ML13364A286

Maximum Extended Load Mine Limit Analysis Plus (Mella+) License Amendment Request-Responses to Requests for Supplemental Information

ML13364A286
Person / Time
Site:	Grand Gulf
Issue date:	12/30/2013
From:	Kevin Mulligan Entergy Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
Shared Package
ML13364A284	List: ML13364A286
References
GNRO-2013-/00100
Download: ML13364A286 (21)

Text

• Entergy

. -.p Entergy Operations, Inc.

P. O. Box 756 Port Gibson. MS 39150 Kevin J. Mulligan Site Vice President TeL (601) 437-7500 Attachment 1 contains PROPRIETARY information.

Withhold per 10 CFR 2.390.

GNRO-2013/00100 December 30,2013 u.s. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001

SUBJECT:

Maximum Extended Load Line Limit Analysis Plus (MELLLA+)

License Amendment Request -

Responses to Requests for Supplemental Information Grand Gulf Nuclear Station, Unit 1 1 Docket No. 50-416 License No. NPF-29

REFERENCES:

1. Entergy Operations, Inc. letter to the NRC, Maximum Extended Load Line Limit Analysis Plus (MELLLA +) License Amendment Request, dated September 25, 2013 (ML13269A140)

2. NRC letter to Entergy Operations, Inc., Grand Gulf Nuclear Station, Unit 1 - Supplemental Information Needed for Acceptance of Licensing Action, Request to Allow Operation in Expanded Maximum Extended Load Line Limit Analysis Plus Domain, dated December 19, 2013 (ML13345A182)

Dear Sir or Madam:

In Reference 1, Entergy Operations, Inc. (Entergy) submitted to the U.S. Nuclear Regulatory Commission (NRC) a license amendment request (LAR) that would allow Grand Gulf Nuclear Station, Unit 1 (GGNS) to operate in the expanded Maximum Extended Load Line Limit Analysis Plus (MELLLA+) domain. In Reference 2, the NRC requested supplemental information to support their acceptance review of the MELLLA+ LAR. Attachment 1 provides responses to these requests for supplemental information (RSls).

General Electric - Hitachi (GEH) considers certain information contained in Attachment 1 to be proprietary and, therefore, exempt from public disclosure pursuant to Title 10 Code of Federal Regulations (10 CFR) 2.390. The associated affidavit for withholding information, executed by GEH, is provided in Attachment 2. The responses to the RSls were provided by GEH to Entergy in a transmittal letter that is referenced in the affidavit. Therefore, on behalf of GEH, Entergy requests Attachment 1 be withheld from public disclosure in accordance with 10 CFR 2.390(b)(1). A non-proprietary version of Attachment 1 is provided in Attachment 3.

When Attachment 1 is removed from this letter, the entire document is NON-PROPRIETARY.

GNRO-2013/00100 Page 2 of 2 Attachment 1 contains PROPRIETARY information.

Withhold per 10 CFR 2.390.

This letter contains no new commitments.

If you have any questions or require additional information, please contact Mr. Jeff Seiter at (601) 437-2344.

I declare under penalty of perjury that the foregoing is true and correct; executed on December 30, 2013.

Sincerely, ~

C;'k<<.A KJM/slw

~ dO~ /(. J I1,J/f1V7 0

Attachments: 1. Responses to NRC Requests for Supplemental Information (Proprietary Version)

2. General Electric - Hitachi Affidavit Supporting Request to Withhold Information from Public Disclosure

3. Responses to NRC Requests for Supplemental Information (Non-Proprietary Version) cc: U.S. Nuclear Regulatory Commission ATTN: Mr. Mark Dapas, (w/2)

Regional Administrator, Region IV 1600 East Lamar Boulevard Arlington, TX 76011-4511 U. S. Nuclear Regulatory Commission ATTN: Mr. Alan Wang, NRRlDORL (w/2)

Mail Stop OWFN 8 B1 Washington, DC 20555-0001 NRC Senior Resident Inspector Grand Gulf Nuclear Station Port Gibson, MS 39150 When Attachment 1 is removed from this letter, the entire document is NON-PROPRIETARY.

Attachment 2 to GNRO-2013/00100 General Electric - Hitachi Affidavit General idavit Supporting Request Disclosure to Withhold Information from Public Disclosure

GE-Hitachi Nuclear Energy Americas LLC AFFIDAVIT I, Linda C. Dolan, state as follows:

(1) I am the Manager of Regulatory Compliance, ofGE-Hilachi Nuclear Energy Americas LLC

("GEH"), and have been delegated the function of reviewing the information described in paragraph (2) which is sought to be withheld, and have been authorized to apply for its withholding.

(2) The information sought to be withheld is contained in Enclosure I of GEH letter, GEH-GGNS-AEP-634, "GEH Respon es to MELLLA Plus EICB Requests for Supplemental Information," dated December 19, 2013. The GEH proprietary information in Enclosure I, which is entitled "Responses to EICB Requests for Supplemental Information in Support of GGNS MFLLLA+ LAR," is identified by a dotted underline inside double square brackets.

((Ib.i~...~~n!~!1~~jL~-'L.~)5..~m.pJ.; ... ~~.'.l] In each case, the superscript notation :3: refers to Paragraph (3) of this affidavit, which provides the basis for the proprietary determination.

(3) In making this application for withholding of proprietary information of which it is the owner or licensee, GEH relies upon the exemption from disclosure set forth in the Freedom ofJriformation Act ("FOIA"), 5 U.S.c. Sec. 552(b)(4), and the Trade Secrets Act, 18 U.S.c.

Sec. 1905, and NRC regulations 10 CFR 9.17(a)(4), and 2.390(a)(4) for trade secrets (Exemption 4). The material for which exemption from disclosure is here sought also qualifies under the narrower definition of trade secret, within the meanings assigned to those terms for purposes of FOIA Exemption 4 in, respectively, Critical Mass Energy Project v. Nuclear Regulatory Commission, 975 F.2d 871 (D.C. Cir. 1992), and Public Citizen Health Research Group v. FDA, 704 F.2d 1280 (D.C. Cir. 1983).

(4) The information sought to be withheld is considered to be proprietary for the reasons set forth in paragraphs (4)a. and (4)b. Some examples of categories of information that fit into the definition ofproprielary information are:

a. Information that discloses a process, method, or apparatus, including supporting data and analyses, where prevention of its use by GEH's competitors without license from GEH constitutes a competitive economic advantage over other companies;

b. Information that, if used by a competitor, would reduce their expenditure of resources or improve their competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing of a similar product;

c. Information that reveals aspects of past, present, or future GEH customer-funded development plans and programs, resulting in potential products to GEH;

d. Information that discloses trade secret or potentially patentable subject matter for which it may be desirable to obtain patent protection.

