ML20236T155

Application for Amend to License NPF-29,revising TSs Re BWROG Enhanced Option I-A Reactor Stability Long Term Solution

ML20236T155
Person / Time
Site:	Grand Gulf
Issue date:	07/20/1998
From:	Hogan J ENTERGY OPERATIONS, INC.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20236T158	List: ML20236T155 ML20236T166
References
GL-94-02, GL-94-2, IEB-88-007, IEB-88-7, NUDOCS 9807270466
Download: ML20236T155 (18)

Text

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ . . _ _ _ _ _ _ _ _ _ _

l Entirgy Oper;ti:ns,Inc.

=Entergy PO. Box 756 w:=~"

Fax 601437 2795

}

Joseph J. Hagan Vice President Operanons Grand Gulf Nuclear Station July 20, 1998 L

U.S. Nuclear Regulatory Commission Mail Station P1-37 Washington, D.C. 20555 Attention: Document Control Desk

Subject:

Grand Gulf Nuclear Station Docket No. 50-416 License No. NPF-29

" Enhanced Option I-A Core Stability" {

Proposed Amendment to the Operating License 3 (LDC 1997-02) m GNRO-98/00053 l

l In accordance with 10 CFR 50.90, Entergy Operations, Inc. (EOl) hereby applies j for amendment of Facility Operating License No. NPF-29, Appendix A - Technical Specifications, for Grand Gulf Nuclear Station (GGNS). The proposed changes implement Boiling Water "aactor Owners Group (BWROG) Enhanced Option I-A (E1 A) Reactor Stability L. Perm Solution as documented in NEDO-32339-A, Revision 1 T.eactor Stam Long-Term Solution, Enhanced Option I-A" and Supplements 1-4, which have been previously reviewed and approved by the NRC.

This request consists of proposed changes required to implement the Boiling Water Reactor Owners Group (BWROG) Enhanced Option I-A (E1 A) Reactor Stability Long Term Solution as documented in NEDO-32339-A, Revision 1,

" Reactor Stability Long-Term Solution, Enhanced Option I-A" and Supplements 1-

4. By letter dated February 25,1998, the NRC staff provided their acceptance of E1 A as a technically acceptable implementation of a Long-Term Stability Solution satisfying the requirements of NRC IE Bulletin 88-07, Supplement 1 and Generic Letter 94-02, "Long-Term Solutions and Upgrade of Interim Operating o\

Recommendations for Thermal-Hydraulic instabilities in Boiling Water Reactors."

With implementation of the E1 A solution, the stability interim Corrective Actions g\

i (ICAs) implemented in response to NRC IE Bulletin 88-07, Supplement 1 and

[ Generic Letter 94-02 are no longer required.

m 9807270466 DR 9807dO ADOCK 05000416 PDR; a__-___--__________--_. _ . _ _ - _ _ _ _ _ _ _ _

'k ;

1 GNRO-98/00053 -

. Page 2 of 3

' The Technical Specification (TS) changes required to implement the Enhanced

. Option I-A solution are identified in NEDO-32339-A, Supplement 4, Revision 1,

" Reactor Stability Long Term Solution: Enhanced Option I-A Generic Technical Specifications." Additional changes to TS 3.3.1.1, " Reactor Protection System (RPS) instrumentation" and TS 3.4.1,'" Recirculation Loops Operating".are required to support implementation of the E1A solution and to remove TS requirements associated with the stability ICAs implemented in response to NRC IE Bulletin 88 .

07, Supplement 1. The TS Bases supporting the above mentioned changes are

- also included with the proposal. Furthermore, EOl confirms that the other required features of the E1 A solution not implemented in TS will be appropriately implemented in the Updated Final Safety Analysis Report and plant operating -

procedures.

~

A note of Affirmation is contained in Attachment 1 of this letter. Attachment 2 provides a description of the proposed changes and the assor.iated justification 1 (including a Basis for No Significant Hazards Considerations). The description of (

' the proposed changes and justifications are in full agreement with the generic

~ Technical Specifications in NEDO-32339-A, Supplement 4, Revision 1.

A marked-up copy of the affected pages from the GGNS Technical Specifications

. (TS) is provided in Attachment 3. The applicable marked-up TS Bases pages are ,

included for your information as Attachment 4. The TS Bases changes will be  !

implemented in accordance with. Technical Specification 5.5.11, Technical

~

Specification Bases Control Program, following approval of the requested i amendment.-

This request has been reviewed and approved by the GGNS Safety Review Committee and the Plant Safety Review Committee. In order to allow time for completion of procedure revisions and training, EOl requests that the proposed ,

amendments, once approved by the NRC, be issued to GGNS allowing Implementation within 120 days of issuance of the NRC's Safety Evaluation  !

Report.. If you have any questions regarding this request or require additional

. information, please contact Charles E. Brooks at (601) 437-6555.

l Yours truly, .

JJH/CEB/

attachments: 1. Affirmation per 10 CFR 50.30

2. Discussion of Justification

3. Mark-up of Affected Technical Specifications Pages

4. Mark .up of Affected Technical Specification Base Pages cc: (See Next Page)

GNRO-98/00053 4 Page 3 of 3 cc: Ms. J. L. Dixon-Herrity, GGNS Senior Resident (w/a)

Mr. L. J. Smith (Wise Carter) (w/a)

Mr. N. S. Reynolds (w/a)

Mr. H. L Thomas (w/o)

Mr. E. W. Merschoff (w/a)

Regional Administrator U.S. Nuclear Regulatory Commission Region IV

' 611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011 Mr. J. N. Donohew, Project Manager (w/2)

Office of Nuclear Reactor Regulation j U.S. Nuclear Regulatory Commission Mail Stop 13H3-Washington, D.C. 20555 Dr. E. F. Thompson (w/a) l State Health Officer State Board of Health P.O. Box 1700  !

