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License Amendment Request - LAR-12-04269 License Basis Changes in Steam Generator Tube Rupture Analysis
	ML14245A408
Person / Time
Site:	Summer
Issue date:	08/27/2014
From:	Gatlin T; South Carolina Electric & Gas Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	LAR-12-04269, RC-14-0112
	Download: ML14245A408 (76)
	v • d • e

Thomas D. Gatlin Vice President, Nuclear Operations 803.345.4342 A SCANA COMPANY August 27, 2014 RC-14-0112 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

Subject:

VIRGIL C. SUMMER NUCLEAR STATION, UNIT 1 DOCKET NO. 50-395 OPERATING LICENSE NO. NPF-12 LICENSE AMENDMENT REQUEST - LAR-12-04269 LICENSE BASIS CHANGES IN STEAM GENERATOR TUBE RUPTURE ANALYSIS

Dear Sir / Madam:

Pursuant to 10 CFR 50.90, South Carolina Electric & Gas Company (SCE&G), acting for itself and as agent for South Carolina Public Service Authority, hereby requests prior NRC review and approval to change the Virgil C. Summer Nuclear Station (VCSNS) licensing basis to incorporate supplemental analyses of a steam generator tube rupture (SGTR) accident which explicitly models operator responses and quantifies their impact on the potential for steam generator overfill and offsite and control room doses. The new transient calculations supplement the VCSNS licensing basis analysis by demonstrating margin to steam generator overfill and providing input to a dose analysis that confirms the licensing basis mass transfer input is conservative.

These proposed changes have been reviewed and approved by both the Plant Safety Review Committee and the Nuclear Safety Review Committee. SCE&G has evaluated the proposed changes in accordance with 10 CFR 50.91(a) using criteria in 10 CFR 50.92(c) and has determined that the proposed changes do not involve a significant hazards consideration.

SCE&G has also determined that the proposed changes satisfy the criteria for a categorical exclusion in accordance with 10 CFR 51.22(c)(9) and does not require an environmental review.

Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment needs to be prepared for this change.

Although this request is neither exigent nor emergency, your review and approval is requested by August 27, 2015. Once approved, the amendment shall be implemented within 120 days.

A new regulatory commitment to revise the VCSNS Final Safety Analysis Report to reflect the updated SGTR analyses can be found in Attachment 3 of this submittal.

In accordance with 10 CFR 50.91, SCE&G is notifying the State of South Carolina of this LAR by transmitting a copy of this letter and enclosure to the designated State Official.

If there are any questions or if additional information is needed, please contact Mr. Bruce L.

Thompson at (803) 931-5042.

. 29065

F(803) 941-9776 Virgil C.Summer Station

Post Office Box 88

Jenkinsville, SC

Document Control Desk LAR 12-04269 CR-04-03328, CR-12-04269 RC-14-0112 Page 2 of 2 I certify under penalty of perjury that the foregoing is correct and true.

rA Executed on Thomas D. Gatlin TS/TDG/pr

Enclosure:

Evaluation of the Proposed Change : Plant Specific Input to Support Use of WCAP-10698-A : Impact Assessment for Fuel Thermal Conductivity Degradation : List of Regulatory Commitments cc: K. B. Marsh S. A. Byrne J. B. Archie N. S. Carns J. H. Hamilton J. W. Williams W. M. Cherry V. M. McCree S. A. Williams NRC Resident Inspector K. M. Sutton S. E. Jenkins P. Ledbetter NSRC RTS CR-04-03328, CR-12-04269 File 813.20 PRSF RC-14-0112

Document Control Desk Enclosure LAR 12-04269 CR-04-03328, CR-12-04269 RC-14-0112 Page 1 of 53 VIRGIL C. SUMMER NUCLEAR STATION (VCSNS) UNIT 1 DOCKET NO. 50-395 OPERATING LICENSE NO. NPF-12 ENCLOSURE EVALUATION OF SUPPLEMENTAL STEAM GENERATOR TUBE RUPTURE ANALYSES FOR VCSNS

Document Control Desk Enclosure LAR 12-04269 CR-04-03328, CR-12-04269 RC-14-0112 Page 2 of 53 ENCLOSURE Evaluation of the Proposed Change

Subject:

Supplemental SGTR Analyses for Virgil C. Summer Nuclear Station (VCSNS) 1.0

SUMMARY

DESCRIPTION 2.0 DETAILED DESCRIPTION

3.0 TECHNICAL EVALUATION

3.1 Introduction 3.2 Margin to Steam Generator Overfill Analysis 3.3 Thermal and Hydraulic Analysis for Radiological Consequences 3.4 Radiological Consequences

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria 4.2 Precedence 4.3 No Significant Hazards Consideration Determination 4.4 Conclusions

5.0 ENVIRONMENTAL CONSIDERATION

6.0 REFERENCES

ATTACHMENTS:

1. Plant Specific Input to Support Use of WCAP-10698-P-A

2. Impact Assessment for Fuel Thermal Conductivity Degradation

3. List of Regulatory Commitments

Document Control Desk Enclosure LAR 12-04269 CR-04-03328, CR-12-04269 RC-14-0112 Page 3 of 53 1.0

SUMMARY

DESCRIPTION This evaluation supports a request to update Virgil C. Summer Nuclear Station's (VCSNS) licensing basis to incorporate supplemental analyses of a Steam Generator Tube Rupture (SGTR) accident which explicitly models operator responses and quantifies their impact on the potential for steam generator overfill and offsite and control room doses. The new transient calculations supplement the VCSNS licensing basis analysis by demonstrating margin to steam generator overfill and providing input to a dose analysis that confirms the licensing basis mass transfer input is conservative.

