ML080220530

Response to Request for Additional Information Regarding Stretch Power Uprate License Amendment Request Response to Questions SCVB-07-0091 Through SCVB-07-0104

ML080220530
Person / Time
Site:	Millstone
Issue date:	01/18/2008
From:	Gerald Bichof Dominion Nuclear Connecticut
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
07-0846C
Download: ML080220530 (16)

Text

Dominion Nuclear Connecticut, Inc.

')OO(! Dominion Boulevard, Glen Allen, Virginia n060 Web !\ddress: www.dom.com January 18, 2008 U. S. Nuclear Regulatory Commission Serial No.: 07-0846C Attention: Document Control Desk NLOS/MAE: RO One White Flint North Docket No.: 50-423 11555 Rockville Pike License No.: NPF-49 Rockville, MD 20852-2378 DOMINION NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION UNIT 3 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION REGARDING STRETCH POWER UPRATE LICENSE AMENDMENT RI:QUEST RESPONSE TO QUESTIONS SCVB-07-0091 THROUGH SCVB-07-0104 Dominion Nuclear Connecticut, Inc. (DNC) submitted a stretch power uprate license amendment request (LAR) for Millstone Power Station Unit 3 (MPS3) in letters dated July 13, 2007 (Serial Nos. 07-0450 and 07-0450A), and supplemented the submittal by letters dated September 12, 2007 (Serial No. 07-04508) and December 13, 2007 (Serial No. 07-0450C).

The NRC staff forwarded requests for additional information (RAls) in October 29, 2007, November 26, 2007, and December 14, 2007 letters. DNC responded to the RAls in letters dated November 19, 2007 (Serial No. 07-0751), December 17, 2007 (Serial No. 07-0799),

January 10, 2008 (Serial Nos. 07-0834, 07-0834A, 07-0834C, and 07-0834F), January 11, 2008 (Serial Nos. 07-08348, 07-0834E, 07-0834G, and 07-0834H), and January 14, 2008 (Serial No. 07-08340). The NRC staff forwarded additional RAls in a December 20, 2007 letter. The responses to Questions SCVB-07-0091 through SCVB-07 -0104 of this RAI are provided in the attachment to this letter.

The information provided by this letter does not affect the conclusions of the significant hazards consideration discussion in the December 13, 2007 DNC letter (Serial No. 07-0450C).

Should you have any questions in regard to this submittal, please contact Ms. Margaret Earle at 804-273-2768.

Sincerely,

~t0!~~6 Vice President - Nuclear Engineering COMMONWEALTH OF VIRGINIA COUNTY OF HENRICO The foregoing document was acknowledged before me, in and for the County and Commonwealth aforesaid, today by Gerald T. Bischof, who is Vice President - Nuclear Engineering 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 /B'H-, day of ~ ,2008.

My Commission Expires: ~ J,,, Ii?'t:'B .

_ _ _ _ _ _ _ _ _ Y!JAA/)",-u.¥;!;.4.urtb

~~otaryPublic

~ARET'.IINNE" Notary Public 3~lJ.30.;l.

Commonwealth of Virginia My Commlllion Expire. Aug 31. 2008

• Serial No. 07-0846C Docket No. 50-423 SPU RAI Question SCV8-07-0091 and SCV8-07-0104 Page 2 Commitments made in this letter: None Attachment cc: U.S. Nuclear Regulatory Commission Region I Regional Administrator 475 Allendale Road King of Prussia, PA 19406-1415 Mr. J. G. Lamb U.S. Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike Mail Stop 0-881A Rockville, MD 20852-2738 Ms. C. J. Sanders Project Manager U.S. Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike Mail Stop 0-883 Rockville, MD 20852-2738 Mr. S. W. Shaffer NRC Senior Resident Inspector Millstone Power Station Director Bureau of Air Management Monitoring and Radiation Division Department of Environmental Protection 79 Elm Street Hartford, CT 06106-5127

Serial No. 07-0846C ATTACHMENT LICENSE AMENDMENT REQUEST STRETCH POWER UPRATE LICENSE AMENDMENT REQUEST RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION RESPONSE TO QUESTIONS SCVB-07-0091 THROUGH SCVB-07-0104 MILLSTONE POWER STATION UNIT 3 DOMINION NUCLEAR CONNECTICUT, INC.

