ML19312C467
| ML19312C467 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 07/14/1977 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19312C459 | List: |
| References | |
| NUDOCS 7912130995 | |
| Download: ML19312C467 (12) | |
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f8 UNITED STATES NUCLEAR REGULATORY COMMISSION 3
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j WASHINGTON, D. C. 20666 0 %.'.... /
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 44T0 FACILITY LICENSE NO. DPR-38_
AMENDMENT N0. 44 TO FACILITY LICENSE NO. DPR-47 AMENDMENT NO. 41 TO FACILITY LICENSE NO. OPR-55 00KE POWER COMPANY OCONEE NUCLEAR STATION, UNIT NOS. 1, 2 AND 3_
00CKETS NOS. 50-269, 50-270, AND 50-287 Introduction By letter dated March 10, 1977, Duke Power Company (licensee) requested that (1) an exemption to Appendix H of 10 CFR 50 be granted for the Oconee Units 1, 2 and 3 which would allow indefinite operation of the Oconee Units with the remainder of the reactor vessel surveillance capsules to be irradiated at Crystal River, Unit 3 rather than at Oconee Station, and' (2) the Oconee Nuclear Station Technical Specifications be revised to allow the Oconee 1, 2 and 3 reactor vessel surveillance capsules to be irradiated at Crystal River, Unit No. 3.
Several modifications to the licensees pro-posal were necessary. These changes were discussed with and agreed to by the licensee.
In addition, information submitted by letter dated January 4,1977 was considered in this evaluation.
The basis for this request is that the surveillance specimen holder tubes in Oconee were damaged and could not be repaired without a complex and expensive repair program and radiation exposure to personnel.
In lieu of putting the surveillance capsules back into the Oconee reactor pressure vessels, they will be placed in a host reactor, Crystal River Unit No. 3, for irradiation.
In addition, data from specimens from other irradiation programs will provide input to the Oconee irradiation program.
This overall program is called an integrated surveillance procram in which all presently operating facilities with B&W 177 fuel assembly reactors will participate.
Discussion & Evaluation The original Oconee design includad three reactor vessel surveillance specircen holder tubes (SSHTs) loc..ed near the reactor inside vessel wall.
Each of these SSHTs housed two capsules containing reactor vessel sur-veillance specimens. When failures of the SSHTs occurred at other Babcock & Wilcox (S&W) designed plants, the licensee shut down the three Oconee facili-ies in succession, starting in March 1976 to inspect the SSHTs. The inspection revealed that all of the SSHTs had suffered some y 9115130%f
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damage.
To prevent further damage all surveillance capsules and all parts of the SSHTs that had failed or were deemed likely to faH during the remainder of that operating cycle were removed from the vessels.
Since the discovery of the damage to the SSHTs, B&W has undertaken the design, manufacture and testing of an improved SSHT.
SSHTs of this im-proved design presently installed in Davis-Besse 1, Crystal River -3 and Three Mile Island-2.
All three of these plants have reactors supplied by B&W and all are or will start initial operation within the next few months.
All of these plants have the same basic B&W 177 fuel assembly reactor design as Oconee 1, 2 and 3.
The acceptacility of the redesigned SSHTs has been demonstrated by a test program reviewed and approved by the NRC staff and conducted in conjunction with the hot functional test performed at Davis-Besse 1.
Installation of the redesigned SSHTs in the Davis Besse 1, Crystal River-3 and TMI-2 reactor vessels did not present any unusual radiological difficulties because installation was prior to neutron activation of the reactor internals.
Studies of methods to install the redesigned SSHTs in the irradiated B&W reactors indicate that substantial installation difficul-ties will be experienced--primarily because precision machining, alignment and inspection must be performed remotely and under water.
Although such problems do not in themselves justify relief from a requirement to re-install the SSHTs in Oconee 1, 2 and 3, they would be likely to cause significant radiation to personnel.
Based on its experience in removing the SSHTs at Three Mile Island-1 and Rancho Seco-1, B&W estimated that installing SSHTs in irradiated reactors could result in personnel exposures totaling about 100 man-rem per reactor.
In the interest of maintaining the radiation exposure of plant personnel as low as reasonably achievable, the licensee, in cooperation with B&W and the owners of other B&W 177 fuel assembly plants, has proposed an alternative program that does not require reinstalling the SSHTs in Oconee 1, 2 and 3 and the other irradiated B&W plants.
The capsules removed from the Oconee vessels which had damaged SSHTs will be placed in a host reactor, Crystal River Unit 3, as part.of the integrated surveillance program discussed herein.
These capsules contain samples of plate or forging material and heat-affected zone material from the vessel beltline as well as weld metal.
