RBG-45940, License Amendment Request (LAR) 2001-43, High Energy Line Break Analysis Method
| ML021410482 | |
| Person / Time | |
|---|---|
| Site: | River Bend  |
| Issue date: | 05/14/2002 |
| From: | Hinnenkamp P Entergy Nuclear South |
| To: | Document Control Desk, Office of Nuclear Security and Incident Response |
| References | |
| RBG-45940 | |
| Download: ML021410482 (95) | |
Text
Entergy Nuclear-South River Bend Station 5485 U.S. Highway 61 P.O. Box 220 St. Francisville, LA 70775
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EntergyTel 225 381 4374 En toWFax 225 381 4872 phinnen@entergy.com Paul D. Hinnenkamp Vice President, Operations River Bend Station RBG-45940 May 14, 2002 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555
Subject:
River Bend Station Docket No. 50-458 License No. NPF-47 License Amendment Request (LAR) 2001-43, "High Energy Line Break Analysis Method"
Dear Sir or Madam:
Pursuant to 100FR50.90, Entergy Operations, Inc. (Entergy) hereby requests the following amendment for River Bend Station, Unit 1. The proposed change revises the method of analysis for the High Energy Line Breaks in the subcompartments inside and outside of containment. This change is the result of a change in the method of analysis code from THREED to GOTHIC.
This is a change in an evaluation methodology according to the current 10CFR50.59 regulation, and a submittal is required by 10CFR50.59(c)(2) (viii). The proposed changes to the Updated Safety Analysis Report are provided for information.
The proposed change has been evaluated in accordance with 10CFR50.91(a)(1) using criteria in 10CFR50.92(c) and it has been determined that this change involves no significant hazards considerations. The bases for these determinations are included in the attached submittal.
The NRC has approved similar changes using GOTHIC for other plants including Joseph M. Farley Nuclear Plant, Units 1 and 2 and Waterford 3.
This amendment is required to implement a modification during Refueling Outage 11 scheduled to begin March 14, 2003.
Entergy requests approval of the proposed amendment prior to this outage. Once approved, the amendment will be implemented prior to startup from the outage.
A c4-
Letter RBG-45940 Page 2 of 2 The proposed change does not include any new commitments.
If you have any questions or require additional information, please contact Barry Burmeister at 225-381-4148.
I declare under penalty of perjury that the foregoing is true and correct. Executed on May 14, 2002.
Sincerely, Paul D. Hinnenkamp Vice President, Operations Attachments:
- 1. Analysis of Proposed change to the method of analysis code
- 2. Proposed Updated Safety Analysis Report Changes (mark-up) cc:
U. S. Nuclear Regulatory Commission Region IV 611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011 NRC Senior Resident Inspector P. O. Box 1050 St. Francisville, LA 70775 Mr. David Wrona U.S. Nuclear Regulatory Commission M/S OWFN 7D1 Washington, DC 20555
bcc:
File Nos.:
G9.5, G9.42 RBEXEC-02-008 RBF1-02-0072 RBG-45940
Aftachment I to RBG-45940 Page 1 of 13
1.0 DESCRIPTION
River Bend Station (RBS) plans to use the GOTHIC (Generation of Thermal-Hydraulic Information for Containments) code to replace the current vendor THREED code for room pressure-temperature analyses due to High Energy Line Breaks (HELB). The reasons for this change are the lack of support for the THREED code by the vendor and the additional capabilities of the GOTHIC code. Use of the GOTHIC code will allow for these analyses to be performed by Entergy personnel with an established code used widely through the nuclear industry. EOI is also considering future use of this code to perform other containment pressure temperature examinations in support of RBS Updated Safety Analysis Report (USAR) Section 6.2 licensing basis analyses, which were originally analyzed with the vendor THREED code.
To address plant operational issues and modifications, the HELB analyses require re-analysis.
The GOTHIC code will be used to perform this analysis. One planned modification will add additional delay time to the initiation logic for the Leak Detection System temperature setpoints, which provide the isolation signals credited to mitigate HELBs in both the Auxiliary and Containment Buildings. To support these activities, GOTHIC models were constructed to perform the HELB analyses. While the modification to add an additional time delay is a change in an input parameter for the analysis, and would not require NRC approval, the change in the analysis code from THREED to GOTHIC does present a deviation in an evaluation methodology according to the current 10 CFR 50.59 regulation. Therefore, NRC approval of this change in methodology is required.
The proposed changes to the Updated Safety Analysis Report (USAR) are provided for information.
Through benchmarking, it has been demonstrated that the use of the GOTHIC computer code for the HELB response analyses produces results that are consistent with the current licensing basis computer code (THREED).
2.0 PROPOSED CHANGE
This amendment request provides the basis for revising the current HELB analysis method for the Auxiliary and the Containment Buildings from the current vendor supplied THREED code to the GOTHIC code.
The changes will affect RBS USAR Appendix 3B and USAR Sections 6.2.1.1.3.2.1 and 6.2.1.2 as shown in Attachment 2.
3.0 BACKGROUND
GOTHIC is a general purpose volumetric 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 response due to Design Basis Accidents such as Loss of Coolant Accidents (LOCA), and containment subcompartment pressurization response to the full spectrum of high energy line breaks. This code is also used for calculation of room temperature response due to failed or degraded room cooling systems, and calculation of temperature profiles for equipment to RBG-45940 Page 2 of 13 qualification, inadvertent system initiation, and degradation or failure of engineered safety features.
Numerical Applications, Inc. (NAI) developed the GOTHIC code for the Electric Power Research Institute (EPRI).
GOTHIC is qualified under the NAI QA program which conforms to the requirements of 10CFR50, Appendix B with error reporting in accordance with 10CFR21. Other plants, such as Joseph M. Farley Nuclear Plant, Units 1 and 2 and Waterford 3 have used the GOTHIC computer code to perform containment response analysis. Other sites have already used GOTHIC for HELB analyses and room heatup analysis. The Waterford 3 GOTHIC models were developed for LOCA analyses. These models were benchmarked against the current licensing containment response analysis code for these plants with good agreement between the two code results.
As part of River Bend initial licensing, pressure response analyses were performed for the various volumes containing high-energy piping.
A detailed discussion of the line breaks selected, vent paths, room volumes, analytical methods, pressure results, etc, has been provided in Updated Final Safety Analysis Report (USAR) section 6.2.1.2 for containment subcompartments and in Appendix 3B for subcompartments located outside the containment.
The NRC staff reviewed the information and performed an independent analysis of the subcompartment environmental conditions following an HELB as discussed in Supplemental Safety Evaluation Report 3 of NUREG-0989.
USAR Section 3.6A defines the complete set of break locations in the high energy piping outside containment from which the design basis breaks for subcompartment pressurization were selected. The definitions for high energy and criteria for protection against dynamic effects associated with postulated rupture of piping are also given in Section 3.6A. The re-analysis did not affect the break locations previously identified.
USAR Appendix 3B provides the design bases, design features, and design evaluation for the pressure response analyses performed for the structural design basis of the main steam tunnel and other subcompartments in the Auxiliary Building for postulated "ruptures of high-energy piping.
In addition to the use of THREED to conduct pressurization analysis, this code was also used to provide equipment qualification (EQ) environmental data. A number of models of Containment and Auxiliary Building areas were constructed to determine the necessary EQ parameters. As with the subcompartment pressurization analysis, GOTHIC will be available for use to conduct future EQ analyses.
The use of the GOTHIC code is proposed for the in-house HELB analyses at River Bend Station since an updated HELB model cannot be maintained with the THREED code. The GOTHIC code also provides improvements in capabilities and modeling when compared to the previous THREED code.
In the new analyses, the mass and energy release rates for the postulated HELBs have been updated to account for as-built plant conditions (leak detection system logic delay times, isolation valve stroke times, etc.). The mass and energy releases also account for the effects of pipe friction; this had only been considered in certain cases before. The HELB to RBG-45940 Page 3 of 13 model description and pressure transient plots in USAR Appendix 3B will be updated correspondingly after NRC approval.
At RBS, a modification was initiated to add additional delay time to the initiation logic for the temperature isolation of high energy lines in the Auxiliary and Containment Buildings.
The HELB analyses for line breaks in Auxiliary and Containment Buildings are impacted due to the additional time delays.
In order to support the proposed modification, Auxiliary and Containment Building GOTHIC models were constructed to perform the HELB analyses.
Although the additional time delay should be treated as an input parameter which does not require explicit NRC approval, the change in the analysis code from THREED to GOTHIC does present a deviation in an evaluation methodology according to the current 10 CFR 50.59 regulation. This deviation in methodology is the result of the detail contained in USAR Section 3.6A "Protection Against Dynamic Effects Associated With The Postulated Rupture Of Piping,"
Appendix 3B "Pressure Analysis For Subcompartments Outside Containment" and USAR Section 6.2 "Containment Systems."
The THREED computer program used in the initial design and licensing is similar to RELAP4 and will give the same results as RELAP4 if similar options are chosen.
THREED was formulated to perform sub-compartment analyses with capabilities and options extended beyond those available in RELAP4. A significant improvement in THREED was that the homogeneous equilibrium model (HEM) was extended to include two-phase, two-component flow that is encountered in sub-compartment analysis.
4.0 TECHNICAL ANALYSIS
The Auxiliary and Containment Building HELB analyses were initially performed using computer code THREED, to support the design basis structural analysis. Several THREED models have been constructed for the Auxiliary and Containment Building HELB cases. The RBS USAR Appendix 6B has a detailed description of the major features of THREED code. The THREED computer program is used to calculate the transient conditions of pressure, temperature, and humidity in various sub-compartments following a postulated rupture in a moderate-or high energy pipeline. The results obtained from THREED analyses are used to calculate loads on structures and to define environmental conditions for equipment qualification.
The new RBS HELB models use the GOTHIC code, which has been qualified at RBS. GOTHIC and THREED codes are similar in most aspects. Both codes use control volumes (i.e., nodes),
flow paths (i.e., junctions), valve/door models, fan models, and thermal conductors (i.e., heat sinks), etc.
Both codes have time dependent boundary condition capabilities.
Thus, no significant difference would be expected between these two codes when evaluating identical configurations.
The GOTHIC code is a general-purpose thermal-hydraulics computer program developed by NAI (Numerical Applications, Inc.) under EPRI sponsorship for design, licensing, safety and operating analysis of nuclear power plant containments and other confinement buildings.
Applications of GOTHIC include evaluation of containment and containment sub-compartment response to the full spectrum of high-energy line breaks within the design basis envelope as to RBG-45940 Page 4 of 13 described in USAR Chapter 6, Section 2. Applications may include pressure and temperature determination, equipment qualification profiles and thermal-hydraulic responses to inadvertent system initiation, and degradation or failure of engineered safety features.
GOTHIC is qualified under the NAI QA program which conforms to the requirements of 10CFR50 Appendix B with error reporting in accordance with 10CFR21. NAI has validated and verified the GOTHIC code for its intended purpose. The code validation and verification is documented in a code Qualification Report prepared by NAI for EPRI.
The validation and verification objective was to demonstrate the applicability of GOTHIC for use as a best-estimate containment analysis code. In addition to the above validation and verification efforts, GOTHIC has been extensively compared to other codes such as CONTEMPT.
The GOTHIC code qualification was performed by the comparison of GOTHIC solver predictions to solutions of analytic problems and to experimental data for containment applications. The objective was to approach qualification on the basis that GOTHIC is intended to be used as a best-estimate containment analysis and volumetric thermal-hydraulic analysis code.
4.1 Differences Between GOTHIC and THREED Based on the description of the GOTHIC and THREED codes, the table below presents a comparison of significant assumptions used in these two codes as applied at RBS. It clearly shows that a more accurate model can be developed by using the GOTHIC code. Due to the improved accuracy in the model, the new analysis results may slightly differ from those obtained with THREED. However, since the GOTHIC code has been extensively studied against both the analytic and experimental problems, no significant change due to the software (vice input parameters or evaluation options utilized) should be expected. The table below is a comparison of assumptions between THREED and GOTHIC:
THREED (USAR App. 6B)
IGOTHIC Homogeneous flow, unless the Moody Inter-phase mass, energy and momentum transfer rates choking option is chosen obtained through constitutive relation.
Thermodynamic equilibrium in each node Separate mass equation solved for each fluid phase, gas component and ice phase. Separate energy equation solved for each fluid phase.
Incompressible form of the momentum Compressible flow for all fluid phases.
equation.
Valve open or close instantaneously Can model valve closure time.
Water, if present, occupies the entire Water in liquid phase can be accumulated at the bottom of a volume, i.e., a homogenous mixture of control volume.
vapor and liquid is assumed Air is assumed to be perfect gas Can model actual air properties. But treat air as ideal gas for mixture calculations.
If air & liquid water are present, the water Can have RH values other than 0% or 100%.
vapor is saturated (RH=100%)
If air is present, liquid water conditions are Water in vapor phase dependent upon momentum, mass the saturated condition and energy equations.
to RBG-45940 Page 5 of 13 Note: In the GOTHIC HELB model, the drop-liquid conversion option in the GOTHIC code is not active for the benchmark model. With this option active, GOTHIC can have a liquid pool on the control volume floor, which will effectively reduce the drop phase fraction inside the control volume. THREED assumes that the air/steam/liquid are mixed uniformly and suspended in the air, which is conservative.
4.2 Benchmark The break locations used in the original analysis remain identical for the benchmark. The mass and energy releases for the benchmark were also identical to those used in the initial analysis.
For benchmark purpose, the GOTHIC model of the 6 inch Reactor Water Cleanup system (RWCU) line double ended rupture (DER) in the heat exchanger room was constructed, which duplicates the inputs in the THREED models. The 6 inch DER is chosen because it is the limiting long term pressurization case. The Pressure/Temperature transients as well as the peak Pressure/Temperature values from both models were compared to verify that the use of GOTHIC code is consistent with the approved THREED code that was used in the original design calculations.
For the HELB benchmark analysis inside the containment, the GOTHIC code used a Homogeneous Equilibrium Model (HEM), which is also used in THREED. The Uchida heat transfer coefficient was applied and the condensate revaporization is 100 percent.
The THREED code was used in previous revisions to obtain the pressure transients for the HELB inside the RWCU heat exchanger room model.
For benchmark purpose, a GOTHIC benchmark model was constructed, which matched the THREED model as closely as possible. All the run parameters in the GOTHIC benchmark model were forced to simulate THREED run parameters. The results obtained in the GOTHIC benchmark model were then compared to the THREED results to verify that the use of GOTHIC is capable of producing results that do not depart from results obtained with THREED.
For conservatism, the vertical ventilation duct in the RWCU heat exchanger room was assumed to remain in place and partially block the flow path out of the RWCU heat exchanger room.
Heat sinks were modeled to consider the effect of concrete and steel slabs inside the containment.
For conservatism, the shield building annulus was included in the model and three external thermal conductors have been modeled to connect the shield building annulus with other containment volumes.
This creates heat conduction paths that could add more energy into the containment volumes, which is conservative.
