NLS2024025, License Amendment Request to Revise Technical Specifications Table 3.3.2.1-1, Control Rod Block Instrumentation: Difference between revisions

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{{#Wiki_filter:H Nebraska       Public   Power District "Always there when you   need us"
{{#Wiki_filter:H Nebraska Public Power District "Always there when you need us"


50.90 NLS2024025 May   9,   2024
50.90 NLS2024025 May 9, 2024


Attention:       Document       Control     Desk U.S. Nuclear     Regulatory       Commission Washington,         D.C. 20555-0001
Attention: Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001


==Subject:==
==Subject:==
License   Amendment         Request       to   Revise   Technical         Specifications       Table     3.3.2.1-1, "Control       Rod   Block   Instrumentation" Cooper   Nuclear     Station,     Docket   No. 50-298,   Renewed       License   No. DPR-46
License Amendment Request to Revise Technical Specifications Table 3.3.2.1-1, "Control Rod Block Instrumentation" Cooper Nuclear Station, Docket No. 50-298, Renewed License No. DPR-46


==Dear   Sir or Madam:==
==Dear Sir or Madam:==
Pursuant to 10 CFR 50.90, "Application for amendment oflicense, construction permit, or early site permit," Nebraska Public Power District (NPPD) is submitting a request for an amendment to Renewed License No. DPR-46 for Cooper Nuclear Station (CNS). The proposed change would relocate cycle specific Minimum Critical Power Ratio values from the CNS Technical Specifications (TS) Table 3.3.2.1-1 to the CNS Core Operating Limits Report.
to this letter provides a description and assessment of the proposed change. provides the existing TS pages marked up to show the proposed change. provides revised (clean) TS pages. The changes to the TS Bases are provided for information only in Attachment 4 and will be incorporated upon implementation of the approved amendment.


Pursuant      to      10  CFR    50.90,   "Application          for  amendment     oflicense,                 construction      permit,     or early site permit,"        Nebraska      Public    Power    District      (NPPD)    is  submitting        a request      for  an amendment to   Renewed        License      No. DPR-46      for  Cooper    Nuclear    Station    (CNS).                The proposed      change would    relocate        cycle  specific    Minimum        Critical    Power  Ratio    values      from  the CNS                                                                      Technical Specifications          (TS)  Table    3.3.2.1-1      to  the  CNS    Core Operating                                                                                                        Limits      Report.
Approval of the proposed amendment is requested by September 30, 2024. Once approved, the amendment shall be implemented by October 30, 2024, to allow plant operation after the completion of a refueling outage.
to  this  letter  provides        a description        and  assessment      of the  proposed      change. provides    the  existing      TS  pages    marked    up  to    show  the  proposed      change. provides      revised      (clean)    TS  pages.                The  changes    to  the  TS  Bases    are provided for information        only  in Attachment        4  and  will  be incorporated      upon    implementation        of the    approved amendment.


Approval      of the proposed       amendment        is  requested    by  September        30,  2024.              Once  approved,      the amendment          shall  be implemented        by  October      30,  2024,    to   allow  plant    operation      after  the completion        of a refueling      outage.
NPPD has determined that there are no significant hazards considerations associated with the proposed change and that the TS change qualifies for a categorical exclusion from environmental review pursuant to the provisions of 10 CFR 51.22( c )(9).


NPPD    has   determined that          there    are  no    significant    hazards      considerations          associated      with  the proposed      change      and that the                                                                                        TS    change    qualifies      for   a categorical        exclusion      from    environmental review    pursuant        to   the provisions        of 10 CFR   51.22( c )(9).
The proposed TS change has been reviewed by the necessary safety review committees (Station Operations Review Committee and Safety Review and Audit Board). In accordance with 10 CFR 50.91, "Notice for public comment; State consultation," a copy of this application, with attachments, is being provided to the designated State of Nebraska Official. Copies to the Nuclear Regulatory Commission Region IV office and the CNS Resident Inspectors are also being provided in accordance with 10 CFR 50.4(b)(l).


The  proposed        TS    change    has  been  reviewed      by the  necessary        safety  review      committees        (Station Operations        Review      Committee        and    Safety    Review    and  Audit    Board).                   In  accordance      with      10 CFR  50.91,       "Notice      for public      comment;        State  consultation,"          a  copy  of this    application,      with attachments,          is  being  provided      to  the  designated        State  of Nebraska        Official.                Copies    to  the Nuclear    Regulatory          Commission        Region      IV  office    and  the  CNS    Resident      Inspectors      are    also being    provided      in  accordance      with      10  CFR    50.4(b)(l).
COOPER NUCLEAR STATION 72676 648A Ave/ P.O. Box 98 / Brownville, NE 68321 http://www.nppd.com NLS2024025 Page 2 of2


COOPER      NUCLEAR    STATION 72676  648A  Ave/    P.O. Box 98 /  Brownville,     NE  68321 http://www.nppd.com NLS2024025 Page    2  of2
There are no regulatory commitments made in this submittal. If you should have any questions regarding this submittal, please contact Linda Dewhirst, Regulatory Affairs and Compliance Manager, at (402) 825-5416.


There      are  no  regulatory            commitments          made    in  this    submittal.                  If you    should      have      any  questions regarding          this      submittal,        please      contact      Linda      Dewhirst,          Regulatory        Affairs        and   Compliance Manager,          at  (402)      825-5416.
I declare under penalty of perjury that the foregoing is true and correct.


I declare      under    penalty      of perjury      that  the    foregoing          is  true    and  correct.
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Attachments:                                             1.                         Description           and   Assessment
Attachments: 1. Description and Assessment
: 2.                           Proposed       Technical           Specifications             Change       (Mark-up)
: 2. Proposed Technical Specifications Change (Mark-up)
: 3.                         Revised       Technical           Specifications             Pages
: 3. Revised Technical Specifications Pages
: 4.                         Proposed         Technical         Specifications             Bases     Changes       (Mark-up)
: 4. Proposed Technical Specifications Bases Changes (Mark-up)


cc:                                                                         Regional         Administrator             w/ attachments USNRC         -           Region       IV
cc: Regional Administrator w/ attachments USNRC - Region IV


Cooper     Project       Manager       w/ attachments USNRC         -         NRR   Plant     Licensing       Branch       IV
Cooper Project Manager w/ attachments USNRC - NRR Plant Licensing Branch IV


Senior     Resident         Inspector       w/ attachments USNRC-CNS
Senior Resident Inspector w/ attachments USNRC-CNS


Nebraska         Health       and   Human       Services     w/ attachments Department           of Regulation           and   Li censure
Nebraska Health and Human Services w/ attachments Department of Regulation and Li censure


NPG   Distribution           w/ attachments
NPG Distribution w/ attachments


CNS     Records       w /   attachments NLS2024025 Page     1 of 6
CNS Records w / attachments NLS2024025 Page 1 of 6


Attachment           1
Attachment 1


Description       and Assessment
Description and Assessment


Cooper Nuclear     Station,   Docket No. 50-298,     Renewed   License No. DPR-46
Cooper Nuclear Station, Docket No. 50-298, Renewed License No. DPR-46


1.0                                                       Summary Description
1.0 Summary Description


2.0                                                       Detailed   Description
2.0 Detailed Description


3.0                                                       Technical   Evaluation
3.0 Technical Evaluation


4.0                                                         Regulatory Analysis
4.0 Regulatory Analysis


4.1                                                           Applicable Regulatory                                                                                                                                                                                                                         Requirements/Criteria 4.2                                                       Precedent 4.3                                                                   No   Significant Hazards                                                                                                                                                                                                                           Consideration     Determination     Analysis 4.4                                                       Conclusion
4.1 Applicable Regulatory Requirements/Criteria 4.2 Precedent 4.3 No Significant Hazards Consideration Determination Analysis 4.4 Conclusion


5.0                                                       Environmental     Evaluation NLS2024025 Page     2 of6
5.0 Environmental Evaluation NLS2024025 Page 2 of6


1.0                                                  
1.0  


==SUMMARY==
==SUMMARY==
DESCRIPTION
DESCRIPTION


In   accordance       with       10 CFR   50.90,       "Application           for amendment         oflicense,                   construction       permit, or early   site   permit,"       Nebraska       Public     Power     District     (NPPD)       is   submitting         a request     for   an amendment         to   Renewed       License     No. DPR-46       for   Cooper     Nuclear       Station     (CNS).               The proposed change     would     amend   the     CNS   Technical         Specifications           (TS)   to   modify       TS   Table   3.3.2.1-1, "Control       Rod   Block     Instrumentation."             The proposed       change     would     relocate       cycle   specific Minimum         Critical     Power     Ratio     (MCPR)       values     to   the   CNS     Core   Operating         Limits   Report (COLR).
In accordance with 10 CFR 50.90, "Application for amendment oflicense, construction permit, or early site permit," Nebraska Public Power District (NPPD) is submitting a request for an amendment to Renewed License No. DPR-46 for Cooper Nuclear Station (CNS). The proposed change would amend the CNS Technical Specifications (TS) to modify TS Table 3.3.2.1-1, "Control Rod Block Instrumentation." The proposed change would relocate cycle specific Minimum Critical Power Ratio (MCPR) values to the CNS Core Operating Limits Report (COLR).


