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{{Adams
#REDIRECT [[LR-N15-0178, Hope Creek - License Amendment Request, Digital Power Range Neutron Monitoring (Prnm) System Upgrade]]
| number = ML15265A224
| issue date = 09/21/2015
| title = Hope Creek - License Amendment Request, Digital Power Range Neutron Monitoring (Prnm) System Upgrade
| author name = Davison P
| author affiliation = PSEG Nuclear, LLC
| addressee name =
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000354
| license number = NPF-057
| contact person =
| case reference number = LARH15-01, LR-N15-0178
| package number = ML15265A223
| document type = Letter, License-Application for Facility Operating License (Amend/Renewal) DKT 50, Technical Specification, Bases Change
| page count = 111
}}
 
=Text=
{{#Wiki_filter:Attachment 1  LAR H15-01 LR-N15-0178 1 of 47 License Amendment Request (LAR) H15-01 -- Digital Power Range Neutron Monitoring (PRNM) System Upgrade
 
Table of Contents
 
1.0DESCRIPTION ...................................................................................................................
22.0PROPOSED CHANGE - Technical Specifications (Section D.11 of DI&C-ISG-06) ........... 43.0BACKGROUND ................................................................................................................ 244.0TECHNICAL ANALYSIS ................................................................................................... 264.1System Description (Section D.1 of DI&C-ISG-06)....................................................... 274.1.1Summary Description ............................................................................................. 274.1.2Detailed System Description .................................................................................. 304.1.2.1PRNM LTR Plant Specific Responses ........................................................... 314.1.2.2OPRM Transition to DSS-CD ......................................................................... 314.1.2.3Transition to Full ARTS .................................................................................. 324.1.2.4Cyber Security Considerations ....................................................................... 334.1.2.5Human Factors Evaluation ............................................................................. 334.1.2.6TSTF-493 ....................................................................................................... 344.1.3System Response Time ......................................................................................... 354.2System (Hardware and Software) Development for the HCGS PRNM System (Section D.2 and D.4 of DI&C-ISG-06) ....................................................................................... 354.2.1Design Analysis Report: Methodology Modifications ............................................. 354.2.2NUMAC System Engineering Development Plan ................................................... 364.2.3NUMAC System Quality Assurance Plan ............................................................... 364.2.4NUMAC System Independent Verification & Validation Plan ................................. 364.2.5Hope Creek Generating Station NUMAC PRNM System Management Plan ........ 364.3Software Architecture / Design Outputs (Section D.3 of DI&C-ISG-06) ....................... 364.3.1System Requirements Specification & APRM Performance Specification ............. 374.3.2APRM Functional Controller System Design Specification .................................... 374.4Environmental Equipment Qualification (Section D.5 of DI&C-ISG-06)
........................ 374.5Defense-In-Depth & Diversity (Section D.6 of DI&C-ISG-06) ....................................... 384.6Communications (Section D.7 of DI&C-ISG-06) ........................................................... 384.7System, Hardware, Software, and Methodology Modifications (Deviations from the Prior LTRs) (Section D.8 of DI&CISG-06) .................................................................... 384.8Compliance with IEEE Standard 603 (Section D.9 of DI&C-ISG-06) ........................... 384.8.1Report on Compliance with IEEE Standards (603-1991 and 7-4.3.2-2003) and Theory of Operations Description ........................................................................... 394.8.2Design Report on Computer Integrity, Test and Calibration, and Fault Detection (IEEE Standard 603-1991 Clause 5.5) ................................................................... 394.8.3Design Analysis Report: Electrical Independence (IEEE Standard 603-1991 Clause 5.6) .........................................................................................................................
394.8.4Setpoint Methodology and Calculations (IEEE Standard 603-1991 Clause 6.8) ... 394.9Conformance with IEEE Standard 7-4.3.2 (Section D.10 of DI&C-ISG-06).................. 404.10Secure Development and Operational Environment (Section D.12 of DI&C-ISG-06) .. 404.11Confirmation of Plant-Specific Actions ......................................................................... 405.0REGULATORY ANALYSIS .............................................................................................. 425.1Applicable Regulatory Requirements/Criteria ............................................................... 425.2No Significant Hazards Consideration .......................................................................... 445.3Conclusions ..................................................................................................................
4
 
==76.0ENVIRONMENTAL CONSIDERATION==
............................................................................ 4
 
==77.0REFERENCES==
................................................................................................................. 47    LAR H15-01 LR-N15-0178 2 of 47  1.0 DESCRIPTION The proposed license amendment request (LAR) would reflect the installation of the General Electric-Hitachi (GEH) digital Nuclear Measurement Analysis and Control (NUMAC) Power Range Neutron Monitoring (PRNM) system. The following Technical Specifications (TS) sections are affected by this change:
TS 2.2  Limiting Safety System Settings  TS 3/4.1.4.3  Rod Block Monitor  TS 3/4.3.1  Reactor Protection System Instrumentation  TS 3/4.3.6  Control Rod Block Instrumentation  TS 3/4.3.11  Oscillation Power Range Monitor  TS 3/4.4.1  Recirculation System  TS 6.9.1.9  Core Operating Limits Report  TS 6.9.3  Special Reports The planned upgrade will replace the existing analog Average Power Range Monitor (APRM)
 
subsystem of the Neutron Monitoring System with the more reliable, digital NUMAC PRNM System during the Spring 2018 refueling outage. This modification will simplify management and maintenance of the system and also includes the following elements:
: 1. The PRNM System design includes an Oscillation Power Range Monitor (OPRM) capability; to detect and suppress reactor instability. The OPRM function continues to
 
satisfy the same regulatory requirements as the currently installed OPRM equipment. 
 
The existing ABB OPRM with BWROG Option III stability solution will change to the GEH OPRM with the Detect and Suppress Solution - Confirmation Density (DSS-CD) stability solution. 
: 2. Full Average power range monitor, Rod block monitor, Technical Specification improvement program (ARTS) implementat ion. Currently, HCGS has implemented 'partial' ARTS. The PRNM system will allow the change to power biased (vs flow biased) Rod Block Monitor (RBM) setpoints. This change allows for Rod Withdrawal Error (RWE) analyses performed for each future reload to take credit for rod blocks during the rod withdrawal transients. 
: 3. Technical Specifications Task Force (TSTF) 493, Revision 4, "Clarify Application of Setpoint Methodology for LSSS Functions". The changes to the TSs include the adoption of the TSTF-493 Option A surveillance notes for the affected PRNM functions.
 
The NRC has issued Interim Staff Guidance (ISG) in digital instrumentation and control (I&C)
DI&C-ISG-06 that describes the licensing process that may be used in the review of LARs associated with digital I&C system modifications. The LAR format and contents of Section 4.0 (Technical Evaluation) of this Attachment 1 are consistent with the guidance provided in Enclosure E and Section C.3 of DI&C-ISG-06. As needed, additional sections have been added to address other aspects of this submittal.
 
LAR H15-01 LR-N15-0178 3 of 47 A similar PRNM system was approved for installation at Columbia Generating Station (CGS) 1, and serves as a precedent for the HCGS installation. 
 
The proposed changes are supported by the following:
 
NRC-approved GEH Licensing Topical Report (LTR) NEDC-32410P-A, Nuclear Measurement Analysis and Control Power Range Neutron Monitor (NUMAC PRNM) Retrofit Plus Option III Stability Trip Function, Volumes 1 and 2, including Supplement 1 (References 1a, 1b, 1c), referred to collectively as the NUMAC PRNM LTR. The NUMAC PRNM LTR provides the primary technical basis for the proposed changes.
NRC-approved GEH Licensing Topical Report (LTR) NEDC-33075P-A; Revision 8, GE Hitachi Boiling Water Reactor Detect and Suppress Solution - Confirmation Density (Reference 2).
Hope Creek Generating Station NUMAC PRNM Upgrade, Enclosures 2 (NEDO-33684, Non-Proprietary) and 3 (NEDC-33684P, Proprietary) of this submittal. These enclosures provide documentation that includes:
o ISG-06 Enclosure B required documentation to support the HCGS PRNM installation.
o HCGS plant-specific responses required (utility action required) by the NUMAC PRNM LTR. Note, that since HCGS is also implementing the OPRM DSS-CD solution, the plant-specific responses also reference DSS-CD.
o Deviations from the NUMAC PRNM LTR and CGS Approval (per ISG-06 Section D.8.2 and ISG-06 Enclosure B Item 1.16).
o Evaluation supporting transition from partial ARTS to full ARTS.
o DSS-CD HCGS Evaluation.
The complete list of the Enclosure documents (Appendices A through T) is provided in Enclosure 2 and 3; a Roadmap is provided cross-referencing the documents to ISG-06 Enclosure B (Enclosure 1). The Enclosure 2 and 3 appendices are correspondingly referenced in Section 4.0 (Technical Analysis) of this Attachment 1.
1  Columbia Generating Station - Issuance of Amendment RE: Implementation of Power Range Neutron Monitoring/Average Power Range Monitor/Rod Block Monitor/Technical Specifications/Maximum Extended Load Line Limit Analysis (PRNM/ARTS/MELLLA) (TAC NO. ME7905) (ADAMS ML13317B623, Non-Proprietary).  (Reference 3) Note that HCGS has previously implemented the ARTS/MELLLA portion.      LAR H15-01 LR-N15-0178 4 of 47 2.0 PROPOSED CHANGE - Technical Specifications (Section D.11 of DI&C-ISG-06)
The proposed TS changes are described below and are indicated on the marked up TS pages provided in Attachment 2 of this submittal. As discussed in ISG-06, setpoint calculations are to be provided with the Phase 2 submittal; however the TS mark-up in this Phase 1 submittal includes the setpoint changes. The surveillance frequency changes discussed/justified in the table below will be applied to the licensee controlled Surveillance Frequency Control Program (SFCP)2, HCGS TS 6.8.4.j.
 
Proposed changes to the TS Bases are provided in Attachment 3 of this submittal for information only; changes to the affected TS Bases pages will be incorporated in accordance with TS 6.15, "Technical Specifications (TS) Bases Control Program." 
 
TS Changes No. Change  Justification  1 Page x, Index Deleted 3/4.3.11, Oscillation Power Range Monitor. 
 
The OPRM function is incorporated into the PRNM system per the NUMAC PRNM LTR (Reference 1); a separate TS section for OPRM
 
is not required. 1b Page xvii, Index
 
Changed page number for TS Bases 3/4.3.7, Monitoring Instrumentation, to B 3/4 3-5.
Administrative change due to addition of Bases text changes for TS 3/4.3.6, Control Rod Block Instrumentation. 2 Page xviii, Index
 
Deleted, 3/4.3.11 Oscillation Power Range Monitor.
The OPRM function is incorporated into the PRNM system per Reference 1; a separate TS section for OPRM is not required. 3 Page 2-4, Table 2.2.1-1, Reactor Protection System Instrumentation Setpoints, Function 2.a. The Neutron Flux-Upscale, Setdown function (Function 2.a) function name is changed to "Neutron Flux-Upscale (Setdown)," consistent with the PRNM LTR. 
 
Updated Trip Setpoint of:  17% of Rated Thermal Power (RTP).
 
Updated Allowable Value of:  19% RTP (no change from current value).
 
Function3 name is updated consistent with Reference 1b Pages H-40 and H-41. 
 
Nominal Trip Setpoints (NTSPs) have been updated consistent with Enclosure 3, Appendix P. (Appendix P provides the Setpoint Methodology and Setpoint Calculations results).
Allowable values (AVs) have been updated consistent with Enclosure 3, Appendix P.
2  TS surveillance frequencies were relocated per TSTF-425, HCGS Amendment 187, February 25, 2011 (ADAMS ML103410243).
3  The term 'function' in this table is used to refer to the 'FUNCTIONAL UNIT' for RPS and the 'TRIP FUNCTION' for Control Rod Block; consistent with the TS tables terminology.      LAR H15-01 LR-N15-0178 5 of 47 No. Change  Justification  4 Page 2-4, Table 2.2.1-1, Reactor Protection System Instrumentation Setpoints, Function 2.b. The Flow Biased Simulated Thermal Power-Upscale (Function 2.b) function name is changed to "Simulated Thermal Power-Upscale," consistent with the PRNM LTR.
Updated Trip Setpoints of: 
: 1. Flow Biased  0.57(w-w) + 59.0% 2. High Flow Clamp  113.5% of RTP (no change from current value)
Updated Allowable Values of: 
: 1. Flow Biased  0.57(w-w) + 61.0% (no change from current value) 2. High Flow Clamp  115.5% of RTP  (no change from current value)
 
Updated note (**) with new w = 10.6% for single recirculation loop operation.
 
Added new note (a). See Item 8 below.
 
Function name is updated consistent with Sections 3.2.5 and 8.3.1.2 of Reference 1a and Pages H-40 and 41 of Reference 1b.
 
NTSPs have been updated consistent with  , Appendix P.
 
AVs have been updated consistent with  , Appendix P. 
 
NTSPs and AVs have been updated consistent with Enclosure 3, Appendix P. 
 
Consistent with Reference 2, note (a) reflects a possible change in the Average Power Range Monitor (APRM) set point due to implementation of the Automated Backup Stability Protection (ABSP) Scram Region. 5 Page 2-4, Table 2.2.1-1, Reactor Protection System Instrumentation Setpoints, Function 2.c. The Fixed Neutron Flux-Upscale (Function 2.c) function name is changed to "Neutron Flux-Upscale," consistent with the PRNM LTR. Updated Trip Setpoint of:  116.3% of RTP.
 
Updated Allowable Value of:  118.3% of RTP.
Function name is updated consistent with Sections 3.2.5 and 8.3.1.2 of Reference 1a and Pages H-40 and 41 of Reference 1b.
NTSPs have been updated consistent with Enclosure 3, Appendix P.
 
AVs have been updated consistent with Enclosure 3, Appendix P. 5a Page 2-4, Table 2.2.1-1, Reactor Protection System Instrumentation Setpoints, Function 2.d.    "lnoperative" trip is retained but is revised to reflect the new NUMAC PRNM system equipment and delete the minimum number of LPRM detector count from this trip. 
 
Change is consistent with Sections 3.2.10 and 8.3.1.2 of Reference 1a.      LAR H15-01 LR-N15-0178 6 of 47 No. Change  Justification  6 Page 2-4, Table 2.2.1-1, Reactor Protection System Instrumentation Setpoints, Function 2.e. Added Function 2.e, 2-Out-Of-4 Voter. 
 
Function for 2-Out-Of-4 Voter added consistent with Sections 8.3.1.2 and 8.3.1.4 of Reference 1a.
There is no Trip Setpoint or Allowable Value associated with this function. This function has been added because all 4 voter channels are required to be operable for this new addition to the logic. Each of the four APRM channels provides signals to the 2-out-of-4 voters for APRM and OPRM trips. 7 Page 2-4, Table 2.2.1-1, Reactor Protection System Instrumentation Setpoints, Function 2.f  Added Function 2.f, OPRM Upscale. 
 
Function for OPRM Upscale added consistent with Reference 1c (Section 8.4.1.4).
This function is relocated from current Limiting Condition of Operation (LCO) 3.3.11 which is being deleted. The Trip Setpoint will be provided in the COLR; there is no Allowable Value associated with this function. 8 Page 2-4, Table 2.2.1-1, Reactor Protection System Instrumentation Setpoints, Notations Changes Added new note (a) stating: "When the Automated BSP Scram Region Setpoints are implemented in accordance with Action 10 of Table 3.3.1-1, the Simulated Thermal Power-Upscale Flow Biased Setpoint will be adjusted per the CORE OPERATING LIMITS REPORT."
 