(5) To address 10 CFR 2.390(b)(4), the information sought to be withheld is being submitted to NRC in confidence. The information is of a sOl1 customarily held in confidence by GEH, Affidavit for GEH-GGNS-AEP-634 Page 1 of3

GE-Hitachi Nuclear Energy Americas LLC and is in fact so held. The information sought to be withheld has, to the best of my knowledge and belief, consistently been held in confidence by GEH, not been disclosed publicly, and not been made available in public sources. All disclosures to third parties, including any required transmittals to the NRC, have been made, or must be made, pursuant to regulatory provisions or proprietary or confidentiality agreements that provide for maintaining the information in confidence. The initial designation of this information as proprietary information, and the subsequent steps taken to prevent its unauthorized disclosure, are as set forth in the following paragraphs (6) and (7).

(6) Initial approval of proprietary treatment of a document is made by the manager of the originating component, who is the person most likely to be acquainted with the value and sensitivity of the information in relation to industry knowledge, or who is the person most likely to be subject to the terms under which it was licensed to GEH. Access to such documents within GEH is limited to a "need to know" basis.

(7) The procedure for approval of external release of such a document typically requires review by the staff manager, project manager, principal scientist, or other equivalent authority for technical content, competitive effect, and determination of the accuracy of the proprietary designation. Disclosures outside GEH are limited to regulatory bod ies, customers, and potential customers, and their agents, suppliers, and licensees, and others with a legitimate need for the information, and then only in accordance with appropriate regulatory provisions or proprietary or confidentiality agreements.

(8) The information identified in paragraph (2) above, is classified as proprietary because it contains the detailed GEH methodology for stability analysis for the GEH Boiling Water Reactor (BWR). These methods, techniques, and data along with their application to the design, modification, and analyses associated with the DSS-CD were achieved at a significant cost to GEH.

The development of the evaluation processes along with the interpretation and application of the analytical resu Its is derived from the extensive experience databases that constitute a major GEH asset.

(9) Public disclosure of the information sought to be withheld is likely to cause substantial harm to GEH's competitive position and foreclose or reduce the availability of profit-making opportunities. The information is part of GEH's comprehensive BWR safety and technology base, and its commercial value extends beyond the original development cost.

The value of the technology base goes beyond the extensive physical database and analytical methodology and includes development of the expertise to determine and apply the appropriate evaluation process. In addition, the technology base includes the value derived from providing analyses done with NRC-approved methods.

The research, development, engineering, analytical and NRC review costs comprise a substantial investment of time and money by GEH. The precise value of the expertise to devise an evaluation process and apply the correct analytical methodology is difficult to quantify, but it clearly is substantial. GEH's competitive advantage will be lost if its Affidavit for GEH-GGNS-AEP-634 Page 2 of3

GE-Hitachi Nuclear Energy Americas LLC competitors are able to use the results of the GEH experience to normalize or verify their own process or if they are able to claim an equivalent understanding by demonstrating that they can arrive at the same or similar conclusions.

The value of this information to GEH would be lost if the information were disclosed to the public. Making such information available to competitors without their having been required to undertake a similar expenditure of resources would unfairly provide competitors with a windfall, and deprive GEH of the opportunity to exercise its competitive advantage to seek an adequate return on its large investment in developing and obtaining these very valuable analytical tools.

I declare under penalty of perjury that the foregoing affidavit and the matters stated therein are true and correct to the best of my knowledge, information, and belief.

Executed on this 19th day of December 2013.

Linda C. Dolan Manager, Regulatory Compliance GE-Hitachi Nuclear Energy Americas LLC 390 I Castle Hayne Rd.

Wilmington, NC 2840]

Linda.Dolan@ge.com Affidavit for GEH-GG S-AEP-634 Page 3 of3

Attachment Attachment 33 to to GNRO-2013/00100 GNRO-2013/00100 Responses Responses to to NRC NRC Requests Requests for for Supplemental Supplemental Information Information (Non-Proprietary (Non-Proprietary Version)

Version)

This is a non-proprietary version of 111"'r~:l""nlrnOlt"'a'" Attachment 11from from which proprietary information which the "",11\+"\11' information has been been removed. The ,.u*t"\,"'U'u~t~I"'\1 proprietary portions n\ ...... I1'\n~ that that have have been been removed removed are are indicated indicated by double square square brackets brackets as shown here:

here: (((( ]1 to GNRO-2013/00100 Page 1 of 14 Non-Proprietary Information - Class I (Public)

MELLLA+ LICENSE AMENDMENT ENDMENT REQUEST RESPONSES TO NRC C REQUESTS FOR SU SUPPLEMENTAL INFORMATION In a letter to the u.s.

U.S. Nuclear Regulatory Commission (NRC) dated September 25, 201 2013, Entergy Operations, Inc. (Entergy) submitted to the NRC a license amendment request (LAR) that would allow Grand Gulf Nuclear Station, Unit 1 (GGNS) to operate in the expanded Maximum Extended Load Line Limit Analysis Plus (MELLLA+) domain. In a letter to Entergy dated December 1 19, 201 2013, the NRC requested supplemental information to support their acceptance review of the MELLLA+ LAR. Responses to these requests for supplemental information (RSls) are provided below.

RSI #1

""",,,.n"'J'' the Page xiii of the September 17, 2007, MELLLA+ topical report (TR) states: "Therefore, the NRC staff concluded that manual backup stability protection is not appropriate and a a NRC-approved automatic backup stability protection must be implemented forrMELLLA+

MELLLA+

operation." Please provide the required supplemental information on the automatic uu ..'v v backup uuvn,uw stability protection.

Response

The NRC approved automatic backup stability protection is implemented for GGNS MELLLA+

Safety Analysis Report (SAR) application (Reference 1.1) as documented in Section 2.4.3 in Reference 1.1. The approval of this automatic backup stability protection is documented in the NRC Safety Evaluation (SE) included in Reference 1.2. All the information about this Automatic Backup Stability Protection (ABSP) is provided in Section 7 of Reference 1.2, which is the approved Detect and Suppress Solution - Confirmation Density (DSS-CD) licensing topical report (LTR) revision applicable to GGNS MELLLA+ SAR.

In the October 2008 MELLLA+ LTR NEDC-33006P-A, Revision 3 (page xiii), NRC stated:

"Therefore, the NRC staff concluded that manual backup stability protection is not appropriate and a NRC-approved automatic backup stability protection must be implemented for MELLLA+ operation." To meet this requirement, GEH has provided in the approved DSS-CD LTR (NEDC-33075P-A, Revision 6, Reference 1.3) a Backup Stability Protection (SSP) (BSP) approach that includes an automatic function and that may be used when the Oscillation Power Range Monitor (OPRM) system is inoperable up to and including operation in the MELLLA+ domain. The BSP SSP solution and the ABSP ASSP function are maintained and approved in the DSS-CD LTR NEDC-33075P-A, Revision 8, that is applied to the GGNS MELLLA+ SAR 1 1).

application (Reference 1.1).

This comprehensive BSP approach provides an alternative means for stability protection by preventing the onset of growing power oscillations in the specific region of the power/flow map identified as likely to develop thermal-hydraulic instabilities. In the June 2011 DSS-CD LTR (NEDC-33075P-A, Revision 6 (page 5), Reference 1.3), 1 the NRC stated that "The asp concept, documented in Section 7 of NEDC-33075P, Revision 5, is a technically acceptable solution to the backup issue.

issue."

