ML19322B700
| ML19322B700 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 01/04/1977 |
| From: | Parker W DUKE POWER CO. |
| To: | Busche B Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 7912050704 | |
| Download: ML19322B700 (29) | |
Text
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ey NRG ror,u 195 U.S. NUCLE AR f.ECULAToHY C 4lSStoN ooCKET NUM*E A 50-269/270/287
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NRC DISTRIBUTION FoR PART 50 DOCKET MATERIAL TO:
FROM:
oATE oF ooCUMENT Duke Nwer Company 1/4/77 Mr. Benard C. Rusche Cl. A wete, North Carolina oATE RECEivEo s
Mr. William O. Parker, Jr.
1/7/77 NETTER ONoToRizEo PROP INPUT FORM NUM8ER oF COPIES RECE6VEo BoRioiNAL ANCLAS$1FIEo One signed OCoPy 45 copies encl. reevd.
oESCRIPTroN ENCLOSURE Ltr. re our 11/16/76 Itr. and their 12/9/76 1tr....trans the following:
Consists of responses to questions constituting the substance of their proposed integrated reactor vessel material surveillance program at the Crystal River site...
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Attention:
Mr. A. Schwencer, Chief C',.
Operating Reactor Branch #1 T:-
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Reference:
Oconee Nuclear Station Docket Nos. 50-269, -270, -287
Dear Mr. Schwencer:
Your letter of November 16, 1976 pointed out that exemptions from the pro-visions of 10CFR50, Appendix H had been issued to permit the operation of Oconee Units 1, 2 and 3, respectively for their present fuel cycles with the reactor vessel material specimens removed from their reactor vessels.
It was stated that review of a course of action other than reinsts11ation of specimens in the vessel from which they were removed may require an extended period of time on the part of the staff. Accordingly, our plans for obtaining compliance with 10CFR50, Appendix H were requested, and if other than reinstallation in the original vessel was chosen, additional information was to be provided.
i Our letter dated December 9, 1976 stated that it is vur intention to continue the irradiation of the Oconee surveillance capsules in a reactor of similar design. To this end, an agreement in principle has been reached with Florida Power Corporation for the irradiation of Oconee capsules in the Crystal River, Unit 3 reactor vessel.
It is expected that this agree-ment will be formalized in the near future. The attached responses to questions constitute the substance of cur proposal to NRC to permit this integrated reactor vessel material surveillance program at the Crystal River site.
In the attached, host reactor and guest reactors are Crystal River 3 and Oconee, respectively.
Formal proposal of this program and the necessary Technical Specifications will-be submitted in the near future.
Your letter also requested a description of any additional programs that we plan to implement to satisfy the fracture toughness requirements of Appendix G to 10CFR50.
Such a program would involve additional surveil-4 lance capsules irradiated at operating reactors and/or test reactor. We
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Mr. Bsnard C. Ruscha Page 2 January 4, 1977 have participated in the B&W User's Group effort toward such a program.
At this time, no formal agreement for such a program has been reached.
Information on this program will be provided when available.
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RESPONSES TO NRC QUESTIONS ON REACTOR VESSEL SURVEILLANCE PROGRAM h.
I 1.
Provide your contingency plans for assuring that your surveillance l
program will not be jeopardized by an extended outage of any other reactor (s) from which you expect to receive data. What time limits will you place on the those reactoris) for a given outage and justi-fy these limits.
RESPONSE
B&W has developed a combined program for irradiating surveillance specimens of welds of interest between operating reactors and test reactors.
Such a synergistic program will offer protection against an extended outage of the host reactor should this occur. Redundancy will be incorporated in the combined program by ensuring that most of the representative welds to be irradiated in operating reactors will also be irradiated in the test reactors. The fluence levels in the test reactor programs should be sufficiently high to ensure that the surveillance material stays ahead of the corresponding reactor vessel beltline region. This, in itself, will allow for somewhat other than nor=al outages at the host reactor. Also, there is redundancy incorporated in the operating reacter program so that if an outage occurs at one host reactor, at least one other host reactor will have representative weld metal in a neutron environment.
