L-05-195, Additional Information Regarding Responses to RAI Dated July 28, 2005 in Support of License Amendment Request No. 320

Additional Information Regarding Responses to RAI Dated July 28, 2005 in Support of License Amendment Request No. 320

ML053460239
Person / Time
Site:	Beaver Valley (DPR-066)
Issue date:	12/06/2005
From:	Lash J FirstEnergy Nuclear Operating Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
L-05-195
Download: ML053460239 (16)

Text

FENOC FirstEnergy Nuclear Operating Company James H. Lash 724-682-5234 Site Vice President Fax: 724-643-8069 December 6, 2005 L-05-195 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555-0001

Subject:

Beaver Valley Power Station, Unit No. 1 BV-1 Docket No. 50-334, License No. DPR-66 Additional Information Regarding Responses to RAI Dated July 28, 2005 in Support of License Amendment Request No. 320 On April 13, 2005, FENOC submitted LAR No. 320 for BVPS Unit No. 1 by letter L 069 (Reference 1).

By letters dated July 28, 2005 and September 28, 2005, the U.S. Nuclear Regulatory Commission (NRC) issued requests for additional information (RAI) pertaining to FirstEnergy Nuclear Operating Company (FENOC) License Amendment Request (LAR)

No. 320 Replacement Steam Generators for Beaver Valley Power Station (BVPS) Unit No. 1. Responses to the July 28, 2005 and September 28, 2005 RAI questions were provided in FENOC letters L-05-137 dated August 26, 2005, and L-05-165 dated November 18, 2005, respectively. and 2 provide updated responses to RAI Questions B. I and C.5 of letter L-05-137 and to RAI Questions 5 and 6 of letter L-05-165. The changes are due to a revision of the analysis for the Steam Generator Tube Rupture (SGTR) Operational Response Case which provides thermal-hydraulic tube rupture data for use in radiological dose consequence analysis. The revised analysis provides additional operator action time for isolation of a failed open steam generator atmospheric dump valve (ADV) on the ruptured steam generator.

Note that the Operational Response Case demonstrates that the licensing basis dose analysis provided in Reference 1 is bounding. As a result, there is no change required to the dose analysis provided in Reference 1.

The updated responses provided by this transmittal have no impact on either the proposed Technical Specification changes or the no significant hazards consideration transmitted by Reference 1.

A

Beaver Valley Power Station, Unit Nos. 1 and 2 Additional Information Regarding Responses to RAI Dated July 28, 2005 in Support of License Amendment Request No. 320 L-05-195 Page 2 No new regulatory commitments are contained in this submittal. If you have questions or require additional information, please contact Mr. Gregory A. Dunn, Manager, Licensing, at 330-315-7243.

I declare under penalty of perjury that the foregoing is true and correct. Executed on December L, 2005.

Sincerely mes H. Lash

Enclosures:

1.

Updated Responses to RAI Questions B.1 and C.5 (from RA -dated July 28, 2005)

2.

Updated Responses to RAI Questions 5 and 6 (from lRAI dated September 28, 2005)

References:

iC C"

1.

FENOC Letter L-05-069, License Amendment Request 320, dated April 13, 2005.

c:

Mr. T. G. Colburn, NRR Senior Project Manager Mr. P. C. Cataldo, NRC Senior Resident Inspector Mr. S. J. Collins, NRC Region I Administrator Mr. D. A. Allard, Director BRP/DEP Mr. L. E. Ryan (BRP/DEP) of L-05-195 UPDATED RESPONSES TO RAI QUESTIONS B.1 AND C.5 (FROM RAI DATED JULY 28 2005)

Additional information is being provided to update responses to RAI Questions B.1 and C.5 relative to the steam generator tube rupture event for BVPS Unit No. 1.

B.1 Original Question (from RAI dated July 28, 2005)

With regard to Enclosure 1 of your July 8, 2005, extended power uprate (EPU) RAI response:

Item X.1, Section 5.4 of Enclosure 1 to your July 8, 2005, RAI response related to your extended power uprate LAR indicated that it would take 51 minutes to terminate the radioactive release from the ruptured tube's SG.

