ML20235U627

Application for Amends to Licenses DPR-57 & NPF-5,revising Fuel Average Planar Linear Heat Generation Rate Limits & ECCS Surveillance Requirements.Proprietary Tech Spec Pages & GE Topical Rept NEDC-31376P Encl.Fee Paid

ML20235U627
Person / Time
Site:	Hatch
Issue date:	10/08/1987
From:	James O'Reilly GEORGIA POWER CO.
To:	NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
Shared Package
ML19304B572	List: ML20235U627
References
SL-3232, TAC-66472, TAC-66473, NUDOCS 8710140171
Download: ML20235U627 (16)

Text

,_

e IGeot0!a Fowhr Cornpany ;

[/ ' 333 Pmdmont Avenue

[  ; Atlanta, Georgia 30308 ,

p Toephone 404 526//8511 x " ' Mailing Address:

Fbot Office Box 4545 -

. Atlanta, Georgia 30302.'

i~

. James P. O'Reilly .  : the southern eixtnc system

. . . Senior Vice President

'F-Nucleat Orierations SL-3232 1679C X7GJ17-H600 October 8, 1987 E .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.DPR-57, NPF-5

. REQUEST TO REVISE TECHNICAL SPECIFICATIONS:

FUEL THERMAL LIMITS AND ECCS SURVEILLANCE RE0VIREMENTS i

. Gentlemen.

l In accordance with the provisions of 10 CFR 50.90, as required 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 to Operating Licenses DPR-57 and NPF-5.

. The- proposed changes involve revisions to the Unit 1 and Unit 2 i Technical' Specifications defining fuel Average Planar Linear Heat Generation - Rate - ( APLHGR) limits and Emergency Core Cooling System (ECCS)

Surveillance Requirements.- Specifically, the proposed Technical Specifications changes would:

1. Revise APLHGR limits for General Electric BP8x8R and P8x8R fuel types (Units 1 and 2).

2. Add an APLHGR limit for fuel types BP8DRB301L and P8DRB301L l (Units 1 and 2). ,

1

3. Revise the minimum flowrate Surveillance Requirement for the i Core Spray System (Units 1 and 2) and the maximum Response Time Surveillance Requirements for the' Core Spray System and the Residual Heat Removal System (Low Pressure Coolant Injection mode) (Unit 2 only).

4. Revise the Bases to reflect the Plant Hatch SAFER /GESTR-LOCA analysis and delete APLHGR limits for the fuel that will not be  !

used (Units 1 and 2).

B710140171 871000 v '

PDR ADOCK O 31 (g40 h%

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U.S. Nuclear Regulatory Commission October 8, 1987 Page.Two.

1 The proposed changes, if approved, would permit greater core design and operating ~ flexibility and would result in nuclear fuel cost savings.

The changes would also remove unnecessary conservatism from the ECCS Surveillance Requirements and ensure that the Technical Specifications are consistent with the new LOCA analysis.

Enclosure.1 provides a detailed -description of the proposed changes and circumstances necessitating the change request.

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

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

The proposed changed Technical Specifications pages follow Enclosure 3.

To support the proposed changes, Enclosure 4 provides documents that have not been submitted to the Nuclear Regulatory Commission (NRC).

Please note that some information in Enclosure 4 has been marked

" Proprietary" to protect the commercial interest of the fuel vendor. In accordance with 10 CFR 2.790, an affidavit is included herein which requests that certain information be withheld from the public domain.

Payment of a 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.

Pursuant to the requirements of 10 CFR 50.91, a copy of this 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.

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U S. Nuclear. Regulatory Commission

. October.8, 1987 Page Three Mr. James - P. O'Reilly states that. he is ' Senior. Vice President 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 By: O>vreA .@ b James P. O'Reilly o

. Sworn to and subscribed befor e this th day of October 1 .

~ '/:

Nk Notary Public k$Y nevry p sc. cia,en county. Gern GKN/lc Uycmune n E

• e5U" 12 *

Enclosures:

1. Basis for Change Request

2. 10 CFR 50.92 Evaluation

3. Page Change Instructions-

.4. "Edwin I. Hatch Nuclear Plant Units 1 and 2 SAFER /GESTR-LOCA

-Loss of Coolant Accident Analysis," NEDC-31376-P.

5. Filing Fee - $150.00 c: Georaia Power Comoany Mr. J. T. Beckham, Jr. w/o enclosures GO-NORMS w/o enclosures U.S. Nuclear Regulatory Commission. Washington. D.C.

