View source for HL-5368, Application for Amend to License DPR-57,revising Safety Limit Minimum Critical Power Ratio in TS 2.1.1.2 to Reflect Results of Cycle Specific Calculation.Proprietary Basis for Change,Also Encl.Encl Withheld,Per 10CFR2.790(a)(4)

{{Adams
| number = ML20141B740
| issue date = 05/09/1997
| title = Application for Amend to License DPR-57,revising Safety Limit Minimum Critical Power Ratio in TS 2.1.1.2 to Reflect Results of Cycle Specific Calculation.Proprietary Basis for Change,Also Encl.Encl Withheld,Per 10CFR2.790(a)(4)
| author name = Sumner H
| author affiliation = SOUTHERN NUCLEAR OPERATING CO.
| addressee name = 
| addressee affiliation = NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
| docket = 05000321
| license number = 
| contact person = 
| document report number = HL-5368, NUDOCS 9705160021
| package number = ML19355F167
| document type = OPERATING LICENSES-APPLIATION TO AMEND-RENEW EXISTING, TEXT-LICENSE APPLICATIONS &amp; PERMITS
| page count = 15
| project = 
| stage = Request
}}

=Text=
{{#Wiki_filter:. _ . .   .
     .    . Lawls Sumner                    Southern Nuclear
.               Vice President              Operating Company. Inc.
Hatch Pr6 ject Sup@rt     40 lnvemess Parkway Post Office Box 1295 Birmingham, Alabama 35201 l
Tel 205.992.7779 Fax 205.992.0341 SOUTHERN h COMPANY Energy:o Serve %rWorld" May 9, 1997 Docket Nos. 50-321                                                                   HL-5368 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555 Edwin I. Hatch Nuclear Plant - Unit 1 Request to Revise Technical Specifications:
Safety Limit Minimum Critical Power Ratios (SIMCPR)
Gentlemen:
In accordance with the provisions of 10 CFR 50.90, as required by 10 CFR 50.59(c)(1),
Southern Nuclear Operating Company (SNC) hereby proposes changes to the Plant Hatch               ,
Unit i Technical Specifications, Appendix A to Operating License DPR-57. This application proposes to change the Safety Limit Minimum Critical Power Ratio (SLMCPR)in Technical Specification (TS) 2.1.1.2 to reflect results of a cycle specific calculation performed for Unit 1 Operating Cycle 18, expected to commence in                      ,
November,1997.
This proposed change for Unit 1 is similar to that previously approved for Unit 2 Cycle 14. The information enclosed to support this change is a Unit I cycle specific version containing the material submitted for Unit 2 in the request for additional                ,
information response dated January 27,1997.
Enclosure I provides a description of the proposed changes and an explanation of the basis for each change. Enclosure 2 details the bases for SNC's determination that the proposed changes do not involve a significant hazards consideration. Enclosure 3 provides page change instructions for incorporating the proposed change for Cycle 18.
Following Enclosure 3 are the revised Technical Specifications page and the
                                                                                                                 )
corresponding' marked-up page.                                                                  /
The information supporting this proposed change was provided by GE Nuclear Energy             /A   )
and is considered to be General Electric Proprietary information as described in                   l 10CFR2.790(a)(4) and the attached aflidavit (Attachment 1). It is requested that this              ;
information be withheld from public disclosure. Proprietary text is denoted by enclosure           l in double brackets and located by brackets in the right hand margins. A nonproprietary version of Enclosure 1 is attached                 w/rrf# fobublic disclosure (Attachment 2
[2 . 7 970516 2 970509                                       l          *
* PDR    ADOCK 05000321                                    ,_........-...'.----
P                              PDR

