2-hour release period case. By assigning the EAB dispersion coefficient from time equal to 0 through the end of the accident, in the 8-hour periodcase, the code would be allowed to identify the peak 2-hour dose period associatedwith the DBA release scenario.

Please provide verification that the methodology used to determine the worst-case 2-hour EAB dose for the design basis MHA scenario, as described in the submittal, was both conservative and accurate.

Response: Although an M-A release is not assumed to be terminated when SDC is established, we understand that this case may miss the peak environmental activity concentration associated with the design basis release period. By assigning the EAB dispersion coefficient from time equal to 0 through the end of the accident, the RADTRAD code would be allowed to identify the peak 2-hour dose period associated with the DBA release scenario. Therefore, the containment and penetration room pathways were rerun with the EAB dispersion coefficient assigned from 0 to 30 days post-LOCA. Note that the MHA containment and vent stack pathway cases were rerun with a linear ramp isotopic release rate and with containment sprays effective on elemental iodine from 90 seconds till 1.7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months post-LOCA per the requirements of NRC question 20 (below). The results are as follows:

" The worst-case containment pathway EAB dose was during the 0.4-2.4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months period and resulted in a decrease in the EAB dose from 1.8988 Rem TEDE to 1.6975 Rem TEDE.

" The worst-case penetration room pathway EAB dose was during the 0.8-2.8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period and resulted in a decrease in the EAB dose from 0.1838 Rem TEDE to 0.1254 Rem TEDE.

These results will be incorporated into the MHA design basis as described in the response to NRC question 20 (below).

24

ATTACHMENT (1)

REQUEST FOR ADDITIONAL INFORMATION -- IMPLEMENTATION OF ALTERNATIVE SOURCE TERM

20. It appears that the licensee assumes an instantaneous release of core activity for the "early in-vessel" phase, as opposed to the 1.3 hour3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months , linear, release specified in the guidance of RG 1.183.

Due to the credit being taken for time-dependent containment removal mechanisms, it is possible that the instantaneous methodology implemented may yield comparatively non-conservative results.

Additionally, this instantaneous early in-vessel phase release methodology will likely skew the determination of the worst-case 2-hour EAB dose. Therefore, provide verification that the instantaneous early in-vessel phase release methodology that has been used is conservative when comparedto the regulatoryguidance.

Response: We believe that regulatory guidance allows both ramp and step increases to model isotopic releases as described in Reference 2, Footnote 12. We initially used a step increase in the MHA calculation for ease of modeling the DF for spray removal of elemental iodine. A DF of 14.04 was calculated for spray removal of elemental iodine. Based on a spray removal rate of 14.816/hr, an effective removal time of 0.178 hours0.00206 days 0.0494 hours 2.943122e-4 weeks 6.7729e-5 months can be calculated for the step increase, starting at 0.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months post-LOCA. For a linear ramp increase, this methodology is inappropriate. A DF of 14.04 means that

-92.88% of elemental iodine must be removed before the sprays become ineffective on elemental iodine.

For a linear ramp increase, 92.88% of the iodine is not injected into Containment until -1.7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months post-LOCA.

However, to respond to the above question, a sensitivity study was done and the MHA containment and vent stack pathway cases were rerun with a linear ramp isotopic release rate and with containment sprays effective on elemental iodine from 90 seconds till 1.7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months post-LOCA. The original and updated results are as follows:

MHA Results Comparisons Results EAB Rem LPZ Rem CR Rem Containment Pathway Step 1.8988 0.4958 3.9682 Ramp 1.6975 0.4227 3.7957 Penetration Room Pathway Step 0.1838 0.0485 0.3968 Ramp 0.1254 0.0350 0.3699 Note that the isotopic ramp increase is less conservative than the original isotopic step increase. The 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months EAB doses for the ramp cases are worst-case taken over 0-30 days per the Regulatory Guidance provided in NRC question 19 above.

Since this NRC question notes that the regulatory guidance prefers a linear ramp increase, the MHA results are modified to incorporate the preferred NRC regulatory guidance. The new MHA results are as follows:

MHA Results Results EAB Rem LPZ Rem CR Rem Containment Pathway 1.6975 0.4227 3.7957 Penetration Room Pathway 0.1254 0.0350 0.3699 Refueling Water Tank Pathway 1.9676E-05 1.8560E-03 3.2979E-01 Hydrogen Purge Line Pathway I 6.4918E-05 1.5283E-05 7.7048E-05 25

ATTACHMENT (1)

REQUEST FOR ADDITIONAL INFORMATION -- IMPLEMENTATION OF ALTERNATIVE SOURCE TERM Containment Shine 0.0547 Plume Shine 0.0030 Control Room Filter Shine 0.0139 Total 1.8229 0.4595 4.5670 Regulatory Limits 25 25 5 Section 4.1.1 of Attachment I (Reference 1) is revised as follows:

4.1.1 Maximum Hypothetical Accident (MHA)

Section 14.24 of the CCNPP UFSAR describes the MHA. The MHA is a non-mechanistic scenario which evaluates the Containments' capability to contain released radioisotopes. Safety system effectiveness is not considered; the quantity of radioisotopes released to the containment atmosphere is dependent on the power level (MWt) of the reactor. The criteria for this release are established so that the magnitude of the release bounds all credible accident releases.

Enclosure (1) contains the detailed M1HA radiological consequences design basis calculation using the AST. The following supporting calculations are also provided: Atmospheric Dispersion Coefficient (X/Q) calculation (Enclosure 9), Source Terms calculation (Enclosure 10), and Primary and Secondary Isotopic calculations (Enclosure 12).

The MHA re-analysis for AST implementation considered the following radiological release pathways for offsite and Control Room doses.

Containment pathway

Hydrogen Purge line pathway

Ventilation stack pathway

Refueling Water Tank pathway

Containment shine

Plume shine

Control Room Filter shine Major assumptions and required plant modifications considered in the MHA re-analysis to meet regulatory requirements are:

" A bounding Control Room in-leakage value of 3,500 cfm,

Modification of the CREVS to a nominal 10,000 cfm flow with 90 percent filtration efficiency for elemental and organic iodine and 99 percent for particulate iodine was credited,

Installation of automatic isolation dampers and radiation monitors at Access Control HVAC Unit RTU-1 and Access Control Air Conditioning Unit 13 on the Auxiliary Building roof,

Revision to the TS 3.4.15 limit for RCS activity from 1.0 jiCi/gm to 0.5 gCi/gm,

Revision to the TS 5.5.16 maximum allowable containment leakage rate, La, from 0.20 percent of containment air weight per day at Pa to 0.16 percent of containment air volume per day at Pa, 26

ATTACHMENT (1)

REQUEST FOR ADDITIONAL INFORMATION -- IMPLEMENTATION OF ALTERNATIVE SOURCE TERM Sump pH neutrality and iodine re-evolution from the sump is controlled by neutral pH which is maintained by adequate trisodium phosphate dodecahydrate (TSP) in the Containment. The mass of TSP required to neutralize the containment sump pH following a loss-of-coolant accident (LOCA) is calculated assuming a boron level of 3105.5 ppm. This mass of TSP is converted into a TSP volume of 289.3 ft3 (TS 3.5.5 LCO limit for TSP volume). Note that the dodecahydrate form of TSP is used because of the high humidity in the Containment during normal operation.

Since the TSP is hydrated, it is less likely to absorb large amounts of water from the humid atmosphere and will undergo less physical and chemical change than the anhydrous form of TSP, and

Additional sources of acid in the containment sump post-LOCA were identified and analyzed per NUREG/CR-5950 "Iodine Evolution and pH Control" including nitric acid generated by irradiation of air and water in Containment, hydrochloric acid generated by the breakdown of certain cable jacketing when it is exposed to high temperatures and radiation, hydriodic acid generated by a reaction of iodine in water, and carbonic acid generated by water absorbing CO2 from the air or from limestone concrete. The increase in TSP required to neutralize the additional acid sources was found to be insignificant and to have a negligible effect on pH.

Table 1 summarizes the results of the MHA re-analysis for the EAB, LPZ, and Control Room doses. The corresponding regulatory dose limits are shown in the last row of the table. All calculated doses are less than the regulatory limit.

Table 1 MHA Results Control Room Cases EAB (Rem) LPZ (Rem) (Rem)

Containment Pathway 1.6975+-189M 0.42270.4958 3.79573-.968-2 Penetration Room Pathway 0.12540.t838 0.03500-.048-5 0.36990-.N" Refueling Water Tank Pathway 1.9676E-05 1.8560E-03 3.2979E-01 Hydrogen Purge Pathway 6.4918E-05 1.5283E-05 7.7048E-05 Containment Shine 0.0547 Plume Shine 0.0030 Control Room Filter Shine 0.0139 Total 1.82292.0827 0.45950-.-546 4.56704-.7664 Regulatory Limits 25 25 5 The above section replaces Section 4.1.1 in Attachment 1 of Reference 1.

21. Were both trains of the penetration room emergency ventilation system (PREVS) assumed in the analysis to be in operation? If so, verify that both trains will be availablefollowing the loss of offsite power associatedwith the DBA analysis.

Response: Per TS 3.7.12 "PREVS," two PREVS trains shall be operable in Modes 1, 2, and 3. Following a LOCA, a containment isolation signal will realign the PREVS dampers and start both of the PREVS fans to initiate filtration. Following a LOOP associated with a DBA, the emergency diesel generators (EDGs) will start, and both penetration room exhaust fans will be immediately sequenced onto the EDGs.

27

ATTACHMENT (1)

REQUEST FOR ADDITIONAL INFORMATION -- IMPLEMENTATION OF ALTERNATIVE SOURCE TERM Assuming that the worst single active failure is the failure of an EDG, only one EDG will start, and thus only one PREVS train will start. As described in the UFSAR Section 6.6, this is acceptable since the PREVS flow rate for a single train is designed to exceed the containment leakage rate into the penetration room.

Therefore, only a single train of the PREVS is assumed to operate in the alternate source term (AST) calculations.

22. Do the licensee's procedures allow for the refueling water tank (RWT) to be refilled after an accident? If so, provide verification that this was accountedfor in the analysis of MHA RWT release pathway, i.e., was the motive force to flushing the RWT consideredalong with pH considerations?

Response: The LOCA response emergency operating procedure makes no allowance for refilling the RWT post-recirculation actuation signal (RAS). In addition, after a design basis LOCA, including a LOOP and the single failure of an EDG, pumping capacity to refill the RWT will be severely limited.

Note that the LOOP and the single failure of an EDG is conservatively assumed to last the entire duration of the accident. However, it was assumed in this analysis that the recirculating sump water could leak back into the RWT and flash to steam, releasing 10% of the isotopics into the RWT air volume.

Note also that if we were to refill the RWT after an accident, any airborne release would be quickly terminated due to submergence of the water inlet piping early during the refilling process. The RWT post-RAS has approximately 2' of water in the bottom and since the recirculating water inlet and makeup water pipes and the water inlet and outlet pipes are located at about this level, 3" more water would completely submerge all of the piping. Therefore, per regulatory guidance, the leakage can be assumed to mix with the water without flashing during periods of total pipe submergence. So the addition of 3" of water into the RWT would terminate the airborne isotopic release into the RWT, decreasing the dose contribution of the RWT pathway.

As noted above, we conservatively assume that recirculating sump water leaks back into the RWT and that 10% of the isotopics in the water are released to the RWT air volume despite the probable submergence of the pipes.

For pH considerations, reference is made to NUREG/CR-5950 "Iodine Evolution and pH Control."

Based on the results of experimental studies, the formation of 12 during radiolysis can be summarized as:

At pH<3, virtually all iodine is converted to 12.

" At pH>7, only a tiny fraction is converted.

" For 3<pH<7, conversion is a function of iodine concentration in gm/liter, pH, and temperature. See Figure 3.1 in NUREG/CR-5950.

Based on the following assumptions:

minimum RWT pH is 4.5 due to the high boric acid content

" RWT temperature of 85'F

" 50% of the core iodines are deposited in the sump 3

sump volume of 68329 ft

" RWT water volume of 2705.3 ft3

" sump to RWT leakrate of 1.7720E-03 cfm 28

ATTACHMENT (1)

REQUEST FOR ADDITIONAL INFORMATION -- IMPLEMENTATION OF ALTERNATIVE SOURCE TERM one can calculate the following iodine source in gm/l as a function of time.

Gram per Liter of Iodine in RWT vs Time Time(day) Time(min) source(gm/i) 32 0 1 1440 1.4904E-07 2 2880 2.6124E-07 5 7200 4.8347E-07 9 12960 6.1173E-07 10 14400 6.2301E-07 11 15840 6.2819E-07 12 17280 6.2819E-07 13 18720 6.2382E-07 14 20160 6.1580E-07 20 28800 5.2177E-07 30 43200 3.2758E-07 The peak iodine concentration is 6.3E-7 gm/l at -41I days post-LOCA. With a pH of 4.5, the fraction of iodine as 12 is less than 2%.

Referring to Section 3.3 of NUREG/CR-5950, the iodine partition coefficient, which is the ratio of 12 aqueous to 12 gas, is 9 PC = 10^(6.29 - 0.0149

T(K)) = 60.44 which again indicates that less than 2% of the iodine will convert to 12 and re-evolve.

Thus, the assumption of a 10% flashing fraction is bounding and conservative.

23. Clarify and provide additionaldetails as to the RWT release .methodologythat is implemented Such additional details may be included in the text of Reference 34 of Design Analysis CA06449, Revision 0.

Response: Per Reference 2, engineered safety feature (ESF) systems that recirculate sump water outside of the primary containment are assumed to leak during their intended operation. This release source may include leakage through valves isolating interfacing systems. There is a potential for an unmonitored release pathway resulting from the post-LOCA leakage of isolation valves in the Safety Injection or Containment Spray System recirculation lines to the RWT, which is vented directly to the atmosphere.

During the recirculation phase, sump water is recirculated through the Emergency Core Cooling System (ECCS) pumps and could leak through various valves and reach the RWT. The two pathways include the two valves in series in the minimum flow recirculation line header (MOV659/660) and the valve from the containment spray pumps (S1459). This was modeled by utilizing the RADTRAD computer code.

The isotopic activities released from the failed fuel per unit power were generated by the isotope generation and depletion computer code SAS2H/ORIGEN-S and multiplied by the relevant power level and release fractions. The core inventory release fractions by radionuclide groups for the gap release and early in-vessel damage phases were extracted from Reference 2 Table 2; however, per Reference 2 29

ATTACHMENT (1)

REQUEST FOR ADDITIONAL INFORMATION -- IMPLEMENTATION OF ALTERNATIVE SOURCE TERM Appendix A Section 5.3, with the exception of iodine, all radioactive materials in the recirculating liquid should be assumed to be retained in the liquid phase. The minimum time to RAS of 32 minutes determines the beginning of the sump recirculation phase. During the recirculation phase, sump water is recirculated through the ECCS pumps and could leak through various valves and reach the RWT. This leak rate is doubled per the requirements of Reference 2. Ten percent of the fluid leaking into the RWT is assumed to flash to steam and release its iodine content into the RWT atmosphere. Per Reference 2, reduction in release activity by dilution or holdup within buildings or by ESF ventilation filtration systems, may be credited where applicable. In order to determine the amount of RWT release to the atmosphere, the breathing of the RWT through the vent due to diurnal temperature variations was calculated. The breathing was determined by performing a 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months transient thermal analysis of the RWT using the computer code TSAP and assuming RWT atmospheric temperature characteristics based on the maximum solar load during summer solstice and a minimum ambient nighttime temperature. This methodology results in an average 4.2 cfm leakrate from the RWT to the environment. This evaluation is included as Appendix 2.

24. Clarify how PREVS is associatedwith the hydrogen (H2)purge line leakage.

Response: While the PREVS is a standby system, it may also operate during normal unit operations, e.g., venting the containment through the 4" containment vent (hydrogen purge) line. Note that if the containment venting is active at the start of a LOCA, it will be terminated by a containment radiation signal, closing the valve in each vent penetration. The valve closure time is controlled by TS and the UFSAR and is included in the analysis as described below.

25. Provide additionaldetails as to how the H2 purge line leakrate was determined, and verify that there are no other driving forces (i.e., operatingfans, etc.) associatedwith a post-LOCA release through the H2 purge lines.

Response: Per Reference 2, if the primary containment is routinely vented during power operations, releases via the vent system prior to containment isolation should be analyzed and the resulting doses summed with the postulated doses from other release paths. The containment vent system is used for pressure and containment radioactivity control purposes, and the vent isolation valves may also be opened for surveillance testing. The 4" vent line isolates on a safety injection actuation signal and containment radiation signal and is assumed to close fully within 30 seconds.

The maximum total time of 30 seconds for valve closure is based on 2.4 seconds for containment pressure buildup, instrument response, and safety injection actuation signal delay, 10 seconds for EDG startup, and 17.6 seconds for valve stroke time. The failure of one EDG would not affect the results, because the vent lines have redundant isolation valves powered from separate EDGs. The vent system is isolated before the onset of the gap release phase, thus release fractions associated with gap release and early in-vessel phases need not be considered.

The 4" containment vent line flow rate was calculated assuming a peak containment pressure of 64 psia for the full 30 seconds that the line is open as shown below.

Variable Value Units Description N2/N1 0.2296 N2/N1 =P2/P1 Mole fraction of air MWair 29 Molecular weight of air (Crane P.A-8)

MWwater 18 Molecular weight of water (Crane P.A-8) 30

ATTACHMENT (1)

REQUEST FOR ADDITIONAL INFORMATION -- IMPLEMENTATION OF ALTERNATIVE SOURCE TERM Variable Value Units Description MWmix 20.5259 Molecular weight of mixture=MWair*N2/N 1 +MWwater*(1 -N2/N 1)

R 75.2708 Gas constant=1545/M (Crane p.1.0)

Absolute temperature in degrees Rankine (288+460) (UFSAR T 748 R Figures 14.20-5/6)

Rho 0.1637 Ibm/cf Rho=144*P/(R*T)

Ratio of specific heat at constant pressure to specific heat at constant k 1.3 volume (steam)

P1 64 psia Containment design pressure (UFSAR Table 14.20-3)

P2 14.696 psia STP pressure (Steam Tables) dP/P 0.7704 psia (P1-P2)/P1 rho 0.1637 Ibm/cf Containment density (M-89-33) d 4 inches Pipe diameter (M-89-33)

K 15.3 Flow resistance coefficient (M-89-33)

Y 0.735 Expansion factor (Crane p.A-22) w 4.4840 Ibm/sec Darcy Equation: w=0.525*Y*dA2*sqrt(dp*rho/K) (Crane Eq 1-11)

F 1643.84 cfm F=w*60/rho A more exact derivation of the containment vent flow would consider the following:

The major driving force associated with a containment vent is the containment pressure, which starts at 16.5 psia at the beginning of the LOCA per UFSAR Table 14.20-3, rises to <59 psia at -10 sec, and remains <59 psia till valve closure at 30 sec per UFSAR Figures 14.20-4 and 14.20-5.