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This comprehensive BSP is described in Section 7 and implemented per the Technical Specification (TS) changes documented in Appendix A of the approved DSS-CD LTR (NEDC-33075P-A, Revision 6, Reference 1.3) and includes two BSP options that are based on selected elements from three distinct constituents. The three constituents are:

4. BSP Manual Regions that comprise plant-specific scram (Region I) and Controlled Entry (Region II) regions in the licensed power/flow operating domain and associated manual operator actions (Section 7.2 of Reference 1.3).

5. BSP Boundary that defines the operating domain portion where potential instability events can be effectively addressed by specific operator actions (Section 7.3 of Reference 1 1.3).

6. ABSP Scram Region, which comprises an automatic reactor scram region initiated by the Average Power Range Monitor (APRM) flow-biased scram setpoint (Section 7.4 of Reference 1.3).

The two BSP options are:

Option 1: Consists of the BSP BSP Manual Manual Regions, BSP Boundary and associated L.lr\I**. . r\t\l

operator actions.

Option 2: Consists of the ABSP Scram Region, as implemented by the APRM flow-biased scram setpoint, Region II and ""'_"""'_"_"""_

"'1""""""-'1 associated operator actions.

The TS changes contained in Reference 1.3 and in the GGNS MELLLA+ LAR delineate specific implementation requirements for both BSP options in the unlikely event the OPRM system is declared inoperable. In such instance, the operators have 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> to manually implement the ASBP function. In the interim, instability protection is implemented via the Manual BSP regions, which are administratively implemented. With the ABSP (Option 2), a scram is automatically generated if the reactor enters a pre-determined scram region regardless of whether or not a thermal-hydraulic instability occurs.

In case the ABSP function cannot be implemented or is inoperable, the licensed stability solution becomes the Manual BSP region with the BSP Boundary, which is manually implemented through administrative actions. ((

))

Therefore, NRC concluded in the DSS-CD LTR NEDC-33075P-A, Revision 6 (page 6), in Reference 1.3 1 "that the proposed asp methodology is an acceptable solution, because it provides sufficient protection against Safety Limit Minimum Critical Power Ratio (SLMCPR) violations commensurate with the probability of an instability event in the short period of time that they are active."active.

All these elements and functions of the BSP SSP solution, including the ABSP function, are maintained and approved in the DSS-CD LTR NEDC-33075P-A, Revision 8, and SE of to GNRO-2013/00100 Page 3 of 14 Non-Proprietary Information - Class I (Public)

NEDC-33075P, Revision 7, (Reference 1.2) that is applied to the GGNS MELLLA+ SAR application (Reference 1.1).

((

))

References:

1 1.11 GE Hitachi Nuclear Energy, "Safety Analysis Report for Grand Gulf Nuclear Station -

Maximum Extended Load Line Limit Analysis Plus," NEDC-3361 NEDC-33612P, September 2013 (Attachment 4 to ML13269A140).

1 1.2 GE Hitachi Hitachi Nuclear _~c',rc", "GE Hitachi Boiling Water Reactor Detect and Suppression Nuclear Energy, Solution - Confirmation Density," NEDC-33075P-A, Revision 8, November 2013 (ML11161 (ML111610593).

11.3 GE Hitachi Nuclear Energy, "General Electric Boiling Water Reactor Detect and Suppression Solution - Confirmation Density," NEDC-33075P-A, Revision 6, January 2008.

RSI #2

..... 4"",..,..*'r\,.. in Please provide hardware/software technical information as r\1"Y'l1"'\

emphasized in the reference document General Electric - Hitachi (GEH) NEDC-33075P, pages 1 1-2. This information should specifically document the hardware and software and/or firmware designs as per any variations from MELLLA to the MELLLA+ algorithm.

Response

As noted in page 1-1 of the approved licensing topic report (LTR) for Detect and Suppress Solution - Confirmation Density (DSS-CD), NEDC-33075P-A (Reference 2.1):

'The DSS-CD is based on the same hardware design as Option III, which is described in References 1 through 3. 3." 4 Also, in page 2-2:

4 The references 1 through 3 cited are from NEDC-33075P-A.

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'The DSS-CD solution introduces a number of changes relative to the Option III solution. In addition, it introduces a number of modifications and restrictions to the successive confirmation period element of the Period Based Detection Algorithm (PBDA) to improve its ability for early recognition of reactor oscillations. These changes only affect the system software/firmware, and therefore, may be able to be implemented on-line."

The NRC staff previously reviewed and approved the implementation of DSS-CD using the approved approved GEH GEH Option III hardware Option III hardware and and software.

software. The The DSS-CD DSS-CD solution solution is is not not generically generically approved for use with non-GEH hardware. The hardware components required to implement DSS-CD are DSS-CD are those those currently used for currently used for the the approved Option III.

approved Option III. IfIf the the DSS-CD DSS-CD hardware hardware implementation implementation deviates deviates from from the Option III approved Option the approved solution, a III solution, a hardware hardware review review byby the the NRC NRC staff will be required. Implementations on other Option III platforms will require plant-specific reviews. The DSS-CD stability solution (CDA, ABSP, ABSP,5 and Option III) for GGNS is implemented in the GGNS Power Range Neutron Monitor implemented in the GGNS Power Range Neutron Monitor (PRNM)

(PRNM) system.

Limitation Limitation and Condition 5.1 in the approved approved DSS-CD DSS-CD LTR LTR NEDC-33075P-A NEDC-33075P-A (Reference (Reference 2.1) 1) states that:

UThe NRC "The NRC staff staff previously previously reviewed reviewed and and approved approved the implementation of DSS-CD using the III approved GEH Option 11/ hardware and software. The DSS-CD S-CD solution is not approved for

,,",,'-'I\..AU,-,I use use with non-GEH hardware. The hardware components required to with non-GEH hardware. The hardware components required to implement implement DSS-CD DSS-CD areare expected to be those currently used for the approved Option III. 11/. If the DSS-CD hardware implementation ae"'Jalf~S from Irnr,'orno""-J1'j,..." deviates the approved Option from the Option III solution, a III solution, a hardware review by by the the NRC NRC staff will be required. Implementations Implementations on on other other Option Option III11/ platforms platforms will will require require plant-specific plant-specific reviews."

The GGNS MELLLA+ safety analysis report (SAR), NEDC-33612P (Reference 2.2) documents in Appendix C compliance with this Limitation and Condition because the DSS-CD solution for GGNS is implemented on GEH hardware that is currently installed and approved by the NRC for the Option III solution.

Therefore, the technical information provided for the GGNS PRNM (Reference 2.3) is applicable to the DSS-CD design. Specifically, the PRNM LTR (References 2.4 and 2.5) provided technical information of the approved PRNM system. Reference 2.6 identified the plant-specific differences from the PRNM LTR. Reference 2.7 (RAI 5) provided additional explanation of the PRNM configuration for the GGNS PRNM PRNM system. The implementation C"\.IC"ll"£:lrTl strategy for the GGNS PRNM is to provide the system with all the required functionalities except that the DSS-CD trip will be inactive (using a jumper) until the implementation of the MELLLA+. Additional information about this approach is provided in Reference I-101'~~ronjr"o 2.8 (RAI 4).

