ML062090071
| ML062090071 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 07/21/2006 |
| From: | Crouch W Tennessee Valley Authority |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| TAC MC3743, TAC MC3744, TAC MC3812, TVA-BFN-TS-418, TVA-BFN-TS-431 | |
| Download: ML062090071 (27) | |
Text
Tennessee Valley Authority, Post Office Box 2000, Decatur, Aabama 35609-2000 July 21, 2006 TVA-BFN-TS-431 TVA-BFN-TS-418 10 CFR 50.90 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Mail Stop OWFN, Pl-35 Washington, D. C. 20555-0001 Gentlemen:
In the Matter of ) Docket Nos. 50-259 Tennessee Valley Authority ) 50-260 50-296 BROWNS FERRY NUCLEAR PLANT (BFN) - UNITS 1, 2, AND 3 -
TECHNICAL SPECIFICATIONS (TS) CHANGES TS-431 AND TS-418 -
EXTENDED POWER UPRATE (EPU) - RESPONSE TO ROUND 6 REQUEST FOR ADDITIONAL INFORMATION (TAC NOS. MC3812, MC3743, AND MC3744)
By letters dated June 28, 2004 (ADAMS Accession No. ML041840109) and June 25, 2004 (ML041840301), TVA submitted applications to the NRC for EPU of BFN Unit 1 and BFN Units 2 and 3, respectively. On June 26, 2006, the NRC staff issued the Round 6 requests for additional information (RAIs) (ADAMS Accession Nos. ML061730002 and ML061680003 for BFN Unit 1 and BFN Units 2 and 3, respectively). By letter dated July 6, 2006, TVA provided a partial response to questions regarding General Electric fuel methods that support BFN Unit l's EPU. to this letter provides responses to sixteen of the RAI questions. TVA is preparing responses to the remaining nine Round 6 RAI questions, which will be provided to the NRC in the near future.
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U.S. Nuclear Regulatory Commission Page 2 July 21, 2006 NRC RAI questions APLA.23/25 (Unit 1/Units 2 and 3),
ACVB.37/35, ACVB.39/37, ACVB.40/38, ACVB.41/39, ACVB.42/40, ACVB.49/47, ACVB.56/54, and ACVB.58/56 request detailed information pertaining to analyses associated with crediting available containment overpressure to ensure adequate net positive suction head (NPSH) for the low pressure emergency core cooling system (ECCS) pumps during analyzed events.
TVA requires further time to prepare these responses due to the issues discussed during the June 28, 2006, meeting as further clarified during the telephone conference call held on July 19, 2006. TVA is revising certain analyses and associated calculations to resolve these issues and is planning periodic meetings and phone calls with the NRC staff to provide results as they are generated. TVA will provide the response to the RAI questions following the completion of the remaining analyses.
TVA has determined that the additional information provided by this letter does not affect the no significant hazards considerations associated with the proposed TS changes.
The proposed TS changes still qualify for a categorical exclusion from environmental review pursuant to the provisions of 10 CFR 51.22(c) (9).
No new regulatory commitments have been made in this submittal.
If you have any questions regarding this letter, please contact me at (256)729-2636.
I declare under penalty of perjury that the foregoing is true and correct. Executed on this 2 1 ST day of July, 2006.
Sincerely, William D. Crouch Manager of Licensing and Industry Affairs
U.S. Nuclear Regulatory Commission Page 3 July 21, 2006
Enclosures:
- 1. Response to Round 6 Requests for Additional Information
U.S. Nuclear Regulatory Commission Page 4 July 21, 2006 cc (w. Enclosures):
State Health Officer Alabama Dept. of Public Health RSA Tower - Administration Suite 1552 P.O. Box 303017 Montgomery, AL 36130-3017 U.S. Nuclear Regulatory Commission Region II Sam Nunn Atlanta Federal Center 61 Forsyth Street, SW, Suite 23T85 Atlanta, Georgia 30303-3415 Mr. Malcolm T. Widmann, Branch Chief U.S. Nuclear Regulatory Commission Region II Sam Nunn Atlanta Federal Center 61 Forsyth Street, SW, Suite 23T85 Atlanta, Georgia 30303-8931 NRC Senior Resident Inspector Browns Ferry Nuclear Plant 10833 Shaw Road Athens, Alabama 35611-6970 NRC Unit 1 Restart Senior Resident Inspector Browns Ferry Nuclear Plant 10833 Shaw Road Athens, Alabama 35611-6970 Margaret Chernoff, Project Manager U.S. Nuclear Regulatory Commission (MS 08G9)
One White Flint, North 11555 Rockville Pike Rockville, Maryland 20852-2739 Ms. Eva A. Brown, Project Manager U.S. Nuclear Regulatory Commission (MS 08G9)
One White Flint, North 11555 Rockville Pike Rockville, Maryland 20852-2739
ENCLOSURE 1 TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT (BFN)
UNITS 1, 2, AND 3 TECHNICAL SPECIFICATIONS (TS) CHANGES TS-431 AND TS-418 -
EXTENDED POWER UPRATE (EPU) - RESPONSE TO ROUND 6 REQUESTS FOR ADDITIONAL INFORMATION (TAC NOS. MC3812, MC3743, AND MC3744)
RESPONSES TO REQUEST FOR ADDITIONAL INFORMATION This enclosure provides TVA's response to sixteen of the NRC staff's June 26, 2006, Round 6 Requests for Additional Information (ADAMS Accession Nos. ML061730002 and ML061680003 for BFN Unit 1 and BFN Units 2 and 3, respectively). Because the same information was requested for all BFN units, the responses to the two sets of NRC Round 6 RAIs are combined below for all three BFN units. The following numbering of the RAI questions and responses corresponds to Unit 1, followed by Units 2 and 3 in the format of "(x/y)."
