ML12040A075

NEDO-33690, Revision 0, Columbia Generating Station Power Range Neutron Monitoring System Response Time Analysis Report

ML12040A075
Person / Time
Site:	Columbia
Issue date:	11/30/2011
From:	GE-Hitachi Nuclear Energy Americas
To:	Office of Nuclear Reactor Regulation
References
DRF Section 0000-0138-0226 R2 NEDO-33690, Rev. 0
Download: ML12040A075 (20)

Text

{{#Wiki_filter:LICENSE AMENDMENT REQUEST TO CHANGE TECHNICAL SPECIFICATIONS IN SUPPORT OF PRNM AND ARTS / MELLLA IMPLEMENTATION Enclosure 2 -Attachment 10 NEDO-33690, Revision 0 Columbia Generating Station Power Range Neutron Monitoring System Response Time Analysis Report November 2011 (non-proprietary version) WHITACHI GE Hitachi Nuclear Energy NEDO-33690 Revision 0 DRF Section 0000-0138-0226 R2 November 2011 Non-Proprietary Information-Class I(Public)Columbia Generating Station Power Range Neutron Monitoring System Response Time Analysis Report Copyright 2011 GE-Hitachi Nuclear Energy Americas LLC All Rights Reserved NEDO-33690 Revision 0 INFORMATION NOTICE This is a non-proprietary version of the document NEDC-33690P, Revision 0, which has the proprietary information removed. Portions of the document that have been removed are indicated by an open and closed bracket as shown here [[IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT Please Read Carefully The design, engineering, and other information contained in this document is furnished for the purpose of supporting the Columbia Generating Station license amendment request for an extended power uprate in proceedings before the U.S. Nuclear Regulatory Commission. The only undertakings of GEH with respect to information in this document are contained in the contracts between GEH and its customers or participating utilities, and nothing contained in this document shall be construed as changing that contract. The use of this information by anyone for any purpose other than that for which it is intended is not authorized; and with respect to any unauthorized use, GEH makes no representation or warranty, and assumes no liability as to the completeness, accuracy, or usefulness of the information contained in this document.ii NEDO-33690P Revision 0 TABLE OF CONTENTS Page EX ECU TIV E SU M M A RY ........................................................................................................ v 1. IN T R O D U C T IO N ............................................................................................................ .1 2. CGS PRNM RESPONSE TIME ANALYSIS .................................................................. 2 2.1 A nalysis A pproach .................................................................................................. ..2 2.2 Evaluation per BTP 7-21 Requirements .................................................................. 2 2.3 Evaluation per Staff Positions 1.19 and 1.20 of DI&C-ISG-04 ............................... 8 3.

SUMMARY

& CONCLUSIONS 

..................................................................................... 13 4 .R E F E R E N C E S ..................................................................................................................... 14 Table 2.2-1 Table 2.2-2 Table 2.2-3 Table 2.3-1 Table 2.3-2 Table 2.3-3 TABLES Safety Function Response Times ........................................................ 3 APRM Channel Flux Trip Delay Time Analysis ...................................... 6 OPRM Trip Delay Time Analysis .......................................................... 7 FDDI Link Capacity Usage .............................................................. 9 Data Error Rate Impact to PRNM System Response Time ......................... 11 Maximum Expected PRNM System Response Time ............................... 12 FIGURES Figure 2.2-1 APRM Channel Response Time Performance Block Diagram ...................... 5 iii NEDO-33690P Revision 0 ACRONYMS AND ABBREVIATIONS Term Definition APRM Average Power Range Monitor ASP Automatic Signal Processor BTP Branch Technical Position CGS Columbia Generating Station FDDI Fiber Direct Data Interface FSAR Final Safety Analysis Report GEH GE-Hitachi Nuclear Energy Americas LLC GGNS Grand Gulf Nuclear Station IEEE Institute of Electrical and Electronics Engineers I&C Instrumentation & Control ISG Interim Staff Guidance LPRM Local Power Range Monitor NUMAC Nuclear Measurement Analysis and Control NRC Nuclear Regulatory Commission NUREG Nuclear Regulatory Commission Regulation OPRM Oscillation Power Range Monitor PRNMS Power Range Neutron Monitoring System RAI Request for Additional Information RPS Reactor Protection System STP Simulated Thermal Power V&V Verification & Validation iv NEDO-33690P Revision 0 EXECUTIVE