2.0                                                     DETAILED       DESCRIPTION
2.0 DETAILED DESCRIPTION


NPPD     proposes       the   following       changes       to   the   CNS   TS:
NPPD proposes the following changes to the CNS TS:
: 1. Revise     the   notes     associated       with   TS   Table     3.3.2.1-1,         "Control       Rod   Block Instrumentation,"                   as   shown     below:
: 1. Revise the notes associated with TS Table 3.3.2.1-1, "Control Rod Block Instrumentation," as shown below:
(a)       THERMAL         POWER             2: 27.5%     and<     62.5%     RTP     and MCPR ~                           less than the limit specified     in   the   COLR   and no   peripheral           control     rod   selected.
(a) THERMAL POWER 2: 27.5% and< 62.5% RTP and MCPR ~ less than the limit specified in the COLR and no peripheral control rod selected.
( d)       THERMAL         POWER             2: 62.5%     and < 82.5% RTP                                                                                                                               and   MCPR ~ less than the limit specified     in   the   COLR   and   no   peripheral           control     rod   selected.
( d) THERMAL POWER 2: 62.5% and < 82.5% RTP and MCPR ~ less than the limit specified in the COLR and no peripheral control rod selected.
(e)       THERMAL         POWER             2: 82.5%   and<     90%   RTP     and   MCPR ~                           less   than     the limit specified     in   the   COLR   and   no   peripheral           control     rod   selected.
(e) THERMAL POWER 2: 82.5% and< 90% RTP and MCPR ~ less than the limit specified in the COLR and no peripheral control rod selected.
(t)             THERMAL         POWER       2: 90%   RTP     and MCPR     <     1 .40 less than   the limit specified       in   the   COLR   and no   peripheral       control     rod   selected.
(t) THERMAL POWER 2: 90% RTP and MCPR < 1.40 less than the limit specified in the COLR and no peripheral control rod selected.
(g)     THERMAL         POWER               2: 27.5%     and<     90%   RTP     and   MCPR ~           less than     the limit specified     in   the   COLR   and   no peripheral           control     rod   selected.
(g) THERMAL POWER 2: 27.5% and< 90% RTP and MCPR ~ less than the limit specified in the COLR and no peripheral control rod selected.
: 2. Revise     TS     5.6.5,     CORE     OPERATING           LIMITS     REPORT         (COLR)       as   shown   below:
: 2. Revise TS 5.6.5, CORE OPERATING LIMITS REPORT (COLR) as shown below:
: a.             Core   operating       limits     shall   be established       prior     to   each   reload     cycle,     or prior     to any remaining       portion     of a reload       cycle,     and   shall   be documented       in the   COLR for   the   following:
: a. Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:
: 1.                                                     The   Average       Planar     Linear     Heat   Generation         Rates     for   Specifications 3.2.1     and   3.7.7.
: 1. The Average Planar Linear Heat Generation Rates for Specifications 3.2.1 and 3.7.7.
: 2.                                                       The   Minimum         Critical       Power     Ratio     for     Specifications           3.2.2   and   3.7.7, and MCPR99.9% for   Specification         3.2.2.
: 2. The Minimum Critical Power Ratio for Specifications 3.2.2 and 3.7.7, and MCPR99.9% for Specification 3.2.2.
: 3.                                                       The   Linear     Heat Generation                                                                                                           Rates     for   Specifications             3 .2.3   and   3. 7. 7.
: 3. The Linear Heat Generation Rates for Specifications 3.2.3 and 3. 7. 7.
: 4.                                                     The   three     Rod   Block     Monitor     Upscale     Allowable         Values     for Specification         3.3.2.1.
: 4. The three Rod Block Monitor Upscale Allowable Values for Specification 3.3.2.1.
: 5.                                                       The power/flow         map   defining       the   Stability       Exclusion       Region     for Specification           3 .4.1.
: 5. The power/flow map defining the Stability Exclusion Region for Specification 3.4.1.
: 6.                                                                         The Minimum         Critical   Power   Ratios     in       Table 3.3.2.1-1 for Specification       3.3.2.1.
: 6. The Minimum Critical Power Ratios in Table 3.3.2.1-1 for Specification 3.3.2.1.


Relocation         of the   cycle     specific     MCPRs       to   the   COLR,     which   is   controlled       by TS     5.6.5,   would provide       NPPD     the   flexibility       to   revise     cycle     specific     MCPRs     in accordance       with   Nuclear Regulatory           Commission           (NRC)     approved     methodologies             without       the need for                                                                                                         a license NLS2024025 Page     3   of 6
Relocation of the cycle specific MCPRs to the COLR, which is controlled by TS 5.6.5, would provide NPPD the flexibility to revise cycle specific MCPRs in accordance with Nuclear Regulatory Commission (NRC) approved methodologies without the need for a license NLS2024025 Page 3 of 6


amendment.           The     COLR,       including         any   mid-cycle           revisions         or supplements,               is required       to   be submitted           to   the   NRC     for   each   reload       cycle     per TS     5.6.5.
amendment. The COLR, including any mid-cycle revisions or supplements, is required to be submitted to the NRC for each reload cycle per TS 5.6.5.


3.0                                                   TECHNICAL         EVALUATION
3.0 TECHNICAL EVALUATION


NRC     Generic         Letter       (GL)     88-16,       "Removal         of Cycle-Specific             Parameter           Limits     From     Technical Specifications,"                 provides         guidance         to   licensees             for   the removal         of cycle     dependent       parameter limits       from     the   TS   provided       these     values       are   included       in a COLR     and   are   determined         with   NRC approved         methodologies               referenced           in the   TS.             The   specific     values       of these     limits     may be modified         by licensees,         without       affecting         nuclear         safety,     provided         that   these     changes are determined           using     an NRC     approved       methodology               and   consistent         with       all   applicable         limits     of the plant     safety       analysis         that are                                                                             addressed         in the   Final       Safety     Analysis           Report.       If any   of the applicable           limits     of the   safety   analysis         are   not met,     prior   NRC     approval         of the   change     is required.
NRC Generic Letter (GL) 88-16, "Removal of Cycle-Specific Parameter Limits From Technical Specifications," provides guidance to licensees for the removal of cycle dependent parameter limits from the TS provided these values are included in a COLR and are determined with NRC approved methodologies referenced in the TS. The specific values of these limits may be modified by licensees, without affecting nuclear safety, provided that these changes are determined using an NRC approved methodology and consistent with all applicable limits of the plant safety analysis that are addressed in the Final Safety Analysis Report. If any of the applicable limits of the safety analysis are not met, prior NRC approval of the change is required.


Control       Rod   Block       Instrumentation             MCPR     values       are   calculated           as part   of the   reload       core   design licensing           analyses         in accordance         with   NRC     approved       methods         in NEDE-24011-P-A,                     General Electric       Standard         Application             for   Reactor         Fuel.                 Because     the     analyses         are   completed         only   two   to three   months         prior     to   the   start   of the   next     refueling           outage,       a need     to   revise     MCPR     values       for   the next   operating           cycle     would     require       submitting             a license       amendment           request       with     a quick turnaround,             placing         an unnecessary           burden       on NPPD       and NRC   resources.           Relocating         the   MCPR values       to   the   COLR     will     allow     NPPD       to make       cycle-specific             changes         that   are   consistent         with NRC     approved         methodologies               and   within         limits     of the   safety     analysis         without       the burdensome process       of amending         the   TS. The   TS   will     continue         to   establish         limits       for MCPR     for the   Control Rod   Block       Instrumentation while                     the   specific       values       for   MCPR     are   relocated           to   the   COLR.
Control Rod Block Instrumentation MCPR values are calculated as part of the reload core design licensing analyses in accordance with NRC approved methods in NEDE-24011-P-A, General Electric Standard Application for Reactor Fuel. Because the analyses are completed only two to three months prior to the start of the next refueling outage, a need to revise MCPR values for the next operating cycle would require submitting a license amendment request with a quick turnaround, placing an unnecessary burden on NPPD and NRC resources. Relocating the MCPR values to the COLR will allow NPPD to make cycle-specific changes that are consistent with NRC approved methodologies and within limits of the safety analysis without the burdensome process of amending the TS. The TS will continue to establish limits for MCPR for the Control Rod Block Instrumentation while the specific values for MCPR are relocated to the COLR.
Therefore,           the   requested           changes       are   essentially           administrative             in nature,       and   the required         level of safety     will     be maintained.
Therefore, the requested changes are essentially administrative in nature, and the required level of safety will be maintained.


4.0                                                   REGULATORY         ANALYSIS
4.0 REGULATORY ANALYSIS


4.1                                                     Applicable             Regulatory         Requirements/Criteria
4.1 Applicable Regulatory Requirements/Criteria


NRC     GL   88-16     discusses         that   processing             TS     changes       to   update       cycle-specific             parameter           limits each   fuel     cycle     places         an unnecessary           burden       on the   licensee         and   the   NRC   if these     limits       are developed         using       an NRC     approved       methodology.                         The   GL   provides         an alternative         that   relocates the   specific       parameter           values     to   the   COLR     provided       the   values       are   determined           using     an NRC approved       methodology                 and   the   TS   require       plant     operation         in accordance           with   the   limits       specified in the   COLR.
NRC GL 88-16 discusses that processing TS changes to update cycle-specific parameter limits each fuel cycle places an unnecessary burden on the licensee and the NRC if these limits are developed using an NRC approved methodology. The GL provides an alternative that relocates the specific parameter values to the COLR provided the values are determined using an NRC approved methodology and the TS require plant operation in accordance with the limits specified in the COLR.


10 CFR   50.36,       Technical           specifications               -         requires         that   the   TS     contain       limiting         conditions           for operation,           which       are   the   lowest       functional             capability         or performance             levels       of equipment required for                                                                                                                                                             safe   operation         of the   facility.
10 CFR 50.36, Technical specifications - requires that the TS contain limiting conditions for operation, which are the lowest functional capability or performance levels of equipment required for safe operation of the facility.


The proposed             changes         are   consistent         with     the   above     regulatory           guidance         and   regulation.
The proposed changes are consistent with the above regulatory guidance and regulation.
NLS2024025 Page   4 of 6
NLS2024025 Page 4 of 6


4.2                                                     Precedent
4.2 Precedent


In March   2018,     Duane   Arnold     Energy     Center   received       approval     to relocate     the   Control     Rod Block   Instrumentation         MCPR     values     to the COLR                                                                       (ML18011A059).                     Also     in January     2014, Columbia       Generating       Station   received       approval     to relocate     the Control                                                                     Rod   Block Instrumentation         MCPR   values     to   the COLR                                                                                 as part   of a change   to   implement         a digital instrumentation           system   (ML133 l 7B623).
In March 2018, Duane Arnold Energy Center received approval to relocate the Control Rod Block Instrumentation MCPR values to the COLR (ML18011A059). Also in January 2014, Columbia Generating Station received approval to relocate the Control Rod Block Instrumentation MCPR values to the COLR as part of a change to implement a digital instrumentation system (ML133 l 7B623).