2.8 (RAI 4).

The same software development process was used for the PRNM, Option III, DSS-CD (CDA and ABSP). The NRC approval of the GEH software development process is noted in the SER (Reference 2.3).

5 The stability protection for MELLLA+ is DSS-CD which includes Confirmation Density Algorithm (GOA)

(CDA) and and Automatic Automatic Backup Stability Stability Solution (ABSP).

(ABSP). Option Option IIIIII is is maintained maintained for for defense-in-depth defense-in-depth purpose. Unless otherwise specified, the term DSS-CD in this request for supplement information purpose. Unless otherwise the term DSS-GD in this for supplement information (RSI) response response would mean both would mean both GOA CDA andand ABSP.

ABSP.

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(Public)

\11",I~."_t""lnnr'I"11" ,....++"... "I'"\I"I"'\l"" evaluated in Uni'nl"'nnl"l"'\

The plant-specific differences evaluated in Reference 2.6 were due to the plant-specific configuration and configuration and other other prior prior licensing licensing commitments.

commitments. There was no deviation from the MELLLA to the MELLLA to the MELLLA+ algorithm. MELLLA+ algorithm.

References:

2.11 GE GE Hitachi Hitachi Nuclear Nuclear Energy, Energy, "GE Hitachi Hitachi Boiling Boiling Water Water Reactor Reactor Detect Detect and and Suppression Suppression Solution - Confirmation Density, NEDC-33075P-A, Solution - Confirmation Density," NEDC-33075P-A, Revision 8, November 2013 Revision 8, November 2013 (ADAMS (ADAMS Accession Accession No, ML111610593).No, ML111610593).

GE Hitachi 2.2 GE Hitachi Nuclear Nuclear Energy, Energy, "Safety "Safety Analysis Analysis Report Report for for Grand Grand Gulf Gulf Nuclear Nuclear Station Station Maximum Extended Load Line Limit Analysis Maximum Extended Load Line Limit Analysis Plus," NEDC-33612P, September 2013. Plus, NEDC-33612P, September 2013.

A.Wang (NRC) 2.3 A.Wang (NRC) to to VP, VP, Operations Operations (Entergy (Entergy Operation, Operation, Inc), Inc), "Grand "Grand Gulf Gulf Nuclear Nuclear Station, Station, Unit 1 of Amendment Power Unit 1 -Issuance of Amendment RE: Power Range Neutron Monitoring System Range Neutron Monitoring System Replacement (TAC Replacement (TAC No.No. ME2531),"

ME2531)," March March 28,2012 28,2012 (ADAMS (ADAMS Accession Accession No. No.

ML120400319).

ML120400319).

2.4 GE NuclearNuclear Energy, Energy, "Nuclear "Nuclear Measurement Measurement Analysis Analysis and and Control Control Power Power Range Range Neutron Neutron Monitor (NUMAC PRNM) Retrofit Plus Option Monitor (NUMAC PRNM) Retrofit Plus Option III Stability Trip Function," NEDC-32410P-A, III Stability Trip Function," NEDC-32410P-A, October 1995.

October 1 NEDC-32410P-A, "Nuclear 2.5 NEDC-32410P-A, "Nuclear Measurement Measurement Analysis Analysis and and Control Control PowerPower Range Range Neutron Neutron Monitor (NUMAC PRNM) Retrofit Plus Option Monitor (NUMAC PRNM) Retrofit Plus Option III Stability Trip Function," Supplement 1, III Stability Trip Function," Supplement 1, 1\.11".ur'Vih ... 1 November 1997.

Gulf Nuclear _1"'1." **

2.6 GE Hitachi Nuclear Energy Report, "Grand Gulf Nuclear Station - Plant-Specific by Responses Required by NUMAC PRNM Retrofit Plus NUMAC PRN Plus Option Option III III Stability Trip Function

....,""'-4I...III1"Y (NEDC-32410P-A),

Topical Report (NEDC-32410P-A)," GE-NE-0000-01 02-0888. GE-NE-0000-0102-0888.

2.7 M. A. Krupa (Entergy Operations Inc.) to U.S Nuclear Regulatory Commission Document Control Desk, "Response to NRC Request for Additional Additional Information Information Pertaining to to License Amendment Request for Power Range Neutron Neutron Monitoring Monitoring System (TAC (TAC No. No. ME2531 ME2531)," ),"

14, GNRO-2010/00070, dated December 14, 2010 (ML103490095) - Response to RAI 2010 (ML103490095) - Response to RAI 5.

2.8 M.A. Krupa (Entergy Operations, Operations, Inc.) Inc.) to to U.S. Nuclear Nuclear Regulatory Regulatory Commission Commission Document Document D k, to NRC . _"",..._....,,,....,

Control Desk, "Responses to NRC Requests for Additional Information Pertaining to for Additional Information Pertaining to Amendment Request for License Amendment Request for Power Range Neutron Monitoring System (TAC No.Range Neutron Monitoring System (TAC No.

• .--r,. ..... ~ ... ), GNRO-2011/00032, May 3,2011 (ML111230756) - Response to RAI 4.4.

ME2531)," GNRO-2011/00032, dated May 3,2011 (ML111230756) - Response to RAI RSI#3 RSI #3 Demonstrate the common-cause Demonstrate common-cause failure failure vulnerabilities vulnerabilities and and the the defense-in-depth defense-in-depth for for the the algorithms and the backup stability solution.

detection algorithms and the backup stability solution. It is unclear if the primary and backup It is unclear if the primary and backup s bility trip functions for MELLLA+ use the same stability trip functions for MELLLA+ use the same software and are therefore subject to software and are therefore subject to common-cause failure.

software common-cause failure. Please provide provide aa discussion discussion of of the the postulated postulated Software Software Common-Cause Failure (SWCCF) with possible Common-Cause Failure (SWCCF) with its possible consequences on diversity and defense- consequences on diversity and in-depth.

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Response

As documented in the response for RSI #2 above, the DSS-CD (Detect and Suppress Solution - Confirmation Density) (Confirmation Density Algo Algorithm m (CDA) and Automatic Backup Stability Protection (ABSP)) and Option III are all software modules within the Power Range Neutron Monitoring (PRNM) system that would provide the stability solution. Prior to the implementation of MELLLA+, the Option III is the licensed solution for instability protection. With the implementation of MELLLA+, DSS-CD would become the licensed stability solution. The Option III trips (Period Based Algorithm (PBA), Amplitude Based Algorithm (ABA), and Growth Rate Algorithm (GRA)) are maintained in the PRNM system to provide defense-in-depth capability capabilitl. 6. However, However, under under the the DSS-CDDSS-CD solution solution the the Option Option III III trips trips are not credited for licensing basis. In the unlikely event that the CDA becomes inoperable, such as not being able to meet the minimum number of cells requirements in more than one channel, the ABSP will be manually activated to provide the stability protection. Therefore, the ABSP is an alternative stability solution to CDA only in those rare instances. As described in the Section 7 and in the Technical Specification (TS) changes documented in the approved DSS-CD licensing topical report (LTR), NEDC-33075P-A (Reference 3.1), 1), there are two Backup Stability Protection (SSP)

(BSP) options that are based on selected elements from three distinct constituents. The three constituents are:

4. BSP Manual Regions that comprise plant-specific scram (Region I) and Controlled Entry (Region II) regions in the licensed power/flow I"\nL~r~t'lnn 1I1""l,.... ...... t"r"t"'t n.!"'\\AII" ... /1'III"'\\A' operating domain domain and associated manual operator actions (Section 7.2 of Reference 3.1). 1).

