ML20217M852
| ML20217M852 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 08/19/1997 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20217M848 | List: |
| References | |
| NUDOCS 9708250330 | |
| Download: ML20217M852 (8) | |
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NUCLEAR REGULATORY COMMISSION 2
WASHINoTON, D.C. 30MH001 4.....,&
4 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 207 TO FACILITY OfERATING LICENSE DPR-57 SOUTHERN NUCLEAR OPERATING COMPANY. INC.. ET AL.
EDWIN 1. HATCH NUCLEAR plan 7. UNIT 1 DOCKET NO. 50-321
1.0 INTRODUCTION
By letter dated April 29, 1997, as supplemented May 28, 1997, Southern Nuclear Operating Company, Inc. (Southern Nuclear), et al. (the licensee) proposed a license amendment to change the Technical Specifications (TS) for the Edwin 1.
Hatch huclear Plant, Unit 1.
The proposed changes would revise the Unit I reactor vessel pressure and temperature limits to reflect data collected from the material sample recovered during the March 1996 Unit 1 outage.
2.0 EVALUATION By letter dated April 14, 1997, the licensee submitted proposed changes related to the pressure-temperature (P-T) limits in the Hatch Unit 1 TS. The submittal was incomplete and was superseded by a new submittal-dated April 29,-
1997. Additional information regarding initial reference temperatures was also supplied in a letter dated May 28, 1997.
The changes are the result of removal and evaluation of ti.e surveillance capsule at the 120' azimuthal location in the Hatch Unit I reactor vessel.
The capsule was removed at 14.3 effective full power years (EFPY). The licensee revised the P-T limits to provide new limits that are valid to 32 EFPY, Previously, a safety evaluation (SE) for the Hatch Units 1 and 2 P-T limits was completed and issued by letter dated April 4, 1997.
The April 4. SE did not reflect the results from the evaluation of the Hatch Unit I surveillance capsule since the information had not yet been provided.
In addition, separate limits wers, approved for the upper vessel and bottom head regions. The separate curv? were developed from the generic pressure (P) vs. temperature minus RT (i 43,) values frcm a General Electric GE analysis for 3 large boiling,,,w,ater t eactor/6 (BWR/6) reactor pressure ve(sse)l (see Section 2.2 of the April 4 SE for more detail).
9708250330 970819 PDR ADOCK 05000321 P
_ PRESSURE-TEMPERATURE LIMITS The staff evaluates the P-T limits based on the following NRC regulations and guidance: Appendix G to 10 CFR Part 50; Generic letters (GLs) 88-11 and 92-01; Regulatory Guide (RG) 1.99, Revision 2; and Standard Review Plan (SRP)
Secti.on 5.3.2.
Appendix G to 10 CFR Part 50 requires that P-T limits for the reactor vessel must be at least as conservative as those obtained by Appendix G to Section XI of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (ASME Code). GL 88-11 requires that licensees use the methods in RG 1.99, Revision 2, to predict the effect of neutron irradiation on the adjusted reference temperature (ART) of reactor vessel materials.
The ART is defined as the sum of initial nil-ductility transition reference temperature (RTm ) of the material, the increase in RT caused by neutron irradiation (6RT m
uncertainties in the prediction meby)d., and a margin to account for o
TheincreaseinRT, Tactor.is calculated from the product of a chemistry factor (CF) and a fluency The chemistry factor may be calculated using credible surveillance data, obtained by the licensee's surveillance program, as directed by Position 2 of RG 1.99, Revision 2.
If credible surveillance data are not available, the chemistry factor is calculated dependent upon the amount of copper and nickel in the vessel material as s)ecified in Table 1 of RG 1.99, Revision 2.
GL 92-01 requires licensees to su)mit reactor vessel materials data, which the staff uses in the review of the P-T limit submittal s.
Standard Review Plan 5.3.2 provides guidance on calculation of the P-T limits using linear elastic fracture mechanics methodology specified in Appendix G to Section 111 of the ASME Code.
The linear elastic fracture mechanics methodology postulates sharp surface defects that are normal to the direction of maximum stress and have a depth of one-fourth of the reactor vessel beltline thickness (1/4T) and a length of 1-1/2 times the beltline thickness.