In summary, the combined surveillance program offers a double redundancy feature for the irradiation of representative weld metal should the host reactor suffer an extended outage.
There is no time limitation on an outage at the host reactor. The operations of this plant will be monitored as discussed in response to question 5 at the frequencies specified in response to question 6.
Should it be determined that an extended outage has the potential for allowing the fluence on the guest reactor vessel to approach the fluence on the surveillance capsules at the host reactor, a review of alternative sources of surveillance data vill be made, as discussed in response to question 4.
The length of time that the host reactor can remain out of service is, of course, a function of the prior service.
2a. Provide your program and schedule for installing the redesigned sur-veillance capsule holders in your reactor in the event this action becomes necessary.
RESPONSE
Due to the availability of applicable surveillance data from alternate sources it is not expected that installation of surveillance specimen holder tubes (SSHT) will be necessary.
In the remote event it does become necessary. B&W must first complete the development and testing of a substantial amount of required tooling. A tabulation of the required tooling and its current status is given in Table 1.
B&W esti=ates that 26 months will be required to ec,plete the develop-ment and testing of the above tooling.
This 26 months will have to be expended before holder tube installation can be initiated. Once'the tooling is developed an estimated 3 months will be required to install three surveillance capsule holders on an irradiated plant. This time estimate does not include any contingency for an improperly installed tube or failure of any tooling to perform as planned and tested.
At this time, B&W is not proceeding with continued development of tooling for the installation of surveillance capsule holders on irra-diated plants. Once such a requirement is identified, a tima span of at least 29 months must transpire before a plant can be outfitted with installed holder tubes.
2b. What is the schedule for withdrawal of your capsules from the host reactor (s)? Relate the schedule to predicted trends in cdjusted reference temperature and Charpy upper shelf energy. What arrange-ments have been made with the cuners of the host reactors to assure that this withdrawal schedule will be met.
RESPONSE
Table 2 lists the withdrawal schedule for the surveillance capsules, as related to the appropriate cycle at the host reactor. Table 3 presents the basis and justification for this withdrawal schedule.
Table 3 relates the schedule to the actual and predicted trends in adjusted reference temperature and charpy upper shelf energy of the surveillance weld metal.
Duke Power Company and Florida Power Corporation have agreed in principle on arrangements for the irradiation of the Oconee capsules at Crystal River Unit 3.
This agreement cocmits Florida Power to making every reasonable effort to assist Duke in meeting the capsule withdrawal schedule. As of this date, the agreement has not been formalized, pending Florida Power's receipt of a full power operating license for Crystal River Unit 3.
3.
Specify the minimum and maximum radiation lead times for: (a) surveil-lance specimens relative to the vessel beltline inner surface, and (b) surveillance specimens relative to the 1/4T position in vessel wall, which you will require for guest specimens exposed in the host reactor (s).
Justify the values specified.
RESPONSE
The minimum radiation lead time specified is 2.5 equivalent effective full power years for the surveillance specimen capsules at the host reactor relative to the guest reactor vessel surface location. This limit is based on assuring that surveillance data with adequate lead time will be available to use in reviewing the heatup and cooldown l
pressurization limits as required by 10CFR50 Appendix H, Secti(n IV.
l Such data will assure that the requirements of 10CFR50, Appendis G, l
Section V.B. are met for the future service period.
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The minimum lead time relative to the reactor vessel 1/4 thickness location is not specified since it will always be a greater lead time than the surface due to neutron attenuation through the vessel wall.
Maximum lead times are not specified, since the withdrawal schedule discussed in response to question 2b will provide the required surveil-lance data at the proper intervals of service life regardless of the operational status of the guest reactor, Should the guest reactor's service be interrupted for an extended period of time with continued operation of the hosc reactor, the effect would be the ability to verify vessel material properties for a longer period of service than would otherwise be possible. Thus no limits are required on maximum lead time.