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i U.S. Nuclear Regulatory Commission                                                                        Page 2 Southern Nuclear Operating Company requests the proposed amendment for Cycle 18 to
       . be issued with the amendment to be effective prior to the restart from the Plant Hatch Unit 1 outage currently to begin October,1997.                                                                                 ;
In accordance with the requirements of 10 CFR 50.91, the designated State official will                                         .
be sent a copy of this letter and all applicable enclosures.
Mr.11. L. Sumner, Jr. states he is Vice President of Southern Nuclear Operating Company and is authorized to execute this oath on behalf of Southern Nuclear Operating Company, and to the best of his knowledge and belief, the facts set forth in this letter are -                                  i true.
Sincerely, H. L. Sumner, Jr.                                                 l Vd Sworn to and subscribed before meday                                  this of $    Wsw                        1997.
                 'A>ufb                               /n w Notlii9 Pull Commission Expiration Date: Y)#v, c2 /SW                     '
L.

==Enclosures:==
: 1. Basis for Change Request
: 2. 10 CFR 50.92 Evaluation
: 3. Page Change Instructions Attachments:
: 1. Affidavit of Proprietary Information
: 2. Nonproprietary Version of the Basis for Change Request IFUeb cc: (See next page.)                                                                                                              )
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       - IIL-5368 4
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U.S. Nuclear Regulatory Commission                                      Page 3 _
cc: Southern Nuclear Overatine Comnany Mr. P. II. Wells, Nuclear Plant General Manager NORMS l].S Nuclear Reentatorv Commission. Washington. D.C.
Mr. K. Jabbour, Licensing Project Manager - Hatch -
US Nuclear Reculatory Commission. Reelon 11 Mr. L. A. Reyes, Regional Administrator Mr. B. L. liolbrook, Senior Resident Inspector - Hatch State ofGeorgia Mr. J. D. Tanner, Commissioner - Department of Natural Resources
         -IIL-5368
                        . j. . .

Enclosure 1 l.
Edwin I. Hatch Nuclear Plant - Unit 1 Request to Revise Technical Specification:                          i
                                                                                                           ~
Safety Limit Minimum Critical Power Ratios (SLMCPR)
Basis for Chance Reauest PROPOSED CHANGES ~
SNC requests that the Technical Specifications (TS) contained in Appendix A to the Plant Hatch . '
Unit 1 Operating License DPR-57 be amended to revise Technical Specifications Section 2.1.1.2 to reflect changes in the Safety Limit Minimum Critical Power Ratio (SLMCPR).
HACKGROUND t
The proposed change involves revising the SLMCPR contained in Section 2.1.1.2 of the Plant llatch Unit 1 TS. In the course of calculating a cycle-specific SLMCPR for another utility,it was determined that the GESTAR II (General Electric Standard Applicationfor Reactor Fuel,
-         NEDE-24011-P-A-11', and U. S. Supplement NEDE-24011-P-A-11-US', November 17,1995)
       ' fuel type generic SLMCPR may be non-conservative when applied to some core and fuel designs. The U. S. Nuclear Regulatory Commission (USNRC) was informed of this condition in a telephone call by GE on March 27,1996, and it was the subject of a 10 CFR Part 21 notification from GENE dated May 24,1996 (Reference 7). The SLMCPR values were confirmed for the current operating cycle, Cycle 17, in a letter from GENE to SNC, llatch-1 Cycle 17 SLA/CPR, dated June 17,1996.
DISCUSSION OF THE PROPOSED CHANGE 1
GENE's calculation for the plant-specific SLMCPR value for Unit 1 Cycle 18 was performed on the expected Reference Loading Pattern which will be part of the Reload Licensing Analysis for Unit 1 Cycle 18 and is based upon NRC approved methods (General Electric Standard Applicationfor Reactor Fuel, NEDE-24011-P-A-13 and U. S. Supplement NEDE-240ll-P-A-13-US, August,1996) and interim implementing procedures which have been discussed between GENE and the NRC during their meetings with the NRC staff on April 17,1996 and May 6 through 10,1996. The implementing procedures are identical to those used for similar recent analyses (Dodkets 50-324,50 325,50-298,50-277, et. al.) and described in GENE's proposed Amendment 25 to GESTAR 11 (R. J. Reda (GE) to T. E. Collins (NRC), Proposed Amendment 25 to GE Licensing Topical Report NEDE-24011-P-A (GESTAR 11) on Cycle Specific SLAfCPR, December 13,1996.) These procedures incorporate cycle specific                                   !
l
       ' ' Revision 11 has since been superseded by Revision 13, dated August,1996. All the Revision 13 material pertinent to this application is unchanged from Revision 11. For purposes related to  ;
evaluation of this application, Revisions 11 and 13 may be considered equivalent and used interchangeably.
HL-5368                                           El-1                                            ;