The critical locations in the vessel for this methodology are the 1/4T and 3/4T locations, which correspond to the maximum depth of the postulated inside surface and outside surface defects, respectively.
E0VIVALENT MARGIN ANALYSIS Appendix G also requires that the predicted Charpy upper-shelf energy (USE) at end-of-license (EOL) for vessel beltline materials be above 50 ft-lb or that 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. ASME Code Case N-512 and Appendix K contain analytical procedures and acceptance criteria for demonstrating that reactor vessel beltline materials with low Charpy USE will have margins of safety against fracture equivalent to Appendix G of the ASME Code.
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! In a December 9.-1993, letter to L.A. England from J.T. Wiggins staff issued the Safety Evaluation Report of the GE topical repor(USNRC), the t NEDO-32205, Revision 1, "10 CFR 50, Appendix G. Equivalent Margin Analysis for low Up)er Shelf Energy in BWR/2 through BWR/6 Vessels.' The staff concluded that tie reactor pressure vessels of the participating utilities should have margins of safety against ductile fracture in low USE plates and welds until their expiration of licenses (E0L) for level A, B, C, and D conditions, and meet the criteria of ASME Code Case N-512 and Appendix K.
Individual licensees that reference the topical report as their basis for addressing the USE requirements of 10 CFR Part 50. A)pendix G, were requested to confirm the plant-specific applicability of t1e topical report by com)aring the predicted percentage drop in the USE to the allowable decrease in tie USE from the topical report.
The April 4, P-T limits SE that was issued by the staff also requested that the licensee address the plant-specific applicability of the Hatch reactor pressure vessel materials to the GE topical report NED0-32205, Revision 1, "10 CFR 50, Appendix G. Equivalent Margin Analysis for low Upper Shelf Energy in BWR/2 through BWR/6 Vessels.".The plant-specific equivalent margins analysis (EMA) for Hatch Unit I was included in the current submittal.
The results from the EMA were compared to the allowable decrease in USE from topical report NEDO-32205.
PRESSURE-TEMPERATURE LIMITS As part of the review, the basis for the initial reference temperature values for all beltline materials were revisited since many of the values differed from the values that were reported in response to GL 92-01.
In response to GL 92-01, the limiting plate initial'RT of 10'F was conservatively applied m
to all beltline plates even though data were available for the other plates.
Similarly, the limiting wnld initial RT,Nayof -10'F was conservatively applied to all beltline welds.
By letter dated 28, 1997, the licensee provi<ied detailed justification for all initial RT"8,all calculations and raw data that values including the certified material test reports.
The staff reviewe were used to determine the initial RT and found all values acceptable.
Table 1 shows a comparison of the preNous values and the values reported in this submittal. The changes will be included in the next update of the reactor vessel integrity database.
RG 1.99, Revision 2, Position 2.1, requires that the chemistry factor (CF) for welds be adjusted based on credibit surveillance capsule test results.
In this procedure, the adjustment is based on the ratio of the chemistry factor from the surveillance material and the chemistry (copper and nickel) of the weld.
The licensee used a similar procedure for its plate surveillance material. The CF that results from the least squares fit of the surveillance data is 221.7'F.
From Table 2 of RG 1.99, Revision 2, the CF for the surveillance plate with copper = 0.12% and nickel - 0.70% is 84.5'F.
Therefore, the plate adjustment is 2.62 (221.7/84.5). This adjustment was conservatively applied to all of the beltline plates.
No unirradiated data were available for the weld material, so the CFs for the welds were calculated using Fo:,ition 1 of RG 1.99, Revision 2.
b
- - -_ _ -__ For the Hatch Unit I reactor vessel, the licensee determined that the most limiting material at the 1/4T and 3/4T locations is the lower-intermediate plate G-4803-7. This plate was fabricated using plate heat C4337-1.
The licensee cciculated an ART of 153'F at the 1/4T location and ll2*F at the 3/4T location,at 32'EFPY. The neutron fluency used inJhe AgT calculation was 1.3 x 10 n/cm at the 1/4T location and 0.6 x 10 n/cm at the 3/4T location. The initial RT for the limiting plate was -20'F.
The margin term used in calculating die ART for the limiting plate was 17'F as permitted by Position 2.1 of RG 1.99, Revision 2.