" An additional driving force is the fan in the PREVS system, which creates a pressure drop of 6 inwg or 0.218 psia across the PREVS filters per TS 5.5.1 1.d.

" Additionally, the throttling effect of the closing isolation valves would further reduce the calculated discharge rate during the stroke time. Integrating over the diameter in the Darcy equation yields an average flow rate 1/3 of that assumed in this analysis over the stroke time. A throttling of 0.5 will be conservatively assumed over a 15 second stroke duration in the following analysis.

Containment Vent Line Flow Rate vs Time Pen Room Containment Throttling Purge Line Ratio to Time Pressure Pressure Effect Flow Rate 1645 cfm (sec) (psia) (psia) (cfm) 0 14.478 16.5 10 14.478 59 1 1080.2095 0.6567 15 14.478 59 1 1561.8350 0.9494 30 14.478 59 0.5 780.9175 0.4747 Average 1010.8344 0.6145 Thus, the 0-30 sec flow ratio is 61%; that is, the 1645 cfm value used in this analysis is overstated by

-62%. Thus, the containment vent line flow rate value of 1645 cfm used in Enclosure 1 of Reference 1 was conservative.

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REQUEST FOR ADDITIONAL INFORMATION -- IMPLEMENTATION OF ALTERNATIVE SOURCE TERM

26. The treatment of all isotopes (excluding noble gases) as particulate conservatively results in the accumulation of all relevant activity, regardless of chemical form, being accumulated in one location. However, by assuming 100% particulate,a resultantassumption is made that all activity is depositing on the high efficiency particulateairfilter [human effiieny *partiulate air , ,,teriof the train, not the carbonfilter. This may or may not be conservative, depending on the location of the high efficiency particulate air filter [human error pr.babilityfilteri, versus the carbon filter, in relation to the assumed control room operator position and other dose receiver locations.

Additionally, effects such as oblique angle scattering,which are not well modeled by the point-kernel method (Micro Shield) being used, may lead to unpredictable dose contributions that are dependent on the geometry caused by differentfilter locations.

Therefore, if the licensee chooses to continue modeling all activity, regardless of chemicalform, in the samefilter location,provide verification that the filter locations selected are limiting with regard to dose, and well modeled by a point-kernel methodology.

Response: The MicroShield model in Enclosure I of Reference 1 was a simple 24"x33"x59" carbon filter, shielded at the bottom by 6" of air and 2' of concrete, with the dose point located 1" from the concrete on the filter centerline. By assuming a dose point 1" below the Control Room ceiling on the centerline of the filter unit, the maximum operator dose is assured. Additional details concerning dose modeling are contained in Appendix 3.

To demonstrate the conservatism of the MicroShield model, a more detailed three-dimensional representation is constructed with MCNP. This code is a general-purpose Monte-Carlo N-Particle code that can be used for neutron, photon, electron, or coupled neutron/photon/electron transport. This sensitivity study is contained in Appendix 3.

Table 1 below provides a comparison of the MicroShield and MCNP dose results for the same filter medium and source, and identical dose point locations 1" from the concrete below the filter. As can be seen, the MicroShield result is within 1% of the MCNP result. This is within the range of uncertainty of the MCNP result, and therefore it can be concluded that the MicroShield results are essentially the same.

Thus backscatter is not a significant contributor for this dose location, which is conservative, and the point-kernel method is considered acceptable for calculating dose at that location.

Table 1 Unadjusted Zero Decay Dose Rate at 1" from Concrete Below Filter Model Dose Rate (mrem/hr)

MicroShield Carbon Filter 2.906E+7 MCNP Carbon Filter (429675767 particles run) 2.937E+7 +/-3%

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27. Provide verification that the larger dimensions that characterizea 10, 000 cfm filter will not bound those of the smaller 2000 cfm filter being assumed, despite the associated reduction in source concentration.

Response: To verify that the larger filter banks loaded with the same amount of activity are bounding, a sensitivity study of dose rate versus filter size was performed using MicroShield for a single control room filter shine case. The width and length were varied from 24" x 33" to 48" x 66" to 96" x 132", while the height was varied from 59" to 109" to 209". The results are listed in the following table. The geometry required for MicroShield execution is the following:

" The current filter unit is 24" in width by 33" in length by 59" in height and is assumed to be composed of carbon. The unit is located 6" above the 69' elevation concrete floor.

" The Control Room ceiling extends from 67' to 69' elevation and is 2' of concrete.

" The dose point for the MicroShield executions is 1" beyond the concrete floor/ceiling, centered on the filter.

" The current filter units are 2000 cfm, not the 10000 cfm filter units that will replace them. Use of a smaller unit in this analysis should concentrate the radionuclides and thus maximize the dose.

Control Room Filter Dose Rates vs Filter Size Filter Length, Width, Height 59" x 24" x 33" 109" x 24" x 33" 209" x 24" x 33" MS5 JobName MHG720 MHG720C MHG720D Dose Rate 2274 mR/hr 1228 mR/hr 592.7 mR/hr Filter Length, Width, Height 59" x 48" x 66" 109" x 48" x 66" 209" x 48" x 66" MS5 JobName MHG720A MHG720B MHG720H Dose Rate 785.9 mR/hr 424.5 mR/hr 206.3 mR/hr Filter Length, Width, Height 59" x 96" x 132" 109" x 96" x 132" 209" x 96" x 132" MS5 JobName MHG720E MHG720F MHG720G Dose Rite 200.7 mR/hr 108.4 mR/hr 52.72 mR/hr Note that in all cases, a larger filter volume results in a smaller dose rate. Thus, use of the current 2000 cfm filter banks in lieu of the larger 10000 cfm filter banks is conservative.

Waste Processing Incident (WPI) and Waste Gas Incident (WGI)

28. Explain the rationalefor re-analyzing these two accidents, as neither are requiredforfull-scope AST implementation.

Response: In our initial submittal (Reference 1) all UFSAR Chapter 14 accidents were evaluated for AST impact. Since these events are currently contained in UFSAR Chapter 14, they were included in the evaluation.

Later discussions with NRC staff has indicated that re-analysis of these events was optional. We therefore withdraw these events from review in this submittal.

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REQUEST FOR ADDITIONAL INFORMATION -- IMPLEMENTATION OF ALTERNATIVE SOURCE TERM METEOROLOGICAL DATA 29.

400.00 350.00 300.00 8 Average 250.00

200.00 E E-A-C 2 r-G F

ýc 150.00 8 Average 100.00 50.00 8Averago 0.00 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 Hour of Day The NRC staff performed confirmatory analysis of the hourly atmospheric stability values provided and noticed some variations relative to the distributionsfor certain classes. If the stability classes A thru G are combined to form only three different class groups:

Unstable (A-C), Neutral (D-E), and Stable (F-G), the variance becomes apparent. The above figure is a composite view of each year's hourly stability annual occurrences and their variationfrom the annual meanfor all 8 years combined (1991 - 1998):

In the figure, it shows that unstable stability classes A-C have an average standarddeviation of 19.07, a maximum of 53.83, and a minimum of 0.83. Neutral stability classes D-E have an average of 25.17, a maximum of 52.98, and a minimum of 9. 01. Stable stability classes F-G were the most behaved data with an average standard deviation of 8.35, a maximum of 22.18, and a minimum of 1.67. Given the uncertainty in the data, justify why annual calculations of atmospheric dispersionfactors (x/Q values) should not be performed independently (each year separately)providing the most conservative resulting values listedfor accidentalatmosphericreleases.

Response: Per RG 1.194 Section 3.1, "The meteorological data needed for X/Q calculations include wind speed, wind direction, and a measure of atmospheric stability. These data should be obtained from an onsite meteorological measurement program based on the guidance of Safety Guide 23, "Onsite Meteorological Programs," that includes quality assurance provisions consistent with 10 CFR Part 50, Appendix B. The meteorological data set used in these measurements should represent hourly averages as defined in Safety Guide 23. Data should be representative of the overall site conditions and be free from local effects such as building and cooling tower wakes, brush and vegetation, or terrain. Collected data should be reviewed to identify instrumentation problems and missing or anomalous observations."

Calvert Cliffs complies with these requirements per Technical Requirements Manual 15.3.2 Meteorological Instrumentation.

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REQUEST FOR ADDITIONAL INFORMATION -- IMPLEMENTATION OF ALTERNATIVE SOURCE TERM Also per RG 1.194 Section 3.1, "The NRC considers 5 years of hourly observations to be representative of long-term trends at most sites." Calvert Cliffs used eight years (1991-1998) of meteorological observations in its submittal to the NRC. Five additional years (2000-2004) of meteorological observations were processed in response to this RAI. Note that meteorological data for 1999 was not readily available and therefore was not included. The atmospheric dispersion coefficients for the Unit 1 Containment to the West Road Inlet of the Auxiliary Building were calculated for each available year and for the 1991-1998 interval and for the 1991-2004 interval (less 1999). The results are presented in the following table. Note that the atmospheric dispersion coefficients for the eight year interval 1991-1998 and for the thirteen year interval 1991-2004 are almost identical. Thus the 95th percentile x/Q average values are consistent over many year intervals, although they do vary slightly from year-to-year. We know of no regulatory guidance that requires a worst-case single year X/Q calculation in lieu of a multi-year average.

Atmospheric Dispersion Coefficients (sec/m3)

Unit 1 Containment to West Road of Auxiliary Building (Taut String) 0-2 hour 2-8 hour 8-24 hour 24-96 hour 96-720 hour 1991 1.10E-03 7.04E-04 3.13E-04 2.64E-04 1.91E-04 1992 1.14E-03 7.92E-04 3.49E-04 2.58E-04 1.99E-04 1993 1.14E-03 7.54E-04 3.33E-04 2.59E-04 1.94E-04 1994 1.10E-03 7.34E-04 3.19E-04 2.53E-04 1.88E-04 1995 1.09E-03 7.23E-04 3.26E-04 2.23E-04 2.01E-04 1996 1.11E-03 7.71E-04 3.37E-04 2.04E-04 1.71E-04 1997 1.11E-03 6.86E-04 2.72E-04 2.15E-04. 1.89E-04 1998 1.11E-03 6.67E-04 2.93E-04 2.02E-04 1.77E-04 2000 1.08E-03 7.01E-04 2.93E-04 2.23E-04 1.72E-04 2001 1.12E-03 6.92E-04 3.05E-04 2.30E-04 1.95E-04 2002 1.06E-03 6.95E-04 2.91 E-04 2.37E-04 1.74E-04 2003 1.16E-03 8.02E-04 3.39E-04 2.46E-04 2.19E-04 2004 1.09E-03 7.1OE-04 2.96E-04 2.42E-04 2.03E-04 1991-2004 1.11E-03 7.25E-04 3.13E-04 2.35E-04 1.90E-04 1991-1998 1.11E-03 7.29E-04 3.19E-04 2.36E-04 1.98E-04 Note that the standard deviations for the average and 95th percentile velocities at 10 and 60 meters are very small as shown below, indicating consistent meteorological conditions over the eight year period of interest.

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REQUEST FOR ADDITIONAL INFORMATION -- IMPLEMENTATION OF ALTERNATIVE SOURCE TERM 10m and 60m Average and 95th Percentile Velocities in rn/sec lom 10M 60m 60m 10m vave 10M v95 60m vave 60m v95 vave Deviation v95 Deviation vave Deviation v95 Deviation 1991 3.00 -0.12 5.80 0.67 4.89 0.22 8.60 0.90 1992 2.98 0.18 5.90 0.13 4.85 0.73 8.60 0.90 1993 3.01 -0.27 6.20 -1.51 4.91 -0.03 9.20 -1.50 1994 3.07 -1.16 5.90 0.13 5.06 -1.93 8.80 0.10 1995 3.02 -0.41 5.80 0.67 4.85 0.73 8.50 1.30 1996 3.02 -0.41 6.20 -1.51 4.92 -0.16 9.10 -1.10 1997 3.00 -0.12 5.90 0.13 4.97 -0.79 8.90 -0.30 1998 2.84 2.28 5.70 1.22 4.81 1.23 8.90 -0.30 Average 2.99 1 5.92 4.91 1 8.82 1 Std Dev 0.07 1 0.18 0.08 0.25 ATMOSPHERIC DISPERSION FACTORS

30. Confirm that the cross-sectional area (A) used in the EAB x/Q calculation for each accident scenario (MHA, FHA, MSLB, SGTR, SRE, CEAEA, WGI, and WPI) is in fact "0 in" as denoted in Enclosures 1-8 of the application. It was noted in Calvert Cliffs Updated Final Safety Analysis Report (UFSAR), Section 2.3.6.1, that the EAB 0-2 hour relative concentrationvalue was calculated using a wake factor (c x A) of 820 m2 (0.5 x 1640 m2). Justify the reasoningfor changing this value to 0 meters for sources: AD Vs, ventilation stacks (VSs), main steam goosenecks (MSGs), RWTs, and containment outage doors (CODs). Note that the value of "A," is defined as the vertical-planecross-sectional area of the reactor building, measured in m2 (refer to RG 1.145, Section C - 1.3.1).

Response: The Gifford wake model uses the idea of the building wake providing an initial plume dilution proportional to the product of wind speed and projected building area. For containment leakage, the building area is the projected containment area over which the leakage occurs. For the ADVs, ventilation stacks, main steam goosenecks, and RWTs, it was assumed that the effluent plume may not be completely entrapped by the building wake. The entrapment fraction depends on parameters such as the ratio of effluent exit speed to mean wind speed, the ratio of stack height to characteristic vertical dimension, source location, building geometry, atmospheric stability, turbulence intensity, and wind speed and direction. A more detailed discussion of the Gifford wake model and possible deviations may be found in DOE/TIC-27601 "Atmospheric Science and Power Production," Chapter 7 "Flow and Diffusion near Obstacles."

The worst case release for these sources would entail a zero entrapment fraction. Since this was most conservative, it was employed in the Calvert Cliffs submittal.

31. The NRC staff routinely performs confirmatory analysis on data presented by a licensee when submittinga change in the currentlicensing basis. In order to conduct these evaluationsfor the low-populationzone (LPZ) atmosphericdispersion estimates (X/Q values), provide the x/Q values for the time intervals of 2- 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , 8-24 hours, 24- 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months , and 96-720 hours. Accompany these values with any inputs and assumptions made to validate that the most conservative offsite atmospheric dispersion estimate was used.

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REQUEST FOR ADDITIONAL INFORMATION -- IMPLEMENTATION OF ALTERNATIVE SOURCE TERM Response: The current licensing basis for the LPZ atmospheric dispersion coefficient values has not changed. As detailed in Reference 1, the EAB and LPZ atmospheric dispersion coefficients used in the re-analysis of all the radiological consequences DBAs were obtained from Section 2 of the current Calvert Cliffs UFSAR. These atmospheric dispersion coefficients are part of the existing design basis offsite dose calculations and are not expected to change, based on the atmospheric dispersion coefficient results detailed in Question 29. In addition, per RG 1.194 Regulatory Position C.2, "holders of operating licenses may continue to use X/Q values determined with methodologies previously approved by the NRC staff and documented in the facility's final safety analysis report (FSAR)."

The atmospheric dispersion coefficients from the containment to the LPZ (2 miles) were extracted from UFSAR Figure 2.3-3/UFSAR Section 14.24.3.

Time (hours) I/Q (sec/m3) 0-2 3.30E-05 2-24 2.20E-06 24-720 5.40E-07

32. In reviewing the source/receptor distances (Attachment C of Enclosure 9, Atmospheric Dispersion Coefficient (X/Q) CalculationCA06012, Rev. 0) for the eight postulated accidents, the methodology shown uses mostly straight-line calculations and taut-string calculationsfor the containment unit source releases. However, reviewing the plant layout and relative distances raisedsome concernfor the length between-the Access Control 13 (AC13) and north wall of the Auxiliary Building. If drawn to scale, the layout shows that the distance D, listed as 16.3031 feet on page 34, more accurately resembles 25.6667 feet, listed as distance E on page 34, and vice-versa. If this is so, all source/receptor distance calculations performed involving AC13 are inaccurate and need to be correctedwith the appropriatedistance of 25.6667feetfor D and 16.3031feetfor E (relative to page 34 of Attachment C). Note that these values are used throughout the source/receptorcalculationsin Attachment C. Accordingly, all ARCON96 X/Q calculations involving AC13 will need to be reevaluated using these corrected distance values for the source/receptor points ADV1AC13, ADV2AC13, CODIACI3, COD2AC13, CTMTIAC13, CTMT2AC13, MSCIAC13, MSG2AC13, RWTIAC13, RWT2AC13, VSIAC13, and VS2AC13. Evaluate these concerns.

Response: The hand drawn addition (i.e., the location of AC13) to the figure on page 34 of CA06012 is not drawn to scale. The distance "E" (25'8") was manually measured in a walkdown documented in Attachment B of CA06012. Distance "F" (27'6") was extracted from BGE Drawing 61685SH0001 Rev. 29. Distance "C" (45'6") was estimated from the to-scale drawing on page 34, but it can also be extracted from BGE Drawing 62050SH0001 Rev. 8. The distance "C+D+E+F" (115'0") was extracted from BGE Drawing 62037SH0001 Rev. 19. Thus distance "D" can be calculated to be 16'4".

Therefore, the atmospheric dispersion calculations involving AC13 are accurate and do not need to be re-evaluated.

33. Provide a copy of the input and output files used to calculate X/Q EAB and LPZ values for the PA VAN computer code.

Response: As detailed in Reference I and the response to question 31, the EAB and LPZ atmospheric dispersion coefficients used in the re-analysis of all the radiological consequences DBAs were obtained 37

ATTACHMENT (1)

REQUEST FOR ADDITIONAL INFORMATION -- IMPLEMENTATION OF ALTERNATIVE SOURCE TERM from Section 2 of the current Calvert Cliffs UFSAR. Therefore, no PAVAN input and output files are available.

VENTILATION SYSTEMS

34. With respect to deleting TS 3.7.10, "Emergency Core Cooling System (ECCS) Pump Room Exhaust Filtration System (PREFS)," if this TS is deleted and design-basis accident doses are maintained within limits, discuss the requirements and the means by which the system will continue to establish and maintain (i.e., drawdown and maintain) a negative pressure so that materials that is released will be release[d]in a manner consistent with the accident analysis. The discussion should include justificationfor not having a TSfor drawdown and a surveillance requirement (SR) to maintain the system. The NRC staff believes that TS 3.7.10 needs an SR somewhat similar to SR 3.7.11.3, which demonstrates that the ECCS pump room area is maintainedat a negative pressure (e.g., 0.25 inches water gage with respect to adjacentareas).

Response: Leakage of the radioactive sump liquid post-LOCA is not considered a release pathway in this analysis. A release in the ECCS pump rooms is not part of our licensing basis.

In 1968, the Preliminary Safety Evaluation Report contained an offsite dose calculation for a low pressure safety injection pump passive seal failure in the ECCS pump room post-RAS. The offsite dose calculation assumed no credit for the ECCS filtration system.