Mr. L. P. Crocker, Licensing Project Manager - Hatch w/ enclosures U.S. Nuclear Regu1Atory Commission. Reaion II Dr. J. N. Grace, Regional Administrator w/ enclosures Mr..P. Holmes-Ray, Senior Resident Inspector - Hatch w/ enclosures Stitte of Georaj.A Mr. J. L. Ledbetter ,

l 1679C

i-i ENCLOSURE 1-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 ECCS SURVEILLANCE RE0UIREMENTS-BASIS-FOR CHANGE RE0 VEST PROPOSED CHANGE 1:

Eackground: j This change revises the Technical Specifications Average Planar Linear Heat Generation Rate (APLHGR) limits to reflect results from the "Edwin I. Hatch ! Nuclear Plant Units .1 and 2 SAFER /GESTR-LOCA Analysis Report,"- -

NEDC-31376-P (Enclosure 4). The- APLHGR limits presently in the- Technical

- Speci fications were calculated using General Electric SAFE /REFLOOD methodology. The . proposed new APLHGR limits for P 8R and BP8x8R fuel

. covered. .in ' the ' Technical Specifications were calc! 2ted .using approved -

. SAFER /GESTR-LOCA and GEMINI physics methods.

Basis for Procosed Change 1: I The General Electric SAFER /GESTR-LOCA methods are described in References 1, 2, and 3. The NRC approval of these methods is documented in  ;

References 4, 5, .and 6. Plant-specific LOCA/ECCS analyses with l SAFER /GESTR-LOCA have been reviewed and approved by the NRC for Duane Arnold Energy Center.-(Reference 7) and for the James A. Fitzpatrick ,

Nuclear Power Plant (Reference 8).

The Plant Hatch SAFER /GESTR-LOCA analysis (Enclosure 4) demonstrates that the generic peak clad temperature (PCT) versus break-size curves (Reference 3) for both nominal and Appendix K model assumptions are i applicable to both Plant Hatch units. The PCT points calculated for

. Plant Hatch Units 1.and 2 using realistic models demonstrate that the shapes of the PCT versus break-size curves match the shape of the nominal generic curve for the BHR 3/4 plant class. The SAFER /GESTR-LOCA licensing: basis PCT calculations for Plant Hatch Units 1 and 2 with Appendix _ K model assumptions also show that both units' PCT versus break-size curves match the gener.c Appendix K PCT versus break-size curves for the BWR 3/4 class of plants.

The Plant Hatch analysis was performed with many of the Emergency Core Cooling System .(ECCS) parameters conservatively established relative to

. the actual performance of the ECCSs at Plant Hatch. This was done in support of the Technical Specifications program for Plant Hatch so that 1679C El-1 10/8/87 SL-3232

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ENCLOSURE 1 (Continued)

RE0 VEST TO REVISE TECHNICAL SPECIFICATIONS:

FUEL THERMAL LIMITS AND ECCS SURVEILLANCE REQUIREMENTS l BASIS FOR CHANGE RE0UEST future improvements to the plant could be accommodated without requiring extensive reanalysis. The maximum PCT from the Plant Hatch SAFER /GESTR-LOCA analysis for nominal model assumptions was nearly identical to that of the generic BWR 3/4 nominal analysis at 14.4 kW/ft presented in Reference 3. Furthermore, Plant Hatch has similar '

sensitivities, as does the generic plant, to the parameters used to calculate the generic upper-bound PCT. Consequently, the results from the generic BHR 3/4 calculations of the statistical upper bound are applicable to the Plant Hatch units with the assumed ECCS parameters. l The Plant Hatch ECCS configuration for the limiting-break and single- h failure combination is identical to the generic BWR 3/4 ECCS configuration.

Finally, the licensing basis PCTs for each Hatch unit exceeded the statistical upper bound PCT, but remained below the 22000F 10 CFR 50.46 limit, as required by Reference 6. All 10 CFR 50.46 requirements for LOCA ECCS evaluations (maximum cladding oxidation, maximum hydrogen generation, PCT, coolable geometry, and long-term cooling) are met by the Plant Hatch fuel types addressed in the SAFER /GESTR-LOCA analysis.  ;

)

The Plant Hatch SAFER /GESTR-LOCA analysis demonstrates that APLHGR limits proposed for these Technical Specifications are derived on the basis of the maximum allowable fuel pellet power level (MLHGR) used as the fuel design basis. Lattice local power and exposure peaking factors ,

calculated using approved GEMINI methods (Reference 9) are applied to  !

transform MLHGR into APLHGR limits for each bundle type. '

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1679C El-2 10/8/87 SL-3232 1

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ENCLOSURE 1 (Continued)

REOUEST TO REVISE TECHNICAL SPECIFICATIONS:

FUEL THERMAL LIMITS AND ECCS SURVEILLANCE REQUIREMENT _S BASIS FOR CHANGE RE0 VEST PROPOSED CHANGE 2:

Backaround:

This change would add an APLHGR limit curve to the Technical Specifications for both Plant Hatch units to allow for the use of General  ;

Electric fuel types BP8DRB301L and P80RB30ll.  !