Please provide the radiological analyses for this scenario which demonstrates that the 30-minute isolation time of your April 13, 2005, LAR is more limiting.

Updated Response (with changes from previous response provided in italicized bold type):

The licensing basis thermal hydraulic analysis model used to determine the post accident releases and associated dose consequences at the site boundary and control room for the BVPS-1 SGTR with RSGs and at EPU conditions is a simplified model, which was the common industry standard prior to 1980. It is based on a hand calculation that predicts conservative break flows and steam releases, and utilizes a termination time of 30 minutes for the break flow and releases from the ruptured SG. The dose consequences, based on break flows and steam releases associated with the above licensing basis thermal hydraulic analysis, were submitted to NRC in Enclosure 2, Section 5.11.9.8 of the BVPS-1 RSG LAR as it was determined to be bounding.

BVPS-1 has also performed an analysis that documents the dose consequences based on values for break flows and steam releases developed by Westinghouse with LOFTTR2 computer modeling. An "operational response case" was considered that reflects a more realistic EPU SGTR transient analysis with RSGs, and takes into consideration a simulator based operator action time. The operational response case utilized a calculated break flow termination time of 66.5 minutes as opposed to the 30-minute licensing basis model.

The operational response model addresses single failure considerations, and includes margin for steam generator overfill. The operational response case was intended to be more consistent with current assumptions required of a SGTR transient analysis, as opposed to the BVPS-1 licensing basis model. Note that the 51 minute isolation time referred to in the NRC question regarding response to RAI Item X.1 pertained to an evaluation performed to support operation at the current power level.

The table below (Table B.1, Analysis Assumptions and Key Parameter

Values, BVPS-1 Steam Generator Tube Rupture - Operational Response Case) lists the key input parameters from the thermal hydraulic analysis associated with the operational response case that were utilized to develop the radiological consequences following a SGTR at BVPS-1. The description of the dose calculation assumptions and methodology presented in Enclosure 2, Section 5.11.9.8 of the RSG LAR is also applicable to the operational response case. The analysis determined that the site boundary and control room dose estimates using the licensing of L-05-195 Page 2 of 3 basis thermal hydraulic analysis model are conservative and -bound the dose estimates developed utilizing the thermal-hydraulic input data based on the operational response case.

Table B.1 Analysis Assumptions and Key Parameter Values BVPS-1 Steam Generator Tube Rupture(') - Operational Response Case Core Power Level Reactor Coolant Mass Break Flow to Faulted Steam Generator (SG)

Time of Reactor Trip Amount of Break Flow that Flashes Leakage Rate to Intact SG's Failed/Melted Fuel Percentage RCS Tech Spec Iodine & NG Concentration FRCS Equilibrium Iodine Appearance Rates Pre-Accident Iodine Spike Activity Accident Initiated Spike Appearance Rate Duration of Accident Initiated Spike Secondary System Release Parameters:

Intact SG Liquid Mass (min)

Faulted SG Liquid Mass (min)

Initial SG Liquid Mass per SG's Tech Spec Activity in SG liquid Form of All Iodine Released to the Environment via SG's Iodine Partition Coefficient (unflashed portion)

Fraction of Iodine Released (flashed portion)

Fraction of Noble Gas Released from any SG Partition Factor in Condenser Steam Flowrate to Condenser Faulted SG Steam Releases via MSSV/ADVs Intact SG Steam Releases via MSSV/ADVs Termination of Release from SGs Environmental Release Points CR Emergency Ventilation: Initiation SignaliTiming Control Room (CR) Is maintained in normal ventilation mode CR Purge Initiation (Manual) Time and Rate 2918 MWt 373,100 Ibm 0-120 sec (9,500 Ibm) 120-3988 sec (196,500 Ibm) 120 sec 0-120 sec (1810 Ibm) 120-2142.5 sec (8635.3 Ibm) 150 gpd 0 630F and 1 ATM for each SG 0%

Table 5.11.4-1 (0.35 uCVgm DE-1131)

Table 5.11.4-2 (0.351 Ci/gm DE-1131)

