ML20135E155
| ML20135E155 | |
| Person / Time | |
|---|---|
| Site: | Dresden, Quad Cities  |
| Issue date: | 02/28/1997 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20135E152 | List: |
| References | |
| NUDOCS 9703060298 | |
| Download: ML20135E155 (6) | |
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-p UNITED STATES NUCLEAR REGULATORY COMMISSION
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4 WASNINGTON, D.C. 3e064001 t
SAFETY EVALUATION EY THE OFFICE OF NUCLEAR REACTOR REGULATION i
l RELATED TO AMENDMENT NO.153 TO FACILITY OPERATING LICENSE NO. DPR-19.
l AMENDMENT NO.148 TO FACILITY OPERATING LICENSE NO. DPR-25.
AMENDMENT NO. 172TO FACILITY OPERATING LICENSE No. DPR-29 l
j AND AMENDMENT NO.168 TO FACILITY OPERATING LICENSE NO. DPR-30
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C00940NWEALTH EDISON COMPANY AND l
MIDAMERICAN ENERGY COMPANY l
DEESDEN NUCLEAR POWER STATION. UNITS 2 AND 3. AND l
00AD CITIES NUCLEAR POWER STATION. UNITS 1 AND 2 l
DOCKET NOS. 50-237. 50-249. 50-254 AND 50-261 l
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1.0 INTRODUCTION
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By letter dated September 20, 1996, Commonwealth Edison Company (Comed, the i
licensee) submitted changes related to the pressure-temperature (P-T) limits in the Dresden Nuclear Power Station, Units 2 and 3, and the Quad Cities Nuclear Power Station, Units 1 and 2, Technical Specifications (TS). These i
changes include reformatting of TS 3/4.6.K from a single, three part paragraph i
to three separate paragraphs for clarity as well as specification of a time limitation in the ACTION statement for TS 3.6.
The licensee revised the P-T limits to provide new limits that are valid to 22 effective full power years i
(EFPY). The curves are based on the most limiting material among the group of four vessels.
In addition, the licensee requested to amend the TSs by incorporation of separate P-T limits for the vessel bottom head. The NRC 1
staff issued a Request for Additional Information (RAI) on December 9,1996, subsequent to a teleconference that occurred on November 26, 1996. Additional clarifying information that did not change the original proposed no significant hazards consideration determination was supplied by Comed in a letter dated January 21, 1997.
In the response to the RAI, Comed coinnitted to provide a revision to the i
supporting analysis, " General Electric Company (GE) report, GE-NE-B11-00707-l 0IR1, ' Pressure Temperature Curves for Dresden and Quad Cities Stations,'
j dated July 1996," to reflect the addition of one standard deviation to the electroslag (ES) weld chemistry mean values. The licensee also committed to j
revise the adjusted reference temperature (ART) tables for 18, 20 and 22 EFPY i
as a result of the change in the chemistry mean values of the ES welds. Both revisions are expected to be provided by April 18, 1997, at which time the 2
staff will verify all cosmiitments.
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9703060298 970228 PDR ADOCK 05000237 P
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I The staff evaluates the P-T limits based on the following NRC regulations and uidance: -Appendix G to 10 CFR Part 50; Generic Letters (GL) 88-11 and 92-01; ulatory Guide (RG) 1.99, Revision 2; and Standard Review Plan (SRP) Section 5.
2.
Appendix G to 10 CFR Part 50 requires that P-T limits for the reactor
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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 i
Pressure Vessel Code (Code). GL 88-11 recommends that licensees use the i
methods in RG 1.99, Revision 2, to predict the effect of neutron irradiation on the ART of reactor vessel materials. The ART is defined as the sum of initial nil-ductility transition reference temperature (RT ) of the material, the increase in RT.the prediction method. account for uncer on, and a margin to i
The increase in RT is l
calculated' from the product of a chemistry factor and a fluence factor. '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 j
vessel material as specified in Table 1 of RG 1.99, Revision 2.
GL 92-01 requires licensees to submit reactor vessel materials data, which the staff uses in the review of the P-T limits submittals.
i SRP 5.3.2 provides guidance on calculation of the P-T limits using linear elastic fracture mechanics methodology specified in Appendix G to Section III 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.