5. BSP Boundary that defines the operating domain portion where potential instability events can be effectively addressed by specific operator actions (Section 7.3 of Reference 3.1).

6. Automated BSP (ABSP) Scram Region, which comprises an automatic reactor scram region initiated by the APRM flow-biased scram setpoint (Section 7.4 of Reference 3.1).

The two BSP options are:

Option 1: Consists of the BSP Manual Regions, BSP Boundary and associated operator actions.

Option 2: Consists of the ABSP Scram Region, as implemented by the Average Power Range Monitor (APRM) flow-biased scram setpoint, Region II and associated operator actions.

The TS changes I""l ..... "I . . . contained in n\t I""l,,, ...... 1""' ....... ,.\t"'t uo'toronl"o in Reference 3.11 delineate specific implementation requirements for both BSP Options in the unlikely event the Oscillation Power Range Monitor (OPRM) system is svs"[em is declared inoperable. inoperable. In In such such instance, the operators have 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> to manually implement the ASBP function. In the .interim, ASBP function. In . . . +I""l..... instability protection is implemented via the 1"Y\

6 See page 1-3 of NEDC-33075P-A (Reference 3.1). The "defense-in-depth" is meant to be additional algorithms that that are are not not credited credited in in the the licensing basis basis but but provide provide additional additional protection protection against against 11"'\""3,.\tll"\lll"'\""3t~~1"I oscillations.

unanticipated oscillations. It It isis not not meant meant toto be be a a diverse diverse stability stability solution. solution.

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Manual BSP regions, which are administratively implemented. In case the ABSP function cannot be implemented or is inoperable, the licensed stability solution becomes the Manual BSP region with the BSP Boundary, which is manually implemented through administrative actions. Therefore, the ABSP function is not meant to provide diversity for the COA CDA algorithm, but to be an alternative solution to be adopted only in specific instances. In addition to this, even if both the OPRM system (Le., the CDA) and the ASSP ABSP function function are are Iinoperable, r " l " n " ..*,....hll" stability t"'t*,....t'*\II ....."

protection is provided through a Manual BSP solution, which only relies on operator actions and plant procedures. This is essentially the same backup approach utilized in Option III for the Period Based Detection Algorithm (PBDA). In Option III solution there is only one SSP BSP Option, which is provided by the Manual BSP Regions and associated operator actions.

The evaluation of the Common-Cause Failure (CCF), including the SWCCF, for the PRNM and Option III trip functions are provided in References 3.2, 3.3 and 3.4. With the implementation of MELLLA+, there is no change to the other PRNM trip functions. The Option III trip functions are still maintained but not credited. Either the DSS-CD or the ABSP trip function will be activated to provide the stability solution. If neither function can be activated the stability protection is provided, per TS, via Manual BSP regions and BSP Boundary. The DSS-CD software is based on the Option III software for Local Power Range Monitor (LPRM) response within the OPRM cells whereas the ABSP software is based on the APRM software for core wide neutron flux and flow responses. In either case, the consequences of a postulated CCF for DSS-CD or ABSP are same as the response postulated CCF for Option III. Specifically, the discussion of "Undetected Power Oscillations" in Reference 3.2 is applicable to DSS-CD. For clarification, the discussion is presented below with additional emphasis on DSS-CD and ABSP.

Undetected Power Oscillations The OPRM system (supporting either DSS-CD or Option III solution) plays an important role in the detection and suppression of power oscillations. The postulated CCF, assumed to result in comprehensive loss of the PRNM system functionality, would also disable the OPRM system (Le., CDA for DSS-CD and PBDA for Option III). In addition to this, the loss of PRNM functionality would also disable the ABSP function of DSS-CD because the APRM system would no longer be available.

Although the GGNS Final Safety Analysis Report (FSAR) does not include power oscillations among the Anticipated Operational Occurrences (AOO) or Design Basis Accident ((DBA), it is appropriate to discuss them. As discussed in Table 8-1 of Reference 3.2, the postulated CCF in the PRNM system results in the system providing valid indications of plant conditions until the transient, at which time they become anomalous. In the case of power oscillations, thus, 1"'\t"'\1"\1"'\ ......... l""l the PRNM system indications of power and flow would track consistently with other plant indicators as they change to a state point where the potential exists for high growth-rate power oscillations (Le., the upper left corner of the power/flow map), but somehow fail to provide any protection if large amplitude oscillations begin to occur. Nevertheless, even while maintaining the severity of the postulated CCF, the plant has the ability to cope with it in conjunction with power oscillations.

to GNRO-2013/00100 Page 8 of 14 Non-Proprietary Information - Class I (Public)

((

))

GGNS procedures require immediate action to reduce reactor power in order to mitigate possible high growth-rate power oscillations following unanticipated core flow reduction events, such as (( )) The operators would know the statepoint because the status of recirculation pumps is provided independent of the PRNM system; flow information is available from the recirculation flow system, and power level information is available from either the electrical power output or a core thermal power calculation. Furthermore, the reactor recirculation flow system, Rod Control and Information System (RC&IS), and manual scram are unaffected by the CCF.

Thus, the plant is is able to cope with with the CCF because they can can determine that defensive steps Nl"'\'tl"'\I"'\t"lin/l"'\

are necessary and execute those steps via immediate actions, Le., ((

)) Because "V""""~"'I"~ throughout this event, the

_v""',......"" ..,..., the SLMCPR is not exceeded acceptance criteria provided in BTP 7-19

" ........!""\ .... ' " 9 (Reference 3.5) are automatically met.

((

))

to GNRO-2013/00100 Page 9 of 14 Non-Proprietary Non-Proprietary Information Information -- ClassClass II (Public)

(Public)

The ABSP is an alternative stability stability solution solution in in the the remote remote case case where CDA CDA becomes inoperable. However, ABSP is designed to prevent the core from However, ABSP is designed to core from operating operating in in regions regions with with high high potential potential for for THI.

THI. Therefore, Therefore, a a postulated postulated CCF CCF of of the the ABSP ABSP would would mean mean that that the the automatic automatic scram scram would would not not occur occur when when the reactor is the reactor is operating operating in in the the BSP BSP Scram Scram region.

region.

The The procedures procedures for for immediate immediate action action toto reduce reduce reactor reactor power power as as discussed discussed above above would would apply. Thus, plant would still be able apply. Thus, the plant would still be able to cope with the CCF. to cope with the CCF.

In In summary, summary, GGNS GGNS evaluation evaluation of of the the CCF CCF for for the the PRNM PRNM system system with with DSS-CD DSS-CD was was performed performed to disposition undetected power oscillations using the acceptance criteria provided in to disposition undetected power oscillations using the acceptance criteria provided in BTP BTP 7-19.