NRC RAI APLA.22/24 It is recognized that the need to have containment accident pressure for emergency core cooling system (ECCS) net positive suction head (NPSH) should be based on a realistic analysis consistent with current probabilistic risk assessment (PRA) practices, as contrasted to a deterministic, design-basis calculation that employs excessive conservatism. Discuss which typical PRA accident sequences realistically require containment accident pressure in order to ensure that the ECCS pumps remain functional. This should include sequences currently modeled in the Browns Ferry PRA models or similar sequences, not currently modeled, that could be risk-significant if containment accident pressure is necessary and not available. This should also consider realistic fire scenarios, such as those considered in the Individual Plant Evaluation of External Events for Severe Accident Vulnerabilities study.
TVA Response to RAI APLA.22/24 Each of the BFN PRAs was reviewed in detail to specifically determine the initiation events resulting in a suppression pool temperature increase that adversely impacts maintaining adequate NPSH for the RHR and Core Spray pumps. This review identified El-i
four initiating events meeting this criterion: LOCA, ATWS, SBO, and stuck open main steam relief valve (MSRV).
As discussed in the TVA replies to NRC Request ACVB.29 (refer to TVA letters to NRC dated March 7, 2006; ADAMS ML060720248 and ML060680583 for Unit 1 and Units 2 and 3, respectively), the stuck open MSRV event is bounded by the LOCA.
Events outside the scope of the BFN PRA were also reviewed regarding the associated suppression pool heatup. This review identified that the Appendix R event results in an elevated suppression pool temperature. Therefore, the events within the scope of review for adequate availability of COP were LOCA, ATWS, SBO, and Appendix R.
NRC R*I APLA.24/26 For each accident sequence in [NRC Request APLA.23/25] above, estimate the risk associated with the need for that accident pressure (i.e., the risk above the level that would exist if the ECCS pumps could function satisfactorily without the need for containment accident pressure). While a realistic core damage frequency and large early release frequency are the desired metrics for this risk estimate, the licensee may utilize sensitivity studies, bounding analyses or qualitative arguments, where appropriate, provided all conclusions are substantially supported by the discussion.
TVA Response to RAI APLA.24/26 For the ATWS and SBO events, BFN has completed an evaluation to determine the risk impact of utilizing containment overpressure (COP) to satisfy the NPSH requirements for the RHR and core spray pumps. This evaluation was accomplished in accordance with the guidelines contained in RG 1.174, Revision 1, and used the BFN Unit 1 Probabilistic Risk Assessment (PRA) internal events model (including internal flooding). This model was revised previously to account for the core damage frequency (CDF) and large early release frequency (LERF) changes due to crediting COP during the large LOCA events. Refer to TVA's March 23, 2006, letters to the NRC (ADAMS ML060880460 and ML060880395 for Unit 1 and Units 2 and 3, respectively).
The evaluation determined that the change in CDF and LERF is very small when crediting COP for the RHR and core spray pumps during the ATWS and SBO events. The BFN report for COP credit risk assessment is provided as Enclosure 2. The report E1-2
concludes that the use of COP to satisfy the NPSH requirements for the RHR and core spray pumps (during large LOCA, ATWS, and SBO events) represents a very small change in CDF (2.4E-08/yr.)
and LERF (2.4E-08/yr.).
For fire events not modeled in the BFN PSA, a qualitative evaluation of realistic postulated fires shows that it is unlikely that fire damage would result in the need for containment overpressure to maintain adequate NPSH for the ECCS pumps. Each of the three BFN units is segregated into five distinct areas regarding the effects for a postulated fire.
These five areas are the control room, reactor building, turbine building, Units 1 and 2 diesel generator building, and Unit 3 diesel generator building. The fire protection design includes physical separation between these areas that provides reasonable assurance a realistic fire in one area will not propagate to another area. The fire protection system provides detection and suppression specifically designed, based on the contents of the area, to limit the extent of damage from a realistic fire. The consideration of a postulated fire has also provided for a design approach within each area that physically separates redundant and diverse trains of ECCS equipment so the propagation of a postulated fire affecting more than one train of equipment is remote. For a postulated fire, successful mitigation regarding NPSH for ECCS pump operation is accomplished by any one of the following scenarios:
- 1) using the balance of plant equipment to dissipate the reactor heat (results in no suppression pool temperature increase),
- 2) using two or more RHR pumps and associated RHRSW pumps in the suppression pool cooling mode of operation (suppression pool temperature is maintained low enough that containment overpressure is not required), or
- 3) adequate containment overpressure is maintained by containment isolation.
