ML19326C005
| ML19326C005 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 04/01/1977 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19326B999 | List: |
| References | |
| NUDOCS 8004180697 | |
| Download: ML19326C005 (9) | |
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/pa atog*g UNITED STATES NUCLEAR REGULATORY COMMisslON y'
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WASHINGTON. D. C. 20555
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%...e SAFETY EVAL'jATION BY THE OFFICE OF NL2 LEAR REACTOR REGULATION SUPPORTING AMENDMENT N0.
TO FACILITY OPERATING LICENSE NO. DPR-51 ARKANSAS POWER & LIGHT COMPANY ARKANSAS NUCLEAR ONE - UNIT 1 DOCKET NO. 50-313 INTRODUCTION By letter dated August 17, 1976, as supplemented by letters dated December 20 and 22,1976, and January 13. 1977, Arkansas Power & Light Company (AP&L) requested that (1) the exemption to Appendix H of Title 10 of the Code of Federal Regulations Part 50 (10 CFR 50) granted for Arkansas Nuclear One - Unit No.1 (ANO 1) by letter of June 11, 1976, be modified to allow indefinite operation of ANO-1 with the remainder of the reactor vessel surveillance capsules to be irradiated at Davis-Besse Unit No.1 rather than in-situ, and (2) the ANO-1 Technical Specifications be revised to albw the remainder of ANO-1 reactor vessel surveillance capsules to be irradiated at Davis-Besse Unit No.1.
We have determined that several modifications to the AP&L proposal were necessary to meet regulatory requirements. These modifications have been discussed with and agreed to by the AP&L staff.
DISCUSSION AND EVALUATION The original ANO-1 design it.cluded three reactor vessel surveillance specin,en 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 Babcock & Wilcox (B&W) designed plants, the licensee shut down the ANO-1 facility on March 19, 1976, to inspect the SSHTs. The inspection -
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revealed that all of the SSHTs had suffered severe damage and that portions of two SSHTs had fallen to the bottom of the reactor vessel.
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.
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a 2-Since the discovery of the damage to_the SSHTs, Babcock & Wilcox Company (B&W), the reactor supplier, has undertaken the design, manufacture and testing of an improved SSHT.
SS4Ts of this improved design are presently installed in Davis-Besse Un. No. 1 and Crystal River Unit No. 3 and are planned for installation in Three Mile Island Unit No. 2.
All three of these plants have reactors supplied by B&W and all are in the process of beginning initial operation within the next few months.
In addition, all of these reactors are of the same basic B&W 177 fuel assembly vessel design as ANO-1. The acceptability of the redesigned SSHTs has been demonstrated by a test program reviewed and approved by the staff and performed in conjunction with the Hot Functional Test performed at Davis-Besse Unit No.1.
Installation of the redesigned SSHTs in the Davis-Besse Unit No.1 and Crystal River Unit No. 3 reactor vessels did not present any unusual difficulties because it was performed prior to neutron activation of the reactor internals.
This will also be the case st Three Mile Island Unit No. 2.
Studie:; of methods to install the redesigned SSHTs in the irrad:ated B&W reactors indicate that substantial difficulties will be experienced, primarily t:ecause 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 ANO-1, they would cause significant radiation to personnel.
Based on their experience in removing the SSHTs at Three Mile Island Unit No. I and Rancho Seco Unit No.1, B&W estimated that installing SSHTs in irradiated reactors would result in personnel exposure of about 100 man-rem per reactor.
In th 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 re-installing the SSHTs in ANO-1 and the other irradiated B&W plants.
This program is complex and 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 at operating reactors which suffered damage to SSHis 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 surveillance program for reactors such as ANO-1 would be 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'cr more reactors that have been in service will now be irradiated in a new host reactor which has been fitted with tM newly-designed capsule holders on the thermal shield in less time anu with-out radiation exposure of the workmen; 2.
An augmented surveillance feature in which more weld metal specimens and some larger fracture mechanics [ compact tension (CT)] specimens will be included in the program; and 3.
A data-sharing feature in which all available irradiation data for all of the beltline welds of a given reactor vessel will be considered by the licensee or his consultants in predictirg the adjusted reference temperature ad in making any fracture analyses for that vessel.
Typically, several of the welds in any ane 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.
