ML20087Q007

Application for Amend to License NPF-5,revising Tech Specs to Add New Fuel Type MAPLHGR Curve & Change Min Critical Power Ratio Curves for 8x8R & P8x8R Fuel to Support Reload 4 Core Design & Allow Hybrid I Control Rod Operation

ML20087Q007
Person / Time
Site:	Hatch
Issue date:	04/03/1984
From:	Beckham J GEORGIA POWER CO.
To:	Stolz J Office of Nuclear Reactor Regulation
Shared Package
ML20087Q008	List: ML20087Q007 ML20087Q011
References
NED-84-192, TAC-54169, TAC-54607, NUDOCS 8404100202
Download: ML20087Q007 (9)

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, Geofgta Power Company --

333 Psedmont Avenue At:anta, Georg.a 30308 Telephone 404 526;7020 4

, Maihng Address-Post Office Box 4545

. At;anta. Georgia 30302 Georgia Power the southem electnc system J. T. Beckham, Jr.

'Vice President and General Manager Nuclear Generation NED-84-192 April 3, 1984 i

Director of Nuclear Reactor Regulation Attention: Mr. John F. Stolz, Chief Operating Reactors Branch No. 4 Division of Licensing U. S. Nuclear Regulatory Ccanission Washington, D. C. 20555 i NRC DOCKET 50-366 OPERATING LICENSE NPF-5 EDWIN I. HA'ICH NUCLEAR PIANT UNIT 2 REQUEST 'IO CHANGE 'IECHNICAL SPECIFICATIONS i POUR'IH CORE REIDAD

-Gentleen:

In accordance with the provisions of 10 CFR 50.90 as required by the provisions of 10 CPR 50.59 (c) (1) , Georgia Power Capany (GPC) hereby l . proposes mendments to the Edwin I. Hatch Unit 2 Technical Specifications (Appendix A to the Operating License). %e changes would: -1) ' add a new fuel type Maximm Average Planar Linear Heat Generation Rate -(MAPLHGR) curve

. and change the Minimm Critical Power Ratio (MCPR) curves for 8x8R and P8x8R fuel to support operation of the core design for Hatch 2 Reload 4; 2) allow

, - operation with the new Hybrid I Control Rod Assably; and 3) allow up to l

four fuel assemblies to be reloaded around the Source Range Monitor (SIE) so that the required 3 counts . per second (cps) for fuel loading can be established.

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.%e Plant Review Board has evaluated these proposed' Technical Specification changes and has determined that the implementation ' of these proposed changes 'would not constitute;an unreviewed_ safety question for the reasons stated below.

%e MAPIJ1GR curve was generated using the methodology and acceptance criteria specified by the NRC approved licensing docments, GESSAR II and -

GESTAR. %e MCPR curve was chosen to bound the results of the reload core design thermal limits which will be derived in accordance with the reference doceents. ~ Consequently, neither the probability of nor the_ conseguences of

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an accident are. increased above those analyzed in the FSAR. Additionally, the margin of safety is maintained because the acceptance criteria specified I by .those- licensing .docments is met. Because no plant system design .is changed, no new type of accident or malfunction is created.

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t Georgia Power k m Director of Nuclear Reactor Regulation i Attention: Mr. John F. Stolz, 011ef

. Operating Reactors Branch No. 4

. April 3, 1984 Page h o On August 22, 1983, the NRC issued a letter reporting their review and acceptance of NEDE-22290, " Safety Evaluation of the General Electric

, Hybrid I control' Rod Assenbly." % e NRC concluded, " Based on our evaluation of ' the information provided - in (a) NEDE-22290, (b) a meeting with GE

- representatives, and (c) responses to NRC staff questions, we conclude that there . is reasonable assu ance that the substitution of Type I HICRs for .

other approved GE control blades will not result in unacceptable hazards to 4

the public and should, in fact, result in improved control blade performance and a positive contribution to reactor safety. %erefore, NEDE-22290, as mended to incorporate this safety evaluation, is approved as a referential doctment for the GE type I HICR." Thus, this change is not an unreviewed 4 safety issue.