Consistent with Reference 2. 8a Page 3/4 1-18, TS 3.1.4.3, Rod Block Monitor  Modified Applicability to add:
 
"-and less than 90% of RATED THERMAL POWER with MCPR less than the value specified in the CORE OPERATING LIMITS REPORT, or THERMAL POWER greater than or equal to 90% of RATED THERMAL POWER with MCPR less than the value specified in the CORE OPERATING LIMITS REPORT"    Consistent with implementation of Full ARTS as described in Enclosure 3 of this submittal; Appendix S, Supplemental Information for the Average Power Range Monitor, Rod Block Monitor and Technical Specification Improvement (ARTS) Program for Hope Creek Nuclear Generating Station, Section 3.5.      LAR H15-01 LR-N15-0178 7 of 47 No. Change  Justification  9 Page 3/4 3-1, TS 3/4 3.1 Reactor Protection System Instrumentation, Surveillance Requirement (SR) 4.3.1.2, Logic System Functional Test Add: "Functional Unit 2.a, 2.b, 2.c, 2.d, and 2.f do not require separate LOGIC SYSTEM FUNCTIONAL TESTS. The LOGIC SYSTEM FUNCTIONAL TEST for APRM Function 2.e includes simulating APRM and OPRM trip conditions at the APRM channel inputs to the voter channel to check all combinations of two tripped inputs to the 2-Out-Of-4 voter logic in the voter channels." 
 
Logic System Functional Test added for the 2-Out-Of-4 voter only consistent with Section 8.3.5.2 of Reference 1a and page H-31 of Reference 1c. 
 
The only portion of the PRNM system that is not directly confirmed by other tests is the voting logic through and including the voter output relays. Therefore, the logic system functional test for APRM Functions 2.a, 2.b, 2.c, and 2.d will be deleted. Similarly, the proposed APRM Function 2.f, "OPRM Upsc ale," does not require an LSFT SR.
The Logic System Functional Test will remain at a frequency of 18 months. 9a Page 3/4 3-1, TS 3/4 3.1 Reactor Protection System Instrumentation
 
Correct typographical error in TS Title; delete extra "4".
Format change only, correcting typographical error inadvertently introduced via Amendment 187. 10 Page 3/4 3-1, TS 3/4 3.1 Reactor Protection System Instrumentation, Surveillance Requirement (SR) 4.3.1.3, Reactor Protection System Response Time Added requirement for licensee controlled SFCP: "RESPONSE TIME Testing for Function 2.e has a frequency of 18 months and for Function 2.e, "n" equals 8 channels for the purpose of determining the staggered test frequency. Testing of APRM and OPRM outputs shall alternate."
 
Addition consistent with Sections 8.3.4.4 and 8.4.4.4 of References 1a and 1c. 
 
The LPRM detectors, APRM channels, OPRM channels, and 2-Out-of-4 Voter channels digital electronics are exempt from response time testing. The requirement for response time testing of the RPS logic and RPS contactors will be retained by including a response time testing requirement for the new APRM Function 2.e, "2-Out-of-4 Voter."
The Response Time Testing will remain at a frequency of 18 months. Specific details about the staggered test basis will be reflected in the Licensee controlled SFCP.      LAR H15-01 LR-N15-0178 8 of 47 No. Change  Justification  11 Page 3/4 3-1, TS 3/4 3.1 Reactor Protection System Instrumentation, Limiting Condition for Operation, Action a Add the following new *** notation to Action a: 
"For Functional Unit 2.a, 2.b, 2.c, 2.d, and 2.f, inoperable channels shall be placed in the tripped condition to comply with Action a.
Placing a trip system in trip is not applicable since these Functions provide trip inputs to both trip systems." 
 
This change in text is consistent with Reference 1b and 1c. Consistent with Section 8.3.2.2 Reference 1a, each APRM channel provides input, or is shared by each RPS trip system. 12 Page 3/4 3-1, TS 3/4 3.1 Reactor Protection System Instrumentation, Limiting Condition for Operation, Action b. 
 
Add to the ** notation: "N ote, Action b. is not applicable for Functional Unit 2.a, 2.b, 2.c, 2.d, and 2.f." 
 
Change in text is consistent with Reference 1b and 1c. Consistent with Section 8.3.2.2 Reference 1a, each APRM channel provides input, or is shared by each RPS trip system. 13 Page 3/4 3-2, Table 3.3.1-1, Reactor Protection System Instrumentation, Function 2.a  The Neutron Flux-Upscale, Setdown function (Function 2.a) is retained; however, the name is changed to "Neutron Flux-Upscale (Setdown)." 
 
Deleted references to OPCONs 3 and 4 and associated actions. 
 
Note (e), applicable to all APRM functions, is modified revising the required number of LPRM inputs.
The Minimum Operable Channels Per Trip System is changed to three for Function 2.a. 
 
A note (l) is added to the minimum number of operable channels for Function 2.a. 
 
Function name updated consistent with Reference 1b Pages H-40 and H-44.
 
Deletions of OPCONs 3 and 4 are consistent with Sections 8.3.3.2 -8.3.3.4 of Reference 1a. 
 
See Item 22.
 
Minimum number of operable channels is consistent with Section 8.3.2.2 of Reference 1a. 
 
See Item 22.      LAR H15-01 LR-N15-0178 9 of 47 No. Change  Justification  14 Page 3/4 3-2, Table 3.3.1-1, Reactor Protection System Instrumentation, Function 2.b  The Flow Biased Simulated Thermal Power - Upscale function (Function 2.b) is retained; however, the name is changed to "Simulated Thermal Power- Upscale," consistent with the PRNM LTR. 
 
No change to OPCONs or Actions required. 
 
The Minimum number of Operable Channels Per Trip System is changed to three for Function 2.b.
A note (l) is added to the minimum number of operable channels stating that each APRM/OPRM channel provides inputs to both trip systems.
 
Function name updated consistent with Reference 1b Pages H-40 and H-44.
 
OPCONs and Actions in HCGS TS are consistent with Sections 8.3.3.2 -8.3.3.4 of Reference 1. 
 
Minimum number of operable channels is consistent with Section 8.3.2.2 of Reference 1a.
 
See Item 22. 15 Page 3/4 3-2, Table 3.3.1-1, Reactor Protection System Instrumentation, Function 2.c  The Fixed Neutron Flux - Upscale function (Function 2.c) is retained; however, the name has changed to "Neutron Flux-Upscale." 
 
No change to OPCONs or actions required. 
 
The Minimum number of Operable Channels Per Trip System is changed to three for Function 2.c.
A note (l) is added to the minimum number of operable channels stating that each APRM/OPRM channel provides inputs to both trip systems.
 
Function name updated consistent with Reference 1b Pages H-40 and H-44. 
 
OPCONs and Actions in HCGS TS are consistent with Sections 8.3.3.2 -8.3.3.4 of Reference 1. 
 
Minimum number of operable channels is consistent with Section 8.3.2.2 of Reference 1a. 
 
See Item 22.      LAR H15-01 LR-N15-0178 10 of 47 No. Change  Justification  16 Page 3/4 3-2, Table 3.3.1-1, Reactor Protection System Instrumentation, Function 2.d  The Inoperative function (Function 2.d) is retained. 
 
Deleted references to OPCONs 3 and 4 and associated actions. 
 
The Minimum number of Operable Channels Per Trip System is changed to three for Function 2.d.
Inoperative function is retained but is revised to reflect the new NUMAC PRNM system equipment and delete the minimum number of LPRM detector count from this trip.
A note (l) is added to the minimum number of operable channels stating that each APRM/OPRM channel provides inputs to both trip systems.. 
 
Deletions of OPCONs 3 and 4 are consistent with Sections 8.3.3.2 -8.3.3.4 of Reference 1a. 
 
Minimum number of operable channels is consistent with Section 8.3.2.2 of Reference 1a. 
 
Change consistent with Sections 3.2.10 and 8.3.1.2 of Reference 1a.
See Item 22. 17 Page 3/4 3-2, Table 3.3.1-1, Reactor Protection System Instrumentation, Function 2.e  New function 2.e, 2-Out-Of-4 Voter is added.
 
The minimum number of channels is two per trip system.
 
Applicable OPCONs are 1 and 2.
Associated Action 1 to new function 2.e. 
 
The 2-Out-Of-4 Voter has been added as described in Reference 1a, Sections 8.3.1.4 and 8.3.2.4.
Minimum number of operable channels per trip system (of two) is consistent with Sections 8.3.2.2 and 8.4.2.2 of References 1a and 1c. 
 
Applicable OPCONs for Function 2.e supported by Section 8.4.3.2 of Reference 1c.
Applicable action added consistent with Reference 1b, page H-44.      LAR H15-01 LR-N15-0178 11 of 47 No. Change  Justification  18 Page 3/4 3-2, Table 3.3.1-1, Reactor Protection System Instrumentation, Function 2.f. New function 2.f, OPRM Upscale, is added with a minimum number of channels of 3 per trip system. 
 
An applicable operating condition of 19% RTP is added to Function 2.f. 
 
A note (l) is added to the minimum number of operable channels for Function 2.f.
 
A note (m) is added to the applicable operational condition. See Item 22 below.
 
Added Actions 10, 11, 12.
 
Minimum number of operable channels is consistent with Section 8.4.2.2 of Reference 1c. 
 
As noted in Section 3.5 of Reference 2, the DSS-CD system is required to be operable above a power level set at 5% of rated power below the lower boundary of the Armed Region defined by the MCPR monitoring threshold power level.
 
See Item 22.
 
Per Reference 2, note (m) addresses the limited operability requirements during the initial testing phase following DSS-CD implementation.
 
See Items 19, 20, 21. 19 Page 3/4 3-4,Table 3.3.1-1, Reactor Protection System Instrumentation, Action 
 
Added new Action 10. 
 
Action 10 is consistent with Action I of Reference 2. Actions required when OPRM upscale trip capability cannot be maintained. 20 Page 3/4 3-4, Table 3.3.1-1, Reactor Protection System Instrumentation, Action 
 
Added new Action 11. 
 
Action 11 is consistent with Action J of Reference 2. Action requires implementation of the Manual BSP Regions defined in the CORE OPERATING LIMITS REPORT if an automatic trip function for instability events is not maintained per ACTION 10. The BSP Boundary associated actions are not applicable (applies to MELLLA+ plants), per Section 7.3 of Reference
: 2. 21 Page 3/4 3-4, Table 3.3.1-1, Reactor Protection System Instrumentation, Action
 
Added new Action 12. 
 
Action 12 is consistent with Action K of Reference 2.      LAR H15-01 LR-N15-0178 12 of 47 No. Change  Justification  22 Page 3/4 3-5, Table 3.3.1-1, Reactor Protection System Instrumentation, Notation Changes Updated note (e) to state: "An APRM channel is inoperable if there are less than 3 LPRM inputs per level or less than 20 LPRM inputs to an APRM channel." 
 
Added note (l) stating "Each APRM/OPRM channel provides inputs to both trip systems." 
 
Added Note (m): "Following DSS-CD implementation, DSS-CD is not required to be armed while in the OPRM Armed Region during the first reactor startup and during the first controlled shutdown that passes completely through the OPRM Armed Region. However, DSS-CD is considered OPERABLE and shall be maintained OPERABLE and capable of automatically arming for operation at recirculation drive flow rates above the OPRM Armed Region."
 
Note (e) is updated consistent with the new minimum number of LPRMs, and limits on the maximum number that can be bypassed or failed (consistent with Section 8.3.2.2 of Reference 1.a).
Note added consistent with Section 8.3.2.4 of Reference 1a, noting the 4-APRM channel replacement configuration is shared by both trip systems for each APRM function  Note (l) is applicable to APRM Functions 2.a, 2.b, 2.c, 2.d and 2.f.
 
As noted in Reference 2, note (m) addresses the limited operability requirements during the initial testing phase following DSS-CD implementation.      LAR H15-01 LR-N15-0178 13 of 47 No. Change  Justification  23 Page 3/4 3-7, Table 4.3.1.1-1, Reactor Protection System Instrumentation Surveillance Requirements, Function 2.a The Neutron Flux-Upscale, Setdown function (Function 2.a) is retained; however, the name is changed to "Neutron Flux-Upscale (Setdown)."
Deleted references to OPCONs 3 and 4. 
 
Channel Check, revised frequency and retained note (b). 
 
Channel Function Test, revised frequency and retained note (l). 
 
Channel Calibration is retained with revised frequency.
 
Added TSTF-493 Option A notes to Notes page, Channel Calibration notated with Notes (n) and (o).
 
Function name updated consistent with Reference 1b Pages H-40 and H-47. 
 
Deletions of OPCONs 3-4 are consistent with the guidance in Sections 8.3.4.2.2 of Reference 1a.
Current note (b) consistent with Section 8.3.4.1.2 of Reference 1a. The APRM Channel check frequency is updated from once per 12 hours to 24 hours consistent with Section 8.3.4.1.2 of Reference 1a.
Current note (l) consistent with Section 8.3.4.2.2 of Reference 1a. The APRM Channel Functional Test frequency is updated from 31 days (monthly) to every 184 days (semi-annual). 
 
Retained channel calibration consistent with Section 8.3.4.3.2 of Reference 1a. The APRM Channel Calibration is updated from 184 days (semi-annual) to every 18 months consistent with Section 8.3.4.3.4 of Reference 1a.
 
Notes (n) and (o) are applicable to APRM Functions 2.a, 2.b, 2.c. These notes are not specified in the NUMAC PRNM LTR. These notes are consistent with TSTF-493, Option A, for the functions affected by this proposed change.      LAR H15-01 LR-N15-0178 14 of 47 No. Change  Justification  24 Page 3/4 3-7, Table 4.3.1.1-1, Reactor Protection System Instrumentation Surveillance Requirements, Function 2.b The Flow Biased Simulated Thermal Power - Upscale function (Function 2.b) is retained; however, the name is changed to "Simulated Thermal Power- Upscale," consistent with the PRNM LTR. 
 
Channel Check, revised frequency and deleted current note (g). 
 
Channel Functional Test, revised frequency and added new note (e). 
 
Channel Calibration, revised frequency, retained note (d), deleted current notes (e) and (h), and added new note (g). 
 
Added TSTF-493 Option A notes to Notes page, Channel Calibration notated with Notes (n) and (o).
 
Function name updated consistent with Reference 1b Pages H-40 and H-47. 
 
Deleted current note (g) consistent with Section 8.3.4.1.2 of Reference 1a. The APRM Channel check frequency is updated from once per 12 hours to 24 hours consistent with Section 8.3.4.1.2 of Reference 1a. 
 
New note (e) is consistent with Section 8.3.4.2.2 of Reference 1a. The APRM Channel Functional Test frequency is updated from quarterly to every 184 days (semi-annual). 
 
Retained note (d) and deleted current notes (e) and (h) consistent with Section 8.3.4.3.2.2 of Reference 1a. Added new note (g) consistent Section 8.3.4.3.2.2 of Reference 1a.
The APRM Channel Calibration is updated from 184 days (semi-annual) to every 18 months consistent with Section 8.3.4.3.4. Deleted the separate requirement for weekly adjustment of flow hardware (calibration of the flow hardware is included in overall Channel Calibration at 18-month intervals.). Retained the requirement to adjust the APRM gain to match APRM power to thermal power at a frequency of every 7 days (weekly).
 
Consistent with TSTF-493 Option A LAR H15-01 LR-N15-0178 15 of 47 No. Change  Justification  25 Page 3/4 3-7, Table 4.3.1.1-1, Reactor Protection System Instrumentation Surveillance Requirements, Function 2.c The Fixed Neutron Flux - Upscale function (Function 2.c) is retained; however, the name has changed to "Neutron Flux-Upscale." No changes to OPCONs or actions are required. 
 
Channel Check, revised frequency. 
 
Channel Functional Test, revised frequency. 
 
Channel Calibration, revised frequency and retained note (d).
 
Added TSTF-493 Option A notes to Notes page, Channel Calibration notated with Notes (n) and (o).
 
Function name updated consistent with Reference 1b Pages H-40 and H-47. 
 
The APRM Channel check frequency is updated from once per 12 hours to 24 hours consistent with Section 8.3.4.1.2 of Reference 1a. 
 
The APRM Channel Functional Test frequency is updated from quarterly to every 184 days (semi-annual), consistent with Section 8.3.4.2.2 of Reference 1a.
Retained note (d) consistent with Section 8.3.4.3.2 of Reference 1a. The APRM Channel Calibration is updated from 184 days (semi-annual) to every 18 months consistent with Section 8.3.4.3.4. Retained the requirement to adjust the APRM gain to match APRM power to thermal power at a frequency of every 7 days (weekly).
 
Consistent with TSTF-493 Option A. 26 Page 3/4 3-7, Table 4.3.1.1-1, Reactor Protection System Instrumentation Surveillance Requirements, Function 2.d The Inoperative function (Function 2.d) is
 
retained.
Deleted references to OPCONs 3 and 4. 
 