2.0 DETAILED DESCRIPTION The new SGTR transient analyses are performed by Westinghouse using the LOFTTR2 computer program following the methodology developed in WCAP-1 0698-P-A and its Supplement 1, with the exception that a single failure was not considered, and with modifications to address NSAL-07-1 1. The methodology used in WCAP-10698-P-A has been reviewed and accepted by the Nuclear Regulatory Commission (NRC) in safety evaluation report, "Acceptance for Referencing of Licensing Topical Report WCAP-1 0698, 'SGTR Analysis Methodology to Determine Margin to Steam Generator Overfill,' December 1984," dated March 30, 1987.

New SGTR dose analyses are also performed by Westinghouse using, consistent with the Virgil C. Summer Licensing Basis, the RADTRAD computer program following the guidance of Regulatory Guide 1.183. Dose consequences are provided for the mass transfer input from the current licensing basis and new transient analyses.

The use of the WCAP-1 0698-P-A methodology, longer operator action times and resulting dose consequences are being submitted for prior NRC review and approval.

Upon approval, conforming changes will be made to the Virgil C. Summer Nuclear Station Final Safety Analysis Report (FSAR) and submitted to the NRC staff in accordance with 10 CFR 50.71 as part of the regular FSAR update process.

3.0 TECHNICAL EVALUATION

3.1 INTRODUCTION

3.1.1 SGTR Event Description A steam generator tube rupture (SGTR) results in a loss of primary coolant from the reactor coolant system (RCS) to the secondary side of the affected (also called ruptured) steam generator (SG). The event is assumed to take place at full power with the reactor coolant contaminated with fission products based on a conservative assumption regarding cladding breech. The flow of radioactive reactor coolant results in contamination of the secondary system. In the event of a coincident loss of offsite power or failure of the condenser steam dump system, discharge of activity to the environment

Document Control Desk Enclosure LAR 12-04269 CR-04-03328, CR-12-04269 RC-14-0112 Page 4 of 53 takes place via the SG power-operated relief valves (PORVs) or safety valves. Operator actions are required to terminate the primary to secondary break flow and the release of steam to the environment. The SGTR analysis is performed to assure that the radiological consequences resulting from an SGTR are within allowable guidelines.

3.1.2 Licensing Basis SGTR Analysis The licensing basis SGTR analysis for the Virgil C. Summer Nuclear Station (VCSNS), is presented in Chapter 15.4.3 of the Final Safety Analysis Report (FSAR). The SGTR analysis consists of simplified thermal-hydraulic calculations to determine the primary-to-secondary break flow and the steam released to the environment and a calculation of the radiological consequences resulting from the event.

The current SGTR thermal-hydraulic calculations do not include a computer analysis to determine the plant transient behavior following a SGTR. Instead, simplified calculations are performed, based on the expected SGTR transient response, to determine the primary-to-secondary break flow and the steam release to the environment for use in calculating the radiological consequences due to the event. Although the operator actions were not explicitly modeled in the analyses, it was implicitly assumed that the required operator actions to terminate the break flow and the steam release from the ruptured steam generator can be performed within 30 minutes. No single failure is utilized in the current SGTR licensing basis analysis.

3.1.3 Bases for the Proposed Change The supplemental SGTR analyses are provided in recognition of the following:

Improvements in the Emergency Operating Procedures (EOPs) and changes in control room protocols have resulted in longer times to complete the SGTR recovery.

During plant simulator exercises, the operating crews are taking greater than 30 minutes to terminate primary to secondary break flow following a SGTR.

The current SGTR methodology does not explicitly model operator actions and thus is not adequate to examine the effects of extended operator action times.

With an extended duration of primary to secondary break flow, the potential for SG overfill increases.

SG overfill represents a challenge to the structural integrity of the steam line, may lead to water relief out of a main steam safety/relief valve, and potentially create an un-isolable release path should a main steam safety/relief valve fail to fully reset.

As a result of the above, administrative limits on RCS activity are currently in place to ensure radiological consequences are limited should a SGTR occur pending this update to the VCSNS licensing basis to support longer operator action times.

Document Control Desk Enclosure LAR 12-04269 CR-04-03328, CR-12-04269 RC-14-0112 Page 5 of 53 As described in Section 3.1.4, the supplemental analyses model operator actions in accordance with the VCSNS Emergency Operating Procedures with times confirmed by plant-specific simulator exercises. The LOFTTR2 computer program, which is a modified version of Westinghouse's LOFTRAN program, is utilized. The modifications include the capability to model operator actions, an improved steam generator secondary side model, and a more realistic tube rupture break flow model.

3.1.4 Analyses to Supplement the VCSNS SGTR Analysis of Record The supplemental SGTR analyses are performed to demonstrate margin to SG overfill and to confirm with plant-specific analyses that the simplified calculations with a 30-minute break flow and release duration produce conservatively high dose results compared to dose results calculated when realistic operator actions are taken into account. In addition, the doses with the mass transfer data based on the simplified 30-minute break flow and release duration are recalculated, to provide a consistent basis for comparison between the dose calculations performed with the supplemental analysis mass release bases.