%e third change allows up to four fuel assenblies to be reloaded around the SINS in order to establish the 3 cpp required for SIM operability.- his i is very similar to the request made earlier and granted by Anen&nent 26 to the Unit 2 license that allowed'two bundles to be loaded diagonally in order to establish SIM operability. Because the unit will have been in an outage for approximately 6 months when it is restarted, two bundles may not- have been enough to establish' the - requisite counts on the SIN. GE spent fuel pool' studies reported in . Chapters 4 and 9 of GESSAR-NEDO-10741 established that any 2x2 uncontrolled array with maxistan reactivity bundles will always renain subcritical (Koo less than .95). .%e bundles that go back around the SINS are the sane bundles that left those locations.- herefore, they will remain subcritical following reinsertion because they were subcritical before they were .- renoved. Consaluently, the probability of occurrence or

~the consequence of an accident is not increased above those analyzed in the ESAR. Also, the margin of safety has not been reduced by using four instead of two bundles because the arrangenent is significantly subcritical.

Because no new mode of operation' or change in plant design occurs, no new type of accident is introduced.

Included - with this proposal is a determination of anendnent class (attachnent ~ 7) . We have determined 1 this to be a ' class IV anendment.

l Appropriate payment,is enclosed.

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GeorgiaPowerkh Director of Nuclear Reactor Regulation Attention: Mr. John F. Stolz, Chief Operating Reactors Branch No. 4 April 3, 1984 Page 'Ihree Instructions for incorporation of these changes along with copies of the affected Technical Specification pages are enclosed (Attachment 3).

Pursuant to the requirenents of 10 CFR 50.92, J. L. Ledbetter of the Georgia Department of Natural Resources will be sent a copy of this letter and all applicable attachments.

J. T. Beckham,'Jr. states that he is Vice President of Georgia Power Company and is authorized to execute this oath on behalf of Georgia Power Canpany, and that to the best of his knowledge and belief the facts set forth in this letter are true.

GB0fCIA POWER O NPANY By:

[/ J. T. Beckhan, M.

Sworn to and subscribed before me this 3rd day of April,1984.

YNh- _ l.53x = -

Notary Putsc.Georps. Siete at Large Notary Public Ah Commesion Empres Aug.26,1906 DLT/inb t

' Enclosure ,,

xc: H. C. Nix, Jr.

Senior Resident Inspector J. P. O'Reilly, (NRC-Region II)

J. L. Ledbetter

GeorgiaPowerkn ATTACINENT 1 NRC DOCKETS 50-366 OPERATING LICENSE NPF-5 EDWIN I. HA'ICH NUCIEAR PLANT UNIT 2 REQUEST 'IO CHANGE 'IECHNICAL SPECIFICATIONS FOURTH OORE REIDAD Pursuant to 10 CFR 170.12, Georgia Power Cmpany has evaluated the attached proposed mendnents to Operating License NPF-5, and has determined that:

a. 'Ihe proposed mendnents do not require evaluation of a new Safety Analysis Report and rewrite of the facility license;

b. 'Ihe proposed mendnents do not contain several emplex issues, do not involve ACRS review, and do not require an enviromental impact statenent; 4

c. The proposed mendnents do involve more than one safety issue incorporating three changes of the Class III type;

d. Therefore, the proposed mendments result in a Class IV mendment.

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P GeorgiaPower A ATTACIMNT 2 NRC DOCKETS 50-366 OPERATING LICENSE NPF-5 EDWIN I. HATCH NUCIEAR PIANT UNIT 2 ausguesi 70 GANGE TEDINICAL SPECIFICATIONS rvunem CORE REIOAD Pursuant to 10 CER 50.92, the following statenents provide a sunnary of and the basis for the proposed changes:

1. . Add a figure to define the ' Average Planar Linear Heat Generation Rate limit (MAPIRGR) for fuel type P8DRB284H.

BASIS:

Georgia Power Canpany proposes to add a curve of MAPIEGR vs. planar exposure for . P8DRB284H fuel (Hatch 1 Reloads 5 and 6 fuel) derived on the basis of the Unit 2 systen IDCA response. This change is requested in connection with the core reloading of Unit 2 to allow for introduction .of a new fuel ' type. Fuel type P8DRB284H is a standard General Electric design as described in the MtC approved fuel licensing doctanent, GESTAR II (NEDE-240ll-P-A-6) . The calculated peak ; clad temperatures for this fuel type correspond to the criteria specified in 10CFR50, Appendix K and in a letter fran J. F. Stolz (USNRC) to J. T.

Beckhan (GPC), dated February 3, 1982; therefore, MAPIRGRs may be

. . defined for exposures greater than 30,000 Mwd /st.