As shown by the results the THREED and GOTHIC benchmark models are in close agreement.
to RBG-45940 Page 6 of 13 4.2.1 Benchmark Model Results The comparisons of the Pressure/Temperature transient results in both the GOTHIC benchmark and THREED models show no significant difference in peak pressures between the benchmark GOTHIC and the THREED models. The differences in peak pressures are less than 0.5%.
Negligible difference exists between the peak temperatures in the nodes containing the break and those immediately connected for the benchmark GOTHIC and THREED results.
For temperatures in these areas consistent results are obtained in the benchmark GOTHIC model.
A larger (less than 2%) difference exists between the peak temperatures for down stream areas in the benchmark GOTHIC and THREED models where the magnitude of the increase is lower.
This difference could be a result of the small differences in the vent path (junction) modeling between the GOTHIC and THREED codes. The junction modeling in the GOTHIC code is more accurate than the THREED code, but needs more input parameters.
Results Summary for the GOTHIC Benchmark and THREED Models 6 inch DER of RWCU line Node Peak Pressure Peak Pressure Peak Temperature (F)
Peak Temperature (F)
(psia)
(psia)
15.86 15.807 213.35 212.85 2
15.49 15.480 200.43 199.38 3
15.52 15.501 188.05 185.34 4
15.49 15.488 103.29 103.08 As shown in the results the original THREED and benchmark GOTHIC models provide close agreement when modeling the same volumes with identical mass and energy inputs. As a result, the GOTHIC models have been successfully benchmarked against the THREED code for the HELB analysis.
to RBG-45940 Page 7 of 13 4.3 New HELB Models and Revised Results As discussed above, the HELB GOTHIC code has been qualified at RBS. Also, the break locations used in the original analysis remain identical for the revised analysis conducted with GOTHIC.
For the revised analysis in the Auxiliary and Containment Buildings, the mass and energy release include the proposed addition of a 5-second time delay. This will result in the extension of the upstream steady-state blowdown time due to the proposed additional logic delay time for the isolation valves. Credit has also been taken for friction; the use of friction in the HELB analysis is consistent with previous THREED analysis as identified in USAR Appendix 3B. As a result, the magnitudes of the mass and energy blowdown rates are expected to be reduced after crediting friction.
4.3.1 New HELB GOTHIC Models In the revised analysis, all the parameters (control volumes, vent paths, thermal conductors) in the GOTHIC model have been updated with current plant conditions and configurations. The high-energy line break locations remain the same as in the THREED HELB analyses. The mass and energy releases are updated with the new blowdown data assuming the additional time delays and crediting flow friction. GOTHIC, unlike THREED, also has the ability to model break flow as liquid or as drop flow.
The room pressurization due to a HELB has the potential to damage the heating and ventilation ducting which can pass through the subcompartment. As a result, pathways can exist which are not normally in communication with the air volume of the subject room.
If the HELB pressurization transient is sufficient to cause duct destruction, a new penetration can create an opening to an adjacent room. The duct flow paths added to the HELB model use the most restrictive flow area (duct area or register area) for the purpose of calculating flow area and hydraulic diameter. Small duct flow paths are not considered. Two cases for each line break have been modeled: duct-destruction (DD) case and non-duct-destruction (NDD) case. The DD case generates more limiting pressure / temperature transients for the subcompartments close to the break room, while the NDD case generates more limiting pressure / temperature transients for the subcompartments that are not adjacent to the break room. The most limiting pressure I temperature transient was used for each subcompartment.
4.3.2 Revised HELB Analysis Results Using the new HELB model the revised mass and energy blowdown calculations for the Containment Building are crediting friction for the upstream steady-state critical flow only. The mass release rates were calculated based on either Moody critical flow model or Henry-Fauske subcooled critical flow model with conservative assumptions on the fluid conditions. The vent path parameters were set to compressible, Critical Flow Model (HEM), and zero entrainment, which is consistent with the NRC Standard Review Plan guidelines for subcompartment analysis (Standard Review Plan Section 6.2). The peak and differential pressures are 16.286 psia and to RBG-45940 Page 8 of 13 1.627 psid in the RWCU Heat Exchanger room and 24.969 psia and 10.425 psid in the RWCU filter/demineralizer room. The current calculated pressures are in USAR Tables 6.2-26 and 6.2
- 29. The results of the revised analysis, which included the additional instrument delay, remain within the subcompartment design limits of 5.0 psid in the RWCU Heat Exchanger room and 21.0 psid in the RWCU filter/demineralizer room.
In the Auxiliary Building, the most limiting case for the subcompartment pressurization in the revised Auxiliary Building HELB analyses is the 8 inch RHR HELB. The peak pressure of 16.5 psia (i.e., 1.8 psid) is about 0.5 psi lower than the originally calculated peak pressure as in USAR Table 3B-3. This peak pressure is also much lower than the design peak differential pressure, which is 3.30 psid and 2.40 psid for all other zones.
More conservatism could be credited since the differential pressures were calculated by subtracting the calculated peak pressure with the environmental pressure (assumed 14.7 psia) instead of the pressure of other EDC zones.
Therefore, the new results have no significant impact on the subcompartment pressurization analyses.
The 8 inch Residual Heat Removal (RHR) HELB in the Auxiliary Building is not impacted due to the high steam flow isolation signal which can be credited for this line break.
The HELB locations in the Drywell and Main Steam Tunnel were not affected by this change in the leak detection system.
5.0 REGULATORY ANALYSIS
Due to the fact that the assumptions and methodology used in mass and energy release calculations slightly deviate from the original design calculations and the code used for the HELB model has been changed from THREED to GOTHIC, the HELB re-analysis represents a deviation in an evaluation methodology as described in the USAR, thus the 50.59 evaluation results in a License Amendment Request.
5.1 Applicable Regulatory Requirements/Criteria The proposed changes have been evaluated to determine whether applicable regulations and requirements continue to be met.
NRC regulatory guidance applicable to this change includes Standard Review Plan (SRP),
NUREG-0800 Sections 3.6.2 "Plant Design for Protection Against Postulated Piping Failures in Fluid Systems Outside Containment," and 6.2.3 "Secondary Containment Functional Design."
Both of these sections discuss the requirement for the systems and structure to demonstrate compliance with General Design Criteria 4 as it relates to the ability to accommodate the effects of postulated accidents.
The requirements and guidance contained in these documents continue to be applied and no changes are needed.
Additional guidance for the analysis models and calculational methods is provided in SRP Section 6.2.1.2. A comparison of the compliance to this SRP guidance is summarized below.
to RBG-45940 Page 9 of 13 SRP compliance of the THREED and GOTHIC Models:
SRP Section THREED Models (Auxiliary and GOTHIC HELB Models Containment Buildings)
SRP 6.2.1.2, Same as SRP guidelines. To GOTHIC HELB model is the same as Section ll.B.1:
maximize the differential THREED model pressure, the 0% relative humidity is assumed. To maximize the peak temperatures for EQ purpose, 100% relative humidity is assumed.
SRP 6.2.1.2, Different models have been The GOTHIC models are consistent with the Section ll.B.2:
developed to obtain pressure /
THREED models.
temperature responses for both sub-compartment pressurization and EQ purposes.
SRP 6.2.1.2, Conservative assumptions are The GOTHIC HELB models are updated with Section ll.B.3:
used in the THREED HELB the as-built plant configurations.
Models SRP 6.2.1.2, HEM for nodes and vent paths, Same as the THREED models except that Section ll.B.4:
100% water entrainment, HEM the drop-liquid conversion option is used in critical flow model, uniformly Containment. The three phase modeling water-steam mixture which option was used in the Auxiliary Building. A occupies the whole volume, etc.
comparison of the change showed negligible difference.
SRP 6.2.1.2, The peak pressures in the sub-The peak pressures in the sub Section ll.B.5:
compartments and the peak compartments and the peak differential differential pressures across the pressures across the walls have been walls have been verified to be verified to be within the acceptance limits.
within the acceptance limits.
Heat Transfer Uchida Uchida specified in Containment. The Coefficient Type GOTHIC default model (similar to Uchida) has been used in the Auxiliary Building case.
Sensitivity studies indicate negligible impact due to this difference.
Heat Sinks Most THREED models credited Heat sinks credited heat sinks As discussed above, this change to the method of evaluation used break locations consistent with the original basis of the plant. The mass and energy inputs remain consistent with the initial licensing with updates to current plant configuration. The change to the methodology for determining the pressure-temperature response to the HELB is changed to a more current and available code.
Therefore, this change continues to demonstrate compliance with General Design Criteria 4.
Generic Letter (GL) 83-11 Supplement 1, provides guidance regarding licensee qualification for performing their own safety analyses including containment response analysis. This guidance includes a requirement to institute a program which includes training, procedures, comparison calculations (benchmarking) and continued quality controls. EOI application of this version of GOTHIC is controlled through established EOI procedures which include Software Control and to RBG-45940 Page 10 of 13 Calculation Procedures. These procedures include independent verification and review under EOI's Quality Control program. EOI training on GOTHIC has included:
"* The code developer, NAI, has provided training to EOI engineers both in training sessions conducted in conjunction with GOTHIC Advisory Group meetings and in an EOI sponsored training session conducted at corporate headquarters.
"* Example test cases compiled by NAI are modeled and run by engineers as part of the code familiarization process. Before performing calculations using GOTHIC, engineers read and become familiar with the GOTHIC Users Manual and other technical background information for the GOTHIC application.
"* Lessons learned and expertise regarding GOTHIC is shared with EOI plants, including through periodic discussions of GOTHIC issues as part of regular EOI Safety Analysis conference calls. Note that an EOI engineer previously served as the Chairman of the GOTHIC Advisory Group.
"* Consistent with GL 83-11 Supplement 1, Entergy's software control procedure contains provisions for evaluating vendor code, updates and for informing code vendors of any problems or errors discovered while using the code.
Thus, Entergy has established expectations for developing and demonstrating capabilities for use of analysis codes such as GOTHIC which are consistent with Generic Letter 83-11 supplement 1. Additionally, as a member of the GOTHIC Advisory Group, EOI and River Bend have the ability to consult and exercise the GOTHIC code developer (NAI) on GOTHIC model development or detailed coding issues.
Based on the above discussions, Entergy has determined that the proposed changes do not require any exemptions or relief from regulatory requirements, including the Technical Specifications, and do not affect conformance with any GDC differently than described in the SAR.
6.0 NO SIGNIFICANT HAZARDS CONSIDERATION The proposed change will revise Appendix 3B and Section 6.2.1.2 of the Updated Safety Analysis Report pertaining to the method of analysis. The proposed change will replace the current vendor THREED code for room pressure-temperature analyses due to High Energy Line Breaks (HELB) with GOTHIC (Generation of Thermal-Hydraulic Information for Containments).
The proposed change will allow EOI to update the analysis and to evaluate additional changes to the plant.
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:
to RBG-45940 Page 11 of 13
- 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 involves no increase in the probability of the accidents previously evaluated since no physical change to the plant will be made. The change of the High Energy Line Break (HELB) analysis method does not affect the probability of the analyzed event occurring. The line break locations have not been affected and remain as originally designed.
This submittal is required due to the change of HELB analysis code from the vendor code THREED to the modern industry standard analysis code GOTHIC.
This is a change in the methodology for determining the effects of the mass and energy release in the plant as a result of currently postulated events.
The change in the evaluation methodology has been benchmarked and reviewed to confirm the results remain consistent with the current analysis. The changes to the model used for the additional analysis allow the use of new, more physically realistic models for Containment and Auxiliary Building pressure / temperature responses and will demonstrate continued qualification of the equipment in these buildings. Mass and energy releases for some cases have also been recalculated to credit pipe friction, which was only credited for certain cases previously.
With these new results the equipment has been reviewed and remains qualified per current programs established at RBS. Therefore, the plant will continue to function as designed and thus there will be no impact on consequences.
- 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
No physical change to the plant will be made. The HELB locations were identified by reviewing all the possible break locations in each Auxiliary and Containment Building volume containing high-energy lines. The locations of the breaks remain the same as the previous HELB analyses. The HELB analyses have been evaluated for the current plant configuration.
The new HELB analysis has been benchmarked against the previous accepted methods and found to correlate with the previous analysis. Therefore the results can be used to predict plant responses to events. The proposed change uses improved methods for mass and energy release calculation and pressure /
temperature responses to determine the EQ qualification envelopes. Therefore, no new or different interaction would be created.
to RBG-45940 Page 12 of 13
- 3.
Will the operation of the facility in accordance with these proposed changes involve a significant reduction in a margin of safety?
Response
The operation of the facility in accordance with the proposed changes will not involve a significant reduction in a margin of safety.
The GOTHIC code has been successfully benchmarked versus the vendor THREED code, which was used in the original design calculations. The HELB analysis results with the benchmarking GOTHIC model are consistent with the THREED results.
Therefore, the use of GOTHIC code will not involve a reduction in an identified margin of safety.
Given that GOTHIC code is an improved methodology and it has been extensively qualified against the solved analytical problems and testing results, the use of GOTHIC code will produce more accurate pressure / temperature responses for the HELB analyses.
The use of the GOTHIC code has been approved for pressure/temperature responses analysis at various other plants including Joseph M.
Farley Nuclear Plant, Units 1 and 2, and Waterford 3.
The results with the revised methods will be used to show that safety equipment meets the EQ requirements.
The peak temperatures and pressures in the HELB GOTHIC benchmark model are within the existing EDC envelopes. Therefore, the pressure /
temperature responses from the HELB benchmark analyses have no impact on the equipment qualification.
The methodology in the original design calculations is very conservative. The mass and energy releases without crediting friction introduce excessive amount of high-energy fluid into the break rooms, which is unrealistic. Some HELB calculations have credited both the frictional flows and the additional zone to eliminate excessive conservatism in the pressure/temperature responses. There is no reduction in a margin of safety and the design room differential pressure limits continue to be meet.
The use of this method by EOI RBS is consistent with the guidance given in NRC Generic Letter 83-11 and Supplement 1, addressing the performance of safety analyses by licensees.
EOI has implemented this guidance for the GOTHIC methodology consistent with the intended application. The GOTHIC methodology has been verified and validated by the software vendor. In addition this methodology is controlled by EOI procedures and under the EOI quality assurance program. This includes EOI and RBS specific verification and validation of this application of GOTHIC and review of the calculations performed.
Based on the above review, it is concluded that: (1) the proposed 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 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.
to RBG-45940 Page 13 of 13
7.0 ENVIRONMENTAL CONSIDERATION
S The proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.
References
"* NUREG-0800, USNRC Standard Review Plan.
"* USAR Section 3.6.1, Plant Design for Protection Against Postulated Piping Failures in Fluid System Outside Containment.
"* USAR Section 3.6.2, Determination of Rupture Locations and Dynamic Effects Associated with the Postulated Rupture of Piping.
"* USAR Section 6.2.1.2, Containment Subcompartments.
"* NEDO-20533, Mark III Containment System Analytical Model, Appendix B, Pipe Inventory Blowdown, June 1974.
"* Lahey, R.T. and Moody, F.J., The Thermal-Hydraulics of a Boiling Water Nuclear Reactor, ANS, 1977.