The supplemental SGTR transient analyses were performed by Westinghouse using the LOFTTR2 computer program following the methodology developed in WCAP-1 0698-P-A and its Supplement 1 (References 2 and 3) with the exception that a single failure is not considered. Modifications were also made to the margin-to-overfill analysis methodology consistent with WCAP-16948-P (Reference 4) to address NSAL-07-11 (Reference 5),

which identified a potential non-conservative assumption regarding the direction of conservatism for decay heat in the WCAP-1 0698-P-A (Reference 2) methodology for evaluating margin-to-overfill. The analyses include the simulation of the operator actions for recovery from an SGTR with reactor trip based on the VCSNS-specific EOPs (EOP-1.0 and EOP-4.0) which are based on the generic Emergency Response Guidelines.

The LOFTTR2 analyses are performed for the time period from the SGTR until the primary and secondary pressures equalized (break flow termination). In the margin-to-overfill (MTO) analysis presented in Section 3.2, the water volume in the secondary side of the ruptured SG was calculated as a function of time to demonstrate that overfill does not occur. The thermal and hydraulic analysis to develop input to the radiological consequences analysis is presented in Section 3.3. In this analysis, the primary-to-secondary break flow and the steam releases to the environment from the ruptured and intact SGs were calculated for use in determining the activity released to the environment. The mass releases were calculated with the LOFTTR2 computer code from the initiation of the event until break flow termination. From break flow termination until all releases are terminated, steam releases from the intact and ruptured SGs, determined from a mass and energy balance, were obtained from the VCSNS analysis of record. The mass transfer information was used to calculate the radiological consequences at the exclusion area boundary and low population zone and to the operators in the control room. The radiological consequences analysis is presented in Section 3.4.

Document Control Desk Enclosure LAR 12-04269 CR-04-03328, CR-12-04269 RC-14-0112 Page 6 of 53 3.2 MARGIN TO STEAM GENERATOR OVERFILL ANALYSIS 3.2.1 Introduction This section includes the methods, assumptions, and inputs used to analyze the margin-to-overfill (MTO) for the SGTR event as well as the sequence of events for the recovery and the calculated results for the limiting case.

3.2.2 Input Parameters and Assumptions The supplemental analyses support operation of VCSNS at rated thermal power with a full power average temperature (Tavg) operating range from 572.0 degrees Fahrenheit to 587.4 degrees Fahrenheit and with up to 10 percent SG tube plugging within the plant's Delta-75 Steam Generators. The SGTR flow is based on a SG tube inside diameter of 0.608 inches, and the available secondary side fill volume is 5543 cubic feet per SG.

Table 1 provides an overview of the major MTO analysis inputs in comparison to the WCAP-1 0698-P-A values. VCSNS specific sensitivity runs were made for the following inputs to determine the limiting values with respect to the MTO: Tavg, SG tube plugging (SGTP), Safety Injection (SI) enthalpy, Emergency Feedwater (EF) enthalpy, and decay heat. As indicated in Table 1, Safety Injection Flow was based on maximum safeguards; analysis inputs are shown in Table 2.

As described in Section 3.2.2.2, control room actions are taken to cooldown and depressurize the RCS. For cooldown, the SG PORVs on the intact SGs are credited based on a minimum as-built capacity of 803,159 Ibm/hr/valve at 1000 psia. For depressurization, one of the three pressurizer PORVs is credited based on a reference design capacity of 210,000 Ibm/hr of saturated steam at 2235 psig.

3.2.2.1 Design Basis Accident The accident modeled was a double-ended break of one SG tube located at the top of the tube sheet on the outlet (cold leg) side of the SG. The location of the break on the cold side of the SG results in higher primary-to-secondary break flow than a break on the hot side of the SG. A loss of offsite power (LOOP) is assumed to occur at the time of reactor trip, and the highest worth control assembly was assumed to be stuck in its fully withdrawn position at reactor trip.

3.2.2.2 Operator Action Times In the event of an SGTR, the operator is required to take actions to stabilize the plant and terminate the primary-to-secondary break flow. The operator actions for SGTR recovery are provided in the EOPs, and major actions were explicitly modeled in this analysis. The operator actions modeled include isolation of the ruptured SG (RSG),

Document Control Desk Enclosure LAR 12-04269 CR-04-03328, CR-12-04269 RC-14-0112 Page 7 of 53 cooldown of the Reactor Coolant System (RCS), depressurization of the RCS and termination of SI to stop primary-to-secondary break flow. These operator actions are described below.

1. Identify the ruptured SG Based on the RCS transient response, the Operator is required to determine that a SGTR is in progress and to identify the ruptured SG. Pre-trip, the key indicators will be:

decreasing RCS pressure and pressurizer level with maximum makeup, mismatches between steam and feedwater flow with increases in SG narrow range level and high steam line radiation.

2. Isolate emergency feedwater (EF) flow to a suspected ruptured steam generator For a large SGTR, a reactor trip will occur either automatically or manually by the operator. Following reactor trip, main feedwater will be isolated and steam generator water level as indicated on the narrow range instrumentation will decrease in all steam generators. This will be followed by automatic initiation of EF. Once initiated, EF flow will begin to refill the steam generators. Since primary-to-secondary leakage adds additional inventory to the ruptured steam generator, the water level will increase more rapidly in that steam generator. This response, as displayed by the SG water level instrumentation, provides confirmation of the ruptured SG and will result in operator isolation of EF flow to the ruptured steam generator once the SG tubes are covered.

Early termination of EF to the ruptured SG maximizes the SG volume available to accommodate primary-to-secondary leakage until further recovery actions are completed. The MTO analysis assumes that EF flow to the ruptured SG is isolated 6 minutes after reactor trip, by which time the ruptured SG indicated level is higher than the 26 percent narrow range span (NRS) level required for EF isolation (i.e., tubes covered) and there is a significant difference in SG level compared to the level in the intact SGs.

