ML20137W667
| ML20137W667 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 09/23/1985 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20137W651 | List: |
| References | |
| NUDOCS 8510040362 | |
| Download: ML20137W667 (4) | |
Text
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SAFETY EVALUATION BY TFE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 82 TO FACILITY OPERATING LICENSE NO. DPR-72 FLORIDA POWER CORPORATION, ET AL.
CRYSTAL RIVER UNIT NO. 3 NUCLEAR GENERATING PLANT COCKET NO. 50-302 INTRODUCTION In a letter frca G. R. Westafer to H. R. Denton dated February 14, 1985, the l
Florida Power Corporation (the licensee) reouested an amendment to the Technical Specifications for Crystal River Unit 3 (CR-3).
In part, the proposed amendment would update the reactor pressure-temperature limits to eight effective full power years (EFPY) based on the analyses of the first surveillance capsule.
The analyses of the first surveillance capsule are documented in Babcock & Wilcox Report RAW 1679, Rev. 1, entitled, " Analyses of Capsule CP-38 Florida Power Corporation, Crystal River Unit 3. Reactor Vessel Materials Surveillance Program." The portion of the licensee's request dealing with deletion of surveillance requirements for reactor vessel l
irradiation specimens from the Technical Specifications has been addressed by the Safety Evaluation previously issued in support of Amendment 80 to the CR-3 operating license.
DISCUSSION Pressure-temperature limits must be calculated in accordance with the recuirements of Aprendix G,10 CFR 50, which became effective on July 26, 1983.
Pressure-temperature limits that are calculated in accordance with the requirementsofAppendixG,10CFR50,aredependentupcntheinitialRT.;Ne for the limiting materials in the beltline and closure flange regions of' reactor vessel and the increase in RT resulting from neutron irradiation NOT damage to the beltline materials The CR-3 reactor vessel was procured to ASME Code requirements, which did not specify fracture toughness testing to determine the RT,dT for each reactor vessel material.
Hence, the initial RT for materia in the closure flange and beltline regien of the CR-3 hNctor vessel could not be determired in accordance with the test requirements of the ASPE Code.
Therefore, the initial RT.,
for these materials must be estimated from test data from other similar malNials used for fabrication of reactor vessels in the nuclear industry.
The limiting closure flange material is the C1csure Flange Head Plate, which was fabricated to ASPE Cede SA 533 GR.B requirements.
The RT for this material was estimated as 60'F; This value was estimated in b ordance with the criteria in B&W Topical Report BAW-1004EA, Dev.1, July 1977 This manas! OTy P
l l
. topical' report was approved by the NRC staff for referencing in licensing l
applications in a letter from S. A. Varga to J. H. Taylor dated June 22, 1977.
The licensee indicates that the limiting materials in the CR-3 reactor vessel i
beltline region are the weld metals. The unirradiated RT for the weld l
metals were estimated in accordance with the criteria in N Topical Report j
8AW-10046A, Rev. 1 July 1977.
The increase in RT resulting from neutron irradiation damage was estimated by the licensee usk the empirical relationship documented in Regulatory Guide 1.99 Rev.1, " Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials." This method of predicting neutron irradiation damage is dependent upon the predicted amount of neutron fluence and the amounts of copper and phosphorus in the beltline material.
The neutron flux was calculated using a discrete ordinate solution of the Boltzman Transport equation with the two dimensional code, DOT 3.5.
The calculated flux at the peak vessel flux location was normalized to the j
measured activities from the Capsule CR-38 dosimetry. There were six different detector reactions measured from the CR-38 dosimeters. The measured activities from these reactions were PO-30 percent less than the calculated activities. The corresponding normalization factor for the six detector reactions varied from.68 to.81.
Since the.81 normalizing factor would produce the least amount of neutron flux reduction, it was used for predicting neutron fluences.
