RBG-45940, License Amendment Request (LAR) 2001-43, High Energy Line Break Analysis Method.

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

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En EntergyTel toWFax St. Francisville, LA 70775 225 381 4374 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) THREED GOTHIC THREED GOTHIC 1 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.

for Thus, Entergy has established expectations for developing and demonstrating capabilities with Generic Letter 83-11 use of analysis codes such as GOTHIC which are consistent Bend supplement 1. Additionally, as a member of the GOTHIC Advisory Group, EOI and River model have the ability to consult and exercise the GOTHIC code developer (NAI) on GOTHIC development or detailed coding issues.

do not Based on the above discussions, Entergy has determined that the proposed changes require any exemptions or relief from regulatory requirements, including the Technical in the Specifications, and do not affect conformance with any GDC differently than described 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 6 -xi August 1987

RBS USAR CHAPTER 6 LIST OF FIGURES (Cont)

Figure Number Title 6.2-38 NODALIZATION DIAGRAM FEEDWATER LINE BREAK RPV-SHIELD WALL ANNULUS 27 NODE MODEL 6.2-39 NODAL PRESSURES FEEDWATER LINE BREAK RPV-SHIELD WALL ANNULUS 27 NODE MODEL 6.2-40 NODAL PRESSURE DIFFERENTIALS FEEDWATER LINE BREAK RPV-SHIELD WALL ANNULUS 27 NODE MODEL 6.2-41 NODALIZATION DIAGRAM FEEDWATER LINE BREAK RPV-SHIELD WALL ANNULUS 25 NODE MODEL 6.2-42 NODAL PRESSURES FEEDWATER LINE BREAK RPV-SHIELD WALL ANNULUS 25 NODE MODEL 6.2-43 NODAL PRESSURE DIFFERENTIAL FEEDWATER LINE BREAK RPV-SHIELD WALL ANNULUS 25 NODE MODEL 6.2-44 NODALIZATION DIAGRAM RECIRCULATION OUTLET LINE BREAK RPV-SHIELD WALL ANNULUS 26 NODE HALF MODEL 6.2-45a NODAL PRESSURES RECIRCULATION OUTLET LINE BREAK through RPV-SHIELD WALL ANNULUS 26 NODE HALF MODEL 6.2-45e 6.2-46 NODAL PRESSURE DIFFERENTIAL RECIRCULATION RPV-SHIELD WALL ANNULUS 26 NODE HALF MODEL 6.2-47 NODALIZATION DIAGRAM 6-IN RCIC HEAD SPRAY LINE BREAK DRYWELL HEAD SUBCOMPARTMENT 6.2-48 NODAL PRESSURES 6-IN RCIC HEAD SPRAY LINE BREAK DRYWELL HEAD SUBCOMPARTMENT 6.2-49 NODAL PRESSURE DIFFERENTIALS 6-IN RCIC HEAD SPRAY LINE BREAK DRYWELL HEAD SUBCOMPARTMENT 6.2-50 NODALIZATION DIAGRAM 4-IN AND 6-IN RWCU LINE BREAK RWCU HEAT EXCHANGER ROOM 6.2-51 NODAL PRESSUR 6-IN RWCU LINE BREAK I Deleted: s RWCU HEAT EXCHANGER ROOM 6-xvii August 1987

RBS USAR CHAPTER 6 LIST OF FIGURES (Cont)

Figure Number Title pEETED....

6.2-52 . Deleted: NODAL PRESSURE

• DIFFERENTIALS 6-IN RWCU LINEI

, BREAK RWCU HEAT EXCHANAGER 6.2-53 NODALIZATION DIAGRAM 8-IN RWCU LINE BREAK IROOM I RWCU FILTER DEMINERALIZER ROOM 6.2-54 NODAL PRESSURES 8-IN RWCU LINE BREAK RWCU FILTER DEMINERALIZER ROOM 6.2-55 DELETED ... ....................... Deleted: NODAL PRESSURE DIFFERENTIALS 8-IN RWCU LINEIL

.BREAK RWC1J FILTER 6.2-56 RHR SUPPRESSION POOL COOLING MODE SUCTION DEMINERALIZER ROOM AND RETURN (PLAN) 6.2-57 RHR SUPPRESSION POOL COOLING MODE SUCTION AND RETURN (SECTION 1-1) 6.2-58 STANDBY GAS TREATMENT SYSTEM (P&ID) 6.2-59 SCTS FAN PERFORMANCE CURVE 6.2-60 FUEL BUILDING CHARCOAL FILTRATION SYSTEM F PERFORMANCE CURVE 6.2-61a PRESSURE IN SHIELD BUILDING ANNULUS VS TIM 6.2-61b PRESSURE IN AUXILIARY BUILDING VS TIME 6.2-62 PRESSURE IN FUEL BUILDING VS TIME 6.2-63 CRITERION 55 - CONTAINMENT ISOLATION VALVE 6.2-64 CRITERION 56 - CONTAINMENT ISOLATION VALVE 6.2-65 CRITERION 56 - CONTAINMENT ISOLATION VALVE 6.2-66 HYDROGEN MIXING PURGE AND RECOMBINER P&ID 6.2-67 DELETED 6.2-68 HYDROGEN CONCENTRATION VS TIME AFTER LOCA 6-xviii

• August 1987

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. Deleted: in Revision 12 6.2-24 December 1999

RBS USAR Deleted: the analysis, the The postulated DER of the 4-in RWCU pump discharge line between instrument delay time is assumed to be 1 sec. ¶ 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 Deleted: 1 sec after the closed sufficiently such that the isolation valve flow area break. At 5.5 sec, equals the flow restrictor area,,_the critical flow location_[ Deleted: . At that time, changes from the flow restrictor to the isolation valves. Flow Deleted: Subsequent closure from the heat exchanger side of the break is limited to critical of the valves terminates flow through the pipe cross-sectional area and is assumed to flow at 6.0 sec.

terminate when the contents of the h exchangers

.eat and I-Deleted: regenerative 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 Deleted: o.s6 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.

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.