Table 5.11.4-2 (21 yCi/gm DE-1131) 335 times equilibrium 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> 91,953 Ibm 91,953 Ibm 91,953 Ibm Table 5.11.4-1 (0.1 pCi/gm DE-1131) 97% elemental; 3% organic 100 (all tubes submerged) 1.0 (Released without holdup) 1.0 (Released without holdup) 100 elemental iodine 1 organic iodine/Noble Gases 0-120 sec (1202 Ibm/sec from faulted SG) 0-120 see (1188 Ibm/sec per intact SG) 120 sec - 3988 sec (89p500 Ibm) 2 hr-8 hr (42,600 Ibm ))

120 sec - 3988 sec (197,400 Ibm) 3988 sec - 7200 sec (228,900 Ibm) 2 hr - 8 hr (768,700 Ibm) 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 0-120 sec (Condenser Air Ejector) 120 sec -8 hr (MSSVs/ADVs) 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> after DBA

@ 16,200 cfm (min) for 30 min Notes:

(1) Steam generator parameter values reflect the Replacement Steam Generators and Operations Assessment (2) Brief depressurization release in preparation of shutdown cooling of L-05-195 Page 3 of 3 C.5 Original Question What is the basis for assuming that the faulted SG will be isolated in 30 minutes as stated in Table 5.11.9-5?

Updated Response (with changes from previous response provided in italicized bold type):

As described in Section 5.4.1 of Enclosure 2 of the April 13, 2005 RSG LAR and discussed in the July 8, 2005 EPU RAI response to Item X.1, Section 5.4 of Enclosure 1, the BVPS-1 licensing basis analysis for the SGTR event is a mass and energy balance calculation that assumes that the steam release from the ruptured steam generator is terminated at 30 minutes after initiation of the SGTR.

A supplemental BVPS-1 SGTR operational response analysis has been performed to show that the ruptured steam generator will not overfill and to develop thermal and hydraulic SGTR input data for radiological dose analysis. The supplemental SGTR operational response analysis includes consideration of single active failures, and the timing of operator actions in accordance with plant Emergency Operating Procedures (EOPs) and demonstrated performance during simulator exercises. The supplemental analysis case for SGTR input data for radiological dose analysis showed that primary and secondary pressures are equilibrated-and that primary-to-secondary break flow and steam releases from the ruptured steam generator are terminated at 66.5 minutes after initiation of the SGTR. The supplemental analysis SGTR input data was used in supplemental radiological dose analysis that confirmed the conservatism in the licensing basis dose calculations based on the assumed 30 minute termination of break flow and steam releases from the ruptured steam generator.

This supplemental BVPS-1 SGTR operational response and radiological dose analysis confirmed that the 30 minute termination licensing basis analysis is conservative from a radiological dose standpoint even though the break flow termination time is greater than 30 minutes. This is because the 30 minute termination licensing basis analysis includes other conservative assumptions that result in higher break flow that flashes from the ruptured steam generator (which is the dominant contributor to dose consequences) than for the SGTR operational response analysis case where break flow and steam releases are terminated in 66.5 minutes.

of L-05-195 UPDATED RESPONSES TO RAI QUESTIONS 5 AND 6 (FROM RAI DATED SEPTEMBER 28. 2005)

Additional information is being provided to update responses to RAI Questions 5 and 6 relative to the steam generator tube rupture event for BVPS Unit No. 1.

5.

Original Question:

Section 5.4 of the April 13, 2005, RSG LAR states that an operational response analysis of steam generator tube rupture (SGTR) was performed for BVPS-1.

The NRC staff requests that the licensee provide a table listing the sequence of events and times from break initiation to event termination that shows operators can terminate the break flow from the ruptured SG within 51 minutes of accident Initiation for the SG replacement and extended power uprate (EPU) conditions.

Additionally, the staff requests that the licensee provide the results of the SGTR thermal-hydraulic analysis over time showing the pressurizer pressure, Intact and ruptured SGs pressures, and ruptured SG water volume for the analysis to demonstrate that no overfilling of the SG occurs.