USAR Sections PROPOSED (MARKED-UP) USAR SECTIONS: See Attachment 2
ATTACHMENT 2 PROPOSED MARKED-UP USAR SECTIONS
USAR Section 6.2.1.2
RBS USAR CHAPTER 6 LIST OF TABLES (Cont)
Table Number Title 6.2-24 SUBCOMPARTMENT VENT PATH DESCRIPTION 6-IN RCIC HEAD SPRAY LINE BREAK DRYWELL HEAD SUBCOMPARTMENT 6.2-25 BLOWDOWN DATA 6-IN RCIC HEAD SPRAY LINE BREAK DRYWELL HEAD SUBCOMPARTMENT 6.2-26 SUBCOMPARTMENT NODAL DESCRIPTION 4-IN AND 6-IN RWCU LINE BREAK RWCU HEAT EXCHANGER ROOM 6.2-27 SUBCOMPARTMENT VENT PATH DESCRIPTION 4-IN AND 6-IN RWCU LINE BREAK RWCU HEAT EXCHANGER ROOM 6.2-28 BLOWDOWN DATA 6-IN RWCU LINE BREAK RWCU HEAT EXCHANGER ROOM 6.2-29 SUBCOMPARTMENT NODAL DESCRIPTION 8-IN RWCU LINE BREAK RWCU FILTER/DEMINERALIZER ROOM 6.2-30 SUBCOMPARTMENT VENT PATH DESCRIPTION 8-IN RWCU LINE BREAK RWCU FILTER/DEMINERALIZER ROOM 6.2-31 BLOWDOWN DATA 8-IN RWCU LINE BREAK RWCU FILTER/
DEMINERALIZER ROOM 6.2-32 SECONDARY CONTAINMENT 6.2-33 PRIMARY CONTAINMENT OPERATION FOLLOWING A DESIGN BASIS ACCIDENT 6.2-34 SECONDARY CONTAINMENT OPERATION FOLLOWING A DESIGN BASIS ACCIDENT 6.2-35 CRITERION 55 -
INFLUENT LINES, REACTOR COOLANT PRESSURE BOUNDARY 6.2-36 CRITERION 55 -
EFFLUENT LINES, REACTOR COOLANT PRESSURE BOUNDARY 6.2-37 CRITERION 56 -
PRIMARY CONTAINMENT ISOLATION PIPES THAT PENETRATE THE CONTAINMENT AND CONNECT TO THE CONTAINMENT ATMOSPHERE August 1987 6 -xi
RBS USAR Figure Number 6.2-38 6.2-39 6.2-40 6.2-41 6.2-42 6.2-43 6.2-44 6.2-45a through 6.2-45e 6.2-46 6.2-47 6.2-48 6.2-49 6.2-50 6.2-51 6-xvii August 1987 CHAPTER 6 LIST OF FIGURES (Cont)
Title NODALIZATION DIAGRAM FEEDWATER LINE BREAK RPV-SHIELD WALL ANNULUS 27 NODE MODEL NODAL PRESSURES FEEDWATER LINE BREAK RPV-SHIELD WALL ANNULUS 27 NODE MODEL NODAL PRESSURE DIFFERENTIALS FEEDWATER LINE BREAK RPV-SHIELD WALL ANNULUS 27 NODE MODEL NODALIZATION DIAGRAM FEEDWATER LINE BREAK RPV-SHIELD WALL ANNULUS 25 NODE MODEL NODAL PRESSURES FEEDWATER LINE BREAK RPV-SHIELD WALL ANNULUS 25 NODE MODEL NODAL PRESSURE DIFFERENTIAL FEEDWATER LINE BREAK RPV-SHIELD WALL ANNULUS 25 NODE MODEL NODALIZATION DIAGRAM RECIRCULATION OUTLET LINE BREAK RPV-SHIELD WALL ANNULUS 26 NODE HALF MODEL NODAL PRESSURES RECIRCULATION OUTLET LINE BREAK RPV-SHIELD WALL ANNULUS 26 NODE HALF MODEL NODAL PRESSURE DIFFERENTIAL RECIRCULATION RPV-SHIELD WALL ANNULUS 26 NODE HALF MODEL NODALIZATION DIAGRAM 6-IN RCIC HEAD SPRAY LINE BREAK DRYWELL HEAD SUBCOMPARTMENT NODAL PRESSURES 6-IN RCIC HEAD SPRAY LINE BREAK DRYWELL HEAD SUBCOMPARTMENT NODAL PRESSURE DIFFERENTIALS 6-IN RCIC HEAD SPRAY LINE BREAK DRYWELL HEAD SUBCOMPARTMENT NODALIZATION DIAGRAM 4-IN AND 6-IN RWCU LINE BREAK RWCU HEAT EXCHANGER ROOM NODAL PRESSUR 6-IN RWCU LINE BREAK RWCU HEAT EXCHANGER ROOM I Deleted: s
RBS USAR CHAPTER 6 LIST OF FIGURES (Cont)
Figure Number 6.2-52 6.2-53 6.2-54 6.2-55 6.2-56 6.2-57 6.2-58 6.2-59 6.2-60 6.2-61a 6.2-61b 6.2-62 6.2-63 6.2-64 6.2-65 6.2-66 6.2-67 6.2-68
. Deleted: NODAL PRESSURE
- DIFFERENTIALS 6-IN RWCU LINEI
, BREAK RWCU HEAT EXCHANAGER I ROOM I
Deleted: NODAL PRESSURE DIFFERENTIALS 8-IN RWCU LINEIL
.BREAK RWC1J FILTER DEMINERALIZER ROOM 6-xviii August 1987 Title pEETED....
NODALIZATION DIAGRAM 8-IN RWCU LINE BREAK RWCU FILTER DEMINERALIZER ROOM NODAL PRESSURES 8-IN RWCU LINE BREAK RWCU FILTER DEMINERALIZER ROOM DELETED RHR SUPPRESSION POOL COOLING MODE SUCTION AND RETURN (PLAN)
RHR SUPPRESSION POOL COOLING MODE SUCTION AND RETURN (SECTION 1-1)
STANDBY GAS TREATMENT SYSTEM (P&ID)
SCTS FAN PERFORMANCE CURVE FUEL BUILDING CHARCOAL FILTRATION SYSTEM F PERFORMANCE CURVE PRESSURE IN SHIELD BUILDING ANNULUS VS TIM PRESSURE IN AUXILIARY BUILDING VS TIME PRESSURE IN FUEL BUILDING VS TIME CRITERION 55 -
CONTAINMENT ISOLATION VALVE CRITERION 56 -
CONTAINMENT ISOLATION VALVE CRITERION 56 -
CONTAINMENT ISOLATION VALVE HYDROGEN MIXING PURGE AND RECOMBINER P&ID DELETED HYDROGEN CONCENTRATION VS TIME AFTER LOCA
RBS USAR therefore, guard pipes are not provided for these systems. Other process lines with check valves inside the drywell such as RCIC head spray and RHR shutdown cooling have guard pipes because these lines can be used during normal plant operation, after which it could be postulated that the check valve sticks in the open position.
6.2.1.1.3.2.1 Reactor Water Cleanup Break The reactor water cleanup (RWCU) pumps are located outside the containment.
RWCU heat exchangers and filter demineralizers are located inside the containment.
This system, when operating, is in direct communication with the reactor coolant system, taking suction on the recirculation lines inside the drywell and injecting back into the feedwater lines.
Breaks in this system result in the release of high energy fluid into the containment.
The mass loss into the containment is terminated by automatic isolation of the RWCU suction and discharge lines upon detection of the leak.
Isolation valves immediately inboard and outboard of the drywell and containment penetrations are provided to perform this function.
Check valves in the discharge line prohibit back flow from the feedwater line in the event of a break inside the containment.
.-- 12 Automatic isolation of the RWCU system in the event of a
postulated line break is initiated by two separate leak detection systems.
First, leakage is detected by means of flow comparison between RWCU system inlet and outlet.
If the inlet flow exceeds the outlet flow by approximately 7 percent of rated flow, an alarm is actuated and an automatic isolation of the system initiated.
In addition to the flow comparison method, leakage is detected by means of temperature sensing elements.
Redundant temperature sensors are located locally to monitor the ambient temperature in all compartments containing equipment and piping for this system.
Signal times to initiate closure of the system isolation valves are on the order of 1 sec for both detection systems described.
12+-e.---
6 The analyses show that the local temperature in the RWCU heat exchanger room rises from 103 0 F to 153OF in 0.4 sec, and the local temperature in the RWCU filter/demineralizer room rises from 105 0 F to 113 0 F in 0.5 sec.
Thus, the leak detection system high ambient temperature signal to isolate the RWCU system would be generated in less than 1 sec.
6.2-24 December 1999 Deleted: in Revision 12
RBS USAR The postulated DER of the 4-in RWCU pump discharge line between the inboard containment isolation valve and the regenerative heat exchangers is the limiting case for containment pressurization.
This break location is shown schematically on Fig. 6.2-26.
Blowdown from the RWCU pump discharge side of the break is initially choked at the 0.0192-sq ft flow restrictor in the pump discharge line.
The leak detection signal initiates automatic isolation of the system within., When the isolation valves have closed sufficiently such that the isolation valve flow area equals the flow restrictor area,,_the critical flow location_[
changes from the flow restrictor to the isolation valves.
Flow from the heat exchanger side of the break is limited to critical flow through the pipe cross-sectional area and is assumed to terminate when the contents of the h
.eat exchangers and Filter/Demineralizers are exhausted.
For all pipe breaks considered in the RWCU system, the peak subcompartment pressures occur before isolation valve closure begins to limit the blowdown.
It should be noted that the valve closure does not influence the blowdown until the valve open area equals the flow restrictor area of 0.0192 sq ft, as flow is choked at the flow restrictor.
Accordingly, the assumed linear valve closure characteristic is conservative for the gate valves used in this application.
Table 6.2-12 summarizes the 4-in RWCU pump discharge line blowdown used in this analysis.
Based on the initial conditions given in Table 6.2-3, this break produces an increase in containment internal pressure of less than 1.0 psig..which is well below the design internal pressure of 15 psig.
6.2.1.1.3.2.2 Instrument Line Break Instrument lines penetrating the drywell wall are provided with 1/4-in orifices located upstream of the drywell penetrations to preclude containment over-pressurization.
In the event of a
- rupture, containment pressure increases until shortly after the operator starts reactor cooldown.
Under the assumption that the operator takes 1/2 hr to detect an instrument line rupture and start reactor cooldown, the rise in containment pressure is only 0.42 psig for a liquid line.
For a steam line break, the pressure rise is less.
Deleted: the analysis, the instrument delay time is assumed to be 1 sec. ¶ Deleted: 1 sec after the break.
At 5.5 sec, Deleted:.
At that time, Deleted: Subsequent closure of the valves terminates flow at 6.0 sec.
I-Deleted: regenerative Deleted: o.s6 6.2-25 August 1987
RBS USAR within the prescribed limits and the action to be taken if these conditions are exceeded is discussed in Section 9.4.6.
The loss of these systems does not result in exceeding the design operating conditions for the safety-related equipment inside the containment.
The safety-related containment systems described in Sections 6.2.2 and 6.5 maintain required containment atmosphere conditions after a LOCA.
6.2.1.1.3.7.5 Instrumentation Refer to Sections 6.2.1.7, 7.2, 7.3, 7.5, and 7.6 for a
discussion of instrumentation inside the containment used for monitoring various containment parameters.
6.2.1.2 Containment Subcompartments 6.2.1.2.1 Design Bases The containment subcompartments are designed in accordance with the following criteria:
- 1.
A pressure response analysis is given for each containment subcompartment containing high energy piping in which breaks are postulated.
The definition of high energy piping and the criteria for postulating breaks are outlined in Section 3.6.
The break which, by virtue of its size and location, produced the greatest release of blowdown mass and energy into the subcompartment, during normal operation and hot standby condition, is selected for the design evaluation.
The breaks used in the design evaluations are listed in Section 6.2.1.2.3.
- 2.
All circumferential breaks are considered to be fully double-ended and no credit for limiting blowdown generation is taken due to pipe restraint locations.
The effective cross-sectional flow area of the pipe is used in the jet discharge evaluation for breaks.
- 3.
The design pressure differentials for all subcompartments are higher than the calculated peak pressure differentials resulting from the design basis pipe breaks.
August 1987 6.2-42
RBS USAR 6.2.1.2.2 Design Features The containment includes the following four subcompartments:
- 1.
Reactor Pressure Vessel-Shield Wall Annulus The 2 ft thick cylindrical primary shield wall which surrounds the RPV has an outside diameter of 29 ft 10 in and extends from the vessel pedestal to el 147 ft 6 in.
Breaks in the recirculation water outlet piping and feedwater piping are analyzed.
- ---12
- 2.
Drywell Head The drywell head is located above the RPV head and surrounds the RPV head, connecting to the drywell bulkhead at el 162 ft 3 in.
Five normally open ventilation exhaust hatches are located in the bulkhead at azimuths 30, 75,
- 165, 225, and 345 deg venting into the drywell.
(These hatches are closed only during refueling.)
Line Breaks were evaluated for the RCIC head spray line.
Although the head spray line was
- removed, the break analysis will remain in place because the analysis bounds a vessel head vent line break.
12+-.
- 3.
RWCU Heat Exchanger Room The RWCU heat exchanger room, located at el 147 ft 3 inches in the containment, vents through the wire door in the south wall and through two 13 ft x 2 ft 2 in openings in the north wall into the containment.
RWCU line breaks are analyzed in this room.
- 4.
RWCU Filter/Demineralizer Rooms The RWCU filter/demineralizer rooms are located at azimuth 270 deg and el 162 ft 3 in.
The HVAC vent openings [I-Deleted: Piping penetration provide the only vents from the filter/demineralizer sleeves rooms.
RWCU piping is routed to and from the demineralizers through the east wall of the cubicles which separates them from the holding pump room and valve nest area.
Complete circumferential DER of the 8-in diameter RWCU line connected to the bottom of the demineralizer is analyzed in this subcompartment.
Drawings depicting
- piping, equipment, and compartment/venting locations are provided in Section 3.6. The volumes and vent areas are discussed in Section 6.2.1.2.3. -
Deleted:The subcompartments described do not incorporate 6.2.1.2.3 Design Evaluation blowout panels.
No credit
.is taken for vent areas that become available after the The breaks utilized in the design evaluation of the containment Ipipe break occurs.
subcompartments are listed in Table 6.2-13.
The 6.2-43 December 1999 Revision 12
RBS USAR tables and figures which contain the nodal parameters and results for each analysis are also listed in Table 6.2-13.
- -+-14 The containment subcompartment design evaluations use the THREED_
RELAP4/MOD5(8) and GOTHIC computer codes.
Both THREED and RELAP4/MOD5 codes consider two-phase, two-component (steam-water-air) flow through the vents and account for the fluid inertia effects.
A detailed description of the THREED analytical model is provided in Appendix 6B.
The GOTHIC code considers
-he liquid, vapor and drop phases.
The blowdown mass and energy releases for each of the breaks are provided in the tables which are cross-referenced in Table 6.2-13, which are calculated based on the ucrated oower conditions (3100 M4t) and maximum reactor pressure (1090 psia).