6.2-42 August 1987

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 Revision 12 6.2-43 December 1999

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_ i Deleted: and 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 i Deleted: . For all cases, the Iblowdown data is based upon calculated based on the ucrated oower conditions (3100 M4t) and conservative methodology developed maximum reactor pressure (1090 psia). An additional 5-second by GE using the Moody steady-slip flow model with subcooling, as time delay in the isolation logic has been assumed for the RWCU described in Reference 9. The line breaks. blowdown mass and energy used in the subcompartment calculation The assumed initial conditions for the subcompartment volumes are Deleted: 102% of the original reactor conservatively chosen so as to maximize transient pressure [power and original reactor the pressure. Evaluations performed at responses. The initial conditions are given in 102% of current rated power and subcompartment nodal description tables. 1072 psia reactor pressure demonstrated that due to the conservatisms in the methodology, The description of and justification for the subsonic and sonic the break mass and energy flows flow model, and the degree of entrainment used in vent flow calculated at the original reactor power and pressure remain calculations are given in Appendix 6B. conservative for application to current rated power conditions.¶ 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

RBS USAR TABLE 6.2-12 BLOWDOWN DATA 4-IN RWCU PUMP DISCHARGE LINE BREAK CONTAINMENT HIGH ENERGY LINE BREAK ANALYSIS Blowdown ~707 Mass B lowdown Flow Rate E:

(lbm/sec) 956 /

956 /

7707 777 07 448 651 442

/ 651 442 651 367 651 36 173 59 1173 29 0 0

/

rep net,,

rtA -f 1 of 1 August 1987

RBS USAR TABLE 6.2-12 BLOWDOWN DATA 4-IN RWCU PUMP DISCHARGE LINE BREAK CONTAINMENT HIGH ENERGY LINE BREAK ANALYSIS Time Blowdown Mass Enthalpy (sec) Flow Rate (Btu/lbm)

(lbm/sec)

Upstream Blowdown 0.0000 0.0 531.44 0.0001 564.7 531.44 0.5774 564.7 531.44 0.5775 212.1 531.44 13.4380 212.1 531.44 15.0000 0.00 531.44 Downstream Blowdown 0 .0000 0.0 472.02 0 .0001 610 .2 472.02 0.7315 610.2 472.02 0.7316 610 .2 361.60 1.5297 610 .2 361.60 1.5298 610 .2 257.54 2.3788 610 .2 257.54 2.3789 610 .2 150.17 4.6461 610.2 150.17 4 .6462 610 .2 93.91

21. 9969 610.2 93.91 21.9970 610.2 146.46 23.5400 610.2 146.46 23.5401 610.2 252.64 25.0137 610.2 252.64 25.0138 610.2 362.04 26.3952 610.2 362.04 26.3953 610.2 419.00 28.3703 610.2 419.00 28.3704 0.0 0.00 1 of 1 August 1987

RBSUSAR TABLE 6.2-13 CONTAINMENT SUBCOMPARTMENT ANALYSIS

SUMMARY

Design Tables Figures Basis Vent Nodal Line Nodal Path Blowdown Nodalization Nodal Pressure Subcompartment Break Description Description Data Diagram Pressures Differentials RPV - Shield Feedwater 6.2-14 6.2-15 6.2-16 6.2-38 6.2-39 6.2-40 Wall Annulus RPV - Shield Wall Annulus Feedwater 6.2-17 6.2-18 6.2-19 6.2-41 6.2-42 6.2-43 RPV - Shield Wall Annulus Recirculation 6.2-20 6.2-21 6.2-22 6.2-44 6.2-45 6.2-46 water outlet

---

12 Drywell Head RCIC head () 6.2-23 6.2-24 6.2-25 6.2-47 6.2-48 6.2-49 spray 12+-

RWCU Heat Exchanger Room RWCU 6.2-26 6.2-27 6.2-50 6.2-51 6.2-218 RWCU Filter/ ýJWIrw (dCe,2,1 (,.I 1-o4re-Demineralizer Rooms RWCU 6.2-29 6.2-30 6.2-31 6.2-53 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.

1 of I December 1999 Revision 12

RBS USAR TABLE 6.2-13 CONTAINMENT SUBCOMPARTMENT ANALYSIS

SUMMARY

Design Tables Finures Basis Vent Nodal Path Blowdown Nodalization Nodal Pressure Line Nodal Description Description Data Diagram Pressures Differentials Subcompartment Break RPV - Shield Feedwater( ) 6.2-14 6.2-15 6.2-16 6.2-38 6.2-39 6.2-40 Wall Annulus RPV - Shield 6.2-42 6.2-43 Feedwater(l) 6.2-17 6.2-18 6.2-19 6.2-41 Wall Annulus RPV - Shield 6.2-45 6.2-46 Recirculation 6.2-20 6.2-21 6.2-22 6.2-44 Wall Annulus water outlet(2)

• -> 12 6.2-48 6.2-49 RCIC head (3) 6.2-23 6.2-24 6.2-25 6.2-47 Drywell Head spray 12<-

RWCU Heat 6.2-51 N/A RWCU 6.2-26 6.2-27 6.2-28 6.2-50 Exchanger Room 6.2-12 RWCU Filter/

Demineralizer 6.2-54 N/A RWCU 6.2-29 6.2-30 6.2-31 6.2-53 Rooms

(')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

RBS USAR TABLE 6.2-26 SUBCOMPARTMENT NODAL DESCRIPTION 6-IN RWCU LINE BREAK RWCU HEAT EXCHANGER ROOM Calculated")I Initial Conditions DBA Break Conditions Peak

/ Break Pressure Volume Volume Temp. Pressure Humidity  % Break Break ea Break Difference No. (cu ft) (OF) (psia) (%) in Vol. Line Asq ft) type ( s'd

/ 7 RWCU See Tble 62-28 1 13,250 103 14.7 0 100 / I ER .3 2 7,149 103 1i.:7 0 0.0 / Z 0.3

/

3 6,312 103 14.7 0 0.0 0.3 4 1,164,879 103 14.7 0 0.0 0.0 6+-.

(Nodal peak pressure minus pressure in node 4 (Pi - P4 )

Revision 6 1 of 1 August 1993

RBS USAR TABLE 6.2-26 SUBCOMPARTMENT NODAL DESCRIPTION 4-IN and 6-IN RWCU LINE BREAKS RWCU HEAT EXCHANGER ROOM Calculated Initial Conditions DBA Break Conditions Peak Pressure Humidity  % Break Break Break Difference Volume Volume Temp. Pressure (OF) (psia) (%) in Vol. Line (psid)

No. (cu ft)

• -+6 13,250 103 14.7 0 100 RWCU DER 1.627 (4-in) 1 1.488 (6-in) 2 7,059 90 14.7 0 0.0 <0.5 14.7 0 0.0 *0.5 3 6,153 90 0 0.0 *0.5 4 1,165,128 90 14.7 14.7 100 0.0 N/A 5 358,000 120 (see note 1) 6+--o 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 By assuming a high initial temperature for except the break room, which has high temperatures after the break.

Volume No. 5, more heat is transferred into the other volumes, which generates more limiting pressure/temperature responses.

August 1993 1 of 1 Revision 6

RBS USAR TABLE 6.2-27

- .. COMPAR-TMENT-VENT. PATH-DESCRIPTI-Of..