For a postulated fire in the control room, each of the units has a backup control system specifically designed, constructed, and operated to provide isolation of control room electrical circuits. The backup control system assures operation of adequate onsite AC and DC power systems and a dedicated train of equipment to shutdown the reactor, depressurize the reactor, and maintain reactor vessel water level. Two RHR pumps and associated RHRSW pumps are provided with backup control for E1-3
suppression pool cooling. With two RHR pumps in the containment cooling mode of operation, the suppression pool temperature is maintained low enough to maintain adequate NPSH for the operating ECCS pumps without containment overpressure.
For a postulated fire in the reactor building, the offsite AC power and associated control power systems needed to operate the balance of plant equipment remain available. This balance of plant equipment provides reactor pressure control and decay heat removal with the turbine control system, and reactor vessel water level control with the feedwater/condensate control systems. The availability and utilization of the balance of plant equipment avoids addition of heat to the suppression pool and the need for containment overpressure. A fire in the reactor building could adversely impact components associated with the operation of the main steam isolation valves (MSIVs),
potentially resulting in closure of one or more MSIVs. If the fire extends to the point that one of the MSIVs in each of the four steam lines is closed, the impact would be isolation of the reactor vessel from the condenser.
The fire protection system provides detection and suppression specifically designed, based on the contents of the reactor building, to limit the extent of damage from a realistic fire.
The consideration of a postulated fire has also provided for a design approach within the reactor building that physically separates redundant and diverse trains of ECCS equipment. A minimum number of main steam relief valves (MSRVs) would be available to reduce reactor vessel pressure and allow the concurrent use of the condensate system to maintain reactor vessel water level. This combination of equipment would avoid the need for containment overpressure.
For a postulated fire in the turbine building, the control room and reactor building contain the equipment required to support reactor shutdown. This fire could adversely impact the availability of offsite AC power. However, the control room and reactor building contain the equipment for supplying onsite AC and DC power for the ECCS. In this case, at least two RHR pumps and associated RHRSW pumps would be available for suppression pool cooling. With two RHR pumps in the containment cooling mode of operation, the suppression pool temperature is maintained low enough to ensure adequate NPSH for the operating ECCS pumps without containment overpressure.
For a fire in either diesel generator building, normal power would remain available to all three units. The condenser would El-4
be available as a heat sink and the reactor feedwater system would be available to supply water to the reactor vessel. The availability and utilization of the balance of plant equipment avoids addition of heat to the suppression pool and the need for containment overpressure.
NRC RAI ACVB.38/36 The current Updated Final Safety Analyses Report Table 14.6-4 shows a higher drywell volume for Case 3, the limiting case for drywell pressure and temperature, than for Cases 1, 2 and 4.
Discuss why there is a larger drywell volume assumed for this case, and whether the same assumption is made for the extended power uprate (EPU).
TVA Response to RAI ACVB.38/36 A range of drywell volumes was specified for the BFN containment analyses. The limiting (more conservative) volumes were chosen for the different containment analyses. For the short-term DBA-LOCA analyses, performed in support of the hydrodynamic loads assessment, a smaller drywell volume produced limiting results. For the analyses performed to establish a peak drywell pressure, the larger drywell volume produced limiting results. The short-term containment analyses, performed for the EPU, were also performed considering this range of drywell volumes. The EPU short-term DBA-LOCA containment analyses, performed to establish the peak drywell pressure, used a volume of 171,000 ft 3, whereas the analyses performed in support of the hydrodynamic evaluations used the smaller volume of 159,000 ft 3 .
Basis for Use of Smaller (159,000 ft 3) Drywell Volume in Hydrodynamic Loads Evaluations A smaller drywell volume produces a higher initial drywell pressurization rate, which results in a higher pool swell load.
A smaller drywell volume also results in a higher drywell-to-wetwell pressure difference, which results in a higher vent mass flux and therefore a higher vent thrust load and a higher condensation oscillation load.
Basis for Use of Larger (171,000 ft 3) Drywell Volume to Calculate Peak Drywell Pressure The peak drywell pressure is controlled by the break flow into the drywell and the vent flow out from the drywell. The break flow into the drywell is controlled by the critical break flow El-5
from the vessel, which is independent of the drywell pressure conditions. However, the vent flow out from the drywell is controlled by the drywell-to-wetwell pressure difference. A higher wetwell pressure forces a higher drywell pressure to maintain the flow out from the drywell.
The peak drywell pressure occurs after the vents have cleared and a significant portion of the drywell non-condensible mass has been transferred to the wetwell. A larger drywell volume contains more non-condensible gas, which is available for transfer to the wetwell airspace. Therefore, a larger drywell volume results in a higher wetwell airspace pressure at the time of peak drywell pressure.
Because a larger drywell volume produces a higher wetwell pressure, it also produces a greater peak drywell pressure. For this reason, the larger drywell volume Was used for Case 3 to establish the limiting condition for peak drywell pressure.
NRC RAI ACVB.43/41 Describe how the make-up of nitrogen to the drywell and wetwell atmospheres could serve as a verification of containment integrity during normal operation.