Serial No. 07-0846C Docket No. 50-423 Attachment, Page 1 of 13 Containment and Ventilation Branch SCVB-07-0091 Millstone Power Station Unit 3 (MPS3) FSAR lists the following initial containment conditions selected for the subcompartment analyses:

Pressurizer Subcompartment: Temperature 100°F; Air partial pressure 8.9 pounds per square inch absolute (psia); relative humidity 10 percent.

Steam Generator and Upper Reactor Cavity Subcompartments: Temperature 120°F; air partial pressure 9.0 psia; relative humidity 50 percent.

Section 2.6.2 of Attachment 5 of the licensing amendment request (LAR) does not list the initial conditions used for subcompartment analysis. Please provide the initial conditions used for SPU subcornpartrnent analysis. In case any of these conditions deviate from the current licensing bases, provide appropriate justification for the deviation.

ONC Response Since the Stretch Power Uprate (SPU) Mass and Energy releases for the steam generator and reactor cavity subcompartments are bounded by the Mass and Energy releases assumed in the current analyses, no new subcompartment analyses were performed for the steam generator and reactor cavity subcompartments. The assumption on initial conditions (stated in MPS3 FSAR Section 6.2.1.2.1) remains unchanged.

For the pressurizer compartment, the SPU Mass and Energy releases are not bounded by the current analyses. Thus, the pressurizer compartment analyses were re-performed. Sensitivity studies were performed to determine the limiting conditions. Four cases were analyzed:

• 80°F and 9.0 psia (total pressure)

• 100°F and 9.0 psia (total pressure)

• 100°F and 10.4 psia (total pressure)

• 100°F and 14.2 psia (total pressure)

All cases assumed a limiting relative humidity of 10 percent.

The sensitivity studies bound the current assumed conditions.

SCVB-07-0092 Section 2.6.2.2.2, fifth paragraph of Attachment 5 of the LAR states that the pre-SPU analysis for short term mass and energy (M&E) release within the steam

Serial No. 07-0846C Docket No. 50-423 Attachment, Page 2 of 13 generator cubicle is conservatively based on a frictionless Moody critical flow model and bounds the SPU condition. Please justify why the pre-SPU M&E release bounds the M&E release for the SPU condition.

ONe Response As discussed in FSAR Section 6.2.1.2.3 in the paragraph labeled "Break Type Definitions and Areas," the following is stated:

"The mass and energy releases for the feedwater line single-ended split (SES) were determined by a manual calculation using the frictionless Moody correlation for a saturated liquid."

As stated in License Report Section 2.6.2.2.2, thE! same Moody correlation has been used to evaluate the impact of SPU on the short term mass and energy releases from the limiting feedwater line break. The current analysis of record for the feedwater line break assumes an initial Steam Generator (SG) pressure at hot full power of 985 psia. This bounds the SG pressure of 984 psia at hot full power at SPU conditions given in Footnote b to License Report Table 1-1.

However, the current analysis of record does not specifically address the increase in SG pressure at Hot Zero Power. To provide additional assurance that the steam generator compartment pressurization at SPU conditions is bounded for all power levels, the mass and energy releases at Hot Zero Power were calculated assuming a conservative value for the Moody critical flow taking into account a bounding value for pipe exit flow resistance (fLlD). Using a conservative value of 1.0, the mass and energy releases at Hot Zero Power at SPU conditions are bounded by the current Mass and Energy releases at Hot Full Power. Thus, it is concluded that the feedwater line break mass and energy releases at SPU conditions are bounded by the current analysis of record.

For the LOCA short term mass and energy releases, the Zaloudek subcooled break flow model is used as discussed in WCAP-8264-P-A, "Topical Report:

Westinghouse Mass and Energy Release Data for Containment Design."