3. Isolate steam flow from the ruptured steam generator Once the ruptured SG has been identified, the operators continue recovery actions by isolating steam flow from the ruptured SG. In addition to minimizing radiological releases, this also enables the operators to establish a pressure differential between the ruptured and intact SGs as a necessary step toward terminating primary-to-secondary break flow. As long as the ruptured SG main steam isolation valve (MSIV) is closed as part of RSG isolation, the time the operators close the MSIV does not impact the analysis results; so, no specific time assumption is included in the analysis.

4. Cooldown the RCS using the intact SGs After isolation of the ruptured SG MSIV, the RCS is cooled rapidly to less than the saturation temperature corresponding to the ruptured SG pressure by dumping steam from only the intact SGs. This ensures adequate subcooling in the RCS after

Document Control Desk Enclosure LAR 12-04269 CR-04-03328, CR-12-04269 RC-14-0112 Page 8 of 53 depressurization to the ruptured SG pressure in subsequent actions. The margin-to-overfill analysis assumed that 15 minutes elapsed from the time of reactor trip until the cooldown was initiated. Since the analysis assumes that offsite power was lost at reactor trip, the cooldown was performed by dumping steam using the Power Operated Relief Valves (PORVs) on the intact SGs. When the EOP target temperature for cooldown (494 degrees Fahrenheit) was reached, the cooldown was terminated. The PORVs on the intact SGs were then used as necessary to maintain that temperature.

5. Depressurize the RCS to restore inventory When the cooldown is completed, SI flow will tend to increase RCS pressure until break flow matches SI flow. Consequently, SI flow must be terminated to stop primary-to-secondary break flow. However, adequate inventory must first be assured. This includes both sufficient RCS subcooling and pressurizer inventory to maintain a reliable pressurizer level indication after SI flow is stopped. Since break flow from the primary side will continue after SI flow is stopped until RCS and ruptured SG pressures equalize, an "excess" amount of inventory is needed to ensure the pressurizer level remains on span. The "excess" amount required depends on the RCS pressure and reduces to zero when the RCS pressure equals the pressure in the ruptured SG.

The analysis assumed that 4 minutes elapsed from the time the cooldown was terminated until the depressurization was initiated. Since offsite power was assumed to be lost at the time of reactor trip, the RCPs were not running and thus normal pressurizer spray was not available. Therefore, the depressurization was modeled using a pressurizer PORV.

The RCS depressurization is continued until any of the following conditions in the VCSNS EOP for SGTR are satisfied: RCS pressure is less than the ruptured SG pressure and pressurizer level is greater than 10 percent, or pressurizer level is greater than 76 percent, or RCS subcooling is less than 52.5 degrees Fahrenheit.

6. Terminate SI to stop primary-to-secondary break flow The previous actions will have established adequate RCS subcooling, a secondary side heat sink, and sufficient RCS inventory to ensure that the SI flow is no longer needed.

When these actions have been completed, the SI flow must be stopped to terminate primary-to-secondary break flow. The analysis assumed that 4 minutes elapsed from the time the depressurization was terminated until SI could be stopped. SI can be stopped provided the following conditions in the VCSNS EOP for SGTR are satisfied: RCS pressure is stable or rising, pressurizer level is greater than 10 percent, RCS subcooling is greater than 52.5 degrees Fahrenheit, and a secondary heat sink is confirmed.

Consistent with the Reference 2 methodology, the analysis does not model EOP actions after SI termination leading to break flow termination. The primary-to-secondary break flow continues after the SI flow is stopped until the RCS and ruptured SG pressures equalize. This provides a conservative margin-to-overfill result.

Document Control Desk Enclosure LAR 12-04269 CR-04-03328, CR-12-04269 RC-14-0112 Page 9 of 53 The assumed operator action times (to isolate EF flow to the ruptured SG, to initiate RCS cooldown, to initiate RCS depressurization, and to terminate SI) bound actual crew performance based on simulator exercises. Table 3 summarizes the operator action times assumed in the analysis.

3.2.3 Description of Analysis The LOFTTR2 analysis for the limiting margin-to-overfill case is described below. The limiting case with respect to margin to SG overfill considered operation at the maximum operating temperature (587.4 degrees Fahrenheit), with the minimum main feedwater temperature (435.0 degrees Fahrenheit), and the maximum SGTP level (10 percent).

The sequence of events for this transient is presented in Table 4.

Following the tube rupture, water flowed from the primary into the secondary side of the ruptured SG since the primary pressure was greater than the SG pressure. In response to this loss of coolant the RCS pressure decreased as shown in Figure 1, as the steam bubble in the pressurizer expanded. As the RCS pressure decreased due to the continued primary-to-secondary break flow, automatic reactor trip occurred on an Overtemperature-delta T (OTDT) trip signal. After reactor trip, core power rapidly decreased to decay heat levels. The turbine stop valves closed and steam flow to the turbine was terminated. The steam dump system is designed to actuate following reactor trip to limit the increase in secondary pressure, but the steam dump valves remained closed due to the loss of condenser vacuum resulting from the assumed loss of offsite power at the time of reactor trip. Thus, the energy transfer from the primary system caused the secondary side pressure to increase rapidly after reactor trip, as shown in Figure 1, until the SG PORVs (and safety valves if their setpoints are reached) lifted to dissipate the energy. As a result of the assumed loss of offsite power, main feedwater flow was assumed to be terminated and EF flow was assumed to be automatically initiated following reactor trip.