Upon issuance of the original Final Safety Analysis Report in 1972, the passive seal failure analysis in the ECCS pump room had been removed. However, in response to a question from the NRC, Calvert Cliffs Nuclear Power Plant committed to use the charcoal filters when the ECCS pumps are operated after an accident. The original TS, issued about the same time, contained the surveillance requirements for the charcoal filters, but did not contain operability requirements or bases.

The Combustion Engineering Standard TS were adopted in the mid-1970s with the operability requirements and Combustion Engineering bases for the ECCS PREFS. However, the ECCS PREFS were not credited in the dose analyses.

The Improved TS were adopted in the mid-1990s with the same operability requirements and bases as the Combustion Engineering Standard TS. Again, the ECCS PREFS were not credited in the dose analyses.

Therefore the ECCS ventilation system is not used or credited for control of radioactive material following an accident. This system provides for defense in depth and will be treated as such in the UFSAR, as appropriate.

35. The definition of La, as proposed in Attachment 1 of the November 3, 2005, application, is different than the markup of the TS in Attachment 2 of that submittal. In Attachment 1 the requested change is, "La is reduce from 0.20 percent of containment air weight per day at Pa to 0.16 percent of containment air volume per day at Pa. " The marked-up TS in Attachment 2 reduces the maximum allowable containment leakage rate, La, from 0.20 to 0. 16percent of containment air weight per day at Pa. It is not clear what is being requested Clarify this aspect of the request.

Response: The definition of La is revised to reflect the change in the maximum allowable containment leakage rate La used in the Containment Leakage Rate Testing Program (TS 5.5.16). La is reduced from 0.20 percent of containment air weight per day at Pa to 0.16 percent of containment air volume per day at 38

ATTACHMENT (1)

REQUEST FOR ADDITIONAL INFORMATION -- IMPLEMENTATION OF ALTERNATIVE SOURCE TERM Pa. The change from percent of containment air weight to percent of containment air volume is made for consistency with what is assumed in the accident analysis.

On TS pages 1.1-3 and 5.5-24, the reference to "containment air weight" will be changed to "containment volume." The revised marked up pages are contained in Attachment 2.

36. Within the Ventilation Filter Testing Program (VFTP), the ECCS PREFS and SFPEVS acceptance criteriashould have a flow rate. There should also be an SR that links the system 's flow rate to the demonstration of a negative pressure with respect to adjacentareas. Provide this information.

Response: The proposed amendment deletes TS 3.7.10 "ECCS PREFS," thus no flow rate or Surveillance Requirement that links the system flow rate to the demonstration of negative pressure are necessary.

Surveillance Requirement 3.7.11.3 requires verification that each spent fuel pool exhaust ventilation system (SFPEVS) fan can maintain a measurable negative pressure with respect to atmospheric pressure.

Negative pressure is verified via plant procedures which implement Surveillance Requirement 3.7.11.3.

While the SFPEVS flow rate does contribute to the negative pressure in the SFP area, a particular SFPEVS flow rate is not sufficient to guarantee a negative pressure. Boundary closure and ventilation configuration are also important. Thus, measurement of a flow rate as an indication of negative pressure may not be sufficient to guarantee a negative pressure. Note also that inclusion of flow rates in TSs is required only to assure proper filter residence times per RG 1.52. Since the SFP filters will no longer be credited, specification of residence times and flow rates is no longer material to this request.

37. It appears that the acceptability of the control room operator doses is dependent upon the installationof the automatic isolation dampers and radiationmonitors at the Access Control HVAC Unit RTU-1 and Access Control Air Conditioning Unit 13 on the Auxiliary Building roof. The licensee should provide a TS and an SR that incorporates the isolation of these dampers on a high radiationsignal. The requirementshould be for all modes of operation as credit is taken for this function during an FHA.

Response: The Access Control heating, ventilation, and air conditioning (HVAC) Unit RTU-1 and Access Control Air Conditioning Unit 13 on the Auxiliary Building roof provide fresh air and air conditioning for several Auxiliary Building rooms outside of the Control Room. Air flowing into the Auxiliary Building through these units would then have to leak into the control room via doors or walls.

Therefore, the Access Control HVAC Unit RTU-1 and Access Control Air Conditioning Unit 13 on the Auxiliary Building roof do not provide direct access into the control room.

For an accident that includes a LOOP, the Auxiliary Building and access control HVAC fans lose power, and their dampers fail closed. Thus, there should be no flow and thus no differential pressure across the access control units. The inleakage values through the access control units into the Auxiliary Building are negligible. Since the leakage is into the Auxiliary Building and not into the control room envelope, the dose consequences are also negligible.

For an accident in which no LOOP occurs, the dampers and radiation monitors are safety-related and the dampers would close on a high radiation signal. The fans will also deenergize. In addition, forced ventilation from the West Road Inlet would dwarf any minor leakage through the access control units.

38. Calvert Cliffs control room envelope inleakage number is based upon a system configuration and operating characteristics that is no longer relevant. The inleakage test was performed with a 39

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REQUEST FOR ADDITIONAL INFORMATION -- IMPLEMENTATION OF ALTERNATIVE SOURCE TERM CREFS [Control Room Emergency Filtration System] recirculatingrate of 2,000 cdm. The new system will be recirculating10,000 cfm. An inleakage test (e.g., American Society for Testing and Materials[ASTM] E741) should be performed based upon the new CREFS recirculatingflow rate to confirm that the inleakage is less than the 3,500 cfm assumed in accident analyses.

Response: Agreed. When the system modifications are complete, the control room inleakage test will be reperformed. However, control room inleakage is dominated by the Control Room Emergency Temperature System, which has an air flow of 41500 +/-10% cfm per train (BGE Drawing 12782-0035).

Thus, diversion of 10000 cfm instead of 2000 cfm into the filters is not expected to change the control room differential pressure profile and is thus not expected to affect the control room inleakage to any great extent.

39. Each of the accident analyses assumed that the CREFS will begin operation20 minutes following the beginning of the accident. There is no discussion of the operationalstatus of the control room ventilation systems during this 20-minute period Provide a discussion of the configuration of the CREFS during this period and determine the control room envelope inleakage using an acceptable test protocol (e.g., ASTM E741) to confirm that the inleakage is less than the 3,500 cfm assumed in the accident analyses.

Response: The Calvert Cliffs control room ventilation system is operated in continuous recirculation mode with no fresh air inflow other than that which leaks into the system. Prior to initiation of the CREFS, the Control Room Emergency Temperature System is operating at 41500+10% cfm per train (BGE Drawing 12782-0035). When the CREFS is initiated, 2000 cfm of that flow is diverted through the control room filters. The 2000 cfm diversion of air flow into the filters, has negligible effect on the control room inleakage. The control room inleakage with and without CREFS operation with 10000 cfm diversion through the filters will be determined after the control room modifications are complete. The worst case value will be confirmed to be within control room habitability program limits; however, little difference is expected.

40. With regards to the new CREVS operatingscheme and damper isolationfunctions, has the licensee confirmed that sufficient radioactivity is released to cause isolation at the Access Control HVAC Unit RTU-1 and Access Control Air Conditioning Unit 13 on the Auxiliary Building roof?.If not, does the normal system continue to operate? If it does, what is the inleakage rate in this mode and what are the dose consequences to the control room operators?

Response: During procurement of the monitors, the sensitivity of the monitors will be specified such that isolation is assured. The design setpoint for the monitors will be such that accident analysis assumptions are verified. I For an accident that includes a LOOP, the Auxiliary Building supply fans and access control HVAC fans lose power, and their dampers fail closed. Thus, there should be no flow and thus no differential pressure across the access control units. The inleakage values through the access control units into the Auxiliary Building are negligible. Since the leakage is into the Auxiliary Building and not into the control room envelope, the dose consequences are also negligible.

For an accident in which no LOOP occurs, the dampers and radiation monitors are safety-related and would close on a high radiation signal. The fans will also deenergize. In addition, forced ventilation from the West Road Inlet would dwarf any minor leakage through the access control units.

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REQUEST FOR ADDITIONAL INFORMATION -- IMPLEMENTATION OF ALTERNATIVE SOURCE TERM

41. The inplace test criteria in the VFTP for all ESF ventilation system charcoaladsorbers and HEPA

[high efficiency particulateair]filters is 1%. Confirm that all of the accident analyses accountfor this 1% bypass of the adsorberand HEPA filter.

Response: Technical Specification 5.5.11 (a) requires that for each of the ESF systems an inplace test of the HEPA filters shows a penetration and system bypass less than or equal to 1.0% when tested in accordance with Regulatory Position C.5.a and C.5.c of RG 1.52, Revision 2, and American National Standards Institute N510-1975 at the system flowrate specified +/- 10%. Thus, a filter efficiency of 99% is appropriate for the HEPA filters and consistent with TS requirements.

Generic Letter 99-02 "Laboratory Testing of Nuclear Grade Activated Charcoal" requires that "in all cases the results should meet the acceptance criterion that is derived from applying a safety factor of 2 to the charcoal filter efficiency assumed in your design-basis dose analysis." Calvert Cliffs complied with this requirement, determined revised TS filter penetrations based on the design basis dose analyses and the requirements of Generic Letter 99-02, and submitted the revised TS limits to the NRC. The NRC approved the application in License Amendments 238/212 issued on December 7, 2000. Thus the activated charcoal filter efficiencies assumed in the AST submittal are those calculated per the requirements of Generic Letter 99-02 and previously approved by the NRC. Based on the above regulatory guidance, Calvert Cliffs contends that the safety factor incorporates future filter degradation and bypass flow effects.

42. Because the flow ratefor the CREVS [Control Room Emergency Ventilation System] has changed from 2,000 to 10,000, discuss the change in the residence time during emergency operation and when testing the carbon adsorber, and confirm that the residence time during testing is the same as it is during emergency operation.

Response: The control room filters will be sized such that the filter residence time will remain the same.

The requirements of RG 1.52 Rev. 2 will be met: 0.25 seconds residence time per two inches of absorbent bed or a system face velocity less than or equal to 40 ft/min.

REFERENCES:

(1) Letter from Mr. B. S. Montgomery (CCNPP) to Document Control Desk, dated November 3, 2005, License Amendment Request: Revision to Accident Source Term and Associated Technical Specifications (2) Regulatory Guide 1.183, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, dated July 2000 (3) Letter from Mr. S. A. McNeil (NRC) to Mr. J. A. Tiernan (BGE), dated June 30, 1987, Revised Safety Evaluation Supporting Amendment No. 108 to Facility Operating License No. DRP-69 41

APPENDIX 1 Sensitivity Study of Atmospheric Dispersion Coefficients for Releases from ADVs and MSSVs Calvert Cliffs Nuclear Power Plant, Inc.

March 22, 2007

APPENDIX 1 Sensitivity Study of Atmospheric Dispersion Coefficients for Releases from ADVs and MSSVs While the atmospheric dump valves (ADVs) are the primary steam release pathways for the steam generators (SGs), if the SG pressures exceed the main steam safety valve (MSSV) lift setpoints as delineated in Technical Specification 3.7.1 "MSSVs," the MSSVs will lift, and steam will be released through the MSSVs into the atmosphere. While the ADV is the preferred steam release pathway, this can not always be guaranteed. However, it can be shown that the ADV-to-West Road and ADV-to-Turbine Building atmospheric dispersion coefficients generated in Enclosure 9 to Reference 1 are bounding for all ADVs and MSSVs. A sensitivity study was performed and is provided below.

(A) ARCON96 Inputs Each reactor has two SGs, and each SG is equipped with one ADV and eight MSSVs. The location of the ADVs and MSSVs can be determined via Baltimore Gas and Electric Company (BGE) Drawing 62050SH0001 "Auxiliary Building Partial Roof Plans and Details." The distance and angular inputs were calculated and are listed in Table 5A below:

Table 5A ADV & MSSV DISTANCES AND ANGLES FOR ARCON96 INPUT Unit 1- Unit 2- Unit 1- Unit 2-West West Turbine Turbine dWR(m) dTB(m) Road Road Building Building advl 65.128656 39.127818 20 70 269 181 adv2 63.22623 37.97078 22 68 265 185 mssvla 65.4761 37.70125 22 68 268 182 mssvl b 65.207446 37.232722 22 68 268 182 mssvl c 64.944923 36.771016 23 67 267 183 mssvld 64.688605 36.316391 23 67 266 184 mssvle 64.663708 38.348917 21 69 268 182 mssvlf 64.391665 37.8884 22 68 267 183 mssvlg 64.125802 37.434781 22 68 266 184 mssvlh 63.866197 36.988315 23 67 265 185 mssv2a 63.474986 36.308651 24 66 264 186 mssv2b 63.23172 35.881676 25 65 263 187 mssv2c 62.99498 35.462823 25 65 262 188 mssv2d 62.764841 35.052384 26 64 261 189 mssv2e 62.649172 37.001929 24 66 263 187 mssv2f 62.402686 36.583046 24 66 262 188 mssv2g 62.16279 36.172315 25 65 261 189 mssv2h 61.929558 35.770018 25 65 260 190 The first column denotes the ADV and MSSV designations. Note that the Unit I and Unit 2 geometries are symmetric along the West Road-Turbine Building Inlets (i.e., the 450 line from true North). The second and third columns denote the distances from the Unit 1 and 2 ADVs and MSSVs to the West Road and Turbine Building Inlets, designated dWR and dTB. The fourth through seventh columns denote the intake-to-source angles measured against true north for the Unit 1 and Unit 2 ADVs and MSSVs to the West Road Inlet and for the Unit 1 and Unit 2 ADVs and MSSVs to the Turbine Building Inlet.

The above ARCON96 inputs were then utilized in 72 ARCON96 executions to determine the time-dependent atmospheric dispersion coefficients.

1

APPENDIX 1 Sensitivity Study of Atmospheric Dispersion Coefficients for Releases from ADVs and MSSVs (B) Buoyant Plume Energetic Release Analysis Two normally shut ADVs (MS-3938-CV and MS-3939-CV) are connected to the main steam headers between the containment penetrations and the MSSVs. When opened, the ADVs exhaust part of the secondary steam flow to the atmosphere through separate vent enclosures, which extend from the 45' level through the roof of the Auxiliary Building. The ADVs are air operated, 5" globe valves that are made of carbon steel. Utilizing the Darcy equation with a resistance coefficient of 9.820, a limiting diameter of 3.760", and an exit diameter of 10.020", a steam exit velocity in excess of 67 m/sec can be calculated for a full-open ADV for all SG conditions prior to shutdown cooling.

Overpressure protection for the secondary side of the SGs and the main steam header is provided by sixteen spring-loaded American Society of Mechanical Engineers Code MSSVs (MS-3992-RV through MS-4007-RV). The safety valves automatically relieve the secondary plant steam pressure to the atmosphere through separate vent enclosures, which extend from the 27' level up through the roof of the Auxiliary Building. The valves are set to sequentially open two at a time, when the main steam header pressure exceeds 935 psig. Utilizing the Darcy equation with a resistance coefficient of 1.020, a limiting diameter of 4.520", and an exit diameter of 15.250", a steam exit velocity in excess of 52 m/sec can be calculated for a full-open MSSV.

From the met tower meteorological data documented in Enclosure 9 to Reference 1, the average and 95th percentile wind velocities for the years 1991-1998 were calculated and are presented in Table 5B below.

Table 5B 1Om and 60m Average and 95th Percentile Velocities in m/sec 10m 10M 60m 60m lOm vave 10M v95 60m vave 60m v95 vave Deviation v95 Deviation vave Deviation v95 Deviation 1991 3.00 -0.12 5.80 0.67 4.89 0.22 8.60 0.90 1992 2.98 0.18 5.90 0.13 4.85 0.73 8.60 0.90 1993 3.01 -0.27 6.20 -1.51 4.91 -0.03 9.20 -1.50 1994 3.07 -1.16 5.90 0.13 5.06 -1.93 8.80 0.10 1995 3.02 -0.41 5.80 0.67 4.85 0.73 8.50 1.30 1996 3.02 -0.41 6.20 -1.51 4.92 -0.16 9.10 -1.10 1997 3.00 -0.12 5.90 0.13 4.97 -0.79 8.90 -0.30 1998 2.84 2.28 5.70 1.22 4.81 1.23 8.90 -0.30 Average 2.99 5.92 4.91 8.82 Std Dev 0.07 1 0.18 1 0.08 0.25 Per Regulatory Guide 1.194 "Atmospheric Relative Concentrations for Control Room Radiological Habitability Assessments at Nuclear Power Plants" Regulatory Position C.6, "In lieu of mechanistically addressing the amount of buoyant plume rise associated with energetic releases from steam release valves or ADVs, the ground level X/Q value calculated with ARCON96 (on the basis of the physical height of the release point) may be reduced by a factor of

5. This reduction may be taken only if (1) the release point is uncapped and vertically oriented and (2) the time-dependent vertical velocity exceeds the 95th percentile wind speed (at the release point height) by a factor of 5."

2

APPENDIX 1 Sensitivity Study of Atmospheric Dispersion Coefficients for Releases from ADVs and MSSVs Furthermore:

"In order to credit these adjustments, the applicant or licensee must be able to demonstrate that the assumed buoyancy or vertical velocity of the effluent will be maintained throughout the time intervals that plume rise is credited."

Based on the 60m 95th percentile wind velocity data presented in Table 5B, the worst-case five times velocity limit is 44.1 m/sec. While all of the full open ADV exit velocity values exceed this limit, the ADVs may be throttled by the operators to control the cool down rate. Thus, the vertical velocity of the ADV effluent may not be maintained throughout the ADV discharge time intervals, and thus plume rise is not credited for ADV discharges. The MSSV exit velocity values also exceed this limit, but the MSSVs are spring-loaded and are not able to be throttled. Thus, the atmospheric dispersion coefficients for the MSSVs will be reduced by a factor of 5 per Regulatory Position C.6 of Regulatory Guide 1.194.

