ML050950215

Response to Request for Additional Information Regarding Proposed Technical Specification Changes for Implementation of Alternate Source Term

ML050950215
Person / Time
Site:	Millstone
Issue date:	03/23/2005
From:	Matthews W Dominion Nuclear Connecticut
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
05-077
Download: ML050950215 (35)

Text

Dominion Nuclear Connecticut, Inc. Dominionf Millstone Power Station Rope Ferry Road Waterford, CT 06385 March 23, 2005 U. S. Nuclear Regulatory Commission Serial No.: 05-077 Attention: Document Control Desk NLOS/MAE Rev. 0 Washington, DC 20555 Docket No.: 50-423 License No.: NPF-49 DOMINION NUCLEAR CONNECTICUT. INC.

MILLSTONE POWER STATION UNIT 3 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION REGARDING PROPOSED TECHNICAL SPECIFICATION CHANGES FOR IMPLEMENTATION OF ALTERNATE SOURCE TERM In a letter dated May 27, 2004, Dominion Nuclear Connecticut, Inc. (DNC) requested an amendment in the form of changes to the Technical Specifications to Facility Operating License Number NPF-49 for Millstone Power Station Unit 3. The proposed changes were requested based on the radiological dose analysis margins obtained by using an alternate source term consistent with 10 CFR 50.67. In facsimiles dated August 20, 2004, and September 22, 2004, the NRC requested additional information to facilitate the technical review being conducted by the staff. In letters dated September 27, 2004, and October 20, 2004, DNC forwarded responses to the requested information.

In letters dated December 29, 2004 and February 9, 2005, the NRC forwarded two subsequent requests for additional information. Attachment 1 of this letter provides the response to the requests for additional information. Attachment 2 provides the marked-up original retyped pages, and Attachment 3 provides the retyped pages, as described in the RAI response.

The additional information provided in this letter does not affect the conclusions of the safety summary and significant hazards considerations discussion in the DNC letter dated May 27, 2004.

If you should have any questions regarding this submittal, please contact Mr. Paul R.

Willoughby at (804) 273-3572.

Very truly yours, William R. Matthews Senior Vice President - Nuclear Operations

Serial No.: 05-077 Docket No.: 50-423 Response to Request for Additional Information Page 2 of 3 Attachments: (3)

Commitments made in this letter: None.

cc: U. S. Nuclear Regulatory Commission Region I 475 Allendale Road King of Prussia, PA 19406-1415 Mr. G. F. Wunder Project Manager - Millstone Unit 3 U.S. Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike Mail Stop 08-B-1A Rockville, MD 20852-2738 Mr. S. M. Schneider NRC Senior Resident Inspector Millstone Power Station Director Bureau of Air Management Monitoring & Radiation Division Department of Environmental Protection 79 Elm Street Hartford, CT 06106-5127

Serial No.: 05-077 Docket No.: 50-423 Response to Request for Additional Information Page 3 of 3 COMMONWEALTH OF VIRGINIA )

COUNTY OF HENRICO The foregoing document was acknowledged before me, in and for the County and Commonwealth aforesaid, today by William R. Matthews, who is Senior Vice President -

Nuclear Operations of Dominion Nuclear Connecticut, Inc. He has affirmed before me that he is duly authorized to execute and file the foregoing document in behalf of that company, and that the statements in the document are true to the best of his knowledge and belief.

Acknowledged before me this A3 day of ________, 2005.

My Commission Expires: l6z73 / 1O8.

Notary Public (SEAL)

Serial No.: 05-077 Docket No.: 50-423 ATTACHMENT 1 PROPOSED TECHNICAL SPECIFICATION CHANGES IMPLEMENTATION OF ALTERNATE SOURCE TERM RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION DOMINION NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION UNIT 3

Serial No.: 05-077 Docket No.: 50-423 Response to Request for Additional Information Attachment 1 Page 1 of 26 PROPOSED TECHNICAL SPECIFICATION CHANGES IMPLEMENTATION OF ALTERNATE SOURCE TERM RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION In a letter dated May 27, 2004, Dominion Nuclear Connecticut, Inc. (DNC) requested an amendment in the form of changes to the Technical Specifications to Facility Operating License Number NPF-49 for Millstone Power Station Unit 3. The proposed changes were requested based on the radiological dose analysis margins obtained by using an alternate source term consistent with 10 CFR 50.67. In facsimiles dated August 20, 2004, and September 22, 2004, the NRC requested additional information to facilitate the technical review being conducted by the staff. In letters dated September 27, 2004, and October 20, 2004, DNC forwarded responses to the requested information.

In letters dated December 29, 2004 and February 9, 2005, the NRC forwarded two subsequent requests for additional information.

Below is the response to the two requests for additional information:

Resnonse To RAI In Letter Dated December 29,2004 NRC Question Regulatory Guide (RG) 1.52 establishes the criterion for penetration in the laboratory testing of engineered safety feature filter systems. In Revision 2 of RG 1.52, the allowable penetration for a 2-inch bed filter is less than 1 percent. In Revision 3 of RG 1.52, some relaxation was allowed and the penetration criterion for a 2-inch bed during a laboratory test was increased to less than 2.5 percent, based in part on maintaining a safety factor of 2. The current Millstone Power Station, Unit No. 3 technical specification (TS) for the control room emergency ventilation filter allows 2.5 percent penetration and is in compliance with RG 1.52, Revision 3. The submittal requests a TS surveillance change to allow a 5-percent penetration test criteria. The Nuclear Regulatory Commission staff does not have sufficient information on filter performance for test penetration criteria lower than specified in your existing TSs.

The fact that lower filter efficiencies were assumed in the design basis analysis does not necessarily support reducing filter efficiency test requirements. The surveillance requirement is based on system parameters for a 2-inch deep filter considering residence time and other parameters using clean carbon adsorbers.

In addition, consideration is given to the change in efficiency over time so that there will be reasonable assurance that the filter will have the required efficiency at the end of the test interval, as well as at the beginning of the test interval. Performance of carbon adsorbers at other efficiencies as a result of depletion of adsorber sites over the time of the test interval have not been evaluated. Please justify the acceptability of reducing the testing criteria in your existing TS.

Serial No.: 05-077 Docket No.: 50-423 Response to Request for Additional Information Attachment 1 Page 2 of 26 DNC Response Current Licensing Bases:

The current charcoal adsorber efficiencies are in accordance with RG 1.52, Revision 2, as modified by GL 99-02. The current testing and efficiencies are also in accordance with the safety evaluation report (SER) associated with Millstone Unit 3 Amendment No.

184. The SER for Amendment 184 adopts Brookhaven National Laboratory Technical Evaluation Report (TER) which states, "Credited removal efficiency for radioactive organic iodine for (Control Room Emergency Ventilation System (CREVS)) is 95%. The proposed test penetration for radioactive methyl iodide for (CREVS) is less than 2.5%.

The proposed test penetration was obtained by applying a safety factor of 2 to the credited efficiency. The proposed safety factor of 2 for (CREVS) is acceptable because it ensures that the efficiency credited in the accident analysis is still valid at the end of the surveillance interval. This is consistent with the minimum safety factor of 2 specified in GL 99-02."

AST Analysis:

The AST analysis provided in DNC letter dated May 27, 2004 assumes charcoal efficiencies (%) for Control Room Filtered Recirculation and Intake as 90% vs the previous assumed efficiencies of 95%. Based on NRC guidance provided in Generic Letter 99-02 "Laboratory Testing of Nuclear-Grade Activated Charcoal," the allowable penetration for charcoal testing, for those plants that test in accordance with ASTM D3803-1989, is calculated as follows (ref: Attachment 2 of GL 99-02):

Allowable Penetration = (100% - Methyl Iodide Eff. for Charcoal Credited in Accident Analysis)

Safety factor Safety factor > 2 for systems with or without humidity control 100% - 90%(the credited charcoal efficiency in the accident analysis as assumed in AST submittal) divided by a factor of safety of 2 is equal to an allowable penetration of 5%.

Millstone Unit 3 presently tests the Control Room charcoal in accordance with ASTM D3803-1989 and therefore the allowable penetration for charcoal testing, per the NRC guidance provided in GL 99-02, is less than or equal to 5%.