Channel Check remains NA. 
 
Channel Functional Test, revised frequency.
 
Channel Calibration remains NA. 
 
Deletion of OPCONs 3-4 is consistent with Section 8.3.3.4 of Reference 1a. 
 
Consistent with Section 8.3.4.1.4 of Reference 1a.
The APRM Channel Functional Test frequency is updated from quarterly to every 184 days (semi-annual), consistent with Section 8.3.4.2.2 of Reference 1a.   
 
NA is retained consistent with Section 8.3.4.3.2 of Reference 1a.      LAR H15-01 LR-N15-0178 16 of 47 No. Change  Justification  27 Page 3/4 3-7, Table 4.3.1.1-1, Reactor Protection System Instrumentation Surveillance Requirements, Function 2.e New function 2.e, 2-Out-Of-4 Voter, is added with applicable operating OPCONs 1 and 2. 
 
Channel Check applies. 
 
Channel Functional Test applies.
 
Channel Calibration is NA. 
 
The 2-Out-Of-4 Voter has been added as described in Reference 1a, Sections 8.3.1.4 and 8.3.2.4. 
 
The Channel Check frequency is established at once per 24 hours consistent with Section 8.3.4.1.2 of Reference 1a.
Consistent with Section 8.3.4.2.2 of Reference 1a and Section 8.4.4.2.2 of Reference 1c. The requirement for a frequency of every 184 days (6 months) is included, which is the same frequency as used for the APRM and OPRM functions supported by the Voter.
Added Channel Calibration function is NA consistent with Reference 1b, page H-48. 28 Page 3/4 3-7, Table 4.3.1.1-1, Reactor Protection System Instrumentation Surveillance Requirements, Function 2.f 
 
New function 2.f, OPRM Upscale, is added with applicable operating condition of 19% RTP.
 
Channel Check applies. 
 
Channel Functional Test applies, added note (e).     
 
Channel Calibration, added note (g).
 
Per Section 3.5 of Reference 2, the DSS-CD system is required to be operable above a power level set at 5% of rated power below the lower boundary of the Armed Region defined by the MCPR monitoring threshold power level.
 
Consistent with Section 8.4.4.1 of Reference 1c, the Channel Check frequency is added as once per 24 hours consistent with the APRM functions (Section 8.3.4.1.2 of Reference 1a).
Channel Functional Test is added consistent with Section 8.4.4.2 Reference 1c. Added note (e) consistent with Section 8.4.4.2 of Reference 1c. The Channel Functional Test frequency is 184 days (semi-annual) consistent with Section 8.4.4.2.2 of Reference 1.c.
Note that Reference 1c also adds a requirement to "confirm that the OPRM Upscale is enabled when APRM Simulated Thermal Power is >[30]% and recirculation flow is < [60]% rated recirculation flow."  Reference 2 removes this requirement.
 
Channel Calibration is added consistent with Section 8.4.4.3 Reference 1c. The OPRM Channel Calibration is added at a frequency of every 18 months consistent with Section 8.4.4.3.2 of Reference 1c.      LAR H15-01 LR-N15-0178 17 of 47 No. Change  Justification  29 Page 3/4 3-8, Table 4.3.1.1-1, Reactor Protection System Instrumentation Surveillance Requirements, Notation Changes Replaced notes (e) and (g), and deleted note (h). 
 
Added new note (e): "The CHANNEL FUNCTIONAL TEST includes the recirculation flow input function, excluding the flow transmitters."
Added new note (g): "Calibration includes the flow input function."
Note (l) is retained for Function 2.a. 
 
Added TSTF-493 Option A notes (n) and (o).
 
Replaced note (e) and deleted note (h) based on Section 8.3.4.3.2 of Reference 1a.
Deleted current note (g) consistent with Section 8.3.4.1.3 of Reference 1a. 
 
New note (e) is consistent with Section 8.3.4.2.2 of Reference 1a and Section 8.4.4.2 of Reference 1c.
 
New note (g) is consistent Section 8.3.4.2.2 of Reference 1a.
Current note (l) still applies and is consistent with Section 8.3.4.2.2 of Reference 1a. 
 
Added TSTF notes per TSTF-493. Notes (n) and (o) address as found and as left tolerance requirements. 30 Page 3/4 3-57, Table 3.3.6-1, Control Rod Block Instrumentation, Function 1.
 
Minimum operable channels remain at two. 
 
Applicable Operational Condition (OPCON
: 1) remains unchanged; the asterisk (*) note on OPCON 1 is modified to:
 
See TS 3.1.4.3 Applicability 
 
Actions remain unchanged.
Consistent with Section 8.5.2.2 of References 1a and 1c.
Consistent with Section 8.5.3.3 of Reference 1a, the operational conditions remain as-is. 
 
See Item 8a in table. Consistent with implementation of Full ARTS as described in Enclosure 3 Appendix S, 'Supplemental Information for the Average Power Range Monitor, Rod Block Monitor and Technical Specification Improvement (ARTS) Program for Hope Creek Nuclear Generating Station,' Section 3.5.
 
Consistent with Section 8.5.2.2 of Reference 1a, the actions remains as-is.      LAR H15-01 LR-N15-0178 18 of 47 No. Change  Justification  31 Page 3/4 3-57, Table 3.3.6-1, Control Rod Block Instrumentation, Function 2. 
 
Replaced four minimum operable channels with three minimum operable channels for Functions 2.a-2.d.
Renamed Function 2a to Simulated Thermal Power - Upscale.
 
Renamed / modified Function 2d to Simulated Thermal Power - Upscale (Setdown).
 
Operational condition remains unchanged. 
 
Actions remain unchanged.
The APRM related control rod block functions are eliminated from the Improved Technical Specifications. No safety analysis or safety credit is taken for the APRM initiated rod blocks, they are provided to reduce the risk of exceeding RPS trip setpoints (Reference 1a, Section 8.5.1.3). The APRM Control Rod Block functions are not credited in any HCGS UFSAR Chapter 15 accident analyses. HCGS is choosing to maintain the functions in TS for administrative reasons (versus relocating to a licensee controlled document). Consistent with Reference 1a Section 8.5.1.4 the changes to the functions are described below:
Consistent with Section 8.5.2.2 of Reference 1c. 
 
Function name updated consistent with Section 8.3.1.2 and Page H-40 of Reference 1b.
 
Function name updated consistent with the renaming of the associated trip function, Section 8.3.1.2 of Reference 1a. Function is revised from a flux-based signal to a Simulated Thermal Power (STP) signal; a low-pass filter with a six second time constant is applied to the Flux signal to develop the STP signal.
 
Consistent with Section 8.5.3.3 of Reference 1a, the operational conditions remains as-is. 
 
Consistent with Section 8.5.2.2 of Reference 1a, the actions remains as-is. 32 Page 3/4 3-57, Table 3.3.6-1, Control Rod Block Instrumentation, Function 6. 
 
Delete Functions 6.a, 6.b, and 6.c. 
 
For ARTS plants, deletions consistent with Section 8.5.1.3 of Reference 1a. 32a Page 3/4 3-58, Table 3.3.6-1, Control Rod Block Instrumentation, Notes 
 
Revised Note (*).
 
See Item 30      LAR H15-01 LR-N15-0178 19 of 47 No. Change  Justification  33 Page 3/4 3-59, Table 3.3.6-2, Control Rod Block Instrumentation Setpoints, Function 1.
Replace Function 1.a(i) and 1.a(ii) with new 1.a(i), 1.a(ii), and 1.a(iii).
 
Added Notes (a), (b), (c), and (d).
 
Function 1.c, Downscale values are relocated to the COLR.
 
Added Note **.
 
Changes consistent with implementation of Full ARTS as described in Enclosure 3 of this submittal; Enclosure 3, Appendix S, 'Supplemental Information for the Average Power Range Monitor, Rod Block Monitor and Technical Specification Improvement (ARTS) Program for Hope Creek Nuclear Generating
 
Station. Flow biased upscale and High Flow Clamped upscale trips are replaced with Low Trip Setpoint (LTSP), Intermediate Trip Setpoint (ITSP) and High Trip Setpoint (HTSP) and their respective trip setpoints and allowable values are relocated to the COLR.  (Enclosure 3, Appendix S Section 3.3.1). Notes are added to the LTSP, ITSP and HTSP identifying the Low Power Setpoint (LPSP),
 
Intermediate Power Set point (IPSP) and High Power Setpoint (HPSP) and their respective allowable values/ranges. This formatting is consistent with improved Technical Specifications. The LPSP, IPSP, and HPSP are described in Enclosure 3 Attachment S Table 5 and Section 3.3.1. Power setpoints do not change on a cycle-by-cycle basis and are therefore assumed constant. Therefore, these power setpoint ranges (provided in Appendix P of Enclosure 3) are referenced directly in the Technical Specifications.
 
Changes consistent with implementation of Full ARTS as described in Enclosure 3, Appendix S, Table 5, footnote 2 (the DTSP is not important for RWE analysis and may be moved to the COLR). Note identifies values that are located in the COLR.      LAR H15-01 LR-N15-0178 20 of 47 No. Change  Justification  33a Page 3/4 3-59, Table 3.3.6-2, Control Rod Block Instrumentation Setpoints, Function 2.
Renamed Function 2a to Simulated Thermal Power - Upscale
 
Updated Trip Setpoint of:  0.57(w-w) + 54%* with a maximum of  108% of RATED THERMAL POWER 
 
Updated Allowable Value of: 0.57(w-w) + 56%* with a maximum of  111% of RATED THERMAL POWER Function 2c, Downscale:
Updated Trip Setpoint of:  5% of RATED THERMAL POWER Updated Allowable Value of:  3% RATED THERMAL POWER  (no change from current value)
Renamed Function 2d to Simulated Thermal Power - Upscale (Setdown)
 
Updated Trip Setpoint of:  11% of RATED THERMAL POWER  (no change from current value)
 
Updated Allowable Value of:  13% RATED THERMAL POWER  (no change from current value)
 
See Item 31 in table.
 
NTSPs have been updated consistent with  , Appendix P.
 
AVs have been updated consistent with Enclosure 3, Appendix P.
 
NTSPs have been updated consistent with Enclosure 3, Appendix P.
AVs have been updated consistent with Enclosure 3, Appendix P.
 
See Item 31 in table.
 
NTSPs have been updated consistent with  , Appendix P.
 
AVs have been updated consistent with Enclosure 3, Appendix P.
33b Page 3/4 3-59, Table 3.3.6-2, Control Rod Block Instrumentation Setpoints, Function 6.
 
Delete Functions 6.a, 6.b, and 6.c
 
For Full ARTS plants, deletions consistent with Section 8.5.1.3 of Reference 1a.      LAR H15-01 LR-N15-0178 21 of 47 No. Change  Justification  34 Page 3/4 3-60, Table 4.3.6-1, Control Rod Block Instrumentation Surveillance Requirements, Function 1.
Applicable Operational Condition (OPCON 1) remains unchanged; the asterisk (*) note on OPCON 1 is modified to:
 
See TS 3.1.4.3 Applicability
 
Channel Check remains NA. 
 
Channel Functional Test, revised frequency. 
 
Channel Calibration, revised frequency. 
 
Added TSTF-493 Option A notes to Notes page, Function 1.a Channel Calibration notated with Notes (g) and (h) 
 
See Item 8a in table. Consistent with implementation of Full ARTS as described in Enclosure 3 of this submittal; NEDC 33864, Appendix S, 'Supplemental Information for the Average Power Range Monitor, Rod Block Monitor and Technical Specification Improvement (ARTS) Program for Hope Creek Nuclear Generating Station,' Section 3.5.
 
Consistent with Section 8.5.4.1.2 of Reference 1a.
Consistent with Section 8.5.4.2.2 of Reference 1a. The Channel Functional Test frequency is updated from 92 days (quarterly) to every 184 days (semi-annual). 
 
Consistent with Section 8.5.4.3.2 of Reference 1a. The Channel Calibration frequency is updated from 184 days (semi-annual) to every 18 months. Inoperative remains NA.
Consistent with TSTF-493 Option A. 35 Page 3/4 3-60, Table 4.3.6-1, Control Rod Block Instrumentation Surveillance Requirements, Function 2.
 
Renamed Function 2a to Simulated Thermal Power - Upscale.
Renamed Function 2d to Simulated Thermal Power - Upscale (Setdown).
Channel Check remains NA. 
 
Channel Functional Test, revised frequency. 
 
Channel Calibration, revised frequency.
 
See Item 31 in table.
 
See Item 31 in table.
 
Consistent with Section 8.5.4.1.2 of Reference 1a.
Consistent with Section 8.5.4.2.2 of Reference 1a. The Channel Functional Test frequency is updated from 92 days (quarterly) to every 184 days (semi-annual).
Consistent with Section 8.5.4.3.2 of Reference 1a. The Channel Calibration frequency is updated from 184 days (semi-annual) to every 18 months. The Inoperative Function 2.b remains NA.      LAR H15-01 LR-N15-0178 22 of 47 No. Change  Justification  36 Page 3/4 3-60, Table 4.3.6-1, Control Rod Block Instrumentation Surveillance Requirements, Function 6 Delete Functions 6.a, 6.b, and 6.c. For ARTS plants, deletions consistent with Section 8.5.1.3 of Reference 1a. 36a Page 3/4 3-61, TABLE 4.3.6-1 (Continued),Control Rod Block Instrumentation Surveillance Requirements, NOTES:
Revised Note (*).
 
Added TSTF-493 Option A Notes (g) and (h).
 
See Item 34.
 
Add TSTF notes per TSTF-493. Notes (g) and (h) address as found and as left tolerance requirements. 37 Page 3/4, 3-110, TS 3/4.3.11, Oscillation Power Range Monitor 
 
Deleted current OPRM section.
OPRM requirements addressed by the addition of the OPRM Upscale Function 2.f to RPS Instrumentation (consistent with Reference 1c.). 38 Page 3/4 4-1 and 2, TS 3/4 4.1 Recirculation System, Recirculation Loops Page 3/4 4-1, Action a.2 and Action a.3:
modified requirement to declare the channel inoperable and reference the TS 3.3.1 and 3.3.6 actions respectively. 
 
Page 3/4 4-2, Action a.4: Deleted Action a.4.
Consistent with Section 8.3.2.2 Reference (1a),
also refer to item 11 above; similar change made for TS 3.3.1. The APRM system is divided into four APRM channels and four 2-Out-Of-4 voter channels. Each APRM channel provides inputs to each of the four voter channels. The four voter channels are divided into two groups of two voters, with each group of two voters providing inputs to one RPS trip system. The system is designed to allow one APRM channel, but no voter channels, to be bypassed. The proposed changes maintain the requirement to reduce the APRM scram and control rod block setpoints and allowable values within four hours of entering single loop operation (SLO).
RBM changed to power versus flow reference for full ARTS, the reactor coolant recirculation flow functions have been deleted (refer to items 32, 33b, and 36).      LAR H15-01 LR-N15-0178 23 of 47 No. Change  Justification  39 Page 6-20, 6.9.1.9, Core Operating Limits Report (COLR)
Deleted 3/4.3.11 Osc illation Power Range Monitor (OPRM).
Added 2.2 Reactor Protection System Instrumentation Setpoints, 3/4.1.4.3 Rod Block Monitor, 3/4.3.1 Reactor Protection System Instrumentation and 3/4 3.6 Control Rod Block Instrumentation.
References updated.
Deletion consistent with the deletion of this section from the HCGS TS.
Consistent with justified markups of associated section.
 