Updated Response (with changes from previous response provided in italicized bold type):

Note, there were no changes to Table 5-2 or to the Figures 5-1 through 5-6. They are being provided for completeness.

The SGTR operational response analysis for EPU conditions with RSG included cases to provide thermal-hydraulic tube rupture data for use in radiological dose consequence analysis and for evaluation of margin to overfill. The sequence of events for these cases is shown in Tables 5-1 and 5-2, respectively. The sequence of events tables show that termination of the event (i.e., termination of primary-to-secondary, break flow) occurs at 3988 seconds (approximately 66.5 minutes) and 3578 seconds (approximately 60 minutes), respectively, for EPU conditions with RSGs. For B VPS-1, the termination time used in the radiological dose consequence analysis was Increased to allow additional operator action time for the operator to Isolate the failed-open steam generator atmospheric dump valve (ADMV on the ruptured steam generator. Figures 5-1 through 5-6 show the response of pressurizer level, pressurizer pressure, intact and ruptured SG pressure, intact loop RCS temperatures, primary-to-secondary break flow, and ruptured SG water volume for the margin to overfill case. The response to Question 6 provides information regarding the analysis performed in the 1990's for current power conditions that established a break flow termination time of 51 minutes.

of L-05-195 Page2of 11 Table 5-1 BVPS-1 Thermal-Hydraulic Analysis Sequence of Events Event (Thermal & Hydraulics for Doses)

Time (seconds)

Steam generator tube rupture 0

Reactor trip - Overtemperature Delta-T 120 Safety injection initiated 148 Isolate auxiliary feedwater to ruptured steam generator 573 Ruptured steam generator steamrline isolated 1122 Ruptured steam generator atmospheric dump valve (ADV) fails open 1124 Ruptured steam generator ADV block valve closed 1724 Reactor coolant system cooldown initiated 1868 Reactor coolant system cooldown terminated 2934 Reactor coolant system depressurization initiated 3116 Reactor coolant system depressurization terminated 3208 Safety injection terminated 3502 Steam relief to maintain sub-cooling 3570 Break flow terminated 3988

• NOTE:

Operator action time to Isolate the failed-open ADV on the ruptured steam generator was Increased to allow for additional operator action time margin.

of L-05-195 Page 3 of 11 Table 5-2 BVPS-1 Margin to Overfill Analysis Sequence of Events Event (Margin to Overfill)

Time (seconds)

Steam generator tube rupture 0

Reactor trip - Overtemperature Delta-T 106 Auxiliary feedwater initiated 121 Safety injection initiated 197 Isolate auxiliary feedwater to ruptured steam generator 515 Ruptured steam generator steamline isolated 1110 Reactor coolant system cooldown initiated 1254 Reactor coolant system cooldown terminated 2466 Reactor coolant system depressurization initiated 2646 Reactor coolant system depressurization terminated 2752 Steam relief to maintain subcooling 2952 Safety injection terminated 3046 Break flow terminated 3578 Volume (ft3)

Available ruptured steam generator secondary volume 5630 Maximum ruptured steam generator secondary volume 5602 Available ruptured steam generator secondary volume to overfill 28 of L-05-195 Page 4 of 11 Figure 5-1 Pressurizer Level Beaver Valley Unit 1 Steam Generator Tube Rupture Margin To Steam Generator Overfill 100

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• 4 of L-05-195 Page 10 of 11

6.

Original Question:

The NRC staff requests that the licensee provide the technical justification as to why a transient break flow analysis was not performed for the SGTR and why the 30-minute release assumption is conservative for the analysis when termination of the event exceeds 30 minutes.

Updated Response (with changes from previous response provided in bold type):

The BVPS-1 SGTR licensing basis analysis methodology (Section 5.4 of Enclosure 2 of FENOC Letter L-05-069) consists of a thermal-hydraulic analysis to provide tube rupture data (e.g., break flow and steam releases) as input to the BVPS-1 SGTR radiological dose consequence analysis. This licensing basis methodology includes an assumption that the break flow and steam release from the ruptured steam generator are terminated at 30 minutes.