The 2.5 EFPY lead time specified includes consideration of the time required to develop alternative means of obtaining the required surveillance data should the host reactor experience an extended outage. The longect lead time alternative is the ultimate installa-tion of surveillance specimen holder tubes (SSHT) at an irradiated plant and the expected 37 months (2.5 EFPY at 8 capacity factor) provides ample time for the necessary tooling development and installa-tion of SSHT's should this be necessary.
Implementation of the 2.5 EFPY lead time is not necessary until the beginning of Oconee 1 Cycle 7 or Oconee 2 and 3 Cycle 5 or later since the heatup and cooldown pressurization limits are being conservatively reviewed for adequacy based on presently available surveillance data and conservative estimates on materials similar to those of the Oconee reactors vessel exposed to this range of fluence. This review is presently planned to be completed in early 1977 as part of the evalua-tion of the surveillance capsules which ware tested in 1976.
4.
Indicate the corrective action to be undertaken at the guest reactor if the limits specified in response to Question 3 above cannot be met.
If the corrective action does not involve reactor shutdown, justify the proposed alternative.
RESPONSE
Since there are several alternatives, the corrective action will not involve reactor shutdown. As discussed in the answer to Question 1, B6W has developed a synergistic surveillance program in which several welds and base metals vill be irradiated in three operating reactors (Davis Besse 1, Crystal River 3, and Three Mile Island 2).
In addition, data for several of the same welds will be obtained in at least two test reactor irradiation programs. The welds to be irradiated in the test reactor programs are described in the responses to questions 11 thru 16.
The synergistic program asspres that applicable data for the 177 FA, B&W design reactor vessels, will be available through the design service life of the vessels.
In the event that the host reactor has an extended outage which is of sufficient duration to endanger the timeliness of the data availability, several possibilities exist that would minimize the impact of such an outage.
Such possibilities or alternatives are:
- i. '
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s 1.
An evaluation of the applicability of the available data (from that reactor or other reactors including test reactors) to the 4
guest reactor could be made.
Such evaluation may indicate the guest reactor capsules do not need to be irradiated within the expected time of the host reactor outage.
2.
The capsules which will generate applicable data for the guest reactor can be removed from the host reactor that is shutdown and inserted into another host reactor that is in operation.
i 1
3.
The pressure-temperature limit curves of the guest reactor could be developed with material properties conservatively assumed until applicable data is available.
The best alternative can only be chosen at the time at which the extended i
i outage occurs, since all the above options require evaluation of the data j
which is or will be available in a timely manner. The monitoring program described in response to Questions 5 and 6 provides assurance that the potential for untimely data will be identified in time so that none of the alternatives, including eventual installation of SSHT's on an irra-3
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diated plant, is precluded.
5.
Describe how the ope:ating staff of the guest reactor will keep informed of the exposure status of the guest specimens at the host reactor (s) relative to the limits specified in response to Question 3 above.
RESPONSE
A procedure has been developed to determine fluence lead time based on conservatively estimated fluence values as a function of effective full power years (EFPY) for the guest and host reactors. The lead time is defined as the equivalent EFPY of predicted neutron fluence received by the guest reactor at the reactor vessel surface subtracted from the equivalent EFPY of predicted neutron fluence received by the reactor vessel surveillance capsules at the host reactor. The procedure also considers the initial irradiation received by the capsules during their original exposure in the guest reactor prior to SSHT removal.
i The basic equation for determining the lead time (in EFPY) is:
Lead Time = T E,i E
j f
where: T = Actual cumulative EFPY on the guest reactor based on E
licensed core power i
T
= Calculated equivalent EFPY on the capsules in the host E,1 reactor T
is calculated f rom the following equation:
E,1 l
E,1 =
2,1 + C3,j H where:
C 2,1 = Constant dependent on initial equivalent fluence on the capsules 1
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location to the gues-reactor vessel wall T
= EFPY of the host reactot, cumulative H
6.
Submit amended proposed Technical Specification *. that reflect the appropriate portions of your responses to Questions 2, 3, 4 and 5 above.