Basis for ProDosed Change 2:

Average Planar Linear Heat Generation Rate limit data, as a function of Average Planar Exposure for BP8DRB301L fuel, is given in Reference 10.

The Plant Hatch SAFER /GESTR-LOCA analysis, reported in NEDC-31376-P, demonstrated that BP8x8R and P8x8R fuel types are not limited by ECCS considerations. Therefore, the Reference 10 data result solely from fuel thermal mechanical design limits divided by appropriate bundle local i peaking factors to yield average planar power limits. In this case, the i local peaking factors were derived with General Electric's NRC-approved l GEMINI physics methods (Reference 9).

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I 1679C El-3 10/8/87 SL-3232

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h ENCLOSURE 1-(Continued)

REOUEST TO REVISE TECHNICAL ~ SPECIFICATIONS:

LFUEL THERMAL LIMITS AND ECCS SURVEILLANCE RE0VIREMENTS I BASIS FOR CHANGE RE00EST PROPOSED CHANGE 3:

Backaround:

This. change revises .certain Technical Specifications Surveillance

. Requirements for the Core Spray (CS) System and the Low Pressure Coolant  !

Injection -(LPCI) mode of the Residual Heat Removal (RHR) System. The

~ changes are: (1) a decrease in the required CS System flowrate i

. Surveillance Requirement from 4625 gpm to 4250 gpm (Units 1 and 2), )

(2) an increase in .the ECCS maximum Response -Time for the CS System from 27 sec to 34 sec- (Unit 2), and (3) an increase in the ECCS maximum Response Time for the LPCI mode of RHR from 40 sec to 64 set (Unit 2).

Basis for Proposed Change 3:

Analyses of the Design Basis Loss-of-Coolant Accident (LOCA) for Plant.

Hatch Units 1 and 2, performed with the SAFER /GESTR-LOCA methods (Enclosure 4), assumed that the performance of the ECCS equipment was as given in. the proposed Surveillance Requirements. No plant safety

. functions, except LOCA mitigation, are being performed by the equipment with proposed revised Surveillance Requirements. Therefore, for consistency with the analysis requirements, the' Surveillance Requirements are' being revised. Neither the CS System nor the LPCI mode of RHR would be modified as a result of the approval of this proposed change.

Emergency Core Cooling System Response Time, defined in the Plant Hatch Unit 2 Technical Specifications Definition section, is that time interval from when the monitored parameter exceeds its ECCS actuation setpoint at the channel sensor until the ECCS equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). The

' Response Time includes diesel generator starting and sequence loading delays for the RHR and CS pumps. The existing requirements for diesel generator starting and sequence loading for RHR are not being revised at this time. The proposed required Response Times reflect the availability of LPCI and/or CS flow assumed in the SAFER /GESTR-LOCA analysis.

It should be noted that the proposed relaxations in pump flow rates and ECCS response times are consistent with the LOCA analysis assumptions.

1679C El-4 10/8/87 SL-3232

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ENCLOSURE I (Continued) _

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RE00EST-TO REVISE TECHNICAL SPECIFICATIONS:

FUEL THERMAL LIMITS AND ECCS SURVEILLANCE RE0VIREMENTS -;

BASIS FOR CHANGE RE00EST

t This will . allow =GPC greater. flexibility in operation ~ and in the purchasing of . . qualified replacement components. Requirements for

Inservice Inspection and Testing. (including ASME'Section XI and Appendix J) will still track the performance history of . mechanical components (eg., pumps and valves)... Sur_veillance and testing. . requirsnents for-Instrumentation and Control equipment will continue as before.