19. ItIt was was determined determined that that sufficient sufficient redundancy redundancy and and diversity diversity exists exists soso that that the the plant plant has has the the ability to cope with any CCF in the PRNM system with ability to cope with any CCF in the PRNM system with Option III or DSS-CD. The CCF Option III or DSS-CD. The CCF evaluations in evaluations in References References 3.2, 3.3 3.3 and and 3.4 3.4 were were reviewed reviewed by by the the NRCNRC andand the the PRNM PRNM system system was was approved approved by by the the NRC NRC (Reference (Reference 3.7).

References:

3.1 3.1 GE Hitachi Hitachi Nuclear Nuclear Energy, Energy, GE Hitachi Hitachi Boiling Boiling Water Water Reactor Reactor DetectDetect and and Suppression Suppression Solution Solution - Confirmation Density, Density, NEDC-33075P-A, Revision Revision 8, 8, November November 2013 2013 (ML111610593).

(ML111610593).

3.2 M.

M. A. A KrupaKrupa (Entergy (Entergy Operations Operations Inc.) Inc.) toto U.S U.S Nuclear Nuclear Regulatory RegUlatory Commission Commission Document Document Control' Control Desk, "Response to to NRC NRC Request Request for for Additional Additional Information Information Pertaining Pertaining to to License Amendment Request for Power Neutron Amendment Request for Power Range Neutron Monitoring System (TAC No. ME2531)," Monitoring (TAC No. ME2531),"

GNRO-2011/00039, dated GNRO-2011/00039, dated MayMay 26,20112011 (ML111460590)

(ML111460590) -- Response Response to to RAI RAI 8.

8.

3.3 M. A. Krupa (Entergy Operations Inc.) to U.S Nuclear Regulatory Commission Document Control Desk, "Response to NRC Request for Additional Additional Information Information Pertaining Pertaining to to License Amendment Request for Power Range Neutron Monitoring System (TAC No.Neutron Monitoring System (TAC No. ME2531),

ME2531),"

GNRO-2011/00039, dated May 26,2011 2011 (ML111460590)

(ML111460590) -- Response to RAI RAI 9.

9.

3.4 M. A Krupa (Entergy Operations Inc.) to U.S Nuclear Regulatory Commission Document Control Desk, "Response to NRC Request for Additional Information Pertaining to License Amendment Request for Power Range Neutron Monitoring System (TAC No. ME2531),"

GNRO-2011/00039, dated May 26,2011 (No, ML111460590) - Response to RAI 10.

3.5 USNRC USNRC Standard Review Plan, "Guidance for Evaluation of of Diversity Diversity and and Defense-In-Defense-In-Depth in in Digital Digital Computer-Based Computer-Based Instrumentation Instrumentation and and Controls Controls Systems," BTP BTP 7-19 7-19 (NUREG-0800),

(NUREG-0800), Revision 6, July 2012. July 3.6 3.6 GE GE Nuclear Nuclear Energy, "Nuclear "Nuclear Measurement Measurement Analysis Analysis andand Control Control Power Power Range Range Neutron Neutron Monitor (NUMAC PRNM) Retrofit Plus Option III Stability Monitor (NUMAC PRNM) Retrofit Plus Option III Stability Trip Function," NEDC-32410P-A, Trip Function," NEDC-32410P-A, 11""'l1""'l1#"'\~""I"'\'" 1 November 1997.

Supplement 1, November 1997.

NRC (A.Wang) 3.7 NRC (AWang) to Entergy Entergy Operation, Operation, Inc Inc (VP, (VP, Operations),

Operations), "Grand"Grand Gulf Gulf Nuclear Nuclear Station, Station, Unit Unit 1 - Issuance of Amendment RE: Power Range Neutron Monitoring 1 Issuance of Amendment RE: Range Neutron Monitoring System System Replacement Replacement (TAC (TAC No.

No. ME2531),"

ME2531 )," MarchMarch 28, 28, 2012 2012 (ML120400319).

(ML120400319).

1\ 1'l"'... I"" ..... l""r"\l"'\lnl" 3 Attachment 3 to GNRO GNRO-2013/00100 -201 Page10 10ofof14 14 Non-P roprietary Inform ation- I (Public)

Non-Proprietary Information - Class I (Public)

RSI#4 RSI #4 ated SWCC F conditi on disablesan appropriate response, then what IfIfaapostul postulated SWCCF condition an appropriate response, be then what le documented on-cau se failure would availab to provide divers diverse e means means not notsubjec subject t totothethesame samecomm common-cause failure would be available how it to provide meets the proposed design with adequate protection? Please discussthe .....,IO:]CO "",IV',,",,",," "V the proposed design with respect d to analys howis it meets meetin g the the ical Positio n (BTP) 9. Provide a detaile guidance of Branch Techn of Branch Technical Position (BTP) 7-19. Provide a detailed analysis meeting the guidan guidance ce contai contained BTP 7-1 ned inin BTP 7-19.

Respo Response nse tion of the Common-Cause Failure (CCF) for References 4.1, 4.2, and and 4.3 4.3 provid provided ed the the evalua evaluation of the Common-Cause applicaFailure ble to both(CCF) for the Power the Power Range Neutro Neutron n Monito Monitor system. The discussion is (PRNM)) system. The discussion (DSS-r (PRNM is applicable to both ess Solutio n -- Confir mation Density CD) solutions with Option III Option III and Detect and and Suppr Suppress Solution Confirmation Density (DSS-CD) solutions evaluawith tion the additio the additional nal explanation docum documented ented in in the response to RSI #3 above. The the response to RSI #3 above. The CCF evaluation and DSS-CD (both Confirmation Densit y shows that that the PRNM PRNM system with with Option Option III III and DSS-CD (both Confirmation

)) the Density guidan ce of atic Backu p Stabili ty Protec tion (ABSP Algorit Algorithm hm (CDA) (CDA) andand Autom Automatic Backup Stability Protection (ABSP)) of meets the the guidance CCF evalua tionofis 4.4). A summ ary Branch Techn Technical ical Positio Position (BTP) 7-19 n (BTP) 7-19 (Reference 4.4). A summary of the CCF evaluation is presented below.

BTP 7-19 (Reference 4.4) Criterion Evaluation of PRNM with DSS-CD (1) For each anticipated operational occurrence in the design basis occurring in conjunction with each single postulated CCF, the plant response calculated using realistic assumptions (e.g., plant operating at normal power levels, temperatures, pressures, flows, normal alignments of equipment, etc.) analyses should not result Table 8-1 of Reference 4.1 provided an in radiation release exceeding 10 percent of evaluation for each Anticipated Operational the applicable siting dose guideline values Occurrences (AOO) and Design Basis or violation of the integrity of the primary Accident (DBA) in the Grand Gulf UFSAR.

coolant pressure boundary. The Based on the evaluation presented in Table applicant/licensee should 8-1, the proposed upgrade satisfies Acceptance Criteria (1).

