ML051920360
| ML051920360 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 07/01/2005 |
| From: | Korsnick M Constellation Energy Group |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| Download: ML051920360 (18) | |
Text
Maria Korsnick 1503 Lake Road Vice President Ontario, New York 14519-9364 585.771.3494 585.771.3943 Fax maria.korsnick @ constellation.com Constellation Energy
_ R.E. Ginna Nuclear Power Plant July 1, 2005 U. S. Nuclear Regulatory Commission Washington, DC 20555 ATTENTION: Document Control Desk
SUBJECT:
R.E. Ginna Nuclear Power Plant Docket No. 50-244 Transmittal of Revised Analysis Associated with the License Amendment Reauest Reearding Main Feedwater Isolation Valves On April 29, 2005, R.E. Ginna Nuclear Power Plant, LLC submitted a license amendment request regarding the main feedwater isolation valves (Accession Number ML051260236). Subsequently, we were notified by the vendor that input errors have been found in one of the attached supporting analysis (Enclosure 5) for the amendment request. The errors involved the inputs used for containment cooler performance, and were associated with the containment atmosphere flow rate and the heat removal rate while the containment atmosphere is superheated. The resolution resulted in slightly higher containment peak pressures and does not constitute a change in the approved methodology. Corrective actions are being addressed by the vendor.
Enclosure 1 to this submittal provides a revised copy of the main steam line break containment analysis.
There are no new commitments being made in this submittal. Should you have questions regarding the information in this submittal, please contact George Wrobel at (585) 771-3535 or george.wrobel~constellation.com.
trly rs17 ary. snck
Enclosures:
(1) Steam Line Break Mass/Energy Release and Containment Response Analysis Dc)
/0o (rftg
STATE OF NEW YORK
- TO WIT:
COUNTY OF WAYNE I, Mary G. Korsnick, being duly sworn, state that I am Vice President - R.E. Ginna Nuclear Power Plant, LLC (Ginna LLC), and that I am duly authorized to execute and file this response on behalf of Ginna LLC. To the best of my knowledge and belief, the statements contained in this document are true and correct. To the extent that these statements are not based on my personal knowledge, they are based upon information provided by other Ginna LLC employees and/or consultants. Such information has been reviewed in accordance with company practic and I believe it to be reliable.
Subscribed and sworn before me, a Notary Public in and for the State of New York and County of (honrop, this /day of J t ,2005.
WVITNESS my Hand and Notarial Seal: 4 I dL Notary Public My Commission Expires: SHARONLMILUER Ng;y Pft*bi: ofd Now York Hun No. 01MR617M5 Date cc: S. J. Collins, NRC P. D. Milano, NRC Resident Inspector, NRC Mr. Peter R. Smith New York State Energy, Research, and Development Authority 17 Columbia Circle Albany, NY 12203-6399 Mr. Paul Eddy NYS Department of Public Service 3 Empire State Plaza, 10th Floor Albany, NY 12223
ENCLOSURE I R.E. Ginna Nuclear Power Plant Steam Line Break Mass/Energy Release and Containment Response Analysis
Steam Line Break Mass/Energy Release and Containment Response Analysis 1.0 Introduction Steamline ruptures occurring inside a reactor containment structure may result in significant releases of high-energy fluid to the containment environment that could produce high pressure conditions for extended periods of time. The magnitude of the releases following a steamnline rupture is dependent upon the plant initial operating conditions and the size of the rupture as well as the configuration of the plant steam system and the containment design. There are competing effects and credible single failures in the postulated accident scenario used to determine the worst cases for containment pressure following a steamline break.
The Ginna steamline break and containment response analysis considers a spectrum of cases that vary the initial power condition and the postulated single failure. The following sections identify the analysis methodology, the selection of cases, the major plant assumptions and the results of the analysis. Major elements considered in this analysis compared to the current licensing-basis analysis are the automatically-actuated main feedwater isolation valves (MFIVs) as a back-up to the feedwater regulating valves (MFRVs), the extended power uprate (EPU), and a reduction in the minimum required shutdown margin for the core at end of cycle conditions.
2.0 Analysis Methods and Computer Codes This section identifies the methods and computer codes used to calculate the steamline break mass and energy releases and the containment pressure response.
2.1 Mass and Energy Release Methodology and Computer Code The analysis documented herein uses the RETRAN code, which is documented in WCAP-14882-P-A, "RETRAN-02 Modeling and Qualification for Westinghouse Pressurized Water Reactor Non-LOCA Safety Analyses" (Ref. 5).
