ML19319D877
| ML19319D877 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 10/20/1977 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19319D872 | List: |
| References | |
| NUDOCS 8003270534 | |
| Download: ML19319D877 (15) | |
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.e UNITE D STATES NUCLEAR REGULATORY COMMIS$10N WASHINGTON, D. C. 20586 SAFETY EVALUATION BY THE OFFICE NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT N0.15 TO FACILITY OPERATING LICENSE NO. DPR-54 SACRAMENTO fiUNICIPAL UTILITY DISTRICT RANCHO SECO NUCLEAR GENERATING STATION DOCKET NO. 50-312 Introduction By letter dated July 9,1976, as supplemented by letters dated August 24, 1976, and December 17, 1976, the Sacramento Municipal Utility District (SMUD) requested + hat (1) the exemptien to Appendix H of Title 10 of the Code of Federal Regulations Part 50 (10 CFR 50) granted for Rancho Seco Nuclear Generating Station (Rancho Seco) by letter of August 13, 1976, be modified to allow indefinite operation l
of Rancho Seco with the remainder of the reactor vessel surveillance capsules to be irradiated at Davis-Besse Unit No. 1 (Davis-Besse 1) rather than in-situ, and.(2) the Rancho Seco Technical Specifications be revised to allow the remainder of Rancho Seco reactor vessel surveillance capsules to be irradiated at Davis-Besse 1.
Discussicn and Evaluation Tht original Rancho Secc design included three reactor vessel surveillance specimen holder tubes (SSHTs) located near the reactor inside vessel
- wall.
Each of these SSHTs housed two capsules containing reactor vessel surveillance specimens.
When failures of the SSHTs occurred at other 8oosan 5 3
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. Babcock & Wilcox'(B&W) designed plants, SMUD shut down Rancho Seco on April 3, 1976, to inspect the SSHTs. The inspection revealed that the SSHTs had suffered some damage.
To prevent further damage, all surveillance capsules and all parts of the SSHTs that had failed or were deemed likely to fail during the remainder of that operating cycle (Cycle 1) were removed from the vessel.
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 improved design are presently installed in Davis-Besce-1, Crystal River-3 and Three Mile Island-2.
All three of these plants have reactors supplied by B&W and all are in the process of beginning initial operation at the present time or within the next few months.
In addition, all of these reactors have the same basic B&W 177 fuel assembly design as Ranc a Seco.
The acceptability of the redesigned SSHTs has been demonstrated by a _ test program reviewed and approved by the NRC staff atJ 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 Three Mile Island-2 reactor vessels did not present any unusual difficulties because it was perfarmed prior to neutron activation of the' reactor internals.
Studit, of methods to install the redesigned SSHTs ir, the irradiated B&W reactors indicate that substantial difficulties
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.- will be experienced primarily because precision machining, alignment and inspection must be performed remotely and under water.
Although such problems do not themselves justify relief from a requir6 ment to reinstall the SSHTs in Rancho Seco, they would cause significant radiation to personnel.
Based on their experience ii. removing the SSHTs at Three Mile Island-l and Rancho Seco, B&W estimated that installing SSHTs in irradiated reactors would result in personnel exposure of about 100 man rem per reactor.
In the interest of maintaining the radiation exposure of plant personnel as low as reacanably achievable, SMUD, 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 Rancho Seco and the other irradiated B&W piants.
This program is very complex, as it 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 programs into 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 features of these proposed programs:
1.
A host-reactor feature, in which the original surveillance materials from one or more reactors that have been in service will now be
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_4 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 radiation exposure of the workmen, and 2.
An augmented surveillance feature in which there will be more weld metal specimens and some larger fracture me nics (compact tension or CT) specimens placed in the capsules, and 3.
A data-sharing feature in which all available irradiation data fvP all of the beltline welds of a given reactor will be considered in predicting its adjusted reference temperature and in making any fracture 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 from point to point in the weld, and low initial upper shelf energy.
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The specific program proposed for Rancho Seco involves installing the original Rancho Seco surveillance capsules in extra locations provided in the Davis Besse-1 vessel.
This plan will accomplish the original purpose of obtaining information on the effect of radiation on material
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. that is representative of (although not identical to) the material in the Rancho Seco reactor vessel on a schedule that provides an appropriate lead time over the vessel irradiation rate.
The overall integrated program also will provide information 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 Rancho Seco vessel.
