ML18190A522
| ML18190A522 | |
| Person / Time | |
|---|---|
| Site: | PROJ0769, NuScale |
| Issue date: | 07/09/2018 |
| From: | Rad Z NuScale |
| To: | Document Control Desk, Office of New Reactors |
| Shared Package | |
| ML18190A521 | List: |
| References | |
| AF-0718-60794, RAIO-0718-60793 | |
| Download: ML18190A522 (113) | |
Text
RAIO-0718-60793 July 09, 2018 Docket: PROJ0769 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk One White Flint North 11555 Rockville Pike Rockville, MD 20852-2738
SUBJECT:
NuScale Power, LLC Response to NRC Request for Additional Information No.
9374 (eRAI No. 9374) on the NuScale Topical Report, "Non-Loss of Coolant Accident Analysis Methodology," TR-0516-49416, Revision 1
REFERENCES:
- 1. U.S. Nuclear Regulatory Commission, "Request for Additional Information No. 9374 (eRAI No. 9374)," dated May 09, 2018
- 2. NuScale Topical Report, "Non-Loss of Coolant Accident Analysis Methodology," TR-0516-49416, Revision 1, dated August 2017 The purpose of this letter is to provide the NuScale Power, LLC (NuScale) response to the referenced NRC Request for Additional Information (RAI).
The Enclosures to this letter contain NuScale's response to the following RAI Questions from NRC eRAI No. 9374:
15.00.02-25 15.00.02-26 15.00.02-27 15.00.02-29 15.00.02-30 The response schedule for the remaining questions of RAI No. 9374, eRAI 9374 were provided in emails to NRC (Greg Cranston) dated June 19 and July 9, 2018. is the proprietary version of the NuScale Response to NRC RAI No. 9374 (eRAI No. 9374). NuScale requests that the proprietary version be withheld from public disclosure in accordance with the requirements of 10 CFR § 2.390. The enclosed affidavit (Enclosure 3) supports this request. Enclosure 2 is the nonproprietary version of the NuScale response.
This letter and the enclosed responses make no new regulatory commitments and no revisions to any existing regulatory commitments.
NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvalis, Oregon 97330, Office: 541.360.0500, Fax: 541.207.3928 www.nuscalepower.com
RAIO-0718-60793 If you have any questions on this response, please contact Paul Infanger at 541-452-7351 or at pinfanger@nuscalepower.com.
Sincerely, Zackary Za Z ckary W. Rad Director, Regulatory Affairs NuScale Power, LLC Distribution: Gregory Cranston, NRC, OWFN-8G9A Samuel Lee, NRC, OWFN-8G9A Rani Franovich, NRC, OWFN-8G9A : NuScale Response to NRC Request for Additional Information eRAI No. 9374, proprietary : NuScale Response to NRC Request for Additional Information eRAI No. 9374, nonproprietary : Affidavit of Zackary W. Rad, AF-0718-60794 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvalis, Oregon 97330, Office: 541.360.0500, Fax: 541.207.3928 www.nuscalepower.com
NuScale Response to NRC Request for Additional Information eRAI No. 9374, proprietary NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvalis, Oregon 97330, Office: 541.360.0500, Fax: 541.207.3928 www.nuscalepower.com
NuScale Response to NRC Request for Additional Information eRAI No. 9374, nonproprietary NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvalis, Oregon 97330, Office: 541.360.0500, Fax: 541.207.3928 www.nuscalepower.com
Response to Request for Additional Information Docket: PROJ0769 eRAI No.: 9374 Date of RAI Issue: 05/09/2018 NRC Question No.: 15.00.02-25 TR-0516-49416-P supports the conclusions in the NuScale FSAR, which under 10 CFR 52.47 must describe the facility, present the design bases and the limits on its operation, and present a safety analysis of the structures, systems, and components and of the facility as a whole. SRP Section 15.0.2 provides the staff guidance on reviewing analytical models and computer codes used to analyze transient and accident behavior. SRP Section 15.0.2 states that the EM documentation must include a user manual that provides clear and unambiguous guidance, including:
detailed instructions about how the computer code is used, a description of how to choose model input parameters and appropriate code options, guidance about code limitations and options that should be avoided for particular accidents, components, or reactor types, and documented procedures for ensuring complete and accurate transfer of information between different elements of the EM.
As discussed during the non-LOCA audit, NuScale-specific modeling guidance for non-LOCA events is provided in document SwUM-0304-15495, NRELAP5 Version 1.3 Input Data Requirements, and relevant parts of the non-LOCA and LOCA topical reports. TR-0516-49416-P, Table 7-1 provides a typical list of initial conditions that are applicable for non-LOCA transient analysis. The list includes a number of parameters that are either redundant or would result in an over-specification of initial steady-state conditions for the plant. This over- specification may cause unacceptable results (i.e., failure to achieve a unique converged steady-state solution).
Examples include:
Volume-weighted core average fuel temperature - Fuel temperatures are calculated from the power distribution and the state of the reactor coolant system. The volume- weighted core average fuel temperature is a calculated outcome, not an initial condition.
RCS fluid temperature - Given the pressurizer pressure, feedwater flow, feedwater temperature, and reactor power level, the RCS fluid temperature is a calculated outcome, not an initial condition.
Turbine governor valve flow rate - Given the feedwater flow rate, steam NuScale Nonproprietary
generator level, and steam header pressure, the steam flow to the turbine is a calculated outcome, not an initial condition.
During an audit discussion (Round 2, Issue 8), the applicant stated that Table 7-1 is intended to list all parameters that could be set for an analysis and that an analyst would identify items to cut from the list. However, TR-0516-49416-P is not clear about the purpose of Table 7-1 and does not provide clear and unambiguous guidance regarding initial condition specification.
Information Requested:
Clearly state the purpose of Table 7-1 in TR-0516-49416-P as discussed during the audit, and revise Table 7-1 and the associated text in TR Section 7.1.1.2 accordingly. The discussions should include the specific constraints under nominal operating conditions and the technical basis to demonstrate that (a) the specified constraints would not result in an over- or under-specification, (b) changes to specified conditions to bias the calculations also fulfill the conditions under part (a), and (c) the specified constraints result in unique initial conditions.
NuScale Response:
The following parameters, that could lead to over-specification if used without further reflection, were removed from Table 7-1.
Feedwater pressure (redundant with turbine header pressure)
Main steam pressure (redundant with turbine header pressure)
Turbine governor valve flow rate (redundant with main feedwater flow rate)
To further clarify the intent of Table 7-1, the initial conditions were grouped according to type:
direct input as opposed to calculated result. Most of the listed parameters are directly input to NRELAP5. Other parameters listed in the table are identified as a "target" because the analyst targets an acceptable value for the designated parameter. Table 7-1 was revised to indicate the target parameters.
Two additional parameters, that would otherwise be classified as targets (steam generator inventory and steam superheat), were removed from Table 7-1 for the reasons noted below.
Steam Generator Inventory Under best-estimate primary flow conditions and various reactor powers, it has been shown that the steam generator inventory is ((2(a),(c) NuScale Nonproprietary
((
}}2(a),(c).
Consequently, steam generator inventory is not a target during the initialization process. ((
}}2(a),(c)
Steam Superheat For reactor power greater than or equal to 20 percent full power and best-estimate primary flow conditions, the steam superheat increases with power for the defined main feedwater conditions. ((
}}2(a),(c)
((
}}2(a),(c)
Those parameters identified as targets in Table 7-1 are: Volume-weighted core average fuel temperature; Primary system mass flow rate; and, RCS average fluid temperature. Volume-Weighted Core Average Fuel Temperature As described in Section 4.3.1.2.3, the conservatively high or low volume-weighted core average NuScale Nonproprietary
fuel temperature ((
}}2(a),(c). So, even though the calculated fuel temperature is a function of the power distribution and the state of the reactor coolant system; with the ((
}}2(a),(c) is retained as a target.
As noted above, ((
}}2(a),(c)
Primary System Mass Flow Rate For the NPM, which does not employ pumps for circulation of the primary fluid, the primary system flow rate is a calculated value. The primary system flow rate is a target value that is a ((
}}2(a),(c)
RCS Average Fluid Temperature As described in Section 6.3.1.3, the primary fluid temperature for the NPM is based on ((
}}2(a),(c) In this instance, the RCS average fluid temperature is constant for reactor powers greater than or equal to 15 percent full power. In contrast, the RCS average fluid temperature increases with power for reactor powers less than 15 percent full power. Given that the ((
}}2(a),(c), the RCS average fluid temperature is retained as a target.
NuScale Nonproprietary
The inputs required ((
}}2(a),(c). In the absence of any core bypass flow, the same T would exist for the primary side of the tubes when that energy is transferred through the steam generator tubes to the secondary fluid. Completing the transfer of that energy to the secondary fluid also requires certain secondary system conditions to exist. Thus, it is also necessary to have inputs ((
}}2(a),(c)
A typical recirculating steam generator is designed to operate with the tubes submerged at all reactor powers; thereby limiting the heat transfer modes present. Hence, the secondary water level is used as a target during the initialization process. On the other hand, the helical coil steam generator used for the NPM is considered a once-through steam generator. As such, the secondary water level only partially covers the tubes, resulting in heat transfer modes ranging from subcooled boiling to single-phase steam. Although a collapsed water level exists on the secondary at all reactor powers, ((
}}2(a),(c) can be used for model initialization without over-specifying the problem. With respect to having a ((
}}2(a),(c) Thus, the parameters in Table 7-1 can also be used for model initialization without under-specifying the problem.
The ability of NRELAP5 to accurately predict the range of heat transfer modes for the helical coil steam generator is demonstrated in Section 5.3.5.4. Specifically, the comparisons of the ((
}}2(a),(c) Consequently, the accuracy of the NRELAP5 predictions extends to the primary and secondary conditions discussed previously.
Section 7.1.1.2 was modified as indicate at the end of this response to incorporate the information requested. NuScale Nonproprietary
Impact on Topical Report: Topical Report TR-0516-49416, Non-Loss of Coolant Accident Analysis Methodology, has been revised as described in the response above and as shown in the markup provided in this response. NuScale Nonproprietary
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Rev. 12 7.0 Non-LOCA Analysis Methodology 7.1 General The discussion in the following sub-sections is applicable for all transients unless the event-specific methodology states otherwise in Section 7.2. 7.1.1 Achieving Steady State Conditions This section identifies the initial and boundary conditions considered for biasing in non-LOCA analyses, including prioritization during the initialization process. While the majority of parameters identified herein are initial conditions relevant to the steady state, other parameters are considered as bounding input for the plant response during the transient progression. 7.1.1.1 Background Establishing the appropriate initial conditions is paramount to obtaining an appropriate plant response for the transient of interest. To this end, it is important to ensure the NRELAP5 model achieves a valid steady state prior to initiating the transient. The effect of various initial conditions on the response to a specific acceptance criterion is assessed for each non-LOCA transient. Several means are available to perform the assessment. For example, the assessment may consist of a combination of:
- Qualitative engineering assessment in the calculation to identify why an initial condition bias is limiting or non-limiting
- Quantitative assessment by execution of appropriate sensitivity calculations in the calculation
- Reference to applicable regulatory precedent that identifies why a particular initial condition bias is not limiting for a particular transient or type of transient, or why a nominal condition is appropriate 7.1.1.2 Identification of Relevant Parameters A list of initial conditions to be considered for biasing was developed for the non-LOCA analyses, including the target value for the initial plant condition and acceptable tolerance to the target value.
Other parameters needed to obtain a steady state include: certain fuel-related and core-related inputs; measurement uncertainties for various safety-related processes; the plant operational limits assumed in the safety analyses; and the DHRS initial conditions. Table 7-1 identifies a typical list of initial conditions considered for non-LOCA analyses. This list of initial conditions includes both direct inputs and calculated results. Most of these parameters are directly input to NRELAP5. The calculated results are identified as © Copyright 20187 by NuScale Power, LLC 270
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Rev. 12 a target because the analyst uses this parameter as a target during the initialization process. Under best-estimate primary flow conditions and various reactor powers, it has been shown that there is relatively little variation in liquid inventory in the steam generator. The inventory does not change appreciably when considering variations in primary coolant temperature and primary flow rates. Consequently, steam generator inventory is not a target during the initialization process. ((
}}2(a),(c)
For reactor powers greater than or equal to 20 percent full power (FP) and best-estimate primary flow conditions, the steam superheat increases with power for the defined main feedwater conditions. ((
}}2(a),(c)
The removal of any restrictions regarding steam generator inventory and steam superheat allows use of the parameters in Table 7-1 for initialization without over-specifying or under-specifying the problem. © Copyright 20187 by NuScale Power, LLC 271
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Rev. 12 Table 7-1 Typical list of initial conditions considered Region Parameter Power Calorimetric power uncertainty Volume-weighted core average fuel temperature (target) Moderator temperature coefficient Doppler temperature coefficient Effective delayed neutron fraction Ratio of effective delayed neutron fraction to prompt neutron lifetime Core 238 U neutron capture to fission ratio Energy deposition factor Average core axial power shape Scram reactivity (shutdown margin) Control rod bank differential worth Decay heat and decay heat uncertainty Boron concentration Primary system mass flow rate (target) RCS average fluid temperature (target) Pressurizer pressure Pressurizer level RCS Total core bypass flow rate Pressurizer spray bypass flow rate Pressurizer heater power Heat losses to containment from the RPV and from piping inside containment Steam generator secondary, Feedwater pressure feedwater system, and MSS Main steam pressure Feedwater mass flow rate Feedwater temperature Steam generator steam superheat Steam generator level (inventory is used as the surrogate) Steam chest pressure (turbine header pressure) Turbine governor valve flow rate Steam generator tube plugging, fouling factor Pressure Containment Heat losses from containment to reactor pool and reactor building Temperature Temperature Reactor pool Level DHRS Liquid volume © Copyright 20187 by NuScale Power, LLC 272
Response to Request for Additional Information Docket: PROJ0769 eRAI No.: 9374 Date of RAI Issue: 05/09/2018 NRC Question No.: 15.00.02-26 TR-0516-49416-P supports the conclusions in the NuScale FSAR, which under 10 CFR 52.47 must describe the facility, present the design bases and the limits on its operation, and present a safety analysis of the structures, systems, and components and of the facility as a whole. RG 1.203 describes the EMDAP, which the NRC staff considers acceptable for use in developing and assessing EMs used to analyze transient and accident behavior. Step 16 of the EMDAP describes, in part, determining the ability of numeric solutions to approximate equation sets: The numeric solution evaluation considers convergence, property conservation, and stability of code calculations to solve the original equations when applied to the target application. The objective of this evaluation is to summarize information regarding the domain of applicability of the numerical techniques and user options that may impact the accuracy, stability, and convergence features of each component code. TR-0516-49416-P, Section 7.1.1.4 describes the typical initialization process and states: Prior to the initialization process, certain parameters critical to establishing the correct steady state for the event of interest are identified. Once the parameters of interest achieve steady state target values within acceptable tolerances, on the basis of engineering judgement, ((
}}2(a),(c) A steady state solution has been achieved when the change in the target value during the loop transits is within the variance band described for each parameter.