The weld metal is expected to be controlling because it is more radiation sensitive. However, capasules containing other than weld metal will be irradiated also, since the purpose
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. of the surveillance program is to obtain data on materials which would prove to be important later on.
This program includes provisions to provide additional information, if required under Appendix G 10 CFR 50 Paragraph V.C., in addition to the normal requirements of Appendix H.
The proposed plan involves integrating the interrupted surveillance pro-gram at Oconee and other plants with the programs for new plants in a manner generally similar to that covered in Appendix H,10 CFR 50, paragraph II.C.4, except that the plants are at different sites.
There are three distinct featu*es of this plan.
1.
The original surveillance materials from one or more reactors that hdve been in service will now be irradiated in a new host reactor, that can be fitted with the newly-designed capsule holders on the thermal shield in less time and without significant radiation exposure of the workmen, and 2.
There will be more vald metal specimens and some larger fracture mechanics (compact tensior, or CT) specimens placed in the capsules, and 3.
A data-sharing feature in which all available irradiation data for the beltline welds af a given reactor some of which will come from other surveillance programs, will be considered in predicting its adjusted reference temperature and in making any fr7cture analyses for that reactor. Typically, several of the welds in any one vessel were made with the same weld wire and flux as those used on some other reactors. The data sharing feature is required because the welds in these reactors have high radiation sensitivity due to high copper content, large and random variation of copper trom point to point in the weld, and low initial upper shelf energy.
The specific program proposed for Oconee 1, 2:and 3 involves installing the Oconee surveillance capsules in extra locations provided in the Crystal River 3 vessel.
This plan will accomplish the original purpose of obtaining information on the effectof radiation on material that is representative of the material in the Oconee reactor vessels on a schedule that provides an appropriate lead time over the vessel irradiation rate.
The overall integrated program also will provide information relevant to Oconee 1, 2 and 3 from surveillance programs in Crystal River 3, Three Mile Island 2, and Davis Besse 1 on material considered to be essentially identical to the actual welds in the Oconee vessels.
It is also important to note that still more information relevant to the Oconee vessel materials will be obtained from the NRC funded HSST irradiation programs underway.
Details are proviced below.
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Oconee 1 There are two weld materials of primary interest for the Oconee 1 vessel, Procedure Qualification (P.Q.) numbers
- SA-1229 and SA-1585.
These are used in the I.D. portion of the upper circumferential weld, and in the forbothoftheseweldsisestimatedtobe1.2x10gglife(EOL) fluence center circumferential weld respectively. The end nyt, and both have compositions that are expected to make them relatively sensitive to radiatirn damage. Weld P. Q. No. WF-25, used for the 0.D. portion of the upp 3r circumferential weld is radiation sensitive, but fluence at theaz.myghallocationsoftheselongitudinalweldsislower(0.7to 1.0 x 11 ) so they will not become limiting during the service life.
The lowtr circumferential weld will not be limiting, because the fluence at this location is estimated to be at least an order of magnitude lower than th t of the other circumferential welds.
The ori linal surveillance material, WF-ll2, was made using the same heat of filler wire but a different batch of flux as WF-154, one of the control ing welds in Oconee 2.
Metallurgical considerations suggest that the radiction behavior is affected more by the wire than the flux, thus WF-ll2 is expected to respond to radiation much like WF-154.
This data will be a useful part of the data base for B&W vessels.
Table 1 shows where samples of the pertinent weld materials will be irradiated in the proposed integrated program, what kinds of specimens will be used, and when information will be available under the present plan.
The irradiation schedule and withdrawal dates shown will be modified to optimize the information obtained as indicated to be appropriate as initial test results are obtained and evaluated.
Oconee 2 There are only two weld materials of interest for the Oconee 2 vessel, P.Q. numbers WF-25 and WF-154.
They are used in the center and upper circumferential welds.
Fluence is expected to be the same at the two welds, but the higher copper content of WF-25 means that it is predicted to be more radiation sensitive.
Fluence at the lower circumferential weld, W' 112, is too low for it to be of concern.
There are no longitudinal welds *, ti;is vessel beltline.
The original surveillance material, WF-209-1, while not identical to any of the beltline welds in B&W reactors, is predicted to be radiation sen-sitive, based on ts copper and phosphorus contents.
Data from WF-209-1
- Weld materials are specifically identifica by the ASME Code by the procedure Qualification Test nvaber, A prodedure qualification test is required on each combination r/ heat of weld wire and batch of flux,
.s 5-will be a useful addition to the data base for these reactors.
Table 2 shows where samples of the pertinent weld materials will be irradiated in the proposed integrated program, what kinds of specimens will be used, and when information will be available under the present plan.