6-IN RWCU LINE BREAK RWCU HEAT EXCHANGER ROOM From To Description Ven itt Vol. Vol. of Vent Vent Loss Due to Pat h Node Node Path Flow Area L/A Head Loss Coefficient Thick Edged No. No. No. (Choked/Unchoked) (sq ft) (ftI) Friction Turning Expansion Contraction Oriface Total IA 1 2 Unchoked 1.628 0.168 0.036 0.998 0.5 0.04 1.573 2 1 Unchoked 1.628 0.168 0.036 0.998 0.5 0.04 1.573 1 2 Unchoked 15,56 0.168 3.38 0.497 3.877 2 1 Unchoked ;5.56 0.168 3.38 0.978 4 .358 lc .1 2 Unchoked 11.024 0.168 0.013 0.985 0.496 0.766 2.260 2 1 Unchoked 11.024 0.168 0.013 0.985 0.496 0.766 2.260 2A 2 Unchoked 1.628 0.724 0.036 0.998 0.5 0.04 1.573 2 1 Unchoked 1.628 0.724 0.036 0.998 0.5 0.04 1.573

/ 1 1 2 Unchoked 1.628 0.724 1.327 - 0.5 1.827 2c , S 2 1 1 Unchoked 1.628 0.724 1.327 - 0.998 2. 32S S1 2 Unchoked 2.806 0.724 0.963 - 0.5 1.463 2C" 2 1 Unchoked, 2.806 0.724 0.963 - 0.996 1.959 12 2 Unchoked 2.965 0.724 0,09 0.996 0.499 1.585 2 1 Unchoked 2.965 0.724 0.09 0.996 0.499/ 1.585

.2E 2 Unchoked 1.18 0.724

• 0.009 0.998 0.5/ 1.507 1 Unchoked 1.18. 0.724 0.009 0.998 0 1.507

/

3 1 4 Unchoked 15.75 0.83j.. 2.15 0.133 1.0 0.164 0.33 3.780

1. 0 / 0. 164 0.33 4 1 Unchoked 15.75 0.8;3 2.15 0.133 3.780 4 2 3 Unchoked 194.02 /92 0.590 0 90 3 2 Unchoked 194.02 0.092 0.Y43 .436 2 4 Unchoked 148.6 0.233 0.027 0/.933 0.0325 0.148 1.141 4 2 Unchoked 148 6 0.233 0.027 0.0042 0.483 0.0382 0.552 655 / . . .. .

2 4 Unchoked 1486 0.233 0.027 0.933 0.0325 0.38 1.141 4 2 Unchoked 0.233 0.027 0.0042 0.483 0.552 0ý 382 7 2 4 Unchoked 44.94 0.261 0.053 0.996 0.458 1.5 4 2 Unchoked

• 14.94 0.261 0.053 0.831 0.499 Unchoked 172.5 0.162 0.922 0.922 88 3 4 172.5 0.162 0.490 0.490 4 3 Unchoked Y,

q9 1 of 2 August 1987

RBS USAR TABLE 6.2-27 (Cont)

/

From To,/ Description Vent Vol. V)I. of Vent Vent / /2 Loss Due to Path Node Node Path Flow Area L/A Head Loss Coefficient Thick Edged No. No. No. (Choked/Unchoked) (sq ft) (ftl) Friction Turning Expansion Contract on Oriface Total 7

/

./

9 3' 4 Unchoked 172.5 0.162 0.922 0.922 3 Unchoked 172.5 0.162 0.490 0.490 re ." ) jk'/

4 NOTES:

1

1. Vent pa hs . , 1b, and Ic are combine into one vent path (vent pa 1) .

2 2 2 2

2. Ven/taths A, a, c, 2D and E ar combined into one vent 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.

1 of 2 August 1987

RBS USAR TABLE 6.2-28

/ 6-IN RWCU LINE BREAK RWCU HEAT EXCHANGER ROOM Belowdown Blowdown Total

/ Mass Blowdown 17 Energy Effective

/ Time Flow Rate Enthalpy Release Rate Break Area (sec) (lbm/sec) (Btu/lbm) (Btu/sec) (sa ft)

) 0.0 0.0 416/-/

416/*

0.0 0.0 00.000;L 873.6 363,418 0/181 0.02,1 873.6 363,418 .081.

S 0 . 0Q,2 1310.3 /4'16 545,085 0.2715 1*1.110

.ill0 1310.3 416 545,085 0.2715 1259.1 416 523,786 0.2609 1259.1 416 523,786 // 0.2609 4.513 1.514 771.2 416 320,820 / 0.1598

,, 1.888 771.2 416 320,820 0.1598 1.889 385.6 416 160, 0 0.0799 5.997 385.6 416 160,/410 0.0799 5.998 841 416 3/49,981 0.0799 9.442 416 /*49, 981 0 0799 9.443 88 74,035 0.0799 16.657 %41.3 88 74, 035 0.0794ý K 16.658 202.2 88 17,794 0 .0192 23.595 202.2 88 17,794 0.0192 25.157 0.0 0.0

__j r Y (c,("i C." -f-C, C&Fýý 1 of 1 August 1987

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 Calculated(')

DBA Break Conditions Peak Initial Conditions Break Pressure Volume Volume Temp. P essure Humidity  % Break Break Area Bre ak Difference No. (cu ft) (OF) (psia) (%) in ol. Line (sq ft) e (psid) 1 2,165.6 105 14.7 0 100 RWCU (See 21.18

/ Table

/i 6.2.-.

/

2,165.6 105 14.7 0 0 f 0/.

3 8,278.9 105 14.7 .7 0 0.0 4 1, 120,000(2) 105 14.7 0 0 0. 0

/'

6<--

rep("APvýr4 1t tA4df-

"2 )Nodal 1

peak pressure minus pressure in Node 4 (Pi-P 4 ) ,

t )Assumed value to maximize pressure differential across RWCU filter/demineralizer room.

1 of 1 August 1993 Revision 6

RBS USAR TABLE 6.2-29 SUBCOMPARTMENT NODAL DESCRIPTION 8-IN RWCU LINE BREAK RWCU FILTER/DEMINERALIZER ROOM Calculated" )

DBA Break Conditions Peak Initial Conditions Break Pressure Volume Volume Temp. Pressure Humidity  % Break Break Break Difference No. cu ft f__ _tsiaj 0M) in Vol. Line Type (psid) 1 2,163.2 105 14.7 0 100 RWCU DER 10.425 2 2,163.2 105 14.7 0 0 0.0 3 8,085.0 100 14.7 0 0 0.0 4 1,120,000(2) 90 14.7 0 0 0.0 6<--e

(" Maximum differential pressure across the RWCY Filter / Demineralizer rgom walls.

