ML20217K890
| ML20217K890 | |
| Person / Time | |
|---|---|
| Site: | Waterford  |
| Issue date: | 10/18/1999 |
| From: | Dugger C ENTERGY OPERATIONS, INC. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20217K897 | List: |
| References | |
| W3F1-99-0156, NUDOCS 9910270022 | |
| Download: ML20217K890 (18) | |
Text
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Enttrgy Operations. Inc.
, =Entergy
==
17265 River Road Fax 504 739 6678 e re ad $t. Ope a ons Waterford 3 W3F1-99-0156 A4.05 PR October 18,1999 i
U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C. 20555
Subject:
Waterford 3 SES Docket No. 50-382 License No. NPF-38 Technical Specification Change Request NPF-38-224 Containment Cooling System R duction in Operable Containment Fan Coolers Gentlemen:
In accordance with 10CFR50.90, Entergy is hereby proposing to amend Operating License NPF-38 for Waterford 3 by requesting the NRC Staff review and approval of the attached changes to the Technical Specifications (TS). The attached description and safety analysis support the proposed change to the Waterford 3 TS. The proposed change modifies (S 3.6.2.2 Limiting Condition for Operation (LCO) to allow Waterford 3 to operate with two independent trains of containment cooling, j\\
consisting of one fan cooler per train, operable during modes 1,2,3, and 4. The k'
existing Waterford 3 TS 3.6.2.2 LCO requires two independent trains of containment cooling, consisting of two operable fan coolers per train, during modes 1,2,3, and 4.
A change to the TS Bases 3/4.6.2.2 has been included to support this change.
This proposed containment cooling TS change is associated with a change in the 0\\
containment cooling assumptions used in the latest containment pressure and fD temperature response analyses performed for the limiting large break Loss of Coolant Accidents (LOCA) and limiting Main Steam Line Break (MSLB) events. The latest analyses used ABB/CE NRC approved computer codes to calculate the mass and energy release data for LOCA and MSLB events, and used the GOTHIC code (Generation Of Thermal-Hydraulic Information for Containment) to analyze the containment response. The original and current licensing basis containment 9910270022 9N o'ie"
Technical Specification Change Request NPF-38-224 Containment Cooling System Reduction in Operable Containment Fan Coolers
- W3F1-99-0156 Page 2 October 18,1999 response analysis is based on the NRC approved modified version of the CONTEMPT-LT Mod 26 computer code. Through benchmarking and sensitivity studies, it is demonstrated that the use of the GOTHIC computer code for the containment response analysis application produced results that were consistent with the currentlicensing basis computer code. This proposed change has been evaluated in accordance with 10CFR50.91(a)(1), using the criteria in 10CFR50.92(c),
and it has been determined thct this request involves no significant hazards
- consideration.
The circumstances surrounding this change do not meet the NRC's criteria for
. exigent or. emergency review. Entergy Operations requests the effective date for this TS change to be within 60 days of approval.
There are no commitments associated with this request. Should you have any questions or comments concerning this request, please contact Everett Perkins at
-(504) 739-6379 or Ron Williams at (504) 739-6255.
Very truly yours,
/'
sI
' O.M. Dugger Vice President, Operations-Waterford 3 CMD/RLW/rtk Attachmer:ts:
Affidavit NPF-38-224 cc:
E.W. Merschoff, NRC Region IV C.P.. Patel, NRC-NRR J. Smith -
j N.S. Reynolds=
NRC Resident inspectors Office Administrator Radiation Protection Division (State of Louisiana)
'American Nuclear Insurers a
F:
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION in the matter of
)
)
Entergy Operations, incorporated
)
Docket No. 50-382 '
Waterford 3 Steam Electric Station
)
AFFIDAVIT Charles Marshall Dugger, being duly sworn, hereby deposes and says that he is Vice President Operations-Waterford 3 of Entergy Operations, incorporated; that he is duly authorized to sign and se with the Nuclear Regulatory Commission the attached Techrical Specification Change Request NPF-38-224; that he is familiar with the content thereof; and that the matters set forth therein are true and correct to the best of his knowledge, information and belief.
/
'~
Charles Marshall Dugger Vice President Operations 'Waterford 3 STATE OF LOUISlANA
)
) ss PARISH OF ST. CHARLES
')
Suoscribed and sworn to before me, a Notary Public in and for the Parish and State above named this a O day of O. /, /~m.