SUMMARY

This report evaluates the Response Time of the Columbia Generating Station (CGS) Nuclear Measurement Analysis and Control (NUMAC) Power Range Neutron Monitoring System (PRNM) system versus the safety analysis requirements and standard criteria for digital instrumentation and controls. This evaluation demonstrates compliance with the criteria of Branch Technical Position (BTP) 7-21 and Staff Positions 1.19 and 1.20 of Digital Instrumentation & Control-Interim Staff Guidance (DI&C-ISG-04). V NEDO-33690P Revision 0 1. INTRODUCTION This report describes the PRNM system response time in support of the four safety-related trip signals. As defined by Section 3.3.2 of Reference 1, "The safety functions of the PRNM system are:* Average Power Range Monitor (APRM) Neutron Flux -High Trip* APRM Simulated Thermal Power (STP) -High Trip* APRM Neutron Flux -High (Setdown) Trip* Oscillation Power Range Monitor (OPRM) Instability Detect-and-Suppress Trip" The CGS PRNMS uses the OPRM Option III algorithm described in Section 3.3 of Reference 1, which is the licensing basis for the CGS NUMAC PRNM. The approach used for the response time analysis presented in this report is similar to the one previously presented to the U.S.Nuclear Regulatory Commission (NRC) for the Grand Gulf Nuclear Station PRNMS project, via responses to RAIs 18 and 20 within Reference 2.This evaluation addresses the criteria of BTP 7-21, and demonstrates compliance with these requirements. Secondly, this analysis addresses Staff Positions 1.19 and 1.20 of DI&C-ISG-04 for the CGS PRNM.Plant specific requirements for time response issues are directly addressed in table items 8.3.4.4.4, 8.4.4.4.4 and 8.5.4.4.4 of Reference 3.I NEDO-33690P Revision 0 2. CGS PRNM RESPONSE TIME ANALYSIS 2.1 Analysis Approach This evaluation addresses the criteria of BTP 7-21, and demonstrates compliance with these requirements. Secondly, this analysis addresses Staff Positions 1.19 and 1.20 of DI&C-ISG-04 for the CGS PRNM.2.2 Evaluation per BTP 7-21 Requirements Real-Time Performance Methodology The instrument response time performance determination for the PRNM is performed by tracking the signal flow from the Local Power Range Monitor (LPRM) input at PRNM, through the LPRM and APRM instruments, ending at the output of the 2-Out-Of-4 Logic Module's Relay Logic Cards output to the Reactor Protection System (RPS), and calculating the design goal processing time delay for each step of the signal transfer process. Figure 2.2-1 provides a block diagram for a single APRM channel.Response Time Start and End Events The LPRM detector input exceeding the respective setpoint is the start event of the NUMAC PRNM response time, and the Relay Logic Card module trip signals to RPS is the end event for response time determination. 2 NEDO-33690P Revision 0 Response Time Performance Table 2.2-2 provides a tabulation of the PRNM instrument processing delay times associated with supporting each of the safety functions identified above. The total channel response (delay)time for the safety functions are calculated by applying the methodology as described above.The longest path of flux data for inclusion in the channel calculations is at the input of the LPRM instrument, processing through the APRM instrument, then output from the 2-Out-Of-4 Logic Module. [[Table 2.2-1 provides a summary of the PRNM response time requirements (per Section 3.3.2 of Reference

1) and the calculated PRNM response times (per Tables 2.2-2 and 2.2-3). This is the total delay time for PRNM to send a trip signal to RPS in response to a flux input that exceeds the setpoint.Table 2.2-1 Safety Function Response Times PRNM Safety Function PRNM Response Time PANM Calculated":.

Requirement (mSec) Response Time (mSec)APRM Neutron Flux -High Trip 40 APRM STP -High Trip 40 APRM Neutron Flux -High 40 (Setdown) Trip OPRM Instability Detect-and-Suppress Trip 400 Tests were performed during PRNM System Validation and Factory Acceptance Testing, to confirm the Columbia PRNM system configuration meets the response time requirement. Testing was performed on production (non-development) equipment manufactured in accordance with the Columbia design documentation. In summary, the calculated PRNM response times, presented in Talbe 2.2-1, meet the requirement of a response time analysis report as stated in Section D.9.4.2.4 of DI&C-ISG-06. Confirmation of response time was performed during verification and validation (V&V) testing.3 NEDO-33690P Revision 0 RPS Response Tine Requirement As identified in Table 1.3.1.1-1 of Reference 4, the RPS response time requirement for the APRM Fixed Neutron Flux -High scram function is 0.09 seconds. The 0.09 second (90 millisecond) RPS response time for this function is used in the transient analysis for the Columbia Final Safety Analysis Report (FSAR) Chapter 15 Accident Analyses (Table 15.0-2, Item 33 of Reference 5). Chapter 15 does not credit OPRM functions in the transient analyses, so its response time is not discussed in the context of total RPS response time.Per Note 17 on page 22 of Reference 6, the trip logic and scram contactor response time is 50 milliseconds. The sensor contact for the NUMAC PRNM system is the RPS Relay Logic Card contacts in the 2-Out-Of-4 Logic Module. The 40 millisecond response time of NUMAC PRNM plus the 50 millisecond RPS response time support the 90 millisecond total RPS response time requirement assumed in the Reference 5 safety analyses.In summary, the response time for the PRNM has been shown by analysis and testing to be less than the required response times, and thus, the PRNM performs sufficiently to meet safety analysis requirements. There is no change in the total RPS response time requirement; therefore, there is no change in setpoint analyses for the CGS NUMAC PRNM due to response time. The NUMAC PRNM response time is adequate to meet the Limiting Response Time of RPS consistent with the guidance provided in Nuclear Regulatory Commission Regulation (NUREG)-0800 and BTP 7-21.Institute of Electrical and Electronics Engineers (IEEE) 603-1991 Clause 4.10 (Reference 7)requires the safety system design basis to include the critical points in time or the plant conditions, after the onset of a design basis event, including:

1) The point in time or plant conditions for which the protective actions of the safety system shall be initiated.