Applicable references are incorporated via GESTAR reference. Deletion of Reference 2 related to Crossflow Ultrasonic Flow Measurement is administrative; this report should have been deleted as part of HCGS EPU amendment (ADAMS ML081230640). 40 Page 6-21, 6.9.3, Special Reports 
 
Added requirement for the OPRM report in Section 6.9.3 which is required in new Action 10 of Table 3.3.1-1 (see Item 19 above). Consistent with Reference 2.
LAR H15-01 LR-N15-0178 24 of 47 3.0 BACKGROUND The proposed change would reflect the installation of the General Electric-Hitachi (GEH) digital Nuclear Measurement Analysis and Contro l (NUMAC) Power Range Neutron Monitoring (PRNM) system. The PRNM system enables HCGS to implement full ARTS and also incorporates the OPRM function (with HCGS changing to the DSS-CD stability methodology).
The PRNM system replaces the existing APRM system which is part of the Neutron Monitoring System (NMS). The NMS monitors the neutron flux level in the reactor in three separate, overlapping ranges; all using in-core instrum entation systems (refer to Hope Creek Updated Final Safety Analysis Report (UFSAR) Figure 7.6-1). The system provides automatic core protection signals in the event of power transients. The NMS includes the Source Range Monitor (SRM) system, Intermediate Range Monitor (IRM) system, and the power range monitoring system. The power range monitoring is accomplished by the APRM system, which receives core flux level signals from the Local Power Range Monitors (LPRM). Additional information on the safety related elements of the NMS are provided in Section 7.6 of the Hope Creek UFSAR.
 
HCGS is a GE BWR/4. The existing design incorporates six APRM channels. Each APRM channel uses input signals from a number of local power range monitors (LPRMs). The six APRM channels are combined in two groups of three channels each to form two trip channels.
The PRNM modification will replace the six-channel APRM with a four-channel APRM configuration whereby each channel uses one-fourth of the total LPRM detectors. The APRM functions in each channel are the same; however four 2-Out-of-4 Voter logic channels are added. Each APRM provides inputs to all four of the 2-Out-of-4 Voter logic channels. Outputs from two voter logic channels supply inputs to each of two Reactor Protection System (RPS) trip system divisions.
The changes are based on LTRs for PRNM and DSS-CD:
 
The PRNM LTR was reviewed and approved by the NRC staff in 1995 and 1997 (References 1a, b, c). The overall change is further supported by prior operating experience that has been gained from changes to install similar General Electric- Hitachi (GEH) Nuclear Measurement Analysis and Control (NUMAC)-based equipment in U.S. nuclear power plants. The LTRs and their corresponding safety evaluations (SEs) establish utility-specific licensee actions that each referencing license amendment request (LAR) must perform, as applicable. The LTRs provide a series of block diagrams to show a variety of GEH NUMAC PRNMS equipment configurations that could be applied to different General Electric (GE) Boiling-Water Reactor (BWR) designs
 
using GEH NUMAC hardware and software.
The DSS-CD LTR was reviewed and approved by the NRC staff in 2013 (Reference 2). The LTR defines the licensing basis and reload applications for the "Detect and Suppress Solution - Confirmation Density" (DSS-CD) methodology. DSS-CD is a type of long-term stability solution that has features similar to the previously approved Option III. DSS-CD maintains for defense-in-depth the algorithms that were approved for Option III:
the Period Based Detection Algorithm (PBDA), the Amplitude Based Algorithm (ABA),
and the Growth Rate Algorithm (GRA).
LAR H15-01 LR-N15-0178 25 of 47 The GEH NUMAC PRNM development approach includes reliance upon pre-developed hardware and software components. A high-level description of these previously developed components is contained in the LTR (References 1a, b, c). The set of pre-developed software supports interfaces with NUMAC modules and instrument-specific application functions, which are configured to construct plant-specific instrumentation such as the HCGS PRNM system.
Most of this previously developed software was produced to satisfy the applicable regulatory evaluation criteria that the NRC staff used to evaluate the base LTR in 1995. However, since that time, the applicable regulatory evaluation criteria used by the NRC staff to evaluate software-based safety functions within digital safety-related equipment have changed. The evaluations provided with this submittal reflect the current regulatory criteria. 
 
Also included in this submittal is an evaluation of the HCGS PRNM system against the plant specific action items (utility action required items) defined in the PRNM LTR and Safety Evaluation (SE). The implementation of DSS-CD is also reflected in the plant-specific
 
evaluation.
 
To prepare for the PRNM upgrade and to support the required licensing process, PSEG has performed, or is in process of performing, the following activities:
: 1. Critical Digital Review
 
A Critical Digital Review (CDR) 4, of the GE-Hitachi (GEH) NUMAC PRNM System was performed prior to finalizing the approval of the PRNM Upgrade Project. The purpose and scope of a CDR was to determine if a given digital-based product is both capable and suitable for use in a given nuclear application - based upon a predefined set of critical characteristics such as physical characteristics (e.g., size, connector type), performance characteristics (e.g., timing, functions, failure detection), and dependability (i.e., programmatic) characteristics.
 
The CDR utilized a systematic risk-informed appr oach to evaluate a digital product's ability to perform specific functions, and the ability to respond to abnormal conditions and events when operating within the plant. The goal of the review was not only to evaluate the NUMAC PRNM product, but to gain a sufficient understanding of Hope Creek specific design considerations, areas of potential risk, and activities to ensure they were addressed in subsequent planning.
 
The CDR concluded that:
 
The GEH NUMAC PRNM product is a technically suitable replacement for the existing Safety-Related Hope Creek PRNM. GEH has an established regulatory approved Appendix B quality program  The GEH quality program and related processes are suitable to ensure the quality of the design, configuration control, Part 21 reportability, and maintenance throughout the life of the NUMAC PRNM system  The GEH NUMAC PRNM utilizes a basic set of modules and components, but each system is uniquely designed to meet plant-specific needs and constraints.
 
4  Also referred to as an independent System Integrity Review (SIR)      LAR H15-01 LR-N15-0178 26 of 47 2. CGS Benchmarking and Incorporation of Industry Lessons-learned 
 
The CGS PRNM upgrade/amendment is cited as a precedent for the HCGS PRNM upgrade. Close alignment with the CGS project was established during the initial stages of the HCGS project, and has been maintained, including CGS site visits, sharing of information, benchmarking, and lessons learned. Other industry operating experience (OE) has also been incorporated into the HCGS PRNM project.
: 3. Procedure Development Plan PRNM procedures from several other utilities, including CGS, have been obtained as a reference for developing the HCGS PRNM procedures. Draft procedures will be in place prior to the FAT scheduled for the first quarter of 2016. In addition, HCGS is using Nine Mile Point PRNM Manual (GEH supplied system manual) as a reference since the Nine Mile Point PRNM system is similar to HCGS. 
: 4. GEH Setpoint Methodology Audit As part of the PRNM upgrade, GEH is performing the setpoint calculations for the new NUMAC PRNM system. These calculations are proprietary, and a calculation results report is delivered to PSEG as part of the project in lieu of the full setpoint calculations. 
 
For past PRNM projects the NRC typically performs an audit of the GEH calculations.
PSEG audited the GEH setpoint program prior to preparation of the setpoint calculations. Topics covered included the GEH setpoint methodology and the GEH TSTF-493 methodology. The audit concluded that the GEH Instrument Setpoint Methodology - Overview document delivered fo r the Phase 1 licensing submittal met the required NRC guidance. 
: 5. TS Amendment removing APRM OPCON 5 operability requirement A HCGS license amendment was approved in 2013 that changed Technical Specification (TS) 3/4.3.1, "Reactor Protection System Instrumentation," and TS 3/4.3.6, "Control Rod Block Instrumentation" by modifying the operability requirements for the average power range monitoring (APRM) instrumentation system. The amendment eliminated the requirements that the APRM "Upscale" and "Inoperative" scram and control rod withdrawal block functions be operable in the "Refueling" Operational Condition (OPCON) 5. This change will permit a more efficient installation of the PRNM system upgrade during the refueling outage.
4.0 TECHNICAL ANALYSIS The format and contents of this LAR are consistent, as appropriate, with the guidance provided in Enclosure E and Section C.3 of DI&C-ISG-06. As appropriate, additional supporting discussion, analysis and evaluation is provided specific to the HCGS PRNM installation.
 
DI&C-ISG-06, Enclosure B, lists documents that are typically submitted by the licensee in support of a Tier 2 submittal during Phases 1 and 2 of the NRC staff review. The Phase 1 documents that are associated with this submittal are provided in Enclosure 2 (Non-Proprietary)      LAR H15-01 LR-N15-0178 27 of 47 and Enclosure 3 (Proprietary). A roadmap, or cross-reference, between the ISG-06 Enclosure B document name and the equivalent document supporting this application is provided in  of this submittal. Two variations of the roadmap are provided:
: 1. ISG-06 Enclosure B Item mapped to the HCGS PRNM GEH Document(s)
: 2. HCGS PRNM GEH Document mapped to the ISG-06 Enclosure B Item(s) 4.1 System Description (Section D.1 of DI&C-ISG-06) 4.1.1 Summary Description The NUMAC PRNM upgrade is based on the LTR (References 1a, b, c), which was approved by the NRC. The PRNM hardware enables HCGS to implement full ARTS. The PRNM design includes an automatic instability trip function, OPRM, which will be implemented with the GEH Boiling Water Reactor Detect and Suppress Solution - Confirmation Density (DSS-CD) methodology (Reference 2). HCGS will be transitioning to the DSS-CD solution from the current ABB OPRM with BWROG Option III stability solution. 
 
The existing power range monitor functions are retained, including LPRM detector signal processing, LPRM averaging, APRM trips, and RBM logic and interlocks. The existing analog LPRM signal processing electronics, LPRM averaging and APRM trip electronics, LPRM detector power supply hardware and recirculation flow signal processing electronics are being replaced by integrated digital NUMAC chassis based APRM electronics. 
 
The existing six APRM channels will be replaced with four channels of NUMAC APRM, each channel utilizing one-fourth of the total available LPRM detectors. Four 2-Out-of-4 Voter channels are being added between the APRM channels and the existing RPS logic. Each Voter receives input from all four APRM channels and provides input to one Reactor Protection System (RPS) trip logic. This ensures that each input to RPS is a voted result of all four APRMs. 
 
All interfaces with external systems are maintained electrically equivalent using interface sub-assemblies with exception of the interface to the plant computer and plant operator's panel.
Interface to the plant computer system is accomplished by the NUMAC Interface Computer (NIC) system and the interface to the operator panel is accomplished with Operator Display Assemblies (ODAs), which replace the existing meter displays. 
 
The PRNM design includes an automatic instability trip function, OPRM. The existing ABB OPRM with BWROG Option III stability solution will change to the GEH OPRM with the Detect and Suppress Solution - Confirmation Density (DSS-CD) stability solution. DSS-CD is designed to detect power oscillations upon inception and initiate control rod insertion (scram) to terminate the oscillations prior to any significant amplitude growth. DSS-CD introduces an enhanced
 
detection algorithm that detects the inception of power oscillations and generates an earlier power suppression trip signal based on successive period confirmation recognition and an amplitude component. The existing Option III algorit hms are retained (with generic setpoints) to provide defense-in-depth protection for unanticipated reactor instability events.
The existing flow-biased RBM will be replaced by a power dependent RBM. The power dependent RBM will permit HCGS to implement "Full" ARTS versus the current "Partial" ARTS; allowing cycle specific RWE analyses to credit the blocking of rod withdrawals.      LAR H15-01 LR-N15-0178 28 of 47 Each APRM channel consists of a Master APRM instrument, a Slave APRM instrument, and a 2-Out-of-4 Voter, which are safety related. The safety functions performed by each APRM channel involve the processing of sensor inputs to produce a set of trip votes that must then satisfy 2-Out-of-4 coincidence voting logic to cause the PRNM relay outputs to the RPS trip system to change state.. Both the Master and the Slave instruments receive inputs from the associated LPRM detectors. Flow transmitters in each of the recirculation loops provide the loop flow input to the associated APRM Master instruments in each channel. APRM communication with RBM, which is not safety-related, is conducted through Fiber Direct Data Interface (FDDl). The FDDl Module provides electrical and communication isolation of the signals while permitting the data to be transmitted.
Each Master APRM instrument provides interfaces to the LPRM detectors and recirculation loop (Loops A and B) flow transmitters, processes detector signals, performs algorithms to produce a set of trip votes, interfaces with all four 2-Out-of-4 Voters to provide its trip votes, receives bypass and self-test status information from its channel's 2-Out-of-4 Voter, and exchanges data with its channel's Slave APRM and one RBM through two separate FDDl links. 
 
Each Slave APRM instrument provides interfaces to a set of LPRM detectors, processes the signal, and exchanges the data with its Master APRM and one channel of the RBM through two separate FDDI links. 
 
The FDDl module in the Master APRM instrument communicates with one RBM channel whereas a separate FDDl module in the Slave APRM instrument communicates with the other RBM channel. Each RBM channel (RBM A and RBM B) receives input from either the Master APRM or the Slave APRM module of each APRM channel.
 
Each 2-Out-of-4 Voter receives trip votes from all four channels of APRM and provides outputs to its associated RPS trip system based on the voter logic. Each 2-Out-of-4 Voter also receives the bypass switch status and forwards their status to the other three 2-Out-of-4 Voters. Bypass processing is implemented to prohibit more than one channel in bypass.
 
The 2-Out-of-4 Voters associated with APRM channels A, C, B, and D are referred to as A1, A2, B1, and B2, respectively for the HCGS installation. Each voter has an input to its associated channel of the RPS trip system which are also typically referred to as trip system A (A1 and A2) and trip system B (B1 and B2) respectively.
 
The "NUMAC Power Range Neutron Monitoring (PRNM) System Architecture Description" shows the interfaces between safety-related and nonsafety-related portions of the PRNM system (Appendix A of Enclosure 2 and 3).
 
The existing system provides outputs to the Plant Process Computer (PPC) or CRIDS. These points are all presently hardwired individually to computer I/O cabinets. The new system provides a new computer interface by processing all signals serially from the APRM and RBM NUMAC instruments through two redundant NICs to CRIDS. The NIC also sends APRM and LPRM gain adjustment factors from the Core Monitoring System computer, via the RBM's and NIC's, to the APRMs. The NIC consists of computer hardware processors and software programs that perform specific tasks to interface between the NUMAC instruments and CRIDs.
All communication interfaces between the RBM, the NIC, and the PPC are non-safety related.      LAR H15-01 LR-N15-0178 29 of 47 Cyber Security requirements for these interfaces are addressed in accordance with the NRC approved cyber security plan for HCGS
: 5.
The PRNM upgrade includes Operator Display Assemblies (ODAs), which are installed at the operator's panel and provide process parameter s, trip and alarm status from the APRM and RBM channels. There are two APRM ODAs and two RBM ODAs. The APRM master and RBM instruments send this information to the ODAs over a one way fiber optic connection. The APRM ODA serves as an interface between the operator and the remote APRM and RBM instruments. Each APRM ODA is capable of displaying data from two APRM instruments in the same division. Each RBM ODA is capable of displaying data from two RBMs.
 
A single fiber optic bypass switch assembly will be installed on panel 10C651 in the main control room to select an APRM channel for bypass. The bypass switch has mutually exclusive positions, thus assuring that only one APRM/OPRM channel is bypassed at a time. This approach is consistent with the proposed TS operability requirements for three out of four APRM channels; thereby ensuring that no single failure will preclude a scram on a valid signal.
The station power sources to the PRNM system are from two independent battery-backed inverters. These inverters are classified as non-Safety Related since loss of output power due to open, short, or ground causes the PRNM system to trip. Safety Related electrical protection assemblies monitor voltage and frequency of the supply and trip when voltage or frequency are outside of the allowable range.
The PRNM system provides scram contact outputs from the 2-out-of-4 Voter for the following functions:  1) Neutron Flux- Upscale (Setdown) 
: 2) Simulated Thermal Power - Upscale 3) Neutron Flux- Upscale  4) Inoperative 
: 5) OPRM Upscale 5  Approved by the NRC via License Amendment Nos. 189 & 192 (ADAMS ML111861560 and ML12335A221)      LAR H15-01 LR-N15-0178 30 of 47 4.1.2 Detailed System Description The PRNM system is described in detail in "NUMAC Power Range Neutron Monitoring (PRNM) System Architecture Description" provided as Appendix A 6 of Enclosure 2 and 3). Additional discussion on the PRNM power supply and test capability is provided below.
 
The APRM and RBM NUMAC instruments are powered by Quad Low Voltage Power Supplies (QLVPS), which provide auctioneered DC power to the instruments. There are five QLVPS in the system, one for each APRM channel and one that powers both RBM channels. Each QLVPS receives input power from both 120VAC sources to the PRNM system. The 2-out-of-4 voters and RBM interface modules are powered directly from 120VAC and are divisionally separated between the two 120VAC sources.
 