The following limitations in the NRC SER for WCAP-14882-P-A have been adhered to in the use of RETRAN to analyze this event.
- The break flow model is the Moody model.
- Only steam (dry vapor) will exit the break, since perfect steam separation in the steam generators is assumed.
- Westinghouse will not evaluate the superheat in the steam released to the containment. Any superheated conditions will be reset to be equal to the saturation temperature.
A BNFL Group company
2.2 Containment Response Methodology and Computer Code The containment response analysis uses the GOTHIC computer code. The GOTHIC program is rapidly becoming the industry standard for performing containment pressure and temperature design-basis analyses. The GOTHIC Technical Manual (Ref. 6) provides a description of the governing equations, constitutive models, and solution methods in the solver. The GOTHIC Qualifications Report (Ref. 7) provides a comparison of the solver results with both analytical solutions and experimental data.
The Ginna GOTHIC containment evaluation model consists of a single lumped-parameter node; the diffusion layer model (DLM) is used for heat transfer to all structures in the containment. Plant input assumptions (identified in Section 4.4) are the same as, or slightly more restrictive, than in the current licensing-basis analysis performed with the COCO code (Ref. 8). Benchmarking between the Ginna COCO and GOTHIC models was performed to confirm consistency in the implementation of the plant input values.
This steamline break containment response analysis uses GOTHIC version 7.2. The latest code version is used to take advantage of the DLM heat transfer option. This heat transfer option was approved by the NRC (Ref. 9) for use in Kewaunee containment analyses with the condition that mist be excluded from what was earlier termed as the mist diffusion layer model (MDLM). The GOTHIC containment modeling for Ginna has followed the conditions of acceptance placed on Kewaunee. Kewaunee and Ginna both have large, dry containments with similar containment volumes and active heat removal capabilities.
Changes in the GOTHIC code versions are detailed in Appendix A of the GOTHIC User Manual Release Notes (Ref. 10). Version 7.2 is used consistent with the restrictions identified in Ref. 9; none of the user-controlled enhancements added to version 7.2 were implemented in the Ginna containment model.
3.0 Case Definitions and Single Failures There are many factors that influence the quantity and rate of the mass and energy release from the steamline. To encompass these factors, a spectrum of cases varies the initial power level and the single failure. This section summarizes the basis of the cases that have been defined for the Ginna plant.
The power level at which the plant is operating when the steamline break is postulated can cause different competing effects that make it difficult to pre-determine a single limiting case. For example, at higher power levels there is less initial water/steam in the steam generator, which is a benefit. However, at a higher power level there is a higher initial feedwater flowrate, higher feedwater temperature, higher decay heat, and there is a higher rate of heat transfer from the primary side, which are all penalties. Therefore, cases consider initial power levels varying from full power to zero power. The specific initial power levels that are analyzed are 100%, 70%, 30% and 0% as presented in WCAP-8822.
All cases consider the largest possible break a double-ended rupture (DER) immediately downstream of the flow restrictor at the outlet of the steam generator. This break conservatively bounds the plant response to any smaller break size. The effective forward break area is limited by the 1.4 ft2 cross-sectional area of the flow restrictor. The reverse break area is the cross-sectional area of the pipe, which is 4.125 ft2 .
A BNFL Group company
Several single failures can be postulated that would impair the performance of various steamline break protection systems. The single failures either reduce the heat removal capacity of the containment safeguards systems or increase the energy release from the steamline break. The single failures that have been postulated for Ginna are summarized below. The analysis cases separately consider each single failure at each initial power level.
- 1) AFW Throttle Valve Failure The AFW system includes flow control valves that are designed to throttle the AFW flow from the motor-driven pumps. It is postulated that this valve fails full open instead of throttling as intended. This affects the flowrate from a single motor-driven AFW pump. This failure results in an AFW flowrate to the faulted SG of 265 gpm, versus the design outlet flowrate of < 230 gpm.
- 2) Main Feedwater Regulating Valve (MIFRV) Failure The MFRV is a fast-closing (10 second stroke time) valve in the feedwater system that is the preferred (fastest) method for terminating feedwater addition to the faulted SG during a steamline break. If the MFRV on the faulted loop fails open, the back-up main feedwater isolation valve (MFIV), with a 30 second stroke time, is credited to close. The slower closure time creates the possibility of additional pumped feedwater entering the faulted SG. Although the main feedwater pumps trip on a safety injection (SI) signal, the condensate pumps do not trip and can continue to provide pumped flow if the faulted SG depressurizes below approximately 350 psia.