PROPOSED CHANGE 4:

flar,haround:

This change would revise -the Technical Specifications Bcses sectiolu for ',

both Plant Hatch units to reflect the changes. in the Design Basis' LOCA ^.

resulting. from the Plant. Hatch SAFER /GESTR-LOCA analysis. This change +

would. also delete ~ from the Technical Specifications for both units . the-existing APLHGR limits. for all 7x7 fuel, for other non-prepressurized- 5 fuel, and for fuel types BP8DRB284LA and P80RB284LA. >

• i Basis for Change 4: ,

The proposed changes are administrative in nature. Existing information H 1

in.the Bases which is no longer applicable is being deleted or modified; _.

analogous information which is applicable to the new LOCA analysis is being added to the Bases. The APLHGR limits for the.7x7 and other fuel types are being deleted, because they willj 'not be used for anticipated reload core designs.

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ENCLOSURE'1 (Continued)

EE00EST TO REVISE TECHNICAL SPECIFICATIONS:

FUEL THERMAL LIMITS AND ECCS SURVEILLANCE REQUIREMENTS BASIS FOR CHANGE REOUEST REFERENCES

1. . . "The 'GESTR-LOCA and SAFER Models for the Evaluation of the Loss of-Coolant Accident," NEDE-23785-1-P (Proprietary); Volume 1, "GESTR/LOCA -A Model for the Prediction of Fuel Rod Thermal Performance," December 1981.

2. "The GESTR-LOCA and SAFER Models for the Evaluation of the Loss of Coolant Accident," EDE-23785-1-P (Proprietary); Volume 2, " SAFER -

Long: Term Inventory Model for BHR Loss-of-Coolant Analysis," December 1981.

3. "The GESTR-LOCA and SAFER Models for the Evaluation of the Loss of Coolant Accident", NEDE-23785-1-P (Proprietary); Volume 3

" SAFER /GESTR. Application Methodology," March 23, 1984.

4. U.S. Nuclear Regulatory Commission, " Safety Evaluation of the General Electric Company Topical Report, NEDE-23785-1, Volume 1 - GESTR/LOCA, A Model for the Prediction of Fuel Rod Thermal Performance,"

September 1983, report approved November 2, 1983.

5. U.S. Nuclear Regulatory Commission, " Safety Evaluation Report on the MFER Code," July _ 12, 1983.

6. Letter, C. O. Thomas (NRC) to J. F. Quirk (GE), " Acceptance of Referencing of Licensing Topical Report NEDE-23785 Revision 1. Volume III, 'The GESTR LOCA and SAFER Models for the Evaluation of the Loss of Coolant Accioent,'" June 1, 1984.

, 7. Letter, A. J. Cappucci (NRC) to L. Liu (IELP), " License Amendment No.

af 142 - Cycle 9 Reload (TACB568)," May 7, 1987.

8.' Letter H. I. Abelson (NRC) to J. C. Brons (NYPA), " Approval of James A. Fitzpatrick Nuclear Plant (Docket 50-333) Technical Sper.ification Amendment Number 109," April 3,1987.

9. Letter, C. O. Thomas (NRC) to J. S. Charnley (GE), " Acceptance for Referencing of Licensing Topical Report NEDE-30130, ' Steady State Nuclear Methods,'" December 22, 1985.

10. "Edwin I. Hatch Nuclear Plant Units 1 and 2 SAFER /GESTR-LOCA Loss of Coolant Accident Analysis," NEDC-31376-P, as amended, December 1986.

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1679C El-6 10/8/87 SL-3232 1

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ENCLOSURE'2-

. PLANT HATCH UNITS 1, 2 NRC~ DOCKETS 50-321, 50-366

'0PERATING LICENSES DPR NPF-5 .

, . REOUEST TO REVISE TECHNICAL SPECIFICATIONS:

. FUEL THERMAL LIMITS AND ECCS SURVEILLANCE RE0UIREMENTS 10 CFR 50.92 EVALUATION PROPOSEDC/dE1:-

This changef revises .the Technical Specifications Average Planar Linear

Heat Generation Rate (APLHGR) limits to reflect results from' the "Edwin I. - Hatch Nuclear Plant Units 1. and 2 SAFER /GESTR-LOCA Loss of: Coolant Accident Analysis Report, ' NEDC-31376-P. The APLHGR limits presently in the Technical Specifications were calculated using General Electric t

SAFE /REFLOOD . methodology. The proposed new APLHGR limits for' P8x8R and BP8x8R fuel covered in the Technical . Specifications were calculated using-approved SAFER /GESTR-LOCA and GEMINI physics method:,. The revised limits would parmit, greater core design and operating flexibility and would result in nuclear fuel' cost savings.

j Basis far Significant Hazards Determination:

The proposed change does- not involve a significant hazards consideration.

for the following reasons:

1

1. It does not. involve a significant increase in the probability or consequences of- an accident previously evaluated, because no. change in plant design or. procedures will occur as a result of this change.