Two high voltage power supplies are provided in each APRM instrument to supply 0 to 200Vdc for the LPRM detectors. One high voltage power supply provides normal power to the LPRM detectors while the second is available to perform detector IV curves or as a backup to the normal high voltage power supply.
 
The NUMAC instruments automatically execute continuous self-test while the instrument key lock switch is in the operate position. When the instrument key switch is placed in the INOP position the self-test is suspended and may be performed manually. The self-test performs memory and internal microprocessor checks, interrogates internal registers on the circuit modules and measures internal voltages. If a fault is detected it is traced to the module level and displayed. Users can interface with the instrument front panel display for more detailed diagnostic information. Loss of an essential function results in an instrument INOP condition and is alarmed and non-essential faults are alarmed. The NUMAC instruments also include additional manual testing features that are performed with key switch in the INOP position.
The Two-out-of-Four Voter modules include automatic self-testing and manual testing features.
The APRM master instrument monitors the testing features of the Two-out-of-Four Voter.
 
Each NUMAC APRM and RBM instrument has a means to perform calibration of the hardware.
The calibration process is for the most part automatic. Internal voltage and frequency standards are calibrated to National Institute of Standards and Technology traceable standards when in the INOP/Calibrate mode of operation. The following functions are calibrated; clock frequency, analog to digital and digital to analog converters, high voltage power supplies and isolation amplifiers. The calibration correction factors are stored in the instruments nonvolatile memory.
The components and interfaces of the HCGS PRNMS are shown in the diagram below; the shaded boxes represent the new PRNM system:
 
6  The System Architecture Description also includes the Communications evaluation discussed in ISG 06 Section D.7.2      LAR H15-01 LR-N15-0178 31 of 47 To support installation of the PRNM system at HCGS, additional assessments and requirements have been completed as described below.
 
4.1.2.1 PRNM LTR Plant Specific Responses The PRNM LTR (Reference 1) requires utility specific responses for each PRNM installation; these are provided in Appendix R of Enclosure 2 and 3. Appendix R provides a table of the LTR section numbers and Utility Action Required. As discussed previously, HCGS is also implementing the oscillation power range monitor (OPRM) stability trip function using the Detect and Suppress Solution-Confirmation Density (DSS-CD) solution. Thus, the plant-specific responses also include reference to the DSS-CD stability solution. 4.1.2.2 OPRM Transition to DSS-CD The PRNM system includes an Oscillation Power Range Monitor (OPRM) capability; to detect and suppress reactor instability. The OPRM function continues to satisfy the same regulatory requirements as the currently installed OPRM equipment. The existing ABB OPRM with BWROG Option III stability solution will change to the GEH OPRM with the Detect and Suppress Solution - Confirmation Density (DSS-CD) stability solution. Enclosure 2 and 3 Appendix T provides the evaluation and justification for implementing DSS-CD at HCGS.      LAR H15-01 LR-N15-0178 32 of 47 4.1.2.3 Transition to Full ARTS With the PRNM installation HCGS is transitioning from partial ARTS to full ARTS as described in the Supplemental Information for the Average Power Range Monitor, Rod Block Monitor and Technical Specification Improvement (ARTS) Program for Hope Creek Nuclear Generating Station (Appendix S) of Enclosure 2 and 3
: 7. The ARTS methodology was implemented at the Hope Creek Generating Station (HCGS) in 2006 (Amendment 163, ADAMS ML060620500 and ML060620470). In that implementation the
 
hardware portion of ARTS was omitted to avoid the physical plant modifications that would have been required. This non-hardware configuration resulted in RWE analyses performed for each reload with no credit for blocking of rod withdrawals. 
 
With the installation of PRNM, HCGS is transitioning from partial ARTS to full ARTS with the Control Rod Block setpoints changing from flow-biased to power-biased. The RWE event analyses performed for each future fuel cycle will take credit for RBM generated rod blocks during the rod withdrawal error event. The results of the RWE event analysis will be considered in establishing the cycle specific operating limits for the fuel. With the implementation of the fuel cycle specific RWE event analysis, the values the RBM setpoints will be contained in the COLR.
Establishing the fuel cycle specific RBM requirements in the COLR is consistent with Generic Letter 88-16. GL 88-16 permitted the relocation of fuel cycle specific parameter limits to the COLR. The updated TS reflect the deletion of the Table 3.3.6-2 Rod Block Monitor Functions 1.a.i Flow Biased and 1.a.ii High Flow Clamped; the addition of the Low Trip Setpoint (LTSP), Intermediate Trip Setpoint (ITSP), and High Trip Setpoint (HTSP); the addition of the Low Power Setpoint (LPSP), Intermediate Power Setpoint (IPSP), and High Power Setpoint (HPSP) ranges; and the modification of the Downscale Trip Setpoint (DTSP). As described in Section 2.0(a) and 3.2 of the Supplemental Information for ARTS (Enclosure 3 Appendix S), the full ARTS implementation removes the flow biased RBM setpoints and replaces them with power biased trips. As described in Enclosure 3 Appendix S, Section 3.3.1, paragraph 8, the trip setpoints (LTSP, ITSP, and HTSP) are dependent on the MCPR value provided in the reload analysis.
Due to cycle-by-cycle variation, the setpoint values are referenced in the COLR. The LPSP, IPSP, and HPSP are described in Table 5 and Section 3.3.1, paragraph 7, of Enclosure 3 Appendix S. Power setpoints do not change on a cycle-by-cycle basis. Therefore, these power setpoint ranges (provided in Appendix P of Enclosure 3) are referenced directly in the Technical Specifications. 
 
Lastly, as described in Enclosure 3 Appendix S Table 5, footnote 2, the DTSP is not important for RWE analysis and may be moved to the COLR.
7  Prior to implementation of the PRNM system, HCGS plans to transition from GE-14 to GNF2 fuel (Fall 2016). The cycle specific reload evaluations or analyses described in Appendices S and T of Enclosure 3 will be performed for the cycle of PRNM implementation and subsequent cycles for the fuel related PRNM elements, full implementation of ARTS and implementation of the GEH OPRM with the DSS-CD stability solution. The cycle specific reload evaluations or analyses will confirm the requirements described in Enclosure 3 Appendices S and T will be met on a cycle specific basis.      LAR H15-01 LR-N15-0178 33 of 47 4.1.2.4 Cyber Security Considerations HCGS established an NRC approved Cyber Security defensive strategy and program that is compliant with NEI 08-09 and has been approved by the NRC via License Amendment Nos.
189 & 192. The program is implemented via Plant procedure IT-AA-503. The defensive strategy establishes 4 security levels:
Level 4 - Control & Safety System Network - Plant control systems  Level 3 - Data Acquisition Network - Plant computer system network and associated components for collecting plant data  Level 2 - Site Local Area Network  Level 1 - Corporate Wide Area Network Cyber security boundary devices are defined according to Example 1 in NEI 08-09:
Firewall and network intrusion detection system implements boundary between Level 4 and level 3 (allows bi-directional communication, but implements information flow controls in NEI 08-09, Appendix D, Section 1.4 and Appendix E, Section 6)  Data diode implements deterministic boundary between Level 3 and Level 2 (only allows communication from Level 3 to Level 2) 4.1.2.5 Human Factors Evaluation The PRNM LTR Plant Specific Responses (Enclosure 2 and 3, Appendix R) Item 2.3.4 states:
For any changes to the plant operator's panel, document in the submittal the human factors review actions that were taken to confirm compatibility with existing plant commitments and procedures.
 
In addition, Section 5.0 of the SE for the LTR identifies six plant-specific actions that require confirmation; Action 6 requires confirmation that any changes to the plant operator's panel have received human factors reviews per plant-specific procedures.
 
Human Factors engineering is addressed as part of the PRNM design change package (DCP), including changes to the operator panel, as discussed below. The PRNM DCP modifies the 10C651 Operator Console arrangement to accommodate the NUMAC PRNM System ODAs and the change from 6 APRM channels to 4 APRM channels. These modifications include:  1. Removing APRM E/F channels from APRM channel status indicators and recorders  2. Removing Flow Unit Bypass switches 
: 3. Removing RBM Status indicator lights 
: 4. Relocating the Scram Discharge Piping Volume Piping Logic Test and Hi Level Scram Bypass controls. 5. Relocating the 4 remaining APRM Monitor Status lights  6. Relocating the IRM Monitor Status lights and the IRM Bypass switches 
: 7. Installing 2 APRM ODAs and 2 RBM ODAs  8. Replacement of two existing with one new APRM Bypass switch      LAR H15-01 LR-N15-0178 34 of 47 PSEG procedure NC.DE-TS.ZZ-1017 identifies design guidance for incorporating human factors engineering principles into design changes based on the guidance in NUREG-0700. The following design principles are applicable to the changes made by the PRNM DCP:  Functional Grouping  Highlight Component Grouping  Control Display Integration  Legend Pushbutton Guidelines  Key-operated Controls  Labeling  Demarcation  Detailed analysis of compliance with NUREG-0 700 will be documented with the completion of the detailed design. The Phase 2 submittal of this PRNM LAR (provided approximately one year after this Phase 1 submittal) will provide a description of the NUREG-0700 compliance. A discussion of the OE assessed to support the PRNM upgrade will also be provided.
4.1.2.6 TSTF-493 PSEG will implement TSTF-493 Option A for the LSSS functions affected by the PRNM upgrade. PSEG has reviewed the model application and safety evaluation (ADAMS ML100710442) and determined it is applicable to HCGS. 
 
Using the guidance of Appendix A of TSTF-493, the two Option A notes specified in the TSTF are applied to the channel calibration of the following affected LSSS functions:
 
TS Table 4.3.1.1-1:
Function 2. Average Power Range Monitors
: a. Neutron Flux- Upscale (Setdown)
: b. Simulated Thermal Power- Upscale
: c. Neutron Flux- Upscale
 
TS Table 4.3.6-1:
 
Function1. Rod Block Monitor a. Upscale
 
The addition of the two notes to the above functions is discussed in Section 2 of this  , and is reflected in the Attachment 2 TS Markup. 
 
The first note requires evaluation of channel performance for the condition where the as-found setting for the channel setpoint is outside its as-found tolerance, but conservative with respect to the AV. The channel evaluation verifies that channel performance continues to satisfy safety analysis assumptions and channel performance assumptions within the setpoint methodology.
The purpose of the assessment is to ensure confidence in channel performance prior to returning the channel to service.
The second note requires that the as-left setting for the channel be returned to within the as-left tolerance of the Nominal Trip Setpoint (NTSP). Where a setpoint more conservative than the      LAR H15-01 LR-N15-0178 35 of 47 NTSP is used in the plant surveillance procedures, the as-left and as-found tolerances, as applicable, will be applied to the surveillance procedure setpoint. This ensures that sufficient margin is maintained to the Safety Limit (SL) and/or Analytical Limit (AL). If the as-left channel setting cannot be returned to within the as-left tolerance of the NTSP, then the channel shall be declared inoperable. This note also indicates that the methodologies used for calculating the as-found and as-left tolerances are specified in the TS Bases. 4.1.3 System Response Time Enclosure 2(3) Appendix N provides the Response Time Analysis Report for the NUMAC PRNM. HCGS has reviewed the report and determined that the analysis bounds the HCGS requirements for the four trips listed in Table 1 of Appendix N.
4.2 System (Hardware and Software) Development for the HCGS PRNM System (Section D.2 and D.4 of DI&C-ISG-06)
The initial NUMAC PRNM development completed in the early to mid-1990s, and the acceptability of the system level approach, functionality to be provided, and software development processes, including V&V, was determined using the regulatory evaluation criteria applicable at that time. However, the applicable regulatory evaluation criteria changed since these earlier reviews and approvals, and these changes include criteria against which the PRNM development processes had not been previously evaluated.
The CGS PRNM upgrade was the first PRNM project to address the updated regulatory guidance (consistent with ISG-06). GEH has adjusted the NUMAC PRNM system and software development life-cycle to align with the current regulatory guidance. Enclosure 2(3) Appendices B through E and K describe the NUMAC PRNM system life cycle process for the HCGS upgrade. 4.2.1 Design Analysis Report: Methodology Modifications ISG-06 Section D.8.2 requires a design analysis report that identifies deviations to the life cycle methodology from a previous NRC approval. Enclosure 2(3) Appendix K describes the evolution of the NUMAC life cycle and maps the BTP 7-14 plans to the corresponding GEH NUMAC plans:
 
Mapping from BTP 7-14 Planning Documents to NUMAC Planning Documents BTP 7-14 Software Planning Documentation Applicable GEH NUMAC Project Documents Software Management Plan  HCGS System Management Plan Software Development Plan  NUMAC Systems Engineering Development Plan Software Quality Assurance Plan  NUMAC Systems Quality Assurance Plan Software Integration Plan  NUMAC Systems Engineering Development Plan Software Safety Plan  NUMAC Systems Independent Verification and Validation Plan      LAR H15-01 LR-N15-0178 36 of 47 Software Verification and Validation Plan  NUMAC Systems Independent Verification and Validation Plan Software Configuration Management Plan  NUMAC Systems Engineering Development Plan Software Test Plan  NUMAC Systems Independent Verification and Validation Plan The following plans are also named in BTP 7-
 
14:
* Software Installation Plan
* Software Maintenance Plan
* Software Training Plan
* Software Operations Plan These plans are not required to be submitted with the Phase 1 submittal; these are required for implementation of the digital system
 
(Phase 3).
The four GEH plans listed in the table jointl y define a development program for the HCGS NUMAC digital PRNM system that is consistent with NRC requirements for a high quality development process for software used in safety systems of nuclear power plants. NUREG 0800, Standard Review Plan, Branch Technical Position (BTP) 7-14, Guidance on Software Reviews for Digital Computer-Based Instrumentation and Control Systems, provides acceptance criteria for process planning. 4.2.2 NUMAC System Engineering Development Plan The NUMAC Systems Engineering Development Plan (Enclosure 2(3) Appendix B) addresses process planning characteristics defined in BTP 7-14 Section B.3.1.2, Software Development Plan; BTP 7-14 Section B.3.1.4, Software Integration Plan; and BTP 7-14 Section B.3.1.11, Software Configuration Management Plan. 4.2.3 NUMAC System Quality Assurance Plan The NUMAC Systems Quality Assurance Plan (Enclosure 2(3) Appendix C) addresses process planning characteristics defined in BTP 7-14 Section B.3.1.3, Software Quality Assurance Plan. 4.2.4 NUMAC System Independent Verification & Validation Plan The NUMAC Systems Independent Verification and Validation Plan (Enclosure 2(3) Appendix D) addresses process planning characteristics defined in BTP 7-14 Section B.3.1.9, Software Safety Plan; BTP 7-14 Section B.3.1.10, Software Verification and Validation Plan; and BTP 7-14 Section B.3.1.12, Software Test Plan. 4.2.5 Hope Creek Generating Station NUMAC PRNM System Management Plan The HCGS NUMAC PRNM System Management Plan(Enclosure 2(3) Appendix E) addresses process planning characteristics defined in BTP 7-14 Section B.3.1.1, Software Management
 
Plan.
4.3 Software Architecture / Design Outputs (Section D.3 of DI&C-ISG-06) BTP 7-14, Section B.3.3.2 requires a description of the software used in the computer (or platform) and the application software, how the software functions, how the various software components are interrelated, and how the software utilizes the hardware. This information is typically contained in the (a) platform and application software architecture description, (b) the      LAR H15-01 LR-N15-0178 37 of 47 platform and application software requirements specification, and (c) the platform and application software design specification. 
 