During audit discussions, the applicant stated that there are two ways to achieve the biased conditions: 1) changing the control target values or 2) combinations of biasing variables to obtain a desired target value. The applicant also stated that part of the null transient evaluation is to examine the primary and secondary conditions to ensure that they are correct for the biased conditions. In addition, the nominal operating conditions are calculated for a given power level, as described in document EC-A030-2713, Primary and Secondary Steady State Parameters. However, based on the staffs audit of EC-A030-2713, the document did not NuScale Nonproprietary
demonstrate that the conditions shown for secondary side level, superheat, primary RCS flow, and RCS temperature correspond to the unique and correct steady-state conditions. If a unique and correct steady state is not achieved, the results of the transient calculation will be unreliable. Information Requested: Provide a document that describes the process of obtaining a unique and correct steady-state initialization for non-LOCA transients or update TR-0516-49416-P to include such information. The document or update should list the constraints imposed on the steady-state initialization (e.g., reactor power, inlet temperature, outlet temperature, feedwater flow rate, feedwater temperature, etc.). NuScale Response: The discussion in the response to this RAI, Question 15.00.02-25 provides the majority of the revisions made to clarify the initialization process for the NPM. However, those updates relate only to Section 7.1.1.2. Additional updates were required to Section 7.1.1.4 to: 1) denote that the parameters listed are those for the primary and secondary; 2) revise RPV Pressure to PZR Pressure; and, 3) add feedwater temperature to the list. As noted in the response to question 15.00.02-25, there are no restrictions ((
}}2(a),(c) completes the list of parameters critical to establishing the correct steady-state for the event of interest.
Impact on Topical Report: Topical Report TR-0516-49416, Non-Loss of Coolant Accident Analysis Methodology, has been revised as described in the response above and as shown in the markup provided in this response. NuScale Nonproprietary
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Rev. 12 7.1.1.3 Prioritization of Initial Conditions An important design feature of the NPM is the natural circulation of primary reactor coolant. This design feature may also limit the ability of the NRELAP5 non-LOCA transient model to achieve a given set of initial conditions. In particular, changes in core power or parameters that affect the SG secondary side heat removal rate will alter the reactor coolant flow rate and density distribution in the primary coolant. The initial conditions are prioritized based on the greatest impact to the transient of interest. Initial conditions may be treated in the NRELAP5 non-LOCA transient model in different ways. For instance, the initial conditions can be specified directly as input or calculated by the code. Parameters identified as important initial conditions for a particular transient (whether directly input or indirectly determined by the code) are checked as part of the steady state balance to confirm that either: 1) the parameter is within the allowable tolerance to the target value based on design references; or, 2) the parameter conservatively bounds the target design value and is adequately steady (within acceptable tolerances). If the initial condition is calculated by the code but not identified as important for a transient, the parameter may or may not be directly checked as part of the steady state initialization process against design values. 7.1.1.4 Typical Initialization Process Prior to the initialization process, certain parameters critical to establishing the correct steady state for the event of interest are identified. Once the parameters of interest achieve steady state target values within acceptable tolerances, on the basis of engineering judgement, ((
}}2(a),(c) A steady state solution has been achieved when the change in the target value during the loop transits is within the variance band described for each parameter.
A typical list of critical parameters for initializing the primary and secondary systems is provided below.
- reactor power: tolerance (or bounded) and variance
- fuel temperature: tolerance (or bounded) and variance
- RCS temperature: tolerance (or bounded) and variance
- RPV PZR pressure: tolerance (or bounded) and variance
- PZR level: tolerance (or bounded) and variance
- RCS flow: tolerance (or bounded) and variance
- steam pressure: tolerance (or bounded) and variance
- feedwater flow rate: variance
- feedwater temperature: tolerance (or bounded) and variance
© Copyright 20187 by NuScale Power, LLC 273
Response to Request for Additional Information Docket: PROJ0769 eRAI No.: 9374 Date of RAI Issue: 05/09/2018 NRC Question No.: 15.00.02-27 TR-0516-49416-P supports the conclusions in the NuScale FSAR, which under 10 CFR 52.47 must describe the facility, present the design bases and the limits on its operation, and present a safety analysis of the structures, systems, and components and of the facility as a whole. RG 1.203 describes the EMDAP, which the NRC staff considers acceptable for use in developing and assessing EMs used to analyze transient and accident behavior. Step 11 of the EMDAP describes the EM structure, and Item (6), Additional Features, addresses the code capability to model boundary conditions and control systems. TR-0516-49416-P, Section 7.1.2 discusses the treatment of plant control systems based on their impact on the calculated consequences relative to the acceptance criteria and states: The effect of various plant controls on the response to a specific acceptance criterion is assessed for each non-LOCA transient (Emphasis added) When considering operation of the various plant controls, the approach is based on the event consequences for a given acceptance criterion. ((
}}2(a),(c)
For the PCS functions of controlling pressurizer pressure, pressurizer level, core average coolant temperature, steam pressure, turbine load, and containment pressure, the initial limiting bias condition is determined for the transient analysis (i.e., the limiting condition in the band of operation of the PCS is determined for the associated function). However, for the most part, there is no discussion for each specific event and associated acceptance criteria of how the operation of the PCS during the event affects the transient response (i.e., the effect of the PCS operation attempting to maintain a control parameter within the operating band during the transient). The staff notes that control systems treatment could have a significant effect on the NuScale Nonproprietary
severity of a transient. Furthermore, the following statement in TR-0516-49416-P, Section 7.1.2, seems to indicate that credit may be taken for operation of the PCS or operator action. This would contradict the previous statement and Section 7.1.7 of the TR, which states that operator action is not credited for non-LOCA analyses: ((
}}2(a),(c)
Information Requested:
- a. Specify for which events PCS operation during the event is considered, and describe how operation of the PCS affects the transient response.
- b. Explain for which non-LOCA analyses PCS operation or operator actions, such as ((
}}2(a),(c), are credited.
- c. Update TR-0516-49416-P to clarify the discussion of plant controls and operator actions, including the second quoted statement above, as appropriate.
NuScale Response: a) The plant controls are listed in Part b of this information request. The impact to the transient response was generally annotated in the Basis column for those tables entitled Initial Conditions, Biases, and Conservatisms - design-basis event. However, a review of the tables of interest shows that not all control systems were discussed. In most instances, the actions related to steam system pressure control, feedwater and turbine load control, and containment pressure control were missing. Consequently, the affected tables were modified to include the missing information as indicated at the end of this response. b) Listed below are the plant control systems considered in the non-LOCA analyses. The plant control system functions are used to establish initial plant conditions and transient response when the normal operation of the system increases the consequences of the event or does not significantly alter the results. Plant Control System Operation (Section 7.1.2 and Section 6.3.1) Pressurizer pressure control (via pressurizer heaters and CVCS pressurizer spray) Pressurizer level control (via CVCS makeup and letdown) NuScale Nonproprietary
Reactor Coolant System Temperature control (via control rod banks or boron concentration) Steam Pressure Control (via turbine throttle valves or turbine bypass valves) Feedwater and Turbine Load Control (via feedwater pump speed) Containment Pressure Control (via containment evacuation system) Based on a review of the applicable event analysis, the status of each control system is provided in the following table. With respect to RCS Temperature control, only control via the rod control system is considered because changing the boron concentration of the reactor coolant requires a change to the boron concentration of the makeup fluid (a manual action). Yes means the control system is active No means the control system is not active (disabled) N/A means the control system is not applicable Partial means some aspect of the control system is active Plant Control Systems Non-LOCA Design-Basis Pzr RCS Steam Feedwater CNV Events (Criteria) Pressure Pzr Level Temp Pressure Flow Pressure 7.2.1 Decrease in Feedwater Yes Yes No Yes No Yes Temperature (MCHFR) (partial) (partial) 7.2.2 Increase in Feedwater Yes Yes No Yes N/A Yes Flow (MCHFR) (partial) (partial) 7.2.3 Increase in Steam Flow Yes No Yes N/A No Yes (MCHFR) (partial) 7.2.4 Steam System Piping Failure Inside or Outside of Yes Yes CNT No Yes No Yes (partial) (partial) (MCHFR) Yes Yes No Yes No Yes (partial) (partial) (Dose) 7.2.5 Containment Flooding Yes Yes Yes Yes Yes Yes (MCHFR) (partial) (partial) Loss of Containment Vacuum Yes Yes Yes Yes Yes N/A (MCHFR) (partial) (partial) NuScale Nonproprietary
7.2.6 Turbine Trip / Loss of Yes No No No No Yes External Load (pressure) (partial) 7.2.7 Loss of Condenser Yes No No No No Yes Vacuum (pressure) (partial) 7.2.8 Main Steam Isolation Yes Yes No Yes Yes Valve(s) Closure (pressure) (partial) No (partial) 7.2.9 Loss of Nonemergency Yes Yes Yes No Yes Yes AC Power (pressure) (partial) (partial) 7.2.10 Loss of Normal Yes No No No N/A No Feedwater Flow (pressure) (partial) 7.2.11 Inadvertent Decay Heat Removal System Yes Actuation No Yes Yes No (partial) Yes Scenario 1, one DHRS valve (partial) (pressure) Yes Yes Yes Scenario 2, one DHRS train No (partial) No (partial) Yes (pressure) Scenario 3, both DHRS trains No No No Yes Yes No (pressure) (partial) 7.2.12 Feedwater System Pipe Break Inside or Outside Yes CNT Yes No No No Yes (partial) (pressure) 7.2.13 Uncontrolled Control Rod Assembly Bank Withdrawal from Subcritical N/A Yes Yes or Low Power Startup Yes N/A Yes (partial) (partial) Conditions (partial) (MCHFR) Yes Yes Yes (partial) N/A Yes (partial) (partial) N/A (fuel centerline temperature) 7.2.14 Uncontrolled Control Rod Assembly Bank Withdrawal at Power Yes Yes Yes N/A Yes Yes (MCHFR) (partial) (partial) Yes Yes Yes N/A Yes Yes (fuel centerline temperature) (partial) (partial) 7.2.15 Control Rod Misoperation Withdrawal of a Single CRA Yes Yes (MCHFR) Yes N/A Yes Yes (partial) (partial) Yes Yes (fuel centerline temperature) Yes N/A Yes Yes (partial) (partial) Dropping One or More CRAs Yes Yes (MCHFR) No Yes Yes Yes (partial) (partial) Yes Yes (fuel centerline temperature) No Yes Yes Yes (partial) (partial) NuScale Nonproprietary
Misalignment of One or More N/A N/A N/A N/A N/A N/A CRAs 7.2.16 Inadvertent Decrease in Boron Concentration N/A Yes N/A N/A N/A N/A (MCHFR) 7.2.17 Chemical and Volume Control System Malfunction that Increases Reactor Yes Yes No Yes Yes Yes Coolant System Inventory (partial) (pressure) 7.2.18 Failure of Small Lines Yes Yes No No Yes Yes Outside Containment (dose) (partial) (partial) 7.2.19 Steam Generator Yes Yes No No Yes Yes Tube Failure (dose) (partial) (partial) Operator Actions As stated in Section 7.1.7 of the TR-0516-49416-P, there are no occasions where operator action is credited for event mitigation by the non-LOCA transient analyses. With respect to abnormal operating events, operator actions taken consistent with plant operating procedures are excluded from consideration. Text of Section 7.1.7 of the TR-0516-49416-P Operator actions credited for the non-LOCA transient analyses are typically justified and consistent with plant operating procedures. For the NPM, there are no occasions where operator action is credited for event mitigation by the non-LOCA transient analyses. Operator actions taken to prevent abnormal operating events from resulting in more severe events are excluded from consideration. For example, very small leaks of reactor coolant from the CVCS that do not result in automatic reactor trip for more than 30 minutes are considered an abnormal operating event where operators are expected to identify and isolate the leak before it results in a more severe event. c) The statement in question is incomplete because it does not clearly state what NuScale Nonproprietary
the operator does. The text in bold italics is added to clarify the inaction of the operator. Section 7.1.2 For non-LOCA transient analyses where loss of inventory or inventory shrinkage is expected to indicate makeup is needed, the process by which the operator approves makeup is considered. Since approval to initiate makeup is a manual action by the operator, the non-LOCA transient analyses do not credit this action. For spurious inventory addition events where automatic letdown could reduce the event consequences, credit for this action is not taken for the non-LOCA transient analyses. Section 6.3.1.2 For non-LOCA transient analyses where loss of inventory or inventory shrinkage is expected to indicate makeup is needed, the process by which the operator approves makeup is considered. Since approval to initiate makeup is a manual action by the operator, the non-LOCA transient analyses do not credit this action. For spurious inventory addition events where automatic letdown could reduce the event consequences, credit for this action is not taken for the non-LOCA transient analyses. Impact on Topical Report: Topical Report TR-0516-49416, Non-Loss of Coolant Accident Analysis Methodology, has been revised as described in the response above and as shown in the markup provided in this response. NuScale Nonproprietary
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Rev. 12 6.3.1 Module Control System (Nonsafety-related) The MCS as implemented in the NRELAP5 non-LOCA model provides control systems that may be used for model initialization and for simulating normal operating transients. These control systems can be used to initialize the model at new steady state conditions or to evaluate the nominal MCS response to a transient initiator. The MCS model allows the simulation of a variety of prototypic control scheme responses to transient conditions to provide nominal responses of these control schemes. The model also includes the capability to disable these control schemes to allow for conservative modeling applications as necessary. 6.3.1.1 Pressurizer Pressure Control (Nonsafety-related) The RPV pressure is controlled via the use of pressurizer heaters and spray to maintain the pressurizer steam pressure at a target value. When the pressure drops, the power to the PZR heaters increases, and if pressure increases, a portion of the CVCS recirculation flow is diverted to the pressurizer spray nozzles to collapse the steam space via condensation and reduce pressure. 6.3.1.2 Chemical Volume Control System Control (Nonsafety-related) The control systems implemented in the NRELAP5 non-LOCA model include simplified mechanisms for controlling CVCS recirculation flow, injection temperature and RCS inventory control (makeup and letdown). ((
}}2(a),(c) The water level in the pressurizer is controlled to a programmed setpoint by operation of the CVCS makeup pumps and the letdown control valve. While operation of CVCS letdown is automatically initiated by the MCS, control of the CVCS makeup is not. Instead, operator permission is required for CVCS makeup to be initiated.