.   " Encl 6mre 1 -
Basis for Change Request parameters into the analysis which include the preliminary reference core loading and the actual bundle parameters evaluated at the projected exposure distribution based on projected control blade patterns for the rodded burn through the cycle. The analysis considers tFe full cycle exposure range to determine the most limiting point (s). At these exposure point (s), conservative variations of the projected control blade patterns are used to maximize the number of bundles that contribute rods calculated to be susceptible to boiling transition in order to obtain a conservative calculation of the SLMCPR. This calculation resulted in a Cycle 18 SLMCPR valtw c '      ). The current Unit 1 SLMCPR value for Cycle 17 of 1.07 will not be bounding for Cycle 18.
EVALUATION The proposed change revises the Technical Specifications to reflect the change in the SLMCPR due to the plant specific evaluation performed by GENE for Unit 1, Reload 17, Cycle 18. The new SLMCPR was calculated using NRC approved methods (General Electric Standard efpplicationfor Reactor Fuel, NEDE-24011-P-A-11, and U. S. Supplement NEDE-24011-P-A-11-US, November 17,1995) and interim implementing procedures as discussed during the GENE meetings with the NRC on April 17,1996 and May 6 through 10,1996. The interim procedures incorporate plant cycle specific parameters which include: 1) the expected reference loading pattern,2) conservative variations of projected control blade patterns,3) the actual bundle parameters, and 4) the full cycle exposure range.
The SLMCPR is set such that no mechanistic fuel damage is calculated to occur if the limit is not violated. Since the parameters which result in fuel damage are not directly observable during reactor operation, the thermal and hydraulic conditions resulting in a departure from nucleate boiling have been used to mark the beginning of the region where fuel damage could occur.
Although it is recognized that a departure from nucleate boiling would not necessarily result in damage to BWR fuel rods, the critical power at which boiling transition is calculated to occur has been adopted as a convenient limit. However, the uncertainties in monitoring the core operating state and in the procedures used to calculate the critical power result in an uncertainty in the value of the critical power. Therefore, the SLMCPR is defined as the CPR in the limiting fuel assembly for which more the 99.9% of the fuel rods in the core are expected to avoid boiling transition considering the power distribution within the core and all uncertainties. The new SLMCPR for Cycle 18 at Unit 1 is 1.10.
* Control Rod Pattern Development for the Ilatch-1 Cycle 18 SLMCPR Analysis Projected control blade patterns for the burn through the cycle were used to deplete the core to the cycle exposures to be analyzed. At the desired cycle exposures, the bundle exposure distributions and their associated R-factors, determined in accordance with Reference 6, were utilized for the SLMCPR cases to be analyzed. The use of different rod patterns to achieve the llL-5368                                         El-2