This is consistent with the background information provided in GL 99-02, which states that the nuclear industry's understanding of charcoal testing has evolved over the years since the original issuance of Regulatory Guide (RG) 1.52. GL 99-02 further states,

Serial No.: 05-077 Docket No.: 50-423 Response to Request for Additional Information Attachment 1 Page 3 of 26 "Analysis of design-basis accidents assume a particular ESF charcoal filter adsorption efficiency when calculating offsite and control room operator doses. Ucensees then test charcoal filter samples to determine whether the filter adsorber efficiency is greater than that assumed in the design-basis accident analysis. The laboratory test acceptance criteria contain a safety factor to insure that the efficiency assumed in the accident analysis is still valid at the end of the operating cycle. Because ASTM D3803-1989 is a more accurate and demanding test than the older tests, addressees that upgrade their TS to this new protocol will be able to use a safety factor as low as 2 for determining the acceptance criteria for charcoal filter efficiency (see note in Enclosure 2 for further discussion)."

Please note that Enclosure 2 of Generic Letter 99-02 provides the formula noted above for calculating the allowable penetration for ESF charcoal testing. The acceptance penetration criterion of 5% (95% efficient), based on an assumed charcoal efficiency of 90% provided in the AST submittal, is in accordance with the guidance in GL 99-02.

To summarize, the proposed 5%allowable penetration for Control Room charcoal testing for surveillance requirements 3/4.7.7.c.2 and 3/4.7.7.d is in accordance with guidance contained in Regulatory Guide 1.52, Revision 2, as modified by Generic Letter 99-02.

Resmonse To RAI In Letter Dated February 9. 2005 Overall Comments NRC Question 1 What is the basis for assuming that the control room ventilation system timing is the same as the fuel handling accident for the spectrum of Millstone Unit 3 accidents?

DNC Resnonse A Control Building Isolation (CBI) Signal initiates the timing of the Control Room Ventilation System. All accidents lead to the generation of a CBI. Table 1.3-1 of the AST submittal dated May 27, 2004 lists the response time for the Control Room Inlet Radiation Monitor to generate the CBI Signal as 5 seconds. This value was validated for each accident analysis where a Safety Injection (SI) Signal was not generated. The validation involved a calculation using the release rate from the specific accident, the control room X/Q, the alarm setpoint, the detector response conversion factor, and the detector time constant. Once a CBI signal is generated, either by the Control Room Inlet Radiation Monitor or the SI signal, the control room dampers isolate within 5 seconds. Prior to isolation, unfiltered, normal ventilation flow is 1450 +/- 10%. The analyses assume 1595 cfm for conservatism. All the analyses then assume that the Control Room Emergency Ventilation System, which consists of filtered intake air that pressurizes the control room and filtered recirculation, is manually placed in service

Serial No.: 05-077 Docket No.: 50-423 Response to Request for Additional Information Attachment 1 Page 4 of 26 conservatively at 101 minutes after the CBI signal. An Operations Department simulation confirmed the acceptability of the assumed time value, since the simulation only required 33.5 minutes to align the Control Room Emergency Ventilation, even with a loss of instrument air.

NRC Question 2 The control room analyses need to consider the doses to the Unit 2 control room operators. Provide those consequences and the operating mode and conditions of the control room ventilation systems in response to the Unit 3 accident. Provide the manner in which the Unit 2 control room ventilation systems are initiated for the Unit 3 accident and the timing associated with that initiation.

DNC ResDonse The Millstone Unit 3 LOCA and FHA were analyzed for effects on the Millstone Unit 2 Control Room and resulted in doses less than the limits imposed by 10 CFR 50.67.

These are the most limiting accidents from Millstone Unit 3 with respect to the impacts on the Millstone Unit 2 Control Room. The licensing and design basis of the Millstone Unit 2 Control Room has not changed from that which is listed in section 9.9.10 and Table 14.8.4-3 of the Millstone Unit 2 FSAR or that which has been recently approved by the NRC in the partial implementation of the AST for Millstone Unit 2. The Millstone Unit 2 Control Room Inlet Radiation Monitor isolates the Millstone Unit 2 Control Room from all Millstone Unit 3 events within 10 seconds. This value was validated and involved a calculation using the release rate from the specific accident, the control room X/Q, the alarm setpoint, the detector response conversion factor, and the associated time constant. Prior to isolation, the normal ventilation flow rate is 800 cfm. Once isolated, the unfiltered inleakage is 130 cfm. This value is in the Millstone Unit 2 Technical Specification 3/4.7.6 and is verified through surveillances. Once isolated, the control room operators manually align the Control Room Emergency Ventilation System within 10 minutes to recirculate control room air through 90% efficient filters. The recirculation flow rate is 2500 cfm + 10%. For conservatism the analyses assume 2250 cfm. These values are in the Millstone Unit 2 Technical Specification 3/4.7.6 and are verified through surveillances.

NRC Question 3 What is the impact of Unit 2 accidents on the doses to the Unit 3 control room operators as a result of these changes to Unit 3?

DNC Resnonse The limiting Millstone Unit 2 accidents to the Millstone Unit 3 Control Room are the Millstone 2 LOCA and FHA. Only the FHA has been selectively approved for AST at Millstone Unit 2, (Amendment 284, dated September 20, 2004). Therefore, the Millstone Unit 2 FHA dose analysis to the Millstone Unit 3 Control Room is performed using dose criteria from 10 CFR 50.67. The Millstone Unit 2 LOCA analysis is

Serial No.: 05-077 Docket No.: 50-423 Response to Request for Additional Information Attachment 1 Page 5 of 26 performed using the traditional dose criteria from GDC1 9 and SRP 6.4. The method of isolation of the Millstone Unit 3 Control Room is from the Control Room Inlet Radiation Monitor generating the CBI. All responses to the CBI are as described in question No. 1. Both the Millstone Unit 2 LOCA and FHA analyses show acceptable results below each respective dose criteria limit with respect to impact on the Millstone Unit 3 Control Room. In addition, if the LOCA results were converted to TEDE, assuming that the thyroid dose is 3% of TEDE, the TEDE to the Millstone Unit 3 Control Room from a Millstone Unit 2 LOCA would be less than the limit specified in 10 CFR 50.67.

NRC Question 4 What is the basis for two spray operation during the recirculation phase of the LOCA?

DNC Response The basis for crediting combined Quench Spray System (QSS) and Recirculation Spray System (RSS) is that both systems operate together until the RWST is empty during the initial stage of cold leg recirculation. Subsequent to the depletion of the RWST inventory, RSS continues to operate. However, iodine scrubbing is only credited during the period when QSS operates alone and when QSS and RSS operate together.

NRC Question 5 Describe what the neutral operating conditions are for the Unit 3 control room. What is the basis for assuming 350 cfm of unfiltered inleakage? What were the ASTM E741 test results when the control room was tested in this condition?

DNC Response As stated in the response to question No.1, the Millstone Unit 3 Control Room isolates upon receipt of a CBI and remains isolated for 101 minutes. During this time period the Control Room Emergency Ventilation System is not operating, does not provide filtered intake air to pressurize the control room and does not filter recirculated air in the control room. No other system is operating that would provide a positive pressure in the control room. This is the neutral operating condition for the Millstone Unit 3 Control Room. The 350 cfm assumed unfiltered inleakage amount used in the accident analyses was a conservative value chosen above the E741 test results. The E741 test, conducted on June 15 and 16 of 2004, was performed on Train A and B of the Control Room Emergency Ventilation System boundary dampers in the neutral operating condition.

Therefore, the control room was isolated by only one set of dampers to simulate a loss of an emergency train, which prevents closure of one complete train of dampers. The results of the latest were:

Train A Isolation: 91 +/- 3 acfm Train B Isolation: 95 +/- 3 acfm 1.3 Analysis Assumptions & Key Parameter Values

Serial No.: 05-077 Docket No.: 50-423 Response to Request for Additional Information Attachment 1 Page 6 of 26 NRC Question 1 There is a discussion in section 1.3.1 about the radiological consequences of the waste gas system failure and the radioactive liquid waste system leak or failure (atmospheric release). It is stated that these analyses will be retained in Chapter 11 of the Millstone 3 FSAR and that the whole body and thyroid doses will be converted to TEDE. These two analyses originate out of SRPs 15.7.1 and 15.7.2 of NUREG-75/087. The acceptance criteria for these two events was 500 mrem whole body which was the Part 20 limit at the time. Additional guidance on the consequences of a waste gas decay tank leak is provided in BTP 11-5 of SRP 11.3. If there is to be a conversion to TEDE, the acceptance criterion for the conversion should be based upon the criterion for the present Part 20, 100 mrem TEDE. In addition, any change to a TEDE dose criterion should not be limited to only the consequences from noble gases and iodides but also include the other isotopes associated with Regulatory Guide 1.183 and the AST.