- 3) Vital Bus Failure One of the postulated failures is that of a vital bus that powers one safeguards train. The main impact is that the active containment heat removal is reduced by 50% with the loss of one train of fan coolers and one containment spray pump. The failure also causes the loss of one train of safety injection.
- 4) Diesel Failure A diesel failure is postulated only after a loss of offsite power. A loss of offsite power necessitates the start of diesels to power the required safety-related equipment. A diesel failure eliminates one safeguards train, including one train of fan coolers, one containment spray pump, one train of safety injection and one train of motor-driven auxiliary feedwater pumps. This failure is similar to the vital bus failure, above, except there are longer delays because of the time to power equipment from the functioning diesels. In addition, the loss of offsite power has other impacts on the event such as tripping the RCPs and the loss of a motor-driven AFW pump, which would have been powered by the failed diesel.
4.0 Analysis Assumptions 4.1 Protection Logic and Setpoints The pertinent signals and setpoints that are actuated in these analyses are summarized below.
A BNFL Group company
The first safety injection (SI) signal is generated by a low steamline pressure signal in all cases. The assumed setpoint is 372.7 psia, with a lead/lag of 12/2. The SI signal is credited to cause:
- start of SI pumps,
- start of motor-driven auxiliary feedwater pumps,
- closure of MFRVs and MFIVs, and
- trip of main feedwater pumps.
The start of containment fan coolers is credited after the hi containment pressure setpoint of 6 psig is reached. The start of containment sprays is credited after the hi-hi containment pressure setpoint of 33.5 psig is reached.
4.2 Secondary Side Assumptions This section summarizes the major input assumptions associated with the steam generator, the main feedwater system, the auxiliary feedwater system and the steamline.
Initial Steam Generator Inventory A high initial steam generator mass is assumed. The initial level corresponds to 60% narrow range span (NRS) at all power levels. This consists of a nominal level of 52% NRS, 4%NRS measurement uncertainty, and 4%NRS measurement bias.
Main Feedwater System The main feedwater flowrate to the faulted SG is conservatively high to maximize the water mass inventory that will be converted to steam and released from the break. The MFRV on the faulted loop is assumed to quickly open in response to the increased steam flow. The intact loop MFRV is assumed to either be in the nominal position based on the initial power level, or to be closed, which causes the maximum flowrate to the faulted SG. The trip of the main feedwater pumps is credited after an SI signal, with a 2 second delay and a 10 second coastdown. The closure of the faulted loop MFRV is credited after a 2 second delay and a 10 second valve stroke time. The closure of the MEIVs is credited after a 2 second delay and a 30 second valve stroke time.
The feedwater in the unisolable feedline between the MFRV and faulted SG is also considered in the analysis. The hot main feedwater usually reaches saturated conditions as the SG and feedline depressurize. The decrease in density as flashing occurs causes most of the unisolable feedwater to enter the faulted SG. The modeling of the unisolable feedline volume is reduced by the pipe volume that is purged by the auxiliary feedwater, which remains subcooled at pressures down to atmospheric conditions.
A BNFL Group company
Auxiliary Feedwater The auxiliary feedwater flowrate to the faulted SG is conservatively high to maximize the water mass inventory that will be converted to steam and released from the break. The motor-driven auxiliary feedwater pumps are actuated on an SI signal. The maximum flowrate assumed in the analysis is 235 gpm to each SG. A minimum delay of 25 seconds is credited in most cases due to the safeguards sequencer logic; cases initiated from zero power do not credit a delay.
The turbine-driven AFW pump is assumed to deliver a maximum flowrate of 630 gpm to the faulted SG.
The turbine-driven AFW pump is assumed to be actuated due to a low-low SG level signal in both SGs for any case initiated from a power level above 50%. The turbine-driven AFW pump is also actuated due to a loss of offsite power, and is therefore modeled for all diesel failure cases.
Operator action is credited to terminate the auxiliary feedwater flow to the faulted steam generator after 10 minutes.
Unisolable Steamline The initial steam in the steamline between the break and the steamline nonreturn check valve is included in the mass and energy released from the break.
Quality of the Break Effluent The quality of the break effluent is assumed to be 1.0, corresponding to saturated steam that is all vapor with no liquid. Although it is expected that there would be a significant quantity of liquid in the break effluent for a full double-ended rupture, the all-vapor assumption conservatively maximizes the energy addition to the containment atmosphere.
