ML20137N181

Safety Evaluation Supporting Amends 206 & 147 to Licenses DPR-57 & NPF-5,respectively

ML20137N181
Person / Time
Site:	Hatch
Issue date:	04/04/1997
From:	NRC (Affiliation Not Assigned)
To:
Shared Package
ML20137N171	List: ML20137N167 ML20137N175 ML20137N181
References
NUDOCS 9704080280
Download: ML20137N181 (6)

Text

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,e* Rie p *, UNITED STATES j

! :s NUCLEAR REGULATORY COMMISSION f 2 WASHINGTON D.C. 20066-0001

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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION

} RELATED TO AMENDNENT NO. 206 TO FACILITY OPERATING LICENSE DPR-57

AND AMENDMENT NO. 147 TO FACILITY OPERATING LICENSE NPF-5 i

SOUTHERN NUCLEAR OPERATING COMPANY. INC.. ET AL.

j EDWIN I. HATCH NUCLEAR PLANT. UNITS 1 AND 2 j DOCTET NOS. 50-321 AND 50-366 l

1.0 INTRODUCTION

i By letter dated September 19, 1996, as supplemented December 17, 1996,

! January 23 and 31, March 21 and April 4,1997, Georgia Power Company, and l Southern Nuclear Operating Company, Inc., et al.- (the licensee), proposed

license amendments to change the Technical Specifications (TS) for the Edwin i I. Hatch Nuclear Plant, Units 1 and 2 [ Reference 1). The amendments request i the extension of the validity of the Unit I curves to 32 effective full power i years (EFPY) and % allow separate monitoring of the three major regions of j the reactor pren.ne vessel for both units by providing separate pressure and j temperature (P/i) limits for the upper vessel, beltline, and bottom head i regions for the inservice leak and hydrostatic tests curve and the heatup and

! cooldown curve. In atJition, changes to the surveillance requirements in the l TS were requested to clarify the applicable conditions to avoid confusion

experienced in past refueling outages. The December 17, 1996, January 23 l and 31, March 21 and April 4,1997, letters provided clarifying information

! that did not change the initial proposed no significant hazards consideration

! determination.

The staff evaluates the P/T limits based on the following NRC regulations and

, guidance: 10 CFR Part 50, Appendix G; Generic Letter (GL) 88-11; GL 92-01,

, Revision 1; GL 92-01, Revision 1, Supplement 1; Regulatory Guide (RG) 1.99, Revision 2; and Standard Review Plan (SRP) Section 5.3.'2. GL 88-11 advised ,

licensees that the staff would use RG 1.99, Revision 2, to review P/T limit l curves. RG 1.99, Revision 2, contains methodologies for determining the increase in transition temperature and the decrease in upper-shelf energy (USE) resulting from neutron radiation. GL 92-01, Revision 1, requested that licensees submit their reactor pressure vessel (RPV) data for their plants to 1 the staff for review. GL 92-01, Revision 1, Supplement 1, requested that l licensees provide and assess data from other licensees that could affect their  !

RPV integrity evaluations. These data are used by the staff as the basis for i the staff's review of P/T limit submittals, and as the basis for the staff's  !

review of pressurized thermal shock (PTS) assessments (10 CFR 50.61 l assessments). Appendix G to 10 CFR Part 50 requires that P/T limits for the '

RPV be at least as conservative as those obtained by applying the methodology of Appendix G to Section XI of the American Society of Mechanical Engineers Boiler and Fressure Vessel (ASME) Code.

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l l SRP 5.3.2 provides an acceptable method of calculating the P/T limits for

! ferritic materials in the beltline of the RPV based on the linear elastic ,

j fracture mechanics (LEFM) methodology of Appendix G to Section XI of the ASME l l Code. The basic parameter of this methodology is the stress intensity factor

K,, which is a function of the stress state and flaw configuration. The j methods of Appendix G postulate the existence of a sharp surface flaw in the RPV that is normal to the direction of the maximum stress. This flaw is l postulated to have a depth that is equal to one-fourth of the RPV beltline

thickness and a length equal to 1.5 times the RPV beltline thickness. The

critical locations in the RPV beltline region for calculating heatup and cooldown P/T limit curves are the 1/4 thickness (1/4T) and 3/4 thickness i (3/4T) locations, which correspond to the ciepth of the maximum postulated

) flaw, if initiated and grown from the inside and outside surfaces of the RPV, I respectively.