,1999.
Wr h
C Notary Public
'~~
My Commission erpires ddJ'/.
DESCRIPTION AND NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION OF PROPOSED CHANGE NPF-38-224 Summary of Proposed Changes
- Technical Specification (TS) 3.6.2.2 presently requires two independent trains of containment cooling fans to be operable, with two containment fan coolers (CFC) in each train, for modes 1,2,3, and 4. The proposed change would reduce the requirement from two CFCs to one CFC required to be operable in each train of the Containment Cooling System for modes 1,2,3, and 4. A change to the TS Bases 3/4.6.2.2 has also been included to support this change.
i The purpose of this Technical Specification Change Request is to make TS 3.6.2.2 consistent with the containment cooling assumptions of having two independent trains of containment cooling, consisting of only one operable CFC per train, as used in the latest Waterford 3 containment pressure and temperature response reanalysis for the limiting large break Loss of Coolant Accident (LOCA) and limiting Main Steam Line Break (MSLB) event. The latest analyses used ABB/CE NRC approved computer codes to calculate the mass and energy release data for LOCA and MSLB events, and used the GOTHIC code (Generation Of Thermal-Hydraulic Information for Containment) to analyze the containment response. The latest analysis for MSLB events include mass and energy release data based on a change in MSIV closure time. An analysis was performed (reference 11) to detern ine whether the MSIVs would close against the maximum expected differential pressure within the licensing basis closure time of 4 seconds under worst case conditions (includes single active failure of one of two dump valves). The analysis determined that the MSIV would not close within the closure time of 4 seconds, but would close within 6.1 seconds. The 6.1 second closure time was evaluated per Generic Letter 91-18 and found to give acceptable results for all affected accident analyses. While not in full compliance with licensing basis closure time of 4 seconds, the valves were declared operable for current and past conditions in accordance with Generic Letter 91-18. The latest analysis accounts for the slower MSIV closure tirne from 4 seconds to 7 seconds during accident conditions due to differential pressure forces acting across the valve.
Through benchmarking and sensitivity studies, it is demonstrated that the use of the GOTHIC computer code for the containment response analysis application produced results that were consistent with the current licensing basis computer code. The original and current licensing basis containment response analysis is based on the.;RC approved modified version of the CONTEMPT-LT Mod 26 computer code (hereafter referred to as CONTEMPT).
1
1 1, i Existing Specification See Attachment A.
Marked-up Specification See Attachment B -
' Proposed Specification -
- See Attachmer;t C'
- References See Attachment D'
Background
Jystem Description 1
.The function of the Containment Heat Removal System (CHRS) under accident conditions is to remove heat from the containment atmosphere to maintain the
- containment pressure and temperature at acceptable levels. The CHRS also serve to limit offsite radiation levels by reducing the pressure' differential between the containment atmosphere and the external environment, thereby decreasing the driving force for fission product leakage outside the containment.' The CHRS consists of two independent!ains of Containment Cooling System (CCS), and two independent trains
.of the Containment Spray System (CSS). The heat removal capacity of the CCS and CSS is. sufficient to keep the containment temperature and pressure below design conditions for any size break up to and including a double-ended break of the largest reactor coolant pipe. The system is also designed to mitigate the consequences of any
- size break, up to and including a double-ended break of a main steam line. The systems continue to reduce containment pressure and temperature and maintain them at acceptable levels post accident.
The CCS and CSS each consists of two redundant loops, and are' designed such that a single failure does not degrade the systems' ability to provide the required heat removal
-y.
capability.,Two of four containment fan coolers and one CSS loop are powered from an independent safety-related bus. The other two containment fan coolers and CSS
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'!oop are powered from another independent safety related bus. The loss of one bus does hot affect the ability of the Containment Heat Removal Systems to maintain containment temperatures and pressures below the design values. Since the proposed TS change only relates to.the CCS, the description of the CSS will not be discussed in this section.
1 The CCS consists of'four containment fan coolers (CFC) and a ducted air distribution system with associated instrumentation and controls. The CFCs are designed to operate during both normal plant operations and post accident conditions. Each CCS j
train consists of two CFCs both of which discharge into a common duct. The ducts
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from each train then are interconnected into a common ring header and ductwork system that distributes the discharge of the CFCs to different areas of the containment.