Jackson, MS 39205 l

4 I

a

Attachm:nt 1 to GNRO-98/00053 l BEFORE THE UNITED STATES NUCLEAR REGULATORY COMMISSION i

l LICENSE NO. NPF-29 I

DOCKET NO. 50-416 IN THE MATTER OF ENTERGY MISSISSIPPI, INC.

and SYSTEM ENERGY RESOURCES, INC.

and SOUTH MISSISSIP.Pl ELECTRIC POWER ASSOCIATION and ENTERGY OPERATIONS, INC.

AFFIRMATION 1, J. J. Hagan, being duly swom, state that I am Vice President, Operations GGNS of Entergy Operations, Inc.; that on behalf of Entergy Operations, Inc., System Energy Resources, Inc., and South Mississippi Electric Power Association I am authorized by Entergy Operations, Inc. to sign and file with the Nuclear Regulatory Commission, this application for amendment of the Operating License of the Grand Gulf Nuclear Station; l that I signed this application as Vice President, Operations GGNS of Entergy Operations, Inc.; and that the statements made and the matters set forth therein are true and correct to the best of my knowledge, information and belief.

7 C& - -

J. .H3gan STATE OF MISSISSIPPI COUNTY OF CLAIBORNE SUBSCRIBED AND SWORN TO before me, a Notary Public, in and for the County and State above named, this.plo "' day of 'Jo L 'f 1998.

(SNAL)h I

c-Notary Pubib Ah \

l M mission exp. ires:

57ATfnCEN0T m pug's 87ECA2NOTANM l

t t

Att chm:nt 2 to GNRO-98/00053 r pig 31 of 14 ENTERGY OPERATIONS, INC.

GRAND GULF NUCLEAR STATION j DOCKET 50-416/ LICENSE NO. NPF-29 l PROPOSED CHANGE TO THE OPERATING LICENSE A. Affected Technical Specification,s This proposed change affects the following Technical Specification sections:

l. 3.2.4 Fraction of Core Boiling Boundary (FCBB) 3.3.1.1 RPS Instrumentation 3.3.1.3 Period Based Detection System (PBDS) 3.4.1 Recirculation Loops Operating 5.6.5 Core Operating Limits Report (COLR)

B 3.2.4 Bases - Fraction of Core Boiling Boundary (FCBB) l- B 3.3.1.1 Bases - RPS Instrumentation B 3.3.1.3 Bases - Period Based Detection System (PBDS)

B 3.4.1 Bases - Recirculation Loops Operating B. Background The requirements of 10 CFR 50 Appendix A, General Design Criterion 12 (GDC-12) specify that neutronic/ thermal-hydraulic instability is to be prevented by design or be

,. readily and reliably detected and suppressed. After neutronic/ thermal-hydraulic

! instability events occurred in the early 1980s at Boiling Water Reactors (BWRs) outside the United States, it was recognized that some BWR designs did not prevent neutronic/ thermal-hydraulic instability. To improve the ability of the operator to detect and suppress potential neutronic/ thermal-hydraulic instability, General Electric prepared Service Information Letter (SIL) #380, Revision 1. The recommendations of SIL #380, l Revision 1, were developed based on the limited event and test data available.

Following the neutronic/ thermal-hydraulic instability event at LaSalle Unit 2 in early 1988, the NRC issued IE Bulletin 88-07. The BWROG responded to concerns raised by the NRC pertaining to the neutronic/ thermal-hydraulic instability issue by performing studies utilizing newer and more detailed models originally developed for the analysis of other BWR neutronic/ thermal-hydraulic transients. The results of these BWROG studies indicated that there was a potential for neutronic/ thermal-hydraulic instability to exceed specified acceptable fuel design limits established for the anticipated operational I

occurrences routinely evaluated to demonstrate compliance with 10 CFR 50 Appendix A, General Design Criterion 10 (GDC-10). Specifically, it was concluded that neutronic/ thermal-hydraulic instability can result in power oscillations which could result in exceeding the Minimum Critical Power Ratio (MCPR) Safety Limit (SL) prior to automatic actuation of the Reactor Protection System.

Based on these results, the BWROG developed Interim Corrective Actions (ICAs) which identified operator actions to be taken based on Neutron Monitoring System (NMS) response, recirculation loop operation, and power and flow conditions within the licensed operating domain, i

L _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

Atttchmsnt 2 to GNRO-98/00053 Pega 2 of 14 The ICAs were endorsed by the NRC in IE Bulletin 88-07, Supplement 1 and incorporated into the GGNS Technical Specifications to satisfy the requirements of the Bulletin. The ICAs defined a region of the power and flow operating domain to be z

. excluded from normal operation. Subsequently, a neutronic/ thermal-hydraulic instability event occurred at WNP Unit 2 in late 1992, outside the stability regions specifed in the NRC/BWROG ICAs. As a result, upgraded interim operating recommendations were issued by the BWROG consistent with the requirements of NRC Generic Letter 94-02, "Long-Term Solutions and Upgrade of interim Operating Recommendations for Thermal-Hydraulic Instabilities in Boiling Water Reactors," dated July 1994. GGNS upgraded its procedures in a manner consistent with these additional operating recommendations.

Concurrent with the development of the ICAs, the BWROG also initiated efforts to develop generic long term solutions to the neutronic/ thermal-hydraulic instability issue.

As described in IE Bulletin 88-07, Supplement 1, the ICAs were accepted by the NRC as ,

- adequate compensatory measures pending final development and implementation of the long term solutions being developed by the BWROG. One of the solutions initially

' developed by the BWROG was designated Option I-A. The original Option I-A solution is described in NEDO-31960-A and was later enhanced through the efforts of a smaller number of BWROG participants. The result was the Enhanced Option I-A (E1 A) solution described in NEDO-32339-A, Revision 1, " Reactor Stability Long-Term Solution:

Enhanced Option I-A" and Supplements 1-4. The NRC review and approval of the E1A solution found it to be an acceptable long-term stability solution satisfying the requirements of Bulletin 88-07, Supplement 1 and GL 94-02.