It is also important to note that still more information relevant to the Rancho Seco vessel materials will be obtained from the NRC funded Heavy Section Steel Technoloay (HHST) irratiation programs underway.
Details are prc. ' -
m.
There are four weld metals of interest for the Rancho Seco
- vessel, Procedure Qualification (P.Q.) numbers
- WF-29, WF-70, WF-154 and WF-233.
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All are predicted to have the same end-of-life fluence,.l.2 x 10 n/cm (EslMeV).
Weld WF 70, which was used for making the ID portion of one of the longitudinal welds, has the highest reported copper content.
As shown in Table I there will be radiation data for WF 70 from a test reactor and f rom two research capsules each in Davis Besse and Crystal River-3.
Weld WF 154, which was used for the center girth weld, is fairly well represented by the surveillance material, WF 193, because they were made using the same heat of filler wire, but using a different batch of flux.
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AWeld materials are.specifically identified by the ASME Code by the Procedure Qualification test number.
A procedure qualification test is required on each combination of' heat of weld wire and batch of flux.
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. Metallurgical considerations suggest that the radiation behavior is affected more by the wire than the flux, thus the surveillance weld is' expected to respond to radiation most like the center girth weld, WF 154.
As shown in Table 1, weld WF 193 was also the surveillance weld for Arkansas-1.
Capsule AN1-E has already been tested.
Capsule RSI-A, which was withdrawn from the Rancho Seco reactor when the problem with the surveillance holder tubes was discovered, will be tested shortly.
It had a fluence of about 3 x 10 n/cm2 (E>lMeV).
All six of the 17 original surveillance capsules from Rancho Seco had samples of weld WF 193, as detailed in Table I.
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TABLE 1 RADIATION DATA FOR RANCHO SECO REACTOR VESSEL Weld Capsule Reactor Removal Specimen Date*
Types WF-70 R-1 Davis Besse-1 1981 Cv,CT R-2 Davis Besse-1 1989 Cv,CT R-1 Crystal River-3 1982 Cv,CT R-2 Crystal River-3 1989.
Cv,CT H55T-3 Test Reactor 1978 Cv, CT to 1.6T WF-154 none, but has same wire as WF-ll2 and WF 193 (Upper Circumferen-tial Weld)
WF-193 AN1-E Arkansas-1 1977 Cv (tested)
AN1-A Davis Besse-1 1983 Cv AN1-C Davis Besse-1 1987 Cv (removed)
RSI-A Rancho Seco 1977 Cv (removed)
R5l B Davis Besse-1 1980 Cv,1/2 TCT R51-C Davis Besse-1 Standby Cv R5l-D Davis Besse-1 1982 Cv, 1.2 TCT RS1-E Davis Besse-1 Standby Cv R5l-F Davis Besse-1 1989-Cv, 1/2 TCT WF-29 none i
WF-233 none 3
- 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.
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In addition to this integrated program, "research" capsules containing tensile, Charp V-notch (Cv), and several sizes of CT specimens of BfN archive material will be included in the overall BAW power reactor surveillance program.
Samples of the weld most likely to be limiting in Rancho Seco, WR 70, will be irradiated in Davis-Besse, and Crystal River-3.
Samples of weld UF 112, which was made of the same heat of weld wire as WF 154 will be irradiated in Crystal River-3 as the surveillance weld for Oconee-l.
Details of withdrawal schedules will be determined later, and will depend on test results from the other procrems.
The dates given in Table 1 are tentative.
i 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 manrsr.
These programs, H55T-2 and H55T-3, consist of many tensile, Cv, and CT specimens irradiated in a test reactor.
Although will be obtained, the main emphasis information on the increase in RTNJT of the HSST programs is to develop methods that can be used to better evaluate low upper shelf toughness using the rather small specimens used in the power reactor programs.
We have evaluated the effectiveness of this overali program clan, and nava concluded that the information to ba developed that is directly
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and indirectly relevant to the Rancho Seco reactor vessel will be sufficient to provide assurance of safety margins against vessel failure that comply with Appendix G, 10 CFR 50.
Further, it is our epinion that even without additional irradiation surveillance I
programs in the Rancho Seco vessel, the proposed program will provide more useful information than would have been obtained from the original surveillance program.
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Until data become available from the surveillance program, a conservative prediction of radiat'on damage can be made by using R.G. 1.99*.
This Guide is based on the NRC staff's analysis of all data available at the time it was written.