l t     .
  . Enclosure 1 -
Basis-for Ch'ange Request desired cycle exposure has been shown to have a negligible impact on the actual calculated SLMCPR. An estimated SLMCPR was obtained for an exposure point near beginning of cycle (BOC), middle of cycle (MOC), and end of cycle (EOC) in order to establish which exposure point (s) would produce the highest (most conservative) calculated SLMCPR.
For each cycle exposure point ofinterest, the SLMCPR is analyzed with radial power distributions that maximize the number of bundles at, or near, the Operating Limit MCPR during rated power operation. This approach satisfies the stipulation in Reference 1 that the number of rods susceptible to boiling transition be maximized. GENE has established criteria to determine if the control rod patterns and resulting radial power distributions are acceptable. These criteria were discussed with the NRC inspection team during the May 610,1996 inspection and have                _
since been incorporated into the GENE technical design procedures. ((                 - - -
                  ----------. )) Different rod patterns were ana yzed until the criteria on the above parameters were met. The rod pattern search was narrowed by starting from a defined set of patterns known from prior experience to yield the ,1 attest possible MCPR distributions. This was done for the two most limiting exposure points in the cycle. A Monte Carlo analysis was then performed for the MOC peak hot excess point aad the EOC 1000 mwd /STU exposure point to establish the maximum SLMCPR for the cycle.
* Comparison of Edwin I. Ilatch Nudear Plant Unit 1 Cycle 18 SLMCPR versus the Generie GE13 Value Table I summarizes the relevant input parameters and results of the SLMCPR determination for both the generic GE13 core and the Edwin L 11atch Nuclear Plant Unit 1 Cycle 18 core. Both generic and plant / cycle specific evaluations are performed using the methods described in GETAB, Reference [1]. The evaluations yield different calculated SLMCPR values because the inputs that are used are different. The quantities that have been shown to have some impact on the determination of the safety limit MCPR (SLMCPR) are provided. Much of this information is redundant but is provided in this case because it has been provided previously to the NRC to assist them in understanding the differences between plant / cycle specific SLMCPR evaluations and the generic values calculated previously by fuel product line. l[----
                                                   .ll Prior to 1996, GESTAR II, Reference [2] stipulated that the SLMCPR analysis for a new fuel design be performed for a large high power density plant assuming a bounding equilibrium core.
The GE13 product line generic SLMCPR value of 1.09 was determined according to this specification. An amendment to GESTAR II, Reference [8], has been submitted to the NRC to describe how plant / cycle specific SLMCPR analyses are used to confirm the calculated SLMCPR value on a plant /cych specific basis using the uncertainties defined in Reference [4].
IIL.5368                                                El-3

.   ' Enclosure 1 -
Basis for Change R6 quest e                                                                                                                                                                                                                          1 l
In comparing the generic GE13 and Edwin I. Ilatch Nuclear Plant Unit 1 Cycle 18 SLMCPR values it is important to note that the Edwin 1.11atch Nuclear Plant Unit 1 Cycle 18 core is not an equilibrium core. It is a mixed core with GE13 and GE9B fuel with 4 GE12 bundles. The freshest fuel is the latest batch of gel 3 that because of the transition to longer cycle lengths, comprises 33.6% of the bundles in the core. Also, the Edwin I. Hatch Nuclear Plant Unit 1 Cycle 18 core is loaded such that the fresh batch of GE13 has the highest enrichment (3.55%), as compared to a core average enrichment of 3.42%, as shown in Table 1. By way of comparison, the generic GE13 equilibrium core has batch and core average enrichments of 3.29%. liigher enrichment in the fresh fuel for the Edwin I. Hatch Nuclear Plant Unit 1 Cycle 18 core (compared to the rest of the core) produces higher power in the fresh bundles relative to the rest of the core. These enrichment differences result in the GE13 fresh fuel producing a higher relative share of the number of fuel rods calculated to be susceptible to boiling transition (NRSBT). For example, at the cycle exposure ofinterest, EOC-1K mwd /STU, the GE13 fuel accounts for 100.0% NRSBT while the GE9 fuel accounts for 0% NRSBT. Because of the dominance of the GE13 fuel in terms of %NRSBT, the Edwin I. Hatch Nuclear Plant Unit 1 Cycle 18 core is compared in Table I to the core used for the gel 3 generic SLMCPR evaluation.
The core MCPR distribution for the Edwin L Hatch Nuclear Plant Unit 1 Cycle 18 core is flatter                                                                                                              _
than the core MCPR distribution used in the GE13 generic analysis. ((.-..--                                                                                         --
      . . . . . . . . . . -            ...              - .-                                  .                     .         ...                   ....                                      .)) By              ~
each measure the Edwin I.11atch Nuclear Plant Unit 1 Cycle 18 core MCPR distribution is                                                                                                                             .
evaluated as being flatter. ((-                                                                                              ----            --------
      ......                  ......                                  ......-                                ......                               ....-           .-                    . jj
((.....            .........                             ............. ..........                                                                                      .          ......
      .. ... .............                             ... ...... .... . ...........-                                                                               .....                                . jj l[...              .. .... ..............                                ....                                                                                                               ..            .
                                                                                                               =. ll The Edwin 1. Hatch Nuclear Plant Unit 1 l
IIL-5368                                                                                             El.4