DNC Resmonse Millstone will retain the radiological consequences of the waste gas system failure and the radioactive liquid waste system leak or failure (atmospheric release) in Chapter 11 of the Millstone Unit 3 FSAR and leave the doses as whole body and thyroid with the acceptance criteria of 500 mrem whole body. Millstone will not convert the doses to TEDE.

NRC Question 2 Provide the basis for increasing the acceptance criterion for methyl iodine penetration of the control room emergency air filtration system in surveillance requirements 3/4.7.7.c.2 and 3/4.7.7.d from 2.5% to 5%.

DNC Response The response to this question was provided as part of the response to the request for additional information dated December 29, 2004 (refer to page 1 of this Attachment).

NRC Question 3 What is the basis for the HEPA filter and charcoal adsorber efficiencies for the control room emergency ventilation system and why are they inconsistent with the values of Regulatory Guide 1.52?

DNC Resmonse The HEPA filter testing has not changed with this AST submittal and the current testing and efficiencies of the HEPA filters are in accordance with Regulatory Guide (RG) 1.52, Revision 2, Section 5.c.

The current charcoal adsorber efficiencies are in accordance with RG 1.52, Revision 2, as modified by GL 99-02. The current testing and efficiencies are also in accordance

Serial No.: 05-077 Docket No.: 50-423 Response to Request for Additional Information Attachment 1 Page 7 of 26 with safety evaluation report (SER) associated with Millstone Unit 3 Amendment No.

184. The SER for Amendment 184 adopts Brookhaven National Laboratory Technical Evaluation Report (TER), which states, "Credited removal efficiency for radioactive organic iodine for (Control Room Emergency Ventilation System (CREVS)) is 95%. The proposed test penetration for radioactive methyl iodide for (CREVS) is less than 2.5%.

The proposed test penetration was obtained by applying a safety factor of 2 to the credited efficiency. The proposed safety factor of 2 for (CREVS) is acceptable because it ensures that the efficiency credited in the accident analysis is still valid at the end of the surveillance interval. This is consistent with the minimum safety factor of 2 specified in GL 99-02."

The proposed charcoal filter efficiency for CREVS is 95% (5% penetration) based on a credited removal efficiency for radioactive organic iodine of 90% (refer to the response to the separate request for additional information dated December 29, 2004 on page 1 of this Attachment).

Serial No.: 05-077 Docket No.: 50-423 Response to Request for Additional Information Attachment 1 Page 8 of 26 3.0 Radioloailcal Event Re-analyses & Evaluation 3.1 Desian Basis Loss of Coolant Accident (LOCA) Reanalvsis NRC Ouestlon 1 Section 3.1.4 provides a discussion of the manner in which containment sprays are utilized during the course of a LOCA and the manner in which their use was modeled in the determination of the radiological consequences. However, the manner in which the information in this section and in Table 3.1-4 were incorporated in the calculations is unclear. Please provide a Table, which notes the time post-LOCA when the sprays were initiated and stopped, the spray removal coefficient for elemental and particulate forms of iodine utilized during these periods, when the sprays were terminated, and the DF at the time of termination.

DNC Response The following table provides the required information:

Time post- Spray Status Elemental Elemental Particulate Particulate LOCA X(per hr) DF X (per hr) DF (hrs) _ _ _ _ _ _ _ _ _ _

0 No Sprays - - -

0.02014 Quench sprays 20 - 12.73 effective 0.2333 Recirc Sprays 20 - 16.14 -

Effective 1.9 Particulate X 20 - 1.61 49.5 reduced 2.078 Quench sprays - 79 - -

off. Recirc sprays assumed off

Serial No.: 05-077 Docket No.: 50-423 Response to Request for Additional Information Attachment 1 Page 9 of 26 NRC Question 2 Was the Powers model used for aerosols in the sprayed region after the quench spray was secured and after it was assumed that there was no removal of iodine by the sprays?

DNC Resoonse As stated in section 3.4.1.2 of the Millstone Unit 3 AST submittal, the Powers model was only used for aerosol deposition in the unsprayed region and set for the 10 1h percentile. It is assumed to continue for the duration of the accident. After quench spray is secured there are no more iodine removal mechanisms, other than leakage and decay, in the sprayed region.

NRC Question 3 How is the tripping of the non-nuclear safety grade exhaust fan accounted for in the assessment of the control room operator's dose?

DNC Resbonse For the control room analysis, the following five fans are tripped by operator action at 80 minutes after the accident: 3HVQ-FN2 (ESF Bldg normal exhaust), 3HVV-FN1A & B (MSVB exhaust) and 3HVR-FN5 and FN7 (Aux. Bldg. Normal exhaust). At 80 minutes post-LOCA any releases as a result of the operation of these five fans cease. Releases as a result of containment filtered and unfiltered leakage, duct and damper bypass leakage, ECCS leakage, and RWST leakage continue as described. There is no deviation from the methodology employed in Millstone Unit 3 Amendment No. 211, dated September 16, 2002 (and the correction issued on November 25, 2002) regarding the revised Final Safety Analysis Report licensing basis for post-accident operation of the Supplementary Leakage Collection and Release System (TAC No. MB3700).

NRC Question 4 In Table 3.1-4, what is the rationale for concluding that the containment's leakrate will be reduced after 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> for the control room consequences but only after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for the offsite consequences?

DNC Response The reduction in the containment leakrate at 1-hour post-LOCA for calculation of control room doses was based on an evaluation of the containment pressure following a LOCA.

As stated on page 36 of the Millstone Unit 3 AST submittal, the reduction is based on the fact that the Millstone Unit 3 containment pressure is rapidly reduced compared to typical PWR's because of its original design as a negative pressure containment. This exception was approved in Amendment No. 211, dated September 16, 2002 and November 25, 2002 for Millstone Unit 3 regarding the revised Final Safety Analysis Report licensing basis for post-accident operation of the Supplementary Leakage Collection and Release System (TAC No. MB3700). The reduction in the containment

Serial No.: 05-077 Docket No.: 50-423 Response to Request for Additional Information Attachment 1 Page 10 of 26 leakrate at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> post-LOCA was a conservative assumption made for the calculation of off-site doses and is consistent with Reg. Guide 1.183.

NRC Question 5 Provide your calculations demonstrating the time at which the particulate and elemental iodine DF are achieved. If these calculations do not describe how the DF is defined, provide the definition.

DNC Response According to SRP 6.5.2, Rev. 2, the iodine decontamination factor is defined as the maximum iodine concentration in the containment atmosphere divided by the concentration of iodine in the containment atmosphere at some time after decontamination.

According to SRP 6.5.2, Rev. 2, the maximum decontamination factor is 200 for elemental iodine. The effectiveness of the spray in removing elemental iodine shall be presumed to end at that time, post- LOCA, when the maximum elemental iodine DF is reached. The calculation of the DF is accomplished by modeling the containment in RADTRAD using the sprayed and unsprayed volume, the elemental iodine removal coefficient for both quench spray only and recirc and quench spray together, the mixing rates for both quench spray only and recirc and quench spray together, setting the elemental iodine composition to 100%, turning off decay and daughter product in-growth and setting the release rate to zero. To calculate the true DF, the timing of the release into the containment must be set so it is a puff release at t=0. The iodines at t=0 are calculated by adding the curies of 1-131 in the sprayed and unsprayed region. At subsequent times, the remaining 1-131 curies in the sprayed and unsprayed region are compared to the 1-131 curies at t=0 to determine when the DF of 200 is reached. This evaluation results in a DF of 79 when sprays are turned off at 7,480 seconds (2.078 hours9.027778e-4 days <br />0.0217 hours <br />1.289683e-4 weeks <br />2.9679e-5 months <br />).

1-131 curies @ t=0: 4.855E+07 1-131 curies @ t=2.078 hrs: 6.177E+05 DF (4.855E+07 / 6.177E+05)= 79 According to SRP 6.5.2, Rev. 2, the effectiveness of the spray in removing particulate iodine shall be reduced at that time, post- LOCA, when a DF of 50 is reached. The calculation of the DF is accomplished by modeling the containment in RADTRAD using the sprayed and unsprayed volume, the particulate iodine removal coefficient for both quench spray only and recirc and quench spray together, the mixing rates for both quench spray only and recirc and quench spray together, setting the particulate iodine composition to 100%, turning off decay and daughter product in-growth and setting the release rate to zero. To calculate the true DF, the timing of the release into the containment must be set so it is a puff release at t=0. The iodines at t=0 are calculated

Serial No.: 05-077 Docket No.: 50-423 Response to Request for Additional Information Attachment 1 Page 11 of 26 by adding the curies of 1-131 in the sprayed and unsprayed region. By iteration (turning off the sprays at different times), the remaining 1-131 curies in the sprayed and unsprayed region are compared to the 1-131 curies at t=0 to determine when the DF of 50 is reached. The evaluation results in a DF of 49.5 at 1.9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br />. At this time the particulate removal coefficient is reduced by a factor of 10 until 7,480 seconds at which time all sprays are assumed secured.