Any future core loading patterns or core operational conditions facilitated by the proposed change will be evaluated, using approved methods, and shown to meet the approved acceptance criteria for analysis of the limiting accidents that were previously evaluated.

Therefore, the consequences of an accident would not be significantly i increased.

The~ consequences of a hypothetical design basis LOCA have been  !

evaluated, consistent with the proposed fuel limits, and shown to be ,

,less severe than those predicted to occur using existing limits in the earlier SAFE-REFLOOD analysis. The revised analysis is reported in General Electric report NEDC-31376-P, "Edwin I. Hatch Nuclear Plant Units 1 and 2 SAFER /GESTR-LOCA Loss of Coolant Accident Analysis." .

f .2. The proposed change would not create the possibility. of a new or different kind of accident from any previously analyzed, because no

[

o change in plant equipment or procedures will occur as a result of

[ this change.

1679C E2-1 10/8/87

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M ENCLOSURE ~2 (Continued).

REOUEST TO REVISE TECHNICAL SPECIFICATIONS:

FUEL THERMAL LIMITS AND ECCS SURVEILLANCE REQUIREMENTS

_10 CFR 50.92 EVALUATION

3. The proposed change' would not involve a'significant reduction- in- the!-

margin of safety, because thel proposed ~ revised -limits : have .been determined, using approved analysis methods,- and. have been shown to n.< - maintain all existing fuel safety ~ margins associated with- Energency Core Cooling; System performance..

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ENCLOSURE 2 (Continued)  ;

REOUEST TO REVISE TECHNICAL SPECIFICATIONS:

'EUEL THERMAL LIMITS-AND ECCS SURVEILLANCE REQUIREMENTS.

10 CFR 50.92 EVALUATION PROPOSED CHANGE 2i

.This change would add' an APLHGR limit curve to the Technical Specifications for .both Plant Hatch units to allow for the_ use of General Electric fuel types BP8DRB301L and P8DRB301L.

Basis for No Significant Hazards Determination:

L The proposed chhnge does not involve a significant hazards consideration

for the following reasons

1. - It does not involve. a significant increase in the probability of an accident previously evaluated, because no change in plant design or procedures will occur as a result of this proposal. The proposed new fuel' type is' generically licensed for use in either Hatch unit under the " General Electric Standard Application for Reactor Fuel,"

NEDE-24011-P-A. Use of the proposed limits would- not significantly increase the consequences of an accident previously evaluated for the same reasons stated for Proposed Change 1 above.

~2. .It ' does not create. the possibility of a new or different kind of accident -from any previously analyzed, because no significant change in plant equipment or procedures will. occur as a result of this change.

3. It does not involve a significant reduction in the margin of safety for'the same reasons stated for Proposed Change 1 above.

1679C E2-3 10/8/87 SL-3232

ENCLOSURE 2 (Continued)

REQUEST TO REVISF_]ICHNICAL SPECIFICATIONS:

FUEL THERMAL LIMITS AND ECCS SURVEILLANCE RE0VIREMENTS 10 CFR 50.92 EVALUATION PROPOSED CHANGE 3:

This change revises certain Technical Specifications Surveillance Requirements for the Core Spray (CS) System and the Low Pressure Coolant Injection (LPCI) mode of the Residual Heat Removal (RHR) System. The changes are: (1) a decrease in the required CS System flowrate Surveillance Requirement from 4625 gpm to 4250 gpm (Units 1 and 2),

(2) an increase in the Emergency Core Cooling System (ECCS) maximum Response Time for the CS System from 27 sec to 34 sec (Unit 2), and (3) an increase in the ECCS maximum Response Time for the LPCI mode of RHR from 40 sec to 64 sec (Unit 2).

Basis for No Significant Hazards Determination:

The proposed change does not involve a significant hazards consideration for the following reasons:

1. It does not involve a significant increase in the probability of an accident previously evaluated. The proposed revised Surveillance Requirements are unrelated to initiating events for the FSAR Chapter 15 accidents. Thus, the probability of occurrence of those  ;

initiating events would not be affected by the proposed change. The ECCS equipment involved would function to mitigate a hypothetical Design Basis LOCA; other functions of the equipment require less performance capability. The performance of this equipment during a LOCA has been evaluated, consistent with the proposed change. It was !

shown that in such an event, all applicable acceptance criteria would be met. These LOCA acceptance criteria were established to ensure ,

that the consequences of a Design Basis LOCA would not significantly af fect public health and safety. Therefore, the consequences of an j accident would not be significantly increased.