For non-LOCA transient analyses where loss of inventory or inventory shrinkage is expected to indicate makeup is needed, the process by which the operator approves makeup is considered. Since approval to initiate makeup is a manual action by the operator, the non-LOCA transient analyses do not credit this action. For spurious inventory addition events where automatic letdown could reduce the event consequences, credit for this action is not taken for the non-LOCA transient analyses. 6.3.1.3 Reactor Coolant System Temperature Control (Nonsafety-related) The MCS model controls RCS average temperature by changing reactivity in the core to increase or decrease core power. This is accomplished with a control rod controller and a boron controller. The control rod controller uses design data to model a calculated rate of reactivity insertion due to maximum or nominal rod movement rates. ((
}}2(a),(c)
© Copyright 20187 by NuScale Power, LLC 265
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Rev. 12 Null transients are used to ensure the re-initialization cases achieve the desired steady state conditions. ((
}}2(a),(c) In addition, feedback from the point kinetics model is active for the null transients.
7.1.2 Treatment of Plant Controls Control systems are necessary to maneuver the NPM within the power range in accordance with normal operating transients. However, not all of these control systems are relevant to the non-LOCA transient analyses because the action of the control system does not alter the event consequences. This section identifies the normal, nonsafety-related plant controls considered in non-LOCA transient analyses. The effect of various plant controls on the response to a specific acceptance criterion is assessed for each non-LOCA transient. Several means are available to perform the assessment. For example, the assessment may consist of a combination of:
- Providing a qualitative engineering assessment that identifies why operation of a PCS is limiting or non-limiting.
- Performing a quantitative assessment via appropriate sensitivity cases.
- Referencing to applicable regulatory precedent that identifies why a particular normal plant control is not limiting for a specific transient or type of transient.
When considering operation of the various plant controls, the approach is based on the event consequences for a given acceptance criterion. Specifically, if operation of the control system leads to a less severe plant response, then the actions of the control system are not simulated for the transient of interest. Conversely, if operation of the control system causes the event consequences to be more severe, the PCS is assumed to operate as designed. ((
}}2(a)(c)
The normal PCSs to be considered in design basis event analysis are: Pressurizer Pressure Control Pressurizer pressure is controlled via operation of the pressurizer spray and the pressurizer heaters. Pressurizer Water Level Control The water level in the pressurizer is controlled to a programmed setpoint by operation of the CVCS makeup pumps and the letdown control valve. While operation of CVCS letdown is automatically initiated by the MCS, control of the CVCS makeup is not. Instead, operator permission is required for CVCS makeup to be initiated. © Copyright 20187 by NuScale Power, LLC 274
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Rev. 12 For non-LOCA transient analyses where loss of inventory or inventory shrinkage is expected to indicate makeup is needed, the process by which the operator approves makeup is considered. Since approval to initiate makeup is a manual action by the operator, the non-LOCA transient analyses do not credit this action. For spurious inventory addition events where automatic letdown could reduce the event consequences, credit for this action is not taken for the non-LOCA transient analyses. Core Average Coolant Temperature Control The average coolant temperature is controlled by moving the control rods or changing the boron concentration of the reactor coolant. This selection is based on the desired rate of change for core power. Hence, the control rods are moved to achieve faster power changes to meet the target average coolant temperature; slower power changes are accommodated by changing the boron concentration of the reactor coolant. At full power, the rod control system is set to insert only mode to prevent automatic withdrawal of the control rods during a transient. Steam Pressure Control Steam pressure is controlled to the desired value using the turbine throttle valves or the turbine bypass valves. The effect of these valves to a change in steam pressure is considered for the non-LOCA transient analyses. Turbine Load Control The mass flow rate and pressure provided by the feedwater pump is used to meet the desired turbine load, which reflects the power generation rate. The impact of the feedwater pumps continuing to operate until the feedwater line is isolated is considered for the non-LOCA transient analyses when DHRS is actuated. Containment pressure control The containment pressure is established at sub-atmospheric conditions via operation of the containment evacuation system. The impact of this system continuing to operate is considered for the non-LOCA transient analyses. 7.1.3 Loss of Power Conditions This section defines the term loss of normal power as applied to the NPM; describes the various power supplies (AC and DC); and, explains how the loss of these power supplies is treated by the non-LOCA transient analyses. 7.1.3.1 Background Chapter 15 of the SRP (Reference 15) does not ordinarily consider a loss of offsite power for events that require a malfunction of an active system for which power must be available; however, exceptions are made for some reactivity initiated events. The role of offsite power is less defined for the NPM plant than for traditional plants for several © Copyright 20187 by NuScale Power, LLC 275
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Rev. 12 Table 7-7 Initial conditions, biases, and conservatisms - decrease in feedwater temperature Parameter Bias / Conservatism Basis Initial reactor power RTP biased upwards to (( }}2(a)(c) account for measurement uncertainty. Initial RCS average temperature Biased to the high condition. (( }}2(a)(c) Initial RCS flow rate Biased to the low condition. (( }}2(a)(c) Initial PZR pressure Biased to the high condition (( }}2(a)(c) Initial PZR level Biased to the high condition. (( }}2(a)(c) Initial feedwater temperature Biased to the high condition. ((
}}2(a)(c)
Initial fuel temperature Nominal. ((
}}2(a)(c)
Moderator Temperature Biased to EOC conditions. (( Coefficient (MTC)
}}2(a)(c)
Kinetics Biased to the EOC condition. ((
}}2(a)(c)
Automatic rod control Enabled ((
}}2(a)(c)
Decay heat Biased to the low condition. ((
}}2(a)(c)
© Copyright 20187 by NuScale Power, LLC 296
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Rev. 12 Parameter Bias / Conservatism Basis Initial SG pressure(1) Biased to the high condition. ((
}}2(a)(c)
SG heat transfer Nominal ((
}}2(a)(c)
PZR spray Disabled ((
}}2(a)(c)
Letdown Disabled ((
}}2(a)(c)
PZR heaters Nominal ((
}}2(a)(c)
RSV lift setpoint Nominal ((
}}2(a)(c)
SG tube plugging Biased to the low condition. ((
}}2(a)(c)
Minimum feedwater temperature Biased to the low condition. ((
}}2(a)(c)
Feedwater temperature cooldown Varied (( rate }}2(a)(c) Feedwater flow rate Volumetric flow rate held (( constant during transient
}}2(a)(c)
© Copyright 20187 by NuScale Power, LLC 297
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Rev. 12 Parameter Bias / Conservatism Basis RCS Temperature Control Automatic rod control Enabled. (( Boron concentration Not credited.
}}2(a)(c)
PZR Pressure Control PZR spray (normal) Disabled. (( (bypass) Nominal. PZR heaters (non-prop.) Nominal. (prop.) Nominal.
}}2(a)(c)
PZR Level Control Charging Not credited. (( Letdown Disabled.
}}2(a)(c)
© Copyright 20187 by NuScale Power, LLC 298
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Rev. 12 Parameter Bias / Conservatism Basis Steam Pressure Control Turbine throttle valves Varied. (( Turbine bypass valves Disabled.
}}2(a)(c)
Feedwater and Turbine Load Control feedwater pump speed Disabled. ((
}}2(a)(c)
CNV Pressure Control CNV evacuation system Enabled. ((
}}2(a)(c)
- 1. (( }}2(a)(c)
Sensitivity studies are performed as needed to identify the limiting response(s) for the acceptance criteria parameter(s) challenged by the event (i.e., system pressures for overheating events, MCHFR for overcooling events). Consequently, sensitivity studies are performed to identify cases with the lowest CHFR response for this overcooling event. For example, several sensitivity studies are performed to consider the effects of fuel-related parameters (initial fuel temperature, time in life), boundary condition type, and the single active failure of an MSIV to isolate. It is concluded that failure of an MSIV to isolate has no impact on MCHFR because the time of MCHFR occurs prior to RTS/DHRS actuation and the associated isolation signal. Representative results for these studies are presented in Table 7-8 through Table 7-11. Representative results for two studies, which evaluate the effects of fuel exposure and initial fuel temperature over a range of feedwater temperature transients, are presented in Table 7-8 and Table 7-9. ((
}}2(a)(c)
© Copyright 20187 by NuScale Power, LLC 299
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Rev. 12 Table 7-14 Initial conditions, biases, and conservatisms - increase in feedwater flow Parameter Bias / Conservatism Basis Initial reactor power RTP biased upwards to (( }}2(a)(c) account for measurement uncertainty. Initial RCS average temperature Biased to the high condition. (( }}2(a)(c) Initial RCS flow rate Biased to the low condition. (( }}2(a)(c) Initial PZR pressure Biased to the high condition. (( }}2(a)(c) Initial PZR level Biased to the high condition. (( }}2(a)(c) Initial feedwater temperature Low. ((
}}2(a)(c)
Initial fuel temperature Nominal. ((
}}2(a)(c)
MTC Biased to EOC conditions. ((
}}2(a)(c)
Kinetics Biased to the EOC condition. ((
}}2(a)(c)
Automatic rod control Enabled. ((
}}2(a)(c)
© Copyright 20187 by NuScale Power, LLC 305
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Rev. 12 Parameter Bias / Conservatism Basis Decay heat Biased to the low condition. ((
}}2(a)(c)
Initial SG pressure(1) High ((
}}2(a)(c)
SG heat transfer Low ((
}}2(a)(c)
PZR spray Disabled ((
}}2(a)(c)
Letdown Disabled ((
}}2(a)(c)
PZR heaters Nominal ((
}}2(a)(c)
RSV lift setpoint Nominal ((
}}2(a)(c)
SG tube plugging Biased to the low condition. ((
}}2(a)(c)
© Copyright 20187 by NuScale Power, LLC 306
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Rev. 12 Parameter Bias / Conservatism Basis RCS Temperature Control Automatic rod control Enabled. (( Boron concentration Not credited.
}}2(a)(c)
PZR Pressure Control PZR spray (normal) Disabled. (( (bypass) Nominal. PZR heaters (non-prop.) Nominal. (prop.) Nominal.
}}2(a)(c)
PZR Level Control Charging Not credited. (( Letdown Disabled.
}}2(a)(c)
© Copyright 20187 by NuScale Power, LLC 307
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Rev. 12 Parameter Bias / Conservatism Basis Steam Pressure Control Turbine throttle valves Enabled. (( Turbine bypass valves Disabled.
}}2(a)(c)
Feedwater and Turbine Load Control feedwater pump speed N/A. Not applicable to this event. CNV Pressure Control CNV evacuation system Enabled. ((
}}2(a)(c)
- 1. (( }}2(a)(c)
Sensitivity studies are performed as needed to identify the limiting response(s) for the acceptance criteria parameter(s) challenged by the event (i.e., system pressures for overheating events, MCHFR for overcooling events). Consequently, sensitivity studies are performed to identify cases with the lowest CHFR response for this overcooling event. For example, two sensitivity studies are performed to identify the most challenging SG liquid inventory and feedwater flow transient for MCHFR. ((
}}2(a),(c) Representative results for these studies are presented in Table 7-15 and Table 7-16.
© Copyright 20187 by NuScale Power, LLC 308
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Rev. 12 Table 7-19 Initial conditions, biases, and conservatisms - increase in steam flow Parameter Bias / Conservatism Basis Initial reactor power RTP biased upwards to (( account for measurement }}2(a)(c) uncertainty. Initial RCS average temperature Biased to the high condition. ((
}}2(a)(c)
Initial RCS flow rate Biased to the low condition. ((
}}2(a)(c)
Initial PZR pressure Biased to the high condition. ((
}}2(a)(c)
Initial PZR level Biased to the high condition. ((
}}2(a)(c)
Initial feedwater temperature Nominal. ((
}}2(a)(c)
Initial fuel temperature Nominal. ((
}}2(a)(c)
Moderator temperature Biased to EOC conditions. (( coefficientMTC
}}2(a)(c)
Kinetics Biased to the EOC condition. ((
}}2(a)(c)
Automatic rod control Enabled ((
}}2(a)(c)
Decay heat Biased to the low condition. ((
}}2(a)(c)
Initial SG pressure(1) Biased to the high condition. ((
}}2(a)(c)
SG heat transfer Varied ((
}}2(a)(c)
© Copyright 20187 by NuScale Power, LLC 313
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Rev. 12 Parameter Bias / Conservatism Basis PZR spray Disabled ((
}}2(a)(c)
Letdown Disabled ((
}}2(a)(c)
PZR heaters Nominal ((
}}2(a)(c)
RSV lift setpoint Nominal ((
}}2(a)(c)
SG tube plugging Biased to the low condition. ((
}}2(a)(c)
Steam flow increase Varied ((
}}2(a)(c)
RCS Temperature Control Automatic rod control Enabled. (( Boron concentration Not credited.
}}2(a)(c)
© Copyright 20187 by NuScale Power, LLC 314
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Rev. 12 Parameter Bias / Conservatism Basis PZR Pressure Control PZR spray (normal) Disabled. (( (bypass) Nominal. PZR heaters (non-prop.) Nominal. (prop.) Nominal.
}}2(a)(c)
PZR Level Control Charging Not credited. (( Letdown Disabled.