SCVB-07-0093 Pressurizer spray line break loss-of-coolant accident (LOCA) M&E release data are not provided in LAR Attachment 5, Section 2.6.3. Section 2.6.3.1.2.2.3, seventh paragraph states the SPU changes "could increase the spray line mass and energy releases by as much as 3.4 percent." Please verify that the calculated SPU M&E release data for the pressurizer spray line break is within 3.4 percent from current M&E release data provided in FSAR Table 6.2-31 using the same assumptions and initial conditions, and is within the margin.

Serial No. 07-0846C Docket No. 50-423 Attachment, Page 3 of 13 ONC Response The mass and energy release data provided in Table 6.2-31 had been increased by 10% to provide margin. The data in Table 6.2-31 was subsequently used in the current pressurizer cubicle analysis of record. As a result of SPU, Westinghouse has recalculated the mass and energy releases for the pressurizer spray line break. The mass and energy releases will increase because of the lower cold leg temperature at SPU conditions. Specific calculations have been performed by Westinghouse to quantify the mass and energy releases using the same approved methodology as described in WCAP-8264-P-A, "Topical Report:

Westinghouse Mass and Energy Release Data for Containment Design." The cold leg temperature was changed to reflect SPU conditions. The SPU calculation showed an increase of 3.4%. This increase is bounded by the current 10% margin. Thus, the existing mass and energy releases provided in Table 6.2-31 of the current MPS3 FSAR remain bounding at SPU conditions.

SCVB-07-0094 Pressurizer surge line break LOCA M&E release data are not provided in LAR , Section 2.6.3. Section 2.6.3.1.2.2.~~, eighth paragraph states the SPU changes "could increase the surge line mass and energy release by as much as 15.75 percent on mass released and 11.27 percent on energy released." Please verify that the calculated SPU M&E release data for the pressurizer surge line break is within 15.75 percent on mass released and within 11.27 percent on energy released from the M&E n31ease data provided in FSAR Table 6.2-32 using the same assumptions and initial conditions.

ONe Response The mass and energy releases calculated for the surge line break at SPU conditions are not bounded by the mass and energy release data provided in FSAR Tables 6.2-32 and 6.2-32A. As stated in License Report Section 2.6.3.1.2.2.3, the calculation of surge line break short term mass and energy releases will increase as a result of the full possible range of SPU temperature conditions. The minimum hot leg temperature is lower than the current assumption due to the implementation of an allowable range for Tave including coastdown. Specific calculations have been performed by Westinghouse to quantify the mass and energy releases using the same approved methodology as described in WCAP-8264-P-A, "Topical Report: Westinghouse Mass and Energy Release Data for Containment Design." The hot leg temperature was changed to reflect SPU conditions. The SPU mass releases are 15.75% higher and the energy releases are 11.27% higher, exceeding the 10% margin data provided in FSAR Tables 6.2-32A. As a result, the pressurizer subcompartment analyses have been revised to take into account the higher mass and energy releases. The revised analyses were performed using the same approved

Serial No. 07-0846C Docket No. 50-423 Attachment, Page 4 of 13 THREED methodology as described in the current IFSAR Section 6.2.1.2.

SCVB-07-0095 FSAR Section 6.2.1.1.3.4 provides a description of the heat sink model and Tables 6.2-1 and 6.2-2 provide thermal and physical properties of the heat sinks in the current licensing basis analysis. It is stated in LAR Attachment 5, Section 2.6.1.2.2.2 that the passive heat sinks are unchanged by SPU. Please verify that the proposed SPU containment analyses use the same description, thermal and physical properties of the heat sinks as in the above FSAR section and tables.

Provide appropriate justification for any variations from the current licensing bases. Please explain how the changes in the sump strainer were accounted for in the passive heat sinks.