2) The point in time or plant conditions that define the proper completion of the safety function.3) The points in time or the plant conditions that require automatic control of protective actions.4) The point in time or the plant conditions that allow returning a safety system to normal.With respect to system response time, the CGS FSAR (Reference

5) Chapter 15 requirements use the timing requirements from the Licensee Controlled Specification Manual (Reference 4)Table 1.3.1.1-1.

This meets the IEEE-603-1991 Clause 4.10 (Reference

7) requirement for the safety system design basis.4 NEDO-33690P Revision 0 Figure 2.2-1 APRM Channel Response Time Performance Block Diagram 11 NEDO-33690P Revision 0 Table 2.2-2 APRM Channel Flux Trip Delay Time Analysis APRM Neutron Flux -High Trip, APRM Neutron Flux. High -High (Setdown)

Trip& APRM STP -High Trip Processing Step Time (msec)i 6 NEDO-33690P Revision 0 Table 2.2-3 OPRM Trip Delay Time Analysis OPRM Tripi Processing Step Time (msec).. ..I.. '" ll.:: iii : : i .. i .. .... : 1-+7 NEDO-33690P Revision 0 2.3 Evaluation per Staff Positions 1.19 and 1.20 of DI&C-ISG-04 [[Staff Position 1.19: If data rates exceed the capacity of a communications link or the ability of nodes to handle traffic, the system will suffer congestion. All links and nodes should have sufficient capacity to support all functions. The applicant should identify the true data rate, including overhead, to ensure that communication bandwidth is sufficient to ensure proper performance of all safety finctions. Communications throughput thresholds and safety system sensitivity to communications throughput issues should be confirmed by testing.The following explains how the PRNM system satisfies the criteria of Staff Position 1.19. This report discusses the features used to ensure sufficient link capacity, identifies link capacity, and details testing performed to ensure communication links have sufficient capacity to ensure proper performance of safety functions. [I 8 NEDO-33690P Revision 0 Table 2.3-1 FDDI Link Capacity Usage[ [. .....Er]9 NEDO-33690P Revision 0 The PRNM system communication link features, capacity usage information, and testing described above satisfy the criteria of Staff Position 1.19.Staff Position 1.20: The safety system response time calculations should assume a data error rate that is greater than or equal to the design basis error rate and is supported by the error rate observed in design and qualification testing.The following explains how the PRNM system satisfies the criteria of Staff Position 1.20. [[10 NEDO-33690P Revision 0 Table 2.3-2 Data Error Rate Effect to PRNM System Response Time[1 11 NEDO-33690P Revision 0 Table 2.3-3 Maximum Expected PRNM System Response Time The data error rate for each safety-related communication link was established and used to determine the effect of data errors on safety system response time. The established data error rates will be supported by testing a similar PRNM system for GGNS during integration testing.Therefore, the criteria of Staff Position 1.20 are met.12 NEDO-33690P Revision 0 3.

SUMMARY

& CONCLUSIONS This report evaluates the response time of the CGS PRNM versus the safety analysis requirements and standard criteria for digital instrumentation and controls.

This evaluation demonstrates compliance with the criteria of BTP 7-21 and Staff Positions 1.19 and 1.20 of DI&C-ISG-04. Plant specific requirements for time response issues are directly addressed in table items 8.3.4.4.4, 8.4.4.4.4 and 8.5.4.4.4 of Reference 3.13 NEDO-33690P Revision 0 4. REFERENCES

1. GE Nuclear Energy, "Nuclear Measurement Analysis and Control Power Range Neutron Monitor (NUMAC PRNM) Retrofit Plus Option III Stability Trip Function, Licensing Topical Report," NEDC-32410P-A,Volumes 1 & 2, October, 1995 (Including Supplement 1, November, 1997).2. Correspondence from Michael Krupa (Entergy Operations, Inc.), to U.S. Nuclear Regulatory Commission, GNRO-2011/0003 9, dated May 26, 2011 (ML 111460590).

3. GE Hitachi Nuclear Energy, "Columbia Generating Station Plant-Specific Responses Required by NUMAC PRNM Retrofit Plus Option III Stability Trip Function Topical Report (NEDC-32410P-A)," 0000-0101-7647-R3, dated October, 2011.4. Columbia Licensee Controlled Specifications Manual 5. Columbia Final Safety Analysis Report 6. GE Nuclear Energy, Reactor Protection System Design Specification Data Sheet, 23A1877AA, Revision 3.7. IEEE 603-1991, IEEE Standard Criteria for Safety Systems for Nuclear Power Generating Stations.14}}