1-131 curies @ t=O: 4.855E+07 I-131 curies @ t=1.9 hrs: 9.813E+05 DF (4.855E+07 / 9.813E+05)= 49.5 NRC Question 6 At what time does the spray removal coefficients for particulate become 1.27 and 1.61 ?

What regions of containment do these spray removal coefficients apply?

DNC Response At no time does the spray removal coefficient for particulate become 1.27 per hour.

This value is based on quench spray operation only. The recirculation sprays become effective prior to the time at which the particulate DF of 50 is reached. Recirculation sprays become effective at 14 minutes and the composite spray removal coefficient for particulate becomes 16.14 per hour, up from 12.73 per hour. At 1.9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> the DF of 50 is reached and the spray removal coefficient for particulate iodine becomes 1.61 per hour. At 2.078 hours9.027778e-4 days <br />0.0217 hours <br />1.289683e-4 weeks <br />2.9679e-5 months <br /> the quench spray system is secured. It is assumed that the recirculation spray system is secured at the same time. Therefore, there is no longer a spray removal coefficient for particulate iodine after 2.078 hours9.027778e-4 days <br />0.0217 hours <br />1.289683e-4 weeks <br />2.9679e-5 months <br />. All spray removal coefficients apply only to the sprayed area. During quench spray operation only, the spray area represents 49.63% of the containment volume. During quench and recirculation spray system operation, the spray area represents 64.5% of the containment volume.

NRC Question 7 Table 3.1-5 provides information on contaminated inflow to the RWST. Is the time information in the Table the duration of the inflow from a given source or is it the time the leak begins post accident?

DNC Response The time listed in Table 3.1-5 is the time at which contaminated sump water from that leakage path reaches the RWST post-LOCA. Using the methodology approved in Amendment 176 for Millstone Unit 3 regarding RWST backleakage, the time for contaminated sump water to reach the RWST is based on the calculated flow rates and the volume of clean water in the associated piping. As further explained on page 40 of Attachment 1 to the AST submittal, the times in Table 3.1-5 are reduced by 50% to

Serial No.: 05-077 Docket No.: 50-423 Response to Request for Additional Information Attachment 1 Page 12 of 26 account for mixing in the lines. The times used in the radiological analysis are listed in Table 3.1-6.

NRC Question 8 How full are each of the lines to the RWST when the backflow begins? What is the volume of each line?

DNC Restonse All lines are assumed to be full of clean water (ie - no iodine) prior to the start of the accident. As explained on page 40 of the Millstone Unit 3 AST submittal dated May 27, 2004, the time required for contaminated fluid to reach the RWST is determined by how long it takes the leakage to displace the clean fluid in the stagnant and isolated piping sections. This time is then reduced by 50% to account for mixing.

As stated on page 39 of the Millstone Unit 3 AST submittal dated May 27, 2004, all volumes were approved in Amendment 176 for Millstone Unit 3 regarding RWST backleakage. The table below lists the RWST backleakage source and the total volume of all lines associated with that source.

RWST Backleakage Data Source Total Volume of all Lines (gal)

CHS Suction 4065 RHRS A Suction 3984 RHRS B Suction 4202 SIH Recirculation 101 RHRS 1658 Recirculation CHS Recirculation 1277 SIH Suction 3048 NRC Question 9 Table 3.1-6 provides information on contaminated inflow to the RWST. Similar to Table 3.1-5 it provides a summary of times. However, as with Table 3.1-6 it is unclear whether the time information in the Table is the duration of the inflow from a given

Serial No.: 05-077 Docket No.: 50-423 Response to Request for Additional Information Attachment 1 Page 13 of 26 source or is it the time the leak begins post accident. It is also unclear why the volume for the SIS R is shown to be zero. Clarify this Table.

DNC Response Table 3.1-5 contains the seven sources of RWST backleakage. The time listed in the second column is the post-LOCA time at which that source of RWST backleakage reaches the RWST. The flow rate in the third column is the flow rate for that specific source of backleakage.

Table 3.1-6 shortens the time in Table 3.1-5 by 50% to account for mixing in the lines.

Therefore, the contaminated sump fluid reaches the RWST sooner than the times calculated in Table 3.1-5. The flow rates in column 2 are the composite flow rates. As an example, SIS R reaches the RWST at 4.25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> with a flow rate of 0.03 cfm. At 14.91 hours0.00105 days <br />0.0253 hours <br />1.50463e-4 weeks <br />3.46255e-5 months <br /> the SIS R path is joined by the RHS R path, increasing the flow rate by 0.08 cfm for a total flow rate of 0.11 cfm. The volume in column 3 is the volume of contaminated sump fluid that entered the RWST at the time listed in column 1. Since the SIS R path does not reach the RWST until 4.25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br />, the volume of contaminated sump fluid in the RWST at this time is zero. From 4.25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> to 14.91 hours0.00105 days <br />0.0253 hours <br />1.50463e-4 weeks <br />3.46255e-5 months <br /> the flow rate into the RWST is 0.03 cfm from SIS R. The total volume of contaminated sump fluid in the RWST is 17.11 ft3 at 14.91 hours0.00105 days <br />0.0253 hours <br />1.50463e-4 weeks <br />3.46255e-5 months <br />. From 14.91 hours0.00105 days <br />0.0253 hours <br />1.50463e-4 weeks <br />3.46255e-5 months <br /> until 18.46 hours5.324074e-4 days <br />0.0128 hours <br />7.60582e-5 weeks <br />1.7503e-5 months <br /> the flow rate into the RWST is 0.11 cfm. The total volume of contaminated sump fluid in the RWST is 39.88 ft3 at 18.46 hours5.324074e-4 days <br />0.0128 hours <br />7.60582e-5 weeks <br />1.7503e-5 months <br />. This methodology was approved in Amendment 176 for Millstone Unit 3 regarding RWST backleakage.

NRC Question 10 Provide the calculation, which demonstrates how the RWST airflow rate of 8.7 cfm was determined?

DNC Resoonse The breathing rate of the RWST was determined by making use of the ideal gas law and expected volumetric change. The latter was based on a conservative rise in air temperature within the RWST as a result of solar heating. From the ideal gas law, the relative change in volume resulting from a change in temperature at constant pressure is given by:

(Delta Volume / Volume) = (Delta Temperature I Temperature)

Assuming a conservative initial air temperature within the tank of 35 F (495 R) and a rise in the temperature as a result of solar heating by 40 F (40 R), the relative change in the air volume is 0.0808. The minimum amount of water at the end of the injection phase is 47,652 gallons. At the maximum fill level the RWST can hold 1,206,644 gallons or 1.613E+05 ft3. That translates into an air volume at the minimum water volume of 1.159E+06 gallons. It is conservatively assumed that this entire space is occupied by air at the end of the injection phase, then the change in volume due to the

Serial No.: 05-077 Docket No.: 50-423 Response to Request for Additional Information Attachment 1 Page 14 of 26 assumed temperature rise is approximately 1.252E+04 ft3 (0.0808 x 1.1 59E+06 gallons

/7.481 gallons per ft3).

To maintain the same pressure in the tank, this amount of air must be released from the tank during the solar heating interval. Following the end of heatup, and as the RWST air starts to cooldown, outside air is drawn into the tank. The cycle repeats on a daily basis. On average, the RWST releases 1.252E+04 ft3 of air per day, or about 8.7 cfm.

This flow rate is used for the entire duration of the accident even though it will decrease over time due to the decreased volume of air in the tank due to the increasing volume of fluid. The methodology used to calculate the RWST breathing rate is consistent with the methodology used in Amendment 176 dated November 4, 1999, for Millstone Unit 3 regarding RWST backleakage.

NRC Question 11 What iodine isotopes contributed to the value of 10,000 grams of iodine in the core?

Provide the calculation for this number.