The specific program proposed for ANO-1 involves installing the remaining original ANO-1 surveillance capsules (ona has been removed and tested) in extra locations provided in the Davis-Besse Unit No.1 vessel. This plan will accomplish the original purpose of obtaining information on the effect of radiation on material that is represent tive of (although not identical to) the material in the ANO-1 reactor vessel on a schedule that provides an appropriate lead time over the vessel irradiation rate. The overall integrated prcgram also will provide information from surveillance programs in Crystal River Unit No. 3, Three Mile Island Unit No. 2, and Davis-Besse Unit No.1 on material considered to be essentially identical to the actual welds iri the ANO-1 vessel.
It is also importar.c to note that still more information relevant to the ANO-1 vessel materials will
- ' obtained from the NRC sponsored Heavy. Section Steel Technology (HSST) frradiation programs underway. Details are providnd below.
Two weld materials are of primary interest for the ANO-1 vessel, Procedure Qualification (P.Q.) numbers
- WF 112 and WF 182-1. These are used in
- Weld 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.
~ q 4-the top and center circumferential welds. The end 9 life (EOL) fluence for both of these welds is estimated to be 1.1 x 10 ( nyt, and both have compositions that are expected to make them relatively sensitive to radiation damage.
Weld P.Q. No. WF 18, used for the longitudinal welds, has low copper, making it less sensitive to radiation.
Further, the EOL fluence)gt the azimuthal locations of these longitudinal welds is low (7 x 10 ) so they will not become limiting during the service life.
Another shell weld, the lower circerferential, is made of a material that is expected to be radiation sensitive (P.Q. No. SA 1788), but the E0L fluence at this location is estimated to be at least an order of magnitude lower than that of the other circumferential welds, so it is not expected to be limiting.
The orginal ANO-1 surveillance material, WF 193, used the same heat of filler wire as WF 112 but a different batch of flux. Metallurgical cor.siderations suggest that the radiation behavior is affected more by the wire than the flux, thus it is expected to respond to radiation much like WF 112.
The following table Lhmes where samples of these three pertinent weld materials will be irradiated in tne proposed integrated program, what kinds of specimens will be used, and when information will be available under the present plan.
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CAPSULE IRRADIATION INFORMATION SPECIPEN JIEg DESIGNATION
- LOCATIONS AVAILABLE TYPES **
WF 112 OCI-E Oconee Unit No. 1 1977 CV (alrea:'y renc ;2d) 0 I-A Crystal River Unit No. 3 1985 Cv OCI-C Crystal River Unit No. 3 1989 Cv WF 132 THI-2-B Three Mile :sland Unit No. 2 1985 Cv, Ci TMI-2 D Three Mile Island Unit No. 2 1995 Cy TMI-2-F Three Mile Island Unit No. 2 1998 Cv, CT TEl-F Davis-Besse Unit No. 1 1979 Cv, CT TEl-B Davis-Besse Unit No. 1 1983 Cv, CT TEl-D Davis-Besse Unit No. I 1992 Cv, CT
- UF 193
/J!1-E Arkansas Unit No.1 1977 CV (alency AN1-A Davis-Besse Unit No.1 (to be CV AN1-C Davis-Besse Unit No. I determined)
Cy The irradiation schedule and withdrawal dates shown will be modified as initial test results are obtained and evaluated to optimize the information obtained.
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- 0CI-E - means capsule E from the Oconee Unit No.1 reactor TMI-2-B - means capsule B in the Three Mile Island Unit No. 2 reactor TEl-F - means capsule F in Toledo Edison's Davis-Besse Unit No. I reactor ANI-E - means capsule E frcm Arkansas Nuclear One - Unit No.1 reactor
- Cv - means Charpy V-notch specimen CT - means Compact Tension specimen A
. In addition to this integrated program, "research" capsules containing tensile, Charpy V-notch (Cv), and several sizes of CT specimens will be included in the overall B&W power reactor surveillance program.
Samples of the weld most likely to be limiting in ANO-1, P.Q. WF 112, will be irradiated in Davis-Besse, and samples of a weld made of the same heat of weld wire as WF 182-1 will be irradiated in the Crystal River Unit No. 3 program.
Details of withdrawal schedules will be determined later, and will depend on test results from the other programs.
Research programs being sponsored by the NRC will also provide useful 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 the reference temperature for nil-ductility (RT will be obtained, the main emphasis of the HSST programs is to devbp) methods that can be used to better evaluate low shelf toughness using the rather small specimens used in the power reactor programs.