4.3 Reactor Coolant System Assumptions While the mass and energy released from the break is determined from assumptions that have been discussed in the previous section, the rate at which the release occurs is largely controlled by the conditions in the reactor coolant system. The major features of the primary side analysis model are summarized below.
- Continued operation of the reactor coolant pumps maintains a high heat transfer rate to the steam generators. However, the diesel failure cases follow a loss of offsite power, and therefore credit the trip of the RCPs.
- The model includes consideration of the heat that is stored in the RCS metal.
- Reverse heat transfer from the intact steam generator to the RCS coolant is modeled as the temperature in the RCS falls below the steam generator fluid temperature.
A BNFL Group company
- Minimum flowrates are modeled from ECCS injection, to conservatively minimize the amount of boron that provides negative reactivity feedback.
- The assumed NSSS power is 1817 MWt, which includes a maximum pump heat of 10 MWt.
- RCS average temperature is the full-power nominal value of 576.00 F plus an uncertainty of
+4.0F.
- Core residual heat generation is assumed based on the 1979 ANS decay heat plus 2a model (Ref.
11).
- Conservative core reactivity coefficients (e.g. moderator temperature) corresponding to end-of-cycle conditions with the most reactive rod stuck out of the core are assumed. This maximizes the reactivity feedback effects as the RCS cools down as a result of the steamline break.
- All cases have credited a minimum shutdown margin of 1.3% Ak.
4.4 Containment Assumptions This section identifies the major input values that are used in the containment response analysis. The assumed initial conditions and the input assumptions associated with the fan coolers and containment sprays are listed in Table 1 (similar to UFSAR Table 6.2-9). The containment heat sink input is provided in Table 2 (similar to UFSAR Table 6.2-10), and the corresponding material properties are listed in Table 3 (similar to UFSAR Table 6.2-11).
A BNFL Group company
Table 1 Initial Containment Conditions, Fan Cooler and Containment Spray Pump Assumptions Parameter Value Containment net free volume (ft3 ) 1,000,000 Initial containment temperature (0 F) 120.0 Initial containment pressure (psia) 15.7 Initial relative humidity (%) 20 Number of fan coolers
- All 4
- Vital bus or diesel failure 2 Hi containment pressure setpoint (psig) 6.0 Delay (sec) from high containment pressure setpoint to start of fan coolers
- With offsite power available 34.0
- With loss of offsite power 44.0 Heat Temp Removal 85 0 120 398 Containment fan cooler heat removal (BTU/sec per fan cooler) vs. 220 8839 containment steam saturation temperature (0 F) 240 10375 260 11911 280 13446 286 13907 Containment fan cooler air flowrate (ft3 /min per fan cooler) 30,000 Number of spray pumps
- All 2
- Vital bus or diesel failure I Hi-hi containment pressure setpoint (psig) 33.5 Delay (sec) from high-high containment pressure setpoint to start of containment sprays:
- 2 sprays 26.8
- I spray 28.5 Containment spray flowrate (gpm/spray pump) 1300.0 RWST/containment spray water temperature (0 F) 104.0 A BNFL Group company
Table 2 Containment Heat Sink Input Heat Sink Area Thickness Thickness Number l Description l (ft 2) Material (inches) (ft) 1 Insulated Containment Wall (1) 36285 SS 0.019 0.00158 Insulation 1.250 0.1042 Steel 0.375 0.03125 Concrete 42.000 3.5 2 Uninsulated Containment Wall (1) 12370 Overcoat 0.008 Primer 0.002 Steel 0.375 0.03125 Concrete 30.000 2.5 3 Basement Floor ( 6576 Overcoat 0.005 Concrete 24.000 2 Steel 0.250 0.0208 Concrete 24.000 2 4 Wet Sump Wall A ' 8.2 Overcoat 0.004 Primer 0.002 Steel 0.250 0.0208 Concrete 36.000 3 5 Dry Sump Wall A ' 2052.8 Overcoat 0.004 Primer 0.002 Steel 0.250 0.0208 Concrete 36.000 3 6 Sump Floors (1) 366 Overcoat 0.005 Concrete 24.000 2 Steel 0.250 0.0208 Concrete 12.000 1 7 Walls of Sump B ' 189 Overcoat 0.005 Concrete 24.000 2 Steel 0.250 0.0208 Concrete 12.000 1 8 Outer Refueling Cavity Wall 6132 Overcoat 0.005 Concrete 35.280 2.94 9 Inner Refueling Cavity Wall ( 5609 SS 0.250 0.0208 Concrete 24.000 2 10 Bottom Refueling Cavity ' 1143 SS 0.250 0.0208 Concrete 48.000 4 11 Loop Compartments 18846 Overcoat 0.005 Concrete 16.938 1.4115 12 Floor of Intermediate Level 9672 Overcoat 0.005 Concrete 3.000 0.25 A BNFL Group company