The TS     for the   plants   below     contain     notes   comparable         to   those   in the   CNS   TS   that   establish limits   on MCPR     for the   Control     Rod   Block   Instrumentation.             Similar   to the proposed       change   to the   CNS   TS,   the   TS   below   do   not provide   the   specific     value     for the   Control     Rod   Block Instrumentation         MCPR   but   specify   the limit   for MCPR     as     "less than   the limit     specified     in the COLR."
The TS for the plants below contain notes comparable to those in the CNS TS that establish limits on MCPR for the Control Rod Block Instrumentation. Similar to the proposed change to the CNS TS, the TS below do not provide the specific value for the Control Rod Block Instrumentation MCPR but specify the limit for MCPR as "less than the limit specified in the COLR."
* Brunswick       Unit     1 (ML062900525)           and   Unit   2 (ML062900536)
* Brunswick Unit 1 (ML062900525) and Unit 2 (ML062900536)
* Browns     Ferry   Unit   2 (ML052780020)
* Browns Ferry Unit 2 (ML052780020)
* Peach   Bottom     Unit   2 (ML052720266)
* Peach Bottom Unit 2 (ML052720266)
* Susquehanna       Unit       1 (ML052720300)         and   Unit   2   (ML052720301)
* Susquehanna Unit 1 (ML052720300) and Unit 2 (ML052720301)


4.3                                                     No   Significant       Hazards       Consideration       Determination           Analysis
4.3 No Significant Hazards Consideration Determination Analysis


Nebraska     Public     Power   District     (NPPD)     requests       an amendment       to the Cooper                                                                       Nuclear     Station (CNS)   Technical         Specifications         (TS)   to modify     TS   Table     3.3.2.1-1,       "Control     Rod   Block Instrumentation."           The proposed       change   would   relocate       cycle     specific   Minimum         Critical     Power Ratio   (MCPR)       values     to   the CNS                                                                       Core   Operating       Limits     Report   (COLR).
Nebraska Public Power District (NPPD) requests an amendment to the Cooper Nuclear Station (CNS) Technical Specifications (TS) to modify TS Table 3.3.2.1-1, "Control Rod Block Instrumentation." The proposed change would relocate cycle specific Minimum Critical Power Ratio (MCPR) values to the CNS Core Operating Limits Report (COLR).


As required     by   10 CFR   50.91(a),     NPPD   has   evaluated       the proposed     change     to   the CNS                                                                                 TS   using the criteria   in     10 CFR   50.92     and   determined       that   the   proposed       change   does   not   involve     a significant     hazards       consideration.         An analysis   of the   issue   of no   significant     hazards       consideration is presented       below.
As required by 10 CFR 50.91(a), NPPD has evaluated the proposed change to the CNS TS using the criteria in 10 CFR 50.92 and determined that the proposed change does not involve a significant hazards consideration. An analysis of the issue of no significant hazards consideration is presented below.
: 1.         Does   the proposed       change   involve     a significant       increase     in the probability         or consequences         of an accident     previously       evaluated?
: 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?


Response:       No
Response: No


The proposed         change   is   an administrative         change   that   does   not   affect   any plant     systems, structures,       or components       designed       for the   prevention       or mitigation       of previously evaluated       accidents.       No   new   equipment     is   added   nor is   installed     equipment       being     changed or operated       in a different     manner.                 Relocation       of the   Control     Rod   Block     Instrumentation MCPR     values     to   the   COLR   has   no   influence     or impact     on, nor does   it contribute       in any way   to   the probability         or consequences       of transients       or accidents.     The   COLR   will   continue to be controlled     by the   CNS   programs       and procedures         that   comply   with   TS   5.6.5.
The proposed change is an administrative change that does not affect any plant systems, structures, or components designed for the prevention or mitigation of previously evaluated accidents. No new equipment is added nor is installed equipment being changed or operated in a different manner. Relocation of the Control Rod Block Instrumentation MCPR values to the COLR has no influence or impact on, nor does it contribute in any way to the probability or consequences of transients or accidents. The COLR will continue to be controlled by the CNS programs and procedures that comply with TS 5.6.5.
Transient       analyses     addressed       in the Updated       Safety   Analysis       Report   will     continue     to be performed       in the   same manner     with respect       to   changes     in the   cycle   dependent       parameters obtained       from   the use   of Nuclear     Regulatory         Commission         (NRC)     approved     reload     design NLS2024025 Page   5 of 6
Transient analyses addressed in the Updated Safety Analysis Report will continue to be performed in the same manner with respect to changes in the cycle dependent parameters obtained from the use of Nuclear Regulatory Commission (NRC) approved reload design NLS2024025 Page 5 of 6


methodologies,         which     ensures     that the transient       evaluation     of new reloads       are bounded     by previously       accepted       analyses.
methodologies, which ensures that the transient evaluation of new reloads are bounded by previously accepted analyses.


Therefore,     the proposed         TS   change   does   not   involve       an increase     in the probability       or consequences       of an   accident     previously       evaluated.
Therefore, the proposed TS change does not involve an increase in the probability or consequences of an accident previously evaluated.
: 2.         Does   the proposed         change     create   the possibility       of a new or different     kind   of accident from   any accident     previously         evaluated?
: 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?


Response:     No
Response: No


The proposed       administrative           change   does   not   involve     any changes     to   the   operation, testing,     or maintenance         of any safety-related,       or otherwise     important     to   safety     systems.
The proposed administrative change does not involve any changes to the operation, testing, or maintenance of any safety-related, or otherwise important to safety systems.
Systems   important       to   safety will   continue     to be operated       and maintained       within     their design   bases.     Relocation         of the Control     Rod   Block     Instrumentation       MCPR     values     to the COLR has   no   influence       or impact   on new   or different     kind   of accidents.
Systems important to safety will continue to be operated and maintained within their design bases. Relocation of the Control Rod Block Instrumentation MCPR values to the COLR has no influence or impact on new or different kind of accidents.


Therefore,     the proposed         change   does   not   create   the possibility     of a new   or different   kind of accident     from     any   accident   previously       evaluated.
Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
: 3.         Does   the proposed         change     involve   a significant       reduction     in a margin   of safety?
: 3. Does the proposed change involve a significant reduction in a margin of safety?


Response:     No
Response: No


The margin     of safety     is not   affected   by the relocation       of cycle-specific         Control     Rod   Block Instrumentation         MCPR     values     from the TS   to the   COLR. Appropriate       measures         exist to control   the values     of these     cycle-specific       limits       since   it is required     by TS   that   only NRC approved   methods       be used   to   determine     the   limits.                 The proposed       change     continues     to require   operation       within     the   core thermal     limits as                                                                                                                       obtained     from NRC     approved     reload design   methodologies               and the actions                                                                             to be taken   if a limit   is   exceeded   remain     unchanged in accordance       with   existing     TS.
The margin of safety is not affected by the relocation of cycle-specific Control Rod Block Instrumentation MCPR values from the TS to the COLR. Appropriate measures exist to control the values of these cycle-specific limits since it is required by TS that only NRC approved methods be used to determine the limits. The proposed change continues to require operation within the core thermal limits as obtained from NRC approved reload design methodologies and the actions to be taken if a limit is exceeded remain unchanged in accordance with existing TS.


Therefore,     the proposed         change has   no   impact     to   the margin   of safety.
Therefore, the proposed change has no impact to the margin of safety.


Based   on the above,     NPPD       concludes     that the proposed         change   presents     no   significant       hazards consideration       under   the   standards       set forth   in     10 CFR   50.92,     and,   accordingly,         a finding     of "no significant     hazards     consideration"             is justified.
Based on the above, NPPD concludes that the proposed change presents no significant hazards consideration under the standards set forth in 10 CFR 50.92, and, accordingly, a finding of "no significant hazards consideration" is justified.


4.4                                                                       Conclusion
4.4 Conclusion


Based   on the considerations             discussed     above,     (1)   there     is   reasonable       assurance     that the health and   safety   of the   public     will     not be endangered     by operation       in the proposed     manner,       (2)   such activities     will   be conducted         in compliance     with the   Commission's         regulations,       and   (3)   the issuance     of the   amendment           will   not be inimical     to the   common     defense     and   security   or to   the health     and   safety   of the public.
Based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
NLS2024025 Page     6 of6
NLS2024025 Page 6 of6


5.0                                                     ENVIRONMENTAL           EVALUATION
5.0 ENVIRONMENTAL EVALUATION


NPPD     has   determined       that   the proposed       amendment       would   change     a requirement       with respect       to installation       or use of a facility     component         located     within   the restricted         area,     as defined     in   10 CFR 20,   or would   change     an inspection       or surveillance       requirement.                     However,       the proposed amendment       does   not   involve     (i)   a significant       hazards     consideration,         (ii)   a significant     change   in the   types   or significant       increase     in the amounts                                                               of any effluents that                                                                                                                                                           may be released     offsite,     or (iii)     a significant     increase     in individual       or cumulative       occupational         radiation       exposure.
NPPD has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure.
Accordingly,         the proposed       amendment       meets   the eligibility                                                                         criterion       for   categorical     exclusion         set forth     in     10 CFR   51.22( c )(9).               Therefore,       pursuant       to       10 CFR   51.22(b ),   no   environmental       impact statement     or environmental           assessment       need   be prepared     in connection       with   the proposed amendment.
Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22( c )(9). Therefore, pursuant to 10 CFR 51.22(b ), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.
NLS2024025 Page     1 of3
NLS2024025 Page 1 of3


Attachment         2
Attachment 2


Proposed     Technical         Specifications         Changes       (Mark-up)
Proposed Technical Specifications Changes (Mark-up)


Cooper     Nuclear       Station,     Docket     No. 50-298,   Renewed License                                                                                                                                                                                                                 No. DPR-46
Cooper Nuclear Station, Docket No. 50-298, Renewed License No. DPR-46


Marked     Up Pages
Marked Up Pages


3.3-19 5.0-21 Control     Rod   Block       Instrumentation 3.3.2.1
3.3-19 5.0-21 Control Rod Block Instrumentation 3.3.2.1


Table3.3.2.1-1               (page       1 of 1)
Table3.3.2.1-1 (page 1 of 1)
Control     Rod   Block     Instrumentation
Control Rod Block Instrumentation