CFCs remove heat from the containment by drawing containment atmosphere through the CFC coils that are supplied with Component Cooling Water (CCW). The fans operate at two speeds, fast for normal operation and slow speed for accident mitigation.
During normal operation, CFCs are manually started from the main control room and
. operate at the higher speed to maintain containment temperature between 90-120 F.
The receipt of a safety injection actuation signal (SIAS) initiates the post-accident mode
.of operation. Each one of the four CFCs receives a SIAS. Upon receipt of a SlAS, the fan coolers are automatically energized to operate at low speed and the two emergency
' dis' harge dampers on the main discharge riser ducts open automatically to discharge c
cooled air to the top of the containment, without going through the ring header and
. branch ducts, to ensure proper mixing following a DBA.
)
Mass _and Energy Release Analysis Codes
. New mass and energy release data were generated for these analyses using ABB/CE codes as described below:
LOCA:
The' containment related Loss of Coolant Accident (LOCA) mass and energy release data are simulated by three different computer codes: CEFLASH-4A, FLOOD 3 and CONTRANS. The initial blowdown phase is simulatsd by the CEFLASH-4A computer code described in Reference 2. CEFLASH-4A has replaced the CEFLASH-4 code that is specified in NRC Standard Review Plans t
(SRP) (Reference 1) since CEFLASH-4 is no longer available. CEFLASH-4A is F
used in the ECCS performance analysis (10 CFR 50 Appendix K, ECCS Evaluation Models). However, many input and nodalization changes are made for containment related LOCA mass and energy calculations relative to the Appendix K model to
- ensure that the mass and energy analysis is biased conservatively.
3
The CEFLASH-4A con.puter code has been used extensively over the last 10 years for analyses conducted at almost all ABB CE plants. One recent NRC Staff
.SER that addressed the LOCA containment mass.and energy analysis was written for the Arizona Public Service 2% power up-rate effort, Reference 3.
1 The second phase of the LOCA is the reflood/ post-reflood. These phases of the
. LOCA are simulated in accordance wMh the NRC Staff approved FLOODMO_D2
. methodology specified in Reference 1. The FLOOD 3 computer code is used to implement this methodology. The FLOOD 3 code replaced the original FLOOD-MOD 2' code, which was used for the reflood phase, and the FROTH code, which was used for the post-reflood phase.
The FLOOD 3 code has been used exclusively for the last 20 years for the majority of the ABB CE plants. The Safety Evaluation Reports (SER) of References 3 and 4 are two examples of its approval for use in containment mass & energy-analyses.
The CONTRANS computer code (Reference 5) was used for the simulation of the long-term cooldown phase of the LOCA. This code calculated the long term mass
& energy release data. This data along with the short term mass & energy release data from CEFLASH-4A and FLOOD 3 are used in the GOTHIC computer code to calculate the containment pressure and temperature response.
CONTRANS calculates long term mass & energy release data accounting for both RCS and SG stored energy, as well as decay heat. The use of the ;CONTRANS' code for Waterford 3 was in general accordance with its SER (Reference 6). The one exception'was the use of a more realistic decay heat curve. Instead of the
- hardwired "Klotz Curve", which was a conservative predecessor to the ANS 1971 Standard, the 1979 + 2 SIGMA curve was used. Although somewhat more realistic than the base CONTRANS decay heat curve, the 1979 + 2 SIGMA curve accounts for long term actinides and heavy metals which are not included in the BTP ASB 9-2 curve presented in the NRC's Standard Review Plan (SRP).
MSLB:
The SGNill computer code is used for the calculation of mass & energy release data following a Main Steam Line Break. This computer code is the current '
Waterford 3 licensing basis code used for the secondary system pipe break analysis. This computer code is cited as an acceptable modeling tool for this application in Section 6.2.1;4 of the SRP (Reference 1). The SGNill computer 1 code was. submitted to the NRC Staff for approval (Reference 7) and the approval
- documented in the SER per Refererce 4.
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Containment Pressure-Temperature Response Computer Codes The modified version of the CONTEMPT computer code is it,uriginal and current licensing basis code used to analyze Waterford 3 containment pressure and i
temperature transients.
{
4 GOTHIC is the computer code used to perform the latest containment pressure and temperature response analyses.' GOTHIC is a general purpose thermal-hydraulic computer program for design, licensing, safety and operating analysis of nuclear power plant containment and other confinement buildings. GOTHIC has many applications including evaluation of containment and containment sub-compartments pressure and l
temperature response to the full spectrum of high energy line beaks, calculation of room temperature response due to failed or degraded room cooling systems, calculation of temperature profile for equipment qualification, inadvertent system laitiation, and degradation or failure of engineered safety features.