(C) Atmospheric Dispersion Coefficient Results Table 5C Adjusted Unit 1 ADV & MSSV ,/Qs (sec/m3)

West Road 0-2 HR 2-8 HR 8-24 HR 1-4 DAY 4-30 DAY advl 1.29E-03 1.03E-03 4.47E-04 3.30E-04 2.27E-04 adv2 1.40E-03 1.08E-03 4.70E-04 3.47E-04 2.43E-04 mssvla 2.56E-04 2.04E-04 8.88E-05 6.48E-05 4.50E-05 mssvl b 2.58E-04 2.06E-04 8.94E-05 6.54E-05 4.52E-05 mssvlc 2.58E-04 2.08E-04 9.04E-05 6.58E-05 4.64E-05 mssvld 2.60E-04 2.1OE-04 9.1OE-05 6.62E-05 4.70E-05 mssvle 2.62E-04 2.08E-04 9.06E-05 6.68E-05 4.68E-05 mssvlf 2.70E-04 2.10E-04 9.12E-05 6.72E-05 4.72E-05 mssvlg 2.74E-04 2.10E-04 9.16E-05 6.76E-05 4.76E-05 mssvlh 2.76E-04 2.12E-04 9.24E-05 6.80E-05 4.80E-05 mssv2a 2.78E-04 2.14E-04 9.40E-05 6.92E-05 4.82E-05 mssv2b 2.80E-04 2.16E-04 9.50E-05 6.90E-05 4.84E-05 mssv2c 2.80E-04 2.18E-04 9.58E-05 6.96E-05 4.86E-05 mssv2d 2.80E-04 2.20E-04 9.66E-05 6.98E-05 4.88E-05 mssv2e 2.80E-04 2.20E-04 9.66E-05 7.06E-05 4.92E-05 mssv2f 2.82E-04 2.22E-04 9.76E-05 7.1OE-05 4.94E-05 mssv2g 2.82E-04 2.24E-04 9.86E-05 7.12E-05 4.96E-05 mssv2h 2.82E-04 2.26E-04 9.92E-05 7.18E-05 5.02E-05 3

APPENDIX 1 Sensitivity Study of Atmospheric Dispersion Coefficients for Releases from ADVs and MSSVs Table 5D Adiusted Unit 2 ADV & MSSV i/Qs (sec/m3)

West Road 0-2 HR 2-8 HR 8-24 HR 1-4 DAY 4-30 DAY advl 1.24E-03 8.72E-04 3.42E-04 2.29E-04 1.51E-04 adv2 1.34E-03 9.27E-04 3.74E-04 2.45E-04 1.65E-04 mssvla 2.48E-04 1.75E-04 7.02E-05 4.64E-05 3.06E-05 mssvlb 2.48E-04 1.76E-04 7.06E-05 4.68E-05 3.10E-05 mssvlc 2.50E-04 1.79E-04 7.16E-05 4.78E-05 3.18E-05 mssvld 2.52E-04 1.80E-04 7.24E-05 4.80E-05 3.20E-05 mssvle 2.50E-04 1.77E-04 7.08E-05 4.68E-05 3.12E-05 mssvlf 2.54E-04 1.80E-04 7.26E-05 4.80E-05 3.22E-05 mssvlg 2.58E-04 1.80E-04 7.30E-05 4.84E-05 3.24E-05 mssvl h 2.60E-04 1.83E-04 7.42E-05 4.86E-05 3.28E-05 mssv2a 2.66E-04 1.86E-04 7.54E-05 5.OOE-05 3.32E-05 mssv2b 2.68E-04 1.89E-04 7.66E-05 5.1OE-05 3.38E-05 mssv2c 2.70E-04 1.91 E-04 7.70E-05 5.18E-05 3.42E-05 mssv2d 2.72E-04 1.93E-04 7.80E-05 5.26E-05 3.48E-05 mssv2e 2.70E-04 1.92E-04 7.70E-05 5.16E-05 3.42E-05 mssv2f 2.72E-04 1.93E-04 7.78E-05 5.20E-05 3.46E-05 mssv2g 2.74E-04 1.96E-04 7.88E-05 5.30E-05 3.52E-05 mssv2h 2.74E-04 1.98E-04 7.94E-05 5.34E-05 3.54E-05 Table 5E Adjusted Unit 1 ADV & MSSV X/Qs (seclm3)

Turbine Buildinig 0-2 HR 2-8 HR 8-24 HR 1-4 DAY 4-30 DAY advl 3.40E-03 2.52E-03 9.69E-04 7.18E-04 5.70E-04 adv2 3.56E-03 2.70E-03 1.01E-03 7.85E-04 6.18E-04 mssvla 7.24E-04 5.36E-04 2.06E-04 1.55E-04 1.22E-04 mssvlb 7.42E-04 5.46E-04 2.10E-04 1.57E-04 1.25E-04 mssvl c 7.54E-04 5.62E-04 2.14E-04 1.63E-04 1.27E-04 mssvld 7.62E-04 5.78E-04 2.16E-04 1.68E-04 1.31E-04 mssvle 6.96E-04 5.24E-04 2.OOE-04 1.51E-04 1.19E-04 mssvlf 7.16E-04 5.36E-04 2.04E-04 1.55E-04 1.23E-04 mssvl*g 7.38E-04 5.46E-04 2.08E-04 1.59E-04 1.25E-04 mssvl h 7.52E-04 5.62E-04 2.1OE-04 1.65E-04 1.27E-04 mssv2a 7.66E-04 5.88E-04 2.16E-04 1.73E-04 1.34E-04 mssv2b 7.84E-04 6.00E-04 2.22E-04 1.79E-04 1.37E-04 mssv2c 8.14E-04 6.12E-04 2.26E-04 1.84E-04 1.40E-04 mssv2d 8.28E-04 6.30E-04 2.30E-04 1.89E-04 1.46E-04 mssv2e 7.56E-04 5.70E-04 2.10E-04 1.71E-04 1.30E-04 mssv2f 7.62E-04 5.88E-04 2.14E-04 1.74E-04 1.35E-04 mssv2g 7.72E-04 6.02E-04 2.22E-04 1.80E-04 1.38E-04 mssv2h 7.96E-04 6.14E-04 2.28E-04 1.85E-04 1.41E-04 4

APPENDIX 1 Sensitivity Study of Atmospheric Dispersion Coefficients for Releases from ADVs and MSSVs Table 5F Adjusted Unit 2 ADV & MSSV X/Qs (sec/m3)

Turbine Buildin9g 0-2 HR 2-8 HR 8-24 HR 1-4 DAY 4-30 DAY advIl 3.49E-03 2.94E-03 1.19E-03 8.60E-04 6.94E-04 adv2 3.68E-03 3.17E-03 1.28E-03 9.56E-04 7.63E-04 mssvl a 7.52E-04 6.28E-04 2.54E-04 1.86E-04 1.50E-04 mssvl b 7.66E-04 6.42E-04 2.60E-04 1.90E-04 1.53E-04 mssvlc 7.74E-04 6.62E-04 2.66E-04 1.97E-04 1.59E-04 mssvld 7.78E-04 6.84E-04 2.72E-04 2.04E-04 1.63E-04 mssvl e 7.1OE-04 6.20E-04 2.46E-04 1.82E-04 1.47E-04 mssvlf 7.42E-04 6.28E-04 2.54E-04 1.87E-04 1.51 E-04 mssv1g 7.62E-04 6.44E-04 2.62E-04 1.94E-04 1.55E-04 mssvl h 7.72E-04 6.62E-04 2.68E-04 2.OOE-04 1.60E-04 mssv2a 7.82E-04 6.90E-04 2.76E-04 2.08E-04 1.66E-04 mssv2b 8.12E-04 7.OOE-04 2.84E-04 2.14E-04 1.71E-04 mssv2c 8.38E-04 7.14E-04 2.90E-04 2.20E-04 1.76E-04 mssv2d 8.48E-04 7.32E-04 2.98E-04 2.26E-04 1.82E-04 mssv2e 7.74E-04 6.66E-04 2.70E-04 2.04E-04 1.63E-04 mssv2f 7.78E-04 6.86E-04 2.76E-04 2.10E-04 1.68E-04 mssv2g 7.86E-04 7.02E-04 2.84E-04 2.14E-04 1.73E-04 mssv2h 8.26E-04 7.1OE-04 2.90E-04 2.20E-04 1.77E-04 The ADV-to-West Road and ADV-to-Turbine Building atmospheric dispersion coefficients generated in to Reference 1 bound all of the individual ADV and MSSV atmospheric dispersion coefficients that are documented above. Thus, use of the atmospheric dispersion coefficients that were generated in Enclosure 9 to Reference 1 are conservative for both ADV and MSSV effluents. Thus, all effluents released by the ADVs and MSSVs are conservatively assumed to be released by the relevant ADV.

REFERENCE:

(1) Letter from Mr. B. S. Montgomery (CCNPP) to Document Control Desk, dated November 3, 2005, License Amendment Request: Revision to Accident Source Term and Associated Technical Specifications 5

APPENDIX 2 Evaluation of Release from the RWT to Atmosphere Calvert Cliffs Nuclear Power Plant, Inc.

March 22, 2007

APPENDIX 2 Evaluation of Release from the RWT to Atmosphere Per Reference 1, reduction in release activity by dilution or holdup within buildings or by engineered safety feature ventilation filtration systems, may be credited where applicable. In order to determine the amount of refueling water tank (RWT) release to the atmosphere, the breathing of the RWT through the vent due to diurnal temperature variations was calculated. The breathing was determined by performing a 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months transient thermal analysis of the RWT using the computer code TSAP and assuming RWT atmospheric temperature characteristics based on the maximum solar load during summer solstice and a minimum ambient nighttime temperature. This methodology results in an average 4.2 cfm leakrate from the RWT to the environment. This evaluation is described below.

A. INPUT DATA Table A. 1 lists the absorbed solar heat load on the top of the RWT, which is the product of the solar heat flux on the top of the RWT, the RWT top area from Table A.3 Item 03, and the RWT tank emissivity from Table A.3 Item 04. The data is used in the TSAP input file RWT3A.DAT.

Table A.2 lists the absorbed solar heat load on the side of the RWT, which is the product of the solar heat flux on the side of the RWT, the RWT side area from Table A.3 Item 03 (L4*D4), and the RWT tank emissivity from Table A.3 Item 04. The data is used in the TSAP input file RWT3A.DAT.

Table A.3 lists the thermodynamic inputs for the transient TSAP analysis. The data is used in the TSAP input file RWT3A.DAT.

TABLE A. 1 TEMPORAL SOLAR HEAT LOAD ON TOP OF RWT TIME SOLAR FLUX AREA HEAT LOAD HOURS BTU/HR/FT 2 FT 2 EMISSIVITY BTU/HR 1 3 1352.652 0.6 2434.774 2 40 1352.652 0.6 32463.648 3 97 1352.652 0.6 78724.346 4 153 1352.652 0.6 124173.454 5 201 1352.652 0.6 163129.831 6 238 1352.652 0.6 193158.706 7 260 1352.652 0.6 211013.712 8 267 1352.652 0.6 216694.850 9 260 1352.652 0.6 211013.712 10 238 1352.652 0.6 193158.706 11 201 1352.652 0.6 163129.831 12 153 1352.652 0.6 124173.454 13 97 1352.652 0.6 78724.346 14 40 1352.652 0.6 32463.648 15 3 1352.652 0.6 2434.774 16 0 1352.652 0.6 0.000 17 0 1352.652 0.6 0.000 18 0 1352.652 0.6 0.000 19 0 1352.652 0.6 0.000 20 0 1352.652 0.6 0.000 21 0 1352.652 0.6 0.000 I

APPENDIX 2 Evaluation of Release from the RWT to Atmosphere TABLE A. 1 TEMPORAL SOLAR HEAT LOAD ON TOP OF RVVT TIME SOLAR FLUX AREA HEAT LOAD HOURS BTU/HR/FT 2 FT 2 EMISSIVITY BTU/HR 22 0 1352.652 0.6 0.000 23 0 1352.652 0.6 0.000 24 0 1352.652 0.6 0.000 TABLE A.2 TEMPORAL SOLAR HEAT LOAD ON SIDE OF RWT TIME SOLAR FLUX AREA HEAT LOAD HOURS BTU/HR/FT 2 FT 2 EMISSIVITY BTU/HR 1 20 1598.746 0.6 19184.952 2 151 1598.746 0.6 144846.388 3 207 1598.746 0.6 198564.253 4 216 1598.746 0.6 207197.482 5 192 1598.746 0.6 184175.539 6 145 1598.746 0.6 139090.902 7 81 1598.746 0.6 77699.056 8 41 1598.746 0.6 39329.152 9 81 1598.746 0.6 77699.056 10 145 1598.746 0.6 139090.902 11 192 1598.746 0.6 184175.539 12 216 1598.746 0.6 207197.482 13 207 1598.746 0.6 198564.253 14 151 1598.746 0.6 144846.388 15 20 1598.746 0.6 19184.952 16 0 1598.746 0.6 0.000 17 0 1598.746 0.6 0.000 18 0 1598.746 0.6 0.000 19 0 1598.746 0.6 0.000 20 0 1598.746 0.6 0.000 21 0 1598.746 0.6 0.000 22 0 1598.746 0.6 0.000 23 0 1598.746 0.6 0.000 24 0 1598.746 0.6 0.000 Table A.3 - TSAP Input Values (01) Nodes:

1 = Outside RWT Air

2 = RWT Tank Top

3 = RWT Tank Side

4 = Inside RWT Air

5 = Bottom Water and Tank Side Adjacent to Water 2

APPENDIX 2 Evaluation of Release from the RWT to Atmosphere (02) Temperatures:

Node 1 = 95°F for 15 daylight hours and 75°F for 9 night hours

Node 5 = 85°F for all 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months (03) Volumes and Dimensions:

V2 = i * (D2/2)**2 * (3/16") * (1'/12") = 21.13519 FT 3

" D2=41.5FT 2

A2 = ;- *(D2/2)**2 = 1352.652 FT 2

A3 = *L4 *D4 = 5022.609 FT

V3 =

D4 * {(3/16")*(246.5") + (7/32")*(83") + (9/32")*(83") + (1 1/32")*(50")}

3

(1.FT2/144.IN2) = 94.98102 FT

V4 = (389807 GAL) / (7.4805 GAL/CF) = 52109.75 FT 3

D4=41.5FT

L4 = V4/{ ir *(D4/2)**2}= 38.524 FT 2

A4 =A3 + 2.

A5= 7727.913 FT 3 3

V5 = (20237 GAL) / (7.4805 GAL/FT ) = 2705.30 FT

" D5=41.5FT

L5 = V2/{ ir *(D2/2)**2} = 2.000 FT 2

" A5 = )r *(D2/2)**2 = 1352.652 FT (04) RWT Tank Properties

(a) Material = SS304

(b) Emissivity = e = 0.6

(c) Cp = 0.11 BTU/LBM-F 3

(d) p =4881bm/ft

(e) k = 8.53 BTU/HR-FT-F @ 100°F (05) Air Properties

" (a) CP = 0.240 BTU/LBM-F

(b) P = 0.071 lbm/ft3

(c) Pr=0.72

(d) Emissivity = a = 0.41

L=3.6

V4 / A4 = 24.275 Mean Beam Length SPH 20 = 0.94915 psia @ 100°F & 100% Humidity SPH 2 oL = 1.5678 atm-ft Sa (H20) = 0.38 SPC02=0.0003 atm

Pc 02L=0.007283 atm-ft

a (C02) = 0.03

a = 0.41 (06) Radiative Heat Transfer Coefficients

(a) Tank top to outside air:

o G21R = A2

g *(0.1714E-08 BTU/HR-R4 -FT 2) = 1.391 1E-06 BTU/HR-R4

(b) Tank side to outside air:

o G31R = A3

6 *(0.1714E-08 BTU/HR-R4 -FT 2) = 5.1653E-06 BTU/HR-R4 3

APPENDIX 2 Evaluation of Release from the RWT to Atmosphere

(c) Tank top to inside air:

o G24R=A2

c

a *(0.1714E-08 BTU/HR-R4-FT 2)=5.7034E-07 BTU/HR-R4

(d) Tank side to inside air:

o G34R=A3* c

a *(0.1714E-08 BTU/HR-R4 -FT 2)=2.1178E-06 BTU/HR-R4 (07) Specific Heats:

" (a) CppV2 =(0.11 BTU/LBM-F)*(488 LBM/FT 3)*(21.13519 FT 3)=1134.537 BTU/F

(b) CpPV3 =(0.11 BTU/LBM-F)*(488 LBM!FT 3)*(94.98102 FT 3)=5098.581 BTU/F

(c) CppV 4=(0.24 BTU/LBM-F)*(0.071 LBM/FT 3)*(52109.75 FT 3)=887.9501BTU/F (08) Heat Transfer Coefficients:

(a) G23 = (8.53 BTU/HR-FT-F) * (i* *41.5'*.015625') / (0.5*(41.5'+38.524'))

= 0.434287 BTU/HR-F

(b) G35 = (8.53 BTU/HR-FT-F) * (ir *41.5'*.028083') / (0.5*(38.524'+2.0'))

= 1.541375 BTU/HR-F

(c) G24 =h

A2 =48.83413 (A TF)° 2 BTU/HR-F T2>T4 o Heated plate facing downward: h = 0.61 (A Tc/L 2)0 2 W/m 2-C o zl=(3.41214 BTU/HR)/(W) o z2=0.3048 m/ft o z3=1.8F/C o L=0.9

D2 = 37.35' = 11.38428 m o h = 0.61 (A Tv/z3/L 2) 0 2*zl*z22/z3 = 0.0361025 (A TF)°.2 BTU/HR-FT 2-F o G24 = h

A2 = 280.0920 (A TF) 0 313 BTU/HR-F T4>T2 o Cooled plate facing downward: h = 1.43 (A Tc)"333 W/m2-C o h = 1.43 (A TF /z3)"33 3 *zl*z22 /z3 = 0.207069 (A TF) 0 333 BTU/HR-FT 2-F

(d) G45 = h

A2 = 48.83413 (A TF)° 2 BTU/HR-F T4>T5 o Cooled plate facing upward: h = 0.61 (A Tc/L 2)0 2 W/m 2-C o L=0.9

D2 = 37.35' = 11.38428 m o h = 0.61 (A TF/z3/L 2)°2*zl*z2 2/z3 = 0.0361025 (A TF)0 2 BTU/HR-FT 2-F o G45 = h

A2 = 280.0920 (A TF) 0 3 33 BTU/HR-F T5>T4 o Heated plate facing upward: h = 1.43 (A Tc)"33 3 W/m2-C o h = 1.43 (A TF /z3). 3 33 *zl *z2 2 /z3 = 0.207069 (A TF)0 "333 BTU/HR-FT 2-F o Gr=g*13*AT*L3/v 2 = 1.314964E+11 o Pr =Cp* P/k = 0.6788 o T=1000 F = 559.67°R o A T=100 F o f=1/T = 1.786767E-03/R 2

o g=9.80665 m/sec o Cp=1.0392 J/gm-C 3 o p =642.3 gm/m o pu =.02848 gm/m-sec o k=0.04360 W/m-C o v=fllp=4.434E-05 m2/sec o L=0.9 *D2 = 11.38428 m o Gr*Pr = 8.9260E+10 --- Turbulent flow

(e) G21 = h

A2 = 48.83413 (A TF)° 2 BTU/HR-F TI>T2 4

APPENDIX 2 Evaluation of Release from the RWT to Atmosphere o G21 = h

A2 = 280.0920 (A TF)°-"' BTU/HR-F T2>T1 o See d above.