Table 2 Containment Heat Sink Input Heat Sink Area Thickness Thickness Number Description (ft2) Material 4(f)
(inches) 13 Operating Deck 15570 Overcoat 0.005 Concrete 12.000 1 14 Thick Crane Structure 7225 Overcoat 0.004 Primer 0.002 Steel 0.750 0.0625 15 Crane Structure 3374 Overcoat 0.004 Primer 0.002 Steel 0.415 0.03455 16 I-Beam 7678 Overcoat 0.004 Primer 0.002 Steel 0.260 0.0217 17 Thick I-Beam 5536 Overcoat 0.004 Primer 0.002 Steel 0.703 0.0586 18 Crane Support 342 Overcoat 0.004 Primer 0.002 Steel 2.000 0.16667 19 Crane Beams 236 Overcoat 0.004 Primer 0.002 Steel 1.440 0.12 20 Grating and Misc 14000 Overcoat 0.004 Primer 0.002 Steel 0.0625 0.005208 Note:
- 1. The air gaps between concrete and steel and between insulation and steel are modeled.
A BNFL Group company
Table 3 Material Properties Table for Containment Heat Sinks Material Conductivity Volumetric Heat Capacity (BTU/ft 3 -F)
(BTU/hr-ft-0 F)
Concrete 0.81 31.5 Carbon Steel 28.0 54.4 Insulation 0.0208 1.11 Stainless Steel 8.8 54.6 Organic Coating 0.1 20.0 Inorganic Primer 1.0 20.0 5.0 Steamline Break Containment Response Results Sixteen steamline break cases were analyzed varying the initial power level and the assumed single failure. The mass and energy release from the break was calculated using the RETRAN code, while the containment pressure response was determined with the GOTHIC code. The analysis included the effects of the extended power uprate to 1817 MWt, a decrease in the shutdown margin to 1.3%, and the benefit of the modification to automatically close the feedwater isolation valves on an SI signal.
The current limiting steamline break containment pressure case, as documented in FSAR Section 6.2.1.2.3 (Rev. 18), is a full power double-ended break with a MFRV failure. With the faulted loop MFRV assumed to fail open, this analysis models feedwater isolation due to the closure of the main feedwater pump discharge valves with an 80-second stroke time. The total water mass addition from the main feedwater system is over 100,000 Ibm to the faulted SG. A key element in limiting the break release rate is crediting the 2.4% shutdown margin, which prevents a post-trip return-to-power.
Figure 1 and Figure 2 compare this limiting case from the FSAR to the new analysis which has credited feedwater isolation due to the automatically actuated MFIV with a stroke time of 30 seconds. In the first 120 seconds, there is a higher release rate and a higher containment pressure due to the EPU effects and a lower shutdown margin (1.3%). However, the MFIV closure limits the main feedwater addition to less than 20,000 lbm, which substantially improves the longer-term containment pressure transient. The peak containment pressure is reduced from 59.8 psig to 51.7 psig.
A BNFL Group company
With the modified MFIV, an MFRV failure is no longer the limiting single failure for the steamline break containment response analysis. The limiting case definition changes to a vital bus failure with an initial power level of 70%. The sequence of events for this limiting case is listed in Table 4. The break flowrate is shown in Figure 3, the break enthalpy is in Figure 4, and the containment pressure transient is in Figure
- 5. The peak containment pressure is 59.6 psig, which is below the containment design pressure of 60.0 psig.