Numerical Applications, Inc. (NAI) developed the code for the Electric Power Research Institute (EPRI). GOTHIC is fully qualified under the NAl QA program which conforms to the requirements of 10CFR50 Appendix B with error reporting in accordance with 10CFR21. NAl has qualified the GOTHIC code for its intended purpose. The qualification consisted of a comparison of GOTHIC predictions to solutions of analytic problems and to experimental data for containment applications. The code qualification is documented in a code Qualification Report prepared by NAl for EPRI. The qualification objective is to demonstrate the applicability of GOTHIC for use as a best-estimate containment analysis code. In addition to the above qualification efforts, GOTHIC has been extensively compared to other codes such as CONTEMPT.
Entergy Operations, Inc. (EOI) plans to use the GOTHIC code in-house to replace the vendor CONTEMPT code for containment analysis to establish more control over the analysis. For Waterford 3 applications, the code is installed and verified against NAl supplied standard problem results and analysis is performed and documented under
- the EOl quality assurance progreir werning nuclear quality related activities.
The Waterford 3 GOTHIC containment niodels were developed for Loss of Coolant Accidents and Main Steam Line Breaks. These models were benchmatked againct the current Waterford 3 licensing containment response analysis code, CONTEMPT, with good agreement between the two code results. For both LOCA and MSLB peak pressures, GOTHIC predicted values were less than one psig below CONTEMPT values.
Other plants, such as Joseph M. Farley Nuclear Plant, Units 1 and 2 have used the GOTHIC computer code to perform containment response analysis in support of their application for power uprate. License Amendment Nos.137 and 129 for Unit Nos.1 5
and 2, respectively, dated April 29,1998, provided approval for use of the GOTHIC comp 0ter co'de. Benchmarking for these plants also showed the capability of GOTHIC in the containment response calculations.
- Description and Safety Considerations Waterford 3 TS 3.6.2.2 presently requires two independent trains of CCS to be 1
operable with both CFCs per trai.n operable for Modes 1,2,3, and 4. Entergy Operations proposes to revise the existing Waterford 3 TS 3.6.2.2 to allow Waterford 3 to operate with one CFC operable in each train. Specifically, this change replaces the requirement of having two operable fan coolers per train in the specification with one operable fan cooler per train.
This proposed change is based on containment pressure and temperature response analyses performed for the limiting large break Loss of Coolant Accidents (LOCA) and limiting Main Steam Line Break (MSLB) events using the GOTHIC computer code.
Several LOCA and MSL'8 events are analyzed using GOTHIC and the Waterford 3 containment model to determine the limiting cases for:
LOCA containment peak pressure,
-. post-LOCA' containment pressure at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />,-
. MSLB peak containment pressure, and MSLB peak containment temperature.
New LOCA and MSLB mass and energy release data are used in the above GOTHIC
~l analyses. The revised mass and energy data are calculated by ABB Combustion Engineering - Nuclear Operations (ABB CE). The mass and energy data for LOCA
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cases are calculated for: (1) Double Ended Discharge Leg Slot Break (DEDLSB); (2)
{
Doubla Ended Suction Leg Slot Break (DESLSB); and (3) Double Ended Hot Leg Slot Break (DEHLSB). All LOCA cases are analyzed with both maximum and minimum Safety _ injection (SI) flow assumptions. Containment pressure and temperature
- response for LOCA cases are analyzed assuming a loss of offsite power and failure of one train of containment heat removal system with one CFC per train operable, i.e.,
- only one containment spray and one CFC assumed operable, for both minimum and
. maximum Si flow cases. This is a very conservative assumption for LOCA cases with s_
maximum safety injection flow, since the maximum Si flow assumption precludes a single failure to the power bus that would cau'se failure of both a CS train and a CCS
' train, i.e. both emergency diesel generators are assumed operable and therefore a single failure will either be the failure of one CS train or iailure of one CFC, but not both.
MSLB events are analyzed for 102%,75% and 50% power level with the following single failure assumption for each power level:
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failure of one main feedwater isolation valve (MFIV) to isolate;
. failure of one main steam isolation valve (MSIV) to isolate; and failure of one containment heat removal system train to operate.