The RCS pressure and pressurizer level would continue to decrease after reactor trip as energy transfer to the secondary system shrinks the primary coolant and the tube rupture break flow continues to deplete primary inventory. The operators initiated an SI signal in anticipation that it would be required by the decrease in RCS inventory. The SI flow increased the RCS inventory and stabilized the RCS pressure. The RCS pressure then trended toward the equilibrium value where the SI flow rate would equal the break flow rate.

EF flow to the ruptured SG was isolated 6 minutes after the reactor trip. By this time the ruptured SG level was well above the level required for isolation, and well above the level in the intact SGs, as shown in Figure 3. The operators then proceed to isolate steam flow from the ruptured SG, completing isolation prior to initiating the cooldown.

Document Control Desk Enclosure LAR 12-04269 CR-04-03328, CR-12-04269 RC-14-0112 Page 10 of 53 The PORVs on the intact SGs were opened for the RCS cooldown at 15 minutes after the reactor trip. (The analysis models closure of the ruptured SG MSIV immediately before the cooldown was initiated.) The cooldown was continued until the EOP cooldown termination temperature (494 degrees Fahrenheit) was reached. When this condition was satisfied, the operator closed the SG PORVs to terminate the cooldown. This cooldown ensured that there would be adequate subcooling in the RCS after the subsequent depressurization of the RCS to the ruptured SG pressure. The reduction in the intact SG pressure required to accomplish the cooldown is shown in Figure 1. The RCS pressure also decreased during this cooldown process due to shrinkage of the RCS as shown in Figure 1.

After termination of the cooldown, a 4-minute operator action time was imposed prior to the RCS depressurization. In this analysis, the RCS depressurization was terminated when the RCS pressure was reduced to less than the ruptured SG pressure and the pressurizer level was above the required value, since there was adequate subcooling margin and the high pressurizer level setpoint was not reached. The RCS depressurization is shown in Figure 1. The depressurization reduced the break flow as shown in Figure 2.

After termination of the depressurization, a 4-minute operator action time was imposed prior to SI termination. The SI flow was terminated at this time since the requirements for SI termination were satisfied (RCS subcooling was greater than the required allowance for subcooling uncertainty, minimum EF flow was available or at least one intact SG level was in the narrow range, the RCS pressure was stable or increasing, and the pressurizer level was greater than the required value).

After SI termination the RCS pressure began to decrease as shown in Figure 1. Break flow continues for more than 16 minutes after SI flow is stopped, as shown in Table 4 and Figure 2, further reducing the margin-to-overfill. Break flow is quickly terminated once the PORVs on the intact steam generators are re-opened to maintain the core exit temperature.

3.2.4 Results The primary-to-secondary break flow rate throughout the recovery operations is presented in Figure 2. The water volume in the ruptured SG is presented as a function of time in Figure 4. The peak ruptured SG water volume is 5532 cubic feet resulting in 11 cubic feet of margin-to-overfill. Therefore, it is concluded that overfill of the ruptured SG will not occur for a design basis SGTR.

It is recognized that there are a number of conservative inputs that could be relaxed to increase the margin-to-overfill. These include reduced operator action times, improved and more detailed modeling of operator actions to reduce/control RCS pressure after SI termination, less conservative EF flow splits, realistic EF initiation delays, and crediting a portion of the steamline to show greater margin to water release.

Document Control Desk Enclosure LAR 12-04269 CR-04-03328, CR-12-04269 RC-14-0112 Page 11 of 53 3.2.5 Conclusions It is concluded that overfill of the ruptured SG will not occur for a design basis SGTR for VCSNS.

Table I Margin-to-Overfill Inputs WCAP-10698 Modeling VCSNS Parameter Direction of Conservatism SGTR MTO Analysis Initial Conditions Power Full - Power Full - Power

[Nominal + Uncertainty] [Nominal + Uncertainty]('( 2 )

RCS Pressure Minimum Minimum Pressurizer Water Level Maximum Maximum SG Secondary Mass Maximum Maximum(3)

Break Location Cold Leg Cold Leg Offsite Power Availability Offsite Power Loss of Offsite Power (LOOP) LOOP(4)

Protection Setpoints and Errors Reactor Trip Delay Minimum Minimum Turbine Trip Delay Minimum Minimum SG Relief Setpoint Minimum (PORV) Minimum (PORV)

Pressurizer Pressure Low Pressurizer Pressure OTAT Trip Setpoint Pressurizer Pressure SI Maximum Maximum(5)

Setpoint Safeguards Capacity SI Flow Rate Maximum Maximum(5)

SI Delay Minimum Minimum(5)

SI Temperature Maximum(6)

EF Flow Rate (Isolation Maximum Maximum by Operator)

EF System Delay Minimum Minimum(7)

EF Temperature Maximum Minimum(8)

Control Systems CVCS Operation, PZR Not Operating Not Operating Control Turbine Runback Mass Included Included Penalty Decay Heat Decay Heat 1971 Nominal 1979-2o(9)

Single Failure Single Failure Included Not included(1°7

Document Control Desk Enclosure LAR 12-04269 CR-04-03328, CR-12-04269 RC-14-0112 Page 12 of 53 Table I (continued)

Notes:

1) Full power Tavg was assumed to range from 572.0 to 587.4 degrees Fahrenheit.

Plant specific sensitivity studies show it is more conservative (i.e., less margin-to-overfill) to use the high value.

