ML20236B787

Application for Amends to Licenses DPR-57 & NPF-5,revising Tech Specs Re Scram Time Testing in Order to Achieve Consistency W/New NRC Approved GE Transient Analysis Methodology.Fee Paid

ML20236B787
Person / Time
Site:	Hatch
Issue date:	10/21/1987
From:	James O'Reilly GEORGIA POWER CO.
To:	NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
Shared Package
ML20236B791	List: ML20236B787 ML20236B809 ML20236B835
References
SL-3197, TAC-66524, TAC-66525, NUDOCS 8710260355
Download: ML20236B787 (18)

Text

. - _ .

" Georgia Power Company 303 Piedmont txenue J Atlanta, Georgio 30308 Telephone 404 ! 26 7851 j Mailing Address Pbst Ofhce Box 1545 Atlanta, Georgia 30302 h$NRC.Q3 l.

IE OCI 2b A 9 50 James P. O'Relll) >

the southern 14ttreic system Senior Vice President nuciear operanor s SL-3197 1663C 1 X7GJ10-H000  !

October 21, 1987 U. S. Nuclear Regulatory Commission '

ATTN: Document Control Desk Hashington, D. C. 20555 PLANT HATCH - UNITS 1, 2 NRC DOCKETS 50-321, 50-366 OPERATING LICENSES OPR-57, NPF-5 REQUEST TO REVISE TECHNICAL SPECIFICATIONS:

FUEL THERMAL LIMITS AND REFUELING OPERATIONS Gentlemen:

i In accordance with the provisions of 10 CFR 50.90, as requir.ed by 10 CFR 50.59(c)(1), Georgia Power Company hereby proposes changes to the Plant Hatch Units 1 and 2 Technical Specifications, Appendix A of each of the Operating Licenses, DPR-57 and NPF-5.

The proposed changes would involve a revision to the Unit 2 Technical 4 Specifications regarding scram time testing in order to achieve l consistency with the new NRC approved General Electric transient analysis methodology. The changes would also allow for .the insertion of four Lead Fuel Assemblies (LFAs) into the Unit 2 reactor during the upcoming  ?

refueling outage. Also proposed are changes to the requirements concerning the loading of fuel assemblies inserted around the SRMs prior to a full core reload on both Units 1 and 2. .

I Specifically, the proposed changes to the Technical Specifications would:

1. Revise the formula used to determine the statistical scram speed j limit and reference applicable plant procedures for detailed discussion (Unit 2).

2. Revise the Option A MCPR limit for 8x8 fuel and add coverage for LFAs (Unit 2).

3. Revise both the APLHGR limits curve for an existing fuel type and the MAPFACF curve to include coverage for LFAs (Unit 2).

4. Allow the loading of any four fuel assemblies which meet certain analytical criteria around each SRM prior to a full core reload (Units 1 and 2).

hoo;~

L 8710260355 871021 C

, l DR ADOCK 0500 1 1/ W y, Qoo

4 U.S. Nuclear Regulatory Commission October 21, 1987 Page Two These proposed revisions will allow Georgia Power Company to take full advantage of GE's NRC approved GEMINI methodology, improve plant refueling operations, and examine fuel from a new fuel vendor in the Plant Hatch Unit 2 reactor.

Enclosure 1 provides detailed descriptions of the proposed changes and circumstances necessitating the change request.

Enclosure 2 details the bases for our determination that the proposed changes do not involve significant hazards considerations.

Enclosure 3 provides page change instructions for incorporating the proposed changes.

The proposed changed Technical Specifications pages for Unit 1 and Unit 2 follow Enclosure 3.

To support proposed change 3 of Enclosure 1, Enclosure 4 is Advanced Nuclear Fuels Corporation report ANF-87-95, Revision 1, " Hatch 9x9 Lead Fuel Assemblies Safety Analysis Report", July 1987.

Payment of the filing fee in the amount of one hundred and fifty dollars is enclosed.

In order to allow time for procedure revisions and orderly incorporation into copies of the Technical Specifications, we request that the proposed amendment, once approved by the NRC, be issued with an effective date to be no later than 60 days from the date of issuance of the amendment.