DNC Resnonse As stated on page 41 of the Millstone Unit 3 AST submittal dated May 27, 2004, there are 20,000 grams of iodine in the core. The same methodology was used to calculate the grams of iodine in the core as was used to calculate the curies of iodines in the core. This methodology was described in the Millstone Unit 3 submittal for the selective implementation of the AST dated March 4, 2003, and approved in the issuance of Amendment 219, "Regarding Selective Implementation of the AST," dated March 17, 2004. Three ORIGENARP computer runs were used to represent the three regions of the fuel. Inputs included 5% enrichment and batch average burnup for one, two, and three cycle fuel. The ORIGENARP calculation resulted in the once burned fuel producing 4.98E+03 grams, the twice-burned producing 8.76E+03 grams, and the thrice-burned producing 3.26E+03 grams of iodines. This results in 17,000 grams of iodine in the core, which was conservatively rounded to 20,000. The output from ORIGENARP only lists the total grams of iodine and does not break it down by isotope.

NRC Question 12 The discussion of backleakage to the RWST in section 3.1.5.3 of the application describes the amount of leakage and the resulting iodine concentration in the tank. The discussion does not provide a value for the pH of the solution in the RWST or the amount of iodine in the air space. Since the iodine concentration and the pH influence the potential release of radioactive iodine, please discuss the pH in the RWST and the amount of iodine released to the air space in the tank during the 30-day LOCA period.

If pH calculations were performed, please provide the results, inputs, and explanation of the inputs. If pH calculations were not performed, please explain why this was considered unnecessary for assuring that elemental iodine would not be evolved from the tank.

Serial No.: 05-077 Docket No.: 50-423 Response to Request for Additional Information Attachment 1 Page 15 of 26 DNC Restonse The methodology used to determine the iodine evolution in the RWST is consistent with the methodology used in Amendment 176 dated November 4, 1999, for Millstone Unit 3 regarding RWST backleakage. A pH calculation was not performed. The methodology used to perform the calculation was described on page 41 of the Millstone Unit 3 AST submittal dated May 27, 2004. The partition coefficient (PC) applicable to the iodines in the RWST water is based upon information in "Iodine Removal From Containment Atmospheres by Boric Acid Spray," BNP-100, July 1970. The PC was based on an acidic solution. Any sump fluid leaking into the RWST would only raise the pH and increase the PC.

For this application, the RWST was assumed to behave like a closed system for the establishment of equilibrium conditions between the water and air. This is appropriate during the cooldown phase when air drawn into the RWST inhibits the loss of airborne iodine. This is also appropriate during the heat-up phase as the change in air volume is small and any impact on equilibrium conditions is therefore minimal.

The critical factor in the magnitude of the partition coefficient (PC) for iodines is the total iodine concentration in the water. For application of this information it was first necessary to compute the iodine concentration in the RWST. The quantity of backleakage over 30 days is 7.41E+04 gallons. Adding this volume to the minimum volume at the end of the injection phase (47,652 gallons) results in a total volume in the RWST at the end of the 30 days of 1.218E+05 gallons. The concentration of iodine in the RWST increases over time and the maximum occurs at the end of the 30 days because initially the volume of water remaining in the RWST is free of iodine. Therefore the concentration of iodine in the RWST at the end of 30 days is 4.3 mgrams/gallon or 1.2 mgrams/liter.

Due to the maximum concentration of iodines in the RWST at the end of 30 days, the partition coefficient (PC) and Decontamination Factor (DF) will be at the minimum at this time. The smaller PC and DF result in a larger amount of iodine available for release.

For conservatism, the DF at the end of 30 days was used for the entire RWST backleakage period. The PC corresponding to the maximum iodine concentration of 1.2 mgrams/liter was taken from Figure 8 of "Iodine Removal From Containment Atmospheres by Boric Acid Spray," BNP-100, July 1970. The PC is approximately 4000. The DF is calculated using the equation from SRP 6.5.2, Rev. 2 and is:

DF = 1 + (Vliquid / Vair)

• PC where Vliquid and Vair are the volumes in the RWST between which the partitioning takes place.

Serial No.: 05-077 Docket No.: 50-423 Response to Request for Additional Information Attachment 1 Page 16 of 26 The RWST liquid volume at the end of 30 days is 1.218E+05 gallons. The maximum RWST capacity is 1,206,644 gallons, resulting in an air volume at the end of 30 days of 1.085E+06 gallons. The resulting DF is 450. A DF of 100 was used for conservatism.

The RADTRAD model did not model a RWST airspace. The concentration of contaminated sump water leaked into the RWST according to Table 3.1-6. At the same time the RWST liquid was conservatively assumed to discharge directly to the environment at a rate of 8.7 cfm with a conservative DF of 100. As stated previously the calculated DF was 450.

3.2 Fuel Handlina Accident NRC Question 1 What is the control building isolation signal based upon?

DNC Resionse Per Technical Specification Table 3.3-4, the alarm setpoint for the MP3 Control Building Inlet Ventilation Radiation Monitors is 1.5E-05 uCi/cc. To determine the concentration at the rad monitor, the release rate must be determined and then multiplied by the appropriate control room X/Q. For conservatism, only the Xe-1 33 portion of the release is considered. The release rate is determined by taking the maximum airborne Xe-1 33 activity in the building following a FHA, multiply it by the air change rate, which provides the release rate, and then multiplying by the X/Q to calculate the concentration at the control room inlet.

An air change rate of 3.5/hr was used for the release. This ensures that essentially all airborne activity is released from containment or the fuel building to the environment within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. To determine the Xe-133 activity in the fuel building (or containment),

the RADTRAD computer code provided the Xe-133 activity in the building at T = 1 sec post-FHA and it was 1.4182E+05 Ci.

The release rate at T= 1 second is 1.4182E+05 Ci

• 3.5 / hour = 138 Ci/sec. With a X/Q to the control room from the containment of 5.34E-04 sec /m3, the resultant activity concentration at the MP3 control room inlet is 7.4E-02 uCi/cc. The lower X/Q for containment was chosen (as opposed to the fuel building) because it will provide less activity to the control room monthly (and a resultant longer rad monitor response time) and therefore is conservative. This activity level will alarm the inlet rad monitor based on the alarm setpoint of 1.5E-05 uCi/cc, which will isolate the control room.

The response time of the rad monitor must be determined to address impact on MP3 control room isolation time. The Xe-1 33 activity at the control room inlet as determined in the above paragraph is 7.4E-02 uCi/cc. The method below will be used to show how quickly the MP3 control room inlet radiation monitor will detect the release.

Serial No.: 05-077 Docket No.: 50-423 Response to Request for Additional Information Attachment 1 Page 17 of 26 Alarm setpoint: 1.5E-05 uCi/cc Xe-1 33 Time constant: 1 sec for a count rate > 3000 cpm Detector response conversion factor: 1.3E-08 uCi/cc per cpm Xe-1 33 Set-point count rate (Cs) = 1.5E-05 uCi/cc / 1.3E-08 uCi/cc per cpm = 1154 cpm Concentration = 7.4E-02 uCi/cc Step increase (Cf) = 7.4E-02 uCi/cc / 1.3E-08 uCi/cc/cpm = 5.7E+06 cpm Since the detector instantaneous count rate (5.7E+06 cpm) is greater than 3000 cpm, the time constant (RC) is 1 sec. The monitor response time, T, to reach the setpoint is determined by:

Cs = Cf (1-e-TRC)

Cs / Cf = (1-6T/Ic) 1154/5.7E+06 = (1-e TI Se T = 2.2E-04 sec The response time of the rad monitor is 2E-04 seconds. For conservatism, 5 seconds is used, which includes the release at 1 second. In addition, 5 seconds is used for damper closure time. The MP3 Technical Requirements Manual lists a control building isolation time of 6 seconds, which includes rad monitor response time. It credits 3 seconds for damper closure time in footnote 10 of TRM section 3.3.2. Five seconds is credited for damper closure in this calculation for conservatism. Therefore, it is assumed that it takes 10 seconds for the control room to isolate following detection by the MP3 control room air inlet detectors.

NRC Question 2 It is proposed that no credit is taken for the control room envelope pressurization system. Therefore, the system can be deleted from technical specification. In fact, credit is taken for the pressurization system. The analyses, which support this amendment request base the inleakage characteristics of the CRE for the first hour of the fuel handling accident and other accidents upon the pressurization system working. If the pressurization system is not operating, the inleakage characteristics of the CRE will be different. The question becomes, 'What is the inleakage characteristics of the CRE with the normal control room ventilation system operating, with no system operating, or with the pressurization system operating"? Which results in the worst case condition?