ONC Response The SPU containment analysis for MPS3 uses the same general grouping of passive heat sinks as outlined in Table 6.2-2 of the current FSAR to model the containment structure, internal concrete, and miscellaneous metal equipment within the containment. However, the analysis was performed utilizing the most up-to-date structural heat sink data (i.e., slab thickness and surface area) in accordance with the Millstone Mass Tracking Program. Please refer to the response to Question SCVB-07-0099 for more information on this program. The thermal properties of the heat sink material used in the SPU analysis are as follows:

Material Thermal Specific Heat Densit~

Conductivity (Btu/lbrn- OF) (Lbm/ft )

(Btu/hr-ft-OF)

Concrete 0.54 0.156 144 Stainless Steel 9.4 0.12 501 Carbon Steel 30 0.107 491 Note:

Paint is assumed to absorb no heat, but to merely present a resistance to heat transfer with a thermal conductance of 500 Btu/hr-ft2-oF.

These properties were developed in a MPS3 plant specific calculation used to support the current containment analyses and are being applied to the SPU containment analyses. However, the computer code LOCTIC used in the current analyses has material properties that cannot be changed to reflect plant specific input. For the current analyses, the MPS3 specific properties are only used for comparison to the "hardwired" values. These "hardwired" properties are given in FSAR Table 6.2-1, Thermophysical Properties of Passive Heat Sink Materials as follows:

Serial No. 07-0846C Docket No. 50-423 Attachment, Page 5 of 13 Material Thermal Specific Heat Densit¥ Conductivity (Btu/lbm- OF) (Lbm/ft )

(Btu/hr-ft-OF)

Concrete 0.8 0.16 145 Stainless Steel 10.0 0.11 490 Carbon Steel 26 0.11 490 Note:

Paint is assumed to absorb no heat, but to merely ~resent a resistance to heat transfer with a thermal conductance of 1000 Btu/hr-ft _oF.

The following table gives the relative difference [(MPS3 specific-FSAR)/FSAR]

between the two tables:

Material Thermal Specific Heat Densit¥ Conductivity (Btu/libm- OF) (Lbm/ft )

(Btu/hr-ft-OF)

Concrete -32.5% -2.5% -0.7%

Stainless Steel -6% +9% 2%

Carbon Steel 15% -2.7% 0.2%

In general, these differences are small (within round-off) and will have a negligible impact on the results. The largest difference is associated with the concrete thermal conductivity. Using the MPS3 plant specific value is conservative, since a lower thermal conductivity will reduce the heat transfer to the concrete structures and, thus, maximize the peak containment temperature and pressure.

Similarly for paint, the MPS3 GOTHIC analysis uses a lower thermal conductivity that will reduce the heat transfer through the paint and maximize the peak containment temperature and pressure.

Thus, it is concluded that applying the MPS3 specific data to the SPU analyses is appropriate and conservative.

Since the sump strainer design was not yet finalized during the SPU containment reanalysis, its contribution as an additional heat sink to the atmosphere and/or sump water was conservatively neglected in the containment integrity analyses.

For further discussions on modeling of the passive heat sinks and conductor surface heat transfer, refer to Sections 2.6.1.2.2.2.:2 and 2.6.1.2.2.2.3 of the LAR .

Serial No. 07-0846C Docket No. 50-423 Attachment, Page 6 of 13 SCVB-07-0096 LAR Attachment 5, Section 2.6.1.2.2.3, under heading "Depressurization Analysis," states that the initial containment conditions that yield slowest containment depressurization are the maximum pressure, temperature, and relative humidity. This is not consistent with the results presented in Table 2.6.1.2.2-6. The table shows that the slowest depressurization during the first hour gives a pressure of 30.2 psia for the initial condition minimum relative humidity of zero percent, and the slowest depressurization during first five hours gives a pressure of 22.6 psia for the initial condition of minimum temperature and minimum relative humidity. Please clarify or remove the inconsistency.

ONC Response The initial containment conditions identified in Section 2.6.1.2.2.3, under heading "Depressurization Analysis," were obtained from Table 3.6.1 of DOM-NAF-3-0.0-P-A (Reference 1 of LAR). In some cases, these initial conditions do not reflect the conclusions of the plant specific parametric studies that have been performed for MPS3. The pertinent sections will be revised as follows: (Note: changes are underlined.)