(f) G34 = h

A3 = 690.9259 (A TF)°.33 3 BTU/HR-F o Vertical plane or cylinder: h=0.95 (A Tc)0 .3 33 W/m2 -C o h=0.95 (A TF/z3) 033 ' 3 zl*z22/z3 = 0.137563 (A TF) 0 .33 3 BTU/HR-FT 2 -F o Gr g*fl*AT*L3/v 2 = 1.442900E+11 o Pr= Cp* P/k=0.6788 o T=1000 F = 559.67°R o A T=10 0 F o 8=1/T = 1.786767E-03/R 2

o g=9.80665 m/sec o Cp=1.0392 J/gm-C 3 42 o p=6 .3 gm/mr o p =.02848 gm/m-sec o k=0.04360 W/m-C o v= p / p =4.434E-05 m2/sec o L=L4 =38.524 FT = 11.74212 m o Gr*Pr =9.7944E+10 --- Turbulent flow

(g) G31 = h

A3 = 690.9259 (A TF) 0 33 3 BTU/HR-F o Vertical plane or cylinder: h=0.95 (A Tc) 0 3 33 W/m2 -C o h=0.95 (A TF/z3) 03" 33 zl*z2 2/z3 = 0.137563 (A TF) 0 333 BTU/HR-FT 2-F o See f above.

B. TECHNICAL ASSUMPTIONS (01) The external atmosphere is at the maximum temperature of 95°F for the fifteen hours of daylight and 75'F for the nine hours of night during the summer months. This is a conservative assumption.

(02) The RWT water temperature is at a constant temperature of 85°F during the entire 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period during the summer months. This is consistent with the control room logs and is conservative.

(03) All of the release from the RWT atmosphere to the external atmosphere is due to diurnal temperature variations.

C. METHOD OF ANALYSIS In order to determine the amount of RWT release to the atmosphere, the breathing of the RWT through the vent due to diurnal temperature variations was calculated. The breathing was determined by performing a 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months transient thermal analysis of the RWT using the computer code TSAP and assuming RWT atmospheric temperature characteristics based on the maximum solar load during summer solstice and a minimum ambient nighttime temperature.

The Thermal System Analysis Program (TSAP) is a general purpose transient multi-dimensional heat transfer program, which can be used to analyze conduction, convection, and radiation heat transfer phenomena in a lumped-parameter thermal network approach. The lumped-parameter thermal network concept solves the nodal temperatures of a thermal network by solving the energy conservation equations as governed by the first law of thermodynamics. This is accomplished by transforming these equations into a set of coupled simultaneous algebraic equations using the finite-difference numerical technique and by solving these equations by the successive point relaxation method.

5

APPENDIX 2 Evaluation of Release from the RWT to Atmosphere D. CALCULATIONS Employing the inputs of Tables A. 1 - A.3, a TSAP input file was constructed and is listed in Section E.

The resulting TSAP execution is printed in Section F. A maximum RWT air temperature of 130.7294'F is attained at time 5.6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , 120.9836°F at 8.8 hrs, 130.3470' at 12.3 hrs, and 77.3871'F at 24.0 hrs.

(Note that time step 1 corresponds to dawn.) From these temperature swings, the leakage from the RWT can be calculated as A VI = (52109.75 ft 3) * {(459.67+130.7294)/(459.67+77.3871)-1.} = 5175.714 cfd

= 3.59 cfm A V2 = (52109.75 ft3 ) * {(459.67+130.3470)/(459.67+120.9836)-1.} 840.302 cfd

= 0.58 cfm AV = AV1 + AV2 4.1778 cfm E. TSAP INPUT DECK C:\JERRY\TSAP\RWT3A.DAT GEG TSAP INPUT / 5 NODE / TRANSIENT NODE DATA

-1,95.,0.,0.

2,76.,1134.537,0.

3,77.,5098.581,0.

4,80.,887.9501,0.

-5,85.,0.,0.

ENDD COND DATA 1,1,2,0.

2,2,4,0.

3,4,5,0.

4,1,3,0.

5,3,4,0.

6,2,3,0.434287 7,3,5,1.541375

-8,1,2,1.3911 E-06

-9,1,3,5.1653E-06

-10,2,4,5.7034E-07

-11,3,4,2.1178E-06 ENDD CNTL DATA TSTP,.01, BETA,0.5, EROR,0.01, DCNV,0.005, NLOP,999, TSTA,0.,

TIME,0.,

6

APPENDIX 2 Evaluation of Release from the RWT to Atmosphere TEND,24.,

IPRT,10, ISTA, 1, IVAB, 1, RLAX, 1.0, ENDD ARRY DATA 1 1,2,3,4,5,6,7,8,9,10,11,12,13,14,15,16,17,18,19,20,21,22,23,24,END 2 002434.774,032463.648,078724.346,124173.454,163129.831,193158.706, 211013.712,216694.850,211013.712,193158.706,163129.831,124173.454, 078724.346,032463.648,002434.774,0.,0.,0.,0.,0.,0.,0.,0.,0., END 3 019184.952,144846.388,198564.253,207197.482,184175.539,139090.902, 077699.056,039329.152,077699.056,139090.902,184175.539,207197.482, 198564.253,144846.388,019184.952,0.,0.,0.,0.,0.,0.,0.,0.,0., END 4 95.,95.,95.,95.,95.,95.,95.,95.,95.,95.,95.,95.,95.,95.,95.,75.,

75.,75.,75.,75.,75.,75.,75.,75.,END ENDD VAB1 DATA C TANK TOP-EXTERNAL AIR CONVECTION TABSS=ABS(T1 -T2)

IF(T2.GT.T1) G1=280.0920*(TABSS)**0.333 IF(T1 .GE.T2) G1 =048.8341*(TABSS)**0.2 C TANK TOP-INTERNAL AIR CONVECTION TABSS=ABS(T2-T4)

IF(T2.GT.T4) G2=48.8341*(TABSS)**0.2 IF(T2.LE.T4) G2=280.0920*(TABSS)**0.333 C INTERNAL AIR - WATER CONVECTION TABSS=ABS(T4-T5)

IF(T5.GT.T4) G3=280.0920*(TABSS)**0.333 IF(T4.GE.T5) G3=048.8341*(TABSS)**0.2 C TANK SIDE-EXTERNAL AIR CONVECTION TABSS=ABS(T1-T3)

G4=690.9259*(TABSS)**0.333 C TANK SIDE-INTERNAL AIR CONVECTION TABSS=ABS(T3-T4)

G5=690.9259*(TABSS)**0.333 C RWT TOP HEAT SOURCE Q2=TERPL1 (Al ,A2,TIME,TIME,1.,1.)

C RWT SIDE HEAT SOURCE Q3=TERPL1 (Al ,A3,TIME,TIME,1.,1.)

C OUTSIDE AIR TEMPERATURE Ti =TERPL1 (Al ,A4,TIME,TIME, 1., 1.)

ENDD VAB2 DATA ENDD VAB3 DATA ENDD USER DATA ENDD ENDD 7

APPENDIX 2 Evaluation of Release from the RWT to Atmosphere F. TSAP OUTPUT FILE The file name is: RWT3B.DAT 7:53 am, Friday, January 13, 1995

$$$$$$$$$$$$$ BEGIN INPUT LISTING $$$$$$$$$$$$$$$

C:\JERRY\TSAP\RWT3B. DAT GEG TSAP INPUT / 5 NODE / TRANSIENT NODE DATA

-1,95.,0.,0.

2,76.,1134.537,0.

3,77.,5098.581,0.

4,80.,887.9501,0.

-5,85.,0.,0.

ENDD COND DATA 1,1,2,0.

2,2,4,0.

3,4,5,0.

4,1,3,0.

5,3,4,0.

6,2,3,0.434287 7,3,5,1.541375

-8,1,2,1.3911 E-06

-9,1,3,5.1653E-06

-10,2,4,5.7034E-07

-11,3,4,2.1178E-06 ENDD CNTL DATA TSTP,.01, BETA,0.5, EROR,0.01, DCNV,0.005, NLOP,999, TSTA,0.,

TIME,0.,

TEND,24.,

IPRT,10, ISTA,1, IVAB,1, RLAX, 1.0, ENDD ARRY DATA 1 1,2,3,4,5,6,7,8,9,10,11,12,13,14,15,16,17,18,19,20,21,22,23,24,END 2 002434.774,032463.648,078724.346,124173.454,163129.831,193158.706, 211013.712,216694.850,211013.712,193158.706,163129.831,124173.454, 078724.346,032463.648,002434.774,0.,0.,0.,0.,0.,0.,0.,0.,0.,END 3 019184.952,144846.388,198564.253,207197.482,184175.539,139090.902, 8

APPENDIX 2 Evaluation of Release from the RWT to Atmosphere 077699.056,039329.152,077699.056,139090.902,184175.539,207197.482, 198564.253,144846.388,019184.952,0., 0.,0.,0.,0.,0.,0.,0.,0., END 4 95.,95.,95.,95.,95.,95.,95.,95.,95.,95.,95.,95.,95.,95.,95.,75.,

75.,75.,75.,75.,75.,75.,75.,75.,END ENDD VAB1 DATA C TANK TOP-EXTERNAL AIR CONVECTION TABSS=ABS(T1 -T2)

IF(T2.GT.T1) G1 =280.0920*(TABSS)**0.333 IF(T1.GE.T2) G1=048.8341*(TABSS)**0.2 C TANK TOP-INTERNAL AIR CONVECTION TABSS=ABS(T2-T4)

IF(T2.GT.T4) G2=48.8341*(TABSS)**0.2 IF(T2.LE.T4) G2=280.0920*(TABSS)**0.333 C INTERNAL AIR - WATER CONVECTION TABSS=ABS(T4-T5)

IF(T5.GT.T4) G3=280.0920*(TABSS)**0.333 IF(T4.GE.T5) G3=048.8341*(TABSS)**0.2 C TANK SIDE-EXTERNAL AIR CONVECTION TABSS=ABS(T1-T3)

G4=690.9259*(TABSS)**0.333 C TANK SIDE-INTERNAL AIR CONVECTION TABSS=ABS(T3-T4)

G5=690.9259*(TABSS)**0.333 C RWT TOP HEAT SOURCE Q2=TERPL1 (Al,A2,TIME,TIME,1.,1.)

C RWT SIDE HEAT SOURCE Q3=TERPL1 (Al ,A3,TIME,TIME, 1., 1.)

C OUTSIDE AIR TEMPERATURE TI =TERPL1 (Al ,A4,TIME,TIME, 1., 1.)

ENDD VAB2 DATA ENDD VAB3 DATA ENDD USER DATA ENDD ENDD

$$$$$$$$$$$$$ END INPUT LISTING $$$$$$$$$$$$$$$

,TSAP PRE-PROCESSOR IN PROGRESS....

END OF NODE DATA PROCESSING.NDTOTL= 5 NCON= 1600 END OF CONDUCTOR DATA PROCESSING.NCTOTL= 11 END OF CONSTANT DATA PROCESSING END OF ARRAY DATA PROCESSING.NARRY= 4 END OF VAB1 DATA PROCESSING END OF VAB2 DATA PROCESSING END OF VAB3 DATA PROCESSING END OF USER DATA PROCESSING END OF PRE-PROCESSOR C:\JERRY\TSAP\RWT3B. DAT 9

APPENDIX 2 Evaluation of Release from the RWT to Atmosphere GEG TSAP INPUT / 5 NODE / TRANSIE