Certain changes described in Enclosure 1 are necessary to support the i reloading of Plant Hatch Unit 2, scheduled for January 13, 1988, and therefore, we request priority approval of these 2 items:

1. Modification of the formula used to calculate the statistical scram speed limit to reflect the use of GE's NRC approved GEMINI methodology.

2. Revisions to the MAPLHGR, MAPFACF and MCPR thermal limits curves to include coverage of the ANF 9x9 Lead Fuel Assemblies.

1663C

l f t 1

]

s. J U.S. Nuclear Regulatory Commission n October 21, 1987 Page-Three In accordance with the requirements .of.10 CFR. 50.91,, a copy of this; i letter and all . applicable enclosures will be sent to Mr. J. L. Ledbetter .

of the Environmental Protection- Division of .the' Georgia Department of Natural Resources' '

Mr. James . P. O'Reilly states - that. he is Senior Vice President.-l t of Georgia Power Company and is authorized to execute this oath'on behalf of Georgia Power Company, and that to. the best of his knowledge and belief '

the facts set forth in this letter are true..

GEORGIA POWER COMPANY _ j

'By: m' (O' T l James P. O'Reilly Sworn to and subscribed before me t st day of Octob 19A7.-

a nu J Notary Public Notary Pubbc, Chyton County. Geergia My Commission Emun oec. 12,1989 ~;

GDP/lc '

Enclosures:

1. Basis for Change Request

2. 10 CFR 50.92 Evaluation

3. Page Change Instructions

4. Supporting Documentation

5. Filing Fee - $150.00 I

c: Georaia Power Comoany l

{

Mr. J. T. Beckham, Jr., Vice President - Plant Hatch _j GO-NORMS U. S. Nuclear Rea.glatory Commission. Washinaton. D. C.

]

Mr. L. P. Crocker, Licensing Project Manager j Hatch O. S. Nuclear Reaulatory Commission. Reaion II Dr. J. N. Grace, Regional Administrator Mr. P. Holmes-Ray, Senior Resident Inspector - Hatch State of Georaia-Mr. J. L. Ledbetter, Commissioner - Department of Natural Resources 1663C

l

ENCLOSURE 1 ,

4 PLANT HATCH - UNITS 1, 2 NRC DOCKETS 50-321, 50-366 OPERATING LICENSES DPR-57, NPF-5 l REQUEST TO REVISE TECHNICAL SPECIFICATIONS: 1 FUEL THERMAL LIMITS AND REFUELING OPERATIONS 'l BASIS FOR CHANGE REOUEST PROPOSED CHANGE 1:

Proposed Change 1 revises the Minimum Critical Power Ratio (MCPR) scram time parameters given in the current Unit 2 Technical Specifications.

One of these limits,TB, is the statistical scram speed ; limit which ,

forms the basis for a reduction in the Operating Limit MCPR (OLMCPR) due j to faster scram . speeds relative to the minimum speeds required by the  ;

Technical Specifications.  !

At present, the formula used to calculate tB is defined in the  :

Technical Specifications as: j l

i j

1/2  !

TB = 0.834 + 1.65 N 1__ (0.059) n l

I Nj l l 11 '

where: 0.834 = the mean of the scram speed distribution (designated ,

as p) assumed in the transient analyses using NRC l approved GENESIS methods to calculate the Option B MCPR.

0.059 - the standard deviation (designated as o) of the scram speed distribution.

General Electric derived the values listed above by evaluating the BWR industry data base for scram speeds. Since the derivation of the y and a for the GENESIS set of methods, the scram speed data base has expanded, and new _ values for p and a have been determined. To simplify the Technical Specifications, since the formula used for the determination of x is controlled by plant procedures, we propose that the revised formula for calculating TB be placed in the Bases section 1663C El-1 10/21/87 SL-3197

.. q l

ENCLOSURE 1 (Continued)

REQUEST TO REVISE TECHNICAL SPECIFICATIONS FUEL THERMAL LIMITS AND REFUELING 0PERATIONS l BASIS FOR CHANGE RE00EST. '!

q of the Technical Specifications and a footacte be added to the MCPR i Limiting Condition for Operation section of the Technical Specifications l referencing the plant procedures for a detailed discussion of the j determination of t. For consistency, the definitions of x, x ,

and 24 would also be moved to the Bases section of the TechnIN1 Specifications, even though these values are unaffected by-a reevaluation of the scram time data base.