Basedonanormalizingconstantof.81,ge vessgl'smaximumfluenceaftereightEFPYwaspredhtedtgbe1.9x10 n/cm (E > 1 PeV) at the 1/4 location and 4.4 x 10 n/cm (E>1MeV)atthe 3/4 T location.
The limiting materials in the CR-3 reactor vessel beltline are the weld metals. The estimated values for the amounts of copper and phosphorus in the beltline welds are documented in Table 6 of B&W Topical Report BAW-1511P, dated October 1980. The source of these values are chemical analyses from vessel weld dropouts and surveillance weld samples, which were fabricated using the same heats of weld wire as the CP-3 beltline welds. Since the amounts of copper and phosphorus in a weld are based upon the amounts of these elements in the weld wire, the use of chemical analyses from welds fabricated using the same heats of weld wire as the CR-3 beltline welds will provide acceptable values for the amounts of copper and phosphorus in the CP-3 beltline welds.
The CR-38 surveillance capsule contains plate and HAZ material from heat C-4344-1 and atypical weld metal designated as WF-209-1.
In Table 1 of this Safety Evaluation, we have compared the amount of increase in RT predicted using Pegulatory Guide 1.99, Rev.1 to the amount obsbed from the surveillance weld metal and plate material.
The test results from the HAZ material produced large data scatter.
Hence, we have not included the test results frcn this riterial in our evaluation.
Since the amount of increase in RT obsor b predicted usin Regulatory Guide 1.99, Rev. 1, is greater than that on the surveil ance material, the Regulatory Guide 1.99, Rev.1, l
method of predicting neutron irradiation will provide conservative estimates of the increase in RT resulting frca neutron irradiation damage.
NOT
. i EVALUATION We used the method of calculating pressure-temperature limits in USNRC Standard Review Plan 5.3.2, NUREG-0800, Rev. 1. July 1981, to evaluate the proposed pressure-temperature limits.
The amount of neutron irradiation damage to the beltline materials was estimated using the method recomended by the staff in Regulatory Guide 1.99, Rev.1.
The amounts of copper and phosphorus in the limiting CR-3 beltline weld were the values reported in Topical Report BAW-1511P. The neutron fluence used to predict neutron irradiation damage was based on a.81 normalizing factor. Our conclusion is that the proposed pressure-temperature limits meet the safety margins of Appendix G,10 CFR 50, for eight EFPY and are therefore acceptable and may be incorporated into the plant's Technical Specifications.
The licensee has also proposed to delete the criticality curve from Figure 3.4-2.
This is acceptable since the requirements of TS 3.1.1.4 continue to apply and will assure that criticality will not occur at unacceptable conditions of temperature and pressure.
According to B&W Topical Report 1543A Rev. 2, entitled " Integrated Reactor Vessel Material Surveillance Program," surveillance capsules CR-3C and CR-3D have been irradiated and removed from the CR-3 vessel. The dosimeter test results from thes'e capsules should be compared to the test results from capsule CR-3B.
If these test results indicate that the normalizing factor l
from the CR-38 capsule dosimeters are nonconservative, the licensee's pressure-temperature limits should be adfusted accordingly.
ENVIRONMENTAL CONSIDERATION This amendment involves a change in the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20.
We have 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 Comission has previously issued a proposed finding that this amendment involves no significant hazards consideration and there has been no public coment on such finding.
Accordingly, this amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Pursuant to 10 CFR 51.2<.(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of this amendment.
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 condu:ted in compliance with the Comission's regulations and the issuance of this amendment will not be ininical to the conron defense and security or to the health and safety of the public.
Dated: September 23, 1935 Principal Contributor:
B. Elliot
Table I l
1 Increase in RT for Capsule CR-3B Surveillance Material NDT l
Surveillance Material Neutron Fluence Increase in RTNDT (Predicted by F) l x 10ta n/cm2 Observed from (E) 1 MeV)
Capsule Test Data Reg. Guide 1.99, Rev. 1 Plate C4344-1
- 1. 0 21 57 Weld Metal 1.17 28 105 WF-209-1 1
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