The (a) software architecture description is provided in the System Architecture Description document (Enclosure 2(3) Appendix A) as discussed previously in Section 4.1. The following documents describe (b) the software requirements specification and (c) the software design specification. 4.3.1 System Requirements Specification & APRM Performance Specification  The NUMAC PRNM System Requirements Specification (Enclosure 2(3) Appendix F) defines the system requirements for the design and manufacture of a NUMAC based PRNM System. The NUMAC PRNM System is designed to replace the existing APRM and RBM channels in a Neutron Monitoring System (NMS) of a Boiling Water Reactor
 
(BWR) nuclear power plant. The NUMAC PRNM System also provides the OPRM channels required for the detection of reactor instability, the ARTS functions, and the implementation of the 2/4 Logic interface to the Reactor Protection System (RPS).
The NUMAC APRM DSS-CD Performance Specification (Enclosure 2(3) Appendix F) defines the performance characteristics and application limits for the HCGS NUMAC APRM application which includes the OPRM Detect and Suppress Solution -
Confirmation Density (DSS-CD) and automatic Backup Stability Protection (BSP) functions. 4.3.2 APRM Functional Controller System Design Specification  APRM Functional Controller System Design Specification (Enclosure 2(3) Appendix G) comprises the high level design of the NUMAC APRM Functional Controller software. The purpose of this document is twofold:  Define the Functional software design in sufficient detail such that software implementation can be undertaken without need for major design decisions. Provide a means for understanding how the NUMAC Functional Controller Software fulfills design input requirements. 4.4 Environmental Equipment Qualification (Section D.5 of DI&C-ISG-06) Documentation of equipment qualification, that confirms that the equipment qualification envelopes plant-specific requirements, is required in the plant-specific license amendment when referencing the previously approved LTR. The equipment qualification activities on the PRNM system comply with IEEE Standard 603 Clause 5.4 (Reference Enclosure 2(3) Appendix O), in accordance with the requirements of IEEE Standard 323 ("IEEE Standard for Qualifying Class 1E Equipment for Nuclear Power Generating Stations," 1983) and the guidance of IEEE Standard 344 ("IEEE Recommended Practices for Seismic Qualification of Class 1E Equipment for Nuclear Power Generating Stations," 2004). 
 
The Hope Creek NUMAC PRNM System Qualification Program (Enclosure 2(3) Appendix H) identifies the requirements to which the replacement PRNM system equipment will be qualified. The program provides guidance for the qualification of the replacement PRNM system and direction on whether the equipment is to be qualified by type-testing, by analysis, or by a combination of the two. The replacement PRNM system components are to be qualified based on operating in conditions considered mild-environments in which they are located. The equipment qualification includes temperature, humidity, pressure, radiation, seismic, and electromagnetic compatibility (EMC).      LAR H15-01 LR-N15-0178 38 of 47 4.5 Defense-In-Depth & Diversity (Section D.6 of DI&C-ISG-06)
BTP 7-19 and DI&C-ISG-02 8 provide guidance to address diversity and defense-in-depth (D3). The D3 analyses must confirm that vulnerabilities to common-cause failures (CCFs) have been adequately addressed; to provide reasonable assurance that CCFs do not defeat either the protection provided by alternative means (i.e., an independent and diverse safety function) or an echelon of defense that provides defense-in-depth.
The Hope Creek NUMAC PRNM Diversity and Defense in Depth Analysis (Enclosure 2(3)
Appendix I) provides the HCGS D3 analysis demonstrating compliance with the BTP 7-19 criteria. 4.6 Communications (Section D.7 of DI&C-ISG-06) Since the prior review and approval of the NUMAC PRNM LTR further NRC staff guidance has been made available that provides evaluation criteria applicable to safety-to-non-safety interfaces of digital inter-channel communication. DI&C-ISG-04, "Task Working Group #4:
Highly-Integrated Control Rooms-Communications Issues (HICRc)," provides current guidance on addressing communication issues in three areas:
Interdivisional Communications  Command Prioritization  Multi-divisional Control and Display Stations Section 7 of the NUMAC PRNM System Architecture Description (Enclosure 2(3) Appendix A) provides the PRNM communication analysis satisfying the positions in DI&C-ISG-04. 4.7 System, Hardware, Software, and Methodology Modifications (Deviations from the Prior LTRs) (Section D.8 of DI&CISG-06) A Design Analysis Report is required that identifies deviations to the system, hardware, software, or design lifecycle methodology 9 from a previous NRC approval of a digital I&C system or approved topical report. Enclosure 2(3) Appendix J provides the HCGS Design Analysis Report that identifies the deviations to the system, hardware, and software from previous NRC approvals. The "Hope Creek NUMAC PRNM System, Hardware, Software, and Methodology Modifications" Design Analysis Report addresses three specific areas:
 
Hope Creek deviations from the approved NUMAC PRNM LTR (Reference 1)  Hope Creek deviations from the approved GE Boiling Water Reactor DSS-CD LTR (Reference 2). There are no deviations. HCGS differences from the CGS NUMAC PRNM system that was reviewed and approved by the NRC (Reference 3). The proposed HCGS system is very similar to the CGS system which is cited as a precedent. Note that consistent with Reference 1.a Section 5.3.5.7 HCGS does have an RRCS output; CGS does not. 4.8 Compliance with IEEE Standard 603 (Section D.9 of DI&C-ISG-06) Enclosure 2(3) Appendices O, M, L and P discuss how the PRNM System meets the requirements of IEEE Standard 603-1991.
8  ISG-02 is superseded by BTP 7-19 Revision 6 9  Deviations to the design lifecycle methodology are provided in a separate report as discussed in Section 4.2.1 (Design Analysis Report: [Life Cycle] Methodology Modifications)      LAR H15-01 LR-N15-0178 39 of 47 4.8.1 Report on Compliance with IEEE Standards (603-1991 and 7-4.3.2-2003) and Theory of Operations Description Enclosure 2(3) Appendix O. This report discusses how the NUMAC PRNM complies with applicable clauses in IEEE Standard 603-1991 and IEEE Standard 7-4.3.2-2003. The report
 
also provides a discussion on the PRNM Theory of Operations Description (ISG-06 Enclosure B Item 1.20).
4.8.2 Design Report on Computer Integrity, Test and Calibration, and Fault Detection (IEEE Standard 603-1991 Clause 5.5)  Enclosure 2(3) Appendix M. This report addresses DI&C-ISG-06 Sections D.9.4.2.5, D.9.4.2.7, D.9.4.2.10, D.9.4.3.5, D.10.4.2.5, D.10.4.2.5.1, D.10.4.2.5.2, D.10.4.2.5.3 and D.10.4.2.7 for the PRNM System (ISG-06 Enclosure B Item 1.
18). Consequently, the report demonstrates compliance with IEEE Standard 603-1991, Clauses 5.5, 5.7, 5.10 and 6.5, and IEEE Standard 7-4.3.2-2003, Clauses 5.5, 5.5.1, 5.5.2, 5.5.3 and 5.7. 4.8.3 Design Analysis Report: Electrical Independence (IEEE Standard 603-1991 Clause 5.6)  Enclosure 2(3) Appendix L. This report addresses DI&C-ISG-06 Section D.9.4.2 (ISG-06 Enclosure B item 1.16); how the PRNM system complies with IEEE Standard 603-1991 Clause 5.6. 4.8.4 Setpoint Methodology and Calculations (IEEE Standard 603-1991 Clause 6.8)
Enclosure 2(3) Appendix P. This report provides the Setpoint Methodology Overview that addresses DI&C-ISG-06 Section D.9.4.3.8 (ISG-06 Enclosure B item 1.21); how the PRNM system complies with IEEE Standard 603-1991 Clause 6.8. 
 
The proposed TS changes and TS markups (Section 2.0 and Attachment 2) provide the resulting calculated setpoints
: 10. Appendix P also includes the Instrument Limit Calculation documents (for APRM and RBM) that provide the inputs and calculated setpoint results. The supporting calculations are available for NRC audi t; in lieu of submitting the calculations. The data in support of the calculations are contained in large data bases, so it would be difficult to provide the data and it would take more time and resources to review the calculations, if submitted. This is consistent with the approach used for the CGS PRNM upgrade review and
 
approval.
10  As discussed in ISG-06, setpoint calculations are to be provided with the Phase 2 submittal; however the TS mark-up in this Phase 1 submittal includes the setpoint changes.      LAR H15-01 LR-N15-0178 40 of 47 4.9 Conformance with IEEE Standard 7-4.3.2 (Section D.10 of DI&C-ISG-06)  (3) Appendix O discusses how the PRNM system meets the requirements of IEEE Standard 7-4.3-2003. This report discusses how the NUMAC PRNM complies with applicable clauses in IEEE Standard 603-1991 and IEEE Standard 7-4.3.2-2003. This includes discussion of the Software Tool Verification Program (ISG-06 D.10.4.2.3.2), the Software Project Risk Management Program (ISG-06 D.10.4.2.3.6), and the Commercial Grade Dedication Plan (ISG-06 D.10.4.2.4.2).
 
The report also provides a discussion on the PRNM Theory of Operations Description (ISG-06 Enclosure B Item 1.20). 4.10 Secure Development and Operational Environment (Section D.12 of DI&C-ISG-06) Enclosure 2(3) Appendix Q, "Secure Development and Operational Environment and Vulnerability Assessment Report" addresses secure software development and operation throughout the PRNM product development to ensure the system is reliable (ISG-06 Enclosure B Item 1.27). This report also includes the PRNM system Vulnerability Assessment (ISG-06 Enclosure B Item 1.26). 4.11 Confirmation of Plant-Specific Actions Section 5.0 of the SE for the NUMAC PRNM LTR (Reference 1a) identifies six plant-specific actions that are required when a licensee references the LTR as part of a license amendment submittal. This section identifies each of these actions and the HCGS confirmation of each action.  (1) Confirm the applicability of the NUMAC PRNM LTR (NEDC-32410P-A), including clarifications and reconciled differences between the specific plant design and the topical report design descriptions.
This license amendment request identifies the specific HCGS PRNM configuration and the general applicability of NEDC-32410P-A. The differences and deviations from the
 
LTR (including differences from the CGS precedent) are provided, and justified, in the System, Hardware, Software, and Methodology Modifications (Deviations from the Prior LTRs) Report (Enclosure 2(3) Appendix J).
(2) Confirm the applicability of the BWROG topical reports that address the PRNM system and associated instability functions, set points and margins.
The applicability of the BWROG topical reports that address the PRNM system and its associated instability functions, set points and margins is provided in the DSS-CD Evaluation (Enclosure 2(3) Appendix T).
(3) Provide plant-specific revised Technical Specification pages for the PRNM system functions consistent with NEDC-32410P-A, Appendix H.
The HCGS TS changes are identified and justified in Section 2.0 of this Attachment 1 to the LAR; the marked-up TS are provided in Attachment 2 of this submittal. 
(4) Confirm the plant-specific environmental conditions are enveloped by the PRNM system equipment qualifications values.      LAR H15-01 LR-N15-0178 41 of 47 The evaluation of HCGS specific environmental conditions and qualification is provided in the NUMAC Power Range Neutron Monitor System Qualification Program - Hope Creek Generating Station (Enclosure 2(3) Appendix H). 
(5) Confirm that administrative controls ar e provided for manually bypassing APRM/OPRM channels or protective functions, and for controlling access to the panel and the APRM/OPRM channel bypass switch.
The Phase 2 submittal of this PRNM LAR (provided approximately one year after this Phase 1 submittal) will provide confirmation of this action (as discussed in Section 4.1.2.5 above). 
(6) Confirm that any changes to the plant operator's panel have received human factors reviews per plant-specific procedures.
The Phase 2 submittal of this PRNM LAR (provided approximately one year after this Phase 1 submittal) will provide confirmation of this action (as discussed in Section
 
4.1.2.5 above).
 
LAR H15-01 LR-N15-0178 42 of 47 5.0 REGULATORY ANALYSIS The PRNM upgrade incorporates redundancy, independence, and diversity while providing simplified management and maintenance of the system. The effect of the PRNM upgrade on TS and accident analyses has been evaluated. Appropriate setpoints have been evaluated for the new system and the TS accordingly revised. The required Defense-in-Depth and Diversity (D3) report has been prepared confirming that vulnerabilities to common-cause failures (CCFs) have been adequately addressed. The hardware and software development for the PRNM upgrade process complies with the Institute of Electrical and Electronics Engineers (IEEE) Standard 603-1991 Clause 5.3 "Quality," and IEEE Standard 7-4.3.2-2003 Clause 5.3 "Quality," including the digital system development life cycle, in order to provide a high quality and well defined development process. The independent V&V effort for the upgrade utilizes a process and activities that comply with IEEE Standard 7-4.3.2-2003 Clause 5.3.3, "Validation and Verification" to ensure the upgrade meets required specified functional requirements and criteria. Finally, the Software configuration management used for the upgrade complies with IEEE Standard 7-4.3.2-2003 Clause 5.3.5, "Software Configuration Management," control the system and programming throughout its development and use. 
 
Therefore, PSEG concludes the proposed PRNM upgrade complies with the 10 CFR 50 regulations and associated regulatory guidance. 5.1 Applicable Regulatory Requirements/Criteria The following regulations and guidance are applicable to the proposed installation of the GEH
 
NUMAC PRNM equipment:
 
10 CFR 50.36, "Technical Specifications."
Paragraph 10 CFR 50.55a(a)(1), states that Structures, Systems, and Components must be designed, fabricated, erected, constructed, tested, and inspected to quality standards commensurate with the importance of the safety function to be performed.
Paragraph 10 CFR 50.55a(h), "Protection and safety systems," approves the 1991 version of IEEE Standard 603, "IEEE Standard Criteria for Safety Systems for Nuclear Power Generating Stations," for incorporation by reference including the correction sheet dated January 30, 1995.
The following General Design Criteria (GDC) in Appendix A to 10 CFR Part 50:
GDC 1, "Quality standards and records" GDC 2, "Design bases for protection against natural phenomena" GDC 4, "Environmental and dynamic effects design bases" GDC 10, "Reactor design" GDC 12, "Suppression of reactor power oscillations" GDC 13, "Instrumentation and control" GDC 15, "Reactor coolant system design" GDC 19, "Control Room" GDC 20, "Protection system functions"      LAR H15-01 LR-N15-0178 43 of 47 GDC 21, "Protection system reliability and testability" GDC 22, "Protective system independence" GDC 23, "Protection system failure modes" GDC 24, "Separation of protection and control systems" GDC 25, "Protection system requirement s for reactivity control malfunctions" GDC 29, "Protection against anticipated operational occurrences" Regulatory Guide 1.75, Revision 3, "Criteria for Independence of Electrical Safety Systems," February 2005 (ADAMS Accession No. ML043630448).
Regulatory Guide 1.100, Revision 3, "Seismic Qualification of Electrical and Active Mechanical Equipment and Functional Qualification of Active Mechanical Equipment for Nuclear Power Plants," September 2009 (ADAMS Accession No. ML091320468).
Regulatory Guide 1.105, Revision 3, "Setpoints for Safety Related Instrumentation," December 1999 (ADAMS Accession No. ML993560062).
Regulatory Guide 1.152, Revision 3, "Criteria for Use of Computers in Safety Systems of Nuclear Power Plants," July 2011 (ADAMS Accession No. ML102870022).
Regulatory Guide 1.168, Revision 2, "Verification, Validation, Reviews, and Audits for Digital Computer Software Used in Safety Systems of Nuclear Power Plants," July 2013 (ADAMS Accession No. ML13073A210).
Regulatory Guide 1.169, Revision 1, "Configuration Management Plans for Digital Computer Software Used in Safety Systems of Nuclear Power Plants," July 2013 (ADAMS Accession No. ML12355A642).
Regulatory Guide 1.170, Revision 1, "Software Test Documentation for Digital Computer Software Used in Safety Systems of Nuclear Power Plants," July 2013 (ADAMS Accession No. ML13003A216).
Regulatory Guide 1.171, Revision 1,"Software Unit Testing for Digital Computer Software Used in Safety Systems of Nuclear Power Plants," July 2013 (ADAMS Accession No. ML13004A375).
Regulatory Guide 1.172, Revision 1 "Software Requirements Specifications for Digital Computer Software and Complex Electronics Used in Safety Systems of Nuclear Power Plants," July 2013 (ADAMS Accession No. ML13007A173).
Regulatory Guide 1.173, Revision 1 "Developing Software Life-Cycle Processes for Digital Computer Software Used in Safety Systems of Nuclear Power Plants," July 2013 (ADAMS Accession No. ML13009A190).
Regulatory Guide 1.180, Revision 1, "Guidelines for Evaluating Electromagnetic and Radio-Frequency Interference in Safety-Related Instrumentation and Control Systems," October 2003 (ADAMS Accession No. ML032740277).
LAR H15-01 LR-N15-0178 44 of 47  Regulatory Guide 1.209, "Guidelines for Environmental Qualification of Safety-Related Computer-Based Instrumentation and Control Systems in Nuclear Power Plants," March 2007 (ADAMS Accession No. ML070190294).
DI&C-ISG-04, Revision 1, "Task Working Group #4: Highly-Integrated Control Rooms-Communications Issues (HICRc)," March 2007 (ADAMS Accession No. ML083310185).
DI&C-ISG-06, "Task Working Group #6: Licensing Process," Revision 1, dated January 19, 2011 (ADAMS ML110140103).
 