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1
                                                                                                                                                                                                                                                                 \
.   ' Enclosure 1 -                                                                                                                                                                                                                                              ;
1 Basis for Ch'ange RI: quest -                                                                                                                                                                                                                             1 4
Cycle 18 core has a flatter core MCPR distribution than the generic GE13 equilibrium core, but the bundle R. factor distribution is slightly less flat.
((..___
                                                                                                                                                .... j j.
((...__..          ..__.           .....___ .-                            ..                                                 .. .
      .-         .... .......__. .-                              ........==-                                                                           .                ....                        .....___....
      ......__...... ... . ..................................__.. ........__.                                                                                                                                                         ._____ . j j     .
((....__............                     .         ..........                   .                        .-          .......                                   ............                                                     .                    .
                                                                                                                                                          .]l one is lead to the conclusion                                                                  _
that the core MCPR distribution for Edwin 1. Ilatch Nuclear Plant Unit 1 Cycle 18 is significantly flatter than the distribution evaluated for the generic GE13 case.
The uncontrolled bundle pin-by-pin power distributions were compared between the Edwin I.
Ilatch Nuclear Plant Unit 1 Cycle 18 bundles and the GE13 bundle used in the generic SLMCPR llL-5368                                                                                                 El-5

,   --                -.                  ,-                  _. .         .       = . .               . . -.                                   .                       ..                .. ..- _- - .                         -
         ~
  .         Enclosure 1 -
Basisfor Ch'ange Rbquest analysis. Pin-by. pin power distributions are characterized in terms of R. factors using the                                                                                                                            j methodology defined in Reference [6]. For the Edwin I. Hatch Nuclear Plant Unit 1 Cycle 18                                                                                                                               '
bundles, there is a slightly flatter distribution of uncontrolled R. factors for the highest power rods in each bundle, which in the calculation are the rods most likely to be susceptible to boiling transition. This fact is difficult to ascertain graphically since the relative flatnesses are similar and the rods that have an R. factor closer to the R. factor for the lead rod are statistically worth much more than those that have R. factors that are further away. ((                                                                                    -                           .---                        -
           ...........................____...jj                                                                    .
The flatness of the pin R. factor distribution within a particular bundle is characterized ((-.....---
P I
h
           . _ _ . . . . _ _ _ . . . . . . . . _                                         ....... ...._..___..-                              .... ... .__...___ _ j j                ,
b
                                                                                           ....... )) Thus this supports the conclusion that the higher SLMCPR value is entirely attributable to the flatter MCPR distribution for the Edwin I.
Ilatch Nuclear Plant Unit 1 Cycle 18 core.                                                                                                                                                                     j i
HL.5368                                                                                            El.6                                                                                                                 i t
I