}}2(a)(c)
Steam Pressure Control Turbine throttle valves N/A. Not applicable to this event. Turbine bypass valves N/A. Not applicable to this event. Feedwater and Turbine Load Control feedwater pump speed Disabled. ((
}}2(a)(c)
CNV Pressure Control CNV evacuation system Enabled. ((
}}2(a)(c)
- 1. (( }}2(a)(c)
© Copyright 20187 by NuScale Power, LLC 315
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Rev. 12 Table 7-24 Initial conditions, biases, and conservatisms - steam line break Parameter Bias / Conservatism Basis Initial reactor power Biased upwards to account for (( measurement uncertainty. }}2(a)(c) Initial RCS average temperature Biased to the high condition. (( }}2(a)(c) Initial RCS flow rate Biased to the low condition. (( }}2(a)(c) Initial PZR pressure Varied ((
}}2(a)(c)
Initial PZR level Varied ((
}}2(a)(c)
Initial feedwater temperature Varied. ((
}}2(a)(c)
Initial fuel temperature Biased to the low condition. (( . }}2(a)(c) Moderator temperature Both EOC and BOC conditions. (( coefficientMTC
}}2(a)(c)
Kinetics Both EOC and BOC conditions. ((
}}2(a)(c)
Automatic rod control Enabled or disabled ((
}}2(a)(c)
Decay heat Both high and low conditions ((
}}2(a)(c)
Initial SG pressure(1) Biased to the high condition. ((
}}2(a)(c)
SG heat transfer Varied. ((
}}2(a)(c)
PZR spray Nominal ((
}}2(a)(c)
© Copyright 20187 by NuScale Power, LLC 320
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Rev. 12 Parameter Bias / Conservatism Basis Letdown Disabled ((
}}2(a)(c)
PZR heaters Normal heater operation is (( allowed.
}}2(a)(c)
RSV lift setpoint Biased to the high condition. (( . }}2(a)(c) SG tube plugging Varied to consider both the low (( and high ends of the allowable }}2(a)(c) range. RCS Temperature Control Automatic rod control Enabled. (MCHFR) (( Varied. (dose) Boron concentration Not credited.
}}2(a)(c)
© Copyright 20187 by NuScale Power, LLC 321
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Rev. 12 Parameter Bias / Conservatism Basis PZR Pressure Control PZR spray (normal) Disabled. (( (bypass) Nominal. PZR heaters (non-prop.) Enabled. (prop.) Nominal.
}}2(a)(c)
PZR Level Control Charging Not credited. (( Letdown Disabled.
}}2(a)(c)
Steam Pressure Control Turbine throttle valves Enabled. (( Turbine bypass valves Disabled.
}}2(a)(c)
© Copyright 20187 by NuScale Power, LLC 322
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Rev. 12 Parameter Bias / Conservatism Basis Feedwater and Turbine Load Control feedwater pump speed Disabled. ((
}}2(a)(c)
CNV Pressure Control CNV evacuation system Enabled. ((
}}2(a)(c)
- 1. (( }}2(a)(c)
Sensitivity studies are performed as needed to identify the limiting response(s) for the acceptance criteria parameter(s) challenged by the event. Consequently, sensitivity studies are performed to identify cases with the lowest CHFR response and challenging mass releases for this overcooling event. Representative results for this study are presented in Table 7-25. In the cases modeling breaks outside of containment, the MSIV on the train with the break is assumed to fail to close, resulting in the complete emptying of the affected SG. In cases modeling breaks inside of containment, the FWIV on the train with the break is assumed to fail to close, which maximizes the event consequences. Break size is the fraction (percent) of the pipe cross section area. ((
}}2(a),(c)
© Copyright 20187 by NuScale Power, LLC 323
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Rev. 12 Table 7-28 Initial conditions, biases, and conservatisms - containment flooding / loss of containment vacuum Parameter Bias / Conservatism Basis Initial reactor power RTP biased upwards to (( }}2(a)(c) account for measurement uncertainty. Initial RCS average temperature Biased to the high condition. (( }}2(a)(c) Initial RCS flow rate Biased to the low condition. (( }}2(a)(c) Initial PZR pressure Varied ((
}}2(a)(c)
Initial PZR level Nominal ((
}}2(a)(c)
Initial feedwater temperature Nominal ((
}}2(a)(c)
Initial fuel temperature Nominal ((
}}2(a)(c)
Moderator temperature Biased to the EOC condition. (( }}2(a)(c) coefficientMTC Kinetics Biased to the EOC condition. (( }}2(a)(c) Automatic rod control Enabled. ((
}}2(a)(c)
Decay heat Biased to the high condition. ((
}}2(a)(c)
(1) Initial SG pressure Nominal ((
}}2(a)(c)
SG heat transfer Nominal ((
}}2(a)(c)
PZR spray Nominal ((
}}2(a)(c)
© Copyright 20187 by NuScale Power, LLC 328
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Rev. 12 Parameter Bias / Conservatism Basis Letdown Automatic RCS inventory (( control is enabled.
}}2(a)(c)
PZR heaters Normal heater operation is (( allowed.
}}2(a)(c)
RSV lift setpoint Nominal ((
}}2(a)(c)
SG tube plugging Biased to the low condition. ((
}}2(a)(c)
Initial containment pressure Varied ((
}}2(a)(c)
Initial pool temperature Varied ((
}}2(a)(c)
RCCW leak flow Varied ((
}}2(a)(c)
RCCW temperature Varied ((
}}2(a)(c)
RCS Temperature Control Automatic rod control Enabled. (( Boron concentration Not credited.
}}2(a)(c)
© Copyright 20187 by NuScale Power, LLC 329
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Rev. 12 Parameter Bias / Conservatism Basis PZR Pressure Control PZR spray (normal) Enabled. (( (bypass) Nominal. PZR heaters (non-prop.) Enabled. (prop.) Nominal.
}}2(a)(c)
PZR Level Control Charging Not credited. (( Letdown Enabled.
}}2(a)(c)
Steam Pressure Control Turbine throttle valves Enabled. (( Turbine bypass valves Disabled.
}}2(a)(c)
© Copyright 20187 by NuScale Power, LLC 330
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Rev. 12 Parameter Bias / Conservatism Basis Feedwater and Turbine Load Control feedwater pump speed Enabled. ((
}}2(a)(c)
CNV Pressure Control CNV evacuation system Enabled. (CNV flooding) ((
}}2(a)(c)
Not applicable to this event. N/A. (CNV vacuum loss)
- 1. (( }}2(a)(c)
Sensitivity studies are performed as needed to identify the limiting response(s) for the acceptance criteria parameter(s) challenged by the event (i.e., system pressures for overheating events, MCHFR for overcooling events). Consequently, sensitivity studies are performed to identify cases with the lowest CHFR response for this overcooling event. As an example, the initial reactor power, pressurizer pressure, RCS average coolant temperature, RCS flow rate, containment pressure, pool temperature, RCCW leak flow rate, RCCW temperature are varied to achieve the results presented in Table 7-29 for the CNV flooding/loss of CNV vacuum event. The sensitivity studies indicate that the CNV flooding cases are more challenging to MCHFR than loss of CNV vacuum. However, considering Figure 4-1, the sensitivity study results indicate that for a variety of initial RCS conditions, reactor pool conditions, and condition of liquid or air ingress to containment, loss of containment vacuum or containment flooding results in a slow overcooling transient that is non-limiting with respect to MCHFR compared to other AOOs. © Copyright 20187 by NuScale Power, LLC 331
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Rev. 12 Table 7-31 Acceptance criteria - turbine trip / loss of external load Acceptance Criteria Discussion Primary pressure Primary pressure quickly rises to the peak value, then drops as the lowest setpoint RSV lifts to reduce pressure. Secondary pressure Peak secondary pressurization is largely a function of DHRS actuation, in addition to the actual turbine trip or loss of external load. The DHRS heat removal is limited by the DHR condenser so some pressurization is expected for every actuation of this system. Critical heat flux ratio This criterion is evaluated by downstream subchannel analysis. Maximum fuel centerline temperature This criterion is evaluated by downstream subchannel analysis. Containment integrity Containment integrity is evaluated by a separate analysis methodology. Escalation of an AOO to an accident This criterion is satisfied by demonstrating stable RCS flow rates and constant or downward trending RCS and DHRS pressures and temperatures exist at the end of the transient, all acceptance criteria evaluated in the transient analysis are met, and shutdown margin is maintained at the end of the transient. RCS conditions during extended DHRS cooling are addressed in a separate analysis. 7.2.6.3 Biases, Conservatisms, and Sensitivity Studies The biases and conservatisms presented in Table 7-32 are considered in identifying the bounding transient simulation for primary and steam generator pressure. Table 7-32 Initial conditions, biases, and conservatisms - turbine trip / loss of external load Parameter Bias / Conservatism Basis Initial reactor power RTP biased to the high (( condition to account for }}2(a)(c) measurement uncertainty. Initial RCS average temperature Varied. ((
. }}2(a)(c)
© Copyright 20187 by NuScale Power, LLC 335
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Rev. 12 Parameter Bias / Conservatism Basis Initial RCS flow rate Biased to the low condition. ((
}}2(a)(c)
Initial PZR pressure Varied. ((
}}2(a)(c)
Initial PZR level Biased to the high condition. ((
}}2(a)(c)
Initial feedwater temperature Nominal. ((
}}2(a)(c)
Initial fuel temperature Biased to the high condition ((
}}2(a)(c)
Moderator temperature Consistent with BOC (( coefficientMTC conditions. }}2(a)(c) Kinetics Biased to BOC conditions. ((
}}2(a)(c)
Automatic rod control Disabled. ((
}}2(a)(c)
Decay heat Biased to the high condition. ((
}}2(a)(c)
(1) Initial SG pressure Varied. ((
}}2(a)(c)
Steam generator heat transfer Varied. ((
}}2(a)(c)
PZR spray Automatic PZR spray is (( disabled }}2(a)(c) Letdown Automatic RCS inventory (( control is disabled }}2(a)(c) PZR heaters Heat input held constant. ((
}}2(a)(c)
© Copyright 20187 by NuScale Power, LLC 336
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Rev. 12 Parameter Bias / Conservatism Basis RSV lift setpoint Biased to the high condition. ((
}}2(a)(c)
SG tube plugging Biased to the low condition. ((
}}2(a)(c)
RCS Temperature Control Automatic rod control Disabled. (( Boron concentration Not credited.
}}2(a)(c)
PZR Pressure Control PZR spray (normal) Disabled. (( (bypass) Nominal. PZR heaters (non-prop.) Disabled. (prop.) Nominal.
}}2(a)(c)
PZR Level Control Charging Not credited. (( Letdown Disabled.
}}2(a)(c)
Steam Pressure Control Turbine throttle valves Disabled. (( Disabled. Turbine bypass valves
}}2(a)(c)
© Copyright 20187 by NuScale Power, LLC 337
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Rev. 12 Parameter Bias / Conservatism Basis Feedwater and Turbine Load Control feedwater pump speed Disabled. ((
}}2(a)(c)
CNV Pressure Control CNV evacuation system Enabled. ((
}}2(a)(c)
- 1. (( }}2(a)(c)
Sensitivity studies are performed as needed to identify the limiting response(s) for the acceptance criteria parameter(s) challenged by the event (i.e., system pressures for overheating events, MCHFR for overcooling events). Consequently, sensitivity studies are performed to identify cases with the highest pressure responses for this overheating event. For example a sensitivity study is performed to identify the highest primary and secondary side pressures, varying initial primary side conditions, initial steam pressure, loss of power, and single active failure (other parameters are biased as indicated in Table 7-32). Representative results of this sensitivity study are presented in Table 7-33. The results of the sensitivity studies indicate that maximum RPV pressure occurs when PZR pressure is biased to the high condition, average temperature is biased to the low condition, and a loss of power leads to a loss of the secondary side heat sink (since feedwater is lost, the failure of an FWIV to close would have no effect on the event). With respect to secondary side SG pressure, maximum SG pressure occurs when average temperature and SG pressure are biased to the high condition, and all power sources are available - the failure of an FWIV to close increases the peak SG pressure by a very small amount and remains well below the design pressure of 2100 psia. © Copyright 20187 by NuScale Power, LLC 338
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Rev. 12 Table 7-36 Initial conditions, biases, and conservatisms - loss of condenser vacuum Parameter Bias / Conservatism Basis Initial reactor power RTP biased upwards to (( account for measurement }}2(a)(c) uncertainty. Initial RCS average temperature Varied. ((
}}2(a)(c)
Initial RCS flow rate Biased to the low condition. ((
}}2(a)(c)
Initial PZR pressure Varied. ((
}}2(a)(c)
Initial PZR level Biased to the high condition. ((
}}2(a)(c)
Initial feedwater temperature Nominal. ((
}}2(a)(c)
Initial fuel temperature Biased to the high condition ((
}}2(a)(c)
Moderator temperature Consistent with BOC kinetics. (( coefficientMTC }}2(a)(c) Kinetics Biased to BOC conditions. ((
}}2(a)(c)
Automatic rod control Disabled. ((
}}2(a)(c)
Decay heat Biased to the high condition. ((
}}2(a)(c)
Initial SG pressure(1) Varied. ((
}}2(a)(c)
© Copyright 20187 by NuScale Power, LLC 342
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Rev. 12 Parameter Bias / Conservatism Basis Steam generator heat transfer Varied. ((
}}2(a)(c)
PZR spray Automatic PZR spray is (( disabled }}2(a)(c) Letdown Automatic RCS inventory (( control is disabled }}2(a)(c) PZR heaters Heat input held constant. ((
}}2(a)(c)
RSV lift setpoint Biased to the high condition. ((
}}2(a)(c)
SG tube plugging Biased to the low condition. ((
}}2(a)(c)
RCS Temperature Control Automatic rod control Disabled. (( Boron concentration Not credited.
}}2(a)(c)
PZR Pressure Control PZR spray (normal) Disabled. (( (bypass) Nominal. PZR heaters (non-prop.) Disabled. (prop.) Nominal.
}}2(a)(c)
PZR Level Control Charging Not credited. (( Letdown Disabled.
}}2(a)(c)
© Copyright 20187 by NuScale Power, LLC 343
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Rev. 12 Parameter Bias / Conservatism Basis Steam Pressure Control Turbine throttle valves Disabled. (( Turbine bypass valves Disabled.