- T* TIME= .0000 EROR= .0100 LOOP= 1

( 1)= 95.0000 ( 2)= 76.0000( 3)= 77.0000( 4)= 80.0000 ( 5)= 85.0000

- T* TIME= .1000 EROR= .0000 LOOP= 2

( 1)= 95.0000( 2)= 77.9508( 3)= 79.1088( 4)= 79.6412 ( 5)= 85.0000

- T* TIME= .2000 EROR= .0001 LOOP= 2

( 1)= 95.0000 ( 2)= 79.6400( 3)= 80.9680( 4)= 79.9268 ( 5)= 85.0000

- T* TIME= .3000 EROR= .0002 LOOP= 2

( 1)= 95.0000 ( 2)= 81.1333( 3)= 82.5858( 4)= 80.5660 ( 5)= 85.0000

- T* TIME= .4000 EROR= .0002 LOOP= 2

( 1)= 95.0000 ( 2)= 82.4792( 3)= 84.0038( 4)= 81.4031 ( 5)= 85.0000

- T* TIME= .5000 EROR= .0002 LOOP= 2

( 1)= 95.0000 ( 2)= 83.6991( 3)= 85.2581( 4)= 82.3350 ( 5)= 85.0000

- T* TIME= .6000 EROR= .0002 LOOP= 2

( **T*

1)= 95.0000( 2)= 84.8094( 3)= 86.3757(

TIME= .7000 EROR=

4)= 83.2956 ( 5)= 85.0000

.0002 LOOP= 2

( 1)= 95.0000 ( 2)= 85.8224( 3)= 87.3782( 4)= 84.2469 ( 5)= 85.0000

- T* TIME= .8000 EROR= .0002 LOOP= 2

( 1)= 95.0000 ( 2)= 86.7488( 3)= 88.2821( 4)= 85.1721 ( 5)= 85.0000

- T* TIME= .9000 EROR= .0002 LOOP= 2

( 1)= 95.0000 ( 2)= 87.5981( 3)= 89.1016( 4)= 86.0720 ( 5)= 85.0000

- T* TIME= 1.0000 EROR= .0002 LOOP= 2

( 1)= 95.0000 ( 2)= 88.3779( 3)= 89.8478( 4)= 86.9345 ( 5)= 85.0000

- T* TIME= 1.1000 EROR= .0002 LOOP= 2

( 1)= 95.0000 ( 2)= 89.2217( 3)= 90.6475( 4)= 87.7627 ( 5)= 85.0000

- T* TIME= 1.2000 EROR= .0002 LOOP= 2

( 1)= 95.0000 ( 2)= 90.2434( 3)= 91.6055( 4)= 88.6063 ( 5)= 85.0000

- T* TIME= 1.3000 EROR= .0002 LOOP= 2

( 1)= 95.0000 ( 2)= 91.4227( 3)= 92.7031( 4)= 89.5059 ( 5)= 85.0000

- T* TIME= 1.4000 EROR= .0002 LOOP= 2

( 1)= 95.0000 ( 2)= 92.7427( 3)= 93.9274( 4)= 90.4858 ( 5)= 85.0000

- T* TIME= 1.5000 EROR= .0003 LOOP= 2

( 1)= 95.0000 ( 2)= 94.1896( 3)= 95.2720( 4)= 91.5605 ( 5)= 85.0000

- T* TIME= 1.6000 EROR= .0004 LOOP= 2

( 1)= 95.0000 ( 2)= 95.7500( 3)= 96.7310( 4)= 92.7383 ( 5)= 85.0000

- T* TIME= 1.7000 EROR= .0003 LOOP= 2

( 1)= 95.0000 ( 2)= 97.3824( 3)= 98.2821( 4)= 94.0196 ( 5)= 85.0000

- T* TIME= 1.8000 EROR= .0003 LOOP= 2

( 1)= 95.0000 ( 2)= 99.0533( 3)= 99.9092( 4)= 95.3973 ( 5)= 85.0000

- T* TIME= 1.9000 EROR= .0004 LOOP= 2

( 1)= 95.0000 ( 2)= 100.7451( 3)= 101.5989( 4)= 96.8614 ( 5)= 85.0000

- T* TIME= 2.0000 EROR= .0004 LOOP= 2

( 1)= 95.0000 ( 2)= 102.4460( 3)= 103.3400( 4)= 98.4009 ( 5)= 85.0000

- T* TIME= 2.1000 EROR= .0004 LOOP= 2

( 1)= 95.0000 ( 2)= 104.2145( 3)= 105.0560( 4)= 99.9988 ( 5)= 85.0000

- T* TIME= 2.2000 EROR= .0004 LOOP= 2

( 1)= 95.0000 ( 2)= 106.0936( 3)= 106.6849( 4)= 101.6204 ( 5)= 85.0000

- T* TIME= 2.3000 EROR= .0004 LOOP= 2

( 1)= 95.0000 ( 2)= 108.0544( 3)= 108.2373( 4)= 103.2383 ( 5)= 85.0000 10

APPENDIX 2 Evaluation of Release from the RWT to Atmosphere

- T* TIME= 2.4000 EROR= .0004 LOOP= 2

( 1)= 95.0000 ( 2)= 110.0718( 3)= 109.7219 ( 4)= 104.8361 ( 5)= 85.0000

- T* TIME= 2.5000 EROR= .0004 LOOP= 2

( 1)= 95.0000( 2)= 112.1249( 3)= 111.1454( 4)= 106.4042 ( 5)= 85.0000

- T* TIME= 2.6000 EROR= .0004 LOOP= 2

( 1)= 95.0000( 2)= 114.1967( 3)= 112.5135( 4)= 107.9373 ( 5)= 85.0000

- T* TIME= 2.7000 EROR= .0004 LOOP= 2

( 1)= 95.0000 ( 2)= 116.2736( 3)= 113.8311( 4)= 109.4327 S5)= 85.0000

- T* TIME= 2.8000 EROR= .0004 LOOP= 2

( 1)= 95.0000( 2)= 118.3452( 3)= 115.1031( 4)= 110.8898 ( 5)= 85.0000

- T* TIME= 2.9000 EROR= .0004 LOOP= 2

( 1)= 95.0000 ( 2)= 120.4031( 3)= 116.3330( 4)= 112.3093 ( 5)= 85.0000

- T* TIME= 3.0000 EROR= .0004 LOOP= 2

( 1)= 95.0000( 2)= 122.4417( 3)= 117.5242( 4)= 113.6917 ( 5)= 85.0000

- T* TIME= 3.1000 EROR= .0004 LOOP= 2

( 1)= 95.0000 ( 2)= 124.4535( 3)= 118.6382( 4)= 115.0345 ( 5)= 85.0000

- T* TIME= 3.2000 EROR= .0004 LOOP= 2

( 1)= 95.0000 ( 2)= 126.4326( 3)= 119.6458( 4)= 116.3215 ( 5)= 85.0000

- T* TIME= 3.3000 EROR= .0002 LOOP= 2

( 1)= 95.0000 ( 2)= 128.3775( 3)= 120.5608( 4)= 117.5432 ( 5)= 85.0000

- T* TIME= 3.4000 EROR= .0003 LOOP= 2

( 1)= 95.0000 ( 2)= 130.2875( 3)= 121.3948( 4)= 118.6970 ( 5)= 85.0000

- T* TIME= 3.5000 EROR= .0003 LOOP= 2

( 1)= 95.0000 ( 2)= 132.1620( 3)= 122.1573( 4)= 119.7841 ( 5)= 85.0000

- T* TIME= 3.6000 EROR= .0002 LOOP= 2

( 1)= 95.0000 ( 2)= 134.0013( 3)= 122.8569( 4)= 120.8084 ( 5)= 85.0000

- T* TIME= 3.7000 EROR= .0002 LOOP= 2

( 1)= 95.0000 ( 2)= 135.8060( 3)= 123.5004( 4)= 121.7744 ( 5)= 85.0000

- T* TIME= 3.8000 EROR= .0002 LOOP= 2

( 1)= 95.0000 ( 2)= 137.5773( 3)= 124.0943( 4)= 122.6876 ( 5)= 85.0000

- T* TIME= 3.9000 EROR= .0002 LOOP= 2

( 1)= 95.0000 ( 2)= 139.3163 ( 3)= 124.6439 ( 4)= 123.5533 ( 5)= 85.0000

- T* TIME= 4.0000 EROR= .0002 LOOP= 2

( 1)= 95.0000 ( 2)= 141.0247( 3)= 125.1541( 4)= 124.3769 ( 5)= 85.0000

- T* TIME= 4.1000 EROR= .0002 LOOP= 2

( 1)= 95.0000 ( 2)= 142.6780( 3)= 125.5998( 4)= 125.1611 ( 5)= 85.0000

- T* TIME= 4.2000 EROR= .0002 LOOP= 2

( 1)= 95.0000 ( 2)= 144.2598 ( 3)= 125.9614 ( 4)= 125.9001 ( 5)= 85.0000

- T* TIME= 4.3000 EROR= .0002 LOOP= 2

( 1)= 95.0000 ( 2)= 145.7823 ( 3)= 126.2496 ( 4)= 126.5947 ( 5)= 85.0000

- T* TIME= 4.4000 EROR= .0002 LOOP= 2

( 1)= 95.0000 ( 2)= 147.2554( 3)= 126.4758( 4)= 127.2331 ( 5)= 85.0000

- T* TIME= 4.5000 EROR= .0001 LOOP= 2

( 1)= 95.0000 ( 2)= 148.6855( 3)= 126.6495( 4)= 127.8080 ( 5)= 85.0000

- T* TIME= 4.6000 EROR= .0001 LOOP= 2

( 1)= 95.0000 ( 2)= 150.0779 ( 3)= 126.7779 ( 4)= 128.3204 ( 5)= 85.0000

- T* TIME= 4.7000 EROR= .0001 LOOP= 2

( 1)= 95.0000 ( 2)= 151.4368 ( 3)= 126.8666 ( 4)= 128.7743 ( 5)= 85.0000

- T* TIME= 4.8000 EROR= .0001 LOOP= 2

( 1)= 95.0000 ( 2)= 152.7660( 3)= 126.9203( 4)= 129.1749 ( 5)= 85.0000 11

APPENDIX 2 Evaluation of Release from the RWT to Atmosphere

- T* TIME= 4.9000 EROR= .0001 LOOP= 2

( 1)= 95.0000 ( 2)= 154.0682( 3)= 126.9431( 4)= 129.5276 ( 5)= 85.0000

- T* TIME= 5.0000 EROR= .0001 LOOP= 2

( 1)= 95.0000 ( 2)= 155.3464( 3)= 126.9384( 4)= 129.8380 ( 5)= 85.0000

- T* TIME= 5.1000 EROR= .0001 LOOP= 2

( 1)= 95.0000 ( 2)= 156.5668( 3)= 126.8889 ( 4)= 130.1081 ( 5)= 85.0000

- T* TIME= 5.2000 EROR= .0001 LOOP= 2

( 1)= 95.0000 ( 2)= 157.7084( 3)= 126.7816( 4)= 130.3311 ( 5)= 85.0000

- T* TIME= 5.3000 EROR= .0001 LOOP= 2

( 1)= 95.0000 ( 2)= 158.7880 ( 3)= 126.6247 ( 4)= 130.5033 ( 5)= 85.0000

- T* TIME= 5.4000 EROR= .0001 LOOP= 2

( 1)= 95.0000 ( 2)= 159.8181( 3)= 126.4243( 4)= 130.6249 ( 5)= 85.0000

- T* TIME= 5.5000 EROR= .0001 LOOP= 2

( 1)= 95.0000 ( 2)= 160.8080( 3)= 126.1861( 4)= 130.6990 ( 5)= 85.0000

- T* TIME= 5.6000 EROR= .0001 LOOP= 2

( 1)= 95.0000 ( 2)= 161.7653( 3)= 125.9143( 4)= 130.7294 ( 5)= 85.0000

- T* TIME= 5.7000 EROR= .0000 LOOP= 2

( 1)= 95.0000 ( 2)= 162.6953( 3)= 125.6126 ( 4)= 130.7205 ( 5)= 85.0000

- T* TIME= 5.8000 EROR= .0000 LOOP= 2

( 1)= 95.0000 ( 2)= 163.6025( 3)= 125.2846 ( 4)= 130.6765 ( 5)= 85.0000

- T* TIME= 5.9000 EROR= .0001 LOOP= 2

( 1)= 95.0000 ( 2)= 164.4908 ( 3)= 124.9329 ( 4)= 130.6008 ( 5)= 85.0000

- T* TIME= 6.0000 EROR= .0001 LOOP= 2

( 1)= 95.0000 ( 2)= 165.3627( 3)= 124.5601( 4)= 130.4973 ( 5)= 85.0000

- T* TIME= 6.1000 EROR= .0001 LOOP= 2

( 1)= 95.0000 ( 2)= 166.1722( 3)= 124.1536( 4)= 130.3662 ( 5)= 85.0000

- T* TIME= 6.2000 EROR= .0001 LOOP= 2

( 1)= 95.0000 ( 2)= 166.8898( 3)= 123.7034( 4)= 130.1996 ( 5)= 85.0000

- T* TIME= 6.3000 EROR= .0001 LOOP= 2

( 1)= 95.0000 ( 2)= 167.5383( 3)= 123.2154( 4)= 129.9932 ( 5)= 85.0000

- T* TIME= 6.4001 EROR= .0001 LOOP= 2

( 1)= 95.0000 ( 2)= 168.1345 ( 3)= 122.6937 ( 4)= 129.7469 ( 5)= 85.0000

- T* TIME= 6.5001 EROR= .0001 LOOP= 2

( 1)= 95.0000 ( 2)= 168.6902 ( 3)= 122.1419 ( 4)= 129.4630 ( 5)= 85.0000

- T* TIME= 6.6001 EROR= .0001 LOOP= 2

( 1)= 95.0000 ( 2)= 169.2147( 3)= 121.5632 ( 4)= 129.1445 ( 5)= 85.0000

- T* TIME= 6.7001 EROR= .0001 LOOP= 2

( 1)= 95.0000 ( 2)= 169.7147 ( 3)= 120.9600 ( 4)= 128.7947 ( 5)= 85.0000

- T* TIME= 6.8001 EROR= .0001 LOOP= 2

( 1)= 95.0000 ( 2)= 170.1952( 3)= 120.3346 ( 4)= 128.4168 ( 5)= 85.0000

- T* TIME= 6.9001 EROR= .0001 LOOP= 2

( 1)= 95.0000 ( 2)= 170.6605( 3)= 119.6891 ( 4)= 128.0138 ( 5)= 85.0000

- T* TIME= 7.0001 EROR= .0001 LOOP= 2

( 1)= 95.0000( 2)= 171.1132( 3)= 119.0255( 4)= 127.5883 ( 5)= 85.0000

- T* TIME= 7.1001 EROR= .0002 LOOP= 2

( 1)= 95.0000( 2)= 171.5071 ( 3)= 118.3662( 4)= 127.1439 ( 5)= 85.0000

- T* TIME= 7.2001 EROR= .0002 LOOP= 2

( 1)= 95.0000( 2)= 171.8143( 3)= 117.7291( 4)= 126.6881 ( 5)= 85.0000

- T* TIME= 7.3001 EROR= .0001 LOOP= 2

( 1)= 95.0000 ( 2)= 172.0588( 3)= 117.1089( 4)= 126.2268 ( 5)= 85.0000 12

APPENDIX 2 Evaluation of Release from the RWT to Atmosphere

- T* TIME= 7.4001 EROR= .0002 LOOP= 2

( 1)= 95.0000 ( 2)= 172.2574( 3)= 116.5023 ( 4)= 125.7637 ( 5)= 85.0000

- T* TIME= 7.5001 EROR= .0002 LOOP= 2

( 1)= 95.0000 ( 2)= 172.4229( 3)= 115.9063 ( 4)= 125.3008 ( 5)= 85.0000

- T* TIME= 7.6001 EROR= .0002 LOOP= 2

( 1)= 95.0000 ( 2)= 172.5645( 3)= 115.3184 ( 4)= 124.8391 ( 5)= 85.0000

- T* TIME= 7.7001 EROR= .0001 LOOP= 2

( 1)= 95.0000 ( 2)= 172.6893( 3)= 114.7371 ( 4)= 124.3790 ( 5)= 85.0000

- T* TIME= 7.8001 EROR= .0002 LOOP= 2

( 1)= 95.0000( 2)= 172.8019( 3)= 114.1610( 4)= 123.9210 ( 5)= 85.0000

- T* TIME= 7.9001 EROR= .0002 LOOP= 2

( 1)= 95.0000 ( 2)= 172.9057 ( 3)= 113.5889 ( 4)= 123.4650 ( 5)= 85.0000

- T* TIME= 8.0001 EROR= .0001 LOOP= 2

( 1)= 95.0000 ( 2)= 173.0035( 3)= 113.0197 ( 4)= 123.0108 ( 5)= 85.0000

- T* TIME= 8.1001 EROR= .0001 LOOP= 2

( 1)= 95.0000 ( 2)= 173.0519( 3)= 112.5237 ( 4)= 122.5655 ( 5)= 85.0000

- T* TIME= 8.2001 EROR= .0002 LOOP= 2

( 1)= 95.0000( 2)= 173.0249( 3)= 112.1556( 4)= 122.1576 ( 5)= 85.0000

- T* TIME= 8.3001 EROR= .0001 LOOP= 2

( 1)= TIME=

95.0000 8.4001

( 2)= 172.9462( 3)= 111.8958( 4)= 121.8062 ( 5)= 85.0000

- T* EROR= .0001 LOOP= 2

( 1)= 95.0000 ( 2)= 172.8328( 3)= 111.7292( 4)= 121.5183 ( 5)= 85.0000

- T* TIME= 8.5001 EROR= .0001 LOOP= 2

( 1)= 95.0000 ( 2)= 172.6972( 3)= 111.6439( 4)= 121.2951 ( 5)= 85.0000

- T* TIME= 8.6001 EROR= .0001 LOOP= 2

( 1)= 95.0000( 2)= 172.5482( 3)= 111.6296( 4)= 121.1342 ( 5)= 85.0000

- T* TIME= 8.7001 EROR= .0000 LOOP= 2

( 1)= 95.0000 ( 2)= 172.3922( 3)= 111.6779( 4)= 121.0318 ( 5)= 85.0000

- T* TIME= 8.8001 EROR= .0001 LOOP= 2

( 1)= 95.0000 ( 2)= 172.2338 ( 3)= 111.7816( 4)= 120.9836 ( 5)= 85.0000

- T* TIME= 8.9001 EROR= .0000 LOOP= 2

( 1)= 95.0000 ( 2)= 172.0759( 3)= 111.9346( 4)= 120.9850 ( 5)= 85.0000

- T* TIME= 9.0001 EROR= .0001 LOOP= 2

( 1)= 95.0000 ( 2)= 171.9203( 3)= 112.1310( 4)= 121.0314 ( 5)= 85.0000

- T* TIME= 9.1001 EROR= .0000 LOOP= 2

( 1)= 95.0000( 2)= 171.7198( 3)= 112.3874( 4)= 121.1198 ( 5)= 85.0000

- T* TIME= 9.2001 EROR= .0001 LOOP= 2

( 1)= 95.0000 ( 2)= 171.4453( 3)= 112.7159( 4)= 121.2515 ( 5)= 85.0000

- T* TIME= 9.3001 EROR= .0001 LOOP= 2

( 1)= 95.0000( 2)= 171.1190( 3)= 113.1068( 4)= 121.4268 ( 5)= 85.0000

- T* TIME= 9.4001 EROR= .0001 LOOP= 2

( 1)= 95.0000 ( 2)= 170.7571( 3)= 113.5518( 4)= 121.6442 ( 5)= 85.0000

- T* TIME= 9.5001 EROR= .0001 LOOP= 2

( 1)= 95.0000 ( 2)= 170.3711( 3)= 114.0433( 4)= 121.9014 ( 5)= 85.0000

- T* TIME= 9.6001 EROR= .0001 LOOP= 2

( 1)= 95.0000 ( 2)= 169.9694 ( 3)= 114.5760( 4)= 122.1954 ( 5)= 85.0000

- T* TIME= 9.7001 EROR= .0001 LOOP= 2

( 1)= 95.0000 ( 2)= 169.5574 ( 3)= 115.1440( 4)= 122.5230 ( 5)= 85.0000

- T* TIME= 9.8001 EROR= .0001 LOOP= 2

( 1)= 95.0000 ( 2)= 169.1393 ( 3)= 115.7432( 4)= 122.8810 ( 5)= 85.0000 13

APPENDIX 2 Evaluation of Release from the RWT to Atmosphere

- T* TIME= 9.9001 EROR= .0001 LOOP= 2

( 1)= 95.0000 ( 2)= 168.7178 ( 3)= 116.3696( 4)= 123.2663 ( 5)= 85.0000

- T* TIME= 10.0001 EROR= .0001 LOOP= 2

( 1)= 95.0000 ( 2)= 168.2946( 3)= 117.0193( 4)= 123.6761 ( 5)= 85.0000

- T* TIME= 10.1001 EROR= .0001 LOOP= 2

( 1)= 95.0000( 2)= 167.8226( 3)= 117.6744( 4)= 124.1048 ( 5)= 85.0000

- T* TIME= 10.2001 EROR= .0001 LOOP= 2

( 1)= 95.0000 ( 2)= 167.2715 ( 3)= 118.3199 ( 4)= 124.5388 ( 5)= 85.0000

- T* TIME= 10.3001 EROR= .0002 LOOP= 2

( 1)= 95.0000 ( 2)= 166.6630 ( 3)= 118.9573 ( 4)= 124.9695 ( 5)= 85.0000

- T* TIME= 10.4001 EROR= .0001 LOOP= 2

( 1)= 95.0000 ( 2)= 166.0117( 3)= 119.5872( 4)= 125.3925 ( 5)= 85.0000

- T* TIME= 10.5001 EROR= .0001 LOOP= 2

( 1)= 95.0000 ( 2)= 165.3286( 3)= 120.2101( 4)= 125.8058 ( 5)= 85.0000

- T* TIME= 10.6001 EROR= .0001 LOOP= 2

( 1)= 95.0000 ( 2)= 164.6212 ( 3)= 120.8265 ( 4)= 126.2097 ( 5)= 85.0000

- T* TIME= 10.7001 EROR= .0001 LOOP= 2

( 1)= 95.0000 ( 2)= 163.8951( 3)= 121.4365( 4)= 126.6038 ( 5)= 85.0000

- T* TIME= 10.8002 EROR= .0001 LOOP= 2

( 1)= 95.0000 ( 2)= 163.1541( 3)= 122.0403( 4)= 126.9890 ( 5)= 85.0000

- T* TIME= 10.9002 EROR= .0001 LOOP= 2

( 1)= 95.0000 ( 2)= 162.4012 ( 3)= 122.6384 ( 4)= 127.3657 ( 5)= 85.0000

- T* TIME= 11.0002 EROR= .0001 LOOP= 2

( 1)= 95.0000 ( 2)= 161.6385 ( 3)= 123.2310 ( 4)= 127.7349 ( 5)= 85.0000

- T* TIME= 11.1002 EROR= .0002 LOOP= 2

( 1)= 95.0000 ( 2)= 160.8318( 3)= 123.7978( 4)= 128.0936 ( 5)= 85.0000

- T* TIME= 11.2002 EROR= .0001 LOOP= 2

( 1)= 95.0000 ( 2)= 159.9590( 3)= 124.3235( 4)= 128.4308 ( 5)= 85.00o0

- T* TIME= 11.3002 EROR= .0001 LOOP= 2

( 1)= 95.0000 ( 2)= 159.0354 ( 3)= 124.8137 ( 4)= 128.7399 ( 5)= 85.0000

- T* TIME= 11.4002 EROR= .0001 LOOP= 2

( 1)= 95.0000 ( 2)= 158.0721( 3)= 125.2732( 4)= 129.0184 ( 5)= 85.0000

- T* TIME= 11.5002 EROR= .0001 LOOP= 2

( 1)= 95.0000 ( 2)= 157.0769( 3)= 125.7056( 4)= 129.2672 ( 5)= 85.0000

- T* TIME= 11.6002 EROR= .0001 LOOP= 2

( 1)= 95.0000 ( 2)= 156.0557( 3)= 126.1138( 4)= 129.4877 ( 5)= 85.0000

- T* TIME= 11.7002 EROR= .0000 LOOP= 2

( 1)= 95.0000 ( 2)= 155.0125( 3)= 126.5007( 4)= 129.6817 ( 5)= 85.0000

- T* TIME= 11.8002 EROR= .0001 LOOP= 2

( 1)= 95.0000 ( 2)= 153.9503 ( 3)= 126.8690 ( 4)= 129.8521 ( 5)= 85.0000

- T* TIME= 11.9002 EROR= .0001 LOOP= 2

( 1)= 95.0000 ( 2)= 152.8718( 3)= 127.2202( 4)= 130.0008 ( 5)= 85.0000

- T* TIME= 12.0002 EROR= .0001 LOOP= 2

( 1)= 95.0000 ( 2)= 151.7784( 3)= 127.5569( 4)= 130.1299 ( 5)= 85.0000

- T* TIME= 12.1002 EROR= .0001 LOOP= 2

( 1)= 95.0000 ( 2)= 150.6453 ( 3)= 127.8508 ( 4)= 130.2379 ( 5)= 85.0000

- T* TIME= 12.2002 EROR= .0000 LOOP= 2

( 1)= 95.0000 ( 2)= 149.4561( 3)= 128.0813( 4)= 130.3127 ( 5)= 85.0000

- T* TIME= 12.3002 EROR= .0000 LOOP= 2

( 1)= 95.0000 ( 2)= 148.2209( 3)= 128.2578 ( 4)= 130.3470 ( 5)= 85.0000 14

APPENDIX 2 Evaluation of Release from the RWT to Atmosphere

- T* TIME= 12.4002 EROR= .0000 LOOP= 2

( 1)= 95.0000 ( 2)= 146.9478 ( 3)= 128.3878 ( 4)= 130.3386 ( 5)= 85.0000

- T* TIME= 12.5002 EROR= .0000 LOOP= 2

( 1)= 95.0000 ( 2)= 145.6418( 3)= 128.4779( 4)= 130.2886 ( 5)= 85.0000

- T* TIME= 12.6002 EROR= .0001 LOOP= 2

( 1)= 95.0000 ( 2)= 144.3071( 3)= 128.5327( 4)= 130.1991 ( 5)= 85.0000

- T* TIME= 12.7002 EROR= .0000 LOOP= 2

( 1)= 95.0000 ( 2)= 142.9462( 3)= 128.5570( 4)= 130.0735 ( 5)= 85.0000

- T* TIME= 12.8002 EROR= .0001 LOOP= 2

( 1)= 95.0000( 2)= 141.5616( 3)= 128.5543( 4)= 129.9147 ( 5)= 85.0000

- T* TIME= 12.9002 EROR= .0001 LOOP= 2

( 1)= 95.0000 ( 2)= 140.1546( 3)= 128.5280( 4)= 129.7258 ( 5)= 85.0000

- T* TIME= 13.0002 EROR= .0001 LOOP= 2

( 1)= 95.0000 ( 2)= 138.7266( 3)= 128.4812( 4)= 129.5096 ( 5)= 85.0000

- T* TIME= 13.1002 EROR= .0001 LOOP= 2

( 1)= 95.0000 ( 2)= 137.2748( 3)= 128.3745( 4)= 129.2650 ( 5)= 85.0000

- T* TIME= 13.2002 EROR= .0001 LOOP= 2

( 1)= 95.0000 ( 2)= 135.7975( 3)= 128.1779( 4)= 128.9792 (5)= 85.0000

- T* TIME= 13.3002 EROR= .0001 LOOP= 2

( 1)= 95.0000 ( 2)= 134.2951( 3)= 127.9052( 4)= 128.6434 ( 5)= 85.0000

- T* TIME= 13.4002 EROR= .0001 LOOP= 2

( 1)= 95.0000 ( 2)= 132.7678( 3)= 127.5670( 4)= 128.2556 ( 5)= 85.0000

- T* TIME= 13.5002 EROR= .0001 LOOP= 2

( 1)= 95.0000( 2)= 131.2152( 3)= 127.1725( 4)= 127.8162 ( 5)= 85.0000

- T* TIME= 13.6002 EROR= .0001 LOOP= 2

( 1)= 95.0000 ( 2)= 129.6371( 3)= 126.7288( 4)= 127.3279 ( 5)= 85.0000

- T* TIME= 13.7002 EROR= .0001 LOOP= 2

( 1)= 95.0000 ( 2)= 128.0330( 3)= 126.2421( 4)= 126.7944 ( 5)= 85.0000

- T* TIME= 13.8002 EROR= .0002 LOOP= 2

( 1)= 95.0000 ( 2)= 126.4024 ( 3)= 125.7177 ( 4)= 126.2194 ( 5)= 85.0000

- T* TIME= 13.9002 EROR= .0002 LOOP= 2

( 1)= 95.0000 ( 2)= 124.7490 ( 3)= 125.1597 ( 4)= 125.6012 ( 5)= 85.0000

- T* TIME= 14.0002 EROR= .0002 LOOP= 2

( 1)= 95.0000 ( 2)= 123.0893( 3)= 124.5716( 4)= 124.9226 ( 5)= 85.0000

- T* TIME= 14.1002 EROR= .0003 LOOP= 2

( 1)= 95.0000 ( 2)= 121.4910 ( 3)= 123.8887 ( 4)= 124.1821 ( 5)= 85.0000

- T* TIME= 14.2002 EROR= .0004 LOOP= 2

( 1)= 95.0000 ( 2)= 119.9941( 3)= 123.0615( 4)= 123.3737 ( 5)= 85.0000

- T* TIME= 14.3002 EROR= .0004 LOOP= 2

( 1)= 95.0000 ( 2)= 118.5668 ( 3)= 122.1100 ( 4)= 122.4957 ( 5)= 85.0000

- T* TIME= 14.4002 EROR= .0004 LOOP= 2

( 1)= 95.0000 ( 2)= 117.1844( 3)= 121.0510( 4)= 121.5496 ( 5)= 85.0000

- T* TIME= 14.5002 EROR= .0004 LOOP= 2

( 1)= 95.0000 ( 2)= 115.8280( 3)= 119.8987( 4)= 120.5380 ( 5)= 85.0000

- T* TIME= 14.6002 EROR= .0004 LOOP= 2

( 1)= 95.0000 ( 2)= 114.4833( 3)= 118.6641( 4)= 119.4639 ( 5)= 85.0000

- T* TIME= 14.7002 EROR= .0005 LOOP= 2

( 1)= 95.0000( 2)= 113.1393( 3)= 117.3568( 4)= 118.3309 ( 5)= 85.0000

- T* TIME= 14.8002 EROR= .0005 LOOP= 2

( 1)= 95.0000( 2)= 111.7879( 3)= 115.9846( 4)= 117.1423 ( 5)= 85.0000 15

APPENDIX 2 Evaluation of Release from the RWT to Atmosphere

- T* TIME= 14.9002 EROR= .0005 LOOP= 2

( 1)= 95.0000 ( 2)= 110.4229( 3)= 114.5533( 4)= 115.9007 ( 5)= 85.0000

- T* TIME= 15.0002 EROR= .0005 LOOP= 2

( 1)= 94.9950 ( 2)= 109.0396( 3)= 113.0680( 4)= 114.6092 ( 5)= 85.0000

- T* TIME= 15.1003 EROR= .0005 LOOP= 2

( 1)= 92.9950 ( 2)= 107.6000( 3)= 111.5240( 4)= 113.2686 ( 5)= 85.0000

- T* TIME= 15.2003 EROR= .0006 LOOP= 2 I.