BArkaround for Procosed Chanae 1:

General Electric has determined that the. change in .MCPR 'is dependent in part upon the speed of control rod insertion during the reactor scram which occurs during a transient. In the GENESIS methodology for calculating the Option A OLMCPR, GE assumed that all control rods would l be inserted at the speed given in the Technical Specifications. However, '

scram speeds . for most BWRs are significantly faster than the speeds required by the Technical Specifications. Therefore, to take credit for this, the Option B method of calculating OLHCPR was devised. For BHR plants that can demonstrate scram speeds faster than those required by the Technical Specifications, a considerably lower Option B OLMCPR is allowed.

The scram speed associated with Option B, tB, is determined from a statistical distribution of scram speeds which have occurred at BWR plants. Thus, TB, which is defined as the Option. B acceptance criterion for the average time required for rods to travel from their fully withdrawn position to notch position 36, depends upon the mean and the standard deviation of the distribution of scram speeds assumed in the Option B HCPR calculation. The form of the equation is:

tB - p + A(n) o, where:

p - the mean of the scram speed distribution assumed by GE in the Option 8 MCPR calculation.

o - the standard deviation of p.

A(n) - a function of the number of rods tested; as the number increases, A(n) decreases, thus TB decreases.

1663C El-2 10/21/87 SL-3197

l l

ENCLOSURE 1 (Continued)

REQUEST TO REVISE TECHNICAL SPECIFICATIONS 1 FUEL THERMAL LIMITS AND REFUELING OPERATIONS

{

BASIS FOR CHANGE RE00EST a The recently approved GEMINI transient analysis methodology employs (

Option B statistical adjustment factors which were derived using a scram ',

speed data base which was broader than the data base used'in deriving the GENESIS adjustment factors. As a result, the mean and the standard i.

deviation of the . distribution used in the analyses have changed. .To simplify the Technical Specifications, it is' proposed that the formula. l for calculating TB be moved to the Bases section, since the formula f is controlled by plant procedures, and a footnote be added .to the' MCPR ,

Limiting Condition for Operation Section of the Technical . Specifications )

indicating that the. formula is provided' in the plant procedures. The j scram time testing requirements will remain unchanged. l.

Basis for Procosed Chance 1:

The change in the scram speed data base is reflected in GE's submittal to i

the NRC concerning the GEMINI application methodology for deriving OLMCPRs from results of transient calculations (Reference 1). The GEMINI i methods and their applications, including improved scram times, have been l approved by the NRC (References 2 and 3).

The proposed Technical Specifications changes will not alter GPC's method )

of determining that the average scram speeds at Plant Hatch conform to the distribution assumed in the licensing analyses for Option B. If i Option B limits are exceeded (i.e., tave greater than TB), the application of either the GENESIS or the GEMINI methodology contains a ,

provision for increasing the OLMCPR up to a value based -upon Technical l Specifications scram speeds. To date, Plant Hatch operating data indicate that scram speeds used in the GEMINI Option B analyses are conservative (i.e., the Plant Hatch scram speeds are typically faster than the scram speeds assumed in the GEMINI Option B analyses).

Although the value for TB will change because the values of p and o are different, the definitions for t, Tave, and TA Wi l.1 remain the same.

PROPOSED CHANGE 2:

In accordance with the NRC-approved new GEMINI methodology, GPC proposes-to reduce the Unit 2 Option A Operating Limit Minimum Critical Power Ratio (0LMCPR) from 1.37 to 1.33 for all Hatch 2 8x8 fuel. The Option B OLMCPR will remain at 1.29 for all Hatch 2 8x8 fuel. Although Unit 2 1663C El-3 10/21/87 SL-3197

q b

ENCLOSURE 1 (Continued).

