ML18029A872
| ML18029A872 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 09/16/1985 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML18029A871 | List: |
| References | |
| NUDOCS 8509230141 | |
| Download: ML18029A872 (5) | |
Text
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~Ig IP ky*yW UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO.
121 TO FACILITY OPERATING LICENSE NO.
DPR-33 AMENDMENT NO. 116 TO FACILITY OPERATING LICENSE NO.
DPR-52 AMENDMENT NO.
92 TO FACILITY OPERATING LICENSE NO. DPR-68 TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT UNITS 1
2 AND 3 DOCKET NOS. 50-259, 50-260 AND 50-296
1.0 INTRODUCTION
By letter dated September 22, 1983 (TVA BFNP TS 191),
as supplemented March 20, 1985, the Tennessee Valley Authority (licensee/TVA) requested amendments to Facility Operating License Nos.
DPR-33, DPR-52 and DPR-68 for the Browns Ferry Nuclear Plant, Units 1, 2 and 3.
The licensee requested amendments to the Technical Specifications (TS) to revise the present pressure-temperature limit curves (Figure 3.6-1).
Periodic revision of these curves is required to account for the loss of reactor vessel material toughness resulting from the accumulated radiation exposure to the vessel with time.
The licensee indicates that the proposed pressure-temperature limits are applicable for a period corresponding to 12 effective full power years (EFPY).
Submitted in the September 22, 1983 letter were three Southwest Research Institute Reports:
No. 02-4884-001, titled, "Analysis of the Vessel Wall Neutron Dosimeter From Browns Ferry Unit 1 Pressure Vessel; No. 01-4884-002 titled, "Analysis of the Vessel Wall Neutron Dosimeter From Browns Ferry Unit 2 Pressure Vessel; and No. 02-4884-003 titled, "Analysis of the Vessel Wall Neutron Dosimeter From Browns Ferry Unit 3 Pressure Vessel.
Submitted in the March 20, 1985 letter was Babcock 5 Wilcox Report BAW-1845, titled, "Browns Ferry Core Region gaterials Information (Units 1, 2 and 3)."
2.0 EVALUATION Appendix G, "Fracture Toughness Requirements,"
and Appendix H, "Reactor Vessel Material Surveillance Program Requirements,"
10 CFR Part 50, describe the conditions that require pressure-temperature limits for the reactor coolant pressure boundary and provide the gen'eral bases for these limits.
These appendices specifically require that pressure-temperature limits must provide safety margins for the reactor coolant pressure boundary at least as great as the safety margins recommended in the ASME Boiler and Pressure Vessel Code,Section III, Appendix G, "Protection Against Nonductile Failure."
Appendix G, 10 CFR Part 50, requires additional safety margins whenever the reactor core is critical, except for low-level physics tests.
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The fracture toughness of all ferritic steels gradually decreases with exposure to fast neutrons above a threshold value.
To adjust for this, the minimum operating temperature vs. pressure curves need to be revised.
The present curves in Figure 3.6-1, which were based on a shift in RT of 30 F, were approved for use through 4.0 EFPY.
The Browns Ferry rector vessels have exceeded this exposure.
By application dated September 23, 1983, the licensee proposed revised curves.
Our review of this submittal determined that additional information (as described in our letter of January 23, 1984) was needed on the chemical composition and test results on the weld and plate material used to fabricate the reactor vessels.
The information requested was provided by TVA'S letter of March 20, 1985.
The March 20, 1985 submittal provided clarifying information to support the licensee's materials evaluation and did not significantly change the initial application.
Based on the irradiation data, TVA concludes that a 45 F shift in RT>> will not occur for at least 12 EFPY of operation.
This would cover about Phe next decade of operation for each Browns Ferry unit.
Prressure-temperature limits must be calculated in accordance with the requirements of Appendix G, 10 CFR 50, which became effective on July 26, 1983.
Pressure-temperature limits that are calculated in accordance with the requirements of Appendix G, 10 CFR 50 are dependent upon the initial RTN for the limiting materials in the beltline and closure flange regions of Ne reactor vessel and the increase in RT resulting from neutron irradiation damage to the limiting beltline Ilerial.
The Browns Ferry FSAR Section 4.2 indicates that all materials in the closure flange region will have a nil-ductility transition (NDT) temperature, as determined by ASTM E 208, of a maximum of 10'F.
The FSAR indicates that material used to fabricate the closure flange forging met the requirements of ASME Code Specification SA 508 C1.2, Code Case 1332.
According to Table 4.4 of NUREG 0577, "Potential for Low Fracture Toughness and Lamaller Tearing on PWR Steam Generator and Reactor Coolant Pump Supports,"
an upper bound estimate of the RTNDT for this material is 40'F..
The amount of time that pressure-temperature limits are effective depends upon the initial RT and the amount of neutron irradiation damage to the limiting beltline mKrial.
The amount of neutron irradiation damage calculated in accordance with Regulatory Guide 1.99, Rev.
1, "fffects of Residual Elements on the Predicted Radiation Damage to Reactor Vessel Materials" is dependent upon the amount of neutron fluence (E less than lMeV) and the percentage of phosphorus (P) and copper (Cu) in the limiting beltline material.
BAW-1845 indicates that the beltline of each Browns Ferry reactor'essel consists of six plates, six longitudinal weld seams and one circumferential weld seam.