APPLICABLE MODES   OR OTHER SPECIFIED                                   REQUIRED                                 SURVEILLANCE                                       ALLOWABLE FUNCTION                                                                   CONDITIONS                                     CHANNELS                               REQUIREMENTS                                               VALUE
APPLICABLE MODES OR OTHER SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS CHANNELS REQUIREMENTS VALUE
: 1.                     Rod     Block     Monitor
: 1. Rod Block Monitor
: a.                     Low   Power         Range   -       Upscale                                                                                                                                                                                                                                                         (a) 2 SR         3.3.2.1.1 U)
: a. Low Power Range - Upscale (a) 2 SR 3.3.2.1.1 U)
SR         3.3.2.1.4 SR       3.3.2.1.5(b)(c)
SR 3.3.2.1.4 SR 3.3.2.1.5(b)(c)
: b.                       Intermediate         Power     Range   -       Upscale                                                                                                                                                                                                             (d) 2 SR         3.3.2.1.1 U)
: b. Intermediate Power Range - Upscale (d) 2 SR 3.3.2.1.1 U)
SR       3.3.2.1.4 SR       3.3.2.1.5(b)(c)
SR 3.3.2.1.4 SR 3.3.2.1.5(b)(c)
: c.                       High   Power     Range   -       Upscale                                                                                                                                                                                                                                                       ( e),(f) 2 SR         3.3.2.1.1 U)
: c. High Power Range - Upscale ( e),(f) 2 SR 3.3.2.1.1 U)
SR         3.3.2.1.4 SR       3.3.2.1.5(b)(c)
SR 3.3.2.1.4 SR 3.3.2.1.5(b)(c)
: d.                     lnop                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                       (f),(g) 2 SR         3.3.2.1.1 NA
: d. lnop (f),(g) 2 SR 3.3.2.1.1 NA
: e.                     Downscale                                                                                                                                                                                                                                                                                                                                                                                                                                                                           (f),(g) 2 SR         3.3.2.1.1 ~ 92/125   divisions SR         3.3.2.1.5                           of full   scale
: e. Downscale (f),(g) 2 SR 3.3.2.1.1 ~ 92/125 divisions SR 3.3.2.1.5 of full scale
: 2.                     Rod   Worth     Minimizer                                                                         1(h),2(h)                                                                     SR         3.3.2.1.2                             NA SR         3.3.2.1.3 SR         3.3.2.1.6 SR         3.3.2.1.8
: 2. Rod Worth Minimizer 1(h),2(h) SR 3.3.2.1.2 NA SR 3.3.2.1.3 SR 3.3.2.1.6 SR 3.3.2.1.8
: 3.                   Reactor       Mode   Switch     -       Shutdown       Position                                                                                                                                                                                                         (i) 2 SR         3.3.2.1. 7 NA
: 3. Reactor Mode Switch - Shutdown Position (i) 2 SR 3.3.2.1. 7 NA


(a)         THERMAL               POWER~ 27.5%   and   <   62.5%     RTP   and   MCPR -<-4--:+G less   than     the   limit specified       in the   COLR     and   no peripheral control       rod   selected.
(a) THERMAL POWER~ 27.5% and < 62.5% RTP and MCPR -<-4--:+G less than the limit specified in the COLR and no peripheral control rod selected.


(b)           If the   as-found       channel       setpoint       is outside       its   predefined     as-found       tolerance,         then   the channel   shall   be evaluated         to verify that it is functioning           as   required     before     returning       the   channel     to service.
(b) If the as-found channel setpoint is outside its predefined as-found tolerance, then the channel shall be evaluated to verify that it is functioning as required before returning the channel to service.


(c)         The   instrument         channel     setpoint     shall     be   reset     to a value   that   is within     the   as-left     tolerance     around   the   Limiting       Trip   Setpoint (L TSP)   at the   completion       of the surveillance;             otherwise,     the   channel       shall     be declared       inoperable.           Setpoints         more   conservative than     the   L TSP   are   acceptable         provided       that   the as-found     and   as-left     tolerances         apply   to   the actual   setpoint         implemented       in the Surveillance             procedures       (Nominal     Trip   Setpoint)       to confirm   channel       performance.               The   Limiting   Trip   Setpoint       and   the methodologies             used     to determine       the as-found         and   the as-left   tolerances         are specified         in the Technical       Requirements           Manual.
(c) The instrument channel setpoint shall be reset to a value that is within the as-left tolerance around the Limiting Trip Setpoint (L TSP) at the completion of the surveillance; otherwise, the channel shall be declared inoperable. Setpoints more conservative than the L TSP are acceptable provided that the as-found and as-left tolerances apply to the actual setpoint implemented in the Surveillance procedures (Nominal Trip Setpoint) to confirm channel performance. The Limiting Trip Setpoint and the methodologies used to determine the as-found and the as-left tolerances are specified in the Technical Requirements Manual.


(d)         THERMAL               POWER~                   62.5%   and   < 82.5%       RTP   and   MCPR   <-4:-+G-   less   than   the   limit   specified     in   the   COLR     and   no peripheral control       rod   selected.
(d) THERMAL POWER~ 62.5% and < 82.5% RTP and MCPR <-4:-+G-less than the limit specified in the COLR and no peripheral control rod selected.


(e)         THERMAL               POWER~                 82.5%   and<     90%   RTP   and   MCPR ~         less than     the   limit   specified in the COLR     and   no   peripheral control       rod   selected.
(e) THERMAL POWER~ 82.5% and< 90% RTP and MCPR ~ less than the limit specified in the COLR and no peripheral control rod selected.


(f)               THERMAL         POWER~                 90%   RTP   and     MCPR -<-4-A-G-less than   the   limit   specified         in   the   COLR   and   no   peripheral         control     rod selected.
(f) THERMAL POWER~ 90% RTP and MCPR -<-4-A-G-less than the limit specified in the COLR and no peripheral control rod selected.


(g)         THERMAL               POWER~                 27.5%   and   <   90%   RTP   and   MCPR ~                                               less   than     the   limit   specified     in the COLR     and     no peripheral control       rod   selected.
(g) THERMAL POWER~ 27.5% and < 90% RTP and MCPR ~ less than the limit specified in the COLR and no peripheral control rod selected.


(h)         With   THERMAL       POWER :5 9.85     RTP.
(h) With THERMAL POWER :5 9.85 RTP.


(i)                 Reactor       mode   switch       in   the shutdown         position.
(i) Reactor mode switch in the shutdown position.


U)       Less   than     or equal   to the Allowable       Value     specified       in the   COLR.
U) Less than or equal to the Allowable Value specified in the COLR.


3.3-19                                                                                                                                                                                                                                                                                                                                                         Amendment                 No. 242 Reporting           Requirements 5.6
3.3-19 Amendment No. 242 Reporting Requirements 5.6


5.6                                                                   Reporting     Requirements               (continued)
5.6 Reporting Requirements (continued)


5.6.3                                   Radioactive       Effluent           Release     Report
5.6.3 Radioactive Effluent Release Report


The   Radioactive           Effluent       Release   Report     covering         the operation     of the   unit   shall     be submitted         in   accordance             with       10 CFR   50.36a.                 The     report shall     include       a   summary   of the quantities     of radioactive               liquid   and   gaseous         effluents         and   solid   waste       released     from   the unit.           The   material         provided       shall   be   consistent         with     the   objectives       outlined         in   the   ODAM and   the   Process         Control         Program   and   in   conformance           with     10 CFR   50.36a     and       10 CFR 50,   Appendix           I,     Section             IV. 8.1.
The Radioactive Effluent Release Report covering the operation of the unit shall be submitted in accordance with 10 CFR 50.36a. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit. The material provided shall be consistent with the objectives outlined in the ODAM and the Process Control Program and in conformance with 10 CFR 50.36a and 10 CFR 50, Appendix I, Section IV. 8.1.


5.6.4                                 (Deleted)
5.6.4 (Deleted)


5.6.5                                   Core   Operating         Limits       Report   (COLR)
5.6.5 Core Operating Limits Report (COLR)
: a.                                                                                   Core   operating             limits   shall   be   established             prior to   each     reload       cycle,     or prior to any   remaining             portion   of a reload       cycle,       and     shall be   documented               in   the   COLR for   the following:
: a. Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:
: 1.                                                                         The   Average         Planar   Linear     Heat     Generation       Rates   for   Specifications 3.2.1       and     3.7.7.
: 1. The Average Planar Linear Heat Generation Rates for Specifications 3.2.1 and 3.7.7.
: 2.                                                                           The     Minimum         Critical   Power     Ratio   for Specifications         3.2.2     and   3.7.7,   and MCPR99.9% for Specification           3.2.2.
: 2. The Minimum Critical Power Ratio for Specifications 3.2.2 and 3.7.7, and MCPR99.9% for Specification 3.2.2.
: 3.                                                                           The     Linear       Heat Generation         Rates     for   Specifications         3.2.3     and   3.7.7.
: 3. The Linear Heat Generation Rates for Specifications 3.2.3 and 3.7.7.
: 4.                                                                           The     three         Rod   Block   Monitor       Upscale       Allowable     Values       for   Specification 3.3.2.1.
: 4. The three Rod Block Monitor Upscale Allowable Values for Specification 3.3.2.1.
: 5.                                                                           The     power/flow         map   defining       the     Stability Exclusion             Region     for Specification                 3.4.1.
: 5. The power/flow map defining the Stability Exclusion Region for Specification 3.4.1.
: 6.                                                                           The     Minimum         Critical   Power     Ratios         in   Table   3.3.2.1-1         for Specification 3.3.2.1.
: 6. The Minimum Critical Power Ratios in Table 3.3.2.1-1 for Specification 3.3.2.1.
: b.                                                                                 The   analytical             methods       used   to determine         the   core   operating         limits     shall     be   those previously             reviewed         and   approved         by the     NRC,   specifically       those       described       in the   following           documents:
: b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:
: 1.                                                                                 NEDE-24011-P-A,               "General         Electric       Standard     Application         for   Reactor Fuel"       (Revision       specified       in   the     COLR).
: 1. NEDE-24011-P-A, "General Electric Standard Application for Reactor Fuel" (Revision specified in the COLR).