The E1 A solution prevents reactor instability through a combination of features.

Implementation of some of these features require changes to the Technical Specifications (TS). ~ The BWROG E1 A Committee prepared NEDO-32339-A, Supplement 4, Revision 1 " Reactor Stability Long-Term Solution: Enhanced Option I-A i

Generic Technical Specifications" to describe the changes to the improved Standard TSs (ITS) of NUREG-1434 required to implement Enhanced Option I-A and is the basis  !

for the proposed TS changes implementing features of the E1 A solution.- The features of the E1 A solution prevent neutronic/ thermal-hydraulic instability during anticipated reactor h state conditions and transients by limiting reactor operation, including conditions resulting from unexpected transients, to prescribed reactor state conditions. .

1 The portion of the reactor power and flow operating domain that must be excluded from  ;

i ' operations due to the potential for neutronic/ thermal-hydraulic instability is designated i the Exclusion Region. The Exclusion Region is implemented through a modification to l existing Reactor Protection System Average Power Range Monitor (RPS APRM) i setpoints. The limitation on entry into this region is provided through the existing APRM ~

- Flow-Biased Simulated Thermal Power - High function. The APRM flow biased scram sotpoints are redefined to coincide with the Exclusion Region boundary. j The portion of the operating domain outside the Exclusion Region where the potential for neutronic/ thermal-hydraulic instability exists without controls on core power distribution is designated as the Restricted Region. Planned operation in the Restricted Region requires implementation of a stability control prior to entry and the Period Based Detection System (PBOS).

L - - - - ._ .- - . - - - - - - - . - - - - - - - - - - - - - - - ----- - - - - - - - _ - - - - - - - - - - -i

Att: chm nt 2 to GNRO-98/00053

, Pag 3 3 of 14

[.

The stability control is a new limit on the Fraction of Core Boiling Boundary (FCBB). As an operator aid to identify uncontrolled entry into the Restricted Region, automatic indication of uncontrolled entry into the Restricted Region is implemented through a modification to the existing control rod block instrumentation. The APRM Neutron Flux -

Upscale function flow biased control rod tiock setpoints are adjusted to coincide with the lower boundary of the Restricted Region.

The portion of the operating domain outside the Restricted Region where the potential for neutronic/ thermal-hydraulic instability is postulated under unanticipated reactor state conditions or transients is designated the Monitored Region. Operation in the Monitored Region requires implementation of the Period Based Detection System (PBDS). Other.

. features of the E1A solution, in addition to the Monitored Region, provide substantial protection from unanticipated reactor state conditions and transients. These features of the solution are not used to demonstrate compliance to GDC-12 but are required in the implementation of the E1A solution.

i C. Description of Proposed Changes

[ , Existing GGNS Technical Specification Requirements Existing TS 3.4.1 limits the allowed power and flow conditions during operation with different combinations of reactor coolant recirculation loops in operation and requires immediate manual scram with no recirculation loops operating in Mode 1. The immediate actions required, when the power and flow conditions of current TS Figure 3.4.1-1 are not met with different combinations of recirculation loops operating l and with no recirculation loops operating in Mode 1, are consistent with the Boiling Water 3 L Reactor Owners' Group (BWROG) Interim Corrective Actions (ICAs) dated November i 1988, which were instituted in response to Supplement 1 of Bulletin 88-07.

' Operating procedures based on the upgraded interim operating recommendations issued by the BWROG consistent with the requirements of NRC Generic Letter 94-02, are currently in place at GGNS and supplement the existing core stabili+y related requirements specified in TS 3.4.1.

L

. Proposed Changes

[. Implementation of the long term solution to the Boiling Water Reactor (BWR)

L neutronic/ thermal-hydraulic instability issue has been an industry and regulatory objective since the issuance of Supplement 1 of IE Bulletin 88-07. The E1A solution option was identified as the solution proposed to be implemented at GGNS in response to Generic Letter 94-02, "Long-Term Solution and Upgrade of interim Operating Recommendations for Thermal-Hydraulic Instabilities in BWRs." Upon implementation of the E1 A long term stability solution, the administrative controls established to comply with the guidance of the BWROG ICAs will no longer be required at GGNS.

The proposed changes to the Technical Specifications will enable the full implementation

~of the Enhanced Option I-A (E1 A) long term solution to address the neutronic/ thermal-hydraulic instability issue. 1 J -_ _ _ - _ - _ - ___ - - _ _ _ _ _ _ __ -

l -

l Attichm:nt 2 to GNRO-98/00053 Paga 4 of 14 i 4 I

Specifically, the proposed change adds new specifications to establish limits for Fraction of Core Boiling Boundary (FCBB) and the Period Based Detection System (PBDS),

relocates the required RPS APRM trip function to the RPS instrumentation specification in current TS 3.3.1.1 from the plant Updated Final Safety Analysis Report, deletes the limits on power and flow conditions and the requirement for immediate manual scram l with no operating recirculation loops in Mode 1 associated with the implementation of the l BWROG ICAs (current TS 3.4.1 and Figure 3.4.1-1), and modifies the description of the contents of the Core Operating Limits Report (COLR) in current TS 5.6.5. Additional '

j changes to current TS 3.3.1.1 not related to implementation of the E1 A solution are made to support relocation of the required RPS APRM trip function (Flow Biased  ;

Simulated Thermal Power-High) into the Technical Specifications.

i

!' Modification of the RPS instrumentation specification provides for automatic exclusion of the portion of the operating domain susceptible to neturonic/ thermal-hydraulic instability. ,

The two new specifications require maintaining a stability control and the availability of a l stability detection system during operation in regions of the power and flow operating i domain determined susceptible to neutronic/ thermal-hydraulic instability without controls on core power distribution. In a general sense, the current TS 3.4.1 actions required when the power and flow limits are not met and no recirculation loops are in operation in Mode 1 are replaced by the automatic exclusion of the portion of the operating domain i susceptible to neutronic/ thermal-hydraulic instability, the requirements to maintain stability control, associated required operator actions, and the required availability of the stability detection system aver the region of the power and flow operating domain susceptible to instability without stability controls and associated required operator actions. The revised action statement for no recirculation loops in operation in Mode 1 is l consistent with the generic ITS of NUREG-1434.