New data, in particular the results of the augmented integrated surveillance program described above, will be used to periodically update the Guide.
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.
In addition there is an "L;;er Limit" line on each graph, which is to be used when information about the copper and phosphorus contents is ir. adequate.
Because th 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, "Ef fects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials", Revision 1, April 1977.
. We also have considered the uncertainties involved in applying radiation effects information obtained in other reactors to the Rancho Seco vessel.
The major uncertainties involved are:
1.
Accuracy of neutron fluence calculations; Magnitude and effect of variation in neutron spectra between reactors; l
2.
Magnitude and effect of variations in irradiation temperature 3.
between reactors; Magnitude and effect of variations in rate of irradiation on material 4.
properties; The effects of these variables have been studied for at least 20 years.
Although some uncertainties still remain, the effects are f airly 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 'sst neutron flux and fluence in these locations can be calculated to an accuracy of + 20%,
particularly if some dosimetry checks are available.
Dosimeters
. from the original Rancho Seco surveillance program were removed and testeo, so the fluence calculations 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 a.e not highly significant.
We have also considered the fact that the desian of the 4
Rancho Seco vessel, internals, and core is almost identical to that of the other reactors that will be used to obtain radiation effects information.
These considerations are the basis for our conclusion that uncertainties in the calculation of neutron fluence will be small, and the effect of such uncertainties on the assessment of the radiation effects on the vessei material will also be small.
Although differences in neutron energy spectra can cause uncertainties 2.
in the effects of radiation on material when this is evaluated without considering spectrum effects, only very large d.fferences in spectra are significant.
T'ne variations from one B&W reactor to another are claimed to be relatively minor, because they have similar geometry.
. We considered the possible differences in neutron spectra that could occur between the B&W power reactors involved in the integrated Such effects can be dealt with, if necessary, through the program.
use of neutron damage functions that are being developed for that However, the worst expected differences are judged incon-purpose.
sequential based on present knowledge of irradiation effects.
If additional developments (theoretical or experimental) suggest that the neutron spectra effects might be significant under scme conditions, appropriate actions will be taken.
3.
The effect of the temperature of irradiation has also been the subject of considerable research.
It is well known that radiation damage is less severe at 600 F than at 500 F (the temperature range of concern).
The differences in effect on the steel appear to be noticeable and shculd be taken into account if the irradiation temperature difference is over about 25 F.
Enough information is known to permit conservative evaluations of the effect of temperature differences of at least 50 F, and probably even 100 F cr more.
The differences in the temperature of the surveillance capsules and vessel walls between the B&W power reactors involved in the integrated program are expected to be less than 50*F, and can be conservatively evaluated.
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. 4.
The effect of irradiation has also been evaluated by research programs at the Naval Research Laboratory (NRL) and other laboratories. Although the consensus of experts on this subject is that there will be no major differences in materie1 property changes by irradiation rates varying over 2 to 3 orders of magnitude, more data from surveillance programs are needed to provide verification.
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 surveillance program for Rancho Seco as an alternative to the original program that was interrupted by failure of the associated hardware.
We conclude that the proposed program will provide information required to provide safety margins that comply with Appendix G, 10 CFR 50, and that the uncertainties involved in various predictions using data obtained from surveillance specimens irradiated in various other B&W power reactors to establish Rancho Seco vessel operating limitations are small and can be accounted for both easily and conservatively.
. Of equal, if not greater, importance is our assessment of the proposed integrated, augmented program (with possible minor modification yet to be finalized), to wit:
It will provide more useful information than The could have been extracted from the original surveillance program.
proposed program will also give results of the kind required to meet Paragraph V.C of Appendix G,10 CFR 50.
t Until the results of the proposed surveillance program become available, our predictions of radiation damage in the B&W power reactors will be base.d on the current revision of Regulatory Guide 1.99.
At present, t
this is Revision 1.
Environmental Considerations We have determined that the amendment does not authorize a change 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 the amendment involves an action which is insignificant from the standpoint of environmental impact and pursuant to 10 CFR 551.5(d)(4), that an environmental j
impact statement or negative declaration and environmental impact appraisal need not 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 conducted in compliance with the Commission's regulations and the issuance of this amendment will not be inimical to the connon defense and security or to the health and safety of the public.
We have osso concluded that the exemption is authorized by law and vill not endanger life or property or the connon defense or security and is otherwise in the public interest.
Dated:
October 20, 1977 O