RESPONSE
l The following is an example of the Technical Specifications which will be formally submitted in the near future, i'
4.2.13 Starting with initial operation of Oconee 1, Cycle 4 or Oconee 2 and 3, Cycle 3, whichever occurs first, the reactor vessel surveillance specimen capsule lead time shall be determined at the frequencies specifiad in Table 4.2-1.
Subsequent to Oconee 1, Cycle 7 or Oconee 2 and 3, Cycle 5 operation, whichever occurs first, if the lead time is less than 2.5 EFPY, a raport describing the means of providing the necessary reactor vessel surveillance data shall be submitted for NRC review 'within 150 days of calculating a lead time of < 2.5.
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Table 4.2-1 SURVEILLANCE SPECIMEN CAPSULE IRRADIATION COMPARISON FREQUENCY Capsule Lead (EFPY)
Comparison Frequency
< 3.5 quarterly 3.5 < 4.5 Semi-annual
> 4.5 Annual The equivalent effective full power years (EFPY) of predicted neutron j
fluence (E > IMEV) received by the guest reactor vessel at the surface location shall be subtracted from the equivalent EEPY of predicted q
neutron fluence (E > 1 MEV) received by the reactor vessel surveillance j
specimen capsules at the host reactor to determine the capsule lead 4
factor. The comparison frequency is to be determined by the lead factor calculated at the last scheduled comparison.
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Bases To assure the availability of adequate surveillance data for the Oconee reactor vessels, a program has been developed to monitor the irradiation of the surveillance specimen capsules at the host reactor, and compare this to the irradiation of the guest reactor vessel. Fluence estimates w'.ich are conservative in the appropriate direction are used for this comparison. The frequency of monitoring varies depending on the known neutron fluence lead factor between the capsules and the reactor vessel.
This provides ample time for anticipating problems and initiating correc-tive action should operation of the host reactor be seriously delayed.
The requirement that the lead factor be 2.5 EFPY by the end of Oconee 1, Cycle 6 or Oconee 2 and 3 Cycle 4, or corrective action be developed pro-vides assurance that surveillance data will he available in a timely manner to allow revisions to Technical Specification 3.1.2.3.
The lead factor of 2.5 EFPY is based on a.8 capacity factor and thus provides over 3 calendar years for consideration and Laplementation of all alterna-tives. The requirement of a factor of 2.5 EFPY lead time need not be implemented prior to operation for the fuel cycles indicated above since Technical Specification 3.1.2.3 is being reviewed for adequacy and will be updated for a period of approximately 10 EFPY total operation. This is based on preliminary results of recent surveillance capsule test resu.1 ts, which justify a period of operation in excess of the fuel cycles indicated above.
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t 7.
Provide a comprehensive tabulation for the guest reactor and each host reactor, of the values of all parameters, including construction and operating characteristics, that may affect the fracture toughness of the reactor vessel material as it is irradiated. Discuss how all differences in these parameters are accommodated in the integrated surveillance program.
RESPONSE
The reactor parameters which could possibly affect the material proper-ties as the vessel is irradiated are (1) the neutron flux energy spec-trum, (2) the irradiation rate, (3) the irradiation temperature and (4) the material type and initial properties.
Each of these is addressed below.
Energy Spectrum - As discussed in the response to Questions 8, 9 and 10, the relative neutron energy spectrum is primarily a function of the geometry and materials of the reactor internals components. As shown in Table 4, the dimensions and materials of both the host and guest reactors are essentially identical. Thus, there is no difference to be accoccodated.
Irradiation Rate - Any significant difference in dose rate obtained at the guest and host reactors would be due to the variations in power level and power distribution. Since the licensed power levels are comparable, the only dif ference is the variation in load swings as the plant maneuvers. When time averaged over anticipated fuel cycles the variation in power level and power distributicn due to maneuvers is expected to be comparable between plants. The comparability of reactor vessel surveillance results from a number of plants presently available, supports this conclusion.