Additional defense-in-depth features of the solution described in NEDO-32339-A,

! Revision 1 provide substantial protection from unanticipated reactor state conditions and transients. EOl confirms that the E1 A defense-in-depth features not implemented .

through the proposed Technical Specification changes will be appropriately implemented i in licensee controlled documents.

i The additions and modifications of the proposed change are consistent with the description of the E1A solution as described in NEDO-32339-A, Revision 1 and Supplements 1-4.

i

AttrchmInt 2 to GNRO-98/00053 Paga 5 of 14 D. Justification of Proposed Changes Stability Controls The stability control used in the E1 A solution is the Fraction of Core Boiling Boundary (FCBB). The FCBB is the ratio of the power generated in the lower 4 feet of the active reactor core to the power required to produce bulk saturated boiling of the coolant entering the fuel channels. Adherence to the FCBB limit maintains the elevation of core average bulk saturation greater than 4 feet above the bottom of active fuel. The limit is based on analysis described in Section 9 of NEDO-32339-A, Revision 1. The boiling boundary limit is established to restrict allowed reactor state conditions outside the Exclusion Region boundary and to ensure that the core remains stable during normal reactor operations in the Restricted Region of the power and flow operating domain which may otherwise be susceptible to neutronic/ thermal-hydraulic instability. This core average boiling boundary is manipulated by operator actions that affect power distribution. The associated operating limit, FCBB, is required to be met during operation in the Restricted Region and meets the criteria for inclusion in the Technical Specifications as delineated in 10CFR50.36. Therefore, a new specification (GGNS i TS 3.2.4) is appropriately added to the Power Distribution Limits section of the TSs.

APRM Flow Biased Setpoints The design of the hardware required for implementation of the E1 A solution provides the ,

capability for the APRM flow biased control rod block and scram function setpoint values to define ine stability region boundaries for the different operating modes, to provide the

" Setup" setpoints prior to and during operation in the Restricted Region, and to select different trip reference sets. To implement these features, the original Flow Control Trip .

Reference (FCTR) card of the RPS is replaced with a card of a new design specifically I developed and manufactured for implementation of the E1A solution. The design of the E1 A FCTR card includes both analog and digital components as described in NEDC-32339P-A, Supplement 2, Revision 1, " Reactor Stability Long-Term Solution:

Enhanced Option I-A Solution Design." The digital design of some of the components provides the capability to perform all functions required by the E1 A flow mapping methodology, including the mapping and calibration of the Exclusion Region and Restricted Region boundaries,in terms of setpoint versus drive flow.

To facilitate intentional entry into the Restricted Region once the stability control is in place, the APRM flow biased control rod block setpoint is " Setup " With the " Setup" setpoint value selected, the setpoints associated with stability regions are elevated l above the normal or "non-Setup" value. The APRM flow biased scram setpoint is also

! elevated to preserve the margin between the rod block and scram setpoints. The APRM l flow biased control rod block function does not meet the criteria as delineated in 10CFR50.36 for inclusion in the TSs and only serves as an operator aid. Therefore, it is l

not included in the proposed TSs and will be included in licensee controlled documents.

The APRM flow biased scram function as modified by the E1A solution provides automatic reactor scram protection upon entry into the Exclusion Region as implemented through the FCTR scram function.

l 1

l

Attichm:nt 2 to GNRO-98/00053 Piga 6 of 14 I I I'

This feature of the E1A solution can potentially increase the occurrence of automatic

reactor scram since the APRM flow biased scram function causes an automatic reactor l scram upon entry into this region of the operating domain. However, the Exclusion l

Region and the region requiring immediate manual reactor scram upon entry by the recommendations of the BWROG ICAs are similar. Therefore, the overallincidence of reactor scram will not significantly change due to the E1 A modification of the APRM flow biased scram function.

I As described in NEDO-32339-A, Supplement 4, Revision 1, gross violation of the currently licensed operating domain is prevented by the APRM flow biased scram function. In addition, the E1 A APRM flow biased scram function provides a preemptive automatic reactor scram upon entry into the Exclusion Region of the operating domain

and is not used directly to protect the MCPR Safety Limit. Therefore, the APRM flow biased scram function does not meet the criteria for inclusion in the Technical L Specifications as delineated in 10CFR50.36. However, since the APRM flow biased i

i scram function is a feature of the E1A stability solution necessary to ensure compliance

! with 10 CFR 50 Appendix A, General Design Criterion 12, it is added to TS Table l 3.3.1.1-1 as RPS Function 2.d. The applicable modes, required channels per trip system, l conditions and surveillance requirements for this function are taken from TS 3.3.1.1 l Function 2.b of NUREG-1434, " Improved BWR-6 Technical Specifications," dated