( continued)
( continued)


Cooper                                                                                                 5.0-21                                                                                                                                                                                                                                                                                                                                                                                                                                                                     Amendment         No.
Cooper 5.0-21 Amendment No.
NLS2024025 Page       1 of3
NLS2024025 Page 1 of3


Attachment       3
Attachment 3


Revised     Technical       Specifications         Pages
Revised Technical Specifications Pages


Cooper   Nuclear       Station,     Docket     No. 50-298,   Renewed       License     No. DPR-46
Cooper Nuclear Station, Docket No. 50-298, Renewed License No. DPR-46


Revised     Pages
Revised Pages


3.3-19 5.0-21 Control         Rod     Block     Instrumentation 3.3.2.1
3.3-19 5.0-21 Control Rod Block Instrumentation 3.3.2.1


Table     3.3.2.1-1 (page     1 of 1)
Table 3.3.2.1-1 (page 1 of 1)
Control     Rod   Block   Instrumentation
Control Rod Block Instrumentation


APPLICABLE MODES       OR OTHER SPECIFIED                                     REQUIRED                                 SURVEILLANCE                                           ALLOWABLE FUNCTION                                                                     CONDITIONS                                       CHANNELS                                 REQUIREMENTS                                                   VALUE
APPLICABLE MODES OR OTHER SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS CHANNELS REQUIREMENTS VALUE
: 1.                       Rod   Block     Monitor
: 1. Rod Block Monitor
: a.                   Low Power         Range   -       Upscale                                                                                                                                                                                                                                                       (a) 2 SR         3.3.2.1.1 (j)
: a. Low Power Range - Upscale (a) 2 SR 3.3.2.1.1 (j)
SR         3.3.2.1.4 SR       3.3.2.1.5(b)(c)
SR 3.3.2.1.4 SR 3.3.2.1.5(b)(c)
: b.                     Intermediate         Power     Range   -       Upscale                                                                                                                                                               (d) 2 SR         3.3.2.1.1 (j)
: b. Intermediate Power Range - Upscale (d) 2 SR 3.3.2.1.1 (j)
SR         3.3.2.1.4 SR       3.3.2.1.5(b)(c)
SR 3.3.2.1.4 SR 3.3.2.1.5(b)(c)
: c.                       High   Power     Range   -       Upscale                                                                                                                                                                                                                                   (e),(f) 2 SR           3.3.2.1.1 0)
: c. High Power Range - Upscale (e),(f) 2 SR 3.3.2.1.1 0)
SR         3.3.2.1.4 SR       3.3.2.1.5(b)(c)
SR 3.3.2.1.4 SR 3.3.2.1.5(b)(c)
: d.                     lnop                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                             (f),(g) 2 SR           3.3.2.1.1 NA
: d. lnop (f),(g) 2 SR 3.3.2.1.1 NA
: e.                     Downscale                                                                                                                                                                                                                                                                                                                                                                                                                                                         (f),(g) 2 SR           3.3.2.1.1 ~ 92/125 divisions SR         3.3.2.1.5                             of full scale
: e. Downscale (f),(g) 2 SR 3.3.2.1.1 ~ 92/125 divisions SR 3.3.2.1.5 of full scale
: 2.                       Rod   Worth     Minimizer                                                                                                                                                             SR         3.3.2.1.2                             NA SR           3.3.2.1.3 SR         3.3.2.1.6 SR         3.3.2.1.8
: 2. Rod Worth Minimizer SR 3.3.2.1.2 NA SR 3.3.2.1.3 SR 3.3.2.1.6 SR 3.3.2.1.8
: 3.                     Reactor     Mode Switch     -       Shutdown       Position                                                                                                                                                           (i) 2 SR           3.3.2.1. 7 NA
: 3. Reactor Mode Switch - Shutdown Position (i) 2 SR 3.3.2.1. 7 NA


(a)         THERMAL                 POWER~ 27.5% and <   62.5% RTP   and   MCPR     less than   the limit specified     in the   COLR   and   no   peripheral     control rod   selected.
(a) THERMAL POWER~ 27.5% and < 62.5% RTP and MCPR less than the limit specified in the COLR and no peripheral control rod selected.


(b)           If the   as-found     channel     setpoint     is outside     its   predefined     as-found       tolerance,     then   the channel     shall   be evaluated       to verify that   it is functioning         as required     before   returning       the channel     to service.
(b) If the as-found channel setpoint is outside its predefined as-found tolerance, then the channel shall be evaluated to verify that it is functioning as required before returning the channel to service.


(c)           The   instrument       channel     setpoint     shall   be reset   to a value   that   is within     the as-left   tolerance       around     the   Limiting     Trip Setpoint (L TSP)   at the completion       of the surveillance;       otherwise,     the   channel     shall   be declared     inoperable.               Setpoints     more   conservative than   the   L TSP   are acceptable         provided     that the   as-found     and   as-left   tolerances       apply   to the   actual   setpoint     implemented     in the Surveillance         procedures       (Nominal     Trip   Setpoint)     to confirm     channel     performance.             The   Limiting   Trip Setpoint                                                             and the methodologies           used   to determine       the as-found       and   the as-left   tolerances       are specified       in the Technical       Requirements     Manual.
(c) The instrument channel setpoint shall be reset to a value that is within the as-left tolerance around the Limiting Trip Setpoint (L TSP) at the completion of the surveillance; otherwise, the channel shall be declared inoperable. Setpoints more conservative than the L TSP are acceptable provided that the as-found and as-left tolerances apply to the actual setpoint implemented in the Surveillance procedures (Nominal Trip Setpoint) to confirm channel performance. The Limiting Trip Setpoint and the methodologies used to determine the as-found and the as-left tolerances are specified in the Technical Requirements Manual.


(d)         THERMAL                 POWER~                   62.5% and   < 82.5% RTP and   MCPR     less than   the limit specified       in the   COLR   and   no   peripheral     control rod   selected.
(d) THERMAL POWER~ 62.5% and < 82.5% RTP and MCPR less than the limit specified in the COLR and no peripheral control rod selected.


(e)         THERMAL                 POWER~                   82.5% and<     90% RTP and   MCPR   less   than   the   limit specified     in the   COLR   and   no peripheral     control   rod selected.
(e) THERMAL POWER~ 82.5% and< 90% RTP and MCPR less than the limit specified in the COLR and no peripheral control rod selected.


(f)               THERMAL           POWER~                   90%   RTP and   MCPR   less than   the   limit specified         in the   COLR and   no peripheral       control     rod   selected.
(f) THERMAL POWER~ 90% RTP and MCPR less than the limit specified in the COLR and no peripheral control rod selected.


(g)         THERMAL       POWER                         ~   27.5% and   < 90% RTP   and   MCPR   less   than   the   limit specified       in the   COLR   and   no peripheral     control   rod selected.
(g) THERMAL POWER ~ 27.5% and < 90% RTP and MCPR less than the limit specified in the COLR and no peripheral control rod selected.


(h)         With   THERMAL       POWER s 9.85 RTP.
(h) With THERMAL POWER s 9.85 RTP.


(i)                   Reactor     mode   switch       in the   shutdown     position.
(i) Reactor mode switch in the shutdown position.


(j)       Less   than   or equal     to the Allowable       Value   specified       in   the   COLR.
(j) Less than or equal to the Allowable Value specified in the COLR.


Cooper                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                         3.3-19                                                                                                                                                                                                                                                                                                                                                                     Amendment             No.
Cooper 3.3-19 Amendment No.
Reporting     Requirements 5.6
Reporting Requirements 5.6


5.6                                                                     Reporting Requirements     ( continued)
5.6 Reporting Requirements ( continued)


5.6.3                                   Radioactive Effluent     Release Report
5.6.3 Radioactive Effluent Release Report


The Radioactive     Effluent   Release Report   covering   the operation of the   unit shall   be submitted   in   accordance   with     10 CFR 50.36a.               The   report shall include a summary of the quantities of radioactive     liquid and gaseous     effluents and solid waste   released   from the unit.           The   material     provided   shall be   consistent   with   the objectives outlined     in   the   ODAM and the   Process   Control   Program and   in   conformance   with     10 CFR 50.36a   and     10 CFR 50,   Appendix       I,     Section     IV.8.1.
The Radioactive Effluent Release Report covering the operation of the unit shall be submitted in accordance with 10 CFR 50.36a. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit. The material provided shall be consistent with the objectives outlined in the ODAM and the Process Control Program and in conformance with 10 CFR 50.36a and 10 CFR 50, Appendix I, Section IV.8.1.


5.6.4                                 (Deleted)
5.6.4 (Deleted)


5.6.5                                       Core Operating     Limits   Report (COLR)
5.6.5 Core Operating Limits Report (COLR)
: a.                                                                                         Core   operating     limits shall be   established     prior to each   reload   cycle, or prior to any   remaining       portion of a reload   cycle,   and   shall be   documented       in   the   COLR for the   following:
: a. Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:
: 1.                                                                                           The Average     Planar Linear   Heat Generation   Rates for Specifications 3.2.1   and   3.7.7.
: 1. The Average Planar Linear Heat Generation Rates for Specifications 3.2.1 and 3.7.7.
: 2.                                                                                     The   Minimum   Critical Power     Ratio for Specifications   3.2.2   and   3.7.7, and MCPR99.9% for Specification   3.2.2.
: 2. The Minimum Critical Power Ratio for Specifications 3.2.2 and 3.7.7, and MCPR99.9% for Specification 3.2.2.
: 3.                                                                                             The   Linear   Heat Generation     Rates for Specifications   3.2.3   and 3.7.7.
: 3. The Linear Heat Generation Rates for Specifications 3.2.3 and 3.7.7.
: 4.                                                                                             The three     Rod   Block Monitor     Upscale Allowable   Values for   Specification 3.3.2.1.
: 4. The three Rod Block Monitor Upscale Allowable Values for Specification 3.3.2.1.
: 5.                                                                                             The   power/flow   map defining     the Stability   Exclusion Region for Specification     3.4.1.
: 5. The power/flow map defining the Stability Exclusion Region for Specification 3.4.1.
: 6.                                                                                             The   Minimum   Critical Power     Ratios   in   Table   3.3.2.1-1   for Specification 3.3.2.1.
: 6. The Minimum Critical Power Ratios in Table 3.3.2.1-1 for Specification 3.3.2.1.
: b.                                                                                           The   analytical     methods   used to   determine   the   core operating     limits shall be   those previously       reviewed   and approved     by the   NRC, specifically   those   described in the following     documents:
: b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:
: 1.                                                                                           NEDE-24011-P-A,   "General     Electric Standard Application   for Reactor Fuel" (Revision   specified   in   the   COLR).
: 1. NEDE-24011-P-A, "General Electric Standard Application for Reactor Fuel" (Revision specified in the COLR).


