ML20138H402

Application for Amend to License DPR-57,revising TS 3.4.9 Re Reactor Coolant Sys Pressure & Temp Limits

ML20138H402
Person / Time
Site:	Hatch
Issue date:	04/29/1997
From:	Sumner H SOUTHERN NUCLEAR OPERATING CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20138H408	List: HL-5376, Proposed Tech Specs 3.4.9 Re Reactor Coolant Sys Pressure & Temp Limits ML20138H402
References
HL-5376, NUDOCS 9705070166
Download: ML20138H402 (6)

Text

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Lewis Sumner Southern Nuclear Vice Pregident Operating Company. Inc.

• Hatch Project Supprt 40 inverness Parkway

. Post Office Box 1295 Birmingharn. Alabama 35201 Tel 205992.7279 Fax 205.992.0341 SOUTHERN a-Energy to Serve nurWorld" April 29,1997 Docket No. 50-321 HL-5376 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555 Edwin 1. Hatch Nuclear Plant - Unit 1 Technical Specifications Revision Request for:

Unit 1 Pressure / Temperature Limits Gentlemen:

In accordance < .m ..: 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 1 Technical Specifications, Appendix A to Operating License DPR-57.

This supersedes the letter sent tn vou on April 14,1997 on the same subject. That submitta Ts sent to you pre a urely and did not include the revised Technical Specificar %res.

Changes are proposec :o Unit 1 Spec'ification 3.4.9, reactor coolant system pressure and temperature limits. Specifically, changes are proposed to the figures which contain the pressure / temperature limits for 1) reactor vessel pressure test,2) non-nuclear heat-up and cooldown and 3) criticality. The changes are made to reflect the latest information from the removal and evaluation of the Unit I reactor vessel material capsule. The capsule was removed duing the Unit I spring outage of 1996.

The evaluation and results of the material capsule surveillance are presented in General Electric report GE-NE-B1100691-01R1, dated March of 1997. This report was submitted

. to you on March 25,1997.

In accordance with the requirements of 10 CFR 50.91, a copy of this letter and all 0 G 014 2 applicable enclosures will be sent to the designated State oisial of the Environmental Protection Division of the Georgia Department of Natural Resources.

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U.S. Nuclear Regulatory Commission Page 2  ;

April 29,1997. j i

i Mr. H. L. Sumner, Jr., states he is Vice President of Southern Nuclear Operatmg  !

. 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 j true.

Sincerely, l i

& Ah  ;

H. L. Sumner, Jr. l l

Sworn to andsubscribed before me this80 day of 1997.

mA.dna y ary Public I l l

Comniission Expiration Date: [Av. cM /9Y L '

OCV/eb l I

Enclosures:

j

1. ' Description and Justification for Proposed Changes

2. 10 CFR 50.92 Evaluation  !

3 Page Change Instructions and Revised Technical Specifications Pages  !

cc: Southern Nuclear Operatine Company Mr. P. H. Wells, Nuclear Plant General Manager  ;

NORMS l U.S. Nuclear Regulatorv Commission. Washineton. D.C.

Mr. K. Jabbour, Licensing Project Manager - Hatch  !

t l US Nuclear Reeulatorv Commission Region 11 Mr. L. A. Reyes, Regional Administrator

! Mr. B. L. Holbrook, Senior Resident Inspector - Hatch I

a l State of Georeia Mr. J. D. Tanner, Commissioner - Department of Natural Resources 3

3, .

HL-5376 4

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• . Enclosure 1 1 Edwin I. Hatch Nuclear Plant - Unit 1 i Technical Specifications Revision Request for:

Unit 1 Pressure / Temperature Limits j i

Descrintion and Justification for Proposed Changes l l

l Descrintion of Change Southern Nuclear Operating Company (SNC) proposes to revise the Unit I reactor vessel pressure and temperature limits to reflect the data collected from the material sample recovered during the March 1996 Unit 1 Outage. Changes are proposed to Figure 3.4.9-1 (pressure / temperature limits for inservice hydrostatic and inservice leakage tests), i Figure 3.4.9-2 (pressure / temperature limits for non-nuclear heat-up, low power physics tests, and cooldown following a shutdown) and Figure 3.4.9-3 (pressure / temperature limits for criticality). These changes pertain to the Unit 1 Technical Specifications only.

Justificaticn for Change l

Pressure / Temperature (Pfr) limits are provided to prevent brittle fracture of the reactor vessel, and are required per Appendix G of 10 CFR 50. Appe.1?ix H requires that samples of the RPV material be periodically removed from the reactor vessel. The I

samples are evaluated for changes in the fracture toughness properties of ferritic materials in the RPV beltline region, which result from exposure to neutron irradiation and the thermal environment.  ;

l T'se RPV surveillance capsule removed contained flux wires for neutron flux monitoring and Charpy V impact and tensile test specimens. The irradiated material properties were compared to available unirradiated properties to determine the effect of irradiation on material toughness for the base and weld materials through Charpy testing. Irradiated tensile testing results are compared with unirradiated data to determine the effect of l irradiation on th; ess-strain relationship of the materials. These evaluations were perfonned per S methods of 10 CFR 50 Appendix H.

j General Electric report GENE-B1100691-01 R1, " Plant llatch Umt 1 RPV Surveillance l Materials Testing and Analysis," March,1997, documents the data and conclusions from l the review of the Unit 1 material sample pulled from the reactor vessel. From the results of this report, we are revising the Hatch 1 Technical Specification Pressure Temperature limit curves to reflect this latest information. There are no changes proposed to the Lir-i ing t Condition for Operation or to any of the surveillances of specification 3.4.9.