An additional 5-second time delay in the isolation logic has been assumed for the RWCU line breaks.
The assumed initial conditions for the subcompartment volumes are conservatively chosen so as to maximize transient pressure responses.
The initial conditions are given in the subcompartment nodal description tables.
The description of and justification for the subsonic and sonic flow model, and the degree of entrainment used in vent flow calculations are given in Appendix 6B.
The piping systems assumed to rupture in the subcompartments are identified in Table 6.2-13.
Break locations are discussed in Section 3.6.
The need to determine the impact of a RCIC head spray line break inside the drywell head is eliminated with the reroute modification for the RCIC line.
Changing the injection line from the reactor spray nozzle to the
'A' feedwater line eliminates the RCIC break in the drywell head as an event and therefore this break does not need to be evaluated.
Although the RCIC break is eliminated with respect to drywell pressurization, another high energy line, the vessel head drain
- line, is also present in the drywell head.
This line is connected between the vessel head and one of the steam lines and is used to purge non-condensable gases from the vessel.
A break in this line will result in the discharge of high energy steam to the drywell head and cause pressurization of the drywell head.
- However, the break area associated with a break in the vessel drain line is significantly smaller than the break area used to calculate the mass and energy release rates applied in the USAR RCIC break calculation.
The reduction in break flow rate due to the smaller break area is much more significant than the effect Revision 14 6.2-44 September 2001 i Deleted: and i Deleted:.
For all cases, the Iblowdown data is based upon conservative methodology developed by GE using the Moody steady-slip flow model with subcooling, as described in Reference 9.
The blowdown mass and energy used in the subcompartment calculation Deleted: 102% of the original reactor
[power and original reactor pressure.
Evaluations performed at 102% of current rated power and 1072 psia reactor pressure demonstrated that due to the conservatisms in the methodology, the break mass and energy flows calculated at the original reactor power and pressure remain conservative for application to current rated power conditions.¶
RBS USAR TABLE 6.2-12 BLOWDOWN DATA 4-IN RWCU PUMP DISCHARGE LINE BREAK CONTAINMENT HIGH ENERGY LINE BREAK ANALYSIS Blowdown Mass B
Flow Rate E:
(lbm/sec) 956
/
956
/
7707
/
~707 777 07 651 651 651 651 173 1173 0
/
lowdown 448 442 442 367 36 59 29 0
rep net,,
rtA
-f August 1987 1 of 1
RBS USAR TABLE 6.2-12 BLOWDOWN DATA 4-IN RWCU PUMP DISCHARGE LINE BREAK CONTAINMENT HIGH ENERGY LINE BREAK ANALYSIS Time (sec) 0.0000 0.0001 0.5774 0.5775 13.4380 15.0000 0.0000 0.0001 0.7315 0.7316 1.5297 1.5298 2.3788 2.3789 4.6461 4.6462
- 21. 9969 21.9970 23.5400 23.5401 25.0137 25.0138 26.3952 26.3953 28.3703 28.3704 Blowdown Mass Flow Rate (lbm/sec)
Upstream Blowdown 0.0 564.7 564.7 212.1 212.1 0.00 Downstream Blowdown 0.0 610.2 610.2 610.2 610.2 610.2 610.2 610.2 610.2 610.2 610.2 610.2 610.2 610.2 610.2 610.2 610.2 610.2 610.2 0.0 August 1987 Enthalpy (Btu/lbm) 531.44 531.44 531.44 531.44 531.44 531.44 472.02 472.02 472.02 361.60 361.60 257.54 257.54 150.17 150.17 93.91 93.91 146.46 146.46 252.64 252.64 362.04 362.04 419.00 419.00 0.00 1 of 1
RBSUSAR TABLE 6.2-13 CONTAINMENT SUBCOMPARTMENT ANALYSIS
SUMMARY
Design Basis Line Break Subcompartment RPV - Shield Wall Annulus RPV - Shield Wall Annulus RPV - Shield Wall Annulus
12 Drywell Head 12+-
RWCU Heat Exchanger Room RWCU Filter/
Demineralizer Rooms Feedwater Feedwater Recirculation water outlet RCIC head ()
spray RWCU RWCU Tables Vent Nodal Path Description Description 6.2-14 6.2-17 6.2-20 6.2-23 6.2-26 6.2-29 6.2-15 6.2-18 6.2-21 6.2-24 6.2-27 6.2-30 Figures Blowdown Nodalization Data Diagram 6.2-16 6.2-19 6.2-22 6.2-25 6.2-218 6.2-31 6.2-38 6.2-41 6.2-44 6.2-47 6.2-50 6.2-53 Nodal Nodal Pressure Pressures Differentials 6.2-39 6.2-42 6.2-45 6.2-48 6.2-51 6.2-40 6.2-43 6.2-46 6.2-49 6.2-54 c} ),[,+4
(')Model of complete (3600) annulus (2) Model of half (180') of annulus due to summary (3) The RCIC head spray line has been deleted and the associated high energy line breaks are no longer possible. However this fhilure and information is being provided as the bounding conditions that were established as part of the original plant desigq and licensing basis.
Revision 12 1 of I December 1999
ýJWIrw (dCe,2,1
(,.I 1-o4re-
RBS USAR TABLE 6.2-13 CONTAINMENT SUBCOMPARTMENT ANALYSIS
SUMMARY
Subcompartment RPV - Shield Wall Annulus RPV - Shield Wall Annulus RPV - Shield Wall Annulus
- -> 12 Drywell Head 12<-
RWCU Heat Exchanger Room RWCU Filter/
Demineralizer Rooms Design Basis Line Break Feedwater( )
Feedwater(l)
Recirculation water outlet(2)
RCIC head (3) spray RWCU RWCU Nodal Description 6.2-14 6.2-17 6.2-20 6.2-23 6.2-26 6.2-29 Tables Vent Path Description 6.2-15 6.2-18 6.2-21 6.2-24 6.2-27 6.2-30 Finures Blowdown Nodalization Data Diagram 6.2-16 6.2-19 6.2-22 6.2-25 6.2-28 6.2-12 6.2-31 6.2-38 6.2-41 6.2-44 6.2-47 6.2-50 6.2-53 Nodal Nodal Pressure Pressures Differentials 6.2-39 6.2-42 6.2-45 6.2-48 6.2-51 6.2-54 6.2-40 6.2-43 6.2-46 6.2-49 N/A N/A
(')Model of complete (3600) annulus (2) Model of half (180') of annulus due to summary (3) The RCIC head spray line has been deleted and the associated high energy line breaks are no longer possible. However this failure and information is being provided as the bounding conditions that were established as part of the original plant design and licensing basis.
Revision 12 1 of 1 December 1999
(cu ft) 1 2
3 4
6+-.
13,250 7,149 6,312 1,164,879 TABLE 6.2-26 SUBCOMPARTMENT NODAL DESCRIPTION 6-IN RWCU LINE BREAK RWCU HEAT EXCHANGER ROOM Initial Conditions DBA Break Conditions
/
Break Temp.
Pressure Humidity
% Break Break ea (OF)
(psia)
(%)
in Vol.
Line Asq ft) 103 14.7 103 1i.:7
/
103 14.7 103 14.7 0
0 0
0 100 0.0 0.0 7
/
RWCU See Tble 62-28
/
/
Calculated")I Peak Pressure Break Difference type
( s'd ER
.3 Z
0.3 0.3 0.0 0.0
(Nodal peak pressure minus pressure in node 4 (Pi P4 )
August 1993 Revision 6 I
1 of 1
RBS USAR TABLE 6.2-26 SUBCOMPARTMENT NODAL DESCRIPTION 4-IN and 6-IN RWCU LINE BREAKS RWCU HEAT EXCHANGER ROOM Initial Conditions Volume Temp.
(cu ft)
(OF) 13,250 103 7,059 90 6,153 90 1,165,128 90 358,000 120 Pressure (psia) 14.7 14.7 14.7 14.7 14.7 DBA Break Conditions Humidity
% Break Break
(%)
in Vol.
Line 0
0 0
0 100 RWCU Break DER 0.0 0.0 0.0 0.0 100 Calculated Peak Pressure Difference (psid) 1.627 (4-in) 1.488 (6-in)
<0.5
- 0.5
- 0.5 N/A (see note 1)
Note: 1. The Volume No.
5 is included for conservatism.
This volume has no vent path connection with other volumes.
The steel containment is modeled as thermal conductors to connect this volume with other volumes except the break room, which has high temperatures after the break.
By assuming a high initial temperature for Volume No.
5, more heat is transferred into the other volumes, which generates more limiting pressure/temperature responses.
Revision 6 1 of 1 August 1 Volume No.
- -+6 1
2 3
4 5
6+--o 993
-.. COMPAR-TMENT-VENT. PATH-DESCRIPTI-Of..
6-IN RWCU LINE BREAK RWCU HEAT EXCHANGER ROOM Ven Pat No.
IA lc 2A 2c,
2C"
.2E 3
4 655 7
88 Description of Vent Path Flow (Choked/Unchoked)
From t
Vol.
h Node No.
1 2
1 2
.1 2
1 2
1 2
/ 1 S
2 S1 2
1 2
1 4
Unchoked 4
1 Unchoked 2
3 Unchoked 3
2 Unchoked 2
4 Unchoked 4
2 Unchoked 2
4 Unchoked 4
2 Unchoked 2
4 Unchoked 4
2 Unchoked 3
4 Unchoked 4
3 Unchoked Vent Area (sq ft) 1.628 1.628 15,56
- 5.56 11.024 11.024 To Vol.
Node No.
2 1
2 1
2 1
2 1
2 1
2 1
2 1
2 1
L/A (ftI) 0.168 0.168 0.168 0.168 0.168 0.168 0.724 0.724 0.724 0.724 0.724 0.724 0.724 0.724 0.724 0.724 Head Loss Coefficient Friction Turning Expansion Contraction 0.036 0.036 0.013 0.013 0.036 0.036 0,09 0.09
- 0.009 0.009 15.75 0.83j..
2.15 15.75 0.8;3 2.15 194.02
/92 194.02 0.092 148.6 0.233 0.027 148 6 0.233 0.027 1486 0.233 0.027 0.233 0.027 44.94 0.261 0.053
- 14.94 0.261 0.053 172.5 0.162 172.5 0.162 3.38 3.38 1.327 1.327 0.963 0.963 0.133 0.133 0.998 0.998 0.985 0.985 0.5 0.5 0.497 0.978 0.496 0.496 0.998 0.5 0.998 0.5 0.5 0.998 0.5 0.996 0.996 0.499 0.996 0.499/
0.998 0.5/
0.998 0
/
1.0 0.164
- 1. 0
/
- 0. 164 0.590 0.Y43 0/.933 0.0325 0.0042 0.483
/
0.933 0.0042 0.996 0.831 0.922 0.490 0.0325 0.483 0.458 0.499 Loss Due to Thick Edged Oriface 0.04 0.04 0.766 0.766 0.04 0.04 0.33 0.33 0.148 0.0382 0.38 0ý 382 August 1987 Unchoked Unchoked Unchoked Unchoked Unchoked Unchoked Unchoked Unchoked Unchoked Unchoked Unchoked
- Unchoked, Unchoked Unchoked Unchoked Unchoked 1.628 1.628 1.628 1.628 2.806 2.806 2.965 2.965 1.18 1.18.
it Total 1.573 1.573 3.877 4.358 2.260 2.260 1.573 1.573 1.827
- 2. 32S 1.463 1.959 1.585 1.585 1.507 1.507 3.780 3.780 0
90
.436 1.141 0.552 1.141 0.552 1.5 0.922 0.490 Y, q9 1 of 2
From Vent Vol.
Path Node No.
No.
9 3'
/
To,/
Description V)I.
of Vent Node Path Flow No.
(Choked/Unchoked) 4 Unchoked 3
Unchoked Vent Area (sq ft)
/
L/A (ftl)
Head Loss Coefficient Friction Turning Expansion 172.5 0.162 172.5 0.162 0.922 0.490
/2 Contract on 7
./
re
." )
Loss Due to Thick Edged Oriface Total /
0.922 0.490
4 jk'/
NOTES:
- 1.
Vent pa hs 1.
, 1b, and Ic are combine into one vent path
- 2.
Ven/taths 2 A, 2 a, 2 c, 2D and 2 E ar combined into one vent
/
(vent pa
- 1).
path ent path 2).
2 of 2 August 1987
RBS USAR TABLE 6.2-27 SUBCOMPARTMENT VENT PATH DESCRIPTION 4-IN and 6-IN RWCU LINE BREAK RWCU HEAT EXCHANGER ROOM vent I Vol.
Vol.
Vent Forward Reverse Choked /
Junct.
Hydraulic Inertia Path A
B Area Loss Loss unchoked Length Diameter Length No.
No.
No.
(ft 2 )
Coeff.
Coeff.
(ft)
(ft)
(ft) 1 1
2 28.210 3.131 2.902 Choked 2
3.719 11.000 2
1 2
28.210 4.918 4.651 Choked 2
3.719 11.000 3
1 4
23.333 11.708 8.196 Choked 13.292 4.516 39.792 4
2 3
192.260 0.630 0.397 Choked 0
8.670 23.875 5
2 4
162.828 1.706 1.706 Choked 9.014 8.041 39.431 6
2 4
162.828 1.706 1.706 Choked 9.014 8.041 39.431 7
2 4
14.708 2.670 1.550 Choked 1.750 0.655 57.000 8
3 4
166.678 1.000 0.500 Choked 0
11.800 38.917 9
3 4
166.678 1.000 0.500 Choked 0
11.800 38.917 Note:
(1) Vent paths #10 through #13 simulate the break junctions for the upstream and downstream blowdown for the 4-in and 6-in RWCU line breaks in the RWCU heat exchanger room.
August 1987 1 of 2
/
/
Time (sec)
)
0.0 0 0.000;L 0.02,1 S 0. 0Q,2 1.ill0
- 1.110 4.513 1.514
,, 1.888 1.889 5.997 5.998 9.442 9.443 16.657 K
16.658 23.595 25.157 6-IN RWCU LINE BREAK RWCU HEAT EXCHANGER ROOM Belowdown
/ Mass Flow Rate (lbm/sec) 0.0 873.6 873.6 1310.3 1310.3 1259.1 1259.1 771.2 771.2 385.6 385.6 841
%41.3 202.2 202.2 0.0 Blowdown Enthalpy (Btu/lbm) 416/ -/
416/*
/4'16 416 416 416 416 416 416 416 416 416 88 88 88 88 Blowdown 17 Energy Release Rate (Btu/sec) 0.0 363,418 363,418 545,085 545,085 523,786 523,786 320,820 /
320,820
- 160, 0
160,/410 3/49,981
/*49, 981 74,035 74, 035 17,794 17,794 August 1987 Total Effective Break Area (sa ft) 0.0 0/181
.081.