(1) demonstrate that sufficient diversity exists to achieve these goals, (2) identify the vulnerabilities discovered and the corrective actions taken, or (3) identify the vulnerabilities discovered and provide a documented basis that justifies taking no action.

(2)For each postulated (2) postulatedaccident accidentininthe the Table 8-1 of Reference 4.1 provided an Table 8-1 of Reference 4.1and provided DBA inan designbasis design occurr ing in conjun basisoccurring in conjunction with ction with evaluation for each AOO evaluation for each AOO and(GGNS DBA in) the each single singlepostulated CCF, postulatedCCF, the plant the plant Grand Gulf Nuclear Station Grand Gulf Nuclear Station (GGNS) respon response calculated se calcula usingrealistic tedusing realistic Updated Final Safety Analysis Report Updated Final Safety Analysis Report tion assum anal shouldnot notresult resultinin UFSA R. on the evalua assumptions analvses should (UFSAR). Based on the evaluation to GNRO-2013/00100 Page 11 of 14 Non-Proprietary Non-Proprietary Information Information -- Class II (Public) (Public)

BTP 7-19 (Reference 4.4) Criterion Evaluation of PRNM with DSS-CD radiation radiation release exceeding exceeding the applicable presented presented in in Table Table 8-1 8-1,, thethe proposed proposed siting dose guideline values, violation of the upgrade satisfies Acceptance Criteria siting dose guideline values, violation of the upgrade satisfies Acceptance Criteria (2).

(2).

integrity of integrity of the the primary primary coolant coolant pressure boundary, or boundary, or violation violation of of the the integrity integrity of of the the containment containment (i.e.,

(i.e., exceeding exceeding coolantcoolant system system or or containment containment design design limits).

limits). The The applicant/licensee should applicant/licensee should (1) (1) demonstrate demonstrate that that sufficient sufficient diversity diversity exists exists to to achieve achieve these (2) identify these goals, (2) identify thethe vulnerabilities vulnerabilities discovered and discovered and the the corrective corrective actions actions

taken, taken, or (3) identify or (3) identify the vulnerabilities vulnerabilities discovered and discovered and provide provide a a documented documented basis basis that justifies taking no that justifies taking no action.

(3) When a failure of a common element or signal source shared by the control system This criterion requires an evaluation of and reactor trip system (RTS) is postulated potential interaction between the Control and the CCF results in a plant response System and RTS echelons when a that requires reactor trip and also impairs postulated CCF results in a plant response the trip function, then diverse means that that requires a reactor trip and also impairs are not subject to or failed by the postulated the trip function. PRNM system is not used failure should be provided to perform the for automatic control of plant operations, so RTS function. The diverse means should if the postulated CCF occurs, it will not assure that the plant response calculated result in a plant response that requires a using realistic assumptions analyses does reactor trip. Therefore, the type of CCF not result in radiation release exceeding 10 described in this criterion cannot occur in percent of the applicable siting dose the upgrade system. Acceptance guideline values or violation of the integrity Criterion (3) is satisfied.

of the primary coolant pressure boundary.