( continued)
( continued)


Cooper                                                                                     5.0-21                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                       Amendment   No.
Cooper 5.0-21 Amendment No.
NLS2024025 Page       1 of2 Attachment       4
NLS2024025 Page 1 of2 Attachment 4


Proposed       Technical       Specifications         Bases       Changes     (Mark-up)
Proposed Technical Specifications Bases Changes (Mark-up)


Cooper     Nuclear         Station,       Docket     No. 50-298,       Renewed         License     No. DPR-46
Cooper Nuclear Station, Docket No. 50-298, Renewed License No. DPR-46


Marked     Up   Page
Marked Up Page


B   3.3-45 Corih*ol Rod     Block lnslruman!alion B 3,3.2,1
B 3.3-45 Corih*ol Rod Block lnslruman!alion B 3,3.2,1


BASES
BASES


APPLICABLE SAFETY ANALYSESr LCO, and APPLICABl:UTY {OOfltinued)
APPLICABLE SAFETY ANALYSESr LCO, and APPLICABl:UTY {OOfltinued)
: 1.                                                                                       Rod Block Monitor
: 1. Rod Block Monitor


The   RBM is   dasfgned         to prevent       violation       of lh@   MCPR SL and   tM cladding. 1 %     plastjo   strain     fuel design     limit   thitt may resutt from       a sjllgle control rod WB,hdrawat error (RWE) event.           TM anal*ytk:eJ methodiS and assumptions             used       in evaluating             the   RV!JE event   are   summarfzed         in R~ 4.             A statistical analysis       of RWE events was     performed ro determine the RSM resp(lfflse for bolh channel& for aech event.     Fron, these     responses.           lhe     fuel thermal       performance             as   a function of RBM
The RBM is dasfgned to prevent violation of lh@ MCPR SL and tM cladding. 1 % plastjo strain fuel design limit thitt may resutt from a sjllgle control rod WB,hdrawat error (RWE) event. TM anal*ytk:eJ methodiS and assumptions used in evaluating the RV!JE event are summarfzed in R~ 4. A statistical analysis of RWE events was performed ro determine the RSM resp(lfflse for bolh channel& for aech event. Fron, these responses. lhe fuel thermal performance as a function of RBM
                              .AJlowabte Vatue was determined. The Allowab4e Values are* chOGen as a function       of P(IWer revel.       Based     on lhe $pecifted     .Alov,able       Val~.
.AJlowabte Vatue was determined. The Allowab4e Values are* chOGen as a function of P(IWer revel. Based on lhe $pecifted.Alov,able Val~.
operating limits are establf,shed,
operating limits are establf,shed,


The   RBM   Function         satisfies           Criterion           3   of' 10 CFR. 50.36(c)(2)(ij)             (Ref.       5).
The RBM Function satisfies Criterion 3 of' 10 CFR. 50.36(c)(2)(ij) (Ref. 5).


T\\'/0 channers of lhe RSM   are required to be OPERABLE~ 'Nith U,eir setpoints       within       lhe   app.ropriate     Allowable           Values,     to ensure     that no singre instrument failure can preclude a rod block from this FundiOn. The actual setpolnts are callbrated oonslstent           wUh appficable setpolnt m-ethodofogy.
T\\'/0 channers of lhe RSM are required to be OPERABLE~ 'Nith U,eir setpoints within lhe app.ropriate Allowable Values, to ensure that no singre instrument failure can preclude a rod block from this FundiOn. The actual setpolnts are callbrated oonslstent wUh appficable setpolnt m-ethodofogy.


The   RBM is   assumed       to   mitigate         the   oonseqoonces           of an   RWE event when operating ~               30% RT P (anatyticai limit), and a peripheral control rod is ool selected. B&low this PO'Net' -,vel   or If a   peripheral         control rod   is seleci:ed1 the   oonsequences           of an RWE event   wil   not exceed         the MCPR SL and, therefore, lhe RSM Is not ra~ulred to be   OPERABLE (Ref, 4 ).           When     operating           < 90%     RTF\\     analyses         (Ref. 4)   have sh0\\\\l11i that with Sl~           Abio1 the analyse. de,rnonstrate a.n .ini.tia .. * .. M .... * .. cPRa**** .... 1.*.n.**.o. RW.E.' eve.ntW1. lhat Wtien * *u res*u.tt operating in.exceed.*~ ... *.* ... al~ 00% in. g*t*h*e.MC. RTP *p.*,R ....
The RBM is assumed to mitigate the oonseqoonces of an RWE event when operating ~ 30% RT P (anatyticai limit), and a peripheral control rod is ool selected. B&low this PO'Net' -,vel or If a peripheral control rod is seleci:ed1 the oonsequences of an RWE event wil not exceed the MCPR SL and, therefore, lhe RSM Is not ra~ulred to be OPERABLE (Ref, 4 ). When operating < 90% RTF\\ analyses (Ref. 4) have sh0\\\\l11i that with Sl~ Abio1 the analyse. de,rnonstrate a.n.ini.tia.. *.. M.... *.. cPRa****.... 1.*.n.**.o. RW.E.' eve.ntW1. lhat Wtien * *u res*u.tt operating in.exceed.*~... *.*... al~ 00% in. g*t*h*e.MC. RTP *p.*,R....
* with     MCPR ~. oo . * * .         E event   will   refiutt           in   exceeding       the   MCPR     SL (Ref, 4 ).             Thntfoft~                 . . ,       Ulese   oonditions1 the RSM is also not required to be OPERABLE.                                                                   ~*                                 '
* with MCPR ~. oo. * *. E event will refiutt in exceeding the MCPR SL (Ref, 4 ). Thntfoft~.., Ulese oonditions1 the RSM is also not required to be OPERABLE. ~* '


The RWM is a. backup to openatOf" oonlrol       of the     rod sequences.                         The RWM eriforces the bankedl poailion wilhd rawal eeque<iee (BPVIJS) by alerting       tha     ~tor when     Ute rod .paUam is not   jn ao::;ordanct, wHh BPWS.       Compliallee             with     BP\\IVS ensures         that   the   initial     conditions       of the CRDA amdysh.i are   not violated.
The RWM is a. backup to openatOf" oonlrol of the rod sequences. The RWM eriforces the bankedl poailion wilhd rawal eeque<iee (BPVIJS) by alerting tha ~tor when Ute rod.paUam is not jn ao::;ordanct, wHh BPWS. Compliallee with BP\\IVS ensures that the initial conditions of the CRDA amdysh.i are not violated.


The analytical         methods           arJd Hsumptioos             used       in evaluating         the   CRDA are* summarized       in References               6   and   7.           The   BPWS   requires       that conlml
The analytical methods arJd Hsumptioos used in evaluating the CRDA are* summarized in References 6 and 7. The BPWS requires that conlml


Cooper}}
Cooper}}

Revision as of 14:13, 4 October 2024

License Amendment Request to Revise Technical Specifications Table 3.3.2.1-1, Control Rod Block Instrumentation
ML24131A026
Person / Time
Site: Cooper Entergy icon.png
Issue date: 05/09/2024
From: Dia K
Nebraska Public Power District (NPPD)
To:
Office of Nuclear Reactor Regulation, Document Control Desk
References
NLS2024025
Download: ML24131A026 (1)
v • d • e


Text

H Nebraska Public Power District "Always there when you need us"

50.90 NLS2024025 May 9, 2024

Attention: Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001

Subject:

License Amendment Request to Revise Technical Specifications Table 3.3.2.1-1, "Control Rod Block Instrumentation" Cooper Nuclear Station, Docket No. 50-298, Renewed License No. DPR-46

Dear Sir or Madam:

Pursuant to 10 CFR 50.90, "Application for amendment oflicense, construction permit, or early site permit," Nebraska Public Power District (NPPD) is submitting a request for an amendment to Renewed License No. DPR-46 for Cooper Nuclear Station (CNS). The proposed change would relocate cycle specific Minimum Critical Power Ratio values from the CNS Technical Specifications (TS) Table 3.3.2.1-1 to the CNS Core Operating Limits Report.

to this letter provides a description and assessment of the proposed change. provides the existing TS pages marked up to show the proposed change. provides revised (clean) TS pages. The changes to the TS Bases are provided for information only in Attachment 4 and will be incorporated upon implementation of the approved amendment.

Approval of the proposed amendment is requested by September 30, 2024. Once approved, the amendment shall be implemented by October 30, 2024, to allow plant operation after the completion of a refueling outage.

NPPD has determined that there are no significant hazards considerations associated with the proposed change and that the TS change qualifies for a categorical exclusion from environmental review pursuant to the provisions of 10 CFR 51.22( c )(9).

The proposed TS change has been reviewed by the necessary safety review committees (Station Operations Review Committee and Safety Review and Audit Board). In accordance with 10 CFR 50.91, "Notice for public comment; State consultation," a copy of this application, with attachments, is being provided to the designated State of Nebraska Official. Copies to the Nuclear Regulatory Commission Region IV office and the CNS Resident Inspectors are also being provided in accordance with 10 CFR 50.4(b)(l).

COOPER NUCLEAR STATION 72676 648A Ave/ P.O. Box 98 / Brownville, NE 68321 http://www.nppd.com NLS2024025 Page 2 of2

There are no regulatory commitments made in this submittal. If you should have any questions regarding this submittal, please contact Linda Dewhirst, Regulatory Affairs and Compliance Manager, at (402) 825-5416.

I declare under penalty of perjury that the foregoing is true and correct.