The staff has evaluated the effectiveness of this overall program plan, and has concluded that the infonnation to be developed that is directly and indirectly relevant to the ANO-1 reactor vessel will be sufficient to provide assurance of safety margins against. vessel failure that comply with Appendix G,10 CFR 50.
Until data become available from the surveillance program, a conservatise prediction of radiation damage can be made by using R.G.1.99* for at least the next five years of operation. This Regulatory 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 Regulatory Guide.
Predictions of the adjustment of reference temperature and the drop in upper shelf energy are given graphically in R.G.1.99 as functions of copper and phosphorus content and of fluence.
In addition there is g
an " Upper Limit" line on each graph, which is to be used when information about the copper and phosphorus contents is inadequate.
Because the i
chemical analyses of the B&W welds have shown considerable variation, the NRC staff intends to use the Upper Limit lines as the basis for any prediction required at this time.
- Regulatory Guide 1.99, " Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials", July 1975. Revision 1 is to be published in April 1977.
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We have also considered the uncertainties involved in applying radiation effects information obtained in other reactors to the ANO-1 vessel.
'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, l.
Calculational methods for estimating the neutron flux at the reactor vessel wall and at 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. As a result, the fast neutron flux and fluence in these locations can generally be calculated to an accuracy of
+ 20%, particularly if some dosimetry checks are available. Dosimeters from the original ANO-1 surveillance program were removed and tested, so the fluence calculations for the vessel can be verified.
In addition, it should be noted that the effect of neutron radiation on reactor vessel steel varies as the square root of the fluence; hence, uncertainties of 20 to 50% in fluence are not highly significant.
We have also considered the fact that the design of the ANO-1 vessel, internals, and core is nominally identical to that of the other reactors which 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 vessel material will also be small.
4 s 2.
Although differe..ces 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 reactor to another are stated -to be relatively minor, because they have similar geometry.
We :onsidered the possible differences in neutron spectra that cou!d occur between the B&W power reactors involved in the integrated program.
Such effects can be dealt with, if necessary, through methods that are being developed for that purpose.
However, the worst expected differences are judged inconsequential based on present knowledge of irradiation effects. The neutron spectrum uncertainty will be kept under active scrutiny by the NRC staff and if additional developments (theoretical or experimental) suggest that the effect might be significant under some conditions, appropriate adjustments in reference temperature, drop in upper shelf energy or other suitable parameter can be made.
3.
The effect of the temperature of irradiation has also been the subject of considerable resegrch.
It is wgil 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 should be taken into accgunt if the irradiation temperature difference is over about 25 F.
Enough information is known to permit conservative evaluations of the effget of temperature differences of at least 50 F, and probably even 100 F or more. The differences in the temperature of the surveillance capsules and vessel walls between the B&W power reactgrs involved in the integrated program are expected to be less than 50 F, and can be conservatively evaluated.
4 4.
The effect of irradiation rate 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 material 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.
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CONCLUSION We have evaluated the adequacy of the proposed integrated, augmented reactor vessel material irradiation program for ANO-1 as an alternative to the original program that was interrupted by failure of the associated hardware.
We conclude that the proposed program will provide the infomation required to comply with Appendix G,10 CFR 50, and that the uncertainties involved in using data obtained from surveillance specimens i
irradiated in various other B&W pcwer reactors to establish ANO-1 vessel operating limitations are small and can be accounted for by imposition of appropriate margins. We also conclude that the associated Technical Specification changes to implement the program are acceptable.
Additionally, the proposed integrated, augmented program (with possible f
minor modification yet to tm finalized) should provide more useful 1
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 B&W power reactors will be based on the current revision of Regulatory Guide 1.99. At present, this is Revision 1.
ENVIRONMENTAL CONSIDERATION We havt. 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 s.ignificant 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 impact statement or' negative declaration and environmental impact appraisal g
i need not be prepared in connection with the issuance of this amendment.
j CONCLUSIONS We have concluded, based on the considerations discussed above, tnat:
(1) because the amendment does not involve a significant increase in the probability of consequences of accidents previously considered and does not involve a significant decrease in a safety margin, the amendment does 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, and (3) such activities will'be conducted in compliance with the Commission's regulations and the issuance cf this amendment will not be inimical to the commoq defense and security or to the health and safety of the public.
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