This report includes test data from base material removed from actual beltline plates and weld metal representative of that used for fabrication of the beltline welds.
As a result of our review of this data, we conclude that the limiting beltline material for Unit 1 is the circumferential weld seam and the limiting beltline materials for Units 2 and 3 are the longitudinal weld seams.
The circumferential weld seam in Unit 1 was fabricated by Babcock
& Wilcox using the automatic submerged arc process with Linde 80 flux and is identified as WF 154.
The upper bound unirradiated RT for this material is reported in BAW-10046 as +20'F.
Its chemical compo36ion is reported in BAW 1799 as
.31% Cu.
and 0.013%
P.
The longitudinal weld seams in Units 2 and 3 were fabricated by Babcock Wilcox using the electroslag process with Linde 124 Flux.
Samples from actual prolongation material used in fabricating these welds were not tested.
However, the licensee provided test results from four electroslag weld procedure qualifications, which Babcock
& Wilcox indicates are the only ones that would have been applicable for the three Browns Ferry Units.
In addition, eight weld wire chemistries were also available.
Based on these test results, the licensee estimated the electroslag chemical composition as
.25K Cu, and
.016%
P and the unirradiated RTNpT as O'.
We have reviewed the test data submitted by the licensee and additional test data from prolongation material removed from Peach Bottom Units 1 and 2 electroslag welds, which was also fabricated by Babcock Wilcox using Linde 124 flux.
This data is contained in FSAR Appendix F of Oresden Units 2 and 3.
As a result of this evaluation, we have concluded that the estimated chemical composition is acceptable, but the unirradiated RT, ~ should be 10'F.
This value of RTN corresponds to the highest NDT rePIIi"ted from the tests on the material Pom the weld procedure qualification.
Since there were only four weld qualification tests performed, a generic upper bound value should be used in estimating the unirradiated RTNDT for the electroslag welds.
The Unit 1 peak end-of-life neutron fluence at the 1/4 thickness location wa~8calculated from the Unit 1 surveillance capsule dosimetry as 1.0 X
10 n/cm'E less than 1MeV).
The dosimetry data and method used to calculate the peak end-of-life neutron fluence is documented in the Southwest Research Report No. 02-4884-001.
The circumferential weld is reported to be 28 inches below the core midplane, which corresponds to the peak neutron flux location.
Based on the vertical distribution of flux reported in the Southwest Research surveillance capsule report, the licensee indicates that the peak neutron fluence for this circumferential weld at the 1/4 thickness location is 1.887 X 10 n/cm'er EFPY.
The Units 2 and 3 peak end-of-life neutron fluence at the 1/4 thickness lor~tion was calculate) from their surveillance capsule dosimetry as 9.0 X
10 n/cm~
and 8.8 X 10 n/cm (E less then 1MeV), respectively.
The dosimetry data and method used to calculate the peak end-of-life neutron fluence is documented for Units 2 and 3 in Southwest Research Reports No.
02-4884-002 and No. 02-4884-003, respectively.
Since the longitudinal welds cross the core midplane, these peak neutron fluence values will conservatively represent the peak end-of-life values for the longitudinal weld seams for Units 2 and 3.
The conditions under which these pressure-temperature limits must be applied
- including operation with the curves in Figure 3.6 are specified in Section 3.6.A of the TS (Thermal and Pressurization Limitations).
We noted during review of these amendment applications that there was a phrase
4 missing from Section 3.6.A.2 of the Unit 3 TS compared to Units 1 and 2.
Specifically, this section in the Units 1 and 2 TS reads:
"During all operations with a critical core, other than for low level physics tests, except when the vessel is vented, the reactor vessel shell and fluid temperatures shall be at or above the temperature of curve 83 of figure 3.6-1".
The phrase that is missing in Section 3.6.A.2 of the Unit 3 TS is:
"except when the vessel is vented".
The phrase is properly included in Section 3.6.A.3 of the Unit 3 TS.
The NRC resident inspectors had also noted this in a recent inspection report.
This phrase was in the original Unit 3 TS.
Our review indicates the phrase was inadvertently dropped when page 185 was retyped to modify Section 3.6.A.3 in Amendment 56 issued July 22,
- 1982, and was not part of the changes intended to be authorized by that Amendment.
The licensee agreed that this error in omission should be corrected to clarify a possible violation and to have the TS for all there units the same (and correct).
By agreement with the licensee, we are adding the phrase "except when the vessel is vented" back to Section 3.6.A.2 of the Unit 3 TSs to correct this error.
Utilizing the method recommended in Regulatory Guide 1.99 Rev.
1 to predict neutron irradiation damage, the neutron fluence estimates calculated by Southwest Research from surveillance capsule dosimetry, the unirradiated RT T and chemical composition for the limiting beltline welds previously di(Pussed, and RT T for the closure flange region of 40 F-; we have determined that t)Vproposed pressure-temperature limits meet the safety margins of Appendix G, 10 CFR 50 for 12 EFPY, and may be incorporated into the plants'echnical Specifications.
3.0 ENVIRONMENTAL CONSIDERATION
S The amendments involve a change in the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20.
The 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.
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.
- 4. 0 CONCLUSION We have 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, and (2)- such activities will be conducted in compliance with the Commission's regulations, and the issuance of these amendments will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributors:
B. Elliot and R. Clark Dated:
September 16, 1985