(2 ) Assumed value to maximize pressure differential across RWCU filter/demineralizer room.

IIoof] August 1993 Revi'zion 6

RBS USAR TABLE 6.2-30 SUBCOMPARTMENT VENT PATH DESCRIPTION 8-IN RWCU LINE BREAK RWCU FILTER/DEMINERALIZER ROOM I,

From Tp' )escription Vent KPath Vent No.

Vol.

Node No.

yol.

/Node No.

of Vent Path Flow (Chcoked/Unchoked)

Area (sq ft)

L/A (1I)

Friction Head Loss Coefficient Thick Edge Turning Grating Expansion Contraction Total 1 3 Unchoked 1.37 2.579 _/ - 0.996 0.497 1.493

- 0.989 0.1499 1.488 3 1 Unchoked 1.37. 2.579 Unchoked lj. 11 1.957 - 0.986 . 499 1.485 2 3 2

/ /'1.811 - 0.995 // 0.496 1.491 2 / 2 3 Unchoked 1.957 / -/

- 0.994/ 0.498 1.492 3 4 Unchoked 2.6 0.785

/ 0.498 1.491 4 3 Unchoked 2.6 0.785

- / '0.999 0.498 4 3 4 Unchok 1.6 2.338

- / 0.991 0.500 /I1.491 4 3 Unc ked 1.6 2.338 4

0.95 0.494 0.776 0.477 2.890 choked 31.5 0.871 0.687(1) 0.163 5 3 4 0.95 0.494 0.911 0.440 2.995 Unchoked 31.5 0. 871 0.694 (1) 0.170 5 4 3

.q

... k

")Includes losses due to grating and thick edged orifice.

1 of 1 August 1987

RBS USAR TABLE 6.2-30 SUBCOMPARTMENT VENT PATH DESCRIPTION 8-IN RWCU LINE BREAK RWCU FILTER/DEMINERALIZER ROOM 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) 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.

I oflI Aut~ust 1987

RBS USAR TABLE 6.2-31 BLOWDOWN-DATA 8-IN RWCU LINE BREAK RWCU FILTER/DEMINERALIZER ROOM

/ Blowdown Mass Blowdown Blowdown Energy Total 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 0.3 016 0.0006 "2378.4 88 2.0,93 x 10 0.3 016 0.0007 3567.6 88 3 4 x 105 0.4 524 0.0097 3567.6 88 .14 X l0' 0.4 524 0.0098 4756.8 88 /4.186 x ls 0.6 032

0. 1305., 4756.8 88 .' 4.186 x l05 0.6 032 0.1306' 2378.4 88 2.093 x 105 0.3 016 1.3925 2378.4 88 " 2.093 x 105 0.3 016 1.3926 592.1 88 5.21 x 104 0.0 7509 9.9675 592.1 88 5.21 x 104 0.0 7 5I 9.9676 592.1 196.8 1.165 x 105 0.0 15.2075 15.2076 592.1 592.1 196.$

303,..'5 1.165 x 105 0. 7'509 1.797 x 10S5 .0 '509 7

18.6275 592.1 303.5 1.797 x 10' 0.0 7'509

'509 18.6276 592.1 Y89.2- 2.304 x 105 0.0 7'509 18.8375 592.1 /389.2 2.304 x 10 0.0 7

'509 18.8376 592.1 7 472.0 2.795 x 1H 7 19.7275 592.1 472.0 2.795 x10 0.0 7'509 19.7276 447.4 453.6 2 .03 y/10 5 0.0 5 675/

22.8975 447.4 453.6 2.03//x I0' 0.0 567 22.8976 151.4 529.2 8.01' X 104 0.0 1.9 26.6655 151.4 529.2 8.012 X 10 4 0.0 1.92/

28.2275 0.0 - 0.0 0.0 60.0 0.0 0.0 0.0 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 Blowdown Mass Enthalpy Break Flow Rate (Btu/Ibm)

(sec) (lbm/sec)

Upstream Blowdown 0.0000 0 93 .9106 0.0001 2241.04 93 .9106 0.0007 2241.04 93.9106 0.0008 4482.09 93 .9106 1.0048 4482.09 93 .9106 1.0049 1061.53 93 .9106 6.0424 1061.53 93 .9106 6.0425 721.681 93 .9106 7.0940 721.681 93 .9106 7.0941 721.681 135.29

8. 9119 721.681 135.29 8.9120 721.681 149.192 9.7007 721.681 149.192 9.7008 463.649 211.803 11.6003 463.649 211.803 11.6004 418.496 263.453 13.7403 418.496 263.453 13.7404 372.771 305.284 13.8435 372.771 305.284 13.8436 372.771 359.662 16.1949 372.771 359.662 16.1950 372.771 388.65 17.9879 372.771 388.65 17.9880 266.015 331.346 18.5536 266.015 331.346 18.5537 266.015 409.861 22.9688 266.015 409.861 22 .9689 266.015 450.744 29.2812 266.015 450.744 29.9296 75.0972 531.441 32.9880 0 531.441 1 of 2 August 1987

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 2 of 2 August 1987

rq(,tu-,j FIGURE 6.2-50 NODALIZATION DIAGRAM 6-IN RWCU LINE BREAK RWCU HEAT EXCHANGER ROOM RIVER BEND STATION UPDATED SAFETY ANALYSIS REPORT a

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

0.0 0.5 1.0 1.5 2.0 2.5 3.0 TIME AFTER BREAK (SEC)

III NA-1 C(Afftte-FIGURE 6.2-51 NODAL PRESSURES 6-IN RWCU LINE BREAK RWCU HEAT EXCHANGER ROOM RIVER BEND STATION UPDATED SAFETY ANALYSIS REPORT

16.5 16.

.5 ........ .. ...