! September 1992, with the following exceptions. The simulated thermal power time l constant for this function will remain in a licensee controlled document. The iTS l Surveillance Requirement for weekly adjustment of the channel to conform to a '

l calibrated flow signalis replaced by a new Surveillance Requirement (TS SR 3.3.1.1.18) l described below. The APRM recirculation flow transmitters are excluded from the semi-l annual channel calibration by addition of Note 3 to current TS SR 3.3.1.10 and a new SR (TS SR 3.3.1.1.17)is added requiring calibration of these transmitters on an 18 month frequency. The combination of the drift exhibited by these transmitters on an 18-month calibration surveillance frequency with the drift in the APRM recirculation flow channel is less than the allowed drift value identified in NEDO-32339-A, Revision 1, " Reactor Stability Long-Term Solution: Enhanced Option I-A."

l Current TS 5.6.5 lists the specifications for which limits are included in the Core Operating Limits Report. NEDO-32339-A, Supplement 4, Revision 1 places the Allowable Value of the APRM flow biased scram function in the COLR. This placement facilitates the revision of these values as it becomes necessary to update them due to changes in core or fuel designs. Accordingly, the proposed change places the APRM flow biased scram (Table 3.3.1.1-1 Function 2.d) Allowable Value (expressed as a l function of reactor recirculation drive flow)in the COLR. The Function 2.d Allowable Value is referenced by a footnote (b) to Table 3.3.1.1-1, which states that the Allowable Values are specified in the COLR.

l The initial application of the flow mapping methodology described in NEDO-32339-A, Supplement 3, Revision 1, uses plant specific historical operating data to establish the initial relationship between core flow and drive flow.

i l

l L_________________-__________-- - - - - - - - - - - - - - -

u <

l -

L Attachm2nt 2 to GNRO-98/00053 Pagn 7 of 14 After an initial flow alignment process that accommodates potential variations in the drive flow to core flow relationship from that assumed in the initial plant specific application of the flow mapping methodology, only periodic adjustment to the digital components of the FCTR is required. The use of digital components in the E1 A FCTR card also allows the incorporation of self-test features which allows more frequent internal checks. Also, l

- digital components, such as those used in the E1A FCTR card, are highly reliable, and ,

. therefore in combination with the self-test capability, less frequent external checks are necessary. For these reasons, a change is made to add a note (Note 4 to current GGNS TS SR 3.3.1.1.10) to the Channel Calibration Surveillance Requirement for Reactor Protection System Function 2.d which states the digital components of the FCTR card are excluded from the Channel Calibration. The proposed change provides a new TS Surveillance Requirement that is applicable to Reactor Protection System Function 2.d

j. and requires the adjustment of the channel to conform to core flow once within 7 days L after reaching equilibrium conditions following each refueling outage. As a result, a new L

SR (TS SR 3.3.1.1.18) is added.

L Period Based Detection Elimination of the operator actions identified in the BWROG ICAs to monitor neutron flux noise levels is justified based on the operation of a stability detection system. This

, system monitors individual LPRM signals for evidence of an approach to and I

development of neutronic/ thermal-hydraulic instability during operation in a region of the power and flow operating domain that is potentially susceptible to oscillations.

, The Period Based Detection System (PBDS) is a required feature of the E1A solution.

The PBDS uses the neutron flux oscillation period confirmation process of the Period

' Based Algorithm (PBA) described in NEDO-31960-A, "BWR Owners' Group Long-Term Stability Solutions Licensing Methodology" and NEDO-31960-A, Supplement 1. . The E1 A Period Based Algorithm (PBA-E1 A) does not include the amplitude component of -

the PBA because the amplitude is not required to detect the approach to instability. The PBA-E1A is documented in NEDC-32339P-A, Supplement 2 Revision 1. The PBDS has no safety function and is not credited during any design basis accident or transient analysis. However, during operation in regions of the operating domain potentially susceptible to instability under any operating conditions, the PBDS provides an indication that conditions consistent with a significant degradation in the stability performance of the reactor has occurred and the potential for imminent onset of neutronic/ thermal-hydraulic instability may exist. Therefore, a new PBDS Specification (GGNS TS 3.3.1.3) is appropriately added to the Instrumentation section of the TSs.

The requirements of the E1A solution PBDS Specification includes manual reactor scram without delay upon receipt of any valid PBDS channel High-High Decay Ratio (Hi-Hi DR) alarm while operating in regions of the power and flow operating domain potentially susceptible to neutronic/ thermal-hydraulic instability, defined by the Monitored Region boundary. Verification that the Hi-Hi DR alarm is valid may be performed in the' control room prior to the manual reactor' scram. This verification is to be completed without delay using another output from a PBDS card, observable from the reactor controls, or confirmation that the plant is not operating in regions of the power and flow operating domain potentially susceptible to neutronic/ thermal-hydraulic instability. This feature of the E1 A solution may result in a potentialincrease in the incidence of manual reactor scram.

Attichm:nt 2 to GNRO-98/00053 Paga 8 of 14-However, current TS 3.4.1 includes a specific requirement to respond to increased neutron flux noise levels (a characteristic of power oscillations induced by neutronic/ thermal-hydraulic instability) by performing a manual reactor scram.

Therefore, it is concluded that implementation of the requirement to insert a manual reactor scram upon receipt of a Hi-Hi DR alarm from any Operable PBDS channel in regions of the power and flow operating domain potentially susceptible to neutronic/ thermal-hydraulic instability will not significantly increase the incidence of reactor scram.

Recirculation Loops Operating Elimination of the limits on power and flow conditions of current TS 3.4.1 and Figure 3.4.1-1, the requirement for immediate manual scram with no operating recirculation loops in Mode 1, and the upgraded guidance of the BWROG ICAs is justified based on the following attributes of the E1 A long term solution: operation in the region of the power and flow operating domain most susceptible to neutronic/ thermal-hydraulic instability is automatically excluded from the licensed operating domain (Exclusion Region); operation in the region of the power and flow operating domain potentially susceptible to neutronic/ thermal-hydraulic instability in the absence of stability controls (Restricted Region) requires implementation of a stability control prior to entry; operation in the region of the power and flow operating domain potentially susceptible to neutronic/ thermal-hydraulic instability under any operating conditions (defined by the Monitored Region boundary) requires implementation of the PBDS and associated operator actions. The stability regions are established using the NRC accepted E1 A methodology and reflect GGNS plant specific design.