2.6.1.2.2.1 Input Parameters, Assumptions, and Acceptance Criteria Input Parameters and Assumptions The initial containment atmospheric conditions are chosen consistent with the guidance in NUREG-0800, Sections 6.2.1 and 62.1.1.A. The assumptions vary depending on the containment design limit that is being verified. For the MPS3 containment, the influence of the containment initial conditions, as documented in Table 3.6.1 of DOM-NAF-3-0.0-P-A (Reference 1), was confirmed by running parametric studies using the MPS3 specific GOTHIC model that assumes a Technical Specifications limit on total pressure and by varying one input while keeping the others constant. The most conservative settings for containment integrity analyses are summarized below. Note that the assurnptions of maximum temperature for the limiting LOCA Peak Pressure and minimum temperature (later in the event) and humidity for the limiting depressurization analysis differ from Table 3.6.1 of DOM-NAF-3.0.0-P-A (Reference 1). This is discussed in additional detail in Section 2.6.1.2.2.3.

[ Analysis Pressure Temperature Humidity LOCA Peak Pressure MAX MAX MIN LOCA Peak Temperature MAX MAX MAX Containment Depressurization MAX MAXIMIN MIN

SeriaINo.07-0846C Docket No. 50-423 Attachment, Page 7 of 13 2.6.1.2.2.3 Primary Containment Function Design Results Depressurization Analysis The first paragraph will be changed as follows: (Note: changes are underlined.)

The results of the containment depressurization analysis are tabulated in Table 2.6.1.2.2-6. The initial containment conditions that yield the slowest containment depressurization are the maximum pressure, temperature, and minimum relative humidity, and are provided in Section 2.6.1.2.2.1. The assumption of minimum relative humidity is different from Table 3.6.1 of DOM-NAF-3-0.0-P-A (Reference 1). Based on the MPS3 plant specific sensitivity analysis, it was observed that the assumption of minimum initial humidity is onlv slightly limiting. In addition, maximum temperature is limiting until between one and two hours into the event and minimum temperature is limiting thereafter.

SCVB-07-0097 LAR Attachment 5, Section 2.6.1.2.2.3, under heading "Peak Temperature Analysis," mentions the calculated containment temperature profile, but does not provide results for the limiting case which gives the slowest cooldown rate.

Please provide the accident case which gives the slowest cooldown rate and specify for which initial conditions it occurs.

ONC Response The discussion under the heading of "Peak Temperature Analysis" pertains to the verification of containment design limits on peak containment temperature and liner temperature. Figure 2.6.1.2.2-3 was provided to demonstrate the general behavior of typical long-term temperature responses for four large break LOCA scenarios. The construction of the long-term temperature profile for electrical equipment qualification (EEQ) analyses involves tile analysis of a large number of both small and large break LOCA scenarios with different initial conditions and/or single failures. The bounding composite temperature envelope is constructed based on the limiting scenario for the transient time of interest, which might be governed by dissimilar initial conditions, break size, location and/or single failure that were synthesized for the EEQ analysis.

For example, as reported in Section 2.6.1.2.2.3, the peak vapor temperature occurs for the hot leg breaks. The limiting containment initial conditions include maximum pressure, maximum temperature and maximum relative humidity.

However, the analogous GOTHIC case assuming initial minimum pressure is only slightly less limiting (266.1°F vs. 267.1°F). Furthermore, based on the large number of cases that have been analyzed for the development of the bounding EEQ temperature profile, it has been observed that the low initial pressure

Serial No. 07-0846C Docket No. 50-423 Attachment, Page 8 of 13 becomes limiting after the initial temperature peak. As shown on Figure 2.6.1.2.2-3, the hot leg break is limiting early followed by the pump suction breaks (Double Ended Rupture 'DER' and 3ft2 ) with the limiting initial condition of minimum pressure, maximum temperature and maximum relative humidity.