}}2(a)(c)
Feedwater and Turbine Load Control feedwater pump speed Disabled. ((
}}2(a)(c)
CNV Pressure Control CNV evacuation system Enabled. ((
}}2(a)(c)
- 1. (( }}2(a)(c)
Sensitivity studies are performed as needed to identify the limiting response(s) for the acceptance criteria parameter(s) challenged by the event (i.e., system pressures for overheating events, MCHFR for overcooling events). Consequently, sensitivity studies are performed to identify cases with the highest pressure responses for this overheating event. For example a sensitivity study is performed to identify the highest primary and secondary side pressures, varying primary side conditions, steam pressure, loss of power, and single active failure (other parameters are biased as indicated in Table 7-36.) Representative results of this sensitivity study are presented in Table 7-37. Several cases with consistent biases and various combinations of an FWIV failure and a loss of AC power are considered. The peak RPV pressures for all of these cases are within a few tenths of a psi, demonstrating that since feedwater is lost at transient initiation, the results are insensitive to failure of an FWIV to isolate and loss of AC power, all other things being equal. Additionally, the results of the sensitivity studies indicate that maximum RPV pressure occurs when PZR pressure is biased to the high condition, and average temperature is biased to the low condition. With respect to secondary side SG pressure, maximum SG pressure occurs when average temperature and SG pressure are biased to the high condition, and all power sources are available (since feedwater is lost, the failure of an FWIV to close would have no effect on the event). © Copyright 20187 by NuScale Power, LLC 344
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Rev. 12 Table 7-40 Initial conditions, biases, and conservatisms - main steam isolation valve closure Parameter Bias / Conservatism Basis Initial reactor power RTP biased upwards to (( account for measurement }}2(a)(c) uncertainty. Initial RCS average temperature Varied. ((
}}2(a)(c)
Initial RCS flow rate Varied. ((
}}2(a)(c)
Initial PZR pressure Varied. ((
}}2(a)(c)
Initial PZR level Varied. ((
. }}2(a)(c)
Initial feedwater temperature Nominal. ((
}}2(a)(c)
Initial fuel temperature Biased to the high condition ((
. }}2(a)(c)
Moderator temperature Consistent with BOC kinetics. (( coefficientMTC }}2(a)(c) Kinetics Biased to BOC conditions. ((
}}2(a)(c)
Automatic rod control Disabled. ((
}}2(a)(c)
Decay heat Biased to the high condition. ((
}}2(a)(c)
Initial SG pressure(1) Varied. ((
}}2(a)(c)
Steam generator heat transfer Varied. ((
}}2(a)(c)
PZR spray Automatic PZR spray is (( disabled }}2(a)(c) Letdown Automatic RCS inventory (( control is disabled }}2(a)(c) PZR heaters Heat input held constant. ((
}}2(a)(c)
RSV lift setpoint Biased to the high condition. ((
}}2(a)(c)
SG tube plugging Biased to the low condition. ((
}}2(a)(c)
© Copyright 20187 by NuScale Power, LLC 348
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Rev. 12 Parameter Bias / Conservatism Basis RCS Temperature Control Automatic rod control Disabled. (( Boron concentration Not credited.
}}2(a)(c)
PZR Pressure Control PZR spray (normal) Disabled. (( (bypass) Nominal. PZR heaters (non-prop.) Disabled. (prop.) Nominal.
}}2(a)(c)
PZR Level Control Charging Not credited. (( Letdown Disabled.
}}2(a)(c)
Steam Pressure Control Turbine throttle valves Enabled. (( Turbine bypass valves Disabled.
}}2(a)(c)
Feedwater and Turbine Load Control feedwater pump speed Enabled. ((
}}2(a)(c)
© Copyright 20187 by NuScale Power, LLC 349
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Rev. 12 Parameter Bias / Conservatism Basis CNV Pressure Control CNV evacuation system Enabled. ((
}}2(a)(c)
- 1. (( }}2(a)(c)
Sensitivity studies are performed as needed to identify the limiting response(s) for the acceptance criteria parameter(s) challenged by the event (i.e., system pressures for overheating events, MCHFR for overcooling events). Consequently, sensitivity studies are performed to identify cases with the highest pressure responses for this overheating event. For example, a sensitivity study is performed to identify the highest primary and secondary side pressures, varying primary conditions, steam generator heat transfer and pressure, etc. (other parameters are biased as indicated in Table 7-40). Representative results of this sensitivity study are presented in Table 7-41. All results presented consider the closure of two MSIVs, which isolates both steam generators, and no single active failure. Maximum pressures are similar for cases in which the RSV actuation limits the RPV pressurization. For example, the maximum calculated pressure for the peak RPV pressure study is equal (to one decimal place) to the peak RPV pressures for two additional cases, namely, decreased steam generator heat transfer and decreased feedwater temperature. Additionally, the results of the sensitivity studies indicate that maximum RPV pressure occurs when PZR pressure is biased to the high condition, RCS flow is at the nominal value (if maximum RCS pressure is below the RSV lift pressure), and a loss of AC power leads to a loss of the secondary side heat sink (loss of AC power occurs coincident with reactor trip). With respect to secondary side SG pressure, maximum SG pressure occurs when average temperature is biased to the high condition, RCS flow is biased to the low condition, and all power sources are available. © Copyright 20187 by NuScale Power, LLC 350
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Rev. 12 7.2.9.3 Biases, Conservatisms, and Sensitivity Studies The biases and conservatisms presented in Table 7-44 are considered in identifying the bounding transient simulation for primary and secondary pressure. Table 7-44 Initial conditions, biases, and conservatisms - loss of normal AC power Parameter Bias / Conservatism Basis Initial reactor power RTP biased upwards to (( account for measurement }}2(a)(c) uncertainty. Initial RCS average temperature Varied. ((
}}2(a)(c)
Initial RCS flow rate Biased to the low condition. ((
}}2(a)(c)
Initial PZR pressure Varied. ((
}}2(a)(c)
Initial PZR level Biased to the high condition. ((
}}2(a)(c)
Initial feedwater temperature Nominal. ((
}}2(a)(c)
© Copyright 20187 by NuScale Power, LLC 355
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Rev. 12 Parameter Bias / Conservatism Basis Initial fuel temperature Biased to the high condition ((
}}2(a)(c)
Moderator temperature Consistent with BOC kinetics. (( coefficientMTC }}2(a)(c) Kinetics Biased to BOC conditions. ((
}}2(a)(c)
Automatic rod control Disabled. ((
}}2(a)(c)
Decay heat Biased to the high condition. ((
}}2(a)(c)
Initial SG pressure(1) Varied. ((
}}2(a)(c)
SG heat transfer Varied. ((
}}2(a)(c)
PZR spray Automatic PZR spray is (( disabled }}2(a)(c) Letdown Automatic RCS inventory (( control is disabled }}2(a)(c) PZR heaters Lost on loss of power. (( }}2(a)(c) RSV lift setpoint Biased to the high condition. ((
}}2(a)(c)
SG tube plugging Biased to the low condition. ((
}}2(a)(c)
RCS Temperature Control Automatic rod control Enabled (prior to event (( initiation). Boron concentration Not credited.
}}2(a)(c)
© Copyright 20187 by NuScale Power, LLC 356
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Rev. 12 Parameter Bias / Conservatism Basis PZR Pressure Control PZR spray (normal) Enabled. (( (bypass) Nominal. PZR heaters (non-prop.) Enabled. (prop.) Enabled.
}}2(a)(c)
PZR Level Control Charging Not credited. (( Letdown Enabled.
}}2(a)(c)
Steam Pressure Control Turbine throttle valves Enabled. (( Turbine bypass valves Disabled.
}}2(a)(c)
Feedwater and Turbine Load Control feedwater pump speed Enabled. ((
}}2(a)(c)
CNV Pressure Control CNV evacuation system Enabled. ((
}}2(a)(c)
© Copyright 20187 by NuScale Power, LLC 357
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Rev. 12 Table 7-48 Initial conditions, biases, and conservatisms - loss of normal feedwater flow Parameter Bias / Conservatism Basis Initial reactor power RTP biased upwards to (( account for measurement }}2(a)(c) uncertainty. Initial RCS average temperature Varied. ((
}}2(a)(c)
Initial RCS flow rate Varied. ((
}}2(a)(c)
Initial PZR pressure Varied. ((
. }}2(a)(c)
Initial PZR level Varied. ((
. }}2(a)(c)
Initial feedwater temperature Varied. ((
}}2(a)(c)
Initial fuel temperature Biased to the high condition. ((
}}2(a)(c)
Moderator temperature Consistent with BOC kinetics. (( coefficientMTC }}2(a)(c) Kinetics Biased to BOC conditions. ((
}}2(a)(c)
Automatic rod control Disabled. ((
}}2(a)(c)
Decay heat Biased to the high condition. ((
}}2(a)(c)
(1) Initial SG pressure Varied. ((
}}2(a)(c)
SG heat transfer Varied. ((
}}2(a)(c)
PZR spray Automatic PZR spray is (( disabled }}2(a)(c) Letdown Automatic RCS inventory (( control is disabled }}2(a)(c) PZR heaters Held constant until loss of AC (( power. }}2(a)(c) RSV lift setpoint Biased to the high condition. ((
}}2(a)(c)
SG tube plugging Biased to the low condition. ((
}}2(a)(c)
Feedwater flow decrease Varied ((
}}2(a)(c)
© Copyright 20187 by NuScale Power, LLC 361
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Rev. 12 Parameter Bias / Conservatism Basis RCS Temperature Control Automatic rod control Disabled. (( Boron concentration Not credited.
}}2(a)(c)
PZR Pressure Control PZR spray (normal) Disabled. (( (bypass) Nominal. PZR heaters (non-prop.) Disabled. (prop.) Nominal.
}}2(a)(c)
PZR Level Control Charging Not credited. (( Letdown Disabled.
}}2(a)(c)
Steam Pressure Control Turbine throttle valves Enabled. (( Turbine bypass valves Disabled.
}}2(a)(c)
Feedwater and Turbine Load Control feedwater pump speed N/A. Not applicable to this event. © Copyright 20187 by NuScale Power, LLC 362
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Rev. 12 Parameter Bias / Conservatism Basis CNV Pressure Control CNV evacuation system Disabled. ((
}}2(a)(c)
- 1. (( }}2(a)(c)
Sensitivity studies are performed as needed to identify the limiting response(s) for the acceptance criteria parameter(s) challenged by the event. Consequently, sensitivity studies are performed to identify cases with the highest primary and secondary pressures, varying the magnitude of the feedwater flow rate decrease (other parameters are biased as indicated in Table Table 7-48). Representative results of this sensitivity study are presented in Table 7-49. The results of the sensitivity studies indicate that peak primary pressure occurs during a total loss of feedwatersince this is sufficient to actuate the lowest-setpoint RSV, ((
}}2(a),(c) With respect to maximum secondary side SG pressure, maximum SG pressure occurs during a partial loss of feedwater, as the mismatch between primary heat production and secondary heat sink causes the RCS temperature to increase, and eventually actuate the RTS on high RCS riser temperature.
Table 7-49 Sensitivity studies - loss of normal feedwater flow ((
}}2(a),(c) © Copyright 20187 by NuScale Power, LLC 363
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Rev. 12 Table 7-52 Initial conditions, biases, and conservatisms - inadvertent decay heat removal system actuation Parameter Bias / Conservatism Basis Initial reactor power Variedmost challenging (( cases are RTP biased upwards }}2(a)(c) to account for measurement uncertainty. Initial RCS average temperature Varied. ((
}}2(a)(c)
Initial RCS flow rate Varied. ((
}}2(a)(c)
Initial PZR pressure Varied. ((
}}2(a)(c)
Initial PZR level Varied. ((
}}2(a)(c)
Initial feedwater temperature Nominal. ((
}}2(a)(c)
Initial fuel temperature Biased to the high condition ((
}}2(a)(c)
Moderator temperature Consistent with BOC kinetics. (( coefficientMTC }}2(a)(c) Kinetics Variedmost challenging (( cases biased to BOC }}2(a)(c) conditions. Automatic rod control Disabled. ((
}}2(a)(c)
Decay heat Biased to the high condition. ((
}}2(a)(c)
(1) Initial SG pressure Varied. ((
}}2(a)(c)
Steam generator heat transfer Varied. ((
}}2(a)(c)
PZR spray Automatic PZR spray is (( disabled }}2(a)(c) © Copyright 20187 by NuScale Power, LLC 367
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Rev. 12 Parameter Bias / Conservatism Basis Letdown Automatic RCS inventory (( control is disabled }}2(a)(c) PZR heaters Constant heat input. ((
}}2(a)(c)
RSV lift setpoint Biased to the high condition. ((
}}2(a)(c)
SG tube plugging Biased to the low condition. ((
}}2(a)(c)
RCS Temperature Control Automatic rod control Disabled. (( Boron concentration Not credited.
}}2(a)(c)
PZR Pressure Control PZR spray (normal) Enabled. (Scenarios 1 & 2) (( Disabled. (Scenario 3) (bypass) Nominal. (Scenarios 1 & 2) Disabled. (Scenario 3) PZR heaters (non-prop.) Enabled. (prop.) Nominal.
}}2(a)(c)
© Copyright 20187 by NuScale Power, LLC 368
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Rev. 12 Parameter Bias / Conservatism Basis PZR Level Control Charging Not credited. (( Letdown Enabled. (Scenarios 1 & 2) Disabled. (Scenario 3)
}}2(a)(c)
Steam Pressure Control Turbine throttle valves Enabled. (( Turbine bypass valves Disabled.
}}2(a)(c)
Feedwater and Turbine Load Control feedwater pump speed Disabled. ((
}}2(a)(c)
CNV Pressure Control CNV evacuation system Enabled. ((
}}2(a)(c)
- 1. (( }}2(a)(c)
- 2. NPM RSV relieving capacities are sized to be significantly greater than transient induced NPM pressurization rates, thus minimizing pressure overshoot. ((
}}2(a)(c)
Sensitivity studies are performed as needed to identify the limiting response(s) for the acceptance criteria parameter(s) challenged by the event. Consequently, sensitivity studies are performed to identify cases with the highest pressure responses for this event. For example, a sensitivity study is performed to identify the highest primary and secondary side pressures, varying primary conditions, steam generator heat transfer and pressure, etc. (other parameters are biased as indicated in Table 7-52). Representative results of this sensitivity study are presented in Table 7-53. Unless otherwise noted, all cases use BOC kinetics. The results of the sensitivity studies indicate that peak primary pressure occurs when both trains of DHR inadvertently actuate (i.e., total loss of normal heat sink). The lowest-setpoint RSV operates to mitigate peak pressurization. (( }}2(a),(c) © Copyright 20187 by NuScale Power, LLC 369
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Rev. 12 Table 7-56 Initial conditions, biases, and conservatisms - feedwater line break Parameter Bias / Conservatism Basis Initial reactor power RTP biased upwards to (( account for measurement }}2(a)(c) uncertainty. Initial RCS average temperature Varied. ((
}}2(a)(c)
Initial RCS flow rate Biased to the low condition. ((
}}2(a)(c)
Initial PZR pressure Varied. ((
}}2(a)(c)
Initial PZR level Varied. ((
}}2(a)(c)
Initial feedwater temperature Varied. ((
}}2(a)(c)
Initial fuel temperature Biased to the high condition. ((
}}2(a)(c)
Moderator temperature Consistent with BOC kinetics. (( coefficientMTC }}2(a)(c) Kinetics Biased to BOC conditions. ((
}}2(a)(c)
Automatic rod control Disabled. ((
}}2(a)(c)
Decay heat Biased to the high condition. ((
}}2(a)(c)
(1) Initial SG pressure Varied. ((
}}2(a)(c)
SG heat transfer Varied. ((
}}2(a)(c)
PZR spray Varied, lost on loss of power ((
}}2(a)(c)
© Copyright 20187 by NuScale Power, LLC 373
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Rev. 12 Parameter Bias / Conservatism Basis Letdown Automatic RCS inventory (( control is disabled }}2(a)(c) PZR heaters Varied, lost on loss of power. ((
}}2(a)(c)
RSV lift setpoint Biased to the high condition. ((
}}2(a)(c)
SG tube plugging Varied ((
}}2(a)(c)
Break size / location Varied. ((
}}2(a)(c)
RCS Temperature Control Automatic rod control Disabled. (( Boron concentration Not credited.