The' parameters of a reactor system most important in determining stability performance are core power, core flow, core inlet enthalpy, and power distribution. Recirculation system design can impact the calculated stability performance through the coupling of the fluid in the recirculation system piping with the reactor core. For a given core power, core flow, core inlet enthalpy, and power distribution, differences in the calculated -

stability performance c.an be of some significance when large differences in the physical dimensions of the recirculation piping (i.e., jet pump configuration, recirculation pipe length and diameter and pump inertia) are assumed._ However, the E1A methodology p requires modeling the plant specific characteristics of the recirculation system design important in evaluating the stability performance of the reactor system and determining the E1 A regions. Operation with a different number of operating recirculation loops at the same core power, core flow, core inlet enthalpy, and power distribution, has only minor impact on these characteristic values. Furthermore, based on well defined regions in the core power and core flow domain, adherence to the stability control adopted for

implementation with the E1 A solution has been demonstrated, as described in Section 9 l' of NEDO-32339-A, Revision 1, to greatly reduce the sensitivity of reactor staldlity L~ performance to all other parameters. Therefore, replacement of the current ICA power and flow limits and the requirement for immediate manual scram with no operating .

recirculation loops in Mode 1 by the proposed E1 A solution is appropriate.

Attachm:nt 2 to GNRO-98/00053 Pagn 9 of 14 Core Operating Limits Report Current TS 5.6.5 is modified to indicate that the Core Operating Limits Report (COLR) contains limits associated with LCO 3.2.4, " Fraction of Core Boiling Boundary (FCBB),"

LCO 3.3.1.1, "RPS Instrumentation," and LCO 3.3.1.3, " Period Based Detection System -

' (PBDS)." These limits include the Allowable Values of the APRM flow biased scram, the Restricted Region boundary, and the Monitored Region boundary. Current TS 5.6.5 is also modified to provide the methodology for development and revision of these

~

limitations. This methodology is~provided in Licensing Topical Report NEDO-32339-A and its supplements, as referenced therein.

NRC SER Compliance In response to Section 5.0 (Plant-Specific Actions) of the NRC SER for NEDC-32339P-A, Supplement 2, Revision 1 (which required that licensees referencing NEDC-32339P-A for implementation of the E1 A long term solution provide certain information in their license amendment submittals), EOl provides the following:

1. The description of the functions of the FCTR card and the PBDS, in NEDO-32339-A, Revision 1 and NEDC-32339P-A, Supplement 2, Revision 1, are applicable to GGNS. Plant specific analysis performed for GGNS has demonstrated -

stability performance indicating potential susceptibility to neutronic/ thermal-hydraulic instability in regions of the power and flow operating domain. Therefore, the E1A solution is applicable to GGNS. Additional plant specific analysis has been performed to established appropriate setpoints for the FCTR card consistent with the methodology described in NEDO-32339-A, Revision 1 and NEDO-32339-A,

' Supplement 3, Revision 1. There are no plant specific required setpoints associated with the required features of the PBDS. Parameter values for optional features of i the PBDS 'will be established during initial installation and testing to optimize the PBDS performance as described in NEDO-32339-A, Revision 1.

2. The GGNS environmental conditions (temperature, humidity, pressure, seismic, and electromagnetic compatibility) for the areas in which the PBDS and the E1 A FCTR card will be installed have been confirmed to be enveloped by the environmental qualification values.

3. Administrative controls for manually bypassing APRM flow biased scram and control rod block functions are in place at GGNS. Comparable administrative controls will be established for the PBDS channels prior to the full implementation of the E1 A solution.

4. The only changes to the GGNS plant operators' control panels associated with the E1 A solution will be those associated with the addition of alarms and indications for the PBDS instrumentation. The E1 A long term solution modifications associated with changes to the plant operator control panels have received human factors reviews.

1

Att chm:nt 2 to GNRO-98/00053 Paga 10 of 14

References:

1. GE Service Information Letter # 380, Revision 1, "BWR Core Thermal-hydraulic Stability."

2. NEDO-32339-A, Revision 1 " Reactor Stability Long-Term Solution: Enhanced Option I-A."

3. NEDC-32339P A, Supplement 1, " Reactor Stability Long-Term Solution: Enhanced Option I-A." '

4. NEDC-32339P-A, Supplement 2, Revision 1 " Reactor Stability Long-Term Solution:

Enhanced Option I-A: Solution Design."

5. NEDO-32339-A, Supplement 3, Revision 1 " Reactor Long Term Stability Solution E1 A: Flow Mapping Methodology."

6. NEDO-32339-A, Supplement 4, Revision 1 " Reactor Stability Long Term Solution:

Enhanced Option I-A Generic TSs."

7. NEDO- 31960-A, "BWR Owners' Group Long-Term Stability Solutions Licensing i Methodology."

8. NEDO-31960-A, Supplement 1, "BWR Owners' Group Long Term Stability Solutions I Licensing Methodology." '

9. 10 CFR 50, Appendix A, General Design Criterion (GDC) 12, " Suppression of Reactor Power Oscillations."

10. IE Bulletin 88-07, " Power Oscillations in Boiling Water Reactors (BWRs)."

11.10 CFR 50, Appendix A, GDC 10," Reactor Design." )

12. IE Bulletin 88-07, Supplement 1, " Power Oscillations in Boiling Water Reactors (BWRs)."

13. Generic Letter 94-02, "Long-Term Solution and Upgrade of Interim Operating Recommendations for Thermal-Hydraulic Instabilities in BWRs."

l

14. 10 CFR 50.36.