I The Appendix G, ASME Code methodology requires that licensees determine the l and the Charpy USE at the adjusted reference maximum postulatedtemperature flaw depth. The(ARTARTor isRT de7) fined as the sum of the fr.itial l (unirradiated) reference temperature (initial RT ), the mean value of the l ra , and a

~l adjustment margin in reference (M) term. The ART temperature caused by ir.,diation (ART ,)d a fluenci

. factor. The chemistry fa.,

ctor isadependent is product ofupon a chemistry the amount factor an and of copper 1

nickel in the material and may be determined from tables in RG 1.99, J

Revision 2, or from surveillance data. The fluence factor is dependent upon the neutron fluence at the maximum postulated flaw depth. The margin term is dependent upon whether the initial RT , is a plant-specific or a generic

, value and whether the chemistry factor was determined using the tables in i RG 1.99, Revision 2, or surveillance data. The margin term is used to account

! for uncertainties in the values of initial RT ,, copper and nickel contents,

fluence and calculational procedures. RG1.9I, Revision 2,describesthe i methodology to be used in calculating the margin term.

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! 2.0 EVALUATION

! The staff evaluated the effect of neutron irradiation embrittlement on each

! beltline material in the reactor vessels of Hatch Units 1 and 2. The amount i of irradiation embrittlement was calculated in accordance with RG 1.99, i Revision 2. The staff has determined that the material with the highest ART l at 32 EFPY for Unit 1 is the circumferential weld 1-313A, which was fabricated j with weld wire heat number 90099 (weld 90099), with 0.17% copper (Cu), 1.00%

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nickel (Ni), and an initial RT , of -10*F. The material with the highest ART at 32 EFPY for Unit 2 is the lower shell axial weld 101-842, which was i fabricated with weld wire heat number 10137, with 0.23% Cu, 0.50% Ni, and an initial RT The ARTS calculated by the staff are 159.9'F for the limiting material

, of -50*F.

of Unit I and 70.9'F for the limiting material of Unit 2.

Both ARTS are epiculated at 1/4T at 32 EFPY with 2 corresponding neutron fluence of 0.19E19 n/cm for Unit I and 0.154E19 n/cm for Unit 2. The ARTS

, calculated by the licensee, using the Chemistry Factor Table in Section 1.1 of

RG 1.99, Revision 2, are 163.9'F for Unit I and 71.9'F for Unit 2.

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The differences between staff-calculated ARTS and licensee-calculated ARTS are i 4.0*F for Unit I and 1.0*F for Unit 2. These differences between the staff and licensee-calculated ARTS were not significant since the differences are i small and the licensee's values are conservative. Substituting, respectively,

the ARTS of 163.9'F for Unit I and 71.9'F for Unit 2 into equations in 1 SRP 5.3.2, the staff verified that the proposed P/T limits (for 32 EFPY) for j heatup, cooldown, and hydrotest meet the beltline material requirements in

Appendix G of 10 CFR Part 50.

j In the licensee's first response [ Reference 2) to the staff's request for j additional information (RAI), the licensee revised the P/T limits for Unit 1 to include also lim 4s for 16 EFPY. These additional P/T limits for 16 EFPY were also evaluatmi by the staff in the manner previously described. The i 16 EFPY P/T limits met the beltline material requirements in Appendix G of i 10 CFR Part 50. In addition, Reference 2 contains the methodology for j generating the P/T limits for the upper vessel and the bottom head to

substantiate the separate P/T limits for these two regions. This involves l developing the plant-specific P/T limits for the upper vessel and bottom head

from the generic pressure (P) vs. ^seperature minus RT (T-RT 7 limits that were developed by General Electric (GE) for a large boning waIer) reactor (BWR) Type-6 vessel through finite element analyses. Plant-specific P/T limits such as those in the current submittal are then derive from the limits.

9eneric P vs.

by the staff as T-RT.'

part of the P/T limits for the bottom head in an IllinoisThis methodology Power Company su)mittal for the Clinton Power Station dated February 22, 1996.

This methodology was approved by the staff in an October 23, 1996, letter to Illinois Power Company. The staff reviewed the information in Reference 2 and concludes that the GE methodology applies to the upper vessel as well. Also, separating P/T limits into three sets of limits for different regions of the vessel in TS Figures 3.4.9-1 and 3.4.9-2 is acceptable for both units since the licensee can monitor these three regions separately (Reference 5].

In addition to beltline materials, Appendix G of 10 CFR Part 50 also imposes P/T limits based on the reference temperature for the reactor vessel closure flange materials.Section IV.A.2 of Appendix G states that when the pressure exceeds 20% of the preservice system hydrostatic test pressure, the temperature of the closure flange regions highly stressed by the bolt preload must exceed the reference temperature of the material in those regions by at least 120*F for normal operation and by 90*F for hydrostatic pressure tests and leak tests. Based on the flange reference temperatures of 16*F for Unit I and 30*F for Unit 2, the staff has detemined that the proposed P/T limits satisfy the requirements in Section IV.A.2 of Appendix G.