Based upon the existing information, the proposed deletion of the pressurization system from the technical specifications may be inappropriate.

Serial No.: 05-077 Docket No.: 50-423 Response to Request for Additional Information Attachment 1 Page 19 of 26 a loss of power was considered. The most limiting characteristic for unfiltered in-leakage was determined to be with the Control Room habitability boundary at a neutral pressure with no loss of power.

3.3 Steam Generator Tube Ruoture Accident NRC Question 1 Why is there no steam releases from 0.8183 hours0.0947 days <br />2.273 hours <br />0.0135 weeks <br />0.00311 months <br /> till 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />? What precipitates the steaming of the affected steam generator at 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and not before?

DNC Resnonse In the time span from 0.8183 hours0.0947 days <br />2.273 hours <br />0.0135 weeks <br />0.00311 months <br /> to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, the affected steam generator is isolated by the closure of its block valve. Therefore, there are no steam releases from the affected generator.

In the Westinghouse Steam Generator Tube Rupture analysis of record, it was assumed that the affected steam generator is depressurized immediately following the completion of the RCS cooldown. This occurs between 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. The depressurization of the affected steam generator early (at 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> into the event) maximizes the dose consequences.

NRC Question 2 What isolates the control room on a SGTR?

DNC Response The control room isolates based on a radiation alarm signal from the control room inlet radiation monitor.

NRC Question 3 Is the control room envelope pressurization system assumed to operate in the event of a SGTR?

DNC Resnonse The Control Room Envelope Pressurization System (CREPS) is not credited to operate during the SGTR. This system is different from the control room emergency air filtration system, which is credited with pressurizing the control room. Because the envelope pressurization system is not credited, it is assumed in the analysis that pressurization does not occur until approximately 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and 41 minutes post-SGTR, after which the emergency air filtration system pressurizes the control room.

NRC Question 4 Was the iodine spike assumed to occur during the entire duration of the accident?

Serial No.: 05-077 Docket No.: 50-423 Response to Request for Additional Information Attachment 1 Page 20 of 26 DNC Resnonse The concurrent iodine spike was evaluated based on an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> duration as discussed on Page 52 of our submittal dated May 27, 2004. After 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, the iodine remaining in the RCS was released to the environment as a function of the primary to secondary leak rate.

NRC Question 5 Why was it assumed that there was no flashing of the break flow from 0.927 hours0.0107 days <br />0.258 hours <br />0.00153 weeks <br />3.527235e-4 months <br /> till 1.554 hours0.00641 days <br />0.154 hours <br />9.160053e-4 weeks <br />2.10797e-4 months <br />?

DNC Response Since releases from the affected steam generator were terminated as a result of operator action at 20 minutes, pressure in the generator built up to a point (from 0.9279 to 1.554 hours0.00641 days <br />0.154 hours <br />9.160053e-4 weeks <br />2.10797e-4 months <br />) where flashing no longer occurred and break flow into the steam generator was 100% liquid.

3.4 Ma!n Steam Line Break Analysis NRC Question 1 What was the duration of the iodine spike for this accident?

DNC Reswonse The concurrent iodine spike was evaluated based on an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> duration as discussed on Page 64 of the submittal.

3.5 Locked Rotor Accident NRC Question 1 Has the closure of the stuck-open ADV within 20 minutes been demonstrated to be a reasonable action for the control room operators or has it been maintained because it is part of the existing licensing basis?

DNC Response In support of Millstone Unit 3 Amendment No. 172, dated July 2, 1999, the 20-minute closure of the ADV block valve was verified. The Safety Evaluation Report of Amendment 172 concurs with the time required for this operator action.

NRC Questlon 2 Table 3.5-1 provides, as a function of time, the total steam flows to atmosphere and the mass flow rates from the 3 intact steam generators. The total steam flow to atmosphere seems to reflect the release from two steam generators and not three. How are these numbers calculated?

Serial No.: 05-077 Docket No.: 50-423 Response to Request for Additional Information Attachment 1 Page 21 of 26 DNC Resnonse The NSSS vendor, Westinghouse, provides the total steam flows to atmosphere calculation. To calculate the amount of heat that must be dissipated via steam release through the Atmospheric Dump Valves (ADVs), an energy balance is done. The energy balance considers the heat generated in the core, the heat released or absorbed by metal in the RCS and intact steam generators, and the heat released or absorbed within the fluids in the RCS and intact steam generators.

The energy that cannot be stored within the defined boundary of the RCS and intact steam generators is removed via steaming. The calculation considers two different time periods: from 0 to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and from 2 to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. The steam flow from the intact steam generators is determined by subtracting the steam flow for the steam generator with the failed open ADV from the total steam flow calculated by the energy balance summarized above.

3.6 RCCA Election Accident NRC Question 1 What is the basis for assuming that the primary system pressure is less than the secondary side after 20 minutes?

DNC ResDonse The 20-minute value is a conservative input, which is provided by Westinghouse. It is based on Westinghouse's analysis of a 3-inch Small Break LOCA transient.

NRC Question 2 Why are there no steam releases from 20 minutes to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />?

DNC Resnonse During the first 20 minutes, the only steam release is via the secondary safety valve.

When the primary system pressure drops below the secondary side pressure, the safety valve closes. At 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, a cooldown to the Residual Heat Removal System entry conditions is initiated. Thus, the releases start again at 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

7.0 Technical specification and Bases Chanae NRC Question 1 The staff will accept the use of the thyroid dose conversion factors but the proposed definition of dose equivalent 1-131 should be changed. The licensee should consider whether they agree to the following definition change (note the words in bold):

Serial No.: 05-077 Docket No.: 50-423 Response to Request for Additional Information Attachment 1 Page 22 of 26 DOSE EQUIVALENT 1-131 shall be that concentration of 1-131 (micro curie per gram) which alone would produce the same TEDE dose as the quantity and isotopic mixture of 1-131, 1-132, 1-133, 1-134, and 1-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed under Inhalation in Federal Guidance Report No. 11 (FGR 11), "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion and Ingestion,"

DNC Response Dominion agrees with the change except for the term 'TEDE." Dominion believes TEDE should be replaced with "CDE-thyroid," which would more accurately reflect use of thyroid dose conversion factors listed under Inhalation in Figure 11.

NRC Question 2 A change to surveillance requirements c.2 and d of Technical specification 3/4.7.7, "Control Room Emergency Air Filtration System," was proposed which would increase the acceptance criteria for methyl iodide penetration from 2.5% to 5%. The basis for this change appears to be revised AST analysis, which decreased the adsorber efficiency for the elemental and organic forms of iodine to 90% and 70% respectively.

This proposed change in adsorber efficiency and associated change to the surveillance requirements are unacceptable. Regulatory Guide 1.52 details the adsorber efficiencies, which should be assigned based upon the various depths of charcoal. The adsorber efficiencies selected for the elemental and organic forms of iodide are inconsistent with this guidance. Therefore, this decrease in adsorber efficiency cannot be approved nor can the acceptance criteria for surveillance requirements c.2 and d unless justification is provided which satisfactorily explains the basis for the change.

DNC Resnonse Millstone Unit 3 is committed to RG 1.52, Revision 2, as modified by Generic Letter (GL) 99-02. Per GL 99-02, the allowable penetration for charcoal adsorber testing is calculated based on the methyl iodide efficiency for charcoal credited in the accident of record. The current credited methyl iodide efficiency in the Millstone Unit 3 accident analysis is 95%. Using the guidance provided in GL 99-02, the allowable penetration for the Control Room Emergency Ventilation System charcoal testing (Technical Specification 3 / 4.7.7) was calculated to be 2.5 % (97.5% efficient). This was accepted by the NRC in the safety evaluation associated with Amendment 184 of Millstone Unit 3.

The proposed AST analysis revises the credited methyl iodide efficiency in the MP3 accident analysis from 95% to 90%. Based on the guidance in GL 99-02, the proposed allowable penetration for charcoal adsorber testing is calculated to be 5% (95%

efficient). This proposed charcoal testing efficiency change is consistent with a previously approved change for Crystal River Unit 3 Control Room Emergency Ventilation System in NRC letter to Florida Power Corporation, "Crystal River Unit 3 -

Serial No.: 05-077 Docket No.: 50-423 Response to Request for Additional Information Attachment 1 Page 23 of 26 Issuance of Amendment Regarding Alternative Source Term and Control Room Ventilation System (TAC No. MR 0241)," dated September 17, 2001.