. REQUEST TO REVISE TECHNICAL SPECIFICATIONS FUEL THERMAL LIMITS AND REFUELING OPERATIONS

. BASIS FOR CHANGE REOUEST-

'I has 'usually operated .with tav the change assumed in ' the GEMINI analyses'e may:less tB,probability; that.inTave

.than' .the increase TB will be greater than xg at some' time during an operating cycle. In that case, 'the OLMCPR ~w111 fall between .the' Option A and 'the Option B limit. Under these . conditions, Plant ~ Hatch Unit '2 operation, would o

benefit from an Option A limit which is.as small as possible. Therefore, L it is proposed .that Option A be. calculated using the approach..which .is. 1 consistent with the GEMINI methodology. .l ;

l Backaround for Proposed Chance 2:: j To account for the model uncertainties. associated with GENESIS, the NRC t imposed a 4.4-percent penalty on the calculated MCPR. This approach' became known as the Option A procedure for calculating-the MCPR operating limit. However, for those utilities that could demonstrate; scram ' speeds j faster than the Technical Specifications requirements, the' NRC approved i

an alternative procedure, known as Option B, for calculating 'the MCPR operating limit. As - planned, the Option B' limit was lower than the Option A limit. Depending upon the measured average scram speed of .the rods, the OLMCPR was determined to be greater than or equal to the Option  ;

B limit but less than or equal to the Option A limit. l The application of the new GEMINI methodology . improved the approach used to determine the Option A limit for pressurization transients. For all pressurization transients analyzed for Plant Hatch, the Option A limit is 1 never more than 0.04 greater than the corresponding Option B limit. The j details of the new Option A calculation are described in References 1 and j 4.

l Basis for Prooosed Change 2: . j Even though GEMINI' may calculate an.0ption B MCPR which is lower-than the value calculated with GENESIS, the Option B Technical -Specifications f limit will not be changed at this time. Reducing the Option A OLHCPR by 1 0.04 relative to its current.value while maintaining- a 0.04 margin to the  !

current Option B limit, is' expected to provide additional operating j flexibility while still maintaining margin to the results of Unit 2  ;

cycle-specific analyses performed with' GEMINI. jl 1

o l

1663C El-4 10/21/87 SL-3197

ENCLOSURE 1 (Continued)

REQUEST TO REVISE TECHNICAL SPECIFICATIONS  !

FUEL THERMAL LIMITS AND REFUELING OPERATIONS BASIS FOR CHANGE RE00EST y 1

PROPOSED CHANGE 3:

i Georgia Power Company intends to load up to four Advanced Nuclear Fuels I Corporation (ANF) Lead Fuel Assemblies (LFAs) as part of the Hatch 2 f Reload 7 fuel batch. These assemblies, known as 9x9 LFAs, have been j neutronically designed to replace the B/P8DRB284H bundles which would l 1

otherwise be used. Proposed Change 3 will add a new axis to the existing B/P80RB284H HAPLHGR curve (Technical Specification Figure 3.2.1-9) to ,

reflect the fact that the 9x9 LFAs will have APLHGR limits that are i equivalent to the fuel they are designed to match on a planar power basis and will change the figure for MAPFACF to show that it applies to the. i 9x9 LFAs as well . Proposed Change 3 will also revise the caption of the l current 8x8 MCPR curve to indicate that the 9x9 LFAs will have the same l MCPR limits as the GE 8x8 fm1, ,

i l Backaround for Proposed Chance J: l l In order to evaluate the product offering of another fuel vendor, GPC plans to load into Hatch 2 up to four 9x9 fuel assemblies manufactured by Advanced Nuclear Fuels Corporation beginning in. Cycle 8.

Basis for Proposed Chance 3:

1 The 9x9 LFAs have been evaluated by Advanced Nuclear fuels Corporation for use in the Hatch 2 reactor, and the results of their analysis are contained in Enclosure 4 of this submittal. These fuel assemblies were neutronically designed to be similar to the GE B/P80RB284H fuel such that the existing MAPLHGR (when properly adjusted to account for the different number of rods) and MCPR thermal limits for the B/P8DRB284H fuel would be applicable to them as well. As stated in Enclosure 4, these objectives were conservatively met. In addition, the evaluations documented in Enclosure 4 demonstrate that the design bases analyses discussed in the FSAR are applicable to and conservatively bound the use of these LFAs.

1 The axis which will be added to the right side of the existing B/P80RB284H MAPLHGR curve (Figure 3.2.1-9) reflects the difference in the number of fueled rods between 8x8 and 9x9 fuel. Each point on the curve, therefore, represents the same planar power in a fuel assembly regardless of the number of rods. The numbers on Figure 3.2.1-9 without parentheses are the MAPLHGR limits in kw/ft of fueled rod for GE's B/P8DRB284H fuel, and those in parentheses are the corresponding MAPLHGR limits in kw/ft of fueled rod for ANF's 9x9 LFAs.