CL 15 u.o UJ 15 (II 14.5 14 1 2.5 0.0 0.5 1.0 1.5 2.0 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 2.5 a"RE 6.2-52 NODAL PRESSURE DIFFERENTIAL ddeiA 6-IN RWCU LINE BREAK

'2e4' 0., f-44te zL 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" EL-186'3" I EL-256'3" NODE I J6 NODE 2 RWCU FILTER / RWCU FILTER I 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

C , , , I r4 +.A I -14 e- ita 7ýý '-C -

FIGURE 6.2-54 NODAL PRESSURES 8-IN RWCU LINE BREAK RWCU FILTER/ DEMINERALIZATION ROOM RIVER BEND STATION UPDATED SAFETY ANALYSIS REPORT

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) fz'

(-- ,r-de 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

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 3B-i August 1987

RBS USAR q-> 1 APPENDIX 3B LIST OF TABLES Table Number Title 3B-1 HIGH-ENERGY LINE BREAKS AUXILIARY BUILDING, Deleted: NODE MODEL 3B-2 HIGH-ENERGY LINE BREAKS MAIN STEAM TUNNEL NODE MODEL 3B-3 SUBCOMPARTMENT NODAL DESCRIPTION AUXILIARY BUILDING , NODE MODEL 3B-4 SUBCOMPARTMENT NODAL DESCRIPTION MAIN STEAM TUNNEL NODE MODEL 3B-5 DELETED Deleted: SUBCOMPARTMENT VENT PATH DESCRIPTION ¶ AUXILIARY BUILIDNG - 20 3B-6 SUBCOMPARTMENT VENT PATH DESCRIPTION NODE MODEL MAIN STEAM TUNNEL NODE MODEL 3B-7 MASS AND ENERGY RELEASE 3-IN. RWCU DER IN AUXILIARY BUILDING Deleted: -NODE 10 3B-8 MASS AND ENERGY RELEASE 6-IN. RWCU DER IN AUXILIARY BUILDING Deleted: -NODE 6¶ 3B-9 MASS AND ENERGY RELEASE 4-IN. RCIC DER IN AUXILIARY BUILDING, Deleted: -NODE 2 3B-10 MASS AND ENERGY RELEASE 8-IN. RHR DER IN AUXILIARY BUILDING Deleted: -NODE 12 3B-11 MASS AND ENERGY RELEASE 24-IN. MAIN STEAM LINE DER IN STEAM TUNNEL-NODE 2 3B-12 MASS AND ENERGY RELEASE 24-IN. MAIN STEAM LINE SER IN STEAM TUNNEL-NODE 1 3B-13 MASS AND ENERGY RELEASE 8-IN. RCIC STEAM LINE DER IN STEAM TUNNEL-NODE 2 1+--

3B-ii August 1988

RBS USAR APPENDIX 3B LIST OF TABLES (Cont) 3B-14 MASS AND ENERGY RELEASE 8-IN. RCIC STEAM LINE SER IN STEAM TUNNEL-NODE 1 3B-15 PELETED Deleted: HEAT SINK SLAB DESCRIPTION -¶ 3B-16 HEAT SINK SLAB DESCRIPTION .AUXILIARY BUILDING - 20 NODE MODEL MAIN STEAM TUNNEL NODE MODEL I+-.

3B-iii August 1988

RBS USAR 0-41 APPENDIX 3B LIST OF FIGURES Figure Number Title 3B-1 THROUGH 3B-2!A 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" & 6" RWCU AND 4" RCIC)¶

¶ 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" & 6" RWCU AND 4" RCIC)¶

¶ 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" & 6" RWCU AND 4" RCIC)¶

¶ 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" & 6" RWCU AND 4" RCIC)¶

¶ 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" & 6" RWCU AND 4" RCIC)1

¶ 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" & 6" RWCU AND 4" RCIC)¶ I

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 Deleted: 3B-12. PRESSURE TRANSIENTS IN NODE ii¶

.AUXILIARY BUILDING - HIGH LIST OF FIGURES (Cont) ENERGY LINEI

-BREAK ANALYSIS (3" & 6" APPENDIX 3B RWCU AND 4" RCIC)¶

¶ 3B-12A. PRESSURE TRANSIENTS LIST OF FIGURES (Cont) IN NODE II¶

.AUXILIARY BUILDING - HIGH ENERGY LINE¶ I .BREAK ANALYSIS (8" RHR)¶

'3B-22 NODALIZATION DIAGRAM - MAIN STEAM ¶ 3B-13 .PRESSURE TRANSIENTS TUNNEL - 6 NODE MODEL IN NODE 12¶

.AUXILIARY BUILDING - HIGH 3B-23 PRESSURE TRANSIENTS IN NODE 1 ENERGY LINEI

.BREAK ANALYSIS (3" & 6" MAIN STEAM TUNNEL RWCU AND 4" RCIC)¶ HIGH ENERGY LINE BREAK ANALYSIS ¶ 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" & 6" RWCU AND 4" RCIC)¶

¶ 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" & 6" RWCU AND 4" RCIC)¶

¶ 3B-18A.PRESSURE TRANSIENTS IN NODE 17¶

.AUXILIARY BUILDING - HIGH ENERGY LINE¶ BREAK ANALYSIS (8" RHR)¶

• ¶ 3B-19. PRESSURE TRANSIENTS IN NODE 181 AUXILIARY BUILDING - HIGH ENERGY LINE¶ BREAK ANALYSIS (3" & 6" RWCU AND 4" RCIC)¶ I

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 1<--o 3B-vi August 1988

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<---

3B-viii August 1988

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 Deleted: for the all postulated breaks,. The accident prififes were generated to 20 subcompartments bound the most limiting pressure responses. Deleted: by two of the Deleted: and identifies the The main steam tunnel analysis considered feedwater, RCIC, and two breaks that determined main steam line breaks. Main steam line break analyses were the design differential performed assuming a two-phase blowdown. Four combinations of pressures 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 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 to maximize differential pressures across node boundaries. Structural components were selected as node boundaries. The differential Deleted: ¶ pressure transients across node boundaries are used to determine Fig. 3B-1 shows the the structural adequacy and component support design. nodalization scheme used in the auxiliary building analysis and identifies the Table 3B-3 provides the nodal descriptions and gives the peak node numbers referred to in calculated and design differential pressures within the auxiliary the remainder of this section. Fig. 3B-22 building. Table 3B-4 similarly shows the subcompartment nodal similarly shows the descriptions for the main steam tunnel and identifies the nodalization scheme for the main steam tunnel.¶ calculated and design peak differential pressures. Figure 3B-22 shows the nodalization scheme for the main steam tunnel. Table Deleted: Table 3B-5 gives 3B-6 presents the vent path description corresponding to that vent flow path data for the auxiliary building shown on Fig. 3B-22 for the main steam tunnel. corresponding to the nodalization scheme shown on In calculating the pressure differentials across the auxiliary Fig. 3B-1. Table 3B-6 presents the vent path building subcompartment walls, it is possible to take credit for description corresponding to the pressurization of the volume on the opposite side of the wall that shown on Fig. 3B-22 for in question. This procedure, however, leads to slightly the main steam tunnel.¶ different pressure differentials for all walls of the Deleted: frictionless subcompartment in question. To minimize the number of Deleted: For the 4 -in RCIC differential pressures to be considered and for conservatism, a line break, partial credit single differential pressure was calculated for each volume by was taken for the effect of friction on reducing the subtracting 14.7 psia from each of the calculated nodal absolute "rate of blowdown.

pressures. ,Considering only the 4-in diameter portion of the RCIC steam supply line, -the total Peak pressure values for the main steam tunnel subcompartments loss coefficient for the also were calculated by subtracting 14.7 psia from the peak fittings and straight pipe pressure values. 'was determined to be K=5.