Oconne 3 There are three weld materials of primary interest for the Oconee 3 vessel, numbers WF-67 and WF-70 in the center circumferential weld, and WF-200 in the upper circumferential weld. The end of life (EOL) fluence for both of these welds is estimated to be 1.2 x 1019 nyt, and the three weld materials have comcositions that are expected to make them relatively sensitive to radiation damage.
Another shell weld, the lower circumferential, is made of a material that is expected to be radiation sensitive (P.O. No. WF-169), but the EOL fluence at this location is estimated to be at least an order of magnitude lower than that of the other circumferential welds, so it will r.ever be limiting.
- The original surveillance material, WF 209-1, is the same as that used in Oconee 2.
The weld wire used for WF 209-1 was also used for WF-70, one of the controlling welds in Oconee 3.
Table 3 shows where samples of the pertinent weld materials will be irradiated in the proposed integrated program, what kinds of specimens will be used, and when information will be available under the present plan.
Table 4.2-1 of the proposed Technical Specifications specifies the Oconee specimens capsules that are to be irradiated in Crystal River Unit No. 3.
These capsules include the weld material shown in Table 2 herein and other materials such as plate or forging material samples and weld heat affected zone material samples from the Oconee vessels which are not now considered to be controlling material but could become so sometime in the future.
For those welds where no surveillance specimens exist, we will be guided by predictions based on the known chemical composition of those welds.
To be conservative, the predictions will be based on the Upper Limit lines of Regulatory Guide 1.99, Revision 1.
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. In addition to this integrated program, "research" capsules containing tensile, Cv, and severt' sizes of Ct specimens of B&W archive material will be included in the overall B&W power reactor surveillance program.
These are shown as " capsule R-1 and R-2" in Tables 1, 2 and 3.
For example, samples of the weld most likely to be limiting in Oconee 1, SA-1585, will be irrdiated in Crystal River 3, and samples of a weld made of the same heat of weld wire as WF-154 will be irradiated in the Davis Besse program.
Details of withdrawal schedules will be reviewed again later, and will depend on test results from the other programs.
Research programs being funded by the NRC will also provide continual information on the effect of radiation on these specific weld materials and on several additional B&W weld materials expected to respond to radiation in a similar manner.
These programs, HSST-2 and HSST-3, consist of many tensile, Cv and CT specimens irradiated in a test reactor.
Although information on shift in RTNDT will be obtained, the main emphasis of the HSST programs is to develop methods that can be used to better evaluate low shelf toughness using the rather small specimens used in the power reactor programs.
We have evaluated the effectiveness o1.his overall program plan, and have concluded that the information to be developed that is directly and in-directly relevant to the Oconee raactor vessels will be sufficient to provide assurance of safety margins against vessel failure that comply with Appendix G,10 CFR 50.
Further, it is our opinion that even without additional irradiation surveillance programs in Oconee vessels, the proposed program will provide more useful information than would have been obtained from the original surveillance program.
Until data become available from the surveillance program, a conservative prediction of radiation damage can be made by using R.G.1.99* for at least the next 5 yea s of operation.
This Regulatory Guide is based on the staff's analysis of all data available at the time the Guide was written.
New data, in particular the results of the augmented integrated surveillance program described above, will be used to update the Guide periodically.
Predictions of the adjustment of reference temperature and the drop in upper shelf energy are given graphically as functions of copper and phosphorus content and of fluence.
Ir. addition, there is an " Upper Limit" line on each graph, which is to be used when information about the copper and phosphorus contents is inadequate.
Because the chemical analyses of the B&W welds have shown considerable variation, we intend to use the Upper Limit lines as the basis for any prediction required at this time.
- Regulatory Guide 1.99, Revision 1, " Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials", April 1977.