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2. It would not 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.

3. It would not involve a significant reduction in the margin of safety for the same reasons stated under Proposed Change 1 above.

1679C E2-4 10/8/87 SL-3232

ENCLOSURE 2 (Continued)

REOUEST TO REVISE TECHNICAL SPECIFICATI0XS: )

ffEL THERMAL LIMITS AND ECCS SURVEILLANCE REQUIREMENTS 10 CFR 50.92 EVALUATION PROPOSED CHANGE 4 This change would revise the Technical Specifications Bases sections for both Plant Hatch units to reflect the changes in the Design Basis LOCA resulting from the Plant Hatch SAFER /GESTR-LOCA analysis. Existing information which is no longer applicable is being deleted or modified, as necessary. Analogous information applicable to the new LOCA analysis is being added to the Bases. This change would also delete from the Technical Specifications for both units the existing APLHGR limits for all 7x7 fuel, other non-prepressurized fuel, and for one prepressurized fuel type.

Basis for No Significant Hazards Determination:

The proposed change does not involve a significant hazards consideration for the following reasons:

1. It does not involve a significant increase in the probability or consequences of an accident, because it is purely an administrative change. The change is being proposed to: (1) achieve consistency of the Technical Specifications Bases requirements that would be revised by Proposed Changes 1, 2, and 3 and (2) delete from the Technical Specifications unnecessary information for fuel not expected to be used in future reload cores.

2. It would not 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.

3. It would not involve a significant reduction in the margin of safety because the change is purely administrative.

1679C E2-5 10/8/87 SL-3232

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ENCLOSURE 3 PLANT HATCH - UNITS 1, 2 NRC DOCKETS 50-321, 50-366 OPERATING LICENSES DPR NPF-5 RE0 VEST TO REVISE TECHNICAL SPECIFICATIONS:

FUEL THERMAL LIMITS AND ECCS SURVEILLANCE REQUIREMENTS EAGE CHANGE INSTRUCTIONS The proposed changes to Technical Specifications-(Appendix A to Operating Licenses DPR-57 and NPF-5) would be incorporated as follows:

Removed Page UNIT 1 Insert Pagg X X 3.5-1 3.5-1  !

3.5-14 3.5-14 3.5-21 3.5-21 3.11-3 3.11-3 Figure 3.11-1 (Sheet 1) Figure 3.11-1 (Sheet 1)

Figure 3.11.1 (Sheet 2) Figure 3.11-1 (Sheet 2)

Figure 3.11.1 (Sheet 3) Figure 3.11-1 (Sheet 3)

Figure 3.11.1 (Sheet 4) Figure 3.11-1 (Sheet 4)

Figure 3.11-1 (Sheet 5) Figure 3.11-1 (Sheet 5)

Figure 3.11-1 (Sheet 6) Figure 3.11-1 (Sheet 6) 3.11-6 3.11-6 Removed Pagg UNIT 2 Insert Paag 3/4 2-2 3/4 2-2 3/4 2-3 3/4 2-3 3/4 2-4a 3/4 2-4a 3/4 2-4b 3/4 2-4b 3/4 2-4c 3/4 2-4c >

3/4 2-4d 3/4 2-4d 3/4 2-4e 3/4 2-4e 3/4 2-4f 3/4 2-4f 3/4 2-49 3/4 2-4g 3/4 2-4h 3/4 2-4h 3/4 2-41 3/4 2-41 3/4 3-30 3/4 3-30 3/4 5-5 3/4 5-5 ,

B 3/4 2-1 B 3/4 2-1 i B 3/4 2-2 B 3/4 2-2 l B 3/4 2-6 B 3/4 2-6 1

1679C E3-1 10/8/87 SL-3232 u_________-

ENCLOSURE 4 PLANT HATCH - UNITS 1, 2 NRC 00CKETS 50-321, 50-366 ,

. an OPERATING LICENSES DPR NPF-5 REQUEST TO REVISE TECHNICAL SPECIFICATIONS:

FUEL THERMAL LIMITS AND ECCS SURVEILLANCE REQUIREMENTS Enclosed is the following document:

"Edwin I. Hatch Nuclear Plant Units 1 and 2 SAFER /GESTR-LOCA Loss of Coolant Accident Analysis," NEDC-31376-P.

l 1679C E4-1 10/8/87 SL-3232