TVA Response to RAI ACVB.43/41 A discussion of nitrogen makeup monitoring was previously provided by TVA Responses to RAI SPSB-A.11 in the March 23, 2006, letters (ADAMS ML060880460 and ML060880395 for Unit 1 and Units 2 and 3, respectively):
During normal power operations, the containment is inerted with nitrogen. Per TS LCO 3.6.2.6, "The drywell pressure shall be maintained 1.1 psid above the pressure of the suppression chamber." Per TRM LCO 3.6.5, "When the primary containment is inerted the containment shall be continuously monitored for gross leakage by review of the inerting system makeup requirements. Nitrogen makeup to the primary containment, averaged over 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (corrected for drywell temperature, pressure, and venting operations), shall not exceed 542 scfh." Per TRM Surveillance Requirement (TSR) 3.6.5.1, "When the primary containment is inerted, the containment shall be continuously monitored for gross leakage by review of the inerting system makeup requirements." The E1-6
frequency for this TSR is "24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />." Satisfying these requirements would identify any pre-existing leak in the drywell portion of containment.
The following is provided as additional discussion.
During plant operation, the BFN containment is inerted with nitrogen. Drywell pressure is maintained positive with respect to the suppression pool per TS 3.6.2.6. Although normal operating pressures in the drywell and suppression pool atmosphere are less than that resulting from a Design Basis Accident, the fact that the containment is pressurized provides a reliable means of verifying that no large leak paths exist in the containment structure. Specifically, any substantial containment leak path will result in operational difficulties in maintaining positive pressure in the containment and the condition will manifest itself in an excessive nitrogen make-up rate. Monitoring for containment leakage is accomplished by monitoring the average daily nitrogen consumption used by the containment inerting system and is determined daily.
Significant containment leakage would be identified through increased nitrogen usage needed to maintain the required TS pressure.
NRC RAI ACVB.44/42 Describe the measures taken to ensure that all containment penetrations are properly isolated prior to and during operation.
TVA Response to RAI ACVB.44/42 Primary containment integrity including control of primary containment penetrations is strictly detailed by the BFN Technical Specifications (Section 3.6) and implemented via plant procedures.
- The primary containment air lock (TS 3.6.1.2) is a double door with limit switches on both doors that provide control room indication of door position.
" Primary containment isolation valves (TS 3.6.1.3) are controlled under plant procedures that provide strict valve controls. Aspects include valve line-up checklists, locking of specific valves, second party verification or independent verification of valve manipulations, and periodic surveillance of positions for accessible valves.
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Additionally, automatic containment isolation valves include position indications on the control room panels.
NRC RAI ACVB.45/43 Describe any other actions/programs which contribute to assurance that the containment is isolated.
TVA Response to RAI ACVB.45/43 Another sign of loss of integrity would be the presence of oxygen gas in containment. BFN Technical Specification (TS) 3.6.3.2 requires that the primary containment oxygen concentration be maintained less than 4.0 volume percent during reactor power operation. Oxygen monitors provide assurance that the oxygen concentration in containment is less than the TS limit. If a greater concentration of oxygen were detected, the operators would take the appropriate action in accordance with procedures.
NRC RAI ACVB.46/44 Address whether the RHR and core spray pumps can be throttled to increase available NPSH and decrease required NPSH. Discuss what, if any, guidance is provided in the emergency operating instructions (EOIs) or abnormal operating instructions regarding throttling these pumps to preserve NPSH margin during accident conditions.
TVA Response to RAI ACVB.46/44 The BFN RHR and core spray pumps can be throttled to increase available NPSH and decrease required NPSH. A discussion of EOI instructions was previously provided by the TVA Response to RAI ACVB.23 in the March 7, 2006 letter (ADAMS ML060720248 and ML060680583 for Unit 1 and Units 2 and 3, respectively). The following is provided as additional discussion.
RHR NPSH and CS NPSH limit curves are presented in the EOIs to provide the operators with guidance on NPSH margin. These curves are generated in accordance with the EPGs. Separate figures are provided for RHR and CS pumps. Each figure includes a set of curves for 0, 5, 10, and 15 psig containment pressures which correlate acceptable NPSH for varying suppression pool temperature and pump flow. Accordingly, based on containment EI-8
pressure and suppression pool temperature, the operator can determine acceptable pump flows to maintain acceptable NPSH.
NRC RAI ACVB.47/45 Discuss whether any of the units have features to automatically terminate drywell or wetwell spray. Describe the conditions under which the operator would terminate drywell and/or wetwell spray under accident conditions in accordance with the EOIs.
Address those measures put in place to prevent an operator from reducing wetwell pressure below that needed for adequate available NPSH.
TVA Response to RAI ACVB.47/45 The Unit 1 EOIs are being prepared for restart.
The BFN units do not have features to automatically terminate drywell or wetwell sprays.
The BFN EOIs do not contain any NPSH specific conditions under which operator would terminate drywell or wetwell spray. The drywell and wetwell spray approach that will be defined by the EOIs has been used as input to the containment analyses to assure consistency regarding containment spray operation. The containment analyses results demonstrate that following a LOCA, continuous containment spray will not prevent adequate available NPSH.
In response to NRC Requests ACVB.40/38 and 56/54, TVA is performing additional analyses for the Appendix R, ATWS, and SBO events. The analyses will include the use of drywell sprays where appropriate in order to assess the effect on the containment pressure response.