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TABLE 1 RADIATION DATA FOR OCONEE - 1 REACTOR Host Removal Specimen
- Weld Capsul e
- Reactor Date Types WF-25 TMI-lE TMI-l 1976 Cv, Tensiles (UpperCir-TMI-1A TMI-2 1982 Cv, Tensiles cumferential, TMI-1C TMI-2 1990 Cv, Tensiles 39%, on 00)
R-1 TMI-2 1982 Cv, CT
,s-R-2 TMI-2 1989 Cv, CT HSST-2 Test Reactor 1977-78 Cv, CT to HSST-3 Test Reactor 1978 4.0T NRL Test Reactor 1977-78 Cv.CT NRL Test Reactor 1977-78 Cv.CT SA-1229 none (Upper Circum-ferential, 61% on ID)
SA-1585 R-l CR-3 1982 Cv, CT (Center R-2 CR-3 1989 Cv, CT Circumferen-HSST-3 Test Reactor 1978 Cv, CT to tial) 4.0T SA-1073 none
.(Longitudinal, in upper shell course)
SA-1493 none (Longitudinal, in middle shell course)
SA-1430 none (Longitudinal, in lower shell course)
SA-1135 (TopCircumfer-ential Weld)
WF-9 (Lower Circumferential Weld)
- TMI-IE - means capsule E in the Three Mile Island Unit
.No. 1 Re, actor 1
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TABLE 2 RADIATION DATA FOR OCONEE -2 REACTOR Host Weld Capsule
- Reactor Removal Specimen **
Date Types WF-25 TMI-lE TMI-1 1976 Cv, tensile TMI-1A TMI-2 1982 Cv, tensiles (Center Circum-TMI-lc THI-2 1990 Cv, tensile ferential Welc ) R-1 TMI-2 1982 Cv, CT R-2 TMI-2 1989 Cv, CT HSST-2 Test Reactor 1977-78 Cv, CT to HSST-3 Test Reactor 1978 4.0T i
NRL Test. Reactor 1977-78 CV CT NRL Test Reacto-1977-78 Cv, CT WF-154 none, but tas same weld wire as WF-ll2.
(Upper Circumferen-tial Weld)
WF-112 OCI-E Oconee 1 1976 Cv, tensile (Lower OCI-A Crystal R.2 1985 Cv, tensile Circumfer-OCI-C Crystal R.3 1989 Cv, tensile ential Weld)
R2 Davis-Besse 1989 Cv, CT
- 0CI-E - means capsule E from the Oconee Unit No. I reactor
- Cy - means Charpy V-notch specimen CT - means Compact Tension specimen
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TABLE 3 RADIATION DATA FOR OCONEE - 3 REACTOR Host Weld Capsule
- Reactor Removal Specimen **
Date Types WF-67 R-1 Davis Besse -1 1981 Cv.CT R-2 Davis Besse-1 1989 Cv.CT (Center R-1 Crystal River 3 1982 Cv,CT fr" R-2 Crystal River 3 1989 Cv,CT NRL Test Reactor 1977-78 Cv, CT a
e 75%, I.D.
WF-70 R-1 Davis-Besse 1 1981 Cv,CT (Center R-2 Davis-Besse 1 1989 Cv CT Circumferen-R-1 Crystal River 3 1982 Cv,CT tial Weld)
R-2 Crystal River 3 1989 Cv CT 25%: 0.D.
HSST-3 Test Reactor 1978 Cv,CT to 1.6T WF-200 none (Upper Circumferen-tial Weld) l WF-169 none (Lower Circumferen-tial Weld)
- R-1, R denotes "research" capsule
- See Table 2
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We have also considered the uncertainties involved in applying radiation effects information obtained in other reactors to the Oconee vessels.
The major uncertainties involved are:
1.
Accuracy of neutron fluence calculations 2.
Magnitude and effect of variation in neutron spectra between reactors 3.
Magnitude and effect of variations in irradiation temperature between reactors.
4 Magnitude and.effect of variations in rate of irradiation on material properties.
The effects of these variables have been studied for at least 20 years.
Although some uncertainties still remain, the effects are fairly well established and understood as discussed below.
1.
Neutron flux calculations for the reactor vessel wall and irradiation capsule locations have been developed over many years.
The dosimetry used in irradiation capsules has furnished information that was used to check out and refine the calculational methods.
It is generally believed that the fast neutron flux and fluence in these locations can be calculated to an accuracy of + 20%, particularly if sore dosimetry checks are available.
DosTheters from the original Oconee surveillance program were removed and tested, so the fluence calcu-lations for the vessel can be verified.
In this connection, it should be emphasized that the effect of neutron radiation on reactor vessel steel varies as the square root of the fluence, so uncertainties of 20 to 50% in fluence are not highly signi ficant.
We have also considered the fact that the design of the Oconee vessels, internals and cores is almost identical to that of the other reactors that will be used to obtain radiatio; effects information.
These considerations are the basis for our conclusion that un-certainties in the calculation of neutron fluence will be small, and the effect of such uncertainties on the assessment of the radiation effects on the vessel material will also be small.
2.
Although differences in neutron energy spectra can cause uncertainties in the effects of radiation on material when this is evaluated without considering spectrum effects, only very large differences in spectra are significant.
The variations from one B&W 177 fuel assembly reactor to another are claimed to be relatively minor, because they have almost identical geometry.