'Ihe proposed MAPIRGR limits were calculated by General Electric using methods consistent with approved analyses _ of the . loss of coolant-

_ accident and anticipated operational transients given L in the Hatch-2 PSAR, and all applicable requirenents stated therein' are met by 'the proposed values. Application of the MAPIRGR limits will not result in any reduction of the margin of safety. or cause any change in the conseguences for- postulated accidents and transients, because_ all acceptance criteria as ' defined above are met. No design changes to the; -

. plant or procedural changes are involved with this part of the snendment. -Therefore, the- probability of occurrence of previously

. considered events would renain unaffected. Because no new failure modes i

would be.introduIced, the possibility of a new type of accident would not be created.

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k, GeorgiaPowerkh MAPIllGR limits for P8DRB284H are presently in the Hatch Unit 1 Technical Specifications and the fuel type has been irradicated in the Unit 1 core. As described above, no changes have been made to the acceptance criteria for the Technical Specifications or to the analytical methods used to deonstrate conformance with the Technical Specifications and  ;

regulations and the NRC has previously found the methods acceptable, j t

Conseguently, this change is associated with a refueling and thus is consistent with Its lii of the "Exmples of Amendments that are Considered Not Likely to Involve Significant Hazards Considerations" listed on page 14870 of the April 6, 1983 issue of the Federal Register and will not result in a significant hazardds consideration.

2. Increase the operating limit Minimm Critical Power Ratio (OIMCPR) for 8x8R and P8x8R fuel.

BASES:

The Operating Limit MCPRs at rated. core flow are defined for 8X8R and P8X8R fuel in Figures 3.2.3-1 and 3.2.3-2 of the Hatch-2 Technical Specifications. % e Option B (t =0) limits are presently 1.26 for 8X8R and 1.27 for P8X8R fuel; Option A (t =1) limits are 1.32 for 8X8R and 1.35 for P8X8R. %e paraneter tis related to the results of timing of control rod insertion speeds and is defined in Section 3.2.3 of the Technical Specifications. It is proposed to change the OIMCFRs for both of these fuel types to 1.29 for Option B and 1.37 for Option A, with linear interpolation as a function of t .

%e intent of this change is to allow for licensing the fourth Hatch-2 reload under 10CFR50.59 when final design and licensing work are cmpleted. Based on review of the anticipated fuel mix, and on the transient analysis input parameters for the fourth reload, cmpared to the results of transient analyses performed for previous reloads of both Hatch units, it is judged that the proposed limits will conservatively bound the OIMCPRs that recult fra licensing analyses of Hatch-2 reload 4 and subsequent Hatch-2 core reloads.

Conformance with the proposed MCPR operating limits shall be assured prior to reactor startup by analyses of limiting operational transients, using the analytical methods given in the approved fuel licensing

- docment GESTAR II (NEDE-240ll-P-A-6) . Application of these verified OLMCPRs will not cause any reduction of the margin of safety or produce any changes in the consequences of postualted accidents and transients l _because all acceptance criteria as defined above are met. No design changes to the plant or procedural changes are involved with this part j- of the amendment.- %erefore, the probability of occurrence of previously considered events 'would renain unaffected. Because no new failure modes would be introduced, the possibility of a new type of l accident would not be created.

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. GeorgiaPower d

%is change is requested in connection with the core reloading of Unit 2. Consequently, this change is consistent with Iten (iii) of the

" Examples of - Amendnents that are Considered Not Likely to Involve Significant Hazards Considerations" listed on page 14870 of the April 6,

'1983 issue of the Federal Register and .will not result in a significant hazards consideration.

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3. Change the description of the control rod assenblies in Section 5.3.2 (Design ' Features) of the Hatch-2 Technical Specifications to delete

. references to the specific materials and details of construction of the

. control blades.

BASIS:

%is change is intended to support the use in Hatch-2 of an arbitrary ntsober (up to 137) Type I General Electric Hybrid I Control Rod - (HICR) assemblies containing some hafnitun as absorber material in place of the boron carbide control rods presently in use. HICRs are intended to be

. standard replacanent control rod assenblies for the General Electric BWR/4 D-lattice in operating reactors. @ e HICRs form, fit and function are identical to that of the blade it replaces. %e HICR is designed to increase control rod assembly life and to eliminate cracking of absorber tubes containing boron carbide (B4C) . %e essential differences between the HICR and the BWR Z-4 D-lattice controls - rod assemblies currently in use are:

a)- improved B4C absorber rod tube material to eliminate cracking during the lifetime of the control rod assenbly, and b) sane B4 C ~ absorber rods' are replaced with solid hafnium absorber rods to increase blade life.