Add knally, the Unit 1 pressure test curve (Figure 3.4.9-1) is being revised to contain the latest exposure dependencies, starting from 16 EFPY to 32 EFPY. All the curves were generated based on the approved methodologies of 10 CFR 50 Appendix G.

HL-5376 El-1

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Enclosure 2

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l , ~10 CFR 50.92 Evaluation l

i In 10 CFR 50.92c, 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 j 50.22 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 and different kind from j any previously evaluated; or (3) involve a significant reduction in the margin of safety.

Southern Nuclear Operating Company has reviewed the proposed license amendment i request and determined its adoption does not involve a significant hazards consideration based on the following discussion.

Basis for no significant hazards consideration determination:

1. Does the change involve a sigm'ficant increase in the probability or consequences of  ;

an accidentpreviously evaluated? i Pressure and Temperature (P/F) limits for the reactor pressure vessel are established j to the requirements of 10 CFR 50, Appendix G to ensure brittle fracture of the l vessel does not occur.

This revision changes the P/F curves in the Unit 1 Technical Specifications to reflect the material capsule surveillance results from the sample removed during the l Spring outage of 1996.

The RPV surveillance capsule contained flux wires for neutron flux monitoring and l

Charpy V notch impact and tensile test specimens. The irradiated material properties were compared to available unirradiated properties to determine the effect ofirradiation on material toughness for the base and weld materials through Charpy testing. Irradiated tensile testing results are compared with unirradiated data to determine the effect ofirradiation on the stress-strain relationship of the materials.

The P/T curves are modified to reflect the results of the above examination. These curves and their operating limits were evaluated using the approved methodologies i

of 10 CFR 50 Appendix 0 and ASME Code Appendix G. The new curves therefore represent the latest information available on the state of the reactor vessel materials.

The P/T cerves are generated for reactor vessel protection against brittle fracture, they do not affect the recirculation piping. Accordingly, the probability of occurrence of a design basis Loss of Coolant Accident (LOCA) is not increased.

Likewise, no other previously evaluated accident and transients, as defined in IIL-5376 E2-2

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l Enclosure 2  ;

10 CFR 50.92 Evaluation j i

Chapter 14 of the Final Safety Analysis Report (FSAR) are affected by this proposed change to the Unit 1 P/T curves. Additionally, this proposed revision does not affect the design, operation, or maintenance of any safety related system designed for the mitigation or prevention of previously analyzed events.

Since no previously evaluated accidents or transients are being affected by this change, their probability of occurrence is not increased and their consequences are not made worse.

2. Do theproposed changes create the possibility ofa new and dgerent type of accidentfrom anypreviously evaluated?

Implementing the proposed P/r curves into the Unit 1 Technical Specifications does not alter the design or operation of any system or piece of equipment designed for the prevention or mitigation of accidents and transients. As a result, no new

' operating modes are introduced from which a new type accident becomes possible.

Existing systems will continue to be operated per present design basis assumptions.

The proposed P/r limits were generated from the evaluation of the material capsule removed during the Spring Unit 1 outage of 1996. As a result, these limits include the latest available information on the reactor vessel materials. Furtherraore, they will continue to be monitored per the requirements of the Technical Specifications and 10 CFR 50 Appendices G and H. For the above reasons, the changes do not create the possibility of a new type of accident.

3. Do the proposed changes involve a sigmficant reduction in the margin ofsafety?

1 The purpose of the P/r limita is to avoid a brittle fracture of the reactor vessel. As such, material capsules are removed periodically to determine the effects of neutron irradiation on reactor vessel materials. This change to the Unit 1 P/r curves is proposed to incorporate the evaluation results of the latest capsule removed during j the Spring Unit 1 outage of 1996. Accordingly, these curves represent the latest  :

information available on the reactor vessel materials. Also, the curves were  !

generated using the approved methodologies of 10 CFR 50 Appendix G. j The pressure test curve (Figure 3.4.9-1) is also being re"ised to reflect exposure dependencies. These curves were generated for exposures of 16,18,20,24,28, and 32 EFPY. As previously described, each of these curves were generated using approved methodologies and all reflect the results of this latest material capsule report. I l

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Enclosure 2

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'10 CFR 50.92 Evaluation The proposed change does not affect the evaluation of any FSAR Unit 1 Chapter 14 transient and accident. Furthermore, the proposed ch* age does not affect the operatior, of systems or equipment important to safety.  ;

The Limiting Cot , tion for Operation of Specification 3.4.9 will not change. Also, no Technical Specification surveillances or surveillance frequencies are revised as a l result of this Technical Specification submittal, besides the fact that the Pfr ,

surveillances will now refer to the revised curves. Procedures regarding the i monitoring of the P/T limits during reactor startup, cooldown, and leakage testing  !

will not change as a result of this proposed Technical Specification change with  !

respect to frequency of the surveillance or the methods used to perform the  ;

surveillances. Thus, the P/T limits will continue to be surveilled as before per the same procedures and the same frequencies.

No other Technical Specifications are affected by the proposed revision. The margin  ;

of safety to any Technical Specifications safety limit therefore is not reduced. 1 For the above reasons the new curves do not represent a significant reduction in the  !

margin of safety. i l

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