Among the single failure cases, the single ernerqency diesel generator (EDG) failure is generally the most limiting although the 480 V Motor Control Centers (MCC) failure is very slightly limiting long-term. Although the small break results are not presented on Figure 2.6.1.2.2-3, they do become limiting briefly during the long-term portion of the transient prior to activation of the recirculation sprays on low Refueling Water Storage Tank (RWST) level. Upon activation of the recirculation sprays for small break scenarios, double ended pump suction break with MCC failure becomes bounding for the remainder of the analysis with the limiting initial conditions of minimum pressure, maximum temperature and maximum relative humidity.

In summary, no single case exhibits the slowest cooldown rate over the full duration of the accident. The case giving the highest temperature during the cooldown phase will vary as time progresses throuqh the transient. Each limiting case will have a different set of initial conditions and single failure. The temperature profiles for all of the cases were overlayed and a bounding composite EEQ temperature profile was constructed.

SCVB-07-0098 LAR Attachment 5, Section 2.6.2.3, third paragraph, states that for the pressurizer surge line break, the steam generator compartment differential pressure increased by approximately 5 percent from the current licensing basis and is bounded by the current analysis results for the steam generator compartment. What is the margin in the current licensing basis and what will be the new margin with the increased differential pressure of 5 percent in the steam generator compartment?

ONC Response For the current steam generator subcompartment analysis, the limiting break is the Reactor Coolant System (RCS) 707 sq. in hot leg intrados split break.

Current analyses demonstrate that all design requirements are met assuming the mass and energy releases from this 707 sq. in. break size. The pressurizer surge line break area is 196.6 sq. in. The steam generator subcompartment is currently analyzed for a break that is approximately 3.5 times bigger than the surge line break. Thus, the current analyses will bound a 5% increase in the differential pressure for the surge line break.

SeriaINo.07-0846C Docket No. 50-423 Attachment, Page 9 of 13 SCVB-07-0099 LAR Attachment 5, Section 2.6.3.2.2.3, provides the sixteen cases analyzed for M&E release for the SPU main steam line break (MSLB) which are different from the current licensing bases MSLB break cases in FSAR Table 6.2-22. Please provide appropriate justification for the differences.

ONC Response The Main Steam Line Break (MSLB) analysis of record for MPS3 is based on a generic methodology for Model F steam generators that utilized the MARVEL computer code. This methodology is no longer used (since the mid 1980s) for analysis of the MSLB transient. The new WCAP-8822 methodology (LAR 2.6.3.2.4 Reference 1) was used in the SPU analysis.

According to WCAP-8822 the following break areas are to be evaluated:

1. A full double ended (DE) rupture at the steam generator nozzle.

2. A small DE rupture at the steam generator nozzle having an area just larger than that at which water entrainment occurs.

3. A small DE rupture at the steam generator nozzle having an area just smaller than that at which water entrainment occurs.

4. A small split rupture that will neither generate a steam line isolation signal nor result in water entrainment in the break <effluent.

Cases 1 and 4 were analyzed in the MPS3 MSLB inside containment analysis as discussed in Section 2.6.3.2.2.3 and assume no liquid entrainment. As presented in Table 2.6.1.2.3-1, modeling water entrainment lowers the break release enthalpy, which in turn lowers the peak containment pressure and temperature.

Since the consequences of Case 1 with no entrainment were shown to be acceptable for MPS3, the spectrum of cases for uprated mass and energy releases was significantly reduced. It is unnecessary to analyze Cases 2 and 3, since these cases are bounded by Case 1. Thus, the SPU analysis is consistent with the NRC-approved methodology given in WCAP-8822.

SCVB-07-0100 LAR Attachment 5, Table 2.6.5-1 provides the maximum sump water temperature for the current analysis as 260°F and the SPU value as 225°F.

What are the reasons for the difference? For the current value determined using LaCTIC model, and the SPU value determined using GOTHIC model, please verify that same assumptions and input parameters were used in the two analyses. For different assumptions and/or input parameters provide appropriate justification and indicate if any conservatism used in the current analysis was reduced in the SPU analysis.