1663C El-5 10/21/87 SL-3197

i i

q i

l ENCLOSURE 1 (Continued) i 1

REQUEST TO REVISE TECHNICAL SPECIFICATIONS FUEL THERMAL LIMITS AND REFUELING OPERATIONS l BASIS FOR CHANGE REOUEST Geor ,ia Power Company intends to load the 9x9 LFAs in core locations f whic i are analyzed to have sufficient margin so that the LFAs are not j expected to be the limiting assemblies in the core on either a nodal or j bundle power bas" This loading strategy is designed to prevent the l hypothetical mis antation of the 9x9 LFAs from being the limiting i critical power event.

PROPOSED CHANGE 4:

Georgia Power Company proposes to make changes to the Refueling (Sections 3.10 and 3/4.9) and Design Features (Section 5.0) sections of the Hatch 1 (

and 2 Technical Specifications. These changes will allow a maximum }

loading of any four fuel assemblies around each Source Range Monitor (SRM) prior to obtaining a minimum SRM count rate of 3 cps before a full core reload. This is in contrast to the current stipulation that only fuel which has previously been in those locations can be used for this purpose. These changes will also ensure that all fuel designs from any fuel vendor will remain subcritical when any four bundles are loaded around a SRM by providing an analytical acceptance criterion on keft.

Background for Procosed Chanae 4:

During a full core reload, up to four assemblies are inserted next to each SRM in order to produce enough neutron flux to demonstrate SRM i operability so that approach to criticality can be monitored during fuel )

loading. Since criticality during fuel loading is to be avoided, the i fuel assemblies which have traditionally been loaded next to the SRMs are  ;

those which were previously in those locations because they are known to i form a subcritical configuration. This scheme has the disadvantage that since these assemblies are typically scheduled for either discharge or placement in another core location, they must be moved and replaced with l l the next cycle's assemblies. j

\

Georgia Power Company proposes to amend the Hatch 1 and 2 Technical l Specifications because overall plant efficiency can be increased during refueling if these additional fuel handling operations are eliminated.

Likewise, plant safety will be increased because the possibility of a fuel handling accident or fuel loading error will be reduced.

1663C El-6 10/21/87 SL-3197

x ..

m.

l 3 l' t ig i.  !

, s' 1 < ,it 7

, S q

1 m a ENCLOSUREfli(Continued)

REQUEST TO REVISE < TECHNICAL: SPECIFICATIONS

~

FUEL THERMAL LIMITS AND REFUELING OPERATIONS ,

l BASIS FOR CHANGE REOUEST y

l:

Basis for Proposed Chanae'4:' k As - part of:.the original' licensing Lbasis .for. Hatch 2, General Ele'ctvic L

l evaluated several ~ possibleofuel configurations in Ethe Spent Fuel:.. Pool (SFP). as reported- in.-the Hatch-2 FSAR analysis. ..-One Lof1 these '

configurations was a four bundle squa' r e array at,least twelvelinchesifroma '

a SFP rack. Although very conservative assumptions awere made, their analysis showed that a. 2x2 fuel array Will' remainn. subcritical Eby La; considerable ma gin even for the highest ; reactivity fuel (i . e . , ken '

g 1.35)' manufactured'by GE'(Reference 5).

qi This analysis is also applicable to the Hatch reactors sinceJ the 12-inch s6paration from the racks means' that neutron absorption due..to the' racks and the neutron multiplication due'to fuel in-the racks'is. minimal. -

Also, since each 2x2 array around an SRM is more' than 12" from each of 1 the other three SRM arrays the referenced FSAR analysis ' remainsL valid.

Therefore, it is concluded.'that . placing any four fuel assemblies. with' K.