In this case, frictional Moody flow141 with fL/D=5 is

,Tables 3B-7 through 3B-10 provide the mass and differential energy release assumed and yields the data for the breaks that determine the design blowdown time history given in Table 3B-9.1 pressures within the auxiliary building. ¶ For the 8-in RHR line break, credit was also taken for In general, ý4oody"i) or Henry-Fauske"2 I flow was assumed (for friction. Considering saturated and subcooled flows, respectively) at the limiting 'piping from the main steam downstream and upstream flow areas crediting friction. During line to the break and choked flow at the break, the total the inventory period, the mass and energy release data were loss coefficient was calculated using the methodology of NEDO-20533"3 ' , except that calculated to be K = 5.41.

the Henry-Fauske model was used to calculate subcooled flow. 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

• +-I 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 Deleted: The UCHIDA heat transfer coefficient was were modeled similarly in the steam tunnel 6-node model, except applied, and condensate that the concrete slabs were assumed to be 1-ft thick, based on revaporization was assumed actual slabs which are 4-ft thick. Table 3B-16 summarizes these to be limited to 8 percent.

The heat sink slabs for the heat slabs. auxiliary building 20-node model are defined in The initial conditions in each node were assumed to be the Table 3B-15.¶

-*-.. - S-- Break (Connuous)-.---.---

maximum normal temperature, 14.7-psia pressure, and maximum relative humidity based on the Environmental Design Criteria Deleted* 100-percent (EDC).

,Fig. 3B-23 through 3B-26 provide the absolute pressure transient Deleted: I plots for the two main steam tunnel subcompartments within the Fig. 33-2 through 3B-21A

.provide the absolute auxiliary building portion of the tunnel. pressure transient plots for the 20 subcompartments in the auxiliary building.¶

,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: I Deleted: I

• <--1 3B-6 August 1988

• +-I 3B-5 August 1988 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

• +-I 3B-6 August 1988

Page 6: [1] Deleted Unknown 3B-12 PRESSURE TRANSIENTS IN NODE 11 AUXILIARY BUILDING - HIGH ENERGY LINE BREAK ANALYSIS (3" & 6" RWCU AND 4" RCIC) 3B-12A PRESSURE TRANSIENTS IN NODE 11 AUXILIARY BUILDING - HIGH ENERGY LINE BREAK ANALYSIS (8" RHR) 3B-13 PRESSURE TRANSIENTS IN NODE 12 AUXILIARY BUILDING - HIGH ENERGY LINE BREAK ANALYSIS (3" & 6" RWCU AND 4" RCIC) 3B-13A PRESSURE TRANSIENTS IN NODE 12 AUXILIARY BUILDING - HIGH ENERGY LINE BREAK ANALYSIS (8" RHR) 3B-14 PRESSURE TRANSIENTS IN NODE 13 AUXILIARY BUILDING - HIGH ENERGY LINE BREAK ANALYSIS (3" & 6" RWCU AND 4" RCIC) 3B-14A PRESSURE TRANSIENTS IN NODE 13 AUXILIARY BUILDING - HIGH ENERGY LINE BREAK ANALYSIS (8" RHR) 3B-15 PRESSURE TRANSIENTS IN NODE 14 AUXILIARY BUILDING - HIGH ENERGY LINE BREAK ANALYSIS (3" & 6" RWCU AND 4" RCIC) 3B-15A PRESSURE TRANSIENTS IN NODE 14 AUXILIARY BUILDING - HIGH ENERGY LINE BREAK ANALYSIS (8" RHR) 3B-16 PRESSURE TRANSIENTS IN NODE 15 AUXILIARY BUILDING -HIGH ENERGY LINE BREAK ANALYSIS (3" & 6" RWCU AND 4" RCIC) 3B-16A PRESSURE TRANSIENTS IN NODE 15 AUXILIARY BUILDING - HIGH ENERGY LINE BREAK ANALYSIS (8" RHR) 3B-17 PRESSURE TRANSIENTS IN NODE 16 AUXILIARY BUILDING - HIGH ENERGY LINE BREAK ANALYSIS (3" & 6" RWCU AND 4" RCIC)

Page 6: [2] Deleted Unknown 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 LINE BREAK ANALYSIS (3" & 6" RWCU AND 4" RCIC) 3B-18A PRESSURE TRANSIENTS IN NODE 17

AUXILIARY BUILDING - HIGH ENERGY LINE BREAK ANALYSIS (8" RHR) 3B-19 PRESSURE TRANSIENTS IN NODE 18 AUXILIARY BUILDING - HIGH ENERGY LINE BREAK ANALYSIS (311 & 6" RWCU AND 4" RCIC) 3B-19A PRESSURE TRANSIENTS IN NODE 18 AUXILIARY BUILDING - HIGH ENERGY LINE BREAK ANALYSIS (8"RHR) 3B-20 PRESSURE TRANSIENTS IN NODE 19 AUXILIARY BUILDING - HIGH ENERGY LINE BREAK ANALYSIS (3" & 6" RWCU AND 4" RCIC) 3B-20A PRESSURE TRANSIENTS IN NODE 19 AUXILIARY BUILDING - HIGH ENERGY LINE BREAK ANALYSIS (8" RHR) 3B-21 PRESSURE TRANSIENTS IN NODE 20 AUXILIARY BUILDING - HIGH ENERGY LINE BREAK ANALYSIS (3" & 6" RWCU AND 4" RCIC) 3B-21A PRESSURE TRANSIENTS IN NODE 20 AUXILIARY BUILDING - HIGH ENERGY LINE BREAK ANALYSIS (8" RHR)

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 (4

assumed n ) t reage do era s e to d be sig double-ended

.ý ress re ruptures.

or ny ode 2

( ) Subcompartment nodes are defined in Table 3B-3 and on Fig. 3B-l.

3

( )This break also could occur in NodIe 9. Consequently, the results for Node 10 are applied to Node 9 considering symmetry.

1 of 1 August 1988

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.