The staff performed an independent calculation of the ART values for the limiting material using the methodology in RG 1.99, Revision 2.
Based on these calculations, the staff verified that the licensee's limiting material for the Hatch Unit I reactor vessel is the lower-intermediate plate G-4803-7 that was fabricated using plate heat C4337-1.
The staff's calculated ART value for the limiting material agreed with the licensee's calculated ART value.
In support of the initial reference temperature evaluation, the staff reviewed weld wire heat data in the reactor vessel integrity database (RVID).
Other initial reference temperature values from plants with data from welds fabricated using the same heats of weld wire as in the Hatch Unit I welds were reported for heats 13253 and 33A377.
If generic initial reference temperature values are used for weld wire heats 13253 and 33A377, the resulting ART values increase slightly. However, the welds do not become limiting.
Substituting the ART values for the Hatch Unit 1 limiting plate into equations in SRP 5.3.2, the staff verified that the proposed P-T limits satisfy the requirements in Paragraph IV.A.2 of Appendix G of 10 CFR Part 50.
In addition to beltline materials, Appendix G of 10 CFR Part 50 also imposes a minimum temperature at the closure head flange based on the reference temperature for the flange material.
Section IV.A.2 of Appendix G states that when the pressure exceeds 20% of the preservice system hydrostatic test
)ressure, the temperature of the closure flange regions highly stressed by the
)olt 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 RT of 16'F for Hatch Unit 1providedbythelicensee,thestaffhasdetermiIe'dthattheproposed P-T limits have satisfied the requirement for the closure flange region during normal operation and hydrostatic pressure test and leak test.
E0VIVALENT MARGIN ANALYSIS Methods acceptable to the staff for determining the percentage decrease in USE are documented in RG 1.99, Revision 2.
Figure 2 in the RG indicates that the cercentage decrease in USE increases with increasing amounts of copper and neutron fluency.
However, the percentage decrease in USE could be affected by
. surveillance test results.
If surveillance data indicate that the percentage decrease in USE is greater than the amount predicted by Figure 2 in the RG, the percentage decrease in USE for the material must be increased.
If surveillance data indicate that the percentage decrease in USE is less than the amount predicted by Figure 2, the percentage decrease in USE for the material may be decreased from the amount predicted by Figure 2.
In the current submittal, the licensee compared the USE decrease from the surveillance plate to the amount of decrease predicted by RC 1.99, Revision 2, and the allowable limits in NE00-32205. The licensee reported that the percent decrease in USE values for the surveillance plate are less than that from using Figure 2 in the RG.
In addition, the predicted USE decrease of 18%
for the limiting plate is less than the allowable limit of 21% for plates from NEDO-32205. The predicted USE decrease of 28% for the limiting weld is less than the allowable limit of 34% for welds from NED0-32205.
Since no unirradiated data were available for the weld material, the licensee used Figure 2 in the RG to obtain the percent decrease in USE value. The parcent decrease in USE from the first to the second capsule was reported for information only, and was not used to obtain the USE decrease of 28% for the limiting weld. Therefore, both plates and welds meet the allowable limits of NED0-32205.
3.0 STAFF CONCLUSION PRESSURE-TEMPERATURE LIMITS The staff has performed an independent analysis to verify the licensee's proposed P-T limits.
The staff concludes that the proposed P-T limits are valid to 32 EFPY since the limits conform to the requirements of Appendix G of 10 CFR Part 50 and GL 88-11.
Hente, the proposed P-T limits may be incorporated in the Hatch Unit 1 Technical Specifications.
E0VIVALENT MARGIN ANALYSIS Based on its evaluation, the tl'/f concludes that the projected decreases in USE for the beltline materials are less than the allowable decreases in USE from topical report NED0-32205. Consequently, the applicability requirements of NEDO-32205 have been satisfied and the conclusions of the topical report are applicable to the Hatch Unit I reactor vessel. As a result, the Hatch Unit I reactor vessel satisfies the criteria in ASME Code Case N-512 and Appendix K, and is arojected to have margins of safety against fracture that are equivalent to tiose required by Appendix G of the ASME Code at expiration of license.