2) SG tube plugging was assumed to range from 0 to 10 percent. Plant specific sensitivity studies show it is more conservative (i.e., less margin-to-overfill) to use the high value.

3) The plant was assumed to be operating with the feedwater temperature at the low end of the temperature range (435 degrees Fahrenheit) since this results in a higher mass of water in the SG at the start of the event, which limits the amount of break flow and emergency feedwater (EF) that can accumulate in the ruptured SG without forcing water into the steam lines.

4) A Loss of Offsite Power (LOOP) is assumed to occur at reactor trip resulting in a coastdown of the Reactor Coolant Pumps.

5) High Head Safety Injection (HHSI) is assumed to be initiated coincident with reactor trip and loss of offsite power with zero delay. The maximum safeguards flow shown in Table 2 was utilized.

6) HHSI temperature is assumed to range from 40 to 120 degrees Fahrenheit.

Plant specific sensitivity studies show it is more conservative (i.e., less margin-to-overfill) to use the high value.

7) Emergency Feedwater (EF) is assumed to be initiated coincident with reactor trip and loss of offsite power with zero delay. All three pumps are assumed to operate, and the flow split was selected to maximize flow to the ruptured SG. No purge volume was credited to delay delivery of cold EF to the SGs as this minimizes steam release. Flow to the ruptured SG continues until it is isolated by the operator at the defined time after reactor trip, whereas flow to the intact SGs was assumed to be throttled to maintain level at or below 60 percent of the narrow range span.

8) EF temperature is assumed to range from 40 to 120 degrees Fahrenheit. Plant specific sensitivity studies show it is more conservative to use the low value.

9) Plant specific sensitivity studies show it is more conservative to use a low value of decay heat in the MTO analysis. This finding is consistent with NSAL-07-11 (Reference 5).

10) Although a deviation from WCAP-1 0698-P-A (Reference 2), neglecting the effects of a single failure is consistent with the VCSNS licensing basis for SGTR.

Document Control Desk Enclosure LAR 12-04269 CR-04-03328, CR-12-04269 RC-14-0112 Page 13 of 53 Table 2 Safety Injection Flows for Design Basis SGTR Analysis (Maximum Safeguards)

Pressure (psig) Total Injection Flow Rate (Ibm/sec) 1100 93.6 1200 91.0 1300 88.2 1400 85.3 1500 82.4 1600 79.4 1700 76.2 1800 72.9 1900 69.5 2000 65.8 2100 62.0 2200 57.9 2300 53.6 2400 48.8 2500 43.6 2600 37.6 2700 30.6 2800 21.6

Document Control Desk Enclosure LAR 12-04269 CR-04-03328, CR-12-04269 RC-14-0112 Page 14 of 53 Table 3 Operator Action Times For Design Basis SGTR Analysis 1 Action Time Operator action time to isolate EF flow Margin-to-Overfill: 6 minutes from reactor trip or to ruptured SG when level reached 26% NRS, whichever is later 2

Input to Dose: Time for the level to reach 26% NRS Operator action to isolate MSIV on Margin-to-Overfill: just prior to cooldown initiation ruptured SG 3 Input to Dose: immediately after reactor trip Operator action time to initiate cooldown Margin-to-Overfill: 15 minutes from reactor trip 4

Input to Dose: 30 minutes from reactor trip Cooldown Calculated by LOFTTR2 Operator action time to initiate 4 minutes from end of cooldown depressurization Depressurization Calculated by LOFTTR2 Operator action time to terminate SI Maximum of 4 minutes from end of depressurization following depressurization or time to satisfy termination criteria Pressure equalization Calculated by LOFTTR2

1. The operator action times modeled in the analysis have been validated by VCSNS simulator exercises (see Attachment 1) and are used as acceptance criteria for continuing training.

2. The analysis conservatively allows EF isolation to occur immediately after initiation.

3. This does not impose a requirement on the operators.

4 The time used in the input to dose analysis was extended to provide a more conservative result. This extends the duration of break flow, break flow flashing and ruptured SG releases which increases the calculated doses and strengthens the conclusion that the doses based on the simplistic calculation with 30 minute break flow and release duration are conservative.

Document Control Desk Enclosure LAR 12-04269 CR-04-03328, CR-12-04269 RC-14-0112 Page 15 of 53 Table 4 Sequence of Events for Limiting Margin-to-Overfill Analysis Event Time (seconds)

Steam Generator Tube Rupture 0 Reactor Trip (OTDT) and LOOP 84 EF Initiated 85 SI Actuated (Manually) 85 EF Flow to Ruptured SG Isolated 445 Ruptured SG MSIV Closed 986 RCS Cooldown Initiated 988 RCS Cooldown Terminated 1442 RCS Depressurization Initiated 1684 RCS Depressurization Terminated 1752 SI Terminated 1993 Break Flow Terminated 3006

Document Control Desk Enclosure LAR 12-04269 CR-04-03328, CR-12-04269 RC-14-0112 Page 16 of 53 RCS Intact SG Ruptured SG 2500 2000 U) 1500 c_

Cf) c-n C-n 1000 cl n) 500 I .I 0I I35 0 0 500 1000 1500 2000 2500 3000 3500 Time (s)

Figure 1 RCS and Secondary Pressures -

Margin-to-Overfill Analysis

Document Control Desk Enclosure LAR 12-04269 CR-04-03328, CR-12-04269 RC-14-0112 Page 17 of 53 50 50 40 40 Cn 30 30 E CD (1.)