0.2715 0.2715 0.2609
//
0.2609 0.1598 0.1598 0.0799 0.0799 0.0799 0 0799 0.0799 0.0794ý 0.0192 0.0192 0.0
__j r Y (c, ("i C." -f-C, C&
Fýý 1 of 1
RBS USAR TABLE 6.2-28 BLOWDOWN DATA 6-IN RWCU LINE BREAK RWCU HEAT EXCHANGER ROOM Time After Break Mass Flow Rate Revised h (sec)
(lbm/sec)
(Btu/lbm)
Upstream Blowdown 0.0000 0.0 419.00 0.0001 892.5 419.00 0.9446 892.5 419.00 0.9447 394.0 419.00 3.2271 394.0 419.00 3.2272 394.0 419.00 5.6170 394.0 419.00 5.6171 1129.4 419.00 9.6189 1129.4 419.00 9.6190 1129.4 93.91 14.9907 1129.4 93.91 14.9908 1129.4 93.91 17.7895 1129.4 93.91 17.7896 212.1 93.91 31.2276 212.1 93.91 32.7896 0.0 93.91 Downstream Blowdown 0.0000 0.0 419.00 0.0001 446.2 419.00 2.0263 446.2 419.00 2.0264 394.0 419.00 2.7902 394.0 419.00 1 of 1 August 1987
RBS USAR TABLE 6.2-29 SUBCOMPARTMENT NODAL DESCRIPTION 8-IN RWCU LINE BREAK RWCU FILTER/DEMINERALIZER ROOM Volume Volume No.
(cu ft)
Temp.
(OF) 105 2,165.6
/i 2,165.6 105 8,278.9 105 1, 120,000(2) 105
/'
Initial Conditions P essure (psia) 14.7
/
14.7 14.7 14.7 Humidity
% Break
(%)
in ol.
0 100
/
0 0
.7 0
0 0
DBA Break Conditions Break Break Area Line (sq ft)
RWCU Bre (See Table 6.2.-.
0.
Calculated(')
Peak Pressure ak Difference e
(psid) 21.18 0/.
0.0 0
rep("APvýr4 1t tA4df-1 )Nodal peak pressure minus pressure in Node 4 (Pi-P 4) t 2)Assumed value to maximize pressure differential across RWCU filter/demineralizer room.
1 of 1 August 1993 Revision 6 1
3 4
6<--
f
RBS USAR TABLE 6.2-29 SUBCOMPARTMENT NODAL DESCRIPTION 8-IN RWCU LINE BREAK RWCU FILTER/DEMINERALIZER ROOM Volume Volume No.
cu ft 2,163.2 2,163.2 8,085.0 1,120,000(2)
Initial Conditions Temp.
Pressure f__
_tsiaj 105 105 100 90 14.7 14.7 14.7 14.7 Humidity 0M) 0 DBA Break Conditions Break
% Break Break in Vol.
Line 100 0
0 0
RWCU 0
0 0
(" Maximum differential pressure across the RWCY Filter / Demineralizer rgom walls.
(2 ) Assumed value to maximize pressure differential across RWCU filter/demineralizer room.
August 1993 Revi'zion 6 Io 1
2 3
4 6<--e Break Type DER Calculated" )
Peak Pressure Difference (psid) 10.425 0.0 0.0 0.0 I of]
RBS USAR TABLE 6.2-30 SUBCOMPARTMENT VENT PATH DESCRIPTION 8-IN RWCU LINE BREAK RWCU FILTER/DEMINERALIZER ROOM I,
From Tp' Vent Vol.
yol.
KPath Node
/Node No.
No.
No.
(Chc
)escription of Vent Path Flow oked/Unchoked)
Vent Area (sq ft)
L/A (1I)
Head Loss Coefficient Friction Thick Edge Turning Grating Expansion Contraction Total 0.996 0.989 0.986 0.995 0.994/
/
/
'0.999
/
0.991 0.494 0.776 0.494 0.911 Unchoked Unchoked Unchoked Unchoked Unchoked Unchoked Unchok Unc ked choked Unchoked
")Includes losses due to grating and thick edged orifice.
August 1987 2
/
2
/
4 4
5 5
1 3
3 2
3 4
3 4
3 4
3 1
2 3
4 3
4 3
4 3
.q k
1.37 1.37.
lj. 11
/'1.811 2.6 2.6 1.6 1.6 31.5 31.5 2.579 2.579 1.957 1.957 0.785 0.785 2.338 2.338 0.871
- 0. 871
-/
/
0.497 0.1499 499
//
0.496 0.498 0.498 0.498 0.500 0.477 0.440 0.687(1) 0.694 (1) 0.163 0.170 1.493 1.488 1.485 1.491 1.492 1.491
/I1.491 2.890 2.995 0.95 0.95 1 of 1
_/
RBS USAR TABLE 6.2-30 SUBCOMPARTMENT VENT PATH DESCRIPTION 8-IN RWCU LINE BREAK RWCU FILTER/DEMINERALIZER ROOM Vent paths #5 and #9 simulate the break junctions for the upstream and downstream blowdown for the 8-in RWCU line break in the Filter /
Demineralizer room.
Aut~ust 1987 Vent(') Vol.
Vol.
Vent Forward Reverse Choked Junct.
Hyd.
Inerti Path A
B Area Loss Loss
/
Length D.
a No.
No.
No.
(ft 2)
Coeff.
Coeff.
unch-(ft)
(ft)
Length oked (ft) 1 1
3 0.25 1.953 1.927 Choked 3.5 0.5 13.665 2
2 3
0.25 1.953 1.927 Choked 3.5 0.5 13.665 3
3 4
0.25 1.500 1.500 Choked 2
0.167 14.125 4
3 4
31.5 4.742 3.642 Choked 7
4.667 31.917 6
1 2
0.167 2.000 1.500 Choked 5.25 0.4 21.167 7
1 2
0.167 1.500 1.500 Choked 4
0.167 19.917 8
2 3
0.25 2.954 2.954 Choked 16.25 0.5 43.125 Note:
(1)
I oflI
RBS USAR TABLE 6.2-31 BLOWDOWN-DATA 8-IN RWCU LINE BREAK RWCU FILTER/DEMINERALIZER ROOM Blowdown Blowdown Total Mass Blowdown Energy Effective Time Flow Rate Enthalpy Release Rate Break Area (sec)
(ibm/sec)
(Btu/lbm)
(Btu/sec)
(sa ft) 0.0 0.00.0 0.0 0.0 0.0001 2378.4 88 2.09,x 10 0.3 0.0006 "2378.4 88 2.0,93 x 10 0.3 0.0007 3567.6 88 3
4 x 105 0.4 0.0097 3567.6 88
.14 X l0' 0.4 0.0098 4756.8 88
/4.186 x ls 0.6
- 0. 1305.,
4756.8 88 4.186 x l05 0.6 0.1306' 2378.4 88 2.093 x 105 0.3 1.3925 2378.4 88 2.093 x 105 0.3 1.3926 592.1 88 5.21 x 104 0.0 9.9675 592.1 88 5.21 x 104 0.0 9.9676 592.1 196.8 1.165 x 105 0.0 15.2075 592.1 196.$
1.165 x 105
- 0.
15.2076 592.1 303,..'5 1.797 x 10S 5
.0 18.6275 592.1 303.5 1.797 x 10' 0.0 18.6276 592.1 Y89.2-2.304 x 105 0.0 18.8375 592.1
/389.2 2.304 x 10 0.0 18.8376 592.1 7 472.0 2.795 x 1 H 19.7275 592.1 472.0 2.795 x10 0.0 19.7276 447.4 453.6 2.03 y/105 0.0 22.8975 447.4 453.6 2.03//x I0' 0.0 22.8976 151.4 529.2 8.01' X 104 0.0 26.6655 151.4 529.2 8.012 X 10 4 0.0 28.2275 0.0 0.0 0.0 60.0 0.0 0.0 0.0 016 016 524 524 032 032 016 016 7509 7
7 7
7 7
7 7
7 5
5 1
1 5I
'509
'509
'509
'509
'509
'509
'509 675/
567
.9
.92/
Sote: /ata based on assume
/ 7,884 Ibm/sec-ft f critical flow of %aturated liquid at 1,000 ps.a a f
August 1987
/
RBS USAR TABLE 6.2-31 BLOWDOWN DATA 8-IN RWCU LINE BREAK RWCU FILTER/DEMINERALIZER ROOM Time After Break (sec) 0.0000 0.0001 0.0007 0.0008 1.0048 1.0049 6.0424 6.0425 7.0940 7.0941
- 8. 9119 8.9120 9.7007 9.7008 11.6003 11.6004 13.7403 13.7404 13.8435 13.8436 16.1949 16.1950 17.9879 17.9880 18.5536 18.5537 22.9688 22.9689 29.2812 29.9296 32.9880 Blowdown Mass Flow Rate (lbm/sec)
Upstream Blowdown 0
2241.04 2241.04 4482.09 4482.09 1061.53 1061.53 721.681 721.681 721.681 721.681 721.681 721.681 463.649 463.649 418.496 418.496 372.771 372.771 372.771 372.771 372.771 372.771 266.015 266.015 266.015 266.015 266.015 266.015 75.0972 0
August 1987 Enthalpy (Btu/Ibm) 93.9106 93.9106 93.9106 93.9106 93.9106 93.9106 93.9106 93.9106 93.9106 135.29 135.29 149.192 149.192 211.803 211.803 263.453 263.453 305.284 305.284 359.662 359.662 388.65 388.65 331.346 331.346 409.861 409.861 450.744 450.744 531.441 531.441 1 of 2
RBS USAR TABLE 6.2-31 BLOWDOWN DATA 8-IN RWCU LINE BREAK RWCU FILTER/DEMINERALIZER ROOM Time After Blowdown Mass Enthalpy Break Flow Rate (Btu/lbm)
(sec)
(ibm/sec)
Downstream Blowdown 0
0 93.91 0.0001 806.11 93.91 0.0098 806.11 93.91 0.0099 1612.2 93.91 0.1999 1612.2 93.91 0.2000 0
93.91 August 1987 2 of 2
FIGURE 6.2-50 NODALIZATION DIAGRAM 6-IN RWCU LINE BREAK RWCU HEAT EXCHANGER ROOM RIVER BEND STATION UPDATED SAFETY ANALYSIS REPORT a
rq(,tu-,j
EL-256'3" EL-186'3" J5 J1 ANNULAR RWCU SPACE BALANCE OF HX OUTSIDE CONTAINMENT ROOM OF HX FREE VOLUME (NODE 1)
J2 ROOM (NODE 4)
(NODE 2)
EL-147'3" EL-144'3" J4 EL-1 44'3" BELOW HX ROOM (NODE 3)
J' EL-1 37'0" EL-90'0" EL-266'3" Node 5 models the Shield Building Annulus, which only connects to other nodes with thermal conductors.
EL-70'0" FIGURE 6.2-50 NODALIZATION DIAGRAM 4-IN AND 6-IN RWCU LINE BREAKS RWCU HEAT EXCHANGE ROOM RIVER BEND STATION UPDATED SAFETY ANALYSIS REPORT
2.0 2.5 3.0 TIME AFTER BREAK (SEC)
FIGURE 6.2-51 NODAL PRESSURES 6-IN RWCU LINE BREAK RWCU HEAT EXCHANGER ROOM RIVER BEND STATION UPDATED SAFETY ANALYSIS REPORT 0.0 0.5 1.0 1.5 III NA-1 C(Afftte-
16.5
- 16.
u.o 15
.5 (II CL UJ 15 14.5 14 1
0.0 0.5 1.0 1.5 2.0 2.5 TIME AFTER BREAK (SEC)
FIGURE 6.2-51 NODAL PRESSURE 6-IN RWCU LINE BREAK RWCU HEAT EXCHANGER ROOM RIVER BEND STATION UPDATED SAFETY ANALYSIS REPORT
.P1-P4 2.0 ddeiA
'2e4'
0., f-44te
- 2 -zL 2.5 a "RE 6.2-52 NODAL PRESSURE DIFFERENTIAL 6-IN RWCU LINE BREAK RWCU HEAT EXCHANGER ROOM RIVER BEND STATION UPDATED SAFETY ANALYSIS REPORT
EL-256'3" ytplc~vj
ýy P4, f'),e FIGURE 6.2-53 NODALIZATION DIAGRAM 8-IN RWCU LINE BREAK RWCU FILTER/ DEMINERALIZATION ROOM RIVER BEND STATION UPDATED SAFETY ANALYSIS REPORT
EL-186'3" I J6 IEMINERALIZER J7 DIEMINEICALIZIL ROOM SOUTH ROOM NORTH EL-162'3" EL-162'3" NODE 4 JI J2 J8 THE REST OF THE CONTAINMENT EL-I186'3"
-J NODE 3 HOLDING PUMP ROOMJ4 EL-90' EL-162'3" FIGURE 6.2-53 NODALIZATION DIAGRAM 8-IN RWCU LINE BREAK RWCU FILTER / DEMINERALIZER ROOM RIVER BEND STATION UPDATED SAFETY ANALYSIS REPORT NODE I RWCU FILTER /
NODE 2 RWCU FILTER I EL-186'3" EL-256'3"
FIGURE 6.2-54 NODAL PRESSURES 8-IN RWCU LINE BREAK RWCU FILTER/ DEMINERALIZATION ROOM RIVER BEND STATION UPDATED SAFETY ANALYSIS REPORT C,
I r4
+.A I
-14 e-ita 7ýý '-C -
30 25 P1 S20
/.P2 15 P3, P4 10 0
20 40 60 80 100 TIME AFTER BREAK (SEC)
FIGURE 6.2-54 NODAL PRESSURES 8-IN RWCU LINE BREAK RWCU FILTER / DEMINERALIZER ROOM RIVER BEND STATION UPDATED SAFETY ANALYSIS REPORT
40 50 60 y REAK EC) j*-*
pve*#eL ':* T~
- 6. -
2 2' FIGURE 6.2-55 NODAL PRESSURE DIFFERENTIAL 8-IN RWCU LINE BREAK RWCU FILTER/DEMINERALIZER ROOM RIVER BEND STATION UPDATED SAFETY ANALYSIS REPORT
(--,r-de fz'
USAR Appendix 3B
RBS USAR APPENDIX 3B PRESSURE ANALYSIS FOR SUBCOMPARTMENTS OUTSIDE CONTAINMENT TABLE OF CONTENTS Section Title Page 3B.1 DESIGN BASES 3B-l 3B.2 DESIGN FEATURES 3B-2 3B.3 DESIGN EVALUATION 3B-3 August 1987 3B-i
RBS USAR q-> 1 APPENDIX 3B LIST OF TABLES Title HIGH-ENERGY LINE BREAKS AUXILIARY BUILDING, Deleted: -
20-NODE MODEL Table Number 3B-1 3B-2 3B-3 3B-4 3B-5 3B-6 3B-7 3B-8 3B-9 3B-10 3B-11 3B-12 3B-13 NODE MODEL Deleted: SUBCOMPARTMENT VENT PATH DESCRIPTION
¶ AUXILIARY BUILIDNG -
20 NODE MODEL Deleted: -NODE 10 Deleted: -NODE 6¶ Deleted: -NODE 2 Deleted: -NODE 12 3B-ii August 1988 HIGH-ENERGY LINE BREAKS MAIN STEAM TUNNEL -
6-NODE MODEL SUBCOMPARTMENT NODAL DESCRIPTION AUXILIARY BUILDING,
SUBCOMPARTMENT NODAL DESCRIPTION MAIN STEAM TUNNEL -
6-NODE MODEL DELETED SUBCOMPARTMENT VENT PATH DESCRIPTION MAIN STEAM TUNNEL -
6-NODE MODEL MASS AND ENERGY RELEASE 3-IN. RWCU DER IN AUXILIARY BUILDING MASS AND ENERGY RELEASE 6-IN.