The applicable portions of the following branch technical positions within NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition" (SRP), Chapter 7, "Instrumentation and Controls," as follows:
 
Branch Technical Position 7-11, "Guidance on Application and Qualification of Isolation Devices"  Branch Technical Position 7-12, "Guidance on Establishing and Maintaining Instrument Setpoints"  Branch Technical Position 7-14, "Guidance on Software Reviews for Digital Computer-Based Instrumentation and Control Systems"  Branch Technical Position 7-19, "Guidance for Evaluation of Diversity and Defense-In-Depth in Digital Computer-Based Instrumentation and Control Systems"  Branch Technical Position 7-21, "Guidance on Digital Computer Real-Time Performance" 5.2 No Significant Hazards Consideration In accordance with 10 CFR 50.90, PSEG Nuclear LLC (PSEG) requests an amendment to Renewed Facility Operating License No. NPF-57 for Hope Creek Generating Station (HCGS).
The proposed license amendment request (LAR) would reflect the installation of the General Electric-Hitachi (GEH) digital Nuclear Measurement Analysis and Control (NUMAC) Power Range Neutron Monitoring (PRNM) system. The planned upgrade will replace the existing analog Average Power Range Monitor (APRM) sub-sy stem of the existing Neutron Monitoring System with the more reliable, digital NUMAC PRNM System. The system upgrade incorporates the Oscillation Power Range Monitor (OPRM) function and the transition from flow-biased to power biased Rod Block Monitor (RBM).
 
The following Technical Specifications (TS) sections are affected by this change:
* TS 2.2  Limiting Safety System Settings
* TS 3/4.1.4.3  Rod Block Monitor
* TS 3/4.3.1  Reactor Protection System Instrumentation
* TS 3/4.3.6  Control Rod Block Instrumentation
* TS 3/4.3.11  Oscillation Power Range Monitor
* TS 3/4.4.1  Recirculation System
* TS 6.9.1.9  Core Operating Limits Report
* TS 6.9.3  Special Reports
 
LAR H15-01 LR-N15-0178 45 of 47 PSEG has evaluated whether a significant hazards consideration is involved with the proposed amendment by focusing on the three conditions set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:
: 1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No.
 
The probability of accidents occurring is not affected by the PRNM system, as the PRNM system is not the initiator of any accident and does not interact with equipment whose failure could cause an accident. The transition from flow-biased to power-biased RBM does not increase the probability of an accident; the RBM is not involved in the initiation of any accident. The regulatory criteria established for the APRM, OPRM, and RBM systems will be maintained with the installation of the upgraded PRNM system. Therefore, the proposed change does not involve a significant increase in the probability of an accident previously evaluated.
 
The consequences of accidents are not affected by the PRNM system, as the setpoints in the PRNM system will be established so that all analytical limits are met. The unavailability of the new system will be equal to or less than the existing system and, as a result, the scram reliability will be equal to or better than the existing system. No new challenges to safety-related equipment will result from the PRNM system modification. The change to power biased RBM allows for Rod Withdrawal Error (RWE) analyses performed for each future reload to take credit for rod blocks during the rod withdrawal transients. The results of the RWE event analysis will be used in establishing the cycle specific operating limits for the fuel. The proposed change will also replace the currently installed and NRC approved Asea Brown Boveri (ABB) OPRM Option III long-term stability solution with an NRC approved General Electric-Hitachi (GEH) Detect and Suppress Solution - Confirmation Density (DSS-CD) stability solution (reviewed and approved by the NRC in Reference 2, Licensing Topical Report). The OPRM meets the GDC 10, "Reactor Design," and 12, "Suppression of Reactor Power Oscillations," requirements by automatically detecting and suppressing design basis thermal hydraulic oscillations to protect specified fuel design limits. Therefore, the proposed change does not involve a significant increase in the consequences of an accident previously evaluated.
 
Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. 
: 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?
Response: No.
The components of the PRNM system will be supplied to equivalent or better design and qualification criteria than is currently required for the plant. Equipment that could be affected by PRNM system has been evaluated. No new operating mode, safety-related equipment lineup, accident scenario, or system interaction mode was identified. Therefore, the upgraded PRNM system will not adversely affect plant equipment.
LAR H15-01 LR-N15-0178 46 of 47 The new PRNM system uses digital equipment that has software controlled digital processing points and software controlled digital processing compared to the existing PRNM system that uses mostly analog and discrete component processing (excluding the existing OPRM). Specific failures of hardware and potential software common cause failures are different from the existing system. The effects of potential software common cause failure are mitigated by specific hardware design and system architecture as discussed in Section 6.0 of the NUMAC PRNM LTR, and supported by a plant specific evaluation. The transition from a flow-biased RBM to a power dependent RBM does not change its function to provide a control rod block when specified setpoints are reached. The change does not introduce a sequence of events or introduce a new failure mode that would create a new or different type of accident. Failure(s) of the system have the same overall effect as the present design. No new or different kind of accident is introduced. Therefore, the PRNM system will not adversely affect plant equipment.
 
The currently installed APRM System is replaced with a NUMAC PRNM system that performs the existing power range monitoring functions and adds an OPRM to react automatically to potential reactor thermal-hydraulic instabilities.
Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.
: 3. Does the proposed amendment involve a significant reduction in a margin of safety?
Response: No. 
 
The proposed TS changes associated with the NUMAC PRNM system implement the constraints of the NUMAC PRNM system design and related stability analyses. The NUMAC PRNM system change does not impact reactor operating parameters or the functional requirements of the PRNM system. The replacement equipment continues to provide information, enforce control rod blocks, and initiate reactor scrams under appropriate specified conditions. The power dependent RBM will continue to prevent rod withdrawal when the power-dependent RBM rod block setpoint is reached. The MCPR and
 
Linear Heat Generation Rate (LHGR) thermal limits will be developed on a cycle specific basis to ensure that fuel thermal mechanical design bases remain within the licensing limits during a control rod withdrawal error event and to ensure that the MCPR SL will not be violated as a result of a control rod withdrawal error event.
 
The proposed change does not reduce safety margins. The replacement PRNM equipment has improved channel trip accuracy compared to the current analog system, and meets or exceeds system requirements previously assumed in setpoint analysis. The power dependent RBM will support cycle specific RWE analysis ensuring fuel limits are not exceeded. Thus, the ability of the new equipment to enforce compliance with margins of safety equals or exceeds the ability of the equipment which it replaces.
 
Therefore, the proposed change does not involve a significant reduction in a margin of safety.
Based on the above, PSEG concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of no significant hazards consideration is justified.
LAR H15-01 LR-N15-0178 47 of 47 5.3 Conclusions In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
 
==6.0 ENVIRONMENTAL CONSIDERATION==
 
A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or a significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.
 
==7.0 REFERENCES==
: 1. a. GE Nuclear Energy, Nuclear Measurement Analysis and Control Power Range Neutron Monitor (NUMAC PRNM) Retrofit Plus Option III Stability Trip Function,
 
NEDC-32410P-A, Volume 1, dated October 1995. b. GE Nuclear Energy, Nuclear Measurement Analysis and Control Power Range Neutron Monitor (NUMAC PRNM) Retrofit Plus Option III Stability Trip Function,
 
NEDC-32410P-A, Volume 2, dated October 1995. c. GE Nuclear Energy, Nuclear Measurement Analysis and Control Power Range Neutron Monitor (NUMAC PRNM) Retrofit Plus Option III Stability Trip Function, NEDC-32410P-A, Supplement 1, dated November 1997.
: 2. GE Hitachi Boiling Water Reactor Detect and Suppress Solution-Confirmation Density, NEDC-33075P-A, Revision 8, November 2013.
: 3. Columbia Generating Station - Issuance of Amendment RE: Implementation of Power Range Neutron Monitoring/Average Power Rang e Monitor/Rod Block Monitor/Technical Specifications/Maximum Extended Load Line Limit Analysis (PRNM/ARTS/MELLLA) (TAC NO. ME7905) (ADAMS ML ML13317B623, Non-Proprietary).
LAR H15-01 LR-N15-0178 Mark-up of Proposed Technical Specification Pages
 
The following Technical Specifications pages for Renewed Facility Operating License NPF-57 are affected by this change request:
 
Technical Specification    Page Index        x, xvii, xviii
 
2.2, "Limiting Safety System Settings"    2-4
 
3/4.1.4.3, "Rod Block Monitor"    3/4 1-18
 
3/4.3.1, "Reactor Protection System Instrumentation"  3/4 3-1, 2, 4, 5, 7 and 8
 
3/4.3.6, "Control Rod Block Instrumentation"    3/4 3-57, 58, 59, 60 and 61
 
3/4.3.11, "Oscillation Power Range Monitor"    3/4 3-110 3/4.4.1, "Recirculation System"    3/4 4-1 and 4-2
 
6.9.1.9, "Core Operating Limits Report"    6-20
 
6.9.3, "Special Reports"      6-21
 
LAR H15-01 LR-N15-0178 Mark-up of Proposed Technical Specification Bases Pages
 
LAR H15-01 LR-N15-0178 ISG-06 Enclosure B Roadmap
 
Two variations of the roadmap are included in this attachment; both map the documents provided in Enclosure 1 (2) of this submittal against ISG-06 Enclosure B:
: 1. ISG-06 Enclosure B Item mapped to the HCGS PRNM GEH Document(s)
: 2. HCGS PRNM GEH Document mapped to the ISG-06 Enclosure B Item(s)
 