The following provides a description of the LOCA and MSLB containment analyses
. performed using the GOTHlC code and the new mass and energy release data.
Bases for Loss of Coolant Accident Analysis Mass and energy release data used in the new Waterford 3 LOCA containment analyses are calculated by ABB/CE using NRC Staff approved methods and are based on the planned Waterford 3 power uprate application conditions. The Waterford 3 power uprate application conditions assume the core power to be 108% of the current rated core power plus 2% uncertainty (1.08
- 3390
- 1.02= 3734.42 Mwt =
3734.42/3390 = 110.16% of rated core power). A qualitative assessment of the parameters and their values that have a potentially significant impact on the LOCA mass & energy releases and containment response at pre-uprate (102%) and uprate (110.16%) power conditions was performed to justify that the power uprate LOCA bounds t'-e current pre-uprate plant analysis. The assessment compared the parameter values of reactor power, core inlet temperature, RCS flow, stea u generator invento.ry, pressure, and RCS liquid inventory. It was determined that the reactor power was the primary parameter that would have a significant impact on the post-LOCA mass and eaergy releases due to significantly more energy being released to the
. containment at the uprate power conditions. Additionally, SRP 6.2.1.3 requires the LOCA peak containment pressure analysis to be performed from 102% power, supporting that the results of the LOCA are maximized as the power increases.
Therefore, use of the power uprate mass and energy release data for containment analyses was determined to be conservative with respect to the current power level and bounds a similar LOCA at the pre-uprate power conditions of 102%.
Mass and energy release data and containment pressure and temperature response are calculated for the following LOCA break locations:
Double Ended Discharge Leg Slot Break (DEDLSB)
Double Ended Suction Leg Slot Break (DESLSB)
Double Ended Hot Leg Slot Break (DEHLSB)
The above LOCA break spectrum is s!milar to the current Waterford 3 LOCA spectrum used for LOCA peak containment pressure and temperature analysis.
The CFCs on train "B" are located at an elevation in the containment where a small portion of the bottom cooling coils may become partially flooded with the SI sump water.
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The limiting long term post-LOCA containment pressure case is analyzed with a reducsd CFC heat removal rate to account for partial flooding of the CFC. This CFC flooding effect is only a concern for the long term post-LOCA containment pressure, i.e., pressure at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, since the CFC will not be ficoded at the time of peak containment pressure for LOCA or MSLB.
LOCA Results i
The new Waterford 3 limiting LOCA for containment peak pressure was determined to be the Double Ended Hot Leg Slot break (DEHLSB). The peak pressure for the hot leg break occurs near the end of the blowdown phase, which is prior to the start of safety
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injection flow; start of CFC operation; and start of containment spray flow into the containment. The current FSAR licensing analysis determined that the DESLSB with-minimum safety injection flow assumption was the limiting LOCA for containment peak pressure. Table 1 below provides a comparison between the containment peak pressure (PP), timing of the peak pressure (TPP), peak terrperature (PT) and timing of peak temperature (TPT) for the new containment analysis and the current licensing analysis. The difference between the old and new results is primarily due to the new mass and energy data which shows lower mass and energy into containment during the reflood time period. Thus, the reflood peak, which was the maximum pressure for the FSAR ana'yses, is lower in the new analyses such that the blowdown peak becomes the maximum pressure. For blowdown peaks, the highest pressure occurs for the hot leg break, which results in the fastest RCS blowdown rate. The peak pressure is less than the containment design pressure of 44 psig.
1 PP (psig)
TPP(sec)
PT( F)
TPT(sec) 4 New Analysis (DEHLSB) 35.20 12.46 254.5 12.46 FSAR Analysis (Ref.8)
'43.1 330 267.4 330 Table 1 The new Waterford 3 limiting LOCA for containment pressure at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> was
- determined to be the Double Ended Discharge Leg Slot break (DEDLSB) with minimum safety injection flow assumption. This case is analyzed assuming one CS train and one CFC.is operable. The operable CFC is assumed to be the Train "B" CFC located at the lowest elevation and partia"y flooded. The current FSAR licensing analysis determined that the DESLSB with assumptions of maximum safety injection flow and the operation
. of one train of CHRS consisting of one containment spray train and two CFCs was the limiting LOCA for containment pressure at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Table 2 below provides a comparison between peak containment pressure (PP), timing of the peak pressure (TPP), and containment pressure at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (P @ 24 hr) for the new containment analysis and the current licensing analysis. The FSAR analysis only analyzed the long 8
term response for the maximum SI flow. The new analysis included both assumptiens for ths long term response and determined that the minimum Sl flow case is slightly worse.