Irradiation Temperature - There are two differences in irradiation temperature considered. The guest reactor vessel beltline inner surface and the surveillance specimens in the hout reactor are exposed to reactor coolant at e.ssentially inlet conditions. The temperature distribution in the surveillance specimens and capsules is controlled primarily by the temperature of the reactor coolant. This is due to the good heat transfer characteristics of the specimen / capsule configuration. Thus variation in reactor coolant inlet temperatures due b'.
h to design difference and the variation as the plant is maneuvere must be con-sidered. The variation due to design differences between the host and guest reactors is insignificant as shown on Table 4.
Between partial
(%157.) and f ull load conditions, the inlet temperature will vary by about 200F as an inverse function of power level.
Figure 4-9 in the FSAR shows this variation. The duration of this variation due to maneuvering is expected to be comparable between plants over time. This is supported by the comparability of reactor vessel surveillance results presently available from a number of plants.
In any case, the inlet condition te=peratures are considered too low to cause significant annealing.
The inlet temperature will also vary about 400F between the hot zero power condition and partial load. This variation is a direct function of power level (0-15%) and again is not significant due to the low temperature and the expected comparability in duration over the long term.
Material Type and Initial Properties - Both the host and guest reactors are constructed of similar materials as discussed in conjunction with the neutron spectrum consideration. Thus, there is no difference to be accommodated.
8.
Describe analytical techniques that you plan to use to estimate the fluence expected at the various welds of the beltline of your vessel.
How much uncertainty do you expect there to be in the fluence estimates?
RESPONSE
Energy dependent neutron fluxes are detersined by a discrete ordinate solution.of the Boltzmann transport equation.
Specifically, ANISN, a one-dimensional code, and DOT, a two-dimensional code, are used to calculate the flux at the detector position.
In both codes, the system is modeled radially from the core out to the air gap outside the pressure vessel. The model includes the core with a time averaged radial power distribution, core liner, barrel, thermal shield, pressure vessel, and water regions.
Inclusion of the incernal components is necessary to account for the distortions of the required energy spectrum by attenua-tion in these components. The ANISN code uses the CASK 22-group neutron cross section set with an 36 order of angular quadrature and a P3 expan-sion of the scattering matrix. The problem is run along a radius across the core flats. Azimuthal variations are obtained with a DOT r-theta calculation that models one-eighth of a plan-view of the core (at the core midplane) and includes a pin by pin, plant specific time averaged power distribution. The DOT calculation uses S6 quadrature and a P1 cross section set derived from CASK.
Fluxes calculated with this DOT model must be adjusted to account for lack of P3 cross section detail in calculations of anisotropic scattering, a perturbation caused by the presence of the capsule, and the axial power distribution. The first two items are both energy and radial-location dependent whereas the latter is axial location dependent.
A P /P1 correc-3 tion factor is obtained by comparing two ANISN 1-D model calculations in which only the order of scatterin was varied. The capsule perturbation factor is obtained from a comparison of two DOT x-y model calculations, one with a capsule explicitly modeled - SS304 cladding, Al filler region, and carbon steel specimens--and the other with water in those regions.
The effect of axial power distribution is determined from plant specific burnup calculations as a function of axial location for the outer rows of fuel assemblies. The net result from these parameter studies is a flux adjustment factor K which is applicabte to the appropriate dosimeters in the 177-FA surveillance programs.
The calculation described above provides the neutron flux as a function of energy at the dosimeter position. These calculated data are used in the following equations to obtain the calculated activities used for comparison with the experimental values.