NRC Question 3 It has been proposed that Technical specification 3/4.7.8, "Control Room Envelope Pressurization System," be deleted since it is no longer credited in the accident analyses for AST. All of the AST accident analyses have assumed operation of the Control Room Envelope Pressurization System. It appears that the analyses of the control room operators dose assumed that the control room would be in a neutral pressure condition even though the CRE should be pressurized by the Control Room Envelope Pressurization System. The AST amendment request did not elaborate on what control room ventilation systems would be operating if the Control Room Envelope Pressurization System did not operate. The assessment of the control room operators dose must account for the manner of operation during the course of the action. Since it is proposed to delete the Control Room Envelope Pressurization System as an ESF system intended to protect the control room operators in the event of a radiological accident, an assessment needs to be performed which identifies the various configuration of control room ventilation systems that may be functioning following a radiological event. For example, it has been proposed that the Control Room Envelope Pressurization System, while no longer ESF grade, will be functioning during the first hour post-accident. What is the CRE's inleakage characteristics when it is? What if the Control Room Envelope Pressurization System does not operate? Now what control room ventilation systems are operating and what is the CRE's inleakage characteristics when they are operating? What is the effect upon CRE cooling under these scenarios?

DNC Response Millstone Unit 3 does not take credit for the operation of the Control Room Envelope Pressurization System (CREPS). Rather, the operation of the system is used to determine the length of time for the scenario with the worst case unfiltered inleakage.

This time is determined to be 101 minutes (1 minute for actuation of CREPS, 60 minutes for CREPS operation, and 40 minutes to align CREVS) or 1.685 hours0.00793 days <br />0.19 hours <br />0.00113 weeks <br />2.606425e-4 months <br />. The assumed unfiltered inleakage during times of neutral pressure is 350 cfm and during times of positive pressure is 100 cfm. Actual values are less and are listed in response to question No. 5 of the February 9, 2005, above. The definition of neutral pressure condition is when the Control Room Emergency Ventilation System (CREVS) is not in operation, neither providing filtered intake air to pressurize the control room or filtering recirculated air in the control room. No other system is operating that would provide a positive pressure in the control room. For that condition, as stated in Table 3.2-1 of the Millstone Unit 3 AST submittal, the unfiltered inleakage into the Millstone Unit 3 Control Room is 350 cfm during the first 1.685 hours0.00793 days <br />0.19 hours <br />0.00113 weeks <br />2.606425e-4 months <br /> of a FHA. The 1.685 hours0.00793 days <br />0.19 hours <br />0.00113 weeks <br />2.606425e-4 months <br /> includes approximately 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> that the Control Room Envelope Pressurization System (CREPS) is available, but not credited, to operate and the associated 40 minutes to align the Control Room Emergency Ventilation System (CREVS). Specifically, CREPS is not credited in any control room analyses. Without CREPS pressurization, an unfiltered inleakage flow rate of 350 cfm will exist until the CREVS is aligned and operating at 101

Serial No.: 05-077 Docket No.: 50-423 Response to Request for Additional Information Attachment 1 Page 24 of 26 minutes. The unfiltered inleakage flow rate would then be reduced to 100 cfm. Since no design change is being made with this amendment, the Control Room Envelope Pressurization System (CREPS) is not being removed physically, but it is being removed from the Technical Specifications. Therefore, Millstone Unit 3 is not taking credit for the positive pressure in the control room for the operation of the Control Room Envelope Pressurization System (CREPS). Furthermore, Millstone Unit 3 AST analysis does not assume use of the CREVS until 101 minutes although it would take only 40 minutes to align that separate system. Accordingly, this assumed delay increases the time at which a higher unfiltered inleakage exists because it is assumed during this 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> that the control room remains at a neutral pressure. If the time penalty is not taken, the first 40 minutes of unfiltered inleakage would be at a rate of 350 cfm followed by 100 cfm for the remainder of the accident. By assuming an unfiltered inleakage flow rate of 350 cfm for 101 minutes (1 minute for actuation of CREPS, 60 minutes for CREPS operation, and 40 minutes to align CREVS), the methodology used in this calculation is conservative.

The table on the next page provides a comparison of what assumptions exist in the current submittal to assumptions that would exist with or without CREPS. The times listed are for a control room radiation monitor initiated CBI signal. Times for other accidents would vary slightly depending on what signal isolates the control room. For example the LOCA isolates the control room prior to plume arrival.

Serial No.: 05-077 Docket No.: 50-423 Response to Request for Additional Information Attachment 1 Page 25 of 26 Time Post Status Current CREPS CREPS Accident Submittal If Operating If Removed 0 CR intake flow 1595 cfm 1595 cfm 1595 cfm CBI Signal generated Signal Signal 5 seconds generated generated CREPS N/A Receives CBI N/A Control room Isolated Isolated Isolated 10 seconds CR press. Neutral Pressurizing Neutral Unfiltered 350 cfm <350 cfm 350 cfm inleakage CREPS N/A Operating Not available CREVS Not operating Not operating Not operating 65 seconds CR Press Neutral Positive Neutral Unfiltered 350 cfm 100 din 350 cfm inleakage CREPS N/A Operating Not available 41 minutes, CREVS Not operating Not operating Operating 5 seconds CR press. Neutral Positive Positive Unfiltered 350cfm 100 cfm 100 cfm inleakage CREPS N/A Exhausted Not available 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, 65 CREVS Not operating Not operating Operating seconds CR press. Neutral Neutral Positive Unfiltered 350 cfm 350 cfm 100 cfm inleakage CREPS N/A Exhausted Not available 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, 41 CREVS Operating Operating Operating minutes, 5 CR press. Positive Positive Positive seconds__ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ___ _ _ _ _ _ _

Unfiltered 100 cfm 100 cfm 100 cfm inleakage I___

To summarize, the current AST submittal dated May 27, 2004 assumes the worst timing and unfiltered inleakage conditions for either case of CREPS operating or removed.

With or without CREPS, the control room dose is maximized by assuming neutral pressure during the time period when the CREPS would have operated.

Serial No.: 05-077 Docket No.: 50-423 Response to Request for Additional Information Attachment 1 Page 26 of 26 The following are comments on the BASES Section.

NRC Question 1 In 3/4.6.1.1 GDC 19 should not be replaced with Regulatory Guide 1.183. GDC 19 does apply.

DNC Response It is agreed that Reg. Guide 1.183 is not appropriate. GDC 19 will be referenced.

NRC Question 2 In 3/4.7.7 it specifies that the Control Room Emergency Ventilation System may be operated in either the isolation and recirculating mode of operation or in the pressurization mode of operation. However, the AST analyses reflects only a pressurization and recirculating mode of operation. The BASES should reflect the actual intended modes of operation for the system. If these other modes of operation are possible then the dose consequences associated with these other modes need to be provided as well as the inleakage characteristics of the CRE during these modes of operation. In addition, in various sections there is reference to modes of operation which are inconsistent with the mode of operation described in the AST analyses.

Examples include the Background, Surveillance Requirement 4.7.7.e.2.

DNC Response Several procedures direct Operations to EOP 35 GA-18, "Aligning Control Room Ventilation." These procedures include "Loss of Reactor or Secondary Coolant,"

"Reactor Trip or Safety Injection," "Steam Generator Tube Rupture," and uRadiation Monitor Alarm Response." The purpose of EOP 35 GA-18 is to provide the actions necessary to align the Control Room Emergency Ventilation to outside filtered air following the initiation of a CBI. The alternate method of aligning Control Room Emergency Ventilation in the recirculation only mode of operation is not an option in the EOP's. Operations personnel are not trained to the recirculation only mode of operation of Control Room Emergency Ventilation. All references to the recirculation only mode of operation in the Technical Specification Bases, including 3/4.7.7 and 4.7.7.e.2 will be removed using the applicable Millstone Station processes.

Serial No.05-077 Docket No. 50-423 ATTACHMENT 2 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION PROPOSED TECHNICAL SPECIFICATION CHANGES FOR IMPLEMENTATION OF ALTERNATE SOURCE TERM MARKED UP PAGES DOMINION NUCLEAR CONNECTICUT. INC.

MILLSTONE POWER STATION UNIT 3

DEFINITIONS CONTAINMENT INTEGR=TY 1.7 CONTAINMENT INTEGRITY shall exist when:

a. All penetrations required to be closed during accident conditions are either:

1. Capable of being closed by an OPERABLE containment automatic isolation valve system*, or

2. Closed by manual valves, blind flanges, or deactivated automatic valves secured in their closed positions, except for valves that are opened under administrative control as permitted by Specification 3.6.3.

b. All equipment hatches are closed and sealed,

c. Each air lock is in compliance with the requirements of Specification 3.6.1.3,

d. The containment leakage rates are within the limits of the Containment Leakage Rate Testing Program, and

e. The sealing mechanism associated with each penetration (e.g., welds, bellows, or 0-rings) is OPERABLE.