PP (psig)
TPP(sec)
P @ 24 hr (psig)
New Analysis sDEDLSB) 33.27 12.46 15.5 FSAR Analysis (Ref.8) 42.9 253 14.90 Table 2 The containment pressure at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> must be reduced to less than half the containment peak pressure or 16.64 psig. The containment pressure at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is less than half the associated containment peak pressure.
Bases for Main Steam Line Break Analysis Mass and energy release data used in the MSLB containment analyses are calculated by ABB/CE using NRC Staff approved methods and are based on a pre-power uprate powerlevel. The new MSLB mass and energy release data are cbtained by using a MSlV closure time of 8 seconds (7 e ad MSIV closure time plus 1 second for signal delay) versus the current mass and energy release data that are based on a 5 second (4 second MSIV closure time plus 1 second for signal delay) MSIV closure time. This change accounts for the slower MSIV movement during accident conditions due to differential pressure forces across the vaive. Mass and energy release data and containment pressure response for MSLB events are calculated for an initial reactor power level of 102%,75% and 50% of the current rated thermal power. MSLB events are calculated considering the same single failures as the current licensing analysis.
These analyses are similar to the current Waterford a MSLB spectrum analyses except the current spectrum includes MSLB from 25% and 0% power level. However, the experience and the results of the new analyses have demonstrated that the higher power level results in a more severe containment pressure and temperature response.
Therefore, the MSLB event from 25% and 0% power were not included in the new analyses.
MSLB Results The new Waterford 3 limiting MSLB event for containment peak pressure was determined to be MSLB from 102% power with failure of one containment heat removal train consisting of one containment spray pump and one CFC operable. The current FSAR licensing analysis determined that the limiting MSLB event was MSLB from 75%
9
h power with the failure of one train of containment heat removal system consisting of one c6ntainrhent' spray train and two operable CFCs. Table 3 below provides a comparison between the containment peak pressure (PP) and timing of the peak pressure (TPP) for the new analysis and the current licensing analysis. The peak
. pressure is less than the containment design pressure of 44 psig.
PP psia (psig)
TPP(sec)
New Analysis (102%).
42.68 63.04 FSAR Analysis (Ref.8) 42.9 57 Table 3 The new Waterford 3 limiting event for containment peak temperature (EQ case) was determined to be MSLB from 102% power with the failure of one MSIV to close. Mass and energy release data for this case are generated considering steam superheating upon uncovering of the SG tubes to maximize containment temperature. The EQ case
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is allowed to credit 8% re-evaporation of the condensate from the passive heat sinks in the containment (NUREG 0588, Appendix B, Page llB-1). The re-evaporation assumption results in cooldown of the containment vapor region and in turn results in a lower containment pressure and temperature. However, this case has been
. conservatively evaluated without including the 8% re-evaporation. The current FSAR licensing ana!ysis determined that the containment peak temperature analysis results -
. from a MSLB from 102% with both CFCs per train operable and the failure of one train of containment spray. Table 4 below provides a comparison between the containment peak temperature (PT) and timing of peak temperature (TPT) for the new analysis and the current licensing analysis EQ cases without crediting re-evaporation. The calculated peak tempersture is less than the maximum temperature of 413.5*F used for
- EQ evaluations as given in TS 3/4.6.2.1 and 3/4.6.'2.2 bases.
PT( F)
TPT(sec)
New' Analysis (102%)-
'397.4 28.24 FSAR Analysis.(Ref.8)
'409.1 30.5 Table 4 1The major' differences between the new analyses and the FSAR analvses (LOCA and MSLB) include:'(1) the FSAR analyses assume two CFCs per train are operable, but
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the new analyses assume only one CFC per train is operable; (2) the new analyses use (the new mass and energy release data calculated by ABB/CE, and (3) the different values used for the parameters specified in Table 5:
10
9 FSAR (Ref.8)
New Analyses initial Containment Temperature 126 F 120 F CCW Temperature 120 *F 115 F CCW Flow 1100 gpm 1200 gpm Initial CS Riser Level 139.5 ft 149.5 ft The initial containment temperature, CCW flow, and initial CS riser level used in the new analyses are the va!ues provided in the Waterford 3 Technical Specir ations. The CCW design temperature c
is 115 F. CCW is the cooling medium for CFCs and CS heat exchanger in the recirculation mode.