RWCU DER IN AUXILIARY BUILDING MASS AND ENERGY RELEASE 4-IN. RCIC DER IN AUXILIARY BUILDING, MASS AND ENERGY RELEASE 8-IN.
RHR DER IN AUXILIARY BUILDING MASS AND ENERGY RELEASE 24-IN. MAIN STEAM LINE DER IN STEAM TUNNEL-NODE 2 MASS AND ENERGY RELEASE 24-IN. MAIN STEAM LINE SER IN STEAM TUNNEL-NODE 1 MASS AND ENERGY RELEASE 8-IN.
RCIC STEAM LINE DER IN STEAM TUNNEL-NODE 2 1+--
RBS USAR APPENDIX 3B LIST OF TABLES (Cont)
MASS AND ENERGY RELEASE 8-IN.
RCIC STEAM LINE SER IN STEAM TUNNEL-NODE 1 PELETED HEAT SINK SLAB DESCRIPTION MAIN STEAM TUNNEL -
6-NODE MODEL Deleted: HEAT SINK SLAB DESCRIPTION
-¶
.AUXILIARY BUILDING -
20 NODE MODEL I+-.
3B-iii August 1988 3B-14 3B-15 3B-16
RBS USAR 0-41 APPENDIX 3B LIST OF FIGURES Figure Number 3B-1 THROUGH 3B-2!A Title DELETED Deleted: 3-1. NODALIZATION DIAGRAM -
AUXILIARY BUILDING¶
.20 NODE MODELI
¶ 3B-2. PRESSURE TRANSIENTS IN NODE I1
.AUXILIARY BUILDING - HIGH ENERGY LINE¶
.BREAK ANALYSIS (3" & 6" RWCU AND 4" RCIC)¶
¶ 3B-2A. PRESSURE TRANSIENTS IN NODE I1
.AUXILIARY BUILDING - HIGH ENERGY LINE¶
-BREAK ANALYSIS (8" RHR)¶ I
3B-3. PRESSURE TRANSIENTS IN NODE 21
-AUXILIARY BUILDING - HIGH ENERGY LINEI
.BREAK ANALYSIS (3"
¶ 3B-3A. PRESSURE TRANSIENTS IN NODE 2¶
.AUXILIARY BUILDING -
HIGH ENERGY LINEI
.BREAK ANALYSIS (8" RHR)¶
¶ 3B-4. PRESSURE TRANSIENTS IN NODE 31
.AUXILIARY BUILDING -
HIGH ENERGY LINEI
.BREAK ANALYSIS (3" & 6" RWCU AND 41" RCIC)¶ 3B-4A. PRESSURE TRANSIENTS IN NODE 3¶
.AUXILIARY BUILDING - HIGH ENERGY LINEI
.BREAK ANALYSIS (8" RHR)¶
¶ 3B-5. PRESSURE TRANSIENTS IN NODE 41
.AUXILIARY BUILDING
- HIGH ENERGY LINEI
.BREAK ANALYSIS (3" & 6" RWCU AND 4" RCIC)¶
¶ 3B-5A. PRESSURE TRANSIENTS IN NODE 41
.AUXILIARY BUILDING - HIGH ENERGY LINEI
-BREAK ANALYSIS (8" RHR)¶
¶ 3B-6. PRESSURE TRANSIENTS IN NODE 5¶
.AUXILIARY BUILDING -
HIGH ENERGY LINE¶
.BREAK ANALYSIS (3"
& 6" RWCU AND 4" RCIC) 1+-o 3B-iv August 1988
RBS USAR APPENDIX 3B LIST OF FIGURES (Cont)
Deleted: 3B-6A. PRESSURE TRANSIENTS IN NODE 51
.AUXILIARY BUILDING
- HIGH ENERGY LINE¶ BREAK ANALYSIS (8"
RHR)¶
¶ 3B-7.PRESSURE TRANSIENTS IN NODE 69
.AUXILIARY BUILDING - HIGH ENERGY LINE¶ BREAK ANALYSIS (3"
¶ 3B-7A.PRESSURE TRANSIENTS IN NODE G¶
.AUXILIARY BUILDING -
HIGH ENERGY LINE¶
.BREAK ANALYSIS (8"
RHR)¶
¶ 3B-8 PRESSURE TRANSIENTS IN NODE 79
.AUXILIARY BUILDING
- HIGH ENERGY LINE¶ BREAK ANALYSIS (3"
¶ 3B-8A. PRESSURE TRANSIENTS IN NODE 7¶
.AUXILIARY BUILDING
- HIGH ENERGY LINEI
.BREAK ANALYSIS (8"
RHR)¶ 9
3B-9 PRESSURE TRANSIENTS IN NODE 81
.AUXILIARY BUILDING HIGH ENERGY LINE¶ BREAK ANALYSIS (3"
¶ 3B-A. PRESSURE TRANSIENTS IN NODE 81
.AUXILIARY BUILDING
- HIGH ENERGY LINEI
.BREAK ANALYSIS (8" RHR)¶ 91 3B-10. PRESSURE TRANSIENTS IN NODE 91
.AUXILIARY BUILDING HIGH ENERGY LINE¶
.BREAK ANALYSIS (3"
¶ 3B-10A PRESSURE TRANSIENTS IN NODE 91 AUXILIARY BUILDING
- HIGH ENERGY LINE¶ BREAK ANALYSIS (8"
RHR)¶ 3B-11. PRESSURE TRANSIENTS IN NODE i0¶ AUXILIARY BUILDING
- HIGH ENERGY LINEI BREAK ANALYSIS (3"
3B-1iA PRESSURE TRANSIENTS IN NODE 101 AUXILIARY BUILDING -
HIGH ENERGY LINE¶ BREAK ANALYSIS (8"
RHR)
¶ l+--
3B-v August 1988
RBS USAR APPENDIX 3B LIST OF FIGURES (Cont)
APPENDIX 3B LIST OF FIGURES (Cont)
NODALIZATION DIAGRAM -
MAIN STEAM TUNNEL -
6 NODE MODEL PRESSURE TRANSIENTS IN NODE 1 MAIN STEAM TUNNEL HIGH ENERGY LINE BREAK ANALYSIS I ' 3B-22 1<--o 3B-vi August 1988 Deleted: 3B-12. PRESSURE TRANSIENTS IN NODE ii¶
.AUXILIARY BUILDING
- HIGH ENERGY LINEI
-BREAK ANALYSIS (3"
¶ 3B-12A. PRESSURE TRANSIENTS IN NODE II¶
.AUXILIARY BUILDING -
HIGH ENERGY LINE¶
.BREAK ANALYSIS (8"
RHR)¶
¶ 3B-13.PRESSURE TRANSIENTS IN NODE 12¶
.AUXILIARY BUILDING
- HIGH ENERGY LINEI
.BREAK ANALYSIS (3"
¶ 3B-13A. PRESSURE TRANSIENTS IN NODE 12¶
.AUXILIARY BUILDING
- HIGH ENERGY LINE¶
-BREAK ANALYSIS (8"
RHR)(¶ I
3B-14. PRESSURE TRANSIENTS IN NODE 13¶
-AUXILIARY BUILDING HIGH ENERGY LINE¶
.BREAK ANALYSIS (3"
¶ 3B-14A. PRESSURE TRANSIENTS IN NODE 13¶
.AUXILIARY BUILDING
- HIGH ENERGY LINE¶I 111ffl Deleted: 3B-17A. PRESSURE TRANSIENTS IN NODE 16¶
.AUXILIARY BUILDING
- HIGH ENERGY LINE¶
.BREAK ANALYSIS (8"
RHR)¶
¶ 3B-18 PRESSURE TRANSIENTS IN NODE 17¶
.AUXILIARY BUILDING HIGH ENERGY LINEI
.BREAK ANALYSIS (3"
¶ 3B-18A.PRESSURE TRANSIENTS IN NODE 17¶
.AUXILIARY BUILDING -
HIGH ENERGY LINE¶ BREAK ANALYSIS (8" RHR)¶
- HIGH ENERGY LINE¶ BREAK ANALYSIS (3"
3B-19A. PRESSURE TRANSIENTS IN NODE 18¶ AUXILIARY BUILDING -
HIGH ENERGY LINEI BREAK ANALYSIS (8"_HR)¶
¶ 3B-20. PRESSURE TRANSIENTS IN NODE 19¶ AUXILIARY BUILDING HIGH ENERGY LINE¶$yj 3B-23
RBS USAR APPENDIX 3B LIST OF FIGURES (Cont) 3B-24 PRESSURE TRANSIENTS IN NODE 1 MAIN STEAM TUNNEL HIGH ENERGY LINE BREAK ANALYSIS 3B-25 PRESSURE TRANSIENTS IN NODE 2 MAIN STEAM TUNNEL HIGH ENERGY LINE BREAK ANALYSIS 3B-26 PRESSURE TRANSIENTS IN NODE 2 MAIN STEAM TUNNEL HIGH ENERGY LINE BREAK ANALYSIS 3B-27 PRESSURE TRANSIENTS FOR EDC ZONE AB-070-3 3B-28 PRESSURE TRANSIENTS FOR EDC ZONE AB-095-3 3B-29 PRESSURE TRANSIENTS FOR EDC ZONE AB-095-4 3B-30 PRESSURE TRANSIENTS FOR EDC ZONE AB-114-8A & 8B I<---
August 1988 3B-viii
RBS USAR APPENDIX 3B PRESSURE ANALYSIS FOR SUBCOMPARTMENTS OUTSIDE CONTAINMENT 3B.1 DESIGN BASES Pressure response analyses were performed for the structural design basis of the main steam tunnel and other subcompartments in the auxiliary building for postulated ruptures of high-energy piping.
The definitions for high energy and criteria for protection against dynamic effects associated with postulated rupture of piping are given in Section 3.6A.
The analyses were performed using SWEC computer code THREED (Appendix 6B) for the main steam tunnel and the GOTHIC (Generation of Thermal-Hydraulic Information for Containments) code (developed by NAI) for the Auxiliary Building.
The auxiliary building was divided into a
large number of iDee:20 separate subcompartments for the purpose of analysis.
The main steam tunnel was divided into four separate subcompartments for its design evaluation.
A fifth node was used to represent the turbine
- building, and a
sixth node represents the outside atmosphere.
The subcompartment boundaries were chosen to represent physical restrictions to flow and to reflect additional detail in the vicinity of the high-energy lines.
Breaks were postulated in each auxiliary building volume containing a high-energy line.
Breaks were postulated in the main team tunnel on both sides of the jet impingement shield wall which bounds the break exclusion zone.
All breaks were considered to be instantaneous circumferential double-ended ruptures (DER),
i.e.,
the break area was equal to twice the effective cross-sectional flow area of the pipe, except that single-ended ruptures (SER) were considered in the main team tunnel break exclusion zone.
Section 3.6A defines the complete set of break locations in high-energy piping outside containment from which the design basis breaks for subcompartment pressurization were selected.
During isolation valve closure, the flow area used for mass and energy release calculations was assumed to be constant until the valve area equaled the flow limiting area.
Subsequently, the limiting flow area was linearly reduced to zero.
Auxiliary building high-energy lines were identified in the reactor water cleanup (RWCU)
- system, the reactor core isolation cooling (RCIC) system, and the residual heat 3B-1 August 1988
RBS USAR removal (RHR) system.
A total of four break locations were postulated and analyzed.
Peak calculated pressure differentials were generated for,ýIi four postulated breaks.
Table 3B-1 lists all postulated breaks,.
The accident prififes were generated to bound the most limiting pressure responses.
The main steam tunnel analysis considered feedwater,
- RCIC, and main steam line breaks.
Main steam line break analyses were performed assuming a two-phase blowdown.
Four combinations of break locations and blowdown conditions were postulated and analyzed.
Peak differential pressure values were generated by the two-phase blowdown breaks.
Table 3B-2 lists the postulated line breaks and identifies the two breaks that determined the design differential pressures for the steam tunnel.
3B.2 DESIGN FEATURES Fig. 1.2-13 through 1.2-19 show the piping and equipment in the subcompartments.
Fig. 1.2-18 shows the louver arrangement in the main steam tunnel chimney area.
There are six louvered panels, three on the east side and three on the west side of the chimney (el 170'-0").
These louvers open at a differential pressure of 3.25 psi, with an opening time of 0.3 sec.,
All high-energy piping with a
potential for producing high pressure and/or temperature environmental conditions in the auxiliary building is routed from the primary containment through the main steam tunnel.
The RWCU pump rooms and RCIC turbine pump room are located directly below the steam tunnel, thus minimizing the length of high-energy piping outside the tunnel.
Fast closing, motor-operated isolation valves are located inside and outside containment on each high-energy line except feedwater lines, which utilize check valves to isolate reverse flow from the reactor to postulated pipe breaks outside containment.
The outboard isolation valves are located in the steam tunnel break exclusion zone.
The isolation valves are automatically closed by signals from the leak detection system, e.g.,
high local area temperature.
To avoid inadvertent isolation signals, time delay relays have been installed in the isolation logics and an additional 5-second time delay has been assumed for the RCIC
/
RWCU line breaks.
Isolation of pipe breaks is also initiated by system high flow and other signals as described in Section 6.2.4.
Pressure tight doors designed to withstand a 'differential pressure of 3.0 psi are utilized to isolate ECCS equipment cubicles from the effects of high-energy line breaks.
These doors are administratively controlled closed.
I+-e Deleted: for the 20 subcompartments Deleted: by two of the Deleted: and identifies the two breaks that determined the design differential pressures 3B-2 August 1988
RBS USAR Two fire doors, A95/8 and A95/9, are maintained open for pressure relief purposes by fusible links which allow the doors to close at temperatures of 2250 F or more.
The pressure analysis assumed these doors to be only 50-percent
- open, and the maximum temperature in this area after the worst-case high-energy line break is less than 2250 F.
3B.3 DESIGN EVALUATION Subcompartment nodalization schemes were selected differential pressures across node boundaries.
components were selected as node boundaries.
The pressure transients across node boundaries are used the structural adequacy and component support design.
to maximize Structural differential to determine Table 3B-3 provides the nodal descriptions and gives the peak calculated and design differential pressures within the auxiliary building.
Table 3B-4 similarly shows the subcompartment nodal descriptions for the main steam tunnel and identifies the calculated and design peak differential pressures.
Figure 3B-22 shows the nodalization scheme for the main steam tunnel.
Table 3B-6 presents the vent path description corresponding to that shown on Fig. 3B-22 for the main steam tunnel.
In calculating the pressure differentials across the auxiliary building subcompartment walls, it is possible to take credit for the pressurization of the volume on the opposite side of the wall in question.
This procedure,
- however, leads to slightly different pressure differentials for all walls of the subcompartment in question.