}}2(a)(c)
PZR Pressure Control PZR spray (normal) Varied. (( (bypass) Nominal. PZR heaters (non-prop.) Varied. (prop.) Nominal.
}}2(a)(c)
PZR Level Control Charging Not credited. (( Letdown Disabled.
}}2(a)(c)
Steam Pressure Control Turbine throttle valves Enabled. (( Turbine bypass valves Disabled.
}}2(a)(c)
© Copyright 20187 by NuScale Power, LLC 374
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Rev. 12 Parameter Bias / Conservatism Basis Feedwater and Turbine Load Control feedwater pump speed Disabled. ((
}}2(a)(c)
CNV Pressure Control CNV evacuation system Enabled. ((
}}2(a)(c)
- 1. (( }}2(a)(c)
Sensitivity studies are performed as needed to identify the limiting response(s) for the acceptance criteria parameter(s) challenged by the event (i.e., system pressures for overheating events, MCHFR for overcooling events). Consequently, sensitivity studies are performed to identify cases with the highest pressure responses for this overheating event. For example, a sensitivity study is performed to identify the highest primary and secondary side pressures, varying primary conditions, break size and location, steam generator heat transfer and pressure, etc. (other parameters are biased as indicated in Table 7-56). Representative results of this sensitivity study are presented in Table 7-57. Unless otherwise noted, all breaks are located at the bioshield, all RCS flows are initialized at the minimum value, and losses of normal AC power occur at event initiation. The results of the sensitivity studies indicate that peak primary pressures occur during a break coincident with a loss of AC power; under these conditions the lowest setpoint RSV operates to mitigate peak pressurization and initial condition biasing contributions are secondary compared to the pressure response to the total loss of heat sink. With respect to maximum secondary side SG pressure, maximum SG pressure occurs during a 10 percent split break just outside of containment coincident with a loss of AC power when initial SG pressure is biased to the high condition and the nonsafety-related check valve fails to seat upon the onset of flow reversal (FWIVs subsequently close as expected). © Copyright 20187 by NuScale Power, LLC 375
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Rev. 12 Table 7-60 Initial conditions, biases, and conservatisms - uncontrolled control rod bank withdrawal from subcritical or low power startup conditions Parameter Bias / Conservatism Basis Initial reactor power Varied. ((
}}2(a)(c)
Initial RCS average temperature Nominal. ((
}}2(a)(c)
Initial RCS flow rate Biased to the low condition. ((
}}2(a)(c)
Initial PZR pressure Nominal. ((
}}2(a)(c)
Initial PZR level Nominal. ((
}}2(a)(c)
Initial feedwater temperature Nominal. ((
}}2(a)(c)
Initial fuel temperature Nominal. ((
}}2(a)(c)
Moderator temperature Most positive (( coefficientMTC
}}2(a)(c)
Kinetics Biased to BOC conditions. ((
}}2(a)(c)
Automatic rod control Disabled. ((
}}2(a)(c)
Decay heat Biased to the high condition. ((
}}2(a)(c)
(1) Initial SG pressure Nominal. ((
}}2(a)(c)
SG heat transfer Nominal. ((
}}2(a)(c)
PZR spray Disabled. ((
}}2(a)(c)
Letdown Nominal. ((
}}2(a)(c)
© Copyright 20187 by NuScale Power, LLC 380
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Rev. 12 Parameter Bias / Conservatism Basis PZR heaters Disabled. ((
}}2(a)(c)
RSV lift setpoint Biased to the high condition. ((
}}2(a)(c)
SG tube plugging Biased to the low condition. ((
}}2(a)(c)
Reactivity insertion rate Varied. (( Maximum
}}2(a)(c)
RCS Temperature Control Automatic rod control N/A. (( Boron concentration Not credited.
}}2(a)(c)
PZR Pressure Control PZR spray (normal) Disabled. (( (bypass) Disabled. PZR heaters (non-prop.) Disabled. (prop.) Nominal.
}}2(a)(c)
PZR Level Control Charging Not credited. (( Letdown Nominal.
}}2(a)(c)
© Copyright 20187 by NuScale Power, LLC 381
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Rev. 12 Parameter Bias / Conservatism Basis Steam Pressure Control Turbine throttle valves N/A. (( Turbine bypass valves Enabled.
}}2(a)(c)
Feedwater and Turbine Load Control feedwater pump speed N/A. ((
}}2(a)(c)
CNV Pressure Control CNV evacuation system Enabled. ((
}}2(a)(c) 1 (( }}2(a)(c)
Sensitivity studies are performed as needed to identify the limiting response(s) for the acceptance criteria parameter(s) challenged by the event. Consequently, a sensitivity study is generally performed to identify cases for lowest MCHFR and highest fuel centerline temperature for this reactivity event. As an example, the initial core power, reactivity insertion rate, and initial core inlet temperature were varied to achieve the results presented in Table 7-61. These results demonstrate the lack of challenging MCHFR values predicted for this event by NRELAP5, which is further supported by the values predicted with the approved subchannel methodology. ((
}}2(a),(c)
© Copyright 20187 by NuScale Power, LLC 382
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Rev. 12 Acceptance Criteria Discussion Containment integrity Containment integrity is evaluated by a separate analysis methodology. Escalation of an AOO to an accident This criterion is satisfied by demonstrating stable RCS flow rates and constant or downward trending RCS and DHRS pressures and temperatures exist at the end of the transient, all acceptance criteria evaluated in the transient analysis are met, and shutdown margin is maintained at the end of the transient. RCS conditions during extended DHRS cooling are addressed in a separate analysis. 7.2.14.3 Biases, Conservatisms, and Sensitivity Studies The biases and conservatisms presented in Table 7-64 are considered in order to identify the bounding transient simulation for MCHFR and maximum fuel centerline temperature. Table 7-64 Initial conditions, biases, and conservatisms - uncontrolled control rod bank withdrawal at power Parameter Bias / Conservatism Basis Initial reactor power Varied. (( RTP biased upwards to account for measurement }}2(a)(c) uncertainty. Initial RCS average temperature Varied. (( Biased to the high condition.
}}2(a)(c)
Initial RCS flow rate Biased to the low condition. ((
}}2(a)(c)
Initial PZR pressure Varied. ((
}}2(a)(c)
Initial PZR level Varied. ((
}}2(a)(c)
© Copyright 20187 by NuScale Power, LLC 386
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Rev. 12 Parameter Bias / Conservatism Basis Initial feedwater temperature Nominal. ((
}}2(a)(c)
Initial fuel temperature Biased to the low condition. ((
}}2(a)(c)
Moderator temperature Most positive (( coefficientMTC
}}2(a)(c)
Kinetics Biased to BOC conditions. ((
}}2(a)(c)
Automatic rod control Disabled. ((
}}2(a)(c)
Decay heat Biased to the high condition. ((
}}2(a)(c)
(1) Initial SG pressure Nominal. ((
}}2(a)(c)
SG heat transfer Nominal. ((
}}2(a)(c)
PZR spray Varied. ((
}}2(a)(c)
Letdown Varied. ((
}}2(a)(c)
PZR heaters Varied. ((
}}2(a)(c)
RSV lift setpoint Biased to the high condition. ((
}}2(a)(c)
SG tube plugging Biased to the low condition. ((
}}2(a)(c)
Reactivity insertion rate Varied (( Maximum
}}2(a)(c)
© Copyright 20187 by NuScale Power, LLC 387
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Rev. 12 Parameter Bias / Conservatism Basis RCS Temperature Control Automatic rod control N/A. (( Boron concentration Not credited.
}}2(a)(c)
PZR Pressure Control PZR spray (normal) Varied. (( (bypass) Nominal. PZR heaters (non-prop.) Varied. (prop.) Nominal.
}}2(a)(c)
PZR Level Control Charging Not credited. (( Letdown Varied.
}}2(a)(c)
Steam Pressure Control Turbine throttle valves Enabled. (( Turbine bypass valves Disabled.
}}2(a)(c)
© Copyright 20187 by NuScale Power, LLC 388
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Rev. 12 Parameter Bias / Conservatism Basis Feedwater and Turbine Load Control feedwater pump speed Enabled. ((
}}2(a)(c)
CNV Pressure Control CNV evacuation system Enabled. ((
}}2(a)(c)
- 1. (( }}2(a)(c)
Sensitivity studies are performed as needed to identify the limiting response(s) for the acceptance criteria parameter(s) challenged by the event. Consequently, a sensitivity study is generally performed to identify cases for lowest MCHFR for this reactivity event. As an example, the initial core power, reactivity insertion rate, pressurizer pressure, pressurizer level, reactor coolant average temperature, pressurizer spray flow, pressurizer heater status, letdown status, and loss of power were varied to achieve the results presented in Table 7-65. These results demonstrate the MCHFR occurs at initial power levels of 75 percent RTP and above, with AC power available for the event duration, the RCS average temperature biased low, and the spray flow rate tuned to delay reactor trip on high pressurizer pressure. © Copyright 20187 by NuScale Power, LLC 389
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Rev. 12 Table 7-68 Initial conditions, biases, and conservatisms - control rod misoperation, single control rod assembly withdrawal Parameter Bias / Conservatism Basis Initial reactor power Varied. (( RTP biased upwards to account for measurement }}2(a)(c) uncertainty. Initial RCS average temperature Varied. (( Biased to the high condition.
}}2(a)(c)
Initial RCS flow rate Biased to the low condition. ((
}}2(a)(c)
Initial PZR pressure Varied. ((
}}2(a)(c)
Initial PZR level Varied. ((
}}2(a)(c)
Initial feedwater temperature Nominal. ((
}}2(a)(c)
Initial fuel temperature Biased to the low condition. ((
}}2(a)(c)
Moderator temperature Most positive (( coefficientMTC
}}2(a)(c)
Kinetics Biased to BOC conditions. ((
}}2(a)(c)
Automatic rod control Disabled. ((
}}2(a)(c)
Decay heat Biased to the high condition. ((
}}2(a)(c)
(1) Initial SG pressure Nominal. ((
}}2(a)(c)
© Copyright 20187 by NuScale Power, LLC 395
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Rev. 12 Parameter Bias / Conservatism Basis SG heat transfer Nominal. ((
}}2(a)(c)
PZR spray Varied. ((
}}2(a)(c)
Letdown Varied. ((
}}2(a)(c)
PZR heaters Varied. ((
}}2(a)(c)
RSV lift setpoint Biased to the high condition. ((
}}2(a)(c)
SG tube plugging Biased to the low condition. ((
}}2(a)(c)
Reactivity insertion rate Varied. (( Maximum
}}2(a)(c)
RCS Temperature Control Automatic rod control N/A. (( Boron concentration Not credited.
}}2(a)(c)
© Copyright 20187 by NuScale Power, LLC 396
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Rev. 12 Parameter Bias / Conservatism Basis PZR Pressure Control PZR spray (normal) Varied. (( (bypass) Nominal. PZR heaters (non-prop.) Varied. (prop.) Varied.
}}2(a)(c)
PZR Level Control Charging Not credited. (( Letdown Varied.
}}2(a)(c)
Steam Pressure Control Turbine throttle valves Enabled. (( Turbine bypass valves Disabled.
}}2(a)(c)
Feedwater and Turbine Load Control feedwater pump speed Enabled. ((
}}2(a)(c)
© Copyright 20187 by NuScale Power, LLC 397
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Rev. 12 Parameter Bias / Conservatism Basis CNV Pressure Control CNV evacuation system Enabled. ((
}}2(a)(c)
- 1. (( }}2(a)(c)
Sensitivity studies are performed as needed to identify the limiting response(s) for the acceptance criteria parameter(s) challenged by the event. Consequently, a sensitivity study is generally performed to identify cases for lowest MCHFR for this reactivity event. As an example, the initial core power, reactivity insertion rate, pressurizer pressure, pressurizer level, reactor coolant average temperature, pressurizer spray flow, pressurizer heater status, letdown status, and loss of power were varied to achieve the results presented in Table 7-69 for the withdrawal of a single CRA event. These results demonstrate the MCHFR occurs at an initial power level of 75 percent RTP, with AC power available for the event duration, the RCS average temperature biased low, the pressurizer pressure biased low, the pressurizer level biased low, and the spray flow rate tuned to delay reactor trip on high pressurizer pressure. © Copyright 20187 by NuScale Power, LLC 398
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Rev. 12 ((
}}2(a),(c)
Dropping One or More CRAs The biases and conservatisms presented in Table 7-70 are considered in order to identify the bounding transient simulation for MCHFR and maximum fuel centerline temperature for the dropped CRA(s) event. Table 7-70 Initial conditions, biases, and conservatisms - control rod misoperation, dropped control rod assemblies Parameter Bias / Conservatism Basis Initial reactor power RTP biased upwards to (( account for measurement uncertainty. }}2(a)(c) Initial RCS average temperature Biased to the high condition. ((
}}2(a)(c)
Initial RCS flow rate Biased to the low condition. ((
}}2(a)(c)
Initial PZR pressure Biased to the high condition. ((
}}2(a)(c)
Initial PZR level Biased to the high (( conditionNominal.