15. NUREG-1434, " Improved BWR Technical Specifications" r

i

_. _ .. ]

Attzchmsnt 2 to GNRO-98/00053 -

Page 11 of 14 i-No Significant Hazards Consideration

.The Commission has provided standards in 10 CFR 50.92 for determining whether a significant hazards consideration exists.' A proposed amendment to an operating license for a facility

. involves no significant hazards consideration if operation of the facility in accordance with the proposed amendment would not: (1) involve a significant increase in the probability or consequences of an accident previously evaluated, (2) create the possibility of a new or different ,

kind of accident from any accident previously evaluated, or (3) involve a significant reduction in t a margin of safety. EOl has reviewed these proposed license amendment requests and

j. believes that their adoption would not involve a significant hazards consideration. The basis for

i. this determination follows.  !

f l

1. This request does not involve a significant increase in the probability or consequences of an accidentpreviously evaluated.

The proposed amendments allow the implementation of the Enhanced Option I-A (E1 A) long term solution to the neutronic/ thermal-hydraulic instability issue. Current Technical Specification (TS) restrictions on power and flow conditions, number of operating recirculation loops and .

operator actions implemented to reduce the probability of neutronic/ thermal-hydraulic instability I are eliminated and new stability requirements consistent with NEDO-32339-A, Supplement 4, Revision 1, are imposed.- These requirements include restrictions on power and flow conditions and actions associated with the modified Average Power Range Monitor (APRM) flow biased scram and control rod block functions. Required actions include adherence to the boiling boundary limit stability control prior to entry and during operation in the region of the power and flow operating domain which is potentially susceptible to neutronic/ thermal-hydraulic instability in the absence of the stability control. In addition, the proposed amendments require operator actions based upon control room indications generated by a new Period Based Detection System (PBDS). The PBDS is designed to provide alarm indication that conditions consistent with a significant degradation in the stability performance of the reactor has occurred and the potential for imminent onset of neutronic/ thermal-hydraulic instability may exist. The PBDS also provides analog indication of the highest and second highest successive period confirmation count of all of the Local Power Range Monitors (LPRMs) monitored. This provides the plant operators with continuous indication of reactor stability operating conditions.

1 The proposed amendments will permit operation in regions of the power and flow operating i domain postulated to be susceptible to'neutronic/ thermal-hydraulic instability. Operation in

{

these regions does not increase the probability of occurrence of initiators and precursors of  !

previously analyzed accidents when neutronic/ thermal-hydraulic instability is not possible. The ,

proposed amendments permit the implementation of the features of the E1 A solution which  !

prevent neutronic/ thermal-hydraulic instability including preemptive reactor scram upon entry into the regions of the power and flow operating domain most susceptible to neutronic/ thermal-

,. hydraulic instability. The E1 A solution also requires implementation of stability control prior to {

entry into a region of the power and flow operating domain which is potentially susceptible, in the  ;

absence of stability control, to neutronic/ thermal-hydraulic instability. The E1 A solution prevents l neutronic/ thermal-hydraulic instability during operation in regions of the power and flow operating domain previously excluded from operation and therefore does not significantly increase the probability of a previously analyzed accident.

i i

Attrchm nt 2 to GNRO-98/00053 Pags 12 of 14

' Operation in the regions of the power and flow operating domain excluded by current TS 3.4.1 and Figure 3.4.1-1 can occur as a result of anticipated operational occurrences The severity of

. these transients may increase in the absence of operator actions due to the potential occurrence of neutronic/ thermal-hydraulic instability as a result of operation in these regions. The proposed amendments will permit the implementation of the E1A long term solution to the stability issue.

Required features of the E1 A solution include adherence to a boiling boundary limit stability control prior to selection by the operator of APRM flow biased scram and control rod block-function " Setup" setpoints which allow operation in a region of the power and flow operating domain potentially susceptible, in the absence of the stability control, to neutronic/ thermal-hydraulic instability. Upon entry, as a result of an anticipated operational occurrence, into the region most susceptible to neutronic/ thermal-hydraulic instability, the preemptive reactor scram prevents neutronic/ thermal-hydraulic instability. Therefore, the consequences of an accident do {<

not significantly increase while operating with the stability control met.

After exiting the region requiring the stability control to be met, the setpoints can be manually reset to their normal values. Stability controls are required to be in place when setpoints are

" Setup". As a backup E1 A feature, the APRM flow biased sMpoints automatically reset to their normal values above a pre-determined flow condition. This automatic reset to the more conservative setpoints ensures that the preemptive reactor scram will prevent operation as a result of an anticipated operational occurrence into the region most susceptible to neutronic/ thermal-hydraulic instability should the operator not select the more conservative setpoints appropriate for operation following exit from the region requiring stability control.

Other required E1 A features, including the PBDS, control rod block alarms associated with entry into the region susceptible to neutronic/ thermal-hydraulic instabilities in the absence of stability controls, and required operator actions, including manual reactor scram, help ensure prevention of neutronic/ thermal-hydraulic instabilities. Therefore, the proposed amendments prevent the

occurrence of neutronic/ thermal-hydraulic instability as a consequence of an anticipated i operational occurrence and do not significantly increase the consequences of any previously analyzed accident.

2. This request does not create the possibility of a new or different kind of accident from any accidentpreviously evaluated.

The proposed amendments replace current restrictions on power and flow conditions with l alternative restrictions which permit the implementation of the E1 A long term stability solution.