1 of 1 August 1988

RBS USAR TABLE 3B-3 SUBCOMPARTMENT NODAL DESCRIPTION AUXILIARY BUILDING 20-NODE MODEL Absolute Calculated Design Peak Net Peak Peak Pressure Differential Node Volume / Break Break Break Pressure Differential 121 Pressure Number (ft 3) Description of Volume Location Type Line111 (psia) (psid) (sid) s_

1 9,685 RHR 'C' Equipment Room, Node 12 Steam 8" RHR 17.03 2.33 2.40 EDC Zone AB-070-4 2 12,524 RCIC Pump Room, Node 12 Steam 8" RHR 17.03 23 2.40 "EDCZone AB-070-3 3 22,845 RPCCW Equipment Area, Node 12 Steam 8" RHR 17.03 .32 2.40 17.02 2./32 2.40

/ EDC Zone AB-095-8 4 1,181/ East-West Passageway, Node 12 Steam 8" RHR 17.02 2.32 2.40

/ EDC Zone AB-095-4 5 4?80 Unit Cooler Area, Node 12 Steam 8" RHR 17.02 2.32 2.40 EDC Zone AB-095-4 RCIC Access Area, Node 12 Steam 8" RHR 17.02 2.32 2.40 6/ 453

/ EDC Zone AB-095-4 Node 12 Steam 8" RHR 17.02 2.32 2.40 7 2,535 Hoist Area, EDC Zone AB-095-4 /

17.02 23 - 2.40 8 21,864 Elevator Area, Node 12 Steam 8" RHR EDC Zone AB-095 627 RWCU 'A' PumROom, Node 9(3) Liquid 3" RWCU 17.94 3.24 3.30 9

EDC Zone AH/095-3

//

Node 10 Liquid 3" RWCU 17.94 3.24 3.30 10 627 RWCU 'B'-Pump Room, EDC Zo/e AB-095-3

'-+1 /

Steam 8" RHR 17.03 2.33 2.4 11 71,439 RPC"W Equipment Area, Node 12 EDO Zone AB-070-8

/

Node J2 Steam 8" RHR 17.01 2.31 / 2.40 12 86,570 / CC Area (East),

/ EDC Zones AB-1 14-3,5,

/ and 8B /1.

17.01 2.31 2.40 13 90,157 / MCC Area(West), Node 12 Steam 8" RHR EDC Zones AB-1 14-1,6, and 8A Veý(61614 W"Pý n-Z4'j 1 of 2 August 1988

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 3 Location Line___ (psia) (psid) (psid)

Number (ft ) Description of Volume Type 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 S/ EDC Zone AB-1 70-1 / /

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~~reakin 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 Description of Volume Vol. (ft3) Absolute Peak Pressure Calculated Peak Diff.

(psia) Pressure(1) (2)

(psid)

AB-070-1 CSL Area 13992 16.49 1.79 AB-070-2 RHS-P1A Pump Room 22733 16.49 1.79 AB-070-3 ICS Pump Room 12524 16.49 1.79 AB-070-4 RHS-P1C Pump Room 9685 16.48 1.78 AB-070-5 RHS-P1IB Pump Room 22733 16.48 1.78 AB-070-6 HPCS Pump Room 13927 16.48 1.78 AB-070-7 Elevator Area 35720 16.48 1.78 AB-070-8 RPCCW Area 35720 16.49 1.79 AB-095-1 CSL Hatch Area 11548 16.48 1.78 RHS Heat Exchanger Area (West) 16402 16.48 1.78 AB-095-2 WCS Area 1567 16.48 1.78 AB-095-3 12614 16.47 1.77 AB-095-4 Hoist Area (Sub-Volume #1) 2535 16.47 1.77 AB-095-4 Hoist Area (Sub-Volume #2) 16402 16.47 1.77 AB-095-5 RHS Heat Exchanger Area (East)

HPCS Hatch Area 22734 16.47 1.77 AB-095-6 21864 16.47 1.77 AB-095-7 Elevator Area 22845 16.48 1.78 AB-095-8 RPCCW Area 55573 16.47 1.77 AB-114-1 & 8A MCC Area and RHR Equipment Removal Cubicle (west) 26775 14.70 0.00 AB-1 14-2 Main Steam Tunnel (North) 30381 16.46 1.76 AB-1 14-3 MCC Area (East) 1945 16.46 1.76 AB-1 14-4 Post Accident Sampling Station 31873 16.46 1.76 AB-1 14-5 Elevator Room 34584 16.47 1.77 AB-1 14-6 RPCCW Area 24613 16.46 1.76 AB-1 14-8B RHR Equipment Removal Cubicle (East)

I of 2 August 1988

RBS USAR TABLE 3B-3 SUBCOMPARTMENT NODAL DESCRIPTION AUXILIARY BUILDING EDC Zone Description of Volume Vol. (ft3) Absolute Peak Pressure Calculated Peak Diff.

(psia) Pressure (1 (2)

(psid)

AB-141-1 Equipment Area (West) 62074 16.45 1.75 AB-141-2 Equipment Area (East) 70772 16.45 1.75 AB-141-3 Elevator Area 39813 16.45 1.75 AB-141-4 RPCCW Area 40273 16.45 1.75 AB-141-5 Standby Gas Treatment Filter (West) 45330 16.45 1.75 AB-141-6 Standby Gas Treatment Filter (East) 42256 16.45 1.75 AB-1 70-1 Annulus Mixing System Fan Area 12172 16.44 1.74 (Sub-Volume #1)

AB-1 70-1 Annulus Mixing System Fan Area 3336 16.44 1.74 (Sub-Volume #2)

AB-1 70-1 Annulus Mixing System Fan Area 1930 16.43 1.73 (Sub-Volume #3)

AB-1 70-2 Continuous Filter Room 9962 16.44 1.74 Elevator Machine Room 1313 16.43 1.73 AB- 170-3.

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.

2 of 2 August 1988

TABLE 3B-5 SUBCOI'MPARfMENT VENT PATHf-ý RI-PTION-AUXILIARY BUILDING

( 20-NODE MODEL K<

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 /

J()/5 7 21.0 0.'0/8

• /50 0.442 0.781 , 0.906 5 7 3.0 ,0.779 5 6 21.0 0.494 5 6 3. Q 3.223 Jl0 4 6 0.169 Jill 3 4 10.75 0 .5883 J12 7 105.0 0.091

/2 J13 6 114.75 0.059

/

J14(4) 2 1 23 .88 0 .104 3.: 11 271.3 0. 0o/9 8 11 115.0 .032 13 272.43 0.018 3

8 12 115.0 0.031 1 of 2 August 1987

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 2.473 203.2 0.0563 0.282 0.012 - - 0.294 J21 i 16

/,"17 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 .0652 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, 2 orifice, mesh door, and any other form loss blocking the vent path.

(2) Closed door (wit 3.0-ft. 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.