G)-

20 20 Cn 10 10 Co C/)

CD 03 0

0

-10 i I I I I I I I I I , I I I I I I I I I I I I I I - + - 10 0 500 1000 1500 2000 2500 3000 3500 Time (s)

Figure 2 Primary to Secondary Break Flow -

Margin-to-Overfill Analysis

Document Control Desk Enclosure LAR 12-04269 CR-04-03328, CR-12-04269 RC-14-0112 Page 18 of 53 In tl SG Level R P. Upt red SG Level R.

4-0% Narrow Rong SŽpa n 100 80 60 C-c- CD 40 1 . . . . . . . . . . I . I I I . I I2 1 1 . . 0 0 500 1000 1500 2000 2500 3000 3500 Time (s)

Figure 3 Indicated SG Levels -

Margin-to-Overfill Analysis

Document Control Desk Enclosure LAR 12-04269 CR-04-03328, CR-12-04269 RC-14-0112 Page 19 of 53 Rupt LIred SC

. . . . . Ii bl e Vr I u me Avo 6000 5500 5000 4500 E

75 E

0 4000 9 3500 3000 I*I I I II I i ý I 2500 I

0 500 1000 1500 2000 2500 3000 3500 Time (s)

Figure 4 Ruptured SG Water Volume -

Margin-to-Overfill Analysis

Document Control Desk Enclosure LAR 12-04269 CR-04-03328, CR-12-04269 RC-14-0112 Page 20 of 53 3.3 THERMAL AND HYDRAULIC ANALYSIS FOR RADIOLOGICAL CONSEQUENCES 3.3.1 Introduction An additional thermal and hydraulic analysis was performed to determine the input for the radiological consequences analysis for an SGTR event modeling the operator responses and break flow continuing beyond the 30 minutes considered in the licensing basis analysis. The thermal and hydraulic analysis was performed by Westinghouse using the LOFTTR2 program and the methodology developed in References 2 and 3, with the exception that a single failure was not considered. This exception is consistent with the VCSNS licensing basis. Plant-specific parameters and operator actions consistent with the VCSNS EOPs and operator training were used. This section includes the methods, assumptions and input used to analyze the SGTR event, as well as the sequence of events for the recovery and the calculated results.

3.3.2 Input Parameters and Assumptions Plant specific sensitivity runs were made to determine the limiting Tavg and Steam Generator Tube Plugging assumptions with respect to the thermal-hydraulic analysis to determine the mass releases for the radiological consequences analysis. The limiting analysis modeled the plant operating at the high end of the Tavg range (587.4 degrees Fahrenheit) with no Steam Generator Tube Plugging. Plant specific sensitivity studies show that a higher initial secondary water mass resulted in increased steam releases and flashed break flow. Assuming feedwater temperature at the high end of the temperature range (445 degrees Fahrenheit), the initial SG mass was increased to account for uncertainty.

Other plant inputs and assumptions are the same as those described in Section 3.2.2 and Table 1 for the MTO analysis except for the following:

Consistent with WCAP-1 0698-P-A, decay heat was maximized by using the 1971+20 percent ANS decay heat model.

Consistent with WCAP-1 0698-P-A, EF temperature was assumed to be at a maximum (i.e., 120 degrees Fahrenheit).

Minimum EF flow was delivered to the SGs following reactor trip and loss of offsite power with a maximum 60-second delay. The flow split was selected to minimize flow to the ruptured SG. A maximum purge volume of 68 cubic feet was modeled to delay delivery of cold EF to the SGs and maximize steam release.

3.3.2.1 Design Basis Accident The accident modeled was a double-ended break of one SG tube located at the top of the tube sheet on the outlet (cold-leg) side of the SG. The location of the break results in higher primary to secondary break flow than a break on the hot side of the SG, as determined by Reference 2. However, the break flow flashing fraction was

Document Control Desk Enclosure LAR 12-04269 CR-04-03328, CR-12-04269 RC-14-0112 Page 21 of 53 conservatively calculated assuming all of the break flow comes from the hot-leg side of the SG consistent with Reference 3.

3.3.2.2 Operator Action Times The major operator actions required for the recovery from an SGTR are discussed in Section 3.2.2.2, and the operator action times used for the analysis are presented in Table 3. The operator action times assumed for the MTO analysis were also used for the radiological consequences analysis with the following exceptions:

1. Isolate EF flow to the ruptured SG Earlier EF isolation results in higher releases. Consequently, it was assumed that EF flow to the ruptured SG was isolated when level in the SG reached the required level that assures the SG tubes are covered. The analysis conservatively allows EF isolation to occur immediately after initiation if the level in the ruptured SG is sufficiently high.

2. Isolate steam flow from the ruptured SG Once the ruptured SG has been identified, operators continue recovery actions by isolating steam flow from the ruptured SG. The input to dose analysis assumes the MSIV on the ruptured SG is closed immediately after reactor trip so that any steam flow out of the ruptured SG is to the environment.

3. Cooldown the RCS using the intact SGs The input to dose analysis assumed that 30 minutes elapsed from the time of reactor trip until the cooldown was initiated. This additional conservatism strengthens the Section 3.4 conclusion that the licensing basis analysis provides a conservative prediction for the radiological consequences of a SGTR.

3.3.3 Description of Analysis The LOFTTR2 results for the limiting input to dose analysis are presented below. As described above, the limiting case is based on plant operation at the maximum operating temperature (587.4 degrees Fahrenheit), with the maximum main feedwater temperature (445.0 degrees Fahrenheit) and the minimum Steam Generator Tube Plugging level (0 percent).