HCGSPRNMUpgradeISG06EnclosureBRoadmapISG06EncBtoPhase1DocumentMapping(Tier2Submittal)
EncB#1.1HardwareArchitectureDescriptions(D.1.2)ISG06Section:D.1.2Phase1(GEH)DocumentLAREnclosure3(NEDC33864P)Appendix:
A001N2029PRNMSystemArchitectureDescription LARSection:4.1.1,4.1.2RevNo:3EncB#1.3SoftwareArchitectureDescriptions(D.3.2,D.4.4.3.2)
ISG06Section:D.3.2,D.4.4.3.2Phase1(GEH)DocumentLAREnclosure3(NEDC33864P)Appendix:
A001N2029PRNMSystemArchitectureDescription LARSection:4.1.1,4.1.2RevNo:3EncB#1.4SoftwareManagementPlan(D.4.4.1.1)
ISG06Section:D.4.4.1.1Phase1(GEH)DocumentLAREnclosure3(NEDC33864P)Appendix:
E002N4398HopeCreekNUMACPRNMSystemManagementPlanLARSection:4.2.5RevNo:1EncB#1.5SoftwareDevelopmentPlan(D.4.4.1.2)
ISG06Section:D.4.4.1.2Phase1(GEH)DocumentLAREnclosure3(NEDC33864P)Appendix:
BNEDE33834PNUMACSystemsEngineeringDevelopmentPlanLARSection:4.2.2RevNo:0Wednesday,September9,20Page1of9ISG06EnclosureBtoPhase1DocumentMapping EncB#1.6SoftwareQAPlan(D.4.4.1.3,D.10.4.2.3.1)
ISG06Section:D.4.4.1.3,D.10.4.2.3.1Phase1(GEH)DocumentLAREnclosure3(NEDC33864P)Appendix:
CNEDE33836PNUMACSystemsQualityAssurancePlanLARSection:4.2.3RevNo:0EncB#1.7SoftwareIntegrationPlan(D.4.4.1.4)
ISG06Section:D.4.4.1.4Phase1(GEH)DocumentLAREnclosure3(NEDC33864P)Appendix:
BNEDE33834PNUMACSystemsEngineeringDevelopmentPlanLARSection:4.2.2RevNo:0EncB#1.8SoftwareSafetyPlan(D.4.4.1.9)
ISG06Section:D.4.4.1.9Phase1(GEH)DocumentLAREnclosure3(NEDC33864P)Appendix:
DNEDE33835PNUMACSystemsIndependentVerification&ValidationPlanLARSection:4.2.4RevNo:0EncB#1.9SoftwareV&VPlan(D.4.4.1.10)
ISG06Section:D.4.4.1.10Phase1(GEH)DocumentLAREnclosure3(NEDC33864P)Appendix:
DNEDE33835PNUMACSystemsIndependentVerification&ValidationPlanLARSection:4.2.4RevNo:0Wednesday,September9,20Page2of9ISG06EnclosureBtoPhase1DocumentMapping EncB#1.10SoftwareConfigurationManagementPlan(D.4.4.1.11)
ISG06Section:D.4.4.1.11Phase1(GEH)DocumentLAREnclosure3(NEDC33864P)Appendix:
BNEDE33834PNUMACSystemsEngineeringDevelopmentPlanLARSection:4.2.2RevNo:0EncB#1.11SoftwareTestPlan(D.4.4.1.12)
ISG06Section:D.4.4.1.12Phase1(GEH)DocumentLAREnclosure3(NEDC33864P)Appendix:
DNEDE33835PNUMACSystemsIndependentVerification&ValidationPlanLARSection:4.2.4RevNo:0EncB#1.12.1SoftwareRequirementsSpecification(D.4.4.3.1)
ISG06Section:D.4.4.3.1Phase1(GEH)DocumentLAREnclosure3(NEDC33864P)Appendix:
F126A8742SANUMACPRNMSystemRequirementsSpecification LARSection:4.3.1RevNo:3EncB#1.12.2SoftwareRequirementsSpecification(D.4.4.3.1)
ISG06Section:D.4.4.3.1Phase1(GEH)DocumentLAREnclosure3(NEDC33864P)Appendix:
F2000N6426NUMACAPRMDSSCDPerformanceSpecification LARSection:4.3.1RevNo:3Wednesday,September9,20Page3of9ISG06EnclosureBtoPhase1DocumentMapping EncB#1.13SoftwareDesignSpecification(D.4.4.3.3)
ISG06Section:D.4.4.3.3Phase1(GEH)DocumentLAREnclosure3(NEDC33864P)Appendix:
G002N2038APRMFunctionalControllerSDSLARSection:4.3.2RevNo:1EncB#1.14EquipmentQualificationTestingPlans(IncludingEMI,Temperature,Humidity,andSeismic)(D.5.2)ISG06Section:D.5.2Phase1(GEH)DocumentLAREnclosure3(NEDC33864P)Appendix:
H001N6665NUMACQualificationProgramHCGSLARSection:4.4RevNo:2EncB#1.15D3Analysis(D.6.2)ISG06Section:D.6.2Phase1(GEH)DocumentLAREnclosure3(NEDC33864P)Appendix:
I003N0063DiversityandDefenseinDepthAnalysisLARSection:4.5RevNo:0EncB#1.16.1DesignAnalysisReports(D.7.2)Communications ISG06Section:D.7.2Phase1(GEH)DocumentLAREnclosure3(NEDC33864P)Appendix:
A001N2029PRNMSystemArchitectureDescription LARSection:4.1,4.6RevNo:3Wednesday,September9,20Page4of9ISG06EnclosureBtoPhase1DocumentMapping EncB#1.16.2DesignAnalysisReports(D.8.2)ISG06Section:D.8.2Phase1(GEH)DocumentLAREnclosure3(NEDC33864P)Appendix:
J001N5783DesignAnalysisReport:HCNUMACPRNMSystem,Hardware,andSoftwareModifications LARSection:4.7RevNo:2EncB#1.16.3DesignAnalysisReports(D.8.2)ISG06Section:D.8.2Phase1(GEH)DocumentLAREnclosure3(NEDC33864P)Appendix:
K001N8626DesignAnalysisReport:MethodologyModifications LARSection:4.2.1RevNo:2EncB#1.16.4DesignAnalysisReports(D.9.4.2.6,D.10.4.2.6)
ISG06Section:D.9.4.2.6,D.10.4.2.6Phase1(GEH)DocumentLAREnclosure3(NEDC33864P)Appendix:
L001N7851DesignAnalysisReportonElectricalIndependence LARSection:4.8.3RevNo:2EncB#1.17SystemDescription(Toblockdiagramlevel)(D.9.2,D.10.2)ISG06Section:D.9.2,D.10.2Phase1(GEH)DocumentLAREnclosure3(NEDC33864P)Appendix:
A001N2029PRNMSystemArchitectureDescription LARSection:4.1.1,4.1.2RevNo:3Wednesday,September9,20Page5of9ISG06EnclosureBtoPhase1DocumentMapping EncB#1.18DesignReportonComputerintegrity,TestandCalibration,andFaultDetection(D.9.4.2.5,D.9.4.2.7,D.9.4.2.10,D.9.4.3.5,D.10.4.2.5,D.10.4.2.7)
ISG06Section:D.9.4.2.5,D.9.4.2.7,D.9.4.2.10,D.9.4.3.5,D.10.4.2.5,D.10.4.2.7Phase1(GEH)DocumentLAREnclosure3(NEDC33864P)Appendix:
M002N2874DesignReportonComputerIntegrity,TestandCalibration,andFaultDetection LARSection:4.8.2RevNo:1EncB#1.19SystemResponseTimeAnalysisReport(D.9.4.2.4)
ISG06Section:D.9.4.2.4Phase1(GEH)DocumentLAREnclosure3(NEDC33864P)Appendix:
N001N8578PRNMSystemResponseTimeAnalysisReportLARSection:4.1.3RevNo:2EncB#1.20TheoryofOperationDescription(D.9.4.2.8,D.9.4.2.9,D.9.4.2.10,D.9.4.2.11,D.9.4.2.13,D.9.4.2.14,D.9.4.3.2,D.9.4.3.5,D.9.4.3.6,D.9.4.3.7,D.9.4.4)ISG06Section:D.9.4.2.8,D.9.4.2.9,D.9.4.2.10,D.9.4.2.11,D.9.4.2.13,D.9.4.2.14,D.9.4.3.2,D.9.4.3.5,D.9.4.3.6,D.9.4.3.7,D.9.4.4Phase1(GEH)DocumentLAREnclosure3(NEDC33864P)Appendix:
O001N5984ReportonCompliancewithIEEEStandards(6031991and74.3.22003)andTheoryofOperationsDescription LARSection:4.8.1RevNo:1EncB#1.21.1SetpointMethodology(D.9.4.3.8,D.11)ISG06Section:D.9.4.3.8,D.11Phase1(GEH)DocumentLAREnclosure3(NEDC33864P)Appendix:
P001N8046GEHInstrumentSetpointMethodologyOverview,HCGSPRNMLARSection:4.8.4RevNo:1Wednesday,September9,20Page6of9ISG06EnclosureBtoPhase1DocumentMapping EncB#1.21.2SetpointMethodology/Results(D.9.4.3.8,D.11)ISG06Section:D.9.4.3.8,D.11Phase1(GEH)DocumentLAREnclosure3(NEDC33864P)Appendix:
P1002N6483InstrumentLimitsCalculationHCGSNUMACPRNMSystemAPRMLARSection:4.8.4RevNo:0EncB#1.21.3SetpointMethodology/Results(D.9.4.3.8,D.11)ISG06Section:D.9.4.3.8,D.11Phase1(GEH)DocumentLAREnclosure3(NEDC33864P)Appendix:
P2002N7071InstrumentLimitsCalculationHCGSNUMACPRNMSystemRBMLARSection:4.8.4RevNo:0EncB#1.23SoftwareToolVerificationProgram(D.10.4.2.3.2)
ISG06Section:D.10.4.2.3.2Phase1(GEH)DocumentLAREnclosure3(NEDC33864P)Appendix:
O001N5984ReportonCompliancewithIEEEStandards(6031991and74.3.22003)andTheoryofOperationsDescription LARSection:4.8.1RevNo:1EncB#1.24SoftwareProjectRiskManagementProgram(D.10.4.2.3.6)
ISG06Section:D.10.4.2.3.6Phase1(GEH)DocumentLAREnclosure3(NEDC33864P)Appendix:
O001N5984ReportonCompliancewithIEEEStandards(6031991and74.3.22003)andTheoryofOperationsDescription LARSection:4.8.1RevNo:1Wednesday,September9,20Page7of9ISG06EnclosureBtoPhase1DocumentMapping EncB#1.25CommercialGradeDedicationPlan(D.10.4.2.4.2)
ISG06Section:D.10.4.2.4.2Phase1(GEH)DocumentLAREnclosure3(NEDC33864P)Appendix:
O001N5984ReportonCompliancewithIEEEStandards(6031991and74.3.22003)andTheoryofOperationsDescription LARSection:4.8.1RevNo:1EncB#1.26VulnerabilityAssessment(D.12.4.1)
ISG06Section:D.12.4.1Phase1(GEH)DocumentLAREnclosure3(NEDC33864P)Appendix:
Q001N7872SecureDevelopmentandOperationalEnvironmentandVulnerabilityAssessmentReportLARSection:4.10RevNo:2EncB#1.27SecureDevelopmentandOperationalEnvironmentControls(D.12.2)ISG06Section:D.12.2Phase1(GEH)DocumentLAREnclosure3(NEDC33864P)Appendix:
Q001N7872SecureDevelopmentandOperationalEnvironmentandVulnerabilityAssessmentReportLARSection:4.10RevNo:2EncB#NAPRNMPlantSpecificResponse(LTR)ISG06Section:Phase1(GEH)DocumentLAREnclosure3(NEDC33864P)Appendix:
R001N8420HCGSPlantSpecificResponsesRequiredbyPRNMLTRLARSection:4.1.2.1RevNo:1Wednesday,September9,20Page8of9ISG06EnclosureBtoPhase1DocumentMapping EncB#NAARTSJustification ISG06Section:Phase1(GEH)DocumentLAREnclosure3(NEDC33864P)Appendix:
S001N8296SupplementalInformationforARTSforHCGSLARSection:4.1.2.3RevNo:3EncB#NADSSCDEvaluation ISG06Section:Phase1(GEH)DocumentLAREnclosure3(NEDC33864P)Appendix:
T000N3922HCGSThermalHydraulicStability,DSSCDEvaluation LARSection:4.1.2.2RevNo:1Wednesday,September9,20Page9of9ISG06EnclosureBtoPhase1DocumentMapping HCGSPRNMUgradeISG06EnclosureBRoadmapPhase1DocumenttoISG06EnclosureBMapping(Tier2Submittal) 001N2029PRNMSystemArchitectureDescription RevNo:3LAREnclosure3(NEDC33864P)Appendix:
AISG06EncB/SubjectLARSection:4.1.1,4.1.21.1HardwareArchitectureDescriptions(D.1.2)EncB#Section:D.1.21.3SoftwareArchitectureDescriptions(D.3.2,D.4.4.3.2)
EncB#Section:D.3.2,D.4.4.3.21.16.1DesignAnalysisReports(D.7.2)Communications EncB#Section:D.7.21.17SystemDescription(Toblockdiagramlevel)(D.9.2,D.10.2)EncB#Section:D.9.2,D.10.2NEDE33834PNUMACSystemsEngineeringDevelopmentPlanRevNo:0LAREnclosure3(NEDC33864P)Appendix:
BISG06EncB/SubjectLARSection:4.2.21.10SoftwareConfigurationManagementPlan(D.4.4.1.11)
EncB#Section:D.4.4.1.111.5SoftwareDevelopmentPlan(D.4.4.1.2)
EncB#Section:D.4.4.1.21.7SoftwareIntegrationPlan(D.4.4.1.4)
EncB#Section:D.4.4.1.4NEDE33836PNUMACSystemsQualityAssurancePlanRevNo:0LAREnclosure3(NEDC33864P)Appendix:
CISG06EncB/SubjectLARSection:4.2.31.6SoftwareQAPlan(D.4.4.1.3,D.10.4.2.3.1)
EncB#Section:D.4.4.1.3,D.10.4.2.3.1Wednesday,September9,201Page1of6Phase1DocumenttoISG06EnclosureBMapping NEDE33835PNUMACSystemsIndependentVerification&ValidationPlanRevNo:0LAREnclosure3(NEDC33864P)Appendix:
DISG06EncB/SubjectLARSection:4.2.41.9SoftwareV&VPlan(D.4.4.1.10)
EncB#Section:D.4.4.1.101.11SoftwareTestPlan(D.4.4.1.12)
EncB#Section:D.4.4.1.121.8SoftwareSafetyPlan(D.4.4.1.9)
EncB#Section:D.4.4.1.9 002N4398HopeCreekNUMACPRNMSystemManagementPlanRevNo:1LAREnclosure3(NEDC33864P)Appendix:
EISG06EncB/SubjectLARSection:4.2.51.4SoftwareManagementPlan(D.4.4.1.1)
EncB#Section:D.4.4.1.126A8742SANUMACPRNMSystemRequirementsSpecification RevNo:3LAREnclosure3(NEDC33864P)Appendix:
F1ISG06EncB/SubjectLARSection:4.3.11.12.1SoftwareRequirementsSpecification(D.4.4.3.1)
EncB#Section:D.4.4.3.1 000N6426NUMACAPRMDSSCDPerformanceSpecification RevNo:3LAREnclosure3(NEDC33864P)Appendix:
F2ISG06EncB/SubjectLARSection:4.3.11.12.2SoftwareRequirementsSpecification(D.4.4.3.1)
EncB#Section:D.4.4.3.1Wednesday,September9,201Page2of6Phase1DocumenttoISG06EnclosureBMapping 002N2038APRMFunctionalControllerSDSRevNo:1LAREnclosure3(NEDC33864P)Appendix:
GISG06EncB/SubjectLARSection:4.3.21.13SoftwareDesignSpecification(D.4.4.3.3)
EncB#Section:D.4.4.3.3 001N6665NUMACQualificationProgramHCGSRevNo:2LAREnclosure3(NEDC33864P)Appendix:
HISG06EncB/SubjectLARSection:4.41.14EquipmentQualificationTestingPlans(IncludingEMI,Temperature,Humidity,andSeismic)(D.5.2)EncB#Section:D.5.2003N0063DiversityandDefenseinDepthAnalysisRevNo:0LAREnclosure3(NEDC33864P)Appendix:
IISG06EncB/SubjectLARSection:4.51.15D3Analysis(D.6.2)EncB#Section:D.6.2001N5783DesignAnalysisReport:HCNUMACPRNMSystem,Hardware,andSoftwareModifications RevNo:2LAREnclosure3(NEDC33864P)Appendix:
JISG06EncB/SubjectLARSection:4.71.16.2DesignAnalysisReports(D.8.2)EncB#Section:D.8.2001N8626DesignAnalysisReport:MethodologyModifications RevNo:2LAREnclosure3(NEDC33864P)Appendix:
KISG06EncB/SubjectLARSection:4.2.11.16.3DesignAnalysisReports(D.8.2)EncB#Section:D.8.2Wednesday,September9,201Page3of6Phase1DocumenttoISG06EnclosureBMapping 001N7851DesignAnalysisReportonElectricalIndependence RevNo:2LAREnclosure3(NEDC33864P)Appendix:
LISG06EncB/SubjectLARSection:4.8.31.16.4DesignAnalysisReports(D.9.4.2.6,D.10.4.2.6)
EncB#Section:D.9.4.2.6,D.10.4.2.6 002N2874DesignReportonComputerIntegrity,TestandCalibration,andFaultDetection RevNo:1LAREnclosure3(NEDC33864P)Appendix:
MISG06EncB/SubjectLARSection:4.8.21.18DesignReportonComputerintegrity,TestandCalibration,andFaultDetection(D.9.4.2.5,D.9.4.2.7,D.9.4.2.10,D.9.4.3.5,D.10.4.2.5,D.10.4.2.7)
EncB#Section:D.9.4.2.5,D.9.4.2.7,D.9.4.2.10,D.9.4.3.5,D.10.4.2.5,D.10.4.2.7 001N8578PRNMSystemResponseTimeAnalysisReportRevNo:2LAREnclosure3(NEDC33864P)Appendix:
NISG06EncB/SubjectLARSection:4.1.31.19SystemResponseTimeAnalysisReport(D.9.4.2.4)
EncB#Section:D.9.4.2.4 001N5984ReportonCompliancewithIEEEStandards(6031991and74.3.22003)andTheoryofOperationsDescription RevNo:1LAREnclosure3(NEDC33864P)Appendix:
OISG06EncB/SubjectLARSection:4.8.11.23SoftwareToolVerificationProgram(D.10.4.2.3.2)
EncB#Section:D.10.4.2.3.21.24SoftwareProjectRiskManagementProgram(D.10.4.2.3.6)
EncB#Section:D.10.4.2.3.61.25CommercialGradeDedicationPlan(D.10.4.2.4.2)
EncB#Section:D.10.4.2.4.2Wednesday,September9,201Page4of6Phase1DocumenttoISG06EnclosureBMapping 1.20TheoryofOperationDescription(D.9.4.2.8,D.9.4.2.9,D.9.4.2.10,D.9.4.2.11,D.9.4.2.13,D.9.4.2.14,D.9.4.3.2,D.9.4.3.5,D.9.4.3.6,D.9.4.3.7,D.9.4.4)EncB#Section:D.9.4.2.8,D.9.4.2.9,D.9.4.2.10,D.9.4.2.11,D.9.4.2.13,D.9.4.2.14,D.9.4.3.2,D.9.4.3.5,D.9.4.3.6,D.9.4.3.7,D.9.4.4001N8046GEHInstrumentSetpointMethodologyOverview,HCGSPRNMRevNo:1LAREnclosure3(NEDC33864P)Appendix:
PISG06EncB/SubjectLARSection:4.8.41.21.1SetpointMethodology(D.9.4.3.8,D.11)EncB#Section:D.9.4.3.8,D.11002N6483InstrumentLimitsCalculationHCGSNUMACPRNMSystemAPRMRevNo:0LAREnclosure3(NEDC33864P)Appendix:
P1ISG06EncB/SubjectLARSection:4.8.41.21.2SetpointMethodology/Results(D.9.4.3.8,D.11)EncB#Section:D.9.4.3.8,D.11002N7071InstrumentLimitsCalculationHCGSNUMACPRNMSystemRBMRevNo:0LAREnclosure3(NEDC33864P)Appendix:
P2ISG06EncB/SubjectLARSection:4.8.41.21.3SetpointMethodology/Results(D.9.4.3.8,D.11)EncB#Section:D.9.4.3.8,D.11001N7872SecureDevelopmentandOperationalEnvironmentandVulnerabilityAssessmentReportRevNo:2LAREnclosure3(NEDC33864P)Appendix:
QISG06EncB/SubjectLARSection:4.101.27SecureDevelopmentandOperationalEnvironmentControls(D.12.2)EncB#Section:D.12.21.26VulnerabilityAssessment(D.12.4.1)
EncB#Section:
D.12.4.1Wednesday,September9,201Page5of6Phase1DocumenttoISG06EnclosureBMapping 001N8420HCGSPlantSpecificResponsesRequiredbyPRNMLTRRevNo:1LAREnclosure3(NEDC33864P)Appendix:
RISG06EncB/SubjectLARSection:4.1.2.1NAPRNMPlantSpecificResponse(LTR)EncB#Section:
001N8296SupplementalInformationforARTSforHCGSRevNo:3LAREnclosure3(NEDC33864P)Appendix:
SISG06EncB/SubjectLARSection:4.1.2.3NAARTSJustification EncB#Section:
000N3922HCGSThermalHydraulicStability,DSSCDEvaluation RevNo:1LAREnclosure3(NEDC33864P)Appendix:
TISG06EncB/SubjectLARSection:4.1.2.2NADSSCDEvaluation EncB#Section:Wednesday,September9,201Page6of6Phase1DocumenttoISG06EnclosureBMapping    LAR H15-01 LR-N15-0178 NEDO-33864, Hope Creek Generati ng Station NUMAC PRNM Upgrade - Non-Proprietary
 
Contains Proprietary Information to be Withheld from Public Disclosure Pursuant to 10 CFR 2.390 Enclosure 3  LAR H15-01 LR-N15-0178 NEDC-33864P, Hope Creek Generati ng Station NUMAC PRNM Upgrade -
Proprietary
 
This Enclosure contains proprietary information of GE-Hitachi Nuclear Energy (GEH).}}

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