The current restrictions on the power and flow conditions and operating recirculation loops in the h RUN mode do not automatically prevent the entri nb regions of the power and flow operating domain most susceptible to neutronic/ thermal-hydraulic instability and therefore the possibility of neutronic/ thermal-hydraulic instability exists in the absence of operator action. The required features of the E1A solution implement a preemptive scram upon entry into the region most susceptible to neutronic/ thermal-hydraulic instability, without operator action. The accessible operating domain allowed by the proposed amendments is a subset of the power and flow operating domain currently allowed. Current initiators and precursors of accidents and anticipated operational occurrences can not occur with new or different initial conditions as a result of this change. Additionally, there are no new event initiators or precursors of accidents and anticipated operational occurrences created by this change. Therefore, the proposed l amendments do not create the possibility of a new or different kind of accident from that previously evaluated.

_ _ _ _ _ _ _ _ __ _ ___________ ________J

- _ - _ - _ _ _ _ _ - _ _ _ - - _ - _ - _ = - _ _ _ _ - - - _ _ _ _ _ _ _ - _ _ _ _ _ - _ _ _ . _ _ _ _ _ _ _ _ - _ _ _

Attachmsnt 2 to GNRO-98/00053 Pign 13 of 14 Concurrent with the implementation of the proposed amendments, a modified Flow Control Trip l Reference (FCTR) card, the E1A FCTR card, and a new Period Based Detection System (PBDS) will be installed as required by the E1 A solution. The function of the E1 A FCTR card is to aid the operator in the identification of entry into regions of the power and flow operating domain potentially susceptible to neutronic/ thermal-hydraulic instability in the absence of stability l controls and to initiate a preemptive scram upon entry into the regions most susceptible to

- neutronic/ thermal-hydraulic instability. This is accomplished by altering the existing values of

.setpoints of the APRM flow biased scram and the control rod block functions generated by the i E1A FCTR card. The E1A FCTR card design includes components which may be susceptible to i electromagnetic interference or other environmental effects. The plant specific environmental i conditions (temperature, humidity, pressure, seismic, and electromagnetic compatibility) have l been confirmed to be enveloped by the environmental qualification values for the E1 A FCTR cards. Therefore, the potential for spurious scrams or common mode failures induced by environmental effects (e.g., electromagnetic interference) is considered negligible. The installation of the E1 A FCTR card will therefore not create the possibility of a new or different kind of accident from any accident previously evaluated. l The function of the PBDS.is to provide the operator with an indication that conditions consistent with a significant degradation in the stability performance of the reactor has occurred and the potential for imminent onset of neutronic/ thermal-hydraulic instability may exist. This is

-accomplished by the installation of a new PBDS card in the Neutron Monitoring System. The PBDS card takes inputs from individual local power range monitors and provides analog indication of the highest and second highest successive period confirmation count, provides a High Decay Ratio (Hi DR) and High-High Decay Ratio (Hi-Hi DR) alarms, and INOP status indication to the operator in the control room. These displays can not create the possibility of a l new or different kind of accident from any accident previously evaluated. The PBDS card  !

design includes components which may be susceptible to electromagnetic interference or other

- environmental effects. However, the plant specific environmental conditions (temperature, j humidity, pressure, seismic, and electromagnetic compatibility) have been confirmed to be enveloped by the PBDS environmental qualification values. Therefore, the installation of the PBDS card will not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. This request does not involve a significant reduction in a margin to safety.

The proposed amendments permit the implementation of the E1A long term solution to the stability issue. Under certain conditions, existing BWR designs are susceptible to ,

neutronic/ thermal-hydraulic instability. General Design Criterion (GDC) 12 of 10 CFR 50, '

Appendix A, requires thermal-hydraulic instability to be prevented by design or be readily and reliably detected and suppressed. When the design of the reactor system does not prevent the 1

occurrence of neutronic/ thermal-hydraulic instability, instability is an anticipated operational occurrence. GDC 10 of 10 CFR 50, Appendix A, requires that specified acceptable fuel design 4

limits not be exceeded during anticipated operational occurrences.

Analyses performed by the BWROG indicate that neutronic/ thermal-hydraulic instability induced power oscillatiomi could result in conditions exceeding the Minimum Critical Power Ratio (MCPR) Safety Limit (SL) prior to detection and suppression by the current design of the Neutron Monitoring System and Reactor Protection System.

l

Attachmsnt 2 to GNRO-98/00053 Page 14 of 14 To ensure compliance with GDC 12 the BWROG developed Interim Corrective Actions (ICAs) to enhance the capability of the operator to readily and reliably detect and suppress neutronic/ thermal-hydraulic instability. The BWROG ICAs also provided additional guidance for monitoring local power range monitors beyond the requirements of current TS 3.4.1 to ensure adequate margin to the onset of neutronic/ thermal-hydraulic instability. Reliance on operator

' actions to comply with GDC 12 was accepted on an interim basis by the NRC pending final

' implementation of a long term solution to the stability issue. Neutronic/ thermal-hydraulic instability is prevented by implementation of the E1A solution through the modified design of the Reactor Protection System (APRM flow biased scram) and the stability control prior to entry into a region of the power and flow operating domain which is potentially susceptible, in the absence of stability control, to neutronic/ thermal-hydraulic instability. In addition, significant backup protection features, including the PBDS, control rod block alarms associated with entry into the region susceptible to neutronic/ thermal-hydraulic instabilities in the absence of stability controls, and specified operator actions, including manual reactor scram, are required to be implemented.

As a result, the margin to the onset of neutronic/ thermal-hydraulic instability provided by the existing TS requirements and BWROG ICAs recommendations is not significantly reduced by the implementation of the E1A solution. The E1A solution assures compliance with GDC 12 by the prevention of neutronic/ thermal-hydraulic instability and therefore precludes neutronic/ thermal-hydraulic instability from becoming a credible consequence of an anticipated operational occurrence. The consequences of anticipated operational occurrences will not increase and the margin to the MCPR SL will not decrease upon implementation of the E1A solution. Therefore, the proposed amendments do not involve a significant reduction in a margin of safety.

i

'