To minimize the number of differential pressures to be considered and for conservatism, a
single differential pressure was calculated for each volume by subtracting 14.7 psia from each of the calculated nodal absolute pressures.
Peak pressure values for the main steam tunnel subcompartments also were calculated by subtracting 14.7 psia from the peak pressure values.
,Tables 3B-7 through 3B-10 provide the mass and energy release data for the breaks that determine the design differential pressures within the auxiliary building.
In
- general,
ý4oody"i) or Henry-Fauske"2 I flow was assumed (for saturated and subcooled
- flows, respectively) at the limiting downstream and upstream flow areas crediting friction.
During the inventory period, the mass and energy release data were calculated using the methodology of NEDO-20533"3 '
except that the Henry-Fauske model was used to calculate subcooled flow.
Deleted: ¶ Fig. 3B-1 shows the nodalization scheme used in the auxiliary building analysis and identifies the node numbers referred to in the remainder of this section.
Fig. 3B-22 similarly shows the nodalization scheme for the main steam tunnel.¶ Deleted: Table 3B-5 gives vent flow path data for the auxiliary building corresponding to the nodalization scheme shown on Fig. 3B-1.
Table 3B-6 presents the vent path description corresponding to that shown on Fig. 3B-22 for the main steam tunnel.¶ Deleted: frictionless Deleted: For the 4 -in RCIC line break, partial credit was taken for the effect of friction on reducing the "rate of blowdown.
,Considering only the 4-in diameter portion of the RCIC steam supply line, -the total loss coefficient for the fittings and straight pipe
'was determined to be K=5.
In this case, frictional Moody flow141 with fL/D=5 is assumed and yields the blowdown time history given in Table 3B-9.1
¶ For the 8-in RHR line break, credit was also taken for friction.
Considering
'piping from the main steam line to the break and choked flow at the break, the total loss coefficient was calculated to be K = 5.41.
Therefore, frictional Moody flow"* with fL/D = 5.41 is used and the blowdown time history is given in Table 3B-1O.¶ 3B-3 August 1987
3B-5 August 1988 The mass and energy release data used for the postulated main steam tunnel pipe breaks are presented in Tables 3B-11 through 3B-14.
These blowdowns were based entirely on frictionless Moody flow with a constant reservoir pressure.
The blowdown was considered to be all steam for the first second after the accident.
After 1 sec, the two-phase froth level rising in the vessel was assumed to discharge through the main steam lines.
The quality of this part of the blowdown was assumed to be 7 percent.
The exposed surfaces of concrete and steel in each auxiliary building node were modeled as heat sinks in the analysis.
The 2 ft thick concrete walls, ceiling, and floors were assumed to be only 1-ft
- thick, absorbing heat from the transient thermal environment in the respective node and insulated on the other side.
The steel heat sinks include the beams, columns, posts,
- stairs, and platforms in the respective node.
An equivalent steel slab was derived by dividing the total steel volume by the total exposed steel surface area.
Concrete and steel heat sinks were modeled similarly in the steam tunnel 6-node model, except that the concrete slabs were assumed to be 1-ft thick, based on actual slabs which are 4-ft thick.
Table 3B-16 summarizes these heat slabs.
The initial conditions in each node were assumed to be the maximum normal temperature, 14.7-psia
- pressure, and maximum relative humidity based on the Environmental Design Criteria (EDC).
,Fig. 3B-23 through 3B-26 provide the absolute pressure transient plots for the two main steam tunnel subcompartments within the auxiliary building portion of the tunnel.
,Fig. 3B-27 through 3B-30 provide the HELB pressure transients for the most limiting sub-compartments (typically the break rooms) in the Auxiliary Building_,
Deleted: The UCHIDA heat transfer coefficient was applied, and condensate revaporization was assumed to be limited to 8 percent.
The heat sink slabs for the auxiliary building 20-node model are defined in Table 3B-15.¶ S--
Break (Connuous)-.---.---
Deleted* 100-percent Deleted: I Fig. 33-2 through 3B-21A
.provide the absolute pressure transient plots for the 20 subcompartments in the auxiliary building.¶ Deleted: I Deleted: I 3B-6 August 1988
- +-I
- <--1
References -
3B.4
- 1.
- Moody, F.
J.
Maximum Flow Rate of a Single Component Two Phase Mixture, Journal of Heat Transfer, Trans.
- ASME, 87, February 1965, p 134-142.
- 2.
- Henry, R.
E.
and Fauske, H. K.
The Two-Phase Critical Flow of One Component Mixtures in Nozzles, Orifices, and Short Tubes, Journal of Heat Transfer, Trans.
- ASME, 93, May 1971, p 179-187.
- 3.
NEDO-20533, Mark III Containment System Analytical Model, Appendix B, Pipe Inventory Blowdown, June 1974.
0-+l
- 4.
- Lahey, R.
T.
and Moody, F.
J.
The Thermal-Hydraulics of a Boiling Water Nuclear Reactor,
- ANS, 1977.
1+-0 August 1988
- +-I 3B-5 August 1988
- +-I 3B-6
Page 6: [1] Deleted Unknown 3B-12 PRESSURE TRANSIENTS IN NODE 11 AUXILIARY BUILDING -
HIGH ENERGY BREAK ANALYSIS (3"
PRESSURE TRANSIENTS IN NODE 11 AUXILIARY BUILDING -
HIGH ENERGY LINE BREAK ANALYSIS (8"
RHR)
PRESSURE TRANSIENTS IN NODE 12 AUXILIARY BUILDING -
HIGH ENERGY LINE BREAK ANALYSIS (3"
PRESSURE TRANSIENTS IN NODE 12 AUXILIARY BUILDING -
HIGH ENERGY LINE BREAK ANALYSIS (8"
RHR)
PRESSURE TRANSIENTS IN NODE 13 AUXILIARY BUILDING HIGH ENERGY BREAK ANALYSIS (3"
& 6" RWCU AND LINE 4"
RCIC)
PRESSURE TRANSIENTS IN NODE 13 AUXILIARY BUILDING -
HIGH ENERGY LINE BREAK ANALYSIS (8"
RHR)
PRESSURE TRANSIENTS IN NODE 14 AUXILIARY BUILDING HIGH ENERGY BREAK ANALYSIS (3"
PRESSURE TRANSIENTS IN NODE 14 AUXILIARY BUILDING -
HIGH ENERGY LINE BREAK ANALYSIS (8"
RHR)
PRESSURE TRANSIENTS IN NODE 15 AUXILIARY BUILDING
-HIGH ENERGY BREAK ANALYSIS (3"
& 6" RWCU AND LINE 4"
RCIC)
PRESSURE TRANSIENTS IN NODE 15 AUXILIARY BUILDING -
HIGH ENERGY LINE BREAK ANALYSIS (8"
RHR)
PRESSURE TRANSIENTS IN NODE 16 AUXILIARY BUILDING -
HIGH ENERGY BREAK ANALYSIS (3"
& 6" RWCU AND LINE 4"
RCIC)
Page 6: [2] Deleted 3B-17A 3B-18 3B-18A Unknown PRESSURE TRANSIENTS IN NODE 16 AUXILIARY BUILDING -
HIGH ENERGY LINE BREAK ANALYSIS (8"
RHR)
PRESSURE TRANSIENTS IN NODE 17 AUXILIARY BUILDING -
HIGH ENERGY BREAK ANALYSIS (3"
& 6" RWCU AND PRESSURE TRANSIENTS IN NODE 17 LINE 4" RCIC) 3B-12A 3B-13 3B-13A 3B-14 3B-14A 3B-15 3B-15A 3B-16 3B-16A 3B-17
AUXILIARY BUILDING -
HIGH ENERGY LINE BREAK ANALYSIS (8"
RHR)
PRESSURE TRANSIENTS IN NODE 18 AUXILIARY BUILDING -
HIGH ENERGY BREAK ANALYSIS (311
PRESSURE TRANSIENTS IN NODE 18 AUXILIARY BUILDING -
HIGH ENERGY LINE BREAK ANALYSIS (8"RHR)
PRESSURE TRANSIENTS IN NODE 19 AUXILIARY BUILDING -
HIGH ENERGY BREAK ANALYSIS (3"
PRESSURE TRANSIENTS IN NODE 19 AUXILIARY BUILDING -
HIGH ENERGY LINE BREAK ANALYSIS (8"
RHR)
PRESSURE TRANSIENTS IN NODE 20 AUXILIARY BUILDING -
HIGH ENERGY BREAK ANALYSIS (3"
PRESSURE TRANSIENTS IN NODE 20 AUXILIARY BUILDING -
HIGH ENERGY LINE BREAK ANALYSIS (8"
RHR) 3B-19 3B-19A 3B-20 3B-20A 3B-21 3B-21A
RBS USAR TABLE 3B-1 HIGH-ENERGY LINE BREAKS AUXILIARY BUILDING 20-NODE MODEL Brea Brek inDesign Break N.Line(*
Node for Nodes (2)
S1 3"1 RWCU lW9,10,15 2
6" RWCU 6
(4) 3 4" RCIC 2
(4) 4 8"
RHR 12 1,2,3,4,5,6 7,8,11,12,13 14,16,17,18,19 20
./
/"
/
/
/
//
(')All breaks are assumed to be double-ended ruptures.
(2) Subcompartment nodes are defined in Table 3B-3 and on
//
Fig. 3B-l.
(3 )This break also could occur in Nod Ie 9. Consequently, the results for Node 10 are applied to Node 9 considering symmetry.
(4
) rea do s
n t
ge era e
d sig
.ý ress re or ny ode August 1988 1 of 1
RBS USAR TABLE 3B-1 HIGH-ENERGY LINE BREAKS AUXILIARY BUILDING Break No.
Line"'
Break Room 1
3" RWCU The RWCU Pump Room (EDC Zone AB-095-3) 2 6"
RWCU The RWCU Hoist Compartment (EDC Zone AB-095-4) 3 4"
RCIC The RCIC Pump Room (EDC Zone AB-070-3) 4 8"
RHR The RHR Equipment Removal Cubicle (EDC Zone AB-114-8A or 8B)
(')All breaks are assumed to be double-ended ruptures.
August 1988 1 of 1
RBS USAR TABLE 3B-3 SUBCOMPARTMENT NODAL DESCRIPTION AUXILIARY BUILDING 20-NODE MODEL Net Node Volume
/
Break Break Break Number (ft
- 3)
Description of Volume Location Type Line111 1
9,685 RHR 'C' Equipment Room, Node 12 Steam 8" RHR EDC Zone AB-070-4 2
12,524 RCIC Pump Room, Node 12 Steam 8" RHR "EDC Zone AB-070-3 3
22,845 RPCCW Equipment Area, Node 12 Steam 8" RHR
/
EDC Zone AB-095-8 4
1,181/
East-West Passageway, Node 12 Steam 8" RHR
/
EDC Zone AB-095-4 5
4?80 Unit Cooler Area, Node 12 Steam 8" RHR EDC Zone AB-095-4 6/
453 RCIC Access Area, Node 12 Steam 8" RHR EDC Zone AB-095-4
/
7 2,535 Hoist Area, Node 12 Steam 8" RHR EDC Zone AB-095-4
/
8 21,864 Elevator Area, Node 12 Steam 8" RHR EDC Zone AB-095 9
627 RWCU 'A' PumROom, Node 9(3)
Liquid 3" RWCU EDC Zone AH/095-3
//
10 627 RWCU 'B'-Pump Room, Node 10 Liquid 3" RWCU EDC Zo/e AB-095-3
'-+1
/
11 71,439 RPC"W Equipment Area, Node 12 Steam 8" RHR EDO Zone AB-070-8
/
12 86,570
/ CC Area (East),
Node J2 Steam 8" RHR
/ EDC Zones AB-1 14-3,5,
/
and 8B 13 90,157
/
MCC Area(West),
Node 12 Steam 8" RHR EDC Zones AB-1 14-1,6, and 8A Absolute Calculated Design Peak Peak Peak Pressure Differential Pressure Differential 121 Pressure (psia)
(psid) s_
(sid) 17.03 2.33 2.40 17.03 23 2.40 17.03
.32 2.40 17.02 2./32 2.40 17.02 2.32 2.40 17.02 2.32 2.40 17.02 2.32 2.40 17.02 2.32 2.40 17.02 23 2.40 17.94 3.24 3.30 17.94 3.24 3.30 17.03 2.33 2.4 17.01 2.31
/
2.40
/1.
17.01 2.31 2.40 August 1988 Veý(61614 W"Pý n-Z4'j 1 of 2
RBS USAR SUBCOMPARTMENT NODAL DESCRIPTION AUXILIARY BUILDING I
20-NODE MODEL Absolute Calculated Design Pea Net Peak Peak Pressure Differenti 1 Node Volume Break Break Break Pressure Differential 121 Pressure Number (ft 3 )
Description of Volume Location Type Line___
(psia)
(psid)
(psid) 14 212,931 General Area, Node 12 Steam 8" RHR 17.00 2.30 2.40 EDC Zones AB-141-1,2,3, 4, and G
/
15 31ý RWCU Piping Area, Node 10 Liquid 3" RWCU 17.13 2.43 3.30
/
EDC Zone AB-095-3
/
/
16 10,084 Annulus Mixing Fan Area,Node 12 Steam 8" RHR 16.98 2.28 2.40 2.40
/
EDC Zone AB-170-1
/
17
/ 3,443 Stairwell to Elev.
Node 12 Steam 8" RHR 16.98 2.28 2.40
/
Mach. Room,
/
EDC Zone AB-170-1
//
18
/
3,336 Rad. Monitor Area, Node 12 Steam 8" RHR 16.98 2.28, 2.40 EDC Zone AB-1 70-1
/
S/
/
19
/
6,040 Continuous Filter Room, Node 12 Steam
,/
8" RHR 16.99 2.29 2.40
/
EDCZoneAB-172
/2 20 3,922 Continuous Filter Room, Node 12 Steam 8" RHR 16.99 2.29 2.40 EDC Zone A)-1 70-2
/
/
/
1**.
/I
/
S~/
/
/
//
/
/
t ')AIl breaks are double-ended ruptures (i.e., break flow area is twice the pipe cross-sectional area).
.)Calculated by subtracting 14.7 psia from the maximum absolute pressure for each node.
B3~~reak in Node 9 was not analyzed, symmetry the results are sumed to be same as those for Node 10.