Serial No. 07-0846C Docket No. 50-423 Attachment, Page 10 of 13 ONC Response The parameters in License Report Table 2.6.5-1 are the parameters used to evaluate the impact of SPU on the Containment Recirculation Spray System (RSS) pump net positive suction head (NPSH) analysis.

The current maximum sump temperature is determined by use of LaCTIC with inputs based upon current technical specification limits and power level. Four single active failures were assumed: (1) failure of one diesel generator; (2) failure of the MCC preventing the opening of service water outlet valves associated with two of the four RSS heat exchangers and failure of two of the four Quench Spray System (QSS) pumps to operate; (3) sequencer failure preventing operation of two of four RSS pumps and causing one RSS pump to deliver flow solely to a containment spray header while the other RSS pump supplies both spray to the containment and flow to the ECCS injection pump suctions and (4) one service water pump failure preventing cooling to two RSS heat exchangers. The limiting break determined by LaCTIC is the double-ended rupture of RCP suction piping. This analysis determined the maximum containment sump temperature of 260°F, which occurs at Containment Depressurization Actuation (CDA) initiation (2 seconds into the event). This temperature is independent of the actual start of the RSS pumps and consequently very conservative. The 260°F limit has been maintained as the maximum containment sump temperature and was used in determining the current RSS pump NPSH irrespective of modifications made to improve NPSH margin.

The SPU containment analysis utilized GOTHIC with the same or more conservative parameter limits as inputs. The single failure study also encompassed those considered in the current analysis. However, the bounding sump temperature is based upon the maximum temperature at the start of recirculation cooling. RSS initiation occurs automatically on low-low water level in the RWST and will occur no sooner than 1962~ seconds (32.7 minutes) after CDA. The GOTHIC sensitivity studies confirm that the maximum sump temperature at the start of recirculation is bounded by 225°F.

SCVB-07-0101 LAR Attachment 5, Table 2.6.6-1 provides the parameters for emergency core cooling system (ECCS) containment backpressure analysis. Please provide justification for changing the minimum initial containment pressure to 8.9 psia and temperature to 80°F from their corresponding FSAR values of 10.4 psia and 90°F.

Serial No. 07-0846C Docket No. 50-423 Attachment, Page 11 of 13 ONC Response For the ECCS containment backpressure analysis for demonstrating compliance to 10 CFR 50.46 and Appendix K, minimum initial containment pressure and temperature is conservative since a lower containment pressure and temperature will result in less ECCS injection into the core and more spillage in the containment. For the SPU analysis, more limiting initial containment conditions (lower initial pressure and temperature) were used to provide added assurance that the LOCA analysis bounds future cycles. The results of the LOCA analyses described in License Report Section 2.8.5.6.3 show that all LOCA analysis criteria are met assuming the more bounding initial containment conditions.

SCVB-07-0102 LAR Attachment 5, Section 2.6.2.3 states that "The design of the pressurizer cubicle has been evaluated for this pressure increase and determined to be acceptable. Analysis of the pressurizer cubicle walls has demonstrated that the current design pressure for the limiting wall element remains bounding with no net decrease in the design basis margin (Le., the margin between the current design pressure utilized in the structural analysis and the allowable design pressure associated with the limiting wall element)."

Please explain why there is no net decrease in the design margin between the current design pressure utilized in the structural analysis and the allowable design pressure associated with the limiting wall element?

ONC Response Calculations were performed for the floor slab and the lower pressurizer cubicle walls to assure that all structural design basis limits were met.

Floor Slab:

For the floor slab, the increase differential pressure results in an increase in the limiting moment and shear by a factor of 1.057. The increased moment and shear are compared to the current analysis to determine if the current analysis remains bounding.