Table 4 Sequence of Events Steamline Break, Vital Bus Failure, 70% Power Event Time (sec)
Break occurs 0 Low steamline pressure SI setpoint reached < 0.05 Hi containment pressure setpoint reached 2.0 Reactor trip 2.0 Main feedwater pumps trip 2.0 MFRV closes 12.0 Auxiliary feedwater starts 25.0 Containment fan coolers start 36.0 Hi-hi containment pressure setpoint reached 39.7 Containment sprays start 68.2 AFW terminated to faulted SG 600 Peak containment pressure occurs 612 Break releases stop 716 6.0 Conclusion The steamline break mass and energy release and containment response analysis has been done to show the effect of the automatically-actuated MFIVs as a back-up feedwater isolation method if the faulted loop MFRV fails open. This plant modification is a benefit to the containment pressure response, lowering the peak containment pressure by 8.1 psi for this postulated single failure. I The analysis also considered more limiting conditions associated with the EPU and a reduction in the minimum required shutdown margin at end of cycle conditions for a spectrum of cases. With the MFIV plant modification, the limiting case definition changes to a double-ended rupture initiated at 70% power with a vital bus failure. The peak containment pressure is 59.6 psig, which is acceptable because it is below the containment design pressure of 60.0 psig.
A BNFL Group company
Figure 1 Ginna Steamline Break Mass Flowrate Full Power, FRV Failure F-SAR Section 6.2.1.2.3 Rev. 18 (MFWP Discharge Volves)
EPU Analysis (Modi fied FIV) 2000 E
- 1500 1000 o
5500 -
0 0 100 200 300 400 500 600 700 Time (s)
Figure 2 Ginna Steamline Break Containment Pressure Response Full Power. FRV Failure FSAR Section 6.2.1.2.3 Rev. 18 (MFWP Dischorge Volves)
- - EPU Anolysis (Modified FIV) 60-50 - t-40 -
30 -
- 3 U) 21320 -
0) 10 0-0 100 200 300 400 500 600 700 Time (s)
A BNFL Group company
Figure 3 Ginna Steamline Break Mass Flowrate 70% Power, Vital Bus Failure EPU Ana I ys i s 12000 0
On, 10000 E
- o. 8000 a) 6000
° 4000 j° 2000 m
0 0
10 Time (s)
Figure 4 Ginna Steamline Break Enthalpy 70% Power, Vital Bus Failure EPU Ana I ys i s 1210-_
E 1200
Ž-1190 m 1180-1170-c 1160 bIJ C -~1150
@ 1140-130 - I Time (s)
A BNFL Group company
Figure 5 Ginna Steamline Break Containment Pressure Response 70% Power, Vital Bus Failure EPU
- Ana I y s i s 60-50 _
- 01 40-0-
=3 a 20-10 0 - I 11111 1 1 1 1 1 111 I I I 1111 I 1111' Time (s)
A BNFL Group company
7.0 References
- 1. Letter from Cecil 0. Thomas (NRC), 'Acceptance for Referencing of Licensing Topical Report WCAP-882I(P)/8859(NP), "TRANFLO Steam Generator Code Description", and WCAP-8822(P)/8860(NP), "Mass and Energy Release Following a Steam Line Rupture," August 1983.
- 2. Land, R.E., "Mass and Energy Releases Following a Steam Line Rupture," WCAP-8822 (Proprietary), WCAP-8860 (Non-Proprietary), September 1976.
- 3. Burnett, T.W.T., et al. "LOFTRAN Code Description," WCAP-7907-P-A (Proprietary) and WCAP-7907-A (Non-Proprietary), April 1984.
- 4. Osborne, M. P. and D. S. Love, "Mass and Energy Releases Following a Steam Line Rupture, Supplement 1 - Calculations of Steam Superheat in Mass/Energy Releases Following a Steamline Rupture," WCAP-8822-S1-P-A (Proprietary), September 1986.
- 5. Huegel, D. S, et al. "RETRAN-02 Modeling and Qualification for Westinghouse Pressurized Water Reactor Non-LOCA Safety Analyses," WCAP- 14882-P-A (Proprietary), April 1999.
Version 7.1, January 2003.
Verstion 7.1, January 2003.
- 8. Bordelon, F. M., and E. T. Murphy, "Containment Pressure Analysis Code (COCO)," WCAP-8327 (Proprietary), WCAP-8326 (Non-Proprietary), July 1974.
- 9. NRC Letter from Anthony C. McMurtray (NRC) to Thomas Coutu (NMC), Enclosure 2, Safety Evaluation, September 29, 2003.
- 10. NAI 8907-02, Revision 14, "GOTHIC Containment Analysis Package User Manual," Version 7.1, January 2003.
- 11. ANSI/ANS-5. 1-1979, "American National Standard for Decay Heat Power in Light Water Reactors," August 1979.
A BNFL Group company