}}2(a)(c)
Initial feedwater temperature Nominal. ((
}}2(a)(c)
Initial fuel temperature Nominal. ((
}}2(a)(c) © Copyright 20187 by NuScale Power, LLC 400
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Rev. 12 Parameter Bias / Conservatism Basis Moderator temperature Varied. (( coefficientMTC
}}2(a)(c)
Kinetics Varied. ((
}}2(a)(c)
Automatic rod control Enabled. ((
}}2(a)(c)
Decay heat Biased to the high condition. ((
}}2(a)(c)
Initial SG pressure Nominal. ((
}}2(a)(c)
SG heat transfer Nominal. ((
}}2(a)(c)
PZR spray Nominal. ((
}}2(a)(c)
Letdown Nominal. ((
}}2(a)(c)
PZR heaters Nominal. ((
}}2(a)(c)
RSV lift setpoint Biased to the high condition. ((
}}2(a)(c)
SG tube plugging Biased to the low condition. ((
}}2(a)(c)
Dropped CRA worth Minimum ((
}}2(a)(c)
RCS Temperature Control Automatic rod control Enabled. (( Boron concentration Not credited.
}}2(a)(c)
© Copyright 20187 by NuScale Power, LLC 401
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Rev. 12 Parameter Bias / Conservatism Basis PZR Pressure Control PZR spray (normal) Nominal. (( (bypass) Nominal. PZR heaters (non-prop.) Nominal. (prop.) Nominal.
}}2(a)(c)
PZR Level Control Charging Not credited. (( Letdown Nominal.
}}2(a)(c)
Steam Pressure Control Turbine throttle valves Enabled. (( Turbine bypass valves Disabled.
}}2(a)(c)
Feedwater and Turbine Load Control feedwater pump speed Enabled. ((
}}2(a)(c)
CNV Pressure Control CNV evacuation system Enabled. ((
}}2(a)(c)
Sensitivity studies are performed as needed to identify the limiting response(s) for the acceptance criteria parameter(s) challenged by the event. Consequently, a sensitivity study is generally performed to identify cases for lowest MCHFR for this reactivity event. As an example, the initial core power, dropped CRA worth, and core time-in-life were © Copyright 20187 by NuScale Power, LLC 402
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Rev. 12 7.2.16.3 Biases, Conservatisms, and Sensitivity Studies The biases and conservatisms presented in Table 7-74 are considered in order to identify the bounding conditions for MCHFR, which coincide with the quickest loss of shutdown margin. Table 7-74 Initial conditions, biases, and conservatisms - inadvertent decrease in boron concentration Parameter Bias / Conservatism Basis Initial reactor power Excluded. Not part of mixing model. Initial RCS average temperature Biased to the high condition. ((
}}2(a)(c)
Initial RCS flow rate Biased to the low condition. ((
}}2(a)(c)
Initial PZR pressure Nominal. ((
}}2(a)(c)
Initial PZR level Excluded. Not part of mixing model. Initial feedwater temperature Excluded. Not part of mixing model. Initial fuel temperature Excluded. Not part of mixing model. Moderator temperature Excluded. Not part of mixing model. coefficientMTC Kinetics Biased to BOC conditions. ((
}}2(a)(c)
Automatic rod control Excluded. Not part of mixing model. Decay heat Excluded. Not part of mixing model. Initial SG pressure(1) Excluded. Not part of mixing model. SG heat transfer Excluded. Not part of mixing model. PZR spray Excluded. Not part of mixing model. Letdown Biased to the high condition. ((
}}2(a)(c)
PZR heaters Excluded. Not part of mixing model. RSV lift setpoint Excluded. Not part of mixing model. SG tube plugging Excluded. Does not alter active RCS volume. Shutdown margin Biased to the low condition. ((
}}2(a)(c)
Initial boron concentration Biased to the high condition (( }}2(a)(c) Boron worth Biased to the high condition. (( }}2(a)(c) Active RCS volume Biased to the low condition. ((
}}2(a)(c)
Makeup flow rate Biased to the high condition. ((
}}2(a)(c)
© Copyright 20187 by NuScale Power, LLC 409
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Rev. 12 Parameter Bias / Conservatism Basis Makeup temperature Biased to the low condition. ((
}}2(a)(c)
RCS Temperature Control Automatic rod control Excluded. Not part of mixing model. Boron concentration Not credited. ((
}}2(a)(c)
PZR Pressure Control PZR spray (normal) Excluded. Not part of mixing model. (bypass) Excluded. Not part of mixing model. PZR heaters (non-prop.) Excluded. Not part of mixing model. (prop.) Excluded. Not part of mixing model. PZR Level Control Charging Enabled. (( Letdown Enabled.
}}2(a)(c)
Steam Pressure Control Turbine throttle valves Excluded. Not part of mixing model. Turbine bypass valves Excluded. Not part of mixing model. Feedwater and Turbine Load Control feedwater pump speed Excluded. Not part of mixing model. CNV Pressure Control CNV evacuation system Excluded. Not part of mixing model. ((
}}2(a),(c)
© Copyright 20187 by NuScale Power, LLC 410
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Rev. 12 7.2.17.2 Acceptance Criteria Evaluation of the most challenging case(s) relative to the acceptance criteria is presented in Table 7-81. Table 7-81 Acceptance criteria - reactor coolant system inventory increase Acceptance Criteria Discussion Primary pressure Primary pressure rises to the peak value, then drops as the lowest setpoint RSV lifts to reduce pressure. Secondary pressure Secondary pressure increases rapidly to the peak value upon turbine trip, then decreases as the plant cools down via DHR. Critical heat flux ratio This criterion is evaluated by downstream subchannel analysis. Maximum fuel centerline temperature This criterion is evaluated by downstream subchannel analysis. Containment integrity Containment integrity is evaluated by a separate analysis methodology. Escalation of an AOO to an accident This criterion is satisfied by demonstrating stable RCS flow rates and constant or downward trending RCS and DHRS pressures and temperatures exist at the end of the transient, all acceptance criteria evaluated in the transient analysis are met, and shutdown margin is maintained at the end of the transient. RCS conditions during extended DHRS cooling are addressed in a separate analysis. 7.2.17.3 Biases, Conservatisms, and Sensitivity Studies The biases and conservatisms presented in Table 7-82 are considered in identifying the bounding transient simulation for primary pressure. Table 7-82 Initial conditions, biases, and conservatisms - reactor coolant system inventory increase Parameter Bias / Conservatism Basis Initial reactor power RTP biased upwards to (( account for measurement }}2(a)(c) uncertainty. Initial RCS average temperature Varied. ((
}}2(a)(c)
Initial RCS flow rate Varied. ((
}}2(a)(c)
© Copyright 20187 by NuScale Power, LLC 415
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Rev. 12 Parameter Bias / Conservatism Basis Initial PZR pressure Varied. ((
}}2(a)(c)
Initial PZR level Varied. ((
}}2(a)(c)
Initial feedwater temperature Nominal. ((
}}2(a)(c)
Initial fuel temperature Biased to the low condition ((
}}2(a)(c)
Moderator temperature Consistent with EOC kinetics. (( coefficientMTC }}2(a)(c) Kinetics Biased to EOC conditions. ((
}}2(a)(c)
Automatic rod control Enabled. ((
}}2(a)(c)
Decay heat Biased to the high condition. ((
}}2(a)(c)
(1) Initial SG pressure Nominal. ((
}}2(a)(c)
Steam generator heat transfer Nominal. ((
}}2(a)(c)
PZR spray Varied ((
}}2(a)(c)
Letdown Letdown is disabled ((
}}2(a)(c)
PZR heaters Nominal. ((
}}2(a)(c)
RSV lift setpoint Biased to the high condition. ((
}}2(a)(c)
SG tube plugging Biased to the low condition. ((
}}2(a)(c)
Makeup temperature Varied ((
}}2(a)(c)
© Copyright 20187 by NuScale Power, LLC 416
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Rev. 12 Parameter Bias / Conservatism Basis RCS Temperature Control Automatic rod control Enabled. (( Boron concentration Not credited.
}}2(a)(c)
PZR Pressure Control PZR spray (normal) Varied. (( (bypass) Nominal. PZR heaters (non-prop.) Nominal. (prop.) Nominal.
}}2(a)(c)
PZR Level Control Charging N/A. Not applicable to this event. (( Letdown Disabled. }}2(a)(c) Steam Pressure Control Turbine throttle valves Enabled. (( Turbine bypass valves Disabled.
}}2(a)(c)
Feedwater and Turbine Load Control Feedwater pump speed Enabled. ((
}}2(a)(c)
© Copyright 20187 by NuScale Power, LLC 417
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Rev. 12 Parameter Bias / Conservatism Basis CNV Pressure Control CNV evacuation system Enabled. ((
}}2(a)(c)
- 1. (( }}2(a)(c)
- 2. These inputs, in conjunction with least negative Doppler temperature coefficient, are selected to maximize the power response (if any) induced by the addition of colder CVCS water. However, since this event is driven by mass addition, reactivity effects prior to RTS actuation (if any) are small when compared to the pressurization associated with the increase in primary inventory.
Sensitivity studies are performed as needed to identify the limiting response(s) for the acceptance criteria parameter(s) challenged by the event, including cases with the highest pressure responses for this inventory increase event. For example, a sensitivity study is performed to identify the highest primary and secondary side pressures, varying primary conditions, makeup temperature, spray availability (other parameters are biased as indicated in Table 7-82). Representative results of this sensitivity study are presented in Table 7-83. The results of the sensitivity studies indicate that peak primary pressure occurs when initial primary temperature and pressure are biased to the low condition, and makeup temperature is biased to the high condition as the RTS actuates on high PZR pressure and the lowest-setpoint RSV operates to mitigate peak pressurization. With respect to maximum secondary side SG pressure, maximum SG pressure occurs when nonsafety-related spray is used to delay the RTS actuation until high PZR level is reached. © Copyright 20187 by NuScale Power, LLC 418
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Rev. 12 Acceptance Criteria Discussion Consequential loss of system functionality This criterion is satisfied by demonstrating stable RCS flow rates and constant or downward trending RCS and DHRS pressures and temperatures exist at the end of the transient, the water level in the RPV remains above the top of the core throughout the transient, all acceptance criteria evaluated in the transient analysis are met, and shutdown margin is maintained at the end of the transient. Radiological consequences The radiological consequences acceptance criteria are evaluated by downstream radiological analysis using the mass release calculated in the non-LOCA transient analysis. 7.2.18.3 Biases, Conservatisms, and Sensitivity Studies The biases and conservatisms indicated in Table 7-86 are considered in identifying a bounding transient simulation for dose. Table 7-86 Initial conditions, biases, and conservatisms - breaks in small lines carrying primary coolant outside containment Parameter Bias / Conservatism Basis Initial reactor power Varied. ((
}}2(a)(c)
Initial RCS average temperature Biased to the high condition. ((
}}2(a)(c)
Initial RCS flow rate Biased to the low condition. ((
}}2(a)(c)
Initial PZR pressure Biased to the high condition ((
}}2(a)(c)
Initial PZR level Biased to the high condition. ((
}}2(a)(c)
Initial feedwater temperature Biased to the high condition. ((
}}2(a)(c)
Initial fuel temperature Nominal. ((
}}2(a)(c)
© Copyright 20187 by NuScale Power, LLC 424
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Rev. 12 Parameter Bias / Conservatism Basis Moderator temperature Varied. (( coefficientMTC
}}2(a)(c)
Kinetics Varied. ((
}}2(a)(c)
Automatic rod control Disabled. ((
}}2(a)(c)
Decay heat Biased to the high condition. ((
}}2(a)(c)
(1) Initial SG pressure Biased to the high condition. ((
}}2(a)(c)
SG heat transfer Nominal. ((
}}2(a)(c)
PZR spray Disabled. ((
}}2(a)(c)
Letdown Disabled. ((
}}2(a)(c)
PZR heaters Nominal. ((
}}2(a)(c)
RSV lift setpoint Biased to the high condition. ((
}}2(a)(c)
SG tube plugging Biased to the low condition. ((
}}2(a)(c)
CVCS volume outside Biased to the high condition. (( containment
}}2(a)(c)
RCS Temperature Control Automatic rod control Disabled. (( Boron concentration Not credited.
}}2(a)(c)
© Copyright 20187 by NuScale Power, LLC 425
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Rev. 12 Parameter Bias / Conservatism Basis PZR Pressure Control PZR spray (normal) Disabled. (( (bypass) Nominal. PZR heaters (non-prop.) Nominal. (prop.) Nominal.
}}2(a)(c)
PZR Level Control Charging Not credited. (( Letdown Disabled.
}}2(a)(c)
Steam Pressure Control Turbine throttle valves Enabled. (( Turbine bypass valves Disabled.
}}2(a)(c)
Feedwater and Turbine Load Control feedwater pump speed Enabled. ((
}}2(a)(c)
CNV Pressure Control CNV evacuation system Enabled. ((
}}2(a)(c)
- 1. (( }}2(a)(c)
© Copyright 20187 by NuScale Power, LLC 426
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Rev. 12 Table 7-91 Initial conditions, biases, and conservatisms - steam generator tube failure Parameter Bias / Conservatism Basis Initial reactor power RTP biased to the high (( condition. }}2(a)(c) Initial RCS average temperature Varied. ((
}}2(a)(c)
Initial RCS flow rate Biased to the low condition. ((
}}2(a)(c)
Initial PZR pressure Biased to the high condition ((
}}2(a)(c)
Initial PZR level Biased to the high condition. ((
}}2(a)(c)
Initial feedwater temperature Varied. ((
}}2(a)(c)
Initial fuel temperature Biased to the high condition. ((
}}2(a)(c)
Moderator temperature Varied. (( coefficientMTC
}}2(a)(c)
Kinetics Varied. ((
}}2(a)(c)
Automatic rod control Disabled. ((
}}2(a)(c)
Decay heat Biased to the high condition. ((
}}2(a)(c)
© Copyright 20187 by NuScale Power, LLC 434
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Rev. 12 Parameter Bias / Conservatism Basis Initial SG pressure(1) Varied. ((
}}2(a)(c)
SG heat transfer Nominal. ((
}}2(a)(c)
PZR spray Disabled. ((
}}2(a)(c)
Letdown Disabled. ((
}}2(a)(c)
PZR heaters Nominal. ((
}}2(a)(c)
RSV lift setpoint Biased to the high condition. ((
}}2(a)(c)
SG tube plugging Varied. ((
}}2(a)(c)
RCS Temperature Control Automatic rod control Disabled. (( Boron concentration Not credited.