The basic equation for the activity D (in uCi/gm) is given as follows:
I i'j 1( -'j }
D = A, 3. 7 x 104 1
f I o (E)t(E) E F (1-e) e g
E n
j=1 where C = normalizing constant, ratio of measured to calculated flux N = Avogadro's number A = atomic weight of target material i g
f = either weight fraction of target isotope in nth g
material of fission yield of desired isotope o (E) = group-averaged cross sections for material n listed in Table D-3
$(E) = group-averaged fluxes calculated by DOT analysis
= fraction of full power during jth time interval, t)
F A = decay constant of ich material g
t) - interval of power history l'
T = sum of total irradiation time, i.e., residual time in reactor, and wait time between reactor shutdown and counting r = cumulative time from reactor startup to end of jth time period, d
i.e.,
j r) = I h*
t k=1 The normalizing constant C can be obtained by equating the right side of the above equation to the measured activity. With C specified, the neutron fluence greater than 1 Mev can be calculated from 15 Mev M
&(E > 1.0 MEV) = C E
& (E) I F) t E=1 j=1 where M is the number of irradiation time intervals; the other values are 4
l defined above.
The analytical model described above, for calculating fast fluence at the l
surveillance capsule includes the pressure vessel region. Thus each cal-culation produces fluence data at the weld position as well as the capsule location.
Since analytical results currently being documented compare within i 15% to dosimeter measurements from sutveillance capsules from 5 reactors to date, calculated data at the ne. by weld position should have similar reliability. Dosimeter data comparisons from surveillance capsules irradiated at the host reactor will provide further comparisons
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with the analytical model. Because of the similarity of the host and guest reactors, these comparisons will also be applicable to the pres-sure vessel fluence calculation for the guest reactor since it uses the same analytical model.
B&W intends to document the uncertainty based on the contributing factors in both the calculation and the measurements from the present capsule evaluations (the value should be on the order of 2 30%). This documenta-tion will be available following completion of the surveillance capsule results and submittal to NRC in the form of a Topical Report is expected by June 1977.
9.
Describe any dosimetry checks that you plan to make on the analytical results.
RESPONSE
Dosimeter measurements from Oconee 1 (Cycles 1 and 2), Oconee II (Cycle 1),
Oconee III (Cycle 1), TMI-l (Cycle 1) and ANO-1 (Cycle 1) have been compared to the analytical model. A nominal difference of 15% was noted in the 1
fast flux (E>l MEV). Multiple dosimeters in surveillance capsules will be 1
in the host reactors and also in subsequent B&W plants to startup in the 1980's. When each capsule is removed dosimeter activities will be measured and then compared to the plant specific analytical result. This will pro-vide data for further vertification comparisons with the analytical techni-que which will be used for plant specific fluence calculations at both the host and guest reactors. No check is considered necessary for calculated data at the weld locations as noted in response to Question 7.
- 10. What differences in neutron energy spectra and dose rate do you predict for your reactor beltline and your surveillance specimens, wherever they are to be irradiated? Describe the corrections, if any, that will be made to the predicted radiation darage at your beltline welds as a result of these differences. Possible corrections include differences in specimen irradiation temperatures, differences in neutron spectra arising from differences in reactor geometry or a different type of fuel (e.g., mixed oxides), and differences in dose rate if some test reactor data are used.
RESPONSE
For the same fuel type (e.g., low enriched uranium), relative neutron energy spectrum is a function of only the internals components (geometry and materials). The internal components design $s the same for both guest and host reactors as discussed in response to Questien 7.
Thus the relative energy spectrum at the same spatial location should not vary between reactors. Therefore, dose rate will vary directly as the fast flux. The analytical model is a multigroup calculation with the same internals arrangement using plant specific core parameters as dis-cussed in response to Question 7.
Consequently no correction is required between plants since the significant variables are already accounted for in the calculation. The use of mixed oxide fuel would harden the spectrum somewhat but any effect on dose rates should be within the analysis uncer-tainty limits. Possible corrections in using data frcm test reactors will depend on the design of the test reactor program, which is not final.
Since impact, tensile and fracture data on many of the same materials, will be obtained both from test reactors and the surveillance programs, a basis for comparison will be available.
Such comparison will deter-mine if correction would be needed.
- 11. Identify the heats of weld wire and flux used in all beltline welds, and give specific locations where each is used.