As described in the response to NRC RAI X.1 of Enclosure 2 of the July 8, 2005 (L-05-112)

Response to RAls on the EPU LAR, a condition report was written in the 1990's that documented that more than 30 minutes was required to terminate radioactive steam release from the ruptured steam generator. At that time through the corrective action process, the break flow termination time was revised to 51 minutes. Even though the break flow termination time increased, it was determined that the primary-to-secondary break flow based on the assumptions that terminate break flow in 30 minutes actually resulted in a higher primary-to-secondary break flow than the case that terminated break flow in 51 minutes. Consequently, it was concluded that the licensing basis methodology including the assumption of break flow termination at 30 minutes was conservative with respect to a transient break flow analysis that included a longer break flow termination time.

This approach established in the 1990's was retained for the BVPS-1 SGTR analysis for EPU conditions with the RSGs. A SGTR licensing basis methodology analysis was performed including the 30-minute isolation time to provide conservative break flow and steam release data as input to the SGTR radiological dose consequence analysis. The results of this SGTR licensing basis methodology analysis are presented in Section 5.4.1 of Enclosure 2 of the RSG LAR.

To develop operator action information for operator training as well as to confirm that the licensing basis methodology analysis continues to provide conservative estimates for the radiological dose consequences following a SGTR, a SGTR operational response (transient break flow) analysis was performed for BVPS-1 at EPU conditions with the RSGs. As noted in Section 5.4 of Enclosure 2 of the April 13, 2005 (L-05-069) RSG LAR, this operational response analysis demonstrates that the BVPS-1 SGTR licensing basis analysis methodology is conservative.

The SGTR radiological dose consequence analysis using the tube rupture data (e.g., break flow and steam releases) from the BVPS-1 SGTR operational response (transient break flow) analysis was provided in response to NRC RAI B.1 of Attachment A of the August 26, 2005 (L-05-137) Response to RAls on the RSG LAR. This SGTR radiological dose consequence analysis demonstrates that the SGTR licensing basis methodology analysis with a break flow termination time of 30 minutes is more limiting than the operational response analysis with a break flow termination time of 3988 seconds (approximately 66.5 minutes), which is the break of L-05-195 Page 11 of 11 flow termination time calculated for the SGTR operational response analysis radiological dose consequence case at EPU conditions with RSGs.

The margin to overfill analysis from the operational response (transient break flow) analysis is provided in the response to Question 5. This margin to overfill analysis demonstrates that the primary-to-secondary break flow into the ruptured steam generator is terminated prior to overfilling the ruptured steam generator. The operational response analysis for the margin to overfill case at EPU conditions with RSGs shows a break flow termination time of 3578 seconds (approximately 60 minutes). The operator action times associated with this analysis have been provided as input to operator training. Additional information pertaining to operator actions and operator training has been provided in Enclosure 3 of the October 7, 2005 (L-05-154) EPU LAR Supplemental Information.

This information along with the more detailed information in the referenced RAls provide the technical justification as to why the 30-minute SGTR licensing basis methodology analysis is retained for the BVPS-1 SGTR and why the 30-minute release assumption is conservative for the radiological dose consequence analysis when termination of the event exceeds 30 minutes.

The technical justification includes a SGTR operational response (transient break flow) analysis as described in this response.

Note that the operator response times for the BVPS-1 SGTR operational response analysis have been validated as part of the emergency operating procedure (EOP) simulator validation process. One operator action time included in the SGTR operational response analysis case to provide thermal-hydraulic tube rupture data for use in radiological dose consequence analysis is the "local isolation of a failed-open atmospheric dump valve on the ruptured SG within 6.5 minutes after the valve fails open." The atmospheric dump valve (ADV) is assumed to fail open when the main steam line isolation valve for the ruptured SG is closed. This operator action time has been validated and is reflected in the BVPS Corrective Action Program. In order to afford additional operator response time for this local operator action, additional analysis has been performed to allow up to a 10-minute operator action time. The analysis results for 10-minute operator action time still support the conclusion that the BVPS-1 SGTR licensing basis methodology with a break flow termination time of 30 minutes is more limiting with respect to radiological dose consequences than the operational response analysis. The results of the rf--r r rM#r8 irI^

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