ExecutedOn: 5/~ ZnZ>{

ate

Khalil Dia Site Vice President

lbs

Attachments: 1. Description and Assessment

2. Proposed Technical Specifications Change (Mark-up)
3. Revised Technical Specifications Pages
4. Proposed Technical Specifications Bases Changes (Mark-up)

cc: Regional Administrator w/ attachments USNRC - Region IV

Cooper Project Manager w/ attachments USNRC - NRR Plant Licensing Branch IV

Senior Resident Inspector w/ attachments USNRC-CNS

Nebraska Health and Human Services w/ attachments Department of Regulation and Li censure

NPG Distribution w/ attachments

CNS Records w / attachments NLS2024025 Page 1 of 6

Attachment 1

Description and Assessment

Cooper Nuclear Station, Docket No. 50-298, Renewed License No. DPR-46

1.0 Summary Description

2.0 Detailed Description

3.0 Technical Evaluation

4.0 Regulatory Analysis

4.1 Applicable Regulatory Requirements/Criteria 4.2 Precedent 4.3 No Significant Hazards Consideration Determination Analysis 4.4 Conclusion

5.0 Environmental Evaluation NLS2024025 Page 2 of6

1.0

SUMMARY

DESCRIPTION

In accordance with 10 CFR 50.90, "Application for amendment oflicense, construction permit, or early site permit," Nebraska Public Power District (NPPD) is submitting a request for an amendment to Renewed License No. DPR-46 for Cooper Nuclear Station (CNS). The proposed change would amend the CNS Technical Specifications (TS) to modify TS Table 3.3.2.1-1, "Control Rod Block Instrumentation." The proposed change would relocate cycle specific Minimum Critical Power Ratio (MCPR) values to the CNS Core Operating Limits Report (COLR).

2.0 DETAILED DESCRIPTION

NPPD proposes the following changes to the CNS TS:

1. Revise the notes associated with TS Table 3.3.2.1-1, "Control Rod Block Instrumentation," as shown below:

(a) THERMAL POWER 2: 27.5% and< 62.5% RTP and MCPR ~ less than the limit specified in the COLR and no peripheral control rod selected.

( d) THERMAL POWER 2: 62.5% and < 82.5% RTP and MCPR ~ less than the limit specified in the COLR and no peripheral control rod selected.

(e) THERMAL POWER 2: 82.5% and< 90% RTP and MCPR ~ less than the limit specified in the COLR and no peripheral control rod selected.

(t) THERMAL POWER 2: 90% RTP and MCPR < 1.40 less than the limit specified in the COLR and no peripheral control rod selected.

(g) THERMAL POWER 2: 27.5% and< 90% RTP and MCPR ~ less than the limit specified in the COLR and no peripheral control rod selected.

2. Revise TS 5.6.5, CORE OPERATING LIMITS REPORT (COLR) as shown below:
a. Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:
1. The Average Planar Linear Heat Generation Rates for Specifications 3.2.1 and 3.7.7.
2. The Minimum Critical Power Ratio for Specifications 3.2.2 and 3.7.7, and MCPR99.9% for Specification 3.2.2.
3. The Linear Heat Generation Rates for Specifications 3.2.3 and 3. 7. 7.
4. The three Rod Block Monitor Upscale Allowable Values for Specification 3.3.2.1.
5. The power/flow map defining the Stability Exclusion Region for Specification 3.4.1.
6. The Minimum Critical Power Ratios in Table 3.3.2.1-1 for Specification 3.3.2.1.

Relocation of the cycle specific MCPRs to the COLR, which is controlled by TS 5.6.5, would provide NPPD the flexibility to revise cycle specific MCPRs in accordance with Nuclear Regulatory Commission (NRC) approved methodologies without the need for a license NLS2024025 Page 3 of 6

amendment. The COLR, including any mid-cycle revisions or supplements, is required to be submitted to the NRC for each reload cycle per TS 5.6.5.

3.0 TECHNICAL EVALUATION

NRC Generic Letter (GL) 88-16, "Removal of Cycle-Specific Parameter Limits From Technical Specifications," provides guidance to licensees for the removal of cycle dependent parameter limits from the TS provided these values are included in a COLR and are determined with NRC approved methodologies referenced in the TS. The specific values of these limits may be modified by licensees, without affecting nuclear safety, provided that these changes are determined using an NRC approved methodology and consistent with all applicable limits of the plant safety analysis that are addressed in the Final Safety Analysis Report. If any of the applicable limits of the safety analysis are not met, prior NRC approval of the change is required.

Control Rod Block Instrumentation MCPR values are calculated as part of the reload core design licensing analyses in accordance with NRC approved methods in NEDE-24011-P-A, General Electric Standard Application for Reactor Fuel. Because the analyses are completed only two to three months prior to the start of the next refueling outage, a need to revise MCPR values for the next operating cycle would require submitting a license amendment request with a quick turnaround, placing an unnecessary burden on NPPD and NRC resources. Relocating the MCPR values to the COLR will allow NPPD to make cycle-specific changes that are consistent with NRC approved methodologies and within limits of the safety analysis without the burdensome process of amending the TS. The TS will continue to establish limits for MCPR for the Control Rod Block Instrumentation while the specific values for MCPR are relocated to the COLR.

Therefore, the requested changes are essentially administrative in nature, and the required level of safety will be maintained.

4.0 REGULATORY ANALYSIS

4.1 Applicable Regulatory Requirements/Criteria

NRC GL 88-16 discusses that processing TS changes to update cycle-specific parameter limits each fuel cycle places an unnecessary burden on the licensee and the NRC if these limits are developed using an NRC approved methodology. The GL provides an alternative that relocates the specific parameter values to the COLR provided the values are determined using an NRC approved methodology and the TS require plant operation in accordance with the limits specified in the COLR.

10 CFR 50.36, Technical specifications - requires that the TS contain limiting conditions for operation, which are the lowest functional capability or performance levels of equipment required for safe operation of the facility.

The proposed changes are consistent with the above regulatory guidance and regulation.

NLS2024025 Page 4 of 6

4.2 Precedent

In March 2018, Duane Arnold Energy Center received approval to relocate the Control Rod Block Instrumentation MCPR values to the COLR (ML18011A059). Also in January 2014, Columbia Generating Station received approval to relocate the Control Rod Block Instrumentation MCPR values to the COLR as part of a change to implement a digital instrumentation system (ML133 l 7B623).

The TS for the plants below contain notes comparable to those in the CNS TS that establish limits on MCPR for the Control Rod Block Instrumentation. Similar to the proposed change to the CNS TS, the TS below do not provide the specific value for the Control Rod Block Instrumentation MCPR but specify the limit for MCPR as "less than the limit specified in the COLR."

4.3 No Significant Hazards Consideration Determination Analysis

Nebraska Public Power District (NPPD) requests an amendment to the Cooper Nuclear Station (CNS) Technical Specifications (TS) to modify TS Table 3.3.2.1-1, "Control Rod Block Instrumentation." The proposed change would relocate cycle specific Minimum Critical Power Ratio (MCPR) values to the CNS Core Operating Limits Report (COLR).

As required by 10 CFR 50.91(a), NPPD has evaluated the proposed change to the CNS TS using the criteria in 10 CFR 50.92 and determined that the proposed change does not involve a significant hazards consideration. An analysis of the issue of no significant hazards consideration is presented below.

1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No

The proposed change is an administrative change that does not affect any plant systems, structures, or components designed for the prevention or mitigation of previously evaluated accidents. No new equipment is added nor is installed equipment being changed or operated in a different manner. Relocation of the Control Rod Block Instrumentation MCPR values to the COLR has no influence or impact on, nor does it contribute in any way to the probability or consequences of transients or accidents. The COLR will continue to be controlled by the CNS programs and procedures that comply with TS 5.6.5.

Transient analyses addressed in the Updated Safety Analysis Report will continue to be performed in the same manner with respect to changes in the cycle dependent parameters obtained from the use of Nuclear Regulatory Commission (NRC) approved reload design NLS2024025 Page 5 of 6

methodologies, which ensures that the transient evaluation of new reloads are bounded by previously accepted analyses.

Therefore, the proposed TS change does not involve an increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No

The proposed administrative change does not involve any changes to the operation, testing, or maintenance of any safety-related, or otherwise important to safety systems.

Systems important to safety will continue to be operated and maintained within their design bases. Relocation of the Control Rod Block Instrumentation MCPR values to the COLR has no influence or impact on new or different kind of accidents.

Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed change involve a significant reduction in a margin of safety?

Response: No

The margin of safety is not affected by the relocation of cycle-specific Control Rod Block Instrumentation MCPR values from the TS to the COLR. Appropriate measures exist to control the values of these cycle-specific limits since it is required by TS that only NRC approved methods be used to determine the limits. The proposed change continues to require operation within the core thermal limits as obtained from NRC approved reload design methodologies and the actions to be taken if a limit is exceeded remain unchanged in accordance with existing TS.

Therefore, the proposed change has no impact to the margin of safety.

Based on the above, NPPD concludes that the proposed change presents no significant hazards consideration under the standards set forth in 10 CFR 50.92, and, accordingly, a finding of "no significant hazards consideration" is justified.

4.4 Conclusion

Based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

NLS2024025 Page 6 of6

5.0 ENVIRONMENTAL EVALUATION

NPPD has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure.

Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22( c )(9). Therefore, pursuant to 10 CFR 51.22(b ), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

NLS2024025 Page 1 of3

Attachment 2

Proposed Technical Specifications Changes (Mark-up)

Cooper Nuclear Station, Docket No. 50-298, Renewed License No. DPR-46

Marked Up Pages

3.3-19 5.0-21 Control Rod Block Instrumentation 3.3.2.1

Table3.3.2.1-1 (page 1 of 1)

Control Rod Block Instrumentation

APPLICABLE MODES OR OTHER SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS CHANNELS REQUIREMENTS VALUE

1. Rod Block Monitor
a. Low Power Range - Upscale (a) 2 SR 3.3.2.1.1 U)

SR 3.3.2.1.4 SR 3.3.2.1.5(b)(c)

b. Intermediate Power Range - Upscale (d) 2 SR 3.3.2.1.1 U)

SR 3.3.2.1.4 SR 3.3.2.1.5(b)(c)

c. High Power Range - Upscale ( e),(f) 2 SR 3.3.2.1.1 U)

SR 3.3.2.1.4 SR 3.3.2.1.5(b)(c)

d. lnop (f),(g) 2 SR 3.3.2.1.1 NA
e. Downscale (f),(g) 2 SR 3.3.2.1.1 ~ 92/125 divisions SR 3.3.2.1.5 of full scale
2. Rod Worth Minimizer 1(h),2(h) SR 3.3.2.1.2 NA SR 3.3.2.1.3 SR 3.3.2.1.6 SR 3.3.2.1.8
3. Reactor Mode Switch - Shutdown Position (i) 2 SR 3.3.2.1. 7 NA

(a) THERMAL POWER~ 27.5% and < 62.5% RTP and MCPR -<-4--:+G less than the limit specified in the COLR and no peripheral control rod selected.