NRC RAI ACVB.48/46 In a letter dated September 4, 1998 letter, Tennessee Valley Authority (TVA) requested the use of containment overpressure for Units 2 and 3. The letter stated that the short term NPSH analysis assumes a double-ended recirculation pump discharge line break while the long term analysis assumes a double-ended suction line break. Address whether this is the case for the EPU analyses. Any difference in assumptions should be explained.
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TVA Response to RAI ACVB.48/46 The EPU containment calculation for NPSH evaluations, including assumptions, was previously provided by Enclosure 7 to the March 23, 2006, letters to the NRC (ADAMS ML060880460 and ML060880395 for Unit 1 and Units 2 and 3, respectively). As supplied in Table 7-2, "Assumptions for DBA LOCA Short-Term NPSH Evaluation," of that submittal, the EPU short-term NPSH analysis assumes a double-ended recirculation discharge line break. As supplied in Table 7-3, "Assumptions for DBA LOCA Long-Term NPSH Evaluation," of that submittal, the EPU long-term NPSH analysis assumes a double-ended suction line break.
An explanation of the assumptions are included in Tables 7-2 and 7-3 in Enclosure 7 to the March 23, 2006, letters.
NRC RAI ACVB.50/48 Using Figure ACVB 7-1 of the March 7, 2006, letter, explain the physical occurrences which result in (1) the reduction in the steep slope at approximately 2 seconds; (2) the small sudden increase at approximately 8 seconds; and (3) the following steep decrease. Discuss at what time the torus-to-drywell vacuum breakers to actuate.
TVA Response to RAI ACVB.50/48 (1) Reduction in the steep drywell pressure slope at approximately 2 seconds The change in the drywell pressure response near two seconds is driven by changes to the break flow into the drywell and vent flow out from the drywell during the first two seconds.
Break Flow During the initial 2 seconds, the break flow is established by the critical break flow rate at the recirculation line break location which is controlled by the break area and by the pressure and enthalpy conditions at the break. The flows into the break region are established by the critical break flow at the minimum flow areas within the flow paths upstream of the break location, and the conditions (enthalpy and pressure) in the vessel downcomer and lower plenum regions which feed the break. The flow out of the break during this initial 2 second phase is greater than the flow feeding the break since total break area is larger than the sum of the minimum flow areas upstream of the El-10
break. This flow imbalance produces a continuous drop in break flow until the flow from the lower plenum and downcomer regions feeding the break is approximately equal to the flow out of the break. This condition occurs at approximately 2 seconds.
After approximately 2 seconds, the break flow is effectively established by the flows feeding the break from the lower plenum and downcomer regions and is nearly constant until 8 seconds.
Vent Flow During the first two seconds, the drywell pressure along with the drywell-to-wetwell pressure difference increase rapidly due to an initially high break flow rate and initially low vent flow rate. The increasing drywell-to-wetwell pressure difference produces an increasingly higher vent flow rate. By approximately 2 seconds, the vent flow increases to the point where it is sufficient to maintain a near constant difference between the drywell pressure and wetwell pressure.
Combined Effect of Break Flow and Vent Flow The continuous reduction in break flow rate and increase in vent flow rate which occurs until approximately 2 seconds produces the change (reduction) in the drywell pressurization rate seen after 2 seconds.
(2) Occurrence of small sudden increase in drywell pressure at approximately 8 seconds; and (3) the following steep decrease The sudden and temporary increase in drywell pressure at approximately 8 seconds occurs when the drop in vessel inventory produces an initial change in the break flow from all liquid break flow to a two-phase mixture of mostly liquid flow with some steam. The higher energy content associated with this flow mixture initially produces a temporary spike in the drywell pressure and temperature.
However, the steam content of the break flow mixture rapidly increases thereafter. The increasingly higher steam content produces a rapid drop in the critical break flow rate and consequently the break mass flow rate to the drywell. The rapid reduction in the break mass flow rate offsets the effect of a higher enthalpy with a higher steam content on the drywell pressure response. This rapid reduction in the mass and energy release rate to the El-lI
drywell produces the steep drop in drywell pressure after approximately 8 seconds.
Time for torus-to-drywell vacuum breakers to actuate.
The time period used to generate Figure ACVB 7-1 covers the first 30 seconds of the DBA-LOCA. During this time, the drywell pressure is always greater than the wetwell airspace pressure. Therefore, for this analysis the torus-to-drywell vacuum breakers do not actuate. The long-term DBA-LOCA analysis shows that the torus-to-drywell vacuum breakers would open near 435 seconds. (See also Table ACVB. 53/51-1.)
NRC RAI ACVB.51/49 Page E1-3 of the letter dated September 4, 1998, indicates that containment pressure is only needed in the short term for the RHR pump at the maximum flow conditions and that "other pathways are available and functional without containment overpressure being relied upon." Discuss whether this is still true with the EPU NPSH analyses. If still true, elaborate on this statement.
TVA Response to RAI ACVB.51/49 In the EPU NPSH analysis, low pressure ECCS and containment heat removal pumps, RHR and core spray, require credit for containment pressure during some portion of the event scenario.
No credit for other systems is taken in the NPSH analysis.
NRC RAI ACVB.52/50 In the safety evaluation dated September 3, 1999, on the credit for containment accident pressure in determining available NPSH, TVA discussed a 10-year frequency for suppression pool cleaning.
Discuss whether suppression pool cleaning is still done on a 10-year frequency.