Table 5 The above values used in the FSAR analyses include uncertainties associated with measurement of these parameters. However, the new analyses do not explicitly include the measurement uncertainties associated with the parameters specified above.
This is reasonable because of the very large margin available between the design pressure and the actual failure pressure of containment. Containment performance studies used in Individual Plant Examinations show that the containment failure pressure is typically at least 2 to 3 times greater than the design pressure. The slight potential increase in calculated containment pressure that might occur if all parameter values were simultaneously at their worst case greatest uncertainty range is very small compared to this margin to failure.
The impact of these uncertainties on the calculated peak containment pressure was determined to be about 0.6 psia for the limiting steam line break analysis. The latest containment failure probability versus pressure curve from the W3 PRA model was used to determine the change in failure probability for a 0.6 psi increase in pressure from 44 to 44.6 psig. This delta failure probability was then multiplied by the initiating event frequency of a large break LOCA or steam line break, the probability that the measured parameter was at its maximum worst case uncertainty (0.05; assuming that the entire j
0.6 psia impact was due to a single parameter) and the probability of single failure of a
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containment spray train assumed in the limiting design basis analysis. This results in a change in containment failure probability of less than 1X10-". Thus, using nominal Technical Specification values without uncertainty for these parameters in the I
containment pressure analysis is clearly not risk significant.
Also note that the limiting events for LOCA and MSLB for the new analyses are
- different than the current (FSAR) limiting events. The change in the limiting events has been determined to be due to differences in the new and old mass and energy release
)
data. The differences in new and old mass and energy data are due to improvements i
in the new NRC Staff approved analysis codes and assumptions.
)
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No Significant Hazards Consideration Determination The proposed changes described above have been evaluated in accordance with 10CFR 50.92(c). The changes shall be deemed to involve a significant hazards consideration if there is a positive finding in any of the following areas:
1.
Will the operation of the facility in accordance with these proposed changes involve a significant increase in the probability or consequence of an accident previously evaluated?
Response
The proposed change to Technical Specification (TS) 3.6.2.2 reduces the number of Containment Fan Coolers (CFC) from two to one required to be operable in each train of the Containment Cooling System for modes 1,2,3, and 4. This change does not create any new system interactions and has no impact on operation or function of any system or equipment in a way that could cause an accident. The CFCs are not an initiator of any events nor affect any accident initiators of any events analyzed in Chapter 15 of the UFSAR. Therefore this change will not impact the probability of occurrence of an accident.
The results of the reanalysis of the limiting Loss of Coolant Accident (LOCA) and Main Steam Line Break (MSLB) accidents show that the consequences of an accident previously evaluated are not increased by the change in the required number of operable CFCs. The limiting accidents affected by the proposed changes are identified below:
. The peak containment pressure following the limiting LOCA (Double Ended Hot
)
Leg Slot Break with minimum safety injection flow) was determined to be 35.2 psig as compared to the current licensing basis limiting LOCA (Double Ended l
Suction Leg Slot Break with minimum safety injection flow) peak pressure of j
43.1 psig..
The peak containment pressure at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following the start of the limiting
{
LOCA - Double Ended Discharge Leg Slot Break with minimum safety injection j
(
flow) and the operation of one containment spray train and one partially flooded CFC operable was determined to be 15.5 psig with a peak pressure of 33.27 psig as compared to the current licensing basis of 14.9 psig with a peak i
pressure of 42.9 psig. The current licensing basis limiting LOCA is the Double Ended Suction Leg Slot Break with maximum safety injection flow and the operation of one containment spray train and two operable CFCs.
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The, peak containment pressure following the limiting MSLB (102% power with failure of one containment heat removal train consisting of one containment spray pump and one CFC operable) was determined to be 42.68 psig as compared to the current licensing basis peak pressure of 42.9 psig. The current licensing basis limiting MSLB is 75% power with the failure of one train of containment heat removal system consisting of one containment spray train and two operable CFCs.
The peak containment equipment qualification temperature following the limiting MSLB (102% power with the failure of one MSIV to close) was determined te be 397.4 F as compared to the current licensing basis peak temperature of 409.1 F. The current licensing basis limiting MSLB is 102%
power with two CFCs per train operable and the failure of one train of containment spray.