}}2(a)(c)
© Copyright 20187 by NuScale Power, LLC 435
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Rev. 12 Parameter Bias / Conservatism Basis PZR Pressure Control PZR spray (normal) Disabled. (( (bypass) Disabled. PZR heaters (non-prop.) Nominal. (prop.) Nominal.
}}2(a)(c)
PZR Level Control Charging Not credited. (( Letdown Disabled.
}}2(a)(c)
Steam Pressure Control Turbine throttle valves Enabled. (( Turbine bypass valves Disabled.
}}2(a)(c)
Feedwater and Turbine Load Control feedwater pump speed Enabled. ((
}}2(a)(c)
© Copyright 20187 by NuScale Power, LLC 436
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Rev. 12 Parameter Bias / Conservatism Basis CNV Pressure Control CNV evacuation system Enabled. ((
}}2(a)(c)
- 1. (( }}2(a)(c)
A sensitivity study is performed as needed to identify the most challenging break characteristics, core time-in-life, MS pressure, SG tube plugging, loss of power, single failure, feedwater temperature, and reactor coolant temperature for a SGTF event with respect to total mass released and iodine spiking duration. Representative results for this study are presented in Table 7-92. These results indicate the mass release and spiking time are maximized with a DEG break (for mass released) or a smaller split break (for spiking time) at the top of the SG tube, AC power available for the event duration, and a single active failure of the primary MSIV on the affected SG to close. © Copyright 20187 by NuScale Power, LLC 437
Response to Request for Additional Information Docket No. 52-048 eRAI No.: 9374 Date of RAI Issue: 05/09/2018 NRC Question No.: 15.00.02-29 TR-0516-49416-P supports the conclusions in the NuScale FSAR, which under 10 CFR 52.47 must describe the facility, present the design bases and the limits on its operation, and present a safety analysis of the structures, systems, and components and of the facility as a whole. RG 1.203 describes the EMDAP, which the NRC staff considers acceptable for use in developing and assessing EMs used to analyze transient and accident behavior. Step 18 of the EMDAP discusses preparation of input and performing of calculations to assess system interactions: The ability of the EM to model system interactions should also be evaluated in this step, and plant input decks should be prepared for the target applications. Sufficient analyses should be performed to determine parameter ranges expected in the nuclear power plant. The system response to a loss of normal feedwater described in TR-0516-49416-P, Section 8.2.1, and the response to a feedwater line break described in TR-0516-49416-P, Section 8.2.3, are difficult to discern due to the scale of the figures. In particular, the time scales extend to 2500 to 3500 seconds. In order for the staff to understand the transient behavior, figures on different time scales are necessary. Information Requested:
- a. Provide Figures 8-29 through 8-36 on a 0 to 100 second time scale.
- b. Provide Figures 8-38 through 8-46 on a 700 to 900 second time scale
- c. Provide Figures 8-55, 8-56, 8-59, and 8-61 on a 0 to 500 second time scale.
NuScale Response: Response to RAI, part a The TR-0516-49416-P figures with the requested scale are provided below. In addition, Figure NuScale Nonproprietary
8-37 of TR-0516-49416-P is provided as Figure a.9 to show the feedwater flow response. TR-0516-49416-P Part a Response Figure Figure 8-29 a.1 8-30 a.2 8-31 a.3 8-32 a.4 8-33 a.5 8-34 a.6 8-35 a.7 8-36 a.8 8-37 a.9 Figure a.1 Reactor pressure vessel pressure response for the representative loss of normal feedwater flow event - reactor coolant system pressure case NuScale Nonproprietary
Figure a.2 Pressurizer level for the representative loss of normal feedwater flow event - reactor coolant system pressure case Figure a.3 Steam generator 2 pressure response for the representative loss of normal feedwater flow event - reactor coolant system pressure case NuScale Nonproprietary
Figure a.4 Power response for the representative loss of normal feedwater flow event - reactor coolant system pressure case Figure a.5 Reactor coolant system flow rate for the representative loss of normal feedwater flow event - reactor coolant system pressure case NuScale Nonproprietary
Figure a.6 Core inlet temperature for the representative loss of normal feedwater flow event - reactor coolant system pressure case Figure a.7 Core outlet temperature for the representative loss of normal feedwater flow event - reactor coolant system pressure case NuScale Nonproprietary
Figure a.8 Net reactivity for the representative loss of normal feedwater flow event - reactor coolant system pressure case Figure a.9 Steam generator 2 secondary flow for the representative loss of normal feedwater flow event - reactor coolant system pressure case Response to RAI, part b NuScale Nonproprietary
The updated figures with the requested scale are provided below. TR-0516-49416-P Part b Response Figure Figure 8-38 b.1 8-39 b.2 8-40 b.3 8-41 b.4 8-42 b.5 8-43 b.6 8-44 b.7 8-45 b.8 8-46 b.9 Figure b.1 Core inlet temperature for the representative loss of normal feedwater flow event - secondary pressure case NuScale Nonproprietary
Figure b.2 Core outlet temperature for the representative loss of normal feedwater flow event - secondary pressure case Figure b.3 Reactor pressure vessel pressure response for the representative loss of normal NuScale Nonproprietary
feedwater flow event - secondary pressure case Figure b.3 Pressurizer level for the representative loss of normal feedwater flow event - secondary pressure case NuScale Nonproprietary
Figure b.5 Steam generator 2 pressure response for the representative loss of normal feedwater flow event - secondary pressure case Figure b.6 Power response for the representative loss of normal feedwater flow event - secondary pressure case NuScale Nonproprietary
Figure b.7 Reactor coolant system flow rate for the representative loss of normal feedwater flow event - secondary pressure case Figure b.8 Net reactivity for the representative loss of normal feedwater flow event - secondary pressure case NuScale Nonproprietary
Figure b.9 Steam generator 2 secondary flow for the representative loss of normal feedwater flow event - secondary pressure case Response to RAI, part c The updated figures with the requested scale are provided below. TR-0516-49416-P Part c Response Figure Figure 8-55 c.1 8-56 c.2 8-59 c.3 8-61 c.4 NuScale Nonproprietary
Figure c.1 Primary temperature response for the representative feedwater line break event NuScale Nonproprietary
Figure c.2 System pressure response for the representative feedwater line break event Figure c.3 Decay heat removal system response for the representative feedwater line break event NuScale Nonproprietary
Figure c.4 Reactor coolant system flow response for the representative feedwater line break event Impact on DCA: There are no impacts to the DCA as a result of this response. NuScale Nonproprietary
Response to Request for Additional Information Docket: PROJ0769 eRAI No.: 9374 Date of RAI Issue: 05/09/2018 NRC Question No.: 15.00.02-30 SRP Section 15.0.2 provides the staff guidance on reviewing analytical models and computer codes used to analyze transient and accident behavior. SRP Section 15.0.2 states that the evaluation model documentation must be scrutable, complete, unambiguous, accurate, and reasonably self-contained. During an audit discussion (Round 2, Issue 13), the applicant stated that an error was found in the loss of containment vacuum (LOCV) calculation described in TR-0516-49416-P, Section 7.2.5. The peak pressure case did not include ((
}}2(a),(c)
Information Requested: To ensure accurate documentation, please provide a revised LOCV analysis, and incorporate the results in a revision of TR-0516-49416-P. NuScale Response: The case in question (Round 2 Issue 13) is a loss of condenser vacuum (LOCV) event (Section 7.2.7 of TR-0516-49416-P) rather than a loss of containment vacuum event (Section 7.2.5 of TR-0516-49416-P). The event was evaluated including a loss of feedwater. The updated case results were used to revise Table 7-37 of TR-0516-49416-P. No other changes were necessary to Section 7.2.7 of TR-0516-49416-P to reflect the updated case results. Impact on Topical Report: Topical Report TR-0516-49416, Non-Loss of Coolant Accident Analysis Methodology, has been revised as described in the response above and as shown in the markup provided in this response. NuScale Nonproprietary
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Rev. 12 Table 7-37 Representative sensitivity studies - loss of condenser vacuum ((
+
}}2(a),(c) 7.2.8 Main Steam Isolation Valve(s) Closure The methodology used to simulate a postulated main steam isolation valve closure for the NPM, and an evaluation of the resulting representative plant response against the acceptance criteria for an AOO listed in Table 7-4 are presented below.
A description of the event including biases and conservatisms, sensitivity studies, single active failure (SAF) and loss of power (LOP) scenarios, challenging case, and acceptance criteria evaluation are presented in the following sections. 7.2.8.1 General Event Description A closure of one or both main steam isolation valves results in a pressurization of the secondary system and overheating of the RCS. Rising secondary side pressure results in a rapid RTS actuation on the high steam pressure signal. Reactor trip and transition to stable DHRS flow terminates the transient with the NPM in a safe, stable condition. The relevant acceptance criteria, single active failure, and loss of power scenarios are listed in Table 7-38. The MSIV closure event can occur when one or both MSIVs close unexpectedly. The limiting pressure responses occur when the event is initiated from full power conditions, and the initial conditions are biased in the conservative directions. Sensitivity studies on number of MSIVs closing, initial primary temperature and primary/secondary pressures are performed to identify the conditions that maximize peak primary and secondary pressures. Additional sensitivity studies are performed on other parameters, as necessary, to identify the case(s) with the potentially limiting peak primary and secondary pressures.
© Copyright 20187 by NuScale Power, LLC 344
RAIO-0718-60793 : Affidavit of Zackary W. Rad, AF-0718-60794 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvalis, Oregon 97330, Office: 541.360.0500, Fax: 541.207.3928 www.nuscalepower.com
NuScale Power, LLC AFFIDAVIT of Zackary W. Rad I, Zackary W. Rad, state as follows:
- 1. I am the Director, Regulatory Affairs of NuScale Power, LLC (NuScale), and as such, I have been specifically delegated the function of reviewing the information described in this Affidavit that NuScale seeks to have withheld from public disclosure, and am authorized to apply for its withholding on behalf of NuScale.
- 2. I am knowledgeable of the criteria and procedures used by NuScale in designating information as a trade secret, privileged, or as confidential commercial or financial information. This request to withhold information from public disclosure is driven by one or more of the following:
- a. The information requested to be withheld reveals distinguishing aspects of a process (or component, structure, tool, method, etc.) whose use by NuScale competitors, without a license from NuScale, would constitute a competitive economic disadvantage to NuScale.
- b. The information requested to be withheld consists of supporting data, including test data, relative to a process (or component, structure, tool, method, etc.), and the application of the data secures a competitive economic advantage, as described more fully in paragraph 3 of this Affidavit.
- c. Use by a competitor of the information requested to be withheld would reduce the competitor's expenditure of resources, or improve its competitive position, in the design, manufacture, shipment, installation, assurance of quality, or licensing of a similar product.
- d. The information requested to be withheld reveals cost or price information, production capabilities, budget levels, or commercial strategies of NuScale.
- e. The information requested to be withheld consists of patentable ideas.
- 3. Public disclosure of the information sought to be withheld is likely to cause substantial harm to NuScale's competitive position and foreclose or reduce the availability of profit-making opportunities. The accompanying Request for Additional Information response reveals distinguishing aspects about the method by which NuScale develops its non-loss of coolant accident analysis methodology.
NuScale has performed significant research and evaluation to develop a basis for this method and has invested significant resources, including the expenditure of a considerable sum of money. The precise financial value of the information is difficult to quantify, but it is a key element of the design basis for a NuScale plant and, therefore, has substantial value to NuScale. If the information were disclosed to the public, NuScale's competitors would have access to the information without purchasing the right to use it or having been required to undertake a similar expenditure of resources. Such disclosure would constitute a misappropriation of NuScale's intellectual property, and would deprive NuScale of the opportunity to exercise its competitive advantage to seek an adequate return on its investment. AF-0718-60794
- 4. The information sought to be withheld is in the enclosed response to NRC Request for Additional Information No. 9374, eRAI 9374. The enclosure contains the designation "Proprietary" at the top of each page containing proprietary information. The information considered by NuScale to be proprietary is identified within double braces, "(( }}" in the document.
- 5. The basis for proposing that the information be withheld is that NuScale treats the information as a trade secret, privileged, or as confidential commercial or financial information. NuScale relies upon the exemption from disclosure set forth in the Freedom of Information Act ("FOIA"), 5 USC § 552(b)(4), as well as exemptions applicable to the NRC under 10 CFR §§ 2.390(a)(4) and 9.17(a)(4).
- 6. Pursuant to the provisions set forth in 10 CFR § 2.390(b)(4), the following is provided for consideration by the Commission in determining whether the information sought to be withheld from public disclosure should be withheld:
- a. The information sought to be withheld is owned and has been held in confidence by NuScale.
- b. The information is of a sort customarily held in confidence by NuScale and, to the best of my knowledge and belief, consistently has been held in confidence by NuScale.
The procedure for approval of external release of such information typically requires review by the staff manager, project manager, chief technology officer or other equivalent authority, or the manager of the cognizant marketing function (or his delegate), for technical content, competitive effect, and determination of the accuracy of the proprietary designation. Disclosures outside NuScale are limited to regulatory bodies, customers and potential customers and their agents, suppliers, licensees, and others with a legitimate need for the information, and then only in accordance with appropriate regulatory provisions or contractual agreements to maintain confidentiality.
- c. The information is being transmitted to and received by the NRC in confidence.
- d. No public disclosure of the information has been made, and it is not available in public sources. All disclosures to third parties, including any required transmittals to NRC, have been made, or must be made, pursuant to regulatory provisions or contractual agreements that provide for maintenance of the information in confidence.
- e. Public disclosure of the information is likely to cause substantial harm to the competitive position of NuScale, taking into account the value of the information to NuScale, the amount of effort and money expended by NuScale in developing the information, and the difficulty others would have in acquiring or duplicating the information. The information sought to be withheld is part of NuScale's technology that provides NuScale with a competitive advantage over other firms in the industry.
NuScale has invested significant human and financial capital in developing this technology and NuScale believes it would be difficult for others to duplicate the technology without access to the information sought to be withheld. I declare under penalty of perjury that the foregoing is true and correct. Executed on July 9, 2018. Zackary WW. Rad AF-0718-60794}}