TVA Response to RAI ACVB.52/50 TVA designed the BFN suppression pool suction strainers assuming a frequency for cleaning the suppression pool of once every 10 years. The 10-year cleaning frequency is based on a conservatively-assumed sludge generation rate of 150 lbs. of dry sludge per year (Ref. NEDO-32686, "Utility Resolution Guidance for ECCS Suction Strainer Blockage").
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A total of 1,500 lbs. of sludge was used in the design of the BFN ECCS strainers (see TVA letter to NRC dated July 25, 1997, "NRC Bulletin No. 96-03, Potential Plugging of Emergency Core Cooling Suction Strainers By Debris In Boiling-Water Reactors").
To ensure that the 1500 lb. limit is not exceeded, TVA has established a program for maintaining cleanliness and for determining the sludge generation rate for each suppression pool.
BFN suppression pool cleaning frequency and scope are based on either the conservative 150 lbs./year generation rate or a measured and calculated sludge generation rate. Using an assumed 150 lbs./year, cleaning would be required every 10 years. Because of the need to perform cleaning to support protective coating inspections, TVA anticipates cleaning to occur at least as often as a 10-year frequency. The important design consideration is the maintenance of the total debris loading below 1500 lbs.
NRC RAI ACVB.53/51 For Figures ACVB 7-3 and ACVB 7-4 from the March 7, 2006, letter, explain the physical occurrences that produce the significant changes in the shape of the curves as a function of time.
TVA Response to RAI ACVB.53/51 Table ACVB. 53/51-1 contains a chronology of the controlling phenomena and impact on the containment responses curves shown on Figure ACVB.7-3 (Drywell and Wetwell Pressure) and Figure ACVB.7-4 (Drywell and Suppression Pool Temperature).
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Table ACVB.53/51-1 Chronology of Containment Response Curves Time Controlling Impact on Drywell Impact on Wetwell Impact on Suppression Phenomena Pressure and Drywell Pressure Pool Temperature Temperature Figures ACVB 7-3 and 7-4 Figure ACVB 7-3 Figure ACVB 7-4 0- 27 seconds Initial blowdown of Increasing drywell Increasing wetwell Increasing pool vessel. Introduction of pressure and temperature pressure mainly due to temperature due to break hot liquid break flow to due to break flow mass carryover of drywell air flow mass and energy the drywell from a and energy. to the wetwell but also transferred to the pressurized vessel, due to increase in suppression pool through wetwell airspace the vent system temperature and vapor pressure with increasing suppression pool temperature.
27-60 seconds Reactor vessel liquid Reduction in drywell Carryover of drywell air Continued increasing elevation drops to the pressure and temperature to the wetwell is suppression pool break elevation, due to reduced break complete. Wetwell temperature as break Transition from liquid flow. The reduced break pressure continues to flow mass and energy is break flow to steam flow to the drywell rise due to increasing transferred to the break flow results in a creates a temporary wetwell airspace drywell and subsequently reduced break flow with imbalance between the temperature and vapor to the suppression pool intermittent liquid and steam formation rate in pressure with increasing via the vent system.
steam flow to drywell. the drywell and the vent pool temperature.
flow out the drywell at the beginning of this period. The drywell pressure and also the drywell-to-wetwell pressure difference fall which reduces the vent flow. This trend continues until a near constant drywell-to-wetwell pressure difference (and vent flow) is established which balances the reduced break flow (and drywell steam formation rate).
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Table ACVB.53151-1 Chronology of Containment Response Curves Time Controlling Impact on Drywell Impact on Wetwell Impact on Suppression Phenomena Pressure and Drywell Pressure Pool Temperature Temperature Figures ACVB 7-3 and 7-4 Figure ACVB 7-3 Figure ACVB 7-4 60-116 seconds Resumption of continuous Increase in drywell Continued wetwell Continued increasing liquid break flow due to pressure and temperature pressure increase due to pool temperature as mass vessel reflood following due to resumption of increasing airspace and energy transfer from initiation of ECCS continuous flow of temperature and vapor the drywell continues to injection near 60 relatively hot break pressure with increasing the suppression pool.
seconds. The vessel liquid, suppression pool liquid temperature and temperature.
therefore the break flow temperature remain higher than the drywell atmosphere temperature.
116-411 seconds The reactor vessel Drywell pressure and During this time period, During this time, the pressure drops below the temperature fall due to the suppression pool pool temperature drywell pressure due to temporary stop in break water removed by ECCS continues to rise but cooling of the vessel flow. suction is greater than there is a reduction in liquid by ECCS the return vent flow to the rise rate. This is injection. The vessel the suppression pool. attributed to the halt liquid elevation does This reduces the in the break flow to the not provide sufficient suppression pool water drywell which reduces static head to maintain volume and increases the the vent flow to the break flow to drywell. wetwell airspace volume, pool.
Break flow to the The increase in the drywell stops. wetwell airspace volume produces a reduction in Vessel water level the wetwell airspace increases during this pressure time due to ECCS injection.