( 1)= 90.9949( 2)= 106.0742( 3)= 109.9151( 4)= 111.8724 ( 5)= 85.0000

- T* TIME= 15.3003 EROR= .0006 LOOP= 2

( 1)= 88.9949 ( 2)= 104.4775(

- T* TIME=

3)= 108.2493( 4)= 110.4174 ( 5)= 85.0000 15.4003 EROR= .0006 LOOP= 2

( 1)= 86.9949 ( 2)= 102.8219( 3)= 106.5334( 4)= 108.9047 ( 5)= 85.0000

- T* TIME= 15.5003 EROR= .0006 LOOP= 2

( 1)= 84.9948 ( 2)= 101.1165( 3)= 104.7726( 4)= 107.3376 ( 5)= 85.0000

- T* TIME= 15.6003 EROR= .0007 LOOP= 2

( 1)= 82.9948 ( 2)= 99.3688( 3)= 102.9722( 4)= 105.7206 ( 5)= 85.0000

- T* TIME= 15.7003 EROR= .0007 LOOP= 2

( 1)= 80.9947( 2)= 97.5849( 3)= 101.1361( 4)= 104.0582 ( 5)= 85.0000

- T* TIME= 15.8003 EROR= .0007 LOOP= 2

( 1)= 78.9947 ( 2)= 95.7695( 3)= 99.2679 ( 4)= 102.3549 ( 5)= 85.0000

- T* TIME= 15.9003 EROR= .0007 LOOP= 2

( 1)= 76.9946 ( 2)= 93.9268( 3)= 97.3710 ( 4)= 100.6149 ( 5)= 85.0000

- T* TIME= 16.0003 EROR= .0007 LOOP= 2

( 1)= 75.0000 ( 2)= 92.0603( 3)= 95.4483 ( 4)= 98.8419 5)= 85.0000

- T* TIME= 16.1003 EROR= .0007 LOOP= 2

( 1)= 75.0000 ( 2)= 90.3264( 3)= 93.6232 ( 4)= 97.0563 5)= 85.0000

- T* TIME= 16.2003 EROR= .0007 LOOP= 2

( 1)= 75.0000 ( 2)= 88.8387( 3)= 91.9963 ( 4)= 95.3286 5)= 85.0000

- T* TIME= 16.3003 EROR= .0006 LOOP= 2

( 1)= 75.0000 ( 2)= 87.5443( 3)= 90.5380 ( 4)= 93.7041 5)= 85.0000

- T* TIME= 16.4003 EROR= .0005 LOOP= 2

( 1)= 75.0000 ( 2)= 86.4062( 3)= 89.2253 ( 4)= 92.2001 5)= 85.0000

- T* TIME= 16.5003 EROR= .0005 LOOP= 2

( 1)= 75.0000 ( 2)= 85.3979( 3)= 88.0402 ( 4)= 90.8193 5)= 85.0000

- T* TIME= 16.6003 EROR= .0004 LOOP= 2

( 1)= 75.0000 ( 2)= 84.4995( 3)= 86.9676 ( 4)= 89.5569 5)= 85.0000

- T* TIME= 16.7003 EROR= .0004 LOOP= 2

( 1)= 75.0000 ( 2)= 83.6949( 3)= 85.9953( 4)= 88.4052 5)= 85.0000

- T* TIME= 16.8003 EROR= .0004 LOOP= 2

( 1)= 75.0000 ( 2)= 82.9719( 3)= 85.1119( 4)= 87.3550 5)= 85.0000

- T* TIME= 16.9003 EROR= .0004 LOOP= 2

( 1)= 75.0000 ( 2)= 82.3199( 3)= 84.3085 ( 4)= 86.3973 5)= 85.0000

- T* TIME= 17.0003 EROR= .0003 LOOP= 2

( 1)= 75.0000 ( 2)= 81.7305( 3)= 83.5770 ( 4)= 85.5231 5)= 85.0000

- T* TIME= 17.1003 EROR= .0002 LOOP= 2

( 1)= 75.0000 ( 2)= 81.1964( 3)= 82.9099 ( 4)= 84.7244 5)= 85.0000

- T* TIME= 17.2003 EROR= .0002 LOOP= 2

( 1)= 75.0000 ( 2)= 80.7117( 3)= 82.3008 ( 4)= 84.0052 5)= 85.0000

- T* TIME= 17.3003 EROR= .0002 LOOP= 2

( 1)= 75.0000 ( 2)= 80.2720( 3)= 81.7451 ( 4)= 83.3648 5)= 85.0000 16

APPENDIX 2 Evaluation of Release from the RWT to Atmosphere

- T* TIME= 17.4003 EROR= .0002 LOOP= 2

( 1)= 75.0000 ( 2)= 79.8736( 3)= 81.2379 ( 4)= 82.7946 ( 5)= 85.0000

- T* TIME= 17.5003 EROR= .0002 LOOP= 2

( 1)= 75.0000 ( 2)= 79.5128( 3)= 80.7754 ( 4)= 82.2856 ( 5)= 85.0000

- T* TIME= 17.6003 EROR= .0001 LOOP= 2

( 1)= 75.0000 ( 2)= 79.1863( 3)= 80.3535 ( 4)= 81.8304 ( 5)= 85.0000

- T* TIME= 17.7003 EROR= .0001 LOOP= 2

( 1)= 75.0000 ( 2)= 78.8906( 3)= 79.9690 ( 4)= 81.4221 ( 5)= 85.0000

- T* TIME= 17.8003 EROR= .0001 LOOP= 2

( 1)= 75.0000 ( 2)= 78.6228( 3)= 79.6182 ( 4)= 81.0550 ( 5)= 85.0000

- T* TIME= 17.9003 EROR= .0001 LOOP= 2

( 1)= 75.0000 ( 2)= 78.3804( 3)= 79.2983 ( 4)= 80.7245 ( 5)= 85.0000

- T* TIME= 18.0003 EROR= .0001 LOOP= 2

( 1)= 75.0000 ( 2)= 78.1606( 3)= 79.0063 ( 4)= 80.4260 ( 5)= 85.0000

- T* TIME= 18.1003 EROR= .0001 LOOP= 2

( 1)= 75.0000 ( 2)= 77.9612( 3)= 78.7397 ( 4)= 80.1565 ( 5)= 85.0000

- T* TIME= 18.2003 EROR= .0001 LOOP= 2

( 1)= 75.0000 ( 2)= 77.7804( 3)= 78.4963 ( 4)= 79.9124 ( 5)= 85.0000

- T* TIME= 18.3003 EROR= .0001 LOOP= 2

( 1)= 75.0000 ( 2)= 77.6161( 3)= 78.2739 ( 4)= 79.6909 ( 5)= 85.0000

- T* TIME= 18.4003 EROR= .0001 LOOP= 2

( 1)= 75.0000 ( 2)= 77.4670( 3)= 78.0709( 4)= 79.4901 ( 5)= 85.0000

- T* TIME= 18.5003 EROR= .0000 LOOP= 2

( 1)= 75.0000 ( 2)= 77.3312( 3)= 77.8851( 4)= 79.3075 ( 5)= 85.0000

- T* TIME= 18.6003 EROR= .0001 LOOP= 2

( 1)= 75.0000 ( 2)= 77.2078( 3)= 77.7154 ( 4)= 79.1415 ( 5)= 85.0000

- T* TIME= 18.7003 EROR= .0001 LOOP= 2

( 1)= 75.0000 ( 2)= 77.0953( 3)= 77.5601 ( 4)= 78.9907 ( 5)= 85.0000

- T* TIME= 18.8003 EROR= .0001 LOOP= 2

( 1)= 75.0000 ( 2)= 76.9929( 3)= 77.4182 ( 4)= 78.8532 ( 5)= 85.0000

- T* TIME= 18.9003 EROR= .0001 LOOP= 2

( 1)= 75.0000 ( 2)= 76.8994( 3)= 77.2882 ( 4)= 78.7280 ( 5)= 85.0000

- T* TIME= 19.0003 EROR= .0001 LOOP= 2

( 1)= 75.0000 ( 2)= 76.8144( 3)= 77.1694 ( 4)= 78.6137 ( 5)= 85.0000

- T* TIME= 19.1003 EROR= .0001 LOOP= 2

( 1)= 75.0000 ( 2)= 76.7366( 3)= 77.0605 ( 4)= 78.5096 ( 5)= 85.0000

- T* TIME= 19.2003 EROR= .0001 LOOP= 2

( 1)= 75.0000 ( 2)= 76.6656( 3)= 76.9607 ( 4)= 78.4144 ( 5)= 85.0000

- T* TIME= 19.3003 EROR= .0000 LOOP= 2

( 1)= 75.0000 ( 2)= 76.6006( 3)= 76.8691 ( 4)= 78.3272 ( 5)= 85.0000

- T* TIME= 19.4003 EROR= .0001 LOOP= 2

( 1)= 75.0000 ( 2)= 76.5413( 3)= 76.7854 ( 4)= 78.2477 ( 5)= 85.0000

- T* TIME= 19.5004 EROR= .0001 LOOP= 2

( 1)= 75.0000 ( 2)= 76.4872( 3)= 76.7087 ( 4)= 78.1750 ( 5)= 85.0000

- T* TIME= 19.6004 EROR= .0001 LOOP= 2

( 1)= 75.0000 ( 2)= 76.4376( 3)= 76.6382 ( 4)= 78.1085 ( 5)= 85.0000

- T* TIME= 19.7004 EROR= .0000 LOOP= 2

( 1)= 75.0000 ( 2)= 76.3924( 3)= 76.5737 ( 4)= 78.0476 ( 5)= 85.0000

- T* TIME= 19.8004 EROR= .0000 LOOP= 2

( 1)= 75.0000 ( 2)= 76.3508( 3)= 76.5143 ( 4)= 77.9919 ( 5)= 85.0000 17

APPENDIX 2 Evaluation of Release from the RWT to Atmosphere

- T* TIME= 19.9004 EROR= .0000 LOOP= 2

( 1)= 75.0000 ( 2)= 76.3127( 3)= 76.4599 ( 4)= 77.9409 ( 5)= 85.0000

- T* TIME= 20.0004 EROR= .0048 LOOP= 1

( 1)= 75.0000 ( 2)= 76.2778( 3)= 76.4100 ( 4)= 77.8940 ( 5)= 85.0000

- T* TIME= 20.1004 EROR= .0044 LOOP= 1

( 1)= 75.0000 ( 2)= 76.2459( 3)= 76.3643( 4)= 77.8512 ( 5)= 85.0000

- T* TIME= 20.2004 EROR= .0040 LOOP= 1

( 1)= 75.0000 ( 2)= 76.2169( 3)= 76.3222( 4)= 77.8118 ( 5)= 85.0000

- T* TIME= 20.3004 EROR= .0037 LOOP= 1

( 1)= 75.0000 ( 2)= 76.1899( 3)= 76.2833 ( 4)= 77.7759 ( 5)= 85.0000

- T* TIME= 20.4004 EROR= .0034 LOOP= 1

( 1)= 75.0000 ( 2)= 76.1654( 3)= 76.2478 ( 4)= 77.7428 ( 5)= 85.0000

- T* TIME= 20.5004 EROR= .0032 LOOP= 1

( 1)= 75.0000 ( 2)= 76.1428( 3)= 76.2152 ( 4)= 77.7126 ( 5)= 85.0000

- T* TIME= 20.6004 EROR= .0029 LOOP= 1

( 1)= 75.0000 ( 2)= 76.1220( 3)= 76.1853 ( 4)= 77.6848 ( 5)= 85.0000

- T* TIME= 20.7004 EROR= .0026 LOOP= 1

( 1)= 75.0000 ( 2)= 76.1030( 3)= 76.1579 ( 4)= 77.6592 ( 5)= 85.0000

- T* TIME= 20.8004 EROR= .0024 LOOP= 1

( 1)= 75.0000 ( 2)= 76.0856( 3)= 76.1326 ( 4)= 77.6357 ( 5)= 85.0000

- T* TIME= 20.9004 EROR= .0023 LOOP= 1

( 1)= 75.0000 ( 2)= 76.0696( 3)= 76.1095( 4)= 77.6142 ( 5)= 85.0000

- T* TIME= 21.0004 EROR= .0021 LOOP= 1

( 1)= 75.0000 ( 2)= 76.0549( 3)= 76.0881 ( 4)= 77.5944 ( 5)= 85.0000

- T* TIME= 21.1004 EROR= .0019 LOOP= 1

( 1)= 75.0000( 2)= 76.0414( 3)= 76.0685 ( 4)= 77.5762 ( 5)= 85.0000

- T* TIME= 21.2004 EROR= .0017 LOOP= 1

( 1)= 75.0000 ( 2)= 76.0290( 3)= 76.0504 ( 4)= 77.5595 ( 5)= 85.0000

- T* TIME= 21.3004 EROR= .0016 LOOP= 1

( 1)= 75.0000 ( 2)= 76.0175( 3)= 76.0339 ( 4)= 77.5443 ( 5)= 85.0000

- T* TIME= 21.4004 EROR= .0015 LOOP= 1

( 1)= 75.0000 ( 2)= 76.0070( 3)= 76.0188 ( 4)= 77.5302 ( 5)= 85.0000

- T* TIME= 21.5004 EROR= .0013 LOOP= 1

( 1)= 75.0000 ( 2)= 75.9973( 3)= 76.0047 ( 4)= 77.5173 ( 5)= 85.0000

- T* TIME= 21.6004 EROR= .0012 LOOP= 1

( 1)= 75.0000 ( 2)= 75.9883( 3)= 75.9918 ( 4)= 77.5055 ( 5)= 85.0000

- T* TIME= 21.7004 EROR= .0012 LOOP= 1

( 1)= 75.0000 ( 2)= 75.9802( 3)= 75.9799 ( 4)= 77.4944 ( 5)= 85.0000

- T* TIME= 21.8004 EROR= .0010 LOOP= 1

( 1)= 75.0000 ( 2)= 75.9726( 3)= 75.9691( 4)= 77.4843 ( 5)= 85.0000

- T* TIME= 21.9004 EROR= .0009 LOOP= 1

( 1)= 75.0000 ( 2)= 75.9657( 3)= 75.9592( 4)= 77.4750 ( 5)= 85.0000

- T* TIME= 22.0004 EROR= .0009 LOOP= 1

( 1)= 75.0000 ( 2)= 75.9595( 3)= 75.9500 ( 4)= 77.4664 ( 5)= 85.0000

- T* TIME= 22.1004 EROR= .0008 LOOP= 1

( 1)= 75.0000 ( 2)= 75.9536( 3)= 75.9414 ( 4)= 77.4586 ( 5)= 85.0000

- T* TIME= 22.2004 EROR= .0008 LOOP= 1

( 1)= 75.0000 ( 2)= 75.9482( 3)= 75.9335 ( 4)= 77.4514 ( 5)= 85.0000

- T* TIME= 22.3004 EROR= .0007 LOOP= 1

( 1)= 75.0000 ( 2)= 75.9434( 3)= 75.9263 ( 4)= 77.4448 ( 5)= 85.0000 18

APPENDIX 2 Evaluation of Release from the RWT to Atmosphere

- T* TIME= 22.4004 EROR= .0007 LOOP= 1

( 1)= 75.0000 ( 2)= 75.9388( 3)= 75.9197 ( 4)= 77.4388 ( 5)= 85.0000

- T* TIME= 22.5004 EROR= .0006 LOOP= 1

( 1)= 75.0000 ( 2)= 75.9345( 3)= 75.9137 ( 4)= 77.4333 ( 5)= 85.0000

- T* TIME= 22.6004 EROR= .0005 LOOP= 1

( 1)= 75.0000 ( 2)= 75.9307( 3)= 75.9081 ( 4)= 77.4281 ( 5)= 85.0000

- T* TIME= 22.7004 EROR= .0005 LOOP= 1

( 1)= 75.0000 ( 2)= 75.9271( 3)= 75.9028 ( 4)= 77.4232 ( 5)= 85.0000

- T* TIME= 22.8004 EROR= .0004 LOOP= 1

( 1)= 75.0000 ( 2)= 75.9240( 3)= 75.8982 ( 4)= 77.4188 ( 5)= 85.0000

- T* TIME= 22.9004 EROR= .0004 LOOP= 1

( 1)= 75.0000 ( 2)= 75.9210( 3)= 75.8937 ( 4)= 77.4148 ( 5)= 85.0000

- T* TIME= 23.0004 EROR= .0004 LOOP= 1

( 1)= 75.0000 ( 2)= 75.9184( 3)= 75.8896 ( 4)= 77.4111 ( 5)= 85.0000

- T* TIME= 23.1004 EROR= .0004 LOOP= 1

( 1)= 75.0000 ( 2)= 75.9159( 3)= 75.8859 ( 4)= 77.4077 ( 5)= 85.0000

- T* TIME= 23.2004 EROR= .0003 LOOP= 1

( 1)= 75.0000 ( 2)= 75.9135( 3)= 75.8826 ( 4)= 77.4047 ( 5)= 85.0000

- T* TIME= 23.3004 EROR= .0004 LOOP= 1

( 1)= 75.0000( 2)= 75.9111( 3)= 75.8794 ( 4)= 77.4017 ( 5)= 85.0000

- T* TIME= 23.4004 EROR= .0002 LOOP= 1

( 1)= 75.0000 ( 2)= 75.9092( 3)= 75.8766 ( 4)= 77.3992 ( 5)= 85.0000

- T* TIME= 23.5004 EROR= .0003 LOOP= 1

( 1)= 75.0000 ( 2)= 75.9073( 3)= 75.8740( 4)= 77.3969 ( 5)= 85.0000

- T* TIME= 23.6004 EROR= .0002 LOOP= 1

( 1)= 75.0000 ( 2)= 75.9056( 3)= 75.8715( 4)= 77.3947 ( 5)= 85.0000

- T* TIME= 23.7004 EROR= .0002 LOOP= 1

( 1)= 75.0000 ( 2)= 75.9042( 3)= 75.8693 ( 4)= 77.3926 ( 5)= 85.0000

- T* TIME= 23.8004 EROR= .0002 LOOP= 1

( 1)= 75.0000 ( 2)= 75.9029( 3)= 75.8671 ( 4)= 77.3907 ( 5)= 85.0000

- T* TIME= 23.9005 EROR= .0002 LOOP= 1

( 1)= 75.0000 ( 2)= 75.9016( 3)= 75.8653 ( 4)= 77.3889 ( 5)= 85.0000

- T* TIME= 24.0000 EROR= .0002 LOOP= 1

( 1)= 75.0000 ( 2)= 75.9004( 3)= 75.8635 ( 4)= 77.3871 ( 5)= 85.0000

REFERENCE:

(1) Regulatory Guide 1.183, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, dated July 2000 19

APPENDIX 3 Sensitivity Study-of Carbon Filter Dose Model Calvert Cliffs Nuclear Power Plant, Inc.

March 22, 2007

APPENDIX 3 Sensitivity Study of Carbon Filter Dose Model The MicroShield model in Enclosure 1 of Reference 1 was a simple 24'"x33"x59" carbon filter, shielded at the bottom by 6" of air and 2' of concrete, with the dose point located 1" from the concrete on the filter centerline. To demonstrate the conservatism of the MicroShield model, a more detailed three-dimensional representation is constructed with MCNP. MCNP is a general-purpose Monte-Carlo N-Particle code that can be used for neutron, photon, electron, or coupled neutron/photon/electron transport. For this aplication it is used in photon-only mode.

Figure 1 shows the location of the Control Room Heating, Ventilation, and Air Conditioning (CR HVAC)