2 of 2 August 1987

RBS USAR TABLE 3b-7

-qAND ENERGY RELEASE I-NAB DE IN AUXILIARY BUILDING - NODE 10 Total Mass Total Enthalpy Time Flow Rate F w Rate (sec) (lbm/sec) (-tu/sec)

/

0-0 0/0 0./01 522.4 277,000 2 . 120 522.4 277,000

/

/

/

2.121 494.3 262,000

/

4.150 494.3 262,000 4 .151 354.5 188,000

'I

6. 940 354.5 188,000 8.000 255.0 135,000 /

8.500 I 208.5 110, 00~t 19.810 / 208.5 / 110o/.000 22 . 0001 I

0 .0/

/

/i I

-I Af& kA.

1 of 1 August 1987

RBS USAR TABLE 3B-7 MASS AND ENERGY RELEASE 3-IN RWCU DER IN AUXILIARY BUILDING Time (sec) Total Mass Flow Total Energy Flow Rate (lbm/sec) Rate (Btu/sec) 0.000 0.0 0 0.001 357.8 198956 1.900 357.8 198956 2.000 336.9 187344 3.800 336.9 187344 3.900 275.3 153069 12.000 275.2 153059 13.200 275.2 153053 13.700 258.1 143550 14.000 228.3 126940 14.300 211.4 117534 14.400 183.7 102146 14.500 164 .2 91283 14.700 158 .2 87974 15.000 117.2 65169 19.500 117.0 65060 24.300 115. 8 64415 25.300 114 .2 63495 27.300 79.4 44135 27.400 46.9 26106 27.900 40.3 22407 28.300 0.0 0 1 of 1 August 1987

RBS USAR TABLE 3b-8 cd, 1 of I August 1987

RBS USAR TABLE 3B-8 MASS AND ENERGY RELEASE 6-IN RWCU DER IN AUXILIARY BUILDING Time (sec) Total Mass Flow Total Energy Rate (lbm/sec) Flow Rate (Btu/sec) 0.000 0.0 0 0.001 1411.9 785159 0.900 1411.9 785159 1.000 706.0 392579

1. 100 165.2 91845 26.500 165.2 91845 28.300 0.0 0 1 of 1. August 1987

RBS USAR TABLE 3b-9 SS S 4D ENERGYF BUILDIN XTLIARY

/IN A

// 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 1 of i. August 1987

RBS USAR TABLE 3B-9 MASS AND ENERGY RELEASE 4-IN RCIC DER IN AUXILIARY BUILDING Time (sec) Total Mass Flow Total Energy Flow Rate (lbm/sec) Rate (Btu/sec) 0.000 0.0 0 0.001 134.8 160373 0.245 134.8 160373 0.250 72.9 86666 12.000 72. 9 86666 13.000 72. 8 86636 14.000 72. 8 86571 15.000 72. 7 86461 16.000 72.5 86260 17.000 72.2 85861 18.000 71.6 85166 19.000 70.2 83530 19.900 67.6 80442 20.900 55.2 65663 21.000 44.1 52481 21.100 37.6 44733 21.200 33.8 40242 21.400 28.8 34273 21.500 3.1 3737 21.900 0.0 0 1.0f 1 August 1987

RBS USAR 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 0.0 287,845

.001 241.4 287,845 2.0 241.4 273,56 4.5 229.4 7.0 209.1 331

,24 9.5 138.! 864,551 12.0 0.0 1 of I August 1988

RBS USAR TABLE 3B-10 MASS AND ENERGY RELEASE 8-IN RHR DER IN AUXILIARY BUILDING Time (sec) Total Mass Flow Total Energy Flow Rate (lbm/sec) Rate (Btu/sec) 0.000 0.0 0 0.001 509.4 605951 0.100 509.4 605951 0.200 266.8 317333 1.900 266.8 317333 2.500 266.0 316362 3.000 265.1 315348 3.500 263.7 313613 4.000 262 .1 311740 4.500 259.9 309137 5.000 256.6 305216 5.500 252.2 300010 6.000 246.6 293348 6.500 237.9 282938 7.000 227.4 270446 7.500 212.8 253096 7.800 204.0 242686 7 .900 198.8 236440 8.000 192.0 228343 8.200 181.2 215506 8.300 179.4 213416 8 .500 175.9 209237 8.600 174.2 207148 8 .900 167.3 198999 9.800 132.2 157301 9.900 125.7 149558 10 .000 116.9 139004 10 .200 102.9 122340 10 .800 72.8 86547 10 .900 53.5 63645 11.000 43.0 51106 11.300 28.8 34273 11.400 3.1 3737 11.800 0.0 0 1 of 1 August 1988

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 3

5,076 Concrete( ) 1.0 {I(e-3 664 Concretet ) 1.0 3

1,750 Concrete{ 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 3

17,880 Concrete( ) 1.0 /

3 38,632 Concrete( ) 1.0 1,106 Concrete(3) 1 3 35 Concrete(') 1.0 3

1 830 Concrete( ) 1.0 1,287 Concrete 1.0 1,662 Concrete 1.0 2,209 Concrete 1.0 3

743 Carb oSteelt ) 0.0306 3

935 C4rbon Steel( ) 0.0307 1 of 2Auittst 1988

RBS USAR TABLE 3B-15 (Cont) 1)Exposed Slab Node Exposure(l) Surface Thickness No. Left Right Area Mf2) Material 21 / 3 3 1,875 Carbon Steel(3 ) 0.05 22/ 5 . 5 278 Carbon Steel(3) 0.0403 3 6 6 476 Carbon Steel(3) 0.0439 24 8 8 1,214 Carbon Steel(3) 0.06 10 25 11 11 7,344 Carbon Steel(3) 0.0460 (<

3 26 16 / 16 1,894 Carbon Steel0 ) 0.04 C&

27 17 / 17 10/8 Carbon Steel(-) 0.04 t

28(4) 12 / 0 /1,798 Carbon Steel ,). 0.00529 29(4) 1 0 982.5 Carbon,_teel(3) 0.00529 3

30(4) 14 0 2,943 Cýrbon Steel( ) 0.00529

//

('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:

Concrete Carbon Steel 3

Conductivity, tu/hr-°F-ft , 0.8 26.0 3

Volumetric I eat capacity, Btu/'F-ft 23.2 53.9 (4) Heat sinks only applicable to 8" RHR HELB analysis.

"2 of 2 August 1988

'I, S

/

J Q.._d*

NOTE:

• THE NOTATION 070-3 FOR EXAMPLE, REFERS TO ENVIRONMENTAL ZONE 3 ON ELEVATION 70'0" OF THE /

AUXILIARY BUILDING

FIGURE 3B-2A PRESSURE TRANSIENTS IN NODE 1

13132 ,I311ir3 AUXILIARY BUILDING HIGH ENERGY LINE BREAK ANALYSIS RIVER BEND STATION UPDATED SAFETY ANALYSIS REPORT

17.0 16.5 RWCU 16.016" 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