RESPONSE
The heats of weld wire and flux used in all beltline region welds, including the surveillance weld, and their specific locations are given in Tables Sai Sb, and Sc for Oconee I, Oconee II and Oconee III, res-pectively.
- 12. State which weld or welds is expected to be controlling with regard to radiation damage and why, i.e.,
give expected neutron flux, initial RTNDT, Charpy upper shelf energy, and chemical composition for the controlling welds.
RESPONSE
For Oconee I - Table 5a also lists the unirradiated RTNDT and Charpy upper shelf energy (Cv-USE), the weight percent of the pertinent ele-ments, the expected end of service neutron fluence at the'l/4T vessel wall location, the predicted shift and adjusted RT DT, and the predicted N
drop and adjusted Cv-USE.
As shown in Table 5a weld SA 1585 has the highest adjusted RTNDT and the lowest adjusted Cv-USE cf all the beltline region welds. However, welds WF 25, and SA 1229 also have the potential of being the controlling weld because their predicted irradiated proper-ties are similar to those predicted for veld SA 1585. The surveillance weld, WF 112, is considered to be representative of the controlling welds.
the predicted irradiated properties for the surveillance weld are similar to those predicted for WF 25, SA 1229 and SA 1585 at the same fluence value. Note that the unirradiated properties of WF 25 and WF 112 were determined by testing and those for SA 1229 and SA 1585 are estimated.
For Oconee II - Table 5b also lists the unirradiated RT DT and Charpy N
upper' shelf energy (Cv-USE), the weight percent of the pertinent elements, the expected end of service neutron fluence at the 1/4T vessel wall location, the predicted shift and adjusted RTNDT, and the predicted drop and adjusted Cv-USE.
As shown in Table 5b weld WF 25 has the highest adjusted RTNDT and weld WF 154 has the lowest adjusted Cv-USE of all the beltline region welds. The other beltline recion weld, WF 112, is subjected to a very low neutron fluence and its properties are expected to be the same throughout the service life of the plant. Welds WF 25 and WF 154 both have the potential of being the controlling weld. The surveillance weld, WF 209-1, is considered to be representative of the controlling welds. The predicted irradiated properties for the surveil-lance weld are similar to those predicted for.WF 25, and WF 154 at the same fluence value. Note that the unirradiated properties of WF 25, WF 112, and WF 209-1 were determined by testing and those for WF 154 are estimated.
For Oconee III - Table Se also lists the unirradiated RTNDT and Charpy upper shelf energy (Cv-USE), the weight percent of the pertinent elements, the expected end of service neutron fluence at the 1/4T vessel wall loca-tion, the predicted drop and adjusted Cv-USE.
As shown in Table Sc welds WF 67 and WF 70 both have the same predicted irradiated properties. Both welds have the highest adjusted RTNDT and the lowest adjusted Cv-USE of all the beltline region welds. Weld WF 200 also has the potential of being the controlling weld because its copper content is only 0.01% lower than those datermined for WF 67 and WF 70.
The unirradiated properties of all the beltline region welds are estimated. The surveillance weld.
WF 209-1, is considered to be representative of the controlling weld.
The predicted adjusted RTNDT for the surveillance veld is 450F higher than that predicted for WF 67 and WF 70, however, the adjusted Cv-USE is about the same.
Even though the adjusted RTNDT for the surveillance weld is higher than for the controlling weld, the surveillance weld is considered representative because welds WF 70 and WF 209-1 have the same weld wire heat number.
- 13. Which welds are represented in the surveillance capsules irradiated in your reactor?
RESPONSE
See the response to Question 12.
- 14. Which welds, if any, are represented in surveillance programs for other reactors?
RESPONSE
Table 6 lists all the welds that are considered representative which will be irradiated as part of the surveillance program of this and other 177FA B&W design power plants. The welds of Table 6 are considered representa-tive of the beltline region welds.
- 15. List any test reactor programs on radiation damage in which your veld metals are represented.
RESPONSE
Presently there are two test reactor programs in which representative welds will be studied. These programs are:
I 1.