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Table ACVB.53/51-1 Chronology of Containment Response Curves Time Controlling Impact on Drywell Impact on Wetwell Impact on Suppression Phenomena Pressure and Drywell Pressure Pool Temperature Temperature Figures ACVB 7-3 and 7-4 Figure ACVB 7-3 Figure ACVB 7-4 411-600 seconds Near 411 seconds, the The introduction of When the WW-to-DW vacuum The suppression pool water level in the relatively colder break breakers open near 435 temperature rise rate reactor vessel has flow water into a hot seconds, wetwell air is increases again during increased to the point drywell produces a rapid transferred back to the this time due to the where there is drop in both drywell drywell. This produces resumption of break flow sufficient static head temperature and a rapid drop in the to the drywell and in the vessel to allow pressure. The drywell wetwell pressure which consequent increase in resumption of near pressure falls below the follows the drop in the the vent flow to the continuous liquid break wetwell pressure near drywell pressure. pool.
flow to the drywell. 435 seconds which Additionally, by 411 induces the opening of seconds, the vessel the WW-to-DW vacuum liquid temperature has breakers.
dropped below the drywell atmosphere temperature due to the cooling effects of ECCS injection. This produces a liquid break flow with a temperature lower than the drywell atmosphere temperature.
At 435 seconds, the wetwell (WW)- to -
drywell (DW) vacuum breakers open. This allows wetwell air to flow back to the drywell.
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Table ACVB.53/51-1 Chronology of Containment Response Curves Time Controlling Impact on Drywell Impact on Wetwell Impact on Suppression Phenomena Pressure and Drywell Pressure Pool Temperature Temperature Figures ACVB 7-3 and 7-4 Figure ACVB 7-3 Figure ACVB 7-4 600 - 1200 At 600 seconds, the LPCI The drywell spray causes The rapid wetwell The suppression pool seconds pumps are realigned from the drywell temperature depressurization temperature continues to vessel injection mode to to fall below the continues due to the rise at a similar rate containment cooling temperature of the flow of air from the as during the previous mode. The RHR heat vessel liquid break wetwell to the drywell period. During this exchangers are aligned flow. This results in a through the WW-to-DW time the effects of the in containment spray further increase in the vacuum breakers. Since RHR containment cooling mode including drywell drywell depressurization the wetwell pressure on the suppression pool and wetwell sprays. rate. follows the drywell temperature are not yet pressure, the wetwell pronounced.
The re-alignment at 600 The drywell pressure and depressurization is seconds results in the temperature continue to halted when the drywell introduction of cold drop rapidly until depressurization stops spray water to the approximately 1200 near 1200 seconds.
drywell and wetwell seconds.
airspace. This affects the drywell conditions By 1200 seconds, the directly. The effect on difference between the the wetwell conditions drywell temperature and is indirect since the colder drywell spray thermodynamic temperature is reduced equilibrium between the to the point where the suppression pool water heat being transferred and wetwell airspace is from the drywell assumed. This means atmosphere to the cold that the WW airspace drywell spray water is temperature is approximately equal to controlled by the pool the heat transferred temperature and not back to the drywell spray temperature. atmosphere by the relatively hotter vessel The realignment reduces break flow. This causes the ECCS injection flow the halt in the drywell to the vessel. This depressurization seen reduction in ECCS flow near 1200 seconds.
produces a reduction in the break flow to the I drywell.
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Table ACVB.53/51-1 Chronology of Containment Response Curves Time Controlling Impact on Drywell Impact on Wetwell Impact on Suppression Phenomena Pressure and Drywell Pressure Pool Temperature Temperature Figures ACVB 7-3 and 7-4 Figure ACVB 7-3 Figure ACVB 7-4 1200 - 18,700 During this time period The drywell pressure The wetwell pressure The suppression pool seconds (time of there is near continuous slowly increases and slowly increases during temperature slowly rises peak suppression liquid break flow to the peaks near the time of this time period due to during this time period pool drywell equal to the the peak suppression increasing airspace due to the continued temperature). ECCS injection flow pool temperature (and temperature and transfer of vessel rate. peak wetwell pressure). increasing vapor sensible energy, decay The drywell pressure pressure with increasing heat and pump heat to Vessel sensible energy follows the wetwell suppression pool the suppression pool.
and decay heat energy pressure during this temperature. The temperature rise are slowly transferred time period with the rate decreases with time to the suppression pool drywell pressure as the decay heat is in addition to pump approximately equal to reduced.
heat. The vessel the wetwell pressure sensible energy and minus the WW-to-DW During this time, the decay heat fall with vacuum breaker setpoint rate of energy addition time. pressure (0.5 psid). is greater than the rate of energy removal by the The RHR heat exchangers, The drywell temperature RHR heat exchanger.
which were actuated at is controlled by the 600 seconds, continue to combined effects of the At the time of maximum remove heat from the vessel liquid break suppression pool suppression pool. temperature and drywell temperature (near 18,700 During this time, the spray temperature during seconds), the rate of heat removal rate by the this time. The vessel heat addition and heat RHR heat exchanger is temperature (and removal from the pool less than the total heat therefore the break are equal.
addition rate to the liquid flow temperature) suppression pool. decrease with time whereas the drywell spray temperature increases with time.
These effects counteract each other resulting in a near constant drywell temperature response during this time.