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l We considered the possible differences in neutron spectra that could occur between the BW power reactors 50 be involved in the integrated program. Such effects can be dealt with, if necessary, through the use of neutron damage functions that are being developed for that purpose. However, the worst expected differences are judged inconsequential based on present knowledge of irradiation effects.
If additional developments (theoretical or experimental) suggest that the neutron spectra effects might be significant under some conditions, appropriate actions can be taken.
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3.
The effect of the terrperature of irradiation has also been the subject i
of considerable research.
It is well known that radiation damage is less severe at 6000F than at 5000F (the temperature range of concern).
The differences in effect on the steel appear to be noticeable and should be taken into account if the irradiation termerature difference is over about 250F. Enough infonnation is known to pennit conservative evaluations of the effect of temperature differences of at least 500F, and probably even 1000F or more.
The differences in the temperature of the surveillance capsules and vessel walls between the BW power reactors involved in the proposed integrated program are expected to be less than 500F, and can be conservatively evaluated.
4.
The effect of irradiation has also been evaluated by research programs at NRL and other laboratories. Although the consensus of experts on this subject is that there will be no major differences in material property changes by irradiation rates varying over 2 to 3 orders of magnitude, more data from surveillance programs are needed to provide veri fication. However, the differences in the rates of irradiation of specimens in the integrated program and the limiting material in the walls of the affected vessels will be less than one order of magnitude, therefore, we have concluded that there will be no significant uncertainties in this program associated with differences in rate of irradiation.
We have evaluated the adequacy of the proposed integrated, augmented reactor vessel material irradiation program for Oconee 1, 2 and 3 as an alternative to the original program that was interrupted by failure of the assocMed hardware. We conclude that the proposed alternative program will provide the information required to comply with Appendix G, 10 CFR 50, and that the uncertainties involved in using data obtained fron surveillance specimens irradiated in various other BW power reactors to e:,tablish Oconee 1, 2 and 3 vessel operating limitations are small and can be accounted for by imposition of appropriate margins.
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Additionally, the proposed integrated, augmented program (with possible minor modification yet to be finalized) should provide more useful information than could have been extracted from the original surveillance program. The proposed program will also give results of the kind required to meet Paragraph V.C of Appendix G,10 CFR 50.
Until the results of the proposed surveillance program become available, our predictions of radiation damage in the Oconee vessels will be based on the current revie ion of Regulatory Guide 1.99.
At present, this is Revision 1.
Because the chemical analyses of the B&W welds have shown considerable variation, we intend to use the Upper Limit lines as the basis for any prediction required at this time.
Environmental Consideration We have determined that these amendments do not authorize a chance in effluent types or total amounts nor an increase in power level and will not result in any significant environmental impact.
Having made this determination, we have further concluded that these amendments involve an action which is insignificant from the standpoint of environmental impact and pursuant to 10 CFR 551.5(d)(4) that an environmental impact statement, negative declaration or environmental impagt appraisal need not be prepared in connection with the issuance of these amendments.
Conclusion We have concluded, based on the considerations discussed above, that:
(1) because the amendments do not involve a significant increase in the i
probability or consequences of accidents previously considered and do not involve a significant decrease in a safety margin, the amendments do not involve a significant hazards consideration, (2) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, a, d (3) 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.
Date: July 14, 1977 x..
UNITED STATES NUCLEAR REGULATORY COMMISSION DOCKET NOS. 50-269, 50-270 AND 50-287 DUKE POWER COMPANY NOTICE OF ISSUANCE OF AMENDMENTS TO FACILITY OPERATING LICENSES The U. S. Nuclear Regulatory Commission (the Commission) has issued Amendment Nos. 44, 44, and 41 to Facility Operating License Nos. DPR-38, CFR-47, and DPR-55, respectively, issued to Duke Power Company which revised the Technical Specifications for operation of the Oconee Nuclear Station, Unit Nos.1, 2 and 3, located in Oconee County, South Carolina.
The amendments are effective as of the date of issuance.
The amendments authorized changes in the Technical Specifications to pennit irradiation of the remaining Oconee Nuclear Station, Units Nos.
1, 2 and 3 reactor vessel surveillance specimens at Crystal River Unit 3.
An exemption to that provision of Appendix H to Title 10 of the Code of Federal Regulations Part 50, which would have otherwise required irradiation of the Oconee capsules in the Oconee vessels, has also been granted.
Such action was in response to a generic failure of first-generation design Surveillance Specimen Holder Tubes (SSHTs) at Oconee and other operating Babcock & Wilcox 177 fuel assembly reactors.
The application for the amendments complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act),
and the Comission's rules and regulations.
The Comission has made DUPLICATE DOCUMENT Entire document previously entered into system under:
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