It is important to note that the initial RT 7 value for weld 33A277 for Unit I that is currently in the Reactor Ves ,sel Integrity Database (RVID) should be revised from -10*F to -50*F. This revision would cause the identity of the limiting beltline material for Unit I to change from weld 33A277 to weld 90099. The initial RT for weld 33A277 was reported to be -10*F in the licensee's latest response E GL 92-01, Revision 1. However, in the second and third responses (References 3 and 4) to the staff's RAI, the licensee revised this value to -50*F based on the GE RT y estimation method that was approved by the staff on December 16, 1994. Sestaffreviewedtheinitial

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. values for all RPV welds with heat numbers 33A277 and 90099 weld wire in j RT thI ,RVID and concluded that the licensee's values was conservative.

i Therefore, this revision to the RVID data for Unit 1 is acceptable.

! Appendix G also requires that the predicted Charpy upper-shelf energy (USE) at

and-of-license (EOL) for vessel beltline saterials be above 50 ft-lb or that i licensees demonstrate that lower values of Charpy USE will provide margins of safety equivalent to those required by Appendix G of Section XI of the ASME Code. The Boiling Water Reactor Owners Group (BWROG) proposed an equivalent margins analysis in a topical report, NEDO-32205, Revision 1, to demonstrate RPV materials with E0L USE values lower than 50 ft-lb will provide margins of safety equivalent to those required by Appendix G of Section XI of the ASME Code. The plant-specific applicability of the Hatch RPV materials to NEDO-32205 will be addressed in a separate correspondence. The equivalent margins analysis will not affect the Hatch P/T limits.

The staff concludes that the proposed P/T limits for the reactor coolant system for heatup, cooldown, leak test, and criticality are valid as indicated on the curves. The Unit 1 P/T limits satisfy the requirements of Appendix G of 10 CFR Part 50 for 16 EFPY and 32 EFPY and the Unit 2 P/T limits satisfy these requirements for 32 EFPY. The proposed P/T limits also satisfy GL 88-11 because the method in RG 1.99, Revision 2, was used to calculate the ART.

Therefore, the proposed P/T limits may be incorporated into the Hatch Units 1 and 2 Technical Specifications. In addition, the staff has reviewed and accepted the changes to the P/T limit surveillance requirements in the TS, which were made to clarify the applicable requirements and to avoid confusion experienced by the licensee in past refueling outages.

3.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Georgia State official was notified of the proposed issuance of the amendments. The State official had no comments.

4.0 ENVIRONMENTAL CONSIDERATION

The amendments change surveillance requirements. The NRC staff has determined that the amendments involve no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration, and there has been no public comment on such finding (62 FR 128 dated January 2, 1997). Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.

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5.0 CONCLUSION

i The Commission has concluded, based on the considerations discussed above, l

, that: (1) there is reasonable assurance that the health and safety of the l public will not be endangered by operation in the proposed manner, (2) such l activities will be conducted in compliance with the Commission's regulations, '

and (3) the issuance of the amendments will not be inimical to the common i l defense and security or to the health and safety of the public. l Principal Contributor: S. Sheng 1

Date: April 4, 1997 i

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REFERENCES

1. September 19, 1996, letter from J. T. Beckham, (GPC) to USNRC Document Control Desk, subject: "Edwin I. Hatch Nuclear Plant - Request to revise Technical Specifications: Pressure-Temperature Limits."

2. December 17, 1996, letter from J. T. Beckham, (GPC) to USNRC Document Control Desk, subject: "Edwin I. Hatch Nuclear Plant - Response to '

Request for Additional Information: Technical Specifications Revision Request on Pressure-Temperature Limits."

3. January 23, 1997, letter from J. D. Woodard, (GPC) to USNRC Document Control Desk, subject: "Edwin I. Hatch Nuclear Plant - Response to Request for Additional Information on Technical Specification Revision Request: Pressure-Temperature Limits."

4. January 31, 1997, letter from J. D. Woodard, (GPC) to USNRC Document Control Desk, subject: "Edwin I. Hatch Nuclear Plant - Response to Request for Additional Information on Technical Specification Revision Request: Pressure-Temperature Limits." l

5. March 21, 1997, letter from H. L. Sumner, (GPC) to USNRC Document Control Desk, subject: "Edwin I. Hatch Nuclear Plant - Request for Additional Information on Pressure / Temperature Technical Specification Revision Request."

6. April 4,1997, letter from John Lamberski (Troutman Sanders LLP) to the USNRC Document Control Desk, subject: "Edwin I. Hatch Nuclear Plant and  ;

Vogtle Electric Generating Plant - Request Regarding Pending Georgia  !

Power Company Submittals."