• 1.35 around each of the .four SRMs' cannot produce a ' critical!

i configuration. . In addition, i t will be required' that . this 2x2 -array- -

analysis be shown to be applicable to all fuel designs from any. fuel l, vendor by adding 'an additional requirement to the Fuel Storage criteria '

l! of the Design Features section of the Technical Specifications. < -!

h d

i I

l

.1663C. El 10/21/871

' SL-3197 R

1-

l

)

l ENCLOSURE 1 (Continued) l l

REQUEST TO REVISE TECHNICAL SPECIFICATIONS l FUEL THERMAL LIMITS AND REFUELING OPERATIONS  :

BASIS FOR CHANGE REOUEST l I

REFERENCES

1. Letter, J. S. Charnley (GE) to H. N. Berkow (NRC), " Revised j Supplementary Information Regarding Amendment 11 to GE Licensing j Topical Report NEDE-240ll-P-A," (GESTAR II), dated January .16,1986.

l

2. Letter, C. O. Thomas (NRC) to J. S. Charnley (GE), " Acceptance for-Referencing of Licensing Topical Report, NEDE-240ll-P-A, Rev. 6, I Amendment 11, ' General Electric Standard Application for Reactor Fuel' (GESTAR II)," dated November 5, 1985.

3. Letter, G. C. Lainas (NRC) to J. S. Charnley (GE), " Acceptance for Referencing of Licensing Topical Report NEDE-240ll-P-A, ' General Electric Licensing Reload Report,' Supplement to Amendment 11," dated I

March 22, 1986.

1

4. Letter, J. S. Charnley (GE) to G. C. Lainas (NRC), " GEMINI /0DYN Statistical Adders for BHR 4/5 (with RPT) - E0C," dated July 28, 1986.

5. Edwin I. Hatch Nuclear Plant Unit 2 Final Safety Analysis Report (Section 9.1).

l 1663C El-8 10/21/87 SL-3197

=.

l 1

4 ENCLOSURE 2 PLANT HATCH - UNITS 1, 2 NRC 00CKETS 50-321, 50-366 OPERATING LICENSES DPR-57, NPF-5 REQUEST TO REVISE TECHNICAL SPECIFICATIONS FUEL THERMAL LIMITS AND REFUELING OPERATIONS' ,

10 CFR 50.92 EVALUATION t PROPOSED CHANGE 1:

This change would revise the formula for determining the Option B OLMCPR statistical scram speed limit, TB, and move it to the Bases section ,

in order to give Georgia Power Company the maximum flexibility to use BHR '

• ndustry experience with respect to scram time testing. For consistency, the definitions of x, tave, and TA would also be moved, since the value of the Operating Limit Minimum Critical Power Ratio (0LMCPR) depends upon all four variables.

Basis for No Significant Hazards Consideration Determination: ,

The proposed change does not involve a significant hazards consideration I because it would not: .

l

1. Involve a significant inct e?.se in the probability or consequences of i an accident previously evaluated because no change in plant operation I will occur as a result of this change. The definitions of x, i TA. TB. Tave, and OLMCPR Option A and Option B will l remain the same and will be monitored at the site in the same manner.

as before.- The value for TB will change as a result of a ,

reevaluation of the BWR scram time data base. The new scram time I distribution that was used to determine 23 for the Option B MCPR limit was reviewed and approved by the NRC 'In their consideration of the GEMINI application methodology.

2. Create the possibility of a new or different kind of accident from any previously analyzed because no change in plant equipment or operations will occur r a result of this change.

3. Involve a significant reduction. in the margin of safety because the OLHCPR will continue to be based upon either actual measured scram speeds or a conservative assumption relative to scram speeds. Both of these methods have been previously approved by the NRC.

1663C E2-1 10/21/87 SL-3197

l l

l

! 1 l ENCLOSURE 2 (Continued) '

l REQUEST TO REVISE TECHNICAL SPECIFICATIONS FUEL THERMAL LIMITS AND REFUELING OPERATIONS ,

10 CFR 50.92 EVALUATION

]

J PROPOSED CHANGE 2-This proposed change would reduce the Option A MCPR limit for Unit 2 from l 1.37 to 1.33 for all Hatch 2 8x8 fuel. This is being done to maximize  !

the MCPR margin for all scram speeds.

Basis for No Significant Hazards Consideration Determination: -l This proposed change does not involve a significant hazards' consideration l because it would not: 1

1. Involve a significant increase in the probability or consequences of ]

an accident previously evaluated because no change in equipment l operations will occur as a result of this change. The new Option A {

MCPR value derived for use with GEMINI methodology will still ensure l fuel design limits will be met because the initial operating value assumed in the LOCA analyses will be conservative for all operating '

conditions, and that with 50% confidence, 99.9% of the fuel rods will I avoid boiling transition during a core wide transient. j j

2. Create the possibility of a new or different kind of accident from i any previously analyzed because no change in plant equipment or )

operations will occur as a result of this change.  !