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Other minor material and dimensional changes are described in detail in NEDE-22290-A, " Safety Evaluation of the General Electric Hybrid-I Control Rod Assenbly," Septenber 1983.

Adherence to the guidelines established for replacement of the standard B4 C control. blades may require that blades in certain core locations be replaced at each refueling outage. ~

It is expected that use of the HICR . blades in- these locations will allow operation of at least two

~18-month fuel cycles without replacing those blades, thus reducing

outage time, equignent duty and personnel exposure otherwise required

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, for blade replacanent.

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%e' details -of design and materials for the new blades will not be included in the revised text, because those details are unnecessary and 4

inconsistent with other portions of the Design Features Section which do not provide design or materials details. Safety design bases which must be . met by control- rods are enumerated in the Hatch-2 FSAR Chapter 4.

Analyses doceented in the approved topical report,aNEDE-22290-A, have shown that those design lkses are met by the HICR blades, therefore, use i of these blades will cause no reduction in the margin of safety.

For exmple, the HICR weight and rod worth are the see as those for the ,

currently used control rod acaembly. % ecefore, the scran speed and scr e reactivity are also the sm e. It follows then that the IRGR, MCA and MAPIRGR limits are not affected by the HICR.

Because the control rod worth is the sane, the capability of the reactor i to achieve the Design Basis cold shutdown reactivity margin is not affected. In addition, existing methodology for analysis of the control l

rod withdrawal error transient and the control rod drop accident renains valid with HICR assemblies installed. It follows then, that the probability of or conseguences of all accidents and transients

. previously evaluated in the FSAR will not ~ be affected by use of the HICRs.

l %e possibility of occurrence of an accident .different than any evaluated in the FSAR is not created by use of the HICR assenblies,

, because there is no functional change in the' control rods.

t As shown above, use of the HICR assenblies' in Hatch-2 does not increase the probability or consequences of a previously analyzed acccident, nor does it significantly reduce any safety margin. %e result of - this design change is' clearly within all acceptance criteria given in the Hatch-2 FSAR as noted above.-

Consequently, this change will not result. in a significant hazards ,

consideration.

. - 4. Change the number of fuel assenblies that can be loaded around a S M in L

order to assure that 3 counts per second (cps) can be achieved without the use of additional sources or dunking chanbers. ,

,. BASIS:

f%e_~ four Sm detectors are located, one per quadrant, roughly half a core radius fran the center. Although these are incore detectors and thus very sensitive when_ the reactor -us fully loaded, they lose sane of -

l; their _ effectiveness when the reactor- is partially defueled and the detectors are located some distance fran the array of renaining fuel. ,

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GeorgiaPower d GE's spent fuel pool studies, GESSAR - NEDO-10741, Chapters 4 and 9, show that: 1) sixteen or more fuel assenblies (i.e., four or more control cells) must be loaded together before criticality is possible; and 2) for an uncontrolled 2x' array of maximun reactivity bundles, Keim will always be less than .95. .In spiral loading squences in the Hatch core, an array containing four or more control cells will be at most two control cells (i.e., about two feet) away from an SM detector. %e sensitivity loss in such a case is at most one decade of sensitivity (i.e., about one fifth of the Sms logarithnic scale) . % is means that criticality cannot be reached during a spiral reload without an operable SM detecting it. A sprial sequence is any sequence in which the central control cell is last unloaded and first reloaded, all fueled locations are contiguous, and no imbedded cavities or major peripheral 4

concavities are permitted.

%e Hatch 2 Technical Specifications rquire that the fuel assenblies be loaded into their previous core position next to each of the four Sms.

%e loading of the bundles around the Sms before attaining the 3 cps is permissible because these bundles were in subcritical configuration when they were removed and therefore will remain subcritical when placed back in the previous position. Wis request is very similar to an earlier rquest by Plant Hatch Unit 2 which was granted as Amendnent No. 26 for loading 2 bundles next to the SEs. Because Hatch-2 has been in an extended outage, more than 2 bundles may be rquired to establish the requisite 3 cps on the Sms so that fuel loading may proceed.

%e possibility of occurrence of an accident different than any evaluated in the ESAR is not created because there is no design change to any plant systens. %is change does not significantly increase the probability or consquences of a previously analyzed accident because the referenced studies demonstrate inadvertent criticality with 4 bundles is not possible and further the sme subcritical assenblies and arrangenent that was discharged is returned to the sane core location.

Finally, the safety margin is not significantly reduced because the bundles renain significantly suberitical. Consequently, this change does not represent a significant hazards consideration.

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