2 of 2 August 1988
RBS USAR TABLE 3B-3 SUBCOMPARTMENT NODAL DESCRIPTION AUXILIARY BUILDING EDC Zone AB-070-1 AB-070-2 AB-070-3 AB-070-4 AB-070-5 AB-070-6 AB-070-7 AB-070-8 AB-095-1 AB-095-2 AB-095-3 AB-095-4 AB-095-4 AB-095-5 AB-095-6 AB-095-7 AB-095-8 AB-114-1 & 8A AB-1 14-2 AB-1 14-3 AB-1 14-4 AB-1 14-5 AB-1 14-6 AB-1 14-8B Description of Volume CSL Area RHS-P1A Pump Room ICS Pump Room RHS-P1C Pump Room RHS-P1IB Pump Room HPCS Pump Room Elevator Area RPCCW Area CSL Hatch Area RHS Heat Exchanger Area (West)
WCS Area Hoist Area (Sub-Volume #1)
Hoist Area (Sub-Volume #2)
RHS Heat Exchanger Area (East)
HPCS Hatch Area Elevator Area RPCCW Area MCC Area and RHR Equipment Removal Cubicle (west)
Main Steam Tunnel (North)
MCC Area (East)
Post Accident Sampling Station Elevator Room RPCCW Area RHR Equipment Removal Cubicle (East)
Vol. (ft3)
Absolute Peak Pressure (psia) 13992 22733 12524 9685 22733 13927 35720 35720 11548 16402 1567 12614 2535 16402 22734 21864 22845 55573 26775 30381 1945 31873 34584 24613 16.49 16.49 16.49 16.48 16.48 16.48 16.48 16.49 16.48 16.48 16.48 16.47 16.47 16.47 16.47 16.47 16.48 16.47 14.70 16.46 16.46 16.46 16.47 16.46 Calculated Peak Diff.
Pressure(1) (2)
(psid) 1.79 1.79 1.79 1.78 1.78 1.78 1.78 1.79 1.78 1.78 1.78 1.77 1.77 1.77 1.77 1.77 1.78 1.77 0.00 1.76 1.76 1.76 1.77 1.76 August 1988 I of 2
RBS USAR TABLE 3B-3 SUBCOMPARTMENT NODAL DESCRIPTION AUXILIARY BUILDING EDC Zone Description of Volume Equipment Area (West)
Equipment Area (East)
Elevator Area RPCCW Area Standby Gas Treatment Filter (West)
Standby Gas Treatment Filter (East)
Annulus Mixing System Fan Area (Sub-Volume #1)
Annulus Mixing System Fan Area (Sub-Volume #2)
Annulus Mixing System Fan Area (Sub-Volume #3)
Continuous Filter Room Elevator Machine Room Vol. (ft3) 62074 70772 39813 40273 45330 42256 12172 3336 1930 9962 1313 Absolute Peak Pressure (psia) 16.45 16.45 16.45 16.45 16.45 16.45 16.44 16.44 16.43 16.44 16.43 Note:
(1) The calculated peak differential pressures were calculated by subtracting 14.7 psia from the maximum absolute pressure for each node.
(2) The design peak differential pressure acceptance criteria are 3.30 psid for EDC Zone AB-095-3 and 2.40 psid for the rest of Auxiliary Building.
August 1988 Calculated Peak Diff.
Pressure (1 (2)
(psid) 1.75 1.75 1.75 1.75 1.75 1.75 1.74 1.74 1.73 1.74 1.73 AB-141-1 AB-141-2 AB-141-3 AB-141-4 AB-141-5 AB-141-6 AB-1 70-1 AB-1 70-1 AB-1 70-1 AB-1 70-2 AB-170-3.
2 of 2
TABLE 3B-5 SUBCOI'MPARfMENT VENT PATHf-ý RI-PTION-AUXILIARY BUILDING 20-NODE MODEL
(
lFrom To Inertia Vent Vol.
Vol.
Vent
- Factor, Head Loss C(
Path Node Node Area L/A No.
No.
No.
(ft
- 2)
(ft-1)
Contraction Expansion ObstructionI1)
Jil 9
15 21.0 0.204 0.279 0.693 0.747 J2 10 15 21.0 0.204 0.279 0.693 0.747 S~/.
J3 15 5
15.75 0.156 0.234 0.504 0.980 J4
/5 8
13.1 0.4342 J5 5
4 57.0 0.176/
0.464 0.010
/
/50 J()/5 7
21.0 0.'0/8 0.442 0.781 0.906 5
5 5
4 3
7 6
/
2 3.:
8 3
7 6
6 6
4
/2 1
11 11 13 3.0 21.0
- 3. Q 10.75 105.0 114.75 23.88 271.3 115.0 272.43
,0.779 0.494 3.223 0.169 0.5883 0.091 0.059 0.104
- 0. 0o/9
.032 0.018 8
12 115.0 0.031 1 of 2 August 1987 K<
Jl0 Jill J12 J13 J14(4)
TABLE 3B-5 SUBCOMPARTMENT VENT PATH DESCRIPTION AUXILIARY BUILDING 20-NODE MDEL J19 12 14 115.0 0.027 0.484 0.970 1.163 0.024 2.641 From To Inertia Vent Vol.
Vol.'
Vent
- Factor, Head Loss Coefficient Path Node NodA Area L/A Turning No.
No.
No//
(ft
- 2)
(ft"1)
Contraction Expansion Obstruction1')
Friction Loss Total 20 13 4
391.0 0.012 0.446 0.903 1.121 0.003 2.473 J21 16 "17 203.2 0.0563 0.282 0.012 0.294 i
/
J22 14 17 21.0 0.0415 0.454 0.988
/'- 1.442 1/'/
,/
~/,
J23 1
18 146.62 0.0566 0.343 0.118
/-
0.461
//
24 18 20 21.0 0.0 6 52 0.453 0.838 71 7
1.291
/
/
//
J25 20 19 207.0 0.05 0.084 0.141
-0.225 26 14 16 28.0
- 0. 315 0.495 0.856 1.351 (1) This term include grating, orifice, mesh door, and any other form loss blocking the vent path.
(2) Closed door (wit 3.0-ft.
2 ventilation louver) modeled to open at 3.5 psid.
( Door louver mod ed to close at 3.5 psid when door opens.
(4) Watertight door modeled to open at 3.5 sid.
August 1987 2 of 2
-qAND ENERGY RELEASE I-N A B DE IN AUXILIARY BUILDING -
NODE 10
'I I
/
I Total Enthalpy F
w Rate
(-tu/sec) 0/0 277,000 277,000 262,000 262,000 188,000 188,000 135,000 110, 00~t /
110 o/.000
/ i
-I
Af&
kA.
August 1987 Time (sec) /
0./01 2. 120 2.121 4.150 4.151
- 6. 940 8.000 8.500 19.810 22. 0001 Total Mass Flow Rate (lbm/sec) 0-0 522.4 522.4 494.3 494.3 354.5 354.5 255.0 208.5 208.5 /
/
0.0/
/
/
/
/
I 1 of 1
RBS USAR TABLE 3B-7 MASS AND ENERGY RELEASE 3-IN RWCU DER IN AUXILIARY BUILDING Time (sec) 0.000 0.001 1.900 2.000 3.800 3.900 12.000 13.200 13.700 14.000 14.300 14.400 14.500 14.700 15.000 19.500 24.300 25.300 27.300 27.400 27.900 28.300 Total Mass Flow Rate (lbm/sec) 0.0 357.8 357.8 336.9 336.9 275.3 275.2 275.2 258.1 228.3 211.4 183.7 164.2 158.2 117.2 117.0 115. 8 114.2 79.4 46.9 40.3 0.0 Total Energy Flow Rate (Btu/sec) 0 198956 198956 187344 187344 153069 153059 153053 143550 126940 117534 102146 91283 87974 65169 65060 64415 63495 44135 26106 22407 0
August 1987 1 of 1
RBS USAR TABLE 3b-8 cd, August 1987 1 of I
RBS USAR TABLE 3B-8 MASS AND ENERGY RELEASE 6-IN RWCU DER IN AUXILIARY BUILDING Time (sec) 0.000 0.001 0.900 1.000
- 1. 100 26.500 28.300 Total Mass Flow Rate (lbm/sec) 0.0 1411.9 1411.9 706.0 165.2 165.2 0.0 Total Energy Flow Rate (Btu/sec) 0 785159 785159 392579 91845 91845 0
August 1987 1 of 1.
RBS USAR TABLE 3b-9 SS S 4D ENERGYF
/IN A XTLIARY BUILDIN
//
Tot 1 Mass Tim'I w Rate
(,,e)
(cbm/sec)
/
/0.0
/" 0.0 0.001 53.86 0.082 53.86 0.083 71.82 12.738 71.82 13.768 0.0 August 1987 1 of i.
RBS USAR TABLE 3B-9 MASS AND ENERGY RELEASE 4-IN RCIC DER IN AUXILIARY BUILDING Time (sec) 0.000 0.001 0.245 0.250 12.000 13.000 14.000 15.000 16.000 17.000 18.000 19.000 19.900 20.900 21.000 21.100 21.200 21.400 21.500 21.900 Total Mass Flow Rate (lbm/sec) 0.0 134.8 134.8 72.9
- 72. 9
- 72. 8
- 72. 8
- 72. 7 72.5 72.2 71.6 70.2 67.6 55.2 44.1 37.6 33.8 28.8 3.1 0.0 Total Energy Flow Rate (Btu/sec) 0 160373 160373 86666 86666 86636 86571 86461 86260 85861 85166 83530 80442 65663 52481 44733 40242 34273 3737 0
August 1987 1.0f 1
TABLE 3B-10
,MASS AND ENERG REL ASE S8-IN RHR DER IN AUXI ARY BUILDING -,aOD]
12
'Total Enthalpy Flow Rate (Btu/sec) 0.0 287,845 287,845 273,56
,24 331 864,551 0.0 August 1988 RBS USAR
.001 2.0 4.5 7.0 9.5 12.0 0.0 241.4 241.4 229.4 209.1 138.!
1 of I
RBS USAR TABLE 3B-10 MASS AND ENERGY RELEASE 8-IN RHR DER IN AUXILIARY BUILDING Time (sec) 0.000 0.001 0.100 0.200 1.900 2.500 3.000 3.500 4.000 4.500 5.000 5.500 6.000 6.500 7.000 7.500 7.800 7.900 8.000 8.200 8.300 8.500 8.600 8.900 9.800 9.900 10.000 10.200 10.800 10.900 11.000 11.300 11.400 11.800 Total Mass Flow Rate (lbm/sec) 0.0 509.4 509.4 266.8 266.8 266.0 265.1 263.7 262.1 259.9 256.6 252.2 246.6 237.9 227.4 212.8 204.0 198.8 192.0 181.2 179.4 175.9 174.2 167.3 132.2 125.7 116.9 102.9 72.8 53.5 43.0 28.8 3.1 0.0 Total Energy Flow Rate (Btu/sec) 0 605951 605951 317333 317333 316362 315348 313613 311740 309137 305216 300010 293348 282938 270446 253096 242686 236440 228343 215506 213416 209237 207148 198999 157301 149558 139004 122340 86547 63645 51106 34273 3737 0
August 1988 1 of 1
RBS USAR TABLE 3B-15 HEAT SINK SLAB DESCRIPTION AUXILIARY BUILDING 20-NODE MODEL Exposed Surface Thickness Material (ft) 2,877 Concrete(3 )
1.0
/,3,320 Concrete{3 )
/
1.0 5,076 Concrete(3 )
1.0 664 Concretet3) 1.0 1,750 Concrete{
3 1.0 2,347 Con,9rete{3) 1.0 947 Qoncrete(3) 1.0 5780
/
Concrete(3) 1.0 15,949/
Concrete(3) 1.0 19,844
/
Concrete(3) 1.0
//
17,880 Concrete(3 )
1.0
/
38,632 Concrete(3 )
1.0 1,106 Concrete(3) 1 3 35 Concrete(')
1.0 1
830 Concrete(3) 1.0 1,287 Concrete 1.0 1,662 Concrete 1.0 2,209 Concrete 1.0 743 Carb oSteelt3) 0.0306 935 C4rbon Steel(3) 0.0307 1
of 2Auittst 1988
{I(e-
Slab Node Exposure(l)
No.
Left Right 21
/
3 3
22/
5
. 5 3
6 6
24 8
8 25 11 11 26 16
/
16 27 17
/
17 28(4) 12
/
0 29(4) 1 0
30(4) 14 0
('Node numbers are defined in T le 3B-3 and on Fig. 3B-I.
Zero exposure indicates an in lated boundary assumption with zero heat transfer at t s boundary.
( Thermal Properties:
Conductivity, tu/hr-°F-ft 3 Volumetric I eat capacity, Btu/'F-ft3 (4) Heat sinks only applicable to 8" RHR HELB analysis.
Exposed Surface Area Mf2) 1,875 278 476 1,214 7,344 1,894 10/8
/1,798 982.5 2,943 Material Carbon Steel(3)
Carbon Steel(3)
Carbon Steel(3)
Carbon Steel(3)
Carbon Steel(3)
Carbon Steel03)
Carbon Steel(-)
Carbon Steelt,).
Carbon,_teel(3)
Cýrbon Steel(
- 3)
//
Concrete Carbon Steel 0.8 26.0 23.2 53.9 August 1988 Thickness 0.05 0.0403 0.0439 0.06 10 0.0460 0.04 0.04 0.00529 0.00529 0.00529 1)
(<
C&
"2 of 2
NOTE:
- THE NOTATION 070-3 FOR EXAMPLE, REFERS TO ENVIRONMENTAL ZONE 3 ON ELEVATION 70'0" OF THE
/
AUXILIARY BUILDING
'I, S
/
J Q.._ d*
- 13132
,I311ir3 FIGURE 3B-2A PRESSURE TRANSIENTS IN NODE 1 AUXILIARY BUILDING HIGH ENERGY LINE BREAK ANALYSIS RIVER BEND STATION UPDATED SAFETY ANALYSIS REPORT
17.0 16.5 16.016" RWCU 16.0 w
W* 15.5 00 UJ w
15.0 14.5 14.01 1.E-01 1.E+00 1.E+01 1.E+02 1.E+03 1.E+04 1.E+05 TIME AFTER ACCIDENT (SECOND)
FIGURE 3B-27 PRESSURE TRANSIENTS FOR EDC ZONE AB-070-3 RIVER BEND STATION UPDATED SAFETY ANALYSIS REPORT
17.0 16.5 6"* RWCU 3" RWCU 16.0 w
, 15.5 U,
U) w 15.0 4" RCIC 14.5 14.0 I.E-01 1.E+00 1.E+01 1.E+02 1.E+03 1.E+04 1.E+05 TIME AFTER ACCIDENT (SECOND)
FIGURE 3B-28 PRESSURE TRANSIENTS FOR EDC ZONE AB-095-3 RIVER BEND STATION UPDATED SAFETY ANALYSIS REPORT
17.0 16.5 16.5 6" RWCU 16.0 w
LW 15.5 U) w 15.0 3" RWCU 4" RCIC 14.5 14.01 1.E-01 1.E+00 1.E+01 1.E+02 1.E+03 1.E+04 1.E+05 TIME AFTER ACCIDENT (PSIA)
FIGURE 3B-29 PRESSURE TRANSIENTS FOR EDC ZONE AB-095-4 RIVER BEND STATION UPDATED SAFETY ANALYSIS REPORT
17.0 16.5 16.0 U) 0.
l 15.5 w
15.0 15.0 3" RWCU 14.5 4" RCIC 14.0 1.E-01 1.E+00 1.E+01 1.E+02 1.E+03 1.E+04 1.E+05 TIME AFTER ACCIDENT (SECOND)
FIGURE 3B-30 PRESSURE TRANSIENTS FOR EDC ZONE AB-1 14-8A & 8B RIVER BEND STATION UPDATED SAFETY ANALYSIS REPORT