3. Involve a significant reduction in the margin of safety because the -

method used to determine the Option A limit is consistent with the j application of GEMINI. This method has been reviewed and approved by I

the NRC for use by BWR utilities.  ;

l l

i i

I l

1663C E2-2 -10/21/87 SL-3197 l i

_ _ _ _ _ _ _ . _ _ _ _ _ _ E_

. l ENCLOSURE 2 (Continued) l REQUEST TO REVISE TECHNICAL SPECIFICATIONS l FUEL THERMAL LIMITS AND REFUELING OPERATIONS 10 CFR 50.92 EVALUATION j l

150 POSED CHANGE 3: l Georgia Power Company proposes to modify an existing APLHGR limit curve in the Unit 2 Technical Specifications to reflect the thermal-mechanical and Emergency Core Cooling System (ECCS) limits on four 9x9 Lead Fuel Assemblies (LFAs) which are expected to be part of the Hatch 2 Reload 7 l fuel batch. The figure for MAPFACF will be revised to show that it I applies to the 9x9 LFAs as well as the GE 8x8R fuel. The caption for the existing OLMCPR curve for 8x8 fuel will be changed to denote that the same limits apply to the 9x9 LFAs as well.

Basis for No Significant Hazards Consideration Determination:

This proposed change does not involve a significant hazards consideration because it would not:

1. Involve a significant increase in the probability or consequence of an accident previously evaluated because:

a. This does not involve any change in ECCS equipment response since there is no change to ECCS equipment, configuration or setpoints.

l

b. Analyses performed by Advanced Nuclear Fuels Corporation (ANF) have determined that, in comparison to GE 8x8R fuel, the 9x9 fuel has an equivalent or improved ECCS-LOCA response. This is due to the lower stored energy in the fuel . rods, better heat transfer characteristics, and less restrictive countercurrent flow as a result of a more open upper tie plate (Reference 1).

c. The 9x9 assemblies have been designed to withstand the same mechanical forces as the current fuel, and analyses have shown that they meet the operating and design safety limits (Reference 2).

(1. ANF has determined that there will be less radioactivity released from the LFAs than GE's 8x8R fuel in the event of an accident in which fission gas is released (Reference 1).

1663C E2-3 10/21/87 SL-3197

l- s

-Q ,1 7 ;

/

~

ENCLOSURE 2.(Continued) l a

REQUEST TO REVISE TECHNICAL SPECIFICATIONS ,

]

FUEL THERMAL LIMITS AND REFUELING OPERATIONS l

10 CFR 50.92 EVALUATION j t

I

e. The weight and mass distribution of the LFAs is'very similar to l the GE assemblies. (The 9x9 bundles weigh sl_ightly less than I their GE counterparts.)

{

f. The ANF bundles will be fully compatilfre'wlth all fuel handling equipment including the fuel grapple. u,

.t

g. The channel and number of bundle spacers will be exactly the same as the GE fuel.

h. This does not involve any change in the control rods or the l control rod drive system. j

i. ANF has determined that the enthalpy deposited in the 9x9 LFAs l will not exceed 280 cal /gm in the unlikely event of a Control Rod 1 Drop accident (Reference 1).

2. Create the possibility of a new or different kind of accident from any previously analyzed because no' change in plant design or.,

operation is involved except for relatively minor changes in the '

l mechanical, thermal-hydraulic, and nuclear aspects of the fuel design for a small quantity of assemblies. ]

3. Involve a significant reduction in the margin of safety because the MCPR safety limit for the 9x9 LFAs is the same as the 8x8 assemblies and the operating limits for existing fuel are conservative for l j

application to the LFAs. ANF has shown that their fuel complies with  !

all Specified Acceptable Fuel Design Limits, and the performances' delineated in 10 CFR 50.46 are complied with by conservative application of the B/P8DRB284H APLHGR limits to the LFAs (Reference 2). j

)

+

J l

y l 1

1 1663C .E2-4 10/21/87-SL-3197

'4

)

D

+

i a

ENCLOSURE 2 (Continued)

-l REQUEST TO REVI.SE. TECHNICAL SPECIFICATIONS FUEL THERMAL LIMITS AND REFUELING OPERATIONS 10 CFR 50.92 EVALUATU)E  ;

1 <

PRAPhSJJ)_ CHANGE' 4:

, c ,

r c Georg1; Power Company proposes to make changes to the Refueling and

' Design)7/atures ' sections of the Hatch 1 and 2 Technical Specifications. '

'fThese changes will allow a maximum loading of any four fuel assemblies

'around' each SRM prior to a full core reload. These changes will also en:ure that all fuel designs from any fuel vendor will remain subcritical j when any four bundles are loaded around a SRM by providing an analytical <

acceptance criterion on keff.