2.0 EVALUATION For the Dresden,' Units 2 and 3, and the Quad Cities, Units 1 and 2, reactor vessels, the licensee determined that the most limiting material among the 4
group of four vessels at the 1/4T and 3/4T locations is the lower intermediate to lower shell circumferential welds in the Dresden, Unit 3, vessel. This weld was fabricated using weld wire heat 299L44. The licensee calculated an ART of 97.4 degrees Fahrenheit at the 1/4T location and an ART of 81.2 degrees Fahrenheit at the 3/4T locatip at 22 EFPY. The neutron fluence used g the 2
2 ART calculation was 2.43 x 10 n/cm at the 1/4T location and 1.17 x 10 n/cm at the 3/4T location. The initial RT for the limiting weld was -5 degrees The margin term used in.,lculating the ART for the limiting weld Fahrenheit.
ca was 59 degrees Fahrenheit.
The staff performed an independent calculation of the ART values for the limiting materials using the methodology in RG 1.99, Revision 2.
Based on these calculations, the staff verified that the licensee's limiting material for the Dresden, Units 2 and 3, and the Quad Cities, Units 1 and 2, reactor vessels is the lower intermediate to lower shell circumferential welds in the f
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Dresden, Unit 3, vessel (heat 299L44). The staff's calculated ART values for i
j the limiting materials agreed with.the licensee's calculated ART values.
i Substituting the ART values for Dresden, Units 2 and 3, and Quad Cities, j
Units 1 and 2, into equations in SRP 5.3.2, the staff verified that the proposed P-T limits satisfy the requirements in Paragraphs IV.A.2 and IV.A.3 i
of Appendix G of 10 CFR Part 50.
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In addition to beltline materials, Ap>endix'G of 10 CFR Part 50 also imposes a minimum temperature at the closure toad flange based on the reference
' temperature for the flange material.Section IV.A.2 of Appendix G states that 5
when the pressure exceeds 20 percent of the preservice system hydrostatic test pressure, the temperature of the closure flange regions highly stressed by the 4
i solt preload must exceed the reference temperature of the material in those regions by at least 120 degrees Fahrenheit for normal operation and by 90 degrees Fahrenheit for hydrostatic pressure tests and leak tests. Based on the flange RT, of 23 degrees Fahrenheit for Dresden and Quad Cities provided by the licensee, the staff has determined that the proposed P-T limits have t
i satisfied the requirement for the closure flange region during normal j
operation, hydrostatic pressure test and leak test.
GE developed non-beltline P-T curves for a conservatively large BWR/6 vessel.
l These curves may be applied generically to other vessels by adding the highest to the pre., value of the applicable vessel component in the subject vessel initial RT l
T RT values from the GE ssure vs. temperature minus RT ' ap(proE)to develop separate vessel ClintonPowerStationusedthIs analysis.
bottom head curves, and this methodology was accepted by the staff in a letter from D. Pickett to P. Telthorst, dated October 23, 1996. The Dresden and Quad Cities P-T limits for the vessel bottom head were also based on the GE analysis. The highest initial RT for the Dresden and Quad Cities bottom head materials was 60 degrees Fah.,heit. The staff's calculations verified ren that the shift was consistent with the licensee's values, and that the P-T l
curves are at least as cons'ervative as that which would be generated using the methodologies of SRP Section 5.3.2 and Appendix G of the ASME Code.
a The ES welds are not the most limiting material among the group of four vessels and, therefore, were not used to develop the P-T curves. However,
, USE, and Chemical Composition of Core BAW 2258 and 2259, " Evaluation of RT.' Units 2 and 3, and Quad Cities, Region Electroslag Welds for Dresden, Units 1 and 2," January 1996, were referenced by the licensee, and reviewed by the staff in support of the submittal. These documents contain the licensee's of the ES assessment of the best estimate chemistry and the initial RT.,he mean value welds. The licensee's best estimate of the initial RT., was t of surveillance weld data as well as' drop weight data reported in the weld procedure qualifications (PQs) of ES welds fabricated in the same time frame as the Dresden and Quad Cities ES welds. The licensee determined the mean value of the initial RT to be 23.1 degrees Fahrenheit with a standard deviation of 13.0 degre., Fahrenheit. The licensee's best estimate of the es chemistry was also the mean of surveillance weld data and weld PQs of ES welds fabricated in the same time frame as the Dresden and Quad Cities ES welds.
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l The licensee determined the mean values of the copper (Cu) and nickel (Ni) contents to be 0.19 percent and 0.31 percent, respectively, with standard deviations of 0.048 percent for Cu and 0.051 percent for N1.