Enclosure 1 -
Basisfor Ch' ange Rbquest Table 2          Comparison of GE13 Generic and Edwin 1. Ilatch Nuclear Plant Unit 1 Cycle 18 Core and Hundle Quantities that Impact the SLMCPR ((                                                           _
QUANTITY, DESCRIPTION                                                        Generic                  Edwin I.
GE13                     IIntch Nuclear Plant Unit 1 Cycle 18 11                                                                                                                              -
Summary The calculated nominal 1.10 Monte Carlo SLMCPR for Edwin 1. IIatch Nuclear Plant Unit 1                                         _
Cycle 18 is consistent with what one would expect [l----------                            =-=          --     ---          ----
  ............... .................... ....................... .]l the 1.10 S LMCPR val ue is appropriate.
Various quantities ((----------..---------- llhave been used over the last year to compare                                      ~J quantities that impact the calculated SLMCPR value. These other quantities have been provided to the NRC previously for other plant / cycle specific analyses using a format such as that given in Table 1. These other quantities have also been compared for this core / cycle ((-                                     -------
                                          . .)) Note that regardless of what comparisons are made the conclusion is reached that the Edwin I. Hatch Nuclear Plant Unit 1 Cycle 18 core / cycle has a much flatter radial power distribution than was used to perform the GE13 generic SLMCPR evaluation. This conclusion is significant enough to more than compensate for the fact that the IIL.5368                                                         El-7

  ,   ' Encl'osure l-Basisfor Ch'ange Request Edwin I. Ilatch Nuclear Plant Unit 1 Cycle 18 bundles are slightly less flat than the bundles used for the generic GE13 SLMCPR evaluation.
Based on all of the information presented above, it is concluded that the calculated SLMCPR value of 1.10 for the Edwin 1. IIatch Nuclear Plant Unit 1 Cycle 18 transition core is appropriate.
For single loop operations (SLO) the safety limit MCPR is 0.02 greater than the two-loop value as detennined by specific calculations for Edwin I. Ilatch Nuclear Plant Unit 1 Cycle 18. An SLO adder of 0.02 for SLO is also what one would expect based on a generic assessment of SLO adders.
        . Conclusion 1       Based on all of the information presented above, it is concluded that the calculated SLMCPR value of 1.10 for the flatch-1 Cycle 18 transition core is appropriate.

==REFERENCES:==
: 1. General Electric BWR Thermal Analysis Basis (GETAB): Data, Correlation and Design Application, NEDO-10958-A, January 1977.
: 2. General Electric Standard Application for Reactor Fuel (GESTAR II), NEDE-24011-P-A-11, November 1995.                                                                                  1
: 3. General Electric Standard Application for Reactor Fuel (GESTAR II), NEDE-24011-P-A-13, August 1996.
: 4. General Electric Fuel Bundle Designs, NEDE-31152-P, Revision 5, June 1996.
: 5. Methodology and Uncertainties for Safety Limit MCPR Evaluations, NEDC-32601P, Class Ill, December 1996.
: 6. R-Factor Calculation Method for gel 1, GE12 and GE13 Fuel, NEDC-32505P, November 1995.
: 7. General Electric letter to NRC,10CFR Part 21. Reportable Condition, SLMCPR Evaluations, dated May 24,1996.
: 8. General Electric letter to NRC, R. J. Reda to T. E. Collins," Proposed Amendment 25 to GE Licensing Topical Report NEDE-24011-P-A (GESTAR 11) on Cycle-Specific Safety Limit MCPR," dated December 13,1996.
IIL-5368                                         El-8 l