(b) If the as-found channel setpoint is outside its predefined as-found tolerance, then the channel shall be evaluated to verify that it is functioning as required before returning the channel to service.

(c) The instrument channel setpoint shall be reset to a value that is within the as-left tolerance around the Limiting Trip Setpoint (L TSP) at the completion of the surveillance; otherwise, the channel shall be declared inoperable. Setpoints more conservative than the L TSP are acceptable provided that the as-found and as-left tolerances apply to the actual setpoint implemented in the Surveillance procedures (Nominal Trip Setpoint) to confirm channel performance. The Limiting Trip Setpoint and the methodologies used to determine the as-found and the as-left tolerances are specified in the Technical Requirements Manual.

(d) THERMAL POWER~ 62.5% and < 82.5% RTP and MCPR <-4:-+G-less than the limit specified in the COLR and no peripheral control rod selected.

(e) THERMAL POWER~ 82.5% and< 90% RTP and MCPR ~ less than the limit specified in the COLR and no peripheral control rod selected.

(f) THERMAL POWER~ 90% RTP and MCPR -<-4-A-G-less than the limit specified in the COLR and no peripheral control rod selected.

(g) THERMAL POWER~ 27.5% and < 90% RTP and MCPR ~ less than the limit specified in the COLR and no peripheral control rod selected.

(h) With THERMAL POWER :5 9.85 RTP.

(i) Reactor mode switch in the shutdown position.

U) Less than or equal to the Allowable Value specified in the COLR.

3.3-19 Amendment No. 242 Reporting Requirements 5.6

5.6 Reporting Requirements (continued)

5.6.3 Radioactive Effluent Release Report

The Radioactive Effluent Release Report covering the operation of the unit shall be submitted in accordance with 10 CFR 50.36a. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit. The material provided shall be consistent with the objectives outlined in the ODAM and the Process Control Program and in conformance with 10 CFR 50.36a and 10 CFR 50, Appendix I, Section IV. 8.1.

5.6.4 (Deleted)

5.6.5 Core Operating Limits Report (COLR)

a. Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:
1. The Average Planar Linear Heat Generation Rates for Specifications 3.2.1 and 3.7.7.
2. The Minimum Critical Power Ratio for Specifications 3.2.2 and 3.7.7, and MCPR99.9% for Specification 3.2.2.
3. The Linear Heat Generation Rates for Specifications 3.2.3 and 3.7.7.
4. The three Rod Block Monitor Upscale Allowable Values for Specification 3.3.2.1.
5. The power/flow map defining the Stability Exclusion Region for Specification 3.4.1.
6. The Minimum Critical Power Ratios in Table 3.3.2.1-1 for Specification 3.3.2.1.
b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:
1. NEDE-24011-P-A, "General Electric Standard Application for Reactor Fuel" (Revision specified in the COLR).

( continued)

Cooper 5.0-21 Amendment No.

NLS2024025 Page 1 of3

Attachment 3

Revised Technical Specifications Pages

Cooper Nuclear Station, Docket No. 50-298, Renewed License No. DPR-46

Revised Pages

3.3-19 5.0-21 Control Rod Block Instrumentation 3.3.2.1

Table 3.3.2.1-1 (page 1 of 1)

Control Rod Block Instrumentation

APPLICABLE MODES OR OTHER SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS CHANNELS REQUIREMENTS VALUE

1. Rod Block Monitor
a. Low Power Range - Upscale (a) 2 SR 3.3.2.1.1 (j)

SR 3.3.2.1.4 SR 3.3.2.1.5(b)(c)

b. Intermediate Power Range - Upscale (d) 2 SR 3.3.2.1.1 (j)

SR 3.3.2.1.4 SR 3.3.2.1.5(b)(c)

c. High Power Range - Upscale (e),(f) 2 SR 3.3.2.1.1 0)

SR 3.3.2.1.4 SR 3.3.2.1.5(b)(c)

d. lnop (f),(g) 2 SR 3.3.2.1.1 NA
e. Downscale (f),(g) 2 SR 3.3.2.1.1 ~ 92/125 divisions SR 3.3.2.1.5 of full scale
2. Rod Worth Minimizer SR 3.3.2.1.2 NA SR 3.3.2.1.3 SR 3.3.2.1.6 SR 3.3.2.1.8
3. Reactor Mode Switch - Shutdown Position (i) 2 SR 3.3.2.1. 7 NA

(a) THERMAL POWER~ 27.5% and < 62.5% RTP and MCPR less than the limit specified in the COLR and no peripheral control rod selected.

(b) If the as-found channel setpoint is outside its predefined as-found tolerance, then the channel shall be evaluated to verify that it is functioning as required before returning the channel to service.

(c) The instrument channel setpoint shall be reset to a value that is within the as-left tolerance around the Limiting Trip Setpoint (L TSP) at the completion of the surveillance; otherwise, the channel shall be declared inoperable. Setpoints more conservative than the L TSP are acceptable provided that the as-found and as-left tolerances apply to the actual setpoint implemented in the Surveillance procedures (Nominal Trip Setpoint) to confirm channel performance. The Limiting Trip Setpoint and the methodologies used to determine the as-found and the as-left tolerances are specified in the Technical Requirements Manual.

(d) THERMAL POWER~ 62.5% and < 82.5% RTP and MCPR less than the limit specified in the COLR and no peripheral control rod selected.

(e) THERMAL POWER~ 82.5% and< 90% RTP and MCPR less than the limit specified in the COLR and no peripheral control rod selected.

(f) THERMAL POWER~ 90% RTP and MCPR less than the limit specified in the COLR and no peripheral control rod selected.

(g) THERMAL POWER ~ 27.5% and < 90% RTP and MCPR less than the limit specified in the COLR and no peripheral control rod selected.

(h) With THERMAL POWER s 9.85 RTP.

(i) Reactor mode switch in the shutdown position.

(j) Less than or equal to the Allowable Value specified in the COLR.

Cooper 3.3-19 Amendment No.

Reporting Requirements 5.6

5.6 Reporting Requirements ( continued)

5.6.3 Radioactive Effluent Release Report

The Radioactive Effluent Release Report covering the operation of the unit shall be submitted in accordance with 10 CFR 50.36a. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit. The material provided shall be consistent with the objectives outlined in the ODAM and the Process Control Program and in conformance with 10 CFR 50.36a and 10 CFR 50, Appendix I, Section IV.8.1.

5.6.4 (Deleted)

5.6.5 Core Operating Limits Report (COLR)

a. Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:
1. The Average Planar Linear Heat Generation Rates for Specifications 3.2.1 and 3.7.7.
2. The Minimum Critical Power Ratio for Specifications 3.2.2 and 3.7.7, and MCPR99.9% for Specification 3.2.2.
3. The Linear Heat Generation Rates for Specifications 3.2.3 and 3.7.7.
4. The three Rod Block Monitor Upscale Allowable Values for Specification 3.3.2.1.
5. The power/flow map defining the Stability Exclusion Region for Specification 3.4.1.
6. The Minimum Critical Power Ratios in Table 3.3.2.1-1 for Specification 3.3.2.1.
b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:
1. NEDE-24011-P-A, "General Electric Standard Application for Reactor Fuel" (Revision specified in the COLR).

( continued)

Cooper 5.0-21 Amendment No.

NLS2024025 Page 1 of2 Attachment 4

Proposed Technical Specifications Bases Changes (Mark-up)

Cooper Nuclear Station, Docket No. 50-298, Renewed License No. DPR-46

Marked Up Page

B 3.3-45 Corih*ol Rod Block lnslruman!alion B 3,3.2,1

BASES

APPLICABLE SAFETY ANALYSESr LCO, and APPLICABl:UTY {OOfltinued)

1. Rod Block Monitor

The RBM is dasfgned to prevent violation of lh@ MCPR SL and tM cladding. 1 % plastjo strain fuel design limit thitt may resutt from a sjllgle control rod WB,hdrawat error (RWE) event. TM anal*ytk:eJ methodiS and assumptions used in evaluating the RV!JE event are summarfzed in R~ 4. A statistical analysis of RWE events was performed ro determine the RSM resp(lfflse for bolh channel& for aech event. Fron, these responses. lhe fuel thermal performance as a function of RBM

.AJlowabte Vatue was determined. The Allowab4e Values are* chOGen as a function of P(IWer revel. Based on lhe $pecifted.Alov,able Val~.

operating limits are establf,shed,

The RBM Function satisfies Criterion 3 of' 10 CFR. 50.36(c)(2)(ij) (Ref. 5).

T\\'/0 channers of lhe RSM are required to be OPERABLE~ 'Nith U,eir setpoints within lhe app.ropriate Allowable Values, to ensure that no singre instrument failure can preclude a rod block from this FundiOn. The actual setpolnts are callbrated oonslstent wUh appficable setpolnt m-ethodofogy.

The RBM is assumed to mitigate the oonseqoonces of an RWE event when operating ~ 30% RT P (anatyticai limit), and a peripheral control rod is ool selected. B&low this PO'Net' -,vel or If a peripheral control rod is seleci:ed1 the oonsequences of an RWE event wil not exceed the MCPR SL and, therefore, lhe RSM Is not ra~ulred to be OPERABLE (Ref, 4 ). When operating < 90% RTF\\ analyses (Ref. 4) have sh0\\\\l11i that with Sl~ Abio1 the analyse. de,rnonstrate a.n.ini.tia.. *.. M.... *.. cPRa****.... 1.*.n.**.o. RW.E.' eve.ntW1. lhat Wtien * *u res*u.tt operating in.exceed.*~... *.*... al~ 00% in. g*t*h*e.MC. RTP *p.*,R....

  • with MCPR ~. oo. * *. E event will refiutt in exceeding the MCPR SL (Ref, 4 ). Thntfoft~.., Ulese oonditions1 the RSM is also not required to be OPERABLE. ~* '

The RWM is a. backup to openatOf" oonlrol of the rod sequences. The RWM eriforces the bankedl poailion wilhd rawal eeque<iee (BPVIJS) by alerting tha ~tor when Ute rod.paUam is not jn ao::;ordanct, wHh BPWS. Compliallee with BP\\IVS ensures that the initial conditions of the CRDA amdysh.i are not violated.

The analytical methods arJd Hsumptioos used in evaluating the CRDA are* summarized in References 6 and 7. The BPWS requires that conlml

Cooper

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