The sequence of events for this transient is presented in Table 5. Although the times change, the general evolution of the event is as described in Section 3.2.3. Figures 5 through 10 show transient results for primary and secondary pressure, primary to secondary break flow, steam release rate to the environment, break flow flashing fraction, total flashed break flow and ruptured SG water volume.

Document Control Desk Enclosure LAR 12-04269 CR-04-03328, CR-12-04269 RC-14-0112 Page 22 of 53 3.3.4 Results The primary-to-secondary break flow rate throughout the recovery operations is presented in Figure 6. The break flow flashing fraction was calculated using the primary pressure, the ruptured loop hot leg temperature, and the ruptured SG secondary pressure. The flashing fraction is presented in Figure 8, and the total flashed break flow is presented in Figure 9. The ruptured SG PORV steam release rate is presented in Figure 7, along with the total intact SG PORV steam release rate. The ruptured steam water volume is shown in Figure 10. The water volume in the ruptured SG when the break flow is terminated is significantly less than the available SG volume.

For the time period from initiation of the accident until break flow termination, the releases were determined from the LOFTTR2 results. Since the condenser was in service until reactor trip, any radioactivity released to the environment prior to reactor trip would be through the condenser vacuum exhaust. A conservatively high pre-trip steam flow rate is considered for use in the dose analysis. After reactor trip, the releases to the environment were assumed to be from the SG PORVs.

The transfer and release data are presented in Table 6 and Table 7. The mass transfer data tabulated for use in the dose analysis conservatively added 10 percent to the mass transfer data calculated by LOFTTR2 and presented in the figures. This added conservatism strengthens the Section 3.4 conclusion that the licensing basis analysis provides a conservative prediction for the radiological consequences of a SGTR.

Low values for the RCS and SG mass data that bound the LOFTTR2 results were developed for use in the dose analysis. The data is presented in Table 8. Although the mass transfer data presented was obtained from a transient modeling maximum initial secondary side water mass, a low SG mass was selected to bound scenarios modeling the minimum initial secondary side water mass.

3.3.5 Conclusions The analysis performed to calculate the mass transfer data for input to the radiological consequences analysis were completed and the data was tabulated for the limiting cases. The results of the analysis are used as input to the radiological consequences analysis presented in Section 3.4.

Document Control Desk Enclosure LAR 12-04269 CR-04-03328, CR-12-04269 RC-14-0112 Page 23 of 53 Table 5 Sequence of Events for Limiting Input to Radiological Consequences Analysis Event Time (seconds)

Steam Generator Tube Rupture 0 Reactor Trip (OTDT) and LOOP 76 SI Actuated (Manually) 76 Ruptured SG MSIV Closed 77 EF Initiated 136 EF Flow to Ruptured SG Isolated 137 RCS Cooldown Initiated 1878 Break Flow Stops Flashing 2067 RCS Cooldown Terminated 2429 RCS Depressurization Initiated 2670 RCS Depressurization Terminated 2732 SI Terminated 2972 Break Flow Terminated 3411

Document Control Desk Enclosure LAR 12-04269 CR-04-03328, CR-12-04269 RC-14-0112 Page 24 of 53 Table 6 Break Flow and Flashed Break Flow I nteg rated I nteg rated End of Period Brak Fwl edrea Start of Period Flow Flashed Break Flow (e)(e)Break during Period (Ibm) during Period (Ibm) 0 76 3600 610 76 2067 94240 5120 2067 3411 37880 0 3411 7200 0 0 7200 28800 0 0 28800 86400 0 0

Document Control Desk Enclosure LAR 12-04269 CR-04-03328, CR-12-04269 RC-14-0112 Page 25 of 53 Table 7 Intact and Ruptured SG Steam Flow to Environment Integrated Integrated Start of Period End of Period Intact SGs Steam Flow Ruptured SG Steam Flow (sec) (sec) to Environment to Environment during Period (Ibm) during Period (Ibm) 0 76 199120* 99560*

76 3411 348300 73740 3411 7200 256400** 0 7200 28800 924900** 0 28800 86400 1200000** 0 Pre-trip steam releases are through the condenser. The values listed here are based on a bounding steam flow rate of 1310 lbm/sec/SG.

- The LOFTTR2 analysis was performed to provide transient mass transfer data until break flow termination. The operators will then continue the SGTR recovery actions and the plant is cooled and depressurized to cold shutdown conditions.

Intact SG releases after break flow termination were obtained from the licensing basis analysis.

Document Control Desk Enclosure LAR 12-04269 CR-04-03328, CR-12-04269 RC-14-0112 Page 26 of 53 Table 8 RCS and SG Mass Data Liquid Mass* (Ibm)

Initial RCS 3.4E+5 Initial Ruptured SG 9.6E+4 Initial Intact SGs (total) 1.9E+5 These are conservative minimum values. The RCS mass bounds that calculated for the range of Tavg considered. The SG masses are the nominal initial secondary water masses reduced to address uncertainty, even though the transient calculations were performed using the nominal initial secondary water masses increased to address uncertainty as noted in Section 3.3.4.

Document Control Desk Enclosure LAR 12-04269 CR-04-03328, CR-12-04269 RC-14-0112 Page 27 of 53 RCS RIntac SG Ruptured SG 2500 2500 2000. 2000 Cr)

o. 1500. 1500 c-Cn c")

Cr)

(I 1000 1000 Co CL 500 350 0 500 1000 1500 2000 2500 3000 3500 Time (s)

Figure 5 RCS and Secondary Pressure -