These values above demonstrate that the containment design basis prersure and equipment qualification temperature of 44 psig and 413.5 F, respectively, are not exceeded and the containment pressure at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after start of the limiting LOCA is less than 50% of the peak pressure.
i The results of the containment response analysis discussed above satisfy the following NRC Staff Standard Review Plan (SRP) section 6.2.1.1.A guidance document acceptance criteria for a PWR dry containment.
The peak calculated containment pressure following a Loss of Coolant Accident (LOCA) or Main Steam Line Break (MSLB) should be less than the containment design pressure.
To satisfy the requirements of GDC 38 to rapidly reduce the containment pressure, the containment pressure should be reduced to less than 50% of the peak calculated pressure for the design basis LOCA within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after I
the postulated accident.
Thus, revising the containment cooling system TS to require only one operable i
CFC per train results in acceptable containment response and therefore, will not adversely impact the consequences of accidents previously evaluated.
I Therefore, the proposed changes will not involve a significant increase in the probability or consequences of an accident previously evaluated.
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.2.
. Will the, operation of the facility in accordance with these proposed changes create the possibility of a new or different kind of accident from any accident previously evaluated?
Response
The proposed change to reduce the number of operable Containment Fan Coolers (CFC) from two to one in each train of the Containment Cooling System for modes 1,2,3, and 4 does not alter the operation of the CFCs. Although only one of the two CFCs per train is required to be operable, the manner in which the CFCs perform their safety function is not changed. All four CFCs (two per train) will be maintained operable to the extent possible to provide the greatest defense in depth and operating flexibility.
This proposed change does not involve a change in plant design, nor does it involve any potential initiating events that would create any new or different kind of accident. This proposed change does not alter the way in which the plant is operated in a manner that would create a new or different accident. Therefore, since no hardware modifications will be made, the proposed change will not create the possibility of a new or different kind of accident from any accident previously evaluated, i
3.
Will the operation of the facility in accordance with these proposed changes i
involve a significant reduction in a margin of safety?
Response
The proposed change revises TS 3.6.2.2, Containment Cooling System. This change revises the required number of fan coolers from two fan coolers per train i
to one fan cooler per train. As described in the containment depressurization and cooling system Technical Specification Bases, the containment cooling system is
' designed to maintain the post accident containment peak pressure below its design value of 44 psig, The system is also designed to reduce the containment pressure by a factor of 2 from its post-accident peak within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
The analyses that have been pe' rformed to support this Technical Specification change have shown that the peak containment pressure remains below 44 psig, the 24-hour containment pressure is less than half the peak pressure, and the containment peak temperature remains below the maximum temperature of
' 4K YF provided in the Bases for Technical Specifications 3.6.2.1 and 3.6.2.2. In comparison of the current safety margins to the safety margins that would exist if the proposed changes were in effect, the results of the analyses, illustrated below, 14
show an increase in the margin of safety for containment pressure and equipment
<jualification temperature following the associated limiting LOCA and MSLB.
{
The peak containment pressure following the limiting LOCA was determined to be 35.2 psig as compared to the current licensing basis limiting LOCA peak
{
pressure of 43.1 psig.
j The peak containment pressure at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following the start of the limiting LOCA was determined to be 15.5 psig with a peak pressure of 33.27 psig as J
compared to the current licensing basis of 14.9 psig with a peak pressure of 42.9 psig.
The peak containment pressure 'ollowing the limiting MSLB was determined to be 42.68 psig as compared to tha current licensing basis peak pressure of 42.9 psig.
The peak containment equipment qualification temperature following the limiting MSLB was determined to be 397.4 F as compared to the current licensing basis peak temperature of 409.1 F.
I This proposed change does not adversely impact a margin of safety, involve a change in plant design, or have any affect on the plant protective barriers.
Therefore, the proposed change will not involve a significant reduction in the margin of safety.
Safety and No Significant Hazards Consideration Determination j
Based on the above review, it is concluded that: (1) the propmed change does not constitute a significant hazards consideration as defined by 10 CFR 50.92; and (2) there is a reasonable assurance that the health and safety of the public will not be j
endangered by the proposed change; and (3) this action will not result in a condition which significantly alters the impact of the station on the environment as described in the NRC Final Environmental Statement.
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