CONTROLLED LEAKAGE 1.8 CONTROLLED LEAKAGE shall be that seal water flow supplied to the reactor coolant pump seals.

CORE ALTERATIONS 1.9 CORE ALTERATIONS shall be the movement of any fuel, sources, reactivity control components, or other components affecting reactivity within the reactor vessel with the vessel head removed and fuel in the vessel. Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe position.

DOSE EQUIVALENT 1-131 CD - iii c J 1.10 DOSE EQUIVALENT 1-131 hall be that concentration of 1-131 (microCurie/gram) which alone would produce the same dose as the quantity and isotopic mixture of I-131, 1-132, 1-133, I-134, and 1-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listedlo Federal Guidance No. 11, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion and Ingestion." I

• In MODE 4, the requirement for an OPERABLE containment isolation valve system is satisfied by use of the containment isolation actuation pushbuttons.

M[ILLSTONE - UNIT 3 1-2 Amendment No. 28, 0, +86, 246

3/4.6 CONTAINMENT SYSTEMS BASES 3/4.6.1 PRIMARY CONTAINMENT 3/4.6.1.1 CONTAINMENT INTEGRITY Primary CONTAINMENT INTEGRITY ensures that the release of radioactive materials from the containment atmosphere will be restricted to those leakage paths and associated leak rates assumed in the safety analyses. This restriction, in conjunction with the leakage rate limitation, will limit the SITE BOUNDARY radiation doses to within the dose guidelines of 10 CFR 50.67 during accident conditions and the control room operators dose to within the guidelines ofRegulatory Primary CONTAINMENT INTEGRITY is required in MODES 1 through 4. This requires an OPERABLE containment automatic isolation valve system. In MODES 1, 2 and 3 this is satisfied by the automatic containment isolation signals generated by high containment pressure, low pressurizer pressure and low steamline pressure. In MODE 4 the automatic containment isolation signals generated by high containment pressure, low pressurizer pressure and low steamline pressure are not required to be OPERABLE. Automatic actuation of the containment isolation system in MODE 4 is not required because adequate time is available for plant operators to evaluate plant conditions and respond by manually operating engineered safety features components. Automatic actuation logic and actuation relays must be OPERABLE in MODE 4 to support system level manual initiation. Since the manual actuation pushbuttons portion of the containment isolation system is required to be OPERABLE in MODE 4, the plant operators can use the manual pushbuttons to rapidly position all automatic containment isolation valves to the required accident position. Therefore, the containment isolation actuation pushbuttons satisfy the requirement for an OPERABLE containment automatic isolation valve system in MODE 4.

3/4.6.1.2 CONTAINMENT LEAKAGE The limitations on containment leakage rates, as specified in the Containment Leakage Rate Testing Program, ensure that the total containment leakage volume will not exceed the value assumed in the safety analyses at the peak accident pressure, Pa. As an added conservatism, the measured overall integrated leakage rate is further limited to less than 0.75 La during performance of the periodic test to account for possible degradation of the containment leakage barriers between leakage tests.

The Limiting Condition for Operation defines the limitations on containment leakage.

The leakage rates are verified by surveillance testing as specified in the Containment Leakage Rate Testing Program, in accordance with the requirements of Appendix J. Although the LCO specifies the leakage rates at accident pressure, Pas it is not feasible to perform a test at such an exact value for pressure. Consequently, the surveillance testing is performed at a pressure greater than or equal to P. to account for test instrument uncertainties and stabilization changes. This conservative test pressure ensures that the measured leakage rates MILLSTONE - UNIT 3 B 3/4 6-1 Amendment No. -59,89, 4-1, -54, 86, 246

Serial No.05-077 Docket No. 50-423 ATTACHMENT 3 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION PROPOSED TECHNICAL SPECIFICATION CHANGES FOR IMPLEMENTATION OF ALTERNATE SOURCE TERM RETYPED PAGES DOMINION NUCLEAR CONNECTICUT. INC.

MILLSTONE POWER STATION UNIT 3

DEFINITIONS CONTAINMENT INTEGRITY 1.7 CONTAINMENT INTEGRITY shall exist when:

a. All penetrations required to be closed during accident conditions are either:

1. Capable of being closed by an OPERABLE containment automatic isolation valve systemr, or

2. Closed by manual valves, blind flanges, or deactivated automatic valves secured in their closed positions, except for valves that are opened under administrative control as permitted by Specification 3.6.3.

b. All equipment hatches are closed and sealed,

c. Each air lock is in compliance with the requirements of Specification 3.6.1.3,

d. The containment leakage rates are within the limits of the Contaimnent Leakage Rate Testing Program, and

e. The sealing mechanism associated with each penetration (e.g., welds, bellows, or O-rings) is OPERABLE.

CONTROLLED LEAKAGE 1.8 CONTROLLED LEAKAGE shall be that seal water flow supplied to the reactor coolant pump seals.

CORE ALTERATIONS 1.9 CORE ALTERATIONS shall be the movement of any fuel, sources, reactivity control components, or other components affecting reactivity within the reactor vessel with the vessel head removed and fuel in the vessel. Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe position.

DOSE EOUIVALENT 1-131 1.10 DOSE EQUIVALENT 1-131 shall be that concentration of 1-131 (microCurielgram) which alone would produce the same CDE-thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133,1-134, and I-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed under "Inhalation" in Federal Guidance No. 11, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion and Ingestion."

• In MODE 4, the requirement for an OPERABLE containment isolation valve system is satisfied by use of the containment isolation actuation pushbuttons.

MILLSTONE - UNIT 3 1-2 Amendment No. 2&, 407, 486, 26

3/4.6 CONTAINMENT SYSTEMS BASES 3/4.6.1 PRIMARY CONTAINMENT 3/4.6.1.1 CONTAINMENT INTEGRITY Primary CONTAINMENT INTEGRITY ensures that the release of radioactive materials from the containment atmosphere will be restricted to those leakage paths and associated leak rates assumed in the safety analyses. This restriction, in conjunction with the leakage rate limitation, will limit the SITE BOUNDARY radiation doses to within the dose guidelines of 10 CFR 50.67 during accident conditions and the control room operators dose to within the guidelines of GDC-19.

Primary CONTAINMENT INTEGRITY is required in MODES I through 4. This requires an OPERABLE containment automatic isolation valve system. In MODES 1,2 and 3 this is satisfied by the automatic containment isolation signals generated by high containment pressure, low pressurizer pressure and low steamline pressure. In MODE 4 the automatic containment isolation signals generated by high containment pressure, low pressurizer pressure and low steamline pressure are not required to be OPERABLE. Automatic actuation of the containment isolation system in MODE 4 is not required because adequate time is available for plant operators to evaluate plant conditions and respond by manually operating engineered safety features components. Automatic actuation logic and actuation relays must be OPERABLE in MODE 4 to support system level manual initiation. Since the manual actuation pushbuttons portion of the containment isolation system is required to be OPERABLE in MODE 4, the plant operators can use the manual pushbuttons to rapidly position all automatic containment isolation valves to the required accident position. Therefore, the containment isolation actuation pushbuttons satisfy the requirement for an OPERABLE containment automatic isolation valve system in MODE 4.

3/4.6.1.2 CONTAINMENT LEAKAGE The limitations on containment leakage rates, as specified in the Containment Leakage Rate Testing Program, ensure that the total containment leakage volume will not exceed the value assumed in the safety analyses at the peak accident pressure, Pa. As an added conservatism, the measured overall integrated leakage rate is further limited to less than 0.75 La during performance of the periodic test to account for possible degradation of the containment leakage barriers between leakage tests.

The Limiting Condition for Operation defines the limitations on containment leakage. The leakage rates are verified by surveillance testing as specified in the Containment Leakage Rate Testing Program, in accordance with the requirements of Appendix J. Although the LCO specifies the leakage rates at accident pressure, P., it is not feasible to perform a test at such an exact value for pressure. Consequently, the surveillance testing is performed at a pressure greater than or equal to Pa to account for test instrument uncertainties and stabilization changes.

This conservative test pressure ensures that the measured leakage rates MILLSTONE - UNIT 3 B 3/4 6-1 Amendment No. P, 99, 44-, 454,486, 2O6