Equipment room on the 69' elevation relative to the Control Room on the 45' elevation. The CR HVAC Equipment Room extends above the north-east comer of .the Control Room by 25'6" in the E-W direction, and 39'7" in the N-S direction. For the MCNP model, the south west comer of the filter was conservatively placed in this location on the 69', which is as close to the center of the Control Room as is possible (worst case placement). The filter was assumed to have the same dimensions, composition, and source strength and energy spectrum as was utilized in MicroShield case MHFOOO.MS5 from Enclosure 1 of Reference 1. A 2" diameter spherical particle flux tally air cell was located 1" from the bottom of the 2' thick El. 69' concrete floor slab for consistency with the MicroShield model. In addition, cylindrical and Cartesian particle flux mesh tallies were placed at the level of the operators (assumed to be 6' tall) on the 45' elevation. The cylindrical tally was centered on the center line of the filter and used 1' radial ring tally bins. The Cartesian tally used the NW corner of the control room as its origin and used 1'xl' square tally bins. Standard American National Standards Institute (ANSI)/American Nuclear Society (ANS)-6.1.1-1977 dose conversion factors are used to convert photon flux to dose rate in rem/hr, with an additional factor of 1000 applied to the total source strength to produce dose rates in mremihr. Figure 2 shows a three-dimensional view of the resulting MCNP model and Figure 3 shows the locations of the dose tallies relative to the CR HVAC filter.

Figure 1 - E-W Cross-section showing location of CR HVAC Room relative to Control Room I

APPENDIX 3 Sensitivity Study of Carbon Filter Dose Model Figure 2 - MCNP model of CR HVAC filter in closest possible location to center of control room Filter - 69' El.

- Dose point at 1" 6' tall ring mesh tally 45' El.

Figure 3 - Dose tally locations in MCNP model Table I below provides a comparison of the MicroShield and MCNP dose results for the same filter medium and source, and identical dose point locations 1" from the concrete below the filter. As can be seen, the MicroShield result is within 1% of the MCNP result. This is within the range of uncertainty of the MCNP result, and therefore it can be concluded that the MicroShield results are essentially the same.

2

APPENDIX 3 Sensitivity Study of Carbon Filter Dose Model Thus backscatter is not a significant contributor for this dose location, which is very conservative, and the point-kernel method is considered acceptable for calculating dose at that location.

Table 1 Unadjusted Zero Decay Dose Rate at 1" from Concrete Below Filter Model Dose Rate (mrem/hr)

MicroShield Carbon Filter (MHFOOO.MS5) 2.906E+7 MCNP Carbon Filter 2.937E+7 +/-3%

An additional demonstration of the amount of margin present between the dose point location used in the MicroShield model and the actual location of the operators in the Control Room is provided by the 6' tall cylindrical mesh tally located at the 45' elevation. The results of this tally are provided in Figure 4 below. This shows that the dose to an operator standing on the 45' elevation in a worst case location directly below the filter would be significantly less than at the dose point used for the filter shine dose calculation in Enclosure I of Reference 1. This would be true regardless of whether the source was located on a filter medium composed of carbon, or the lower density fiberglass used for high efficiency particulate air filters. Thus, the dose location 1" below the 2' concrete slab below the filter used in the filter shine calculation is conservative regardless of where the filter is placed within the CR HVAC equipment room, or what filter medium the source is captured on.

Unadjusted Zero Decay Control Room HVAC Filter Shine Dose Rate (assumes 6' tall operator at 45' El. and worst case filter placement) 1.E+07

.9!

0 a

1.IE+06 I.0 0 0.0 5.0 10.0 15.0 20.0 25.0 Distance from CR HVAC Filter Centerline (feet)

Figure 4 - MCNP Cylindrical Mesh Tally Results MCNP Model CCNPP Control Room HVAC Filter Dose for AST c CELL CARDS 1 2 -0.001225 110 -120 210 -220 310 -320 500 imp:p=1024 $Control Room Air 3

APPENDIX 3 Sensitivity Study of Carbon Filter Dose Model 5 1 -2.35 110 -120 210 -220 305 -310 imp:p=1024 $Control Room Floor top 10cm 10 1 -2.35 110 -120 210 -220 300 -305 imp:p =l $Control Room Floor 15 1 -2.35 110 -120 210 -220 320 -322 imp: p=1024 $Control Room Ceiling bot 10cm 20 1 -2.35 110 -120 210 -220 322 -324 imp:p=256 $Control Room Ceiling 25 1 -2.35 110 -120 210 -220 324 -326 imp: p= 64 $Control Room Ceiling 30 1 -2.35 110 -120 210 -220 326 -328 imp: p=16 $Control Room Ceiling 35 1 -2.35 110 -120 210 -220 328 -329 imp: p=4 $Control Room Ceiling 40 1 -2.35 110 -120 210 -220 329 -330 imp:p =l $El 69' Floor 50 2 -0.001225 110 -120 210 -220 330 -340 400 imp:p=l $Cont Rm HVAC Eq Room Air 60 4 -2.25 -400 imp:p=l $Carbon Filter 70 2 -0.001225 -500 imp:p=1024 $Tally sphere below filter 80 0 -110:120:-210:220:-300:340 imp:p=0 $Outside world c SURFACE CARDS c X Planes 100 px -10 $10 cm into W. Wall 110 px 0 $Control Room West Wall (Line Md in middle) 120 px 1935.48 $Control Room East Wall (63'6" E-W) - dwg 62042sh0001 130 px 1945.48 $10 cm into E. Wall c Y Planes 200 py -10 $10 cm into wall 210 py 0 $Control Room North Wall 220 py 2743.2 $Control Room South Wall (90' N-S) - dwg 62042sh0001 c Z Planes 300 pz -76.2 $bottom of 2'6" thick slab for El 45' - dwg 61680sh0001 305 pz -10 $10 cm into floor slab for backscatter 310 pz 0 $CR floor at 45' Elevation - dwg 60218 320 pz 670.56 $bottom of 2' thick slab for El 69' - dwg 61685sh0001 322 pz 680 $splitting surface 324 pz 690 $splitting surface 326 pz 700 $splitting surface 328 pz 710 $splitting surface 329 pz 720 $splitting surface 330 pz 731.52 $CR HVAC floor at 69' Elevation - dwg 60218 340 pz 1000 $top of model c Filter c Farthest west filter can go is 26'6" from line La per dwg 60319sh0001 c Farthest souti h filter can go is 40'7" from line 17 per dwg 60319sh0001 c Filter 6" abo*ve floor c Same dimensioins used in CA06449 c

400 BOX 1158.24 1206.5 746.76 -83.82 0 0 0 -60.96 0 0 0 149.86 c Tally Sphere 500 s 1116.33 1176.02 668.02 2.5399 c DATA CARDS c

c MATERIAL CARDS ml 1001 -. 0132 $ concrete ANS 6.4.3-1991 p 107 8000 -1.171 $ d=2.35 g/cc 11000 -. 0456 $ d=l.70 g/cc for masonry 13000 -. 1072 14000 -. 7449 19000 -. 0451 20000 -. 1941 26000 -. 0287 c

m2 18000 -. 013 $ air d=0.001225 g/cc 7000 -. 755 8000 -. 232 c

m3 14000 1 $ Fiberglass (assume Si02) HEPA filter d=0.18 g/cc 8000 2 $ based on CA06260 Att. H info from Amer Air Filter c

m4 6000 1 $ Carbon filter c

c SOURCE SPECIFICATION MODE P SDEF CEL=60 x=Dl y=D2 z=D3 ERG=D4 AXS 0 0 1 SIl 1074.42 1158.24 4

APPENDIX 3 Sensitivity Study of Carbon Filter Dose Model SPI 0 1 S12 1145.54 1206.5 SP2 0 1 S13 746.76 896.62 SP3 0 1 c 20 group t=0 filter spectrum from CA06449 case MHFOOO.MS5 S14 L 0.015 0.02 0.03 0.04 0.05 0.06 0.08 0.1 0.15 0.2 0.3 0.4 0.5 0.6 0.8 1 1.5 2 3 4 SP4 D 1.58E+12 1.44E+15 2.90E+17 5.45E+15 2.57E+16 7.29E+15 3.39E+16 3.06E+16 1.74E+17 3.16E+17 2.75E+17 1.41E+18 2.96E+18 3.35E+18 6.74E+18 2.53E+18 1.91E+18 4.98E+17 2.90E+14 2.40E+13 c

c TALLY SPECIFICATION c Photon/cm^2-history mesh ring tally in control room centered below filter FMESH14:P GEOM=cyl ORIGIN= 1116.33 1176.02 0 IMESH=731.52 IINTS=24 JMESH=182.88 JINTS=1 KMESH= 1 KINTS= 1 OUT=ik c 1'xl'x6' grid photon/cm^2-history mesh tally in control room at 45' El.

FMESH24:P ORIGIN=0 0 0 IMESH=1920.24 IINTS=63 JMESH=2743 JINTS=90 KMESH=i82.88 KINTS=1 OUT=ij c Tally sphere in same location as Microshield dose point F34:p 70 $ (photons/cm^2-history)

C C Tally multipliers FM14 2.055E22 $ total photons/sec

1000 mrem/rem FM24 2.055E22 FM34 2.055E22 c photon flux to dose rate ratios {(rem*cmA2*s)/(hr*photon)}

c from MCNP4C Manual page H-6 ANSI/ANS-6.1.1-1977 deo LOG 0.01 0.03 0.05 0.07 0.1 0.15 0.2 0.25 0.3 0.35 0.4 0.45 0.5 0.55 0.6 0.65 0.7 0.8 1 1.4 1.8 2.2 2.6 2.8 3.25 3.75 4.25 4.75 5.0 5.25 5.75 6.25 6.75 7.5 9 11.0 13.0 15.0 df0 LOG 3.96-6 5.82-7 2.90-7 2.58-7 2.83-7 3.79-7 5.01-7 6.31-7 7.59-7 8.78-7 9.85-7 1.08-6 1.17-6 1.27-6 1.36-6 1.44-6 1.52-6 1.68-6 1.98-6 2.51-6 2.99-6 3.42-6 3.82-6 4.01-6 4.41-6 4.83-6 5.23-6 5.60-6 5.80-6 6.01-6 6.37-6 6.74-6 7.11-6 7.66-6 8.77-6 1.03-5 1.18-5 1.33-5 c

CUT:P 1.E+33 0.019 PRDMP 60 0 1 0 ctme 1440 print -85

REFERENCE:

(1) Letter from Mr. B. S. Montgomery (CCNPP) to Document Control Desk, dated November 3, 2005, License Amendment Request: Revision to Accident Source Term and Associated Technical Specifications 5

ATTACHMENT (2)

MARKED UP TECHNICAL SPECIFICATION PAGES 1.1-3 5.5-17 Calvert Cliffs Nuclear Power Plant, Inc.

March 22, 2007

4 Definitions 1.1 1.1 Definitions disintegration (in MeV) for isotopes, other than iodines, with half lives > 15 minutes, making up at least 95% of the total non-iodine activity in the coolant.

ENGINEERED SAFETY FEATURE The ESF RESPONSE TIME shall be that time interval (ESF) RESPONSE TIME from when the monitored parameter exceeds its ESF actuation setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays, where applicable. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and methodology for verification have been previously reviewed and approved by the NRC.

La The maximu 1._~llwwable containment lea kae te, La, shall b ).Qf containment air c. per O.1)-I&7o1 day at the calculated peak containment pressure (Pa) -. oo e LEAKAGE LEAKAGE shall be:

a. Identified LEAKAGE

1. LEAKAGE, such as that from pump seals or valve packing (except reactor coolant pump (RCP) seal leakoff), that is captured and conducted to collection systems or a sump or collecting tank;

2. LEAKAGE into the containment atmosphere from sources that are both specifically located and known either not to CALVERT CLIFFS - UNIT 1 1.1-3 Amendment No. 279 CALVERT CLIFFS - UNIT 2 Amendment No. 256

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.16 Containment Leakage Rate Testing Program A program shall be established to implement the leakage testing of the containment as required by 10 CFR 50.54(o) and 10 CFR Part 50, Appendix J, Option B. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program," dated September 1995, including errata, as modified by the following exceptions:

a. Nuclear Energy Institute (NEI) 94 1995, Section 9.2.3:

The first Unit 1 Type A test performed after the June 15, 1992 Type A test shall be performed no later than June 14, 2007.

b. Unit 1 is excepted from post-modification integrated leakage rate testing requirements associated with steam generator replacement.

c. Unit 2 is excepted from post-modification integrated leakage rate testing requirements associated with steam generator replacement.

The peak calculated containment internal pressure for the design basis loss-of-coolant accident, P, is 49.4 psig. The containment design pressure is 50 psig.

The maximum allowable containment leakage rate, L., shall b percent of containment air per day at Pa Leakage rate acceptance criteria are:

a. Containment leakage rate acceptance criterion is

1.0 La.

During the first unit startup following testing, in accordance with this program, the leakage rate acceptance criterion are

0.60 La for Types B and C tests and

0.75 L, for Type A tests.

CALVERT CLIFFS - UNIT 1 5.5-17 Amendment No. 278 CALVERT CLIFFS - UNIT 2 Amendment No. 255

v t e Letter Sequence Request
TAC:MC8845, (Open) TAC:MC8846, (Open)
Initiation Request Acceptance, Acceptance Administration Questions 22 December 2006 Answers 17 July 2007 Results Approval... Other:
MONTHYEAR2007-03-22 [Table View]

ML070870110

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