This criterion requires an evaluation of (4) When a failure of a common element or potential interactions between the Control signal source shared by the control system System and Engineered Safety Features and ESFAS is postulated and the CCF echelons when Actuation System (ESFAS) echelons when results in a plant response that requires in a plant a postulated CCF results in a plant engineered safety features (ESF) and also function, then response that requires an

~~~~~*~~~that an ESF response impairs the ESF function, then diverse diverse means not subject to oror failed by and also and also impairs impairs ESF function. PRNM ESF function. PRNM means that are are not subject to failed by is not used for automatic control of of the postulated failure should be provided to system is not used for automatic control C'"C'"rr1orT"ll the postulated failure should be provided to plant operations, so so ifif the the postulated function. The (JIt/F--~r."'R plant operations, postulated CCF CCF perform the ESF function. The diverse means occurs, occurs, itit will will not result in not result in aa plant plant response response means should should assure assure thatthat the plant the plant that requires an ESF response.

that requires an response.

response calculated using realistic calculated using realistic Furthermore, neither the existing nor nor Furthermore, neither the existing assumptions analyses does not result in replacement PRNM system interface with replacement PRNM system interface with radiation release exceeding 10 percent of the applicable siting dose guideline guideline values values the ESFAS. Therefore, the the type of the type of CCF CCF the applicable siting dose in this rt~~""I"U"'\l"'\rt in this criterion criterion cannot cannot occur occur in in or violation the integrity of the primary described or violation of the integrity of the primary the upqrade u rade system. stem. Acceptance Acce tance the

l \ 1'1:':11\1\ It"\""'H'" Inl' 3 to GNRO-201 GNRO-2013/00100 Page 12 12 of 14 14 Non-Proprietary Non-Proprietary Information Information -- Class II (Public) (Public)

STP 7-19 (Reference 4.4) Criterion Evaluation of PRNM with DSS-CD coolant pressure boundary. Criterion (4) is satisfied.

This criterion requires that a failure in the monitoring and display echelon will not not (5) No failure of monitoring or display adversely affect the RTS RTS or ESFAS systems should influence the functioning of echelons.

echelons. PRNM PRNM system system doesdoes notnot rely rely on on the RTS or ESFAS. If a plant monitoring or any input from the monitoring or receive any input from the monitoring system failure induces operators to attempt and display echelon; therefore, a failure in to operate the plant outside safety limits or the monitoring and display systems will not in violation of the limiting conditions of propagate propagate to to PRNM PRNM system.

system. If If the failure in the failure in operation, the analysis should demonstrate the monitoring and display system results in that such operator-induced transients will an operator-induced transient, the be compensated by protection system automatic protective functions of PRNM function.

system are

"'-="'-='IL.lII.

are available available for for compensation.

compensation.

Acceptance Criterion (5) Criterion (5) is is satISTle,a.

satisfied.

(6) For safety systems to satisfy IEEE Std. 603-1991 Clauses 6.2 and 7.2, which are incorporated by reference in 10 CFR 50.55a(h), a safety-related means shall be This This criterion requires requires a a Q~tot\l_rOII~tt::ln safety-related provided in the control room to implement means for manual initiation of the RTS and manual initiation at the division level of the ESFAS functions.

RTS and ESFAS functions. The means provided shall minimize the number of This criterion is not applicable to the PRNM discrete operator manual manipulations and system upgrade. The evaluation performed shall depend on operation of a minimum of for Acceptance Criteria (1) and (2) equipment. If the means is independent demonstrates that if a CCF occurs in PRNM and diverse from the safety-related system, the plant is able to cope without automatically initiated RTS and ESFAS relying on a manual scram or ESF functions, the design meets the system- actuation. It is noted that the manual scram level actuation criterion in Point 4 of this and ESF actuation are retained, if needed BTP. If credit is taken for a manual for other reasons, because they are totally actuation method that meets both the IEEE separate from PRNM system and not Std.603-1991, Clauses 6.2 and 7.2 affected by the proposed upgrade in any requirements and a a need for a diverse way.

manual backup, then the applicant/licensee JJL.l'-""" ..... JJ. then should should demonstrate demonstrate that that the the criteria criteria are are satisfied and satisfied and sufficient diversity exists. sufficient diversity (7)

(7) If If the the 03 03 assessment reveals a potential potential These criteria criteria require require evaluations evaluations of of the the for a CCF, then then the the method method forfor methods for accomplishing the independent accomplishing accomplishing the the independent independent and and rJIV'AT.f;J;A diverse and diverse means of actuating the means means of of actuating actuating the the protective protective safety safety protective safety function when the functions functions can can be accomplished via via either anan Defense-in-Depth (03)

Defense-In-Depth (03) analysis reveals the nn't,t"\I""\'t.'"I for a automated system (see Section 3.4, "Use of automated system Section of potential for a CCF.

Automation Automation in in Diverse Diverse Backu Backup Safety to GNRO-2013/00100 Page 13 of 14 Non-Proprietary Information - Class I (Public)

BTP 7-19 (Reference 4.4) Criterion Evaluation of PRNM with DSS-CD Functions" below), or manual operator The Nuclear Measurement Analysis and actions that meet HFE acceptability criteria pi orm is not present in Control (NUMAC) platform (see Section 3.5, "Use of Manual Action in any part of RTS except the PRNM system, Diverse Backup Safety Functions" below). and is not present in the ESFAS. Their designs are not affected by the proposed upgrade, and these systems are not vulnerable to the postulated CCF in PRNMS. Therefore, Acceptance Criterion (7) is not applicable to the PRNMS upgrade.

(8) If the 03 assessment reveals a potential for a CCF, then the method for accomplishing the independent and diverse means of actuating the protective safety functions should meet the following criteria:

The independent and diverse means should be:

a) at the division level; These criteria require evaluations of the b) initiated from the control room; methods for accomplishing the independent and diverse means of actuating the c) capable of responding with sufficient time protective safety function when the D3 available for the operators to determine the analysis reveals the potential for a CCF.

The NUMAC platform is not present in any need for protective actions even with part of RTS except the PRNM system, and malfunctioning indicators, if credited in the is not present in the ESFAS. Their designs 03 coping analysis; are not affected by the proposed upgrade, d) appropriate for the event; and these systems are not vulnerable to the postulated CCF in PRNMS. Therefore, e) supported by sufficient instrumentation Acceptance Criterion (8) is not applicable to that indicates: the PRNMS upgrade.

1. the protective function is needed, r:::::lT'~TV'-r~jr:::lTt:..~n automated

2. the .safety-related automated system system did did not perform the protective \L1L\L.1\L..ILI II' \L.I function, and

3. the automated backup or manual action is successful in performing the safety function.

(9) If the 03 assessment reveals a potential These criteria require evaluations of the for a CCF, then, in accordance with the methods for accomplishing the independent auamented Quality auidance for the and diverse means of actuatin actuating the

Attachment 3 to GNRO-2013/00100 00 Page 14 of 14 Non-Proprietary Information - Class I (Public)

BTP 7-19 (Reference 4.4) Criterion Evaluation of PRNM with DSS-CD independent independent and and diverse diverse backup backup system system protective safety function when the D3 03 used to cope with a used to cope with a CCF, the CCF, the design design ofof a a analysis reveals the potential for a CCF.

diverse diverse automated automated or or diverse diverse manual manual The NUMAC platform is not present in any backup backup actuation actuation system system should should address address part of RTS except the PRNM system, and how how to to minimize minimize the the potential for a potential for a spurious spurious is is not present in not present in the the ESFAS.

ESFAS. Their designs Their designs actuation of the protective system actuation of the protective system caused caused are not affected by the proposed are not affected by the proposed upgrade, upgrade, by by the the diverse diverse system.

system. UseUse ofof design design and these systems are not vulnerable to the techniques (for example: redundancy, techniques (for example: redundancy, postulated CCF in PRNM system.

conservative conservative setpoint selection, and setpoint selection, and use use of of Therefore, Therefore, Acceptance Acceptance Criterion Criterion (9)

(9) is is not not quality quality components) components) to to mitigate mitigate these these applicable applicable toto the the PRNM PRNM system system upgrade.

upgrade.

concerns concerns is is recommended.

recommended.

In In summary, summary, the the proposed proposed upgrade upgrade was was evaluated evaluated using using the the acceptance acceptance criteria criteria provided provided inin BTP 7-19 (Reference 4.4). The GGNS specific disposition of a CCF for the PRNMS with DSS-CD DSS-CD is documented in is documented in the the response to to RSI RSI #3#3 above above and and itit is is not not repeated repeated in in this this response. It was confirmed with the additional explanation documented for DSS-CD (provided (provided in in the the response response to to RSI RSI #3)

1. 3) that that sufficient sufficient redundancy redundancy and and diversity diversity exists exists soso that that the the plant has the ability to cope with any CCF in PRNMS with Option III or DSS-CD. The CCF evaluation evaluation (References (References 4.1, 4.1, 4.2, 4.2, and and 4.3) 4.3) was was reviewed reviewed byby the the NRCNRC andand the the PRNM PRNM system system was approved by the NRC (Reference 4.5).

References:

4.11 M. A. Krupa (Entergy Operations, Inc.) to U.S. Nuclear Regulatory Commission Document Control Control Desk, "Response to to NRC NRC Request for for Additional Additional Information Information Pertaining Pertaining to to License Amendment Request for Power Range Neutron Monitoring System (TAC No. ME2531),"

GNRO-2011/00039, dated May 26,2011 2011 (ML111460590) - Response to RAI 8.

4.2 M. A. Krupa (Entergy Operations, Inc.) to U.S. Nuclear Regulatory Commission Document Control Desk, "Response to NRC Request for Additional Information Pertaining to License Amendment Amendment Request Request forfor Power Power Range Neutron Neutron Monitoring Monitoring System System (TAC(TAC No.

No. ME2531),"

ME2531),"

GNRO-2011/00039, GNRO-2011/00039, dated May 26,2011 (ML111460590) - Response to RAI 9.

May 1 (ML111460590) - Response to RAI 9.

4.3 M. A. Krupa (Entergy Operations, Inc.) to U.S. Nuclear Regulatory Commission Document Control Control Desk, "Response "Response to to NRC NRC Request Request forfor Additional Additional Information Information Pertaining Pertaining to to License License Amendment Request for Power Range Neutron Monitoring System (TAC No. ME2531),"

GNRO-2011/00039, dated May 26,2011 (ML111460590) - Response to RAI 10.

4.4 4.4 USNRC USNRC Standard Standard Review Review Plan, Plan, "Guidance "Guidance for for Evaluation Evaluation ofof Diversity Diversity and and Defense-In-Defense-In-Depth in Digital Computer-Based Instrumentation and Control Systems," NUREG 0800, BTP BTP 7-19, 7-19, Revision Revision 6,6, July July 201 2012.

4.5 NRC (A.Wang) to Entergy Operations, Inc. (VP, Operations), "Grand Gulf Nuclear Station, Unit 1 - Issuance of Amendment RE: Power Range Neutron Monitoring System Replacement (TAC No. ME2531)," March 28,2012 28, 2012 (ML120400319).