Moment = current limiting moment

• 1.057 = 78.4 ft-k*1.057 = 82.9 ft-k. For a moment of 78.4 ft-k, the required rebar area is 1.:2 square inches. Applying the 1.057 multiplier to the 1.2 square inches, the rebar area is increased to 1.268 sq.

in. The current limit for rebar area is 1.27 sq. in. Thus, the current limit for the minimum rebar area remains bounding for SPU.

Serial No. 07-0846C Docket No. 50-423 Attachment, Page 12 of 13 Similarly for shear, Shear = current limiting shear x 1.057 = 25.1 k x 1.057 = 26.53 k. The current limit is 30.3 k.

Thus, it is concluded that all design requirements are met for the floor slab.

Lower pressurizer cubicle walls For the lower pressurizer cubicle walls, the revised differential pressure is multiplied by the dynamic load factor and compared to the current differential pressure used for the current analyses. For the limiting wall, the SPU calculated differential pressure is 24.5 psi. Applying the correspondinq dynamic load factor of 1.12, the differential pressure becomes 27.53 psi, which is below the current analysis of 28.8 psi.

Thus, it is concluded that all design requirements are met for the pressurizer cubicle walls.

SCVB-OJ-0103 LAR Attachment 5, Section 2.6.6.2 states, "Table 2.6.6-2 provides the structural heat sink data used in the ECCS containment backpressure boundary condition analysis. The structural heat sink data has been updated to reflect re-validation of the data and implemented design changes, including the sump strainer."

Please explain what is meant by "to reflect re-validation of the data and implemented design changes, including the sump strainer."

ONC Response The subject statement is intended to signify that the analysis was performed utilizing the most up-to-date structural heat sink data for MPS3 in accordance with Millstone Mass Tracking Program.

MPS3 Common Engineering Procedure C EN 114, "Containment Mass Tracking," provides instructions for reporting and tracking the amount and type of the identified changes to materials inside containment as well as changes in containment volume as a result of various desiqn modifications, such as the GSI-191 sump strainer installation. The containment volume and energy absorption capabilities of the structures and components are critical design characteristics for design basis containment analysis. For example, the minimum energy absorption capabilities are used to determine the post-accident peak containment pressure and temperature and to develop environmental profiles.

The maximum energy absorption capabilities are used to determine minimum containment backpressure during a LOCA (which maximizes the calculated peak

SeriaINo.07-0846C Docket No. 50-423 Attachment, Page 13 of 13 fuel clad temperature). Thus, both additions and deletions of structures and components inside containment are important and are tracked. The impact of variations in structural heat sink data on the desiqn basis analysis is evaluated for each design modification. In most cases, it can be determined from the existing analysis results that the changes in containment mass inventory have a negligible impact on the current analyses of record. This is possible since the heat structure data used in the containment analyses include margins for future design changes. If required, however, the analyses of record are reanalyzed and the associated documentation updated. For SPU, since it is necessary to revise the containment analyses, these minor changes in containment inventory currently tracked under procedure C EN 114, have been incorporated into the revised analysis to establish a new basis for the containment mass tracking and to re-establish margins.

SCVB-07-0104 NRC Generic Letter (GL) 96-06, issue number 3 states: Thermally induced overpressurization of isolated water-filled piping sections in containment could jeopardize the ability of accident-mitigating systems to perform their safety functions and could also lead to a breach of containment integrity via bypass leakage. Corrective actions may be needed to satisfy system operability requirements. "

Please verify that this issue was reconsidered for the SPU conditions, and confirm that the piping systems that penetrate! the containment which are susceptible to thermal expansion of the fluid and overpressurization, will remain within their design limits.

ONC Response The piping systems that penetrate containment have been evaluated and it was concluded they are not susceptible to thermal overpressurization. Thermal overpressurization of piping penetrating containment is prevented as a result of either 1) installed thermal relief valves, communication with components which are relief protected, having process fluid that is non-liquid, or 2) through administrative controls requiring that the affected piping be drained prior to operation at power. As part of the SPU evaluations, the design pressure and temperature of each system were evaluated. There are no changes to the design pressure and temperature of those portions of piping that penetrates containment and the thermal relief valves will continue to perform their intended function.