L The staff considers the licensee's proposed meth'od of detemining the initial RT, acceptable. The staff verified the initial RT and standard deviation l
vaIuesproposedbythelicensee._Theuncertaintyin.,thisestimateis l
accounted for by using the standard deviation of the initial RT., in the margin tem of the ART calculation.
The staff does not consider the initial proposal by the licensee for 1
i detemining the best-estimate chemistry acceptable because it did not meet the l
criteria in RG 1.99, Revision 2, nor was an alternative provided to the NRC.
This RG indicates that the best-estimate values of copper and nickel for the material will normally be the mean of the measured values for weld samples made with the same weld wire heat number as the vessel weld.
If these data f
are not available, a conservative estimate (mean plus one standard deviation) i based on generic data may be used if justified. The licensee could not j
identify the particular heats of weld wire that were used to fabricate the j
vessel ES welds. The licensee proposed generic best-estimate values based on j
the average of all the available data. However, RG 1.99, Revision 2, i
recommends the best-estimate copper and nickel values should be the mean plus one standard deviation values, when generic values are being used to determine the amount of embrittlement. The staff infomed the licen'see of its position in an RAI dated December 9, 1996.
4 In response to the staff RAI, the licensee reevaluated its position and will l
treat the data as generic. The revised best estimate chemistry will result i
from the addition of one standard deviation to the mean of the data. This methodology is consistent with RG 1.99, Revision 2, Position 1.1.
As mentioned, the licensee has committed to provide a revision to the supporting GE analysis and revised ART tables for 18, 20 and 22 EFPY by April 18, 1997.
However, since the revisions do not affect the limiting material, they will not affect the P-T limits.
3.0 CONCLUSION
The staff has perfomed an independent analysis to verify the licensee's proposed P-T limits. The staff concludes that the proposed P-T limits, including the separate P-T limits for the vessel bottom head, are valid to 22 EFPY since the limits conform to the requirements of Appendix G of 10 CFR Part 50 and GL 88-11. Hence, the proposed P-T limits may be incorporated in the Dresden, Units 2 and 3, and the Quad Cities, Units 1 and 2, TSs.
In addition, the proposed changes in the Bases section of the TSs are consistent with the P-T limits change, therefore, they are acceptable.
4.0 STATE CONSULTATION
In accordance with the Commission's regulations, the Illinois State official was notified of the proposed issuance of the amendments. The State official had no comments.
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5.0 ENVIRONMENTAL CONSIDERATION
The amendments change a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and change surveillance requirements. The NRC staff has determined that the amendments involve no significant increase in the amounts, and no i
significant change in the types, of any effluents that may be released l
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 4
j consideration, and there has been no public comment on such-finding i
(62FR66703). 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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6.0 CONCLUSION
The Commission has concluded, based on the considerations discussed above, that:
(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, I
and (3) the issuance of the amendments will not be inimica) to the common j
defense and security or to the health and safety of the public.
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Principal Contributor:
A. Lee i
Deta: February 28, 1997
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7.0 REFERENCES
1.
Regulatory Guide 1.99, " Radiation Embrittlement of Reactor Vessel 4
Materials," Revision 2, May 1988.
2.
NUREG-0800, Standard Review Plan, Section 5.3.L Pressure-Temperature Limits.
3.
Code of Federal Regulations, Title 10, Part 50, Appendix G, " Fracture Toughness Requirements."
4.
Generic Letter.88-II, "NRC Position on Radiation Embrittlement of Reactor Vessel Materials and its Impact on Plant Operations," July 12, 1988.
5.
American Society of Mechanical Engineers Boiler and Pressure Vessel Code,Section III, Appendix G for Nuclear Power Plant Components, Division 1, " Protection Against Nonductile Failure."
6.
Letter from P. Piet (Comed) to USNRC Document Control Desk, subject:
Dresden Nuclear Power Station Units 2 and 3 Quad Cities Nuclear Power
' Station Units I and 2 Changes to Pressure-Temperature (P-T) Curves, dated September 20, 1996.
7.
Letter from J. S. Perry (Comed) to USHRC Document Control Desk,
Subject:
Dresden Station Units 2 and 3 Quad Cities Unit's 1 and 2 Comed Response to NRC Request for Additional Information: P-T Curves, dated January 21, 1997.
1 8.
Letter from D. Pickett to P. Telthorst,
Subject:
Issuance of Amendment No. 109 to Facility Operating License No. NPF Clinton Power Station, Unit 1 (TACJio. M94887), dated October 23, 1996.
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