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Table ACVB.53/51-1 Chronology of Containment Response Curves Time Controlling Impact on Drywell Impact on Wetwell Impact on Suppression Phenomena Pressure and Drywell Pressure Pool Temperature Temperature Figures ACVB 7-3 and 7-4 Figure ACVB 7-3 Figure ACVB 7-4 18,700 seconds to During this time, there The drywell pressure and The wetwell pressure The suppression pool end of analysis is a near continuous temperature slowly fall falls during this time temperature slowly liquid break flow to the during this time due to due to the decrease in decreases during this drywell which is equal a decreasing break the airspace temperature time due to the slow net to the ECCS injection liquid temperature and and decrease in the decrease in suppression flow rate. The vessel decreasing drywell spray wetwell vapor pressure pool energy.
temperature and temperature. with falling suppression therefore the liquid pool temperature.
break temperature slowly fall during this time due to the reduction in decay heat and the reduction in the ECCS injection water temperature.
Containment spray temperature falls with falling pool temperature.
Vessel sensible energy and decay heat energy are slowly transferred to the suppression pool in addition to pump heat. Decay heat continues to fall.
The heat exchangers continue to remove energy from the suppression pool.
During this time the heat removal rate by the RHR heat exchanger is greater than the total heat addition rate to the suppression pool.
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N'RC RAI ACVB.54/52 Table ACVB 22-1 in response to ACVB 22 from the March 7, 2006, letter, states that the licensing basis calculation of NPSH assumes no heat sinks while the realistic calculation does.
Address whether the reverse should be true to ensure conservatism. Also, see TVA reply to ACVB 27 and Table SPSB-A.1I-2 which states that not crediting heat sinks is conservative.
TVA Response to RAI ACVB.54/52 Table ACVB.22-1, the response to ACVB.27, and Table SPSB-A.11-2 are based upon the efforts taken to provide re-analyses of the suppression pool temperature response to reflect realistic values. The containment analysis case that produces the peak suppression pool temperature (licensing basis case) assumes no credit for heat sinks. This is conservative as it maximizes suppression pool temperature. The realistic assumption would be to credit heat sinks. Table SPSB-A.11-2 includes some results of analyses with credit for heat sinks.
The containment analysis case that minimizes containment pressure includes credit for heat sinks. This is conservative as it will minimize containment pressure. No effort was taken to re-analyze this containment analysis case with realistic values.
NRC RAI ACVB.55/53 Table ACVB 22-1 in response to ACVB 22 from the March 7, 2006, letter, gives values of wetwell airspace and suppression pool volume that sum-to different values for the realistic and the licensing basis values. Discuss whether the sums should be the same and equal to the total volume of the wetwell.
TVA Response to RAI ACVB.55/53 Table ACVB.22-1 was intended to provide an overview of realistic values that could be used in the NPSH calculations. The actual values that were modified in re-evaluating peak suppression pool temperatures utilizing realistic input values for selected parameters are listed in Table SPSB-A.1I-2 of our March 23, 2006, letter.
Although wetwell airspace free volume and suppression pool volume are directly linked, their influence on NPSH calculations is different. A larger value for suppression pool volume would El-20
provide a greater heat sink and result in a lower peak suppression pool temperature. Since this change would reduce the need for containment overpressure for pump NPSH, a larger realistic value based on a nominal value for this parameter was provided in Table ACVB.22-1. A corresponding smaller value for wetwell airspace free volume (to maintain the same total volume of the wetwell) would provide a smaller initial containment volume and could result in a higher available containment overpressure. Since this would not decrease the need for containment overpressure for pump NPSH, the realistic value for this parameter was not changed from the licensing basis value for this parameter in Table ACVB.22-1.
NRC RAI ACVB.57/55 The response to RAI SPSB-A.11 provided Table SPSB-A.11-2, which contains calculations of suppression pool temperature with various assumptions. The cases are identified as either GE or TVA. Describe the analytical methods used for the TVA calculations and the steps taken to ensure a meaningful comparison with SHEX.
TVA Response to RAI ACVB.57/55 The TVA analytical method employed in the sensitivity study cases reported in Table SPSB-A.11-2 uses a simple, one-dimensional model of the suppression pool, the RCS and the RHR system, developed as shown in the attached schematic diagram (Figure ACVB.57/55-1). The mass and energy balance equations for this system were solved for suppression pool temperature.
In order to ensure a meaningful comparison with design basis analysis methods, the model was benchmarked against existing GE results obtained with SHEX as shown in Case la of Table SBSB-A.1I-2. The range of input parameters used to generate additional cases is limited to small changes and basic thermodynamic principles to ensure that the benchmarking remains valid. TVA and GE case results given in Table SPSB-A.lI-2 are consistent over the range of parameters used in the sensitivity analysis.
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Figure ACVB.57155-1 Model Schematic Figure 1 Model Schematic mCS CS Pumps Tsw RIIRSW Pumps El-22
ENCLOSURE 2 TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT (BFN)
UNITS 1, 2, AND 3 TECHNICAL SPECIFICATIONS (TS) CHANGES TS-431 AND TS-418 -
EXTENDED POWER UPRATE (EPU) - RESPONSE TO ROUND 6 REQUESTS FOR ADDITIONAL INFORMATION (TAC NOS. MC3812, MC3743, AND MC3744)
BFN EPU CONTAINMENT OVERPRESSURE (COP) CREDIT RISK ASSESSMENT (SEE ATTACHED)