Basis for Nplianificant Hazards C60sYderation Determination 2:

The proposed changes do not invohe a significant hazards consideration because they would not: 1

1. Invofve a significant increase in the probability or consequence of an accident previously evaluated because there will be a reduction in i fuel ' handling operations and the number of core configurations. This reduction in operations and configurations will reduce the probability of occurrence of a fyel handling accident and a fuel i

loadind' error. It was reported in the Hatch 2 FSAR that any credibly postulated 2x2 fuel assembly array cannot become critical even under the most' lirrAting condiyons. '^

2. Create the possibility o,1r a new or different kind of accident from any previcusly analyzed because no significant change in plant equipment or operation #id11 occur as a result of this change.

3. Involve a significant reduction in margin of safety because the analysis documented in the Hatch 2 FSAR shows that there is considerable margin to criticality for any limiting 2x2 fuel array.

This conclusion haWheen reviewed and approved by the NRC as part of the initial licilhsing submittal for Hatch 2.

r ..

i

's l

.(

t ,

(f ,

(b i 63C # ,/ E2-5 /' 10/21/87 jil-197 <

o 4

s y.iq z

a l s . a

, e n i ny

' p-sio[ $qf J b., L, sn >  ; <'

-" o'.

l s'- b y A.

y( ( . (, '47 y 3 ;,-

'7

) 4 t i

[(, d, g e e \ t i 9 4 .'#\ l 3 g ENCLOSURE 2 (Soni tudd)

N'.

REQUEST TO REVISE TECK'. TIC 81, SPECIFICATIONS ,i' .

3 FUEL THERIAlf6IMITS AND REFUELING OPERATIONS J i

lQ_Cf8 50.92 EVALUATION , .f

<\ 1, , t. ,

.c , < 4 q 1 p

5 REFEREKES ,3 1.I ANF-87-95, Rer. 1, " H'3 '

Report," July 1987. atch 9x9 Lead Fuel Assemblids Safety) Ahlysis s . i, i

-u

3. XN-NF-85-bNO)(A), Rn. (T,' \ " Generic Mechanical Design for Exdn '

Nuclear Jet Pump BHR Rdoad Fuel, Sentember 1986.

<3,

~

r

\

, 3 y

s.

4 / .

< = \ f 1 l

\ ,

\ .

( 1

• ]

-l

\ v

\ -

\ 1

, 11 a

i

\

1 I

1 1

l 1

\

1 3

s <

1603C ,

E2-6 10/21/87

/ - 'j

.p SL-3197

, }

~'h,.,',

- dby a, .

t- ENCLOSURE 3 l A.

7T * '

PLANT HATCH - UNITS 1 , 2 NRC DOCKETS 50-321, 50-366 OPERATING LICENSES DPR-57, NPF-5 t REQUEST TO REVISE TECHNICAL SPECIFICATIONS  ;

FUEL THERMAL LIMITS AND REFUELING OPERATIONS PAGE CHANGE INSTRUCTIONS i

'l UNIT 1 f

i Removed Page Insert Paae 3.10-2 3.10-2 ~ .

3.10-7 3.10-7 l 5.0-1 5.0-1 l UNIT 2 Removed Page Insert Pace'  ;

3/4 2-4g 3/4 2-4g 4 l 3/4 2-4j 3/4 2-4j 3/4 2-6 3/4 2-6 l 3/4 2-7 3/4 2-7 3/4 2-7a 3/4 2-7a 3/4 9-4 3/4 9-4 B 3/4 2-2 B 3/4 2-2 8 3/4 2-4 B 3/4 2-4 B 3/4 9-1 B 3/4 9-1 5-3 5-3 1663C E3-1 10/21/87 SL-3197