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J        .'                               Enclosure 2 Edwin I. Hatch Nuclear Plant - Unit 1 Request to Revise Technical Specifications:
Safety Limit Minimum Critical Power Ratios (SLMCPR) 10 CFR 50.92 Evaluation in 10 CFR 50.92(c), the NRC provides the following standards to be used in determining the existence of a significant hazards consideration:
             ...a proposed amendment to an operating license for a facility licensed under
{50.21(b) or {50.22 or for a testing facility involves no significant hazards consideration, if operation of the facility in accordance with the proposed amendment would not: (1) Involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) Create the possibility of an accident of a new or different kind of accident from any accident previously evaluated; or (3) Involve a significant reduction in a margin of safety.
Southern Nuclear Operating Company has reviewed the proposed license amendment requests and determined their adoption does not involve a significant hazards consideration based on the following discussion.
Basis for no signiReant hazards consideration determination
: 1. The proposed technical specification changes do not involve a significant increase in the probability of an accident previously evaluated.
The derivation of the revised SLMCPR for Plant Hatch Unit 1 Cycle 18 for incorporation into the TS, and its use to determine cycle-specific thermal limits, have been performed using NRC approved methods. Additionally, interim implementing procedures that incorporate cycle specific parameters have been used which result in a more restrictive value for SLMCPR. These calculations do not change the method of operating the plant and have no effect on the probability of an accident initiating event or transient.
The basis of the MCPR Safety Limit is to ensure no mechanistic fuel damage is calculated to occur if the limit is not violated. The new SLMCPR preserves the existing margin to transition boiling and the probability of fuel damage is not increased. Therefore, the proposed changes do not involve an increase in the probability or consequences of an accident previously evaluated.
llL-5368                                        E2-1

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      $nclosu(e 2 10 CFR 50.92 Evaluation
: 2. The proposed TS change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
The proposed changes result only from a revised method of analysis for the Unit 1 Cycle 18 core reload. These changes do not involve any new method for operating the facility and do not involve any facility modifications. No new initiating events or transients result from these changes. Therefore, the proposed TS changes do not create the possibility of a new or difTerent kind of accident from any accident           ,
previously evaluated.
: 3. The proposed TS changes do not involve a significant reduction in a margin of safety.
The margin of safety as defined in the TS bases will remain the same. The new SLMCPR is calculated using NRC approved methods which are in accordance with the current fuel design and licensing criteria. Additionally, interim implementing procedures, which incorporate cycle-specific parameters, have been used. The              ;
SLMCPR remains high enough to ensure that greater than 99.9% of all fuel rods in the core are expected to avoid transition boiling if the limit is not violated, thereby preserving the fuel cladding integrity.
Therefore, the proposed TS changes do not involve a reduction in the margin of safety.
The proposed change has been reviewed and recommended for approval by the Plant flatch Plant Review Board and reviewed by the Safety Review Board.
ENVIRONMENTAL IMPACT The proposed technical specification changes were reviewed against the criteria of             l 10CFR51.22 for environmental considerations. The proposed changes do not involve a             ,
significant hazards consideration, a significant increase in the amounts of effluents that     ;
may be released offsite, or a significant increase in individual or cumulative occupational radiation exposures. Based on the foregoing, Southem Nuclear Operating Company concludes the proposed Technical Specifications meet the criteria given in                     !
10CFR51.22(c)(9) for a categorical exclusion from the requirements for an Environmental Impact Statement.
111.-5368                                     E2-2                                             ,

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   .             ' Snclosure 2                                                                                                   I
 -                 10 CFR '50.92 Evaluation                                                                                       i k
CONCLUSION Based on the evaluation above: (1) there is reasonable assurance the health and safety of j                 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
 -                proposed amendment will not be inimical to the common defense and security or the                            . ,
health and safety of the public.
9 4                                                                                                                                .
SCHEDULE OF CHANGE This amendment is needed to support Unit 1 Operating Cycle 18 following refueling and will be implemented following receipt of NRC approval. Due to its significant impact on
                 - power operation, we request expedited NRC review and approval.
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Enclosure 3 Edwin I. Hatch Nuclear Plant - Unit 1 Request to Revise Technical Specifications:
Safety Limit Minimum Critical Powers Ratios (SLMCPR)
Pace Chance Instructions Cycle 18 Unit 1 Pace                                                Replace 2.0.I                                               2.0-1 IIL-5368                             E3-1 5}}