Therefore, the Hatch Unit I reactor vessel also meets the requirements of Appendix G to 10 CFR Part 50, l
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- TABLE 1 COMPARISON OF PREVIOUS AND CURRENT INITIAL REFERENCE TEMPERATURE VALUES BELTLINE NEAT.
PREVIOUS CURRENT DESCRIPTION / WELD TYPE IDENTIFICATION INITIAL REFERENCE REFERENCE TEMPERATURE TEMPERATURE VALUE('F)
VALUE (*F)
PLATES:
Lower G-4805-1 C4112-1 10 8
G-4805-2 C4112-2 10 10 G-4805-3 C4149-1 10
-10 Lower-Intermediate G-4803-7 C4337-1 10
-20 G-4804-1 C3985-2
-10
-20 G-4804-2 C4114-2 10
-20 WELDS:
Lower Longitudinal 1-307 13253
-10
-50 Lower-Intermediate IP2809
-10
-50 Longitudinal 1-308 IP2815
-10
-50 Lower to Lower-Int.
90099
-10
-10 Girth 1-313 i
33A277
-10
-50
i l
3.0 STATE CONSULTATION
In accordance with the Commission's regulations, the Georgia State official was notified of tbg proposed issuance of the amendment. The State official had no comments.
4.0 ' ENVIRONMENTAL CONSIDERATION The amendment changes a requirement with respect to the installation or use of facility components located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendment involves 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 amendment involves no significant hazards consideration, and there has been no public comment on such finding (62 FR 38138 dated July 16, 1997.
Accordingly, the amendment meets the eligibility criteria for ca)tegorical exclusion set forth in 10 CFR Sl.22(c)(9).
Pursuant to 10 CFR 51.22(b no environmental impact statement or environmental assessment need be prep)ared in connection with the issuance of the amendment.
5.0 CONCLU$1QH The Commission has concladed, based on the considerations discussed above, that: _ ii public w(1)l not be endangered by operation in the proposed manner,y of the there is reasonable assurance that the health and safet activities will be conducted in compliance with the Commission's reg (2) suchulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
Attachment:
References Principal Contributor:
A. D. Lee Date:
August 19, 1997
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. REFERENCES 1.
Regulatory Guide 1.99, Radiation Embrittlement of Reactor Vessel Materials, Revision 2, May 1988.
2.
NUREG-0800, Standard Review Plan, Section 5.3.2:
Pressure-Temperature i
Limits.
' Code of Federal Reaulations, Title 10, Part 50, Appendix G. Fracture 3.
Toughness Requirements.
4.
Generic Letter 88-11, NRC Position on Radiation Embrittlement of Reactor Vessel Materials and its impact on Plant Operations, July 12, 1988.
5.
ASME Boiler and Pressure Vessel Code, Section 111, Appendix G for Nuclear Power Plant Components, Division 1, " Protection Against Nonductile Failure."
6.
October 23, 1996, Letter from D. Pickett to P. Telthorst,
Subject:
Issuance of Amendment No. 109 to Facility Operating License No.
NPF-62-Clinton Power Station, Unit 1 (TAC # M94887).
7.
February 28, 1997, Letter from J.F. Stang to 1. Johnson,
Subject:
Issuance of Amendments (TAC NOS M96898, M96899, M96900 AND M96901).
8.
April 4,1997, Letter from K.H. Jabbour (USNRC) to H.L. Sumner, Jr.,
Subject:
Issuance of Amendments - Edwin I, Hatch Nuclear Plant, Units 1 and 2 (TAC NOS. M96609 and M96610).
9.
April 14, 1997, Letter from H.L. Sumner, Jr., to USNRC Document Control Desk,
Subject:
Edwin 1. Hatch Nuclear Plant - Unit 1 Technical Specifications Revision Request for: Unit 1 Pressure / Temperature Limits.
10.
April 29, 1997, Letter from H.L. Sumner, Jr., to USNRC Document Control Desk,
Subject:
Edwin 1. Hatch Nuclear Plant - Unit 1 Technical Specifications Revision Request for: Unit 1 Pressure / Temperature Limits.
11.
May 28, 1997, Letter from H.L. Sumner, Jr., to USNRC Document Control Desk,
Subject:
Edwin 1. Hatch Nuclear Plant - thit 1 Response to Verbal Request for Information on P-T limits Technical Specification Revision Request.
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