HSST Irradiation Studies Program.
l 2.
NRC-NRL Implace Annealing Studies Program.
Data from these programs is, of course, readily available to NRC.
- 16. List any other test reactor and surveillance programs in which welds that are expected to be in the same category as yours from the standpoint of radiation sensitivity are represented, which you intend to utilize.
RESPONSE
Other than the programs outlined in response to Question 14 and 15 B&W
ic investigating the possibilities of irradiating similar veld metals in a EPRI program to be initiated prior to mid-1977.
1
TABLE 1 STATUS OF REQUIRED SSiiT IRRADIATED PLANT INSTALLATION TOOLING Tool Status Comments 1.
Boring Mill Complete with Backup 2.
Pintle Removal Toc 1 Complete, no Backup No backup Necessary 3.
Drill and Tapper a) Basic tool 98% com-
]
plete.
b) Backup dri.11s and taps must be scaled water-tight and test-ed.
t c) Drilling and tapping at other than pint.le locations has not been developed.
4.
Thread Inspection ccr. cept only Tool 5.
Spot Face Tool Basic tool 20% complete 6.
Spot Face Inspec-Concept only t.i.un Tool 7.
S.S.H.T.
Installa-Concept only tion Tool 8.
Verification of Concept only Byacket Contact Inspection Tool 9.
Crimping Tool Concept only 10.
Free Path Inspec-Complete, no Backup tion Tool l
l r
TABLE 2 INSERT AND WITHDRAWAL SC1!EDULE OF INTEGRATED PROGRA!! AT CRYSTAL RIVER 3 CYCLES Lead I::1d;r Factor Tuba at 1/4T Location Capsule:
0 1
2 3
4 5
6 7
8 9
10 11 12 13 14 I
. 72 WZ 9.6 Upper CR3-L1
-a 1,3 a
2.3 OCII-B
--- yg
- t OCII-D 2.4
-o Lower CR3-L2
- r. - - - - - -
-+2. 2 OCIII-l' 20 68
- - = =
---o ZY 6.9 Upper CR3-C
--*2. 3 OCIII-E 18
--o 2. 2 Lower CR3-A x
,34 2.3 YX 9.6 Upper OCII-A x
o l
o OCI-C 1.17 Lower OCI-A x
--g
- - - - --2. 3 3:
OCI-B IT6 2.3 i
--o l
YX 9.6 Upper OCII-E
- x. --
42.3
- c-- -
OCII-F 18 2.2 o
Lower OCIII-D x
-*2. 3 OCI-D 18
. 31 2.2 XW 6.9 Upper CR3-3 x
Y
-o CR3-E 3.
1.0
-o Lower CR3-D x----
I (cont'd on page 2)
s
~
. cg INSERT AND WITilDRAWAL SCllEDULE OF INTEGRATED PROGRAM AT CRYSTAL RIVER 3 CYCLES Lead 11old:r Factor Tube at 1/4T Location Capsule 0
1 2
3 4
5 6
7I 8
9 1 10 11 12 l 13 14 31 WX 9.6 Upper OCIII-B
);----5 1.3' OCIII-C 3:--
=-
o 2*16 Lower CR-3-F
).
o l
Th3 essumed EFPD per cycle are 450 days for the first cycle and 250 days for the others.
The values to the right of o(Identific.ation of Withdrawal) is the predicted accumulated ncutron fluence x 1019 (n/cm E>1Mev) at the capsule location.
2 x - Capsule Int.ertion o - Capsule Withdrawal t
G e
.s
Table 3a SCHEDULE FOR WITHDRAWAL OF OCONEE I's REACTOR VESSEL SURVEILLANCE CAPSULES FROM T!!E CRYSTAL RIVER UNIT 3 Predicted Impact Properties j_
"Id Approximate Neutron Fluence t
gg) to be Accumulated by Capsule RTNDT Cv USC 2
r Capsule Time of Withdrawal (E > 1 Mev, n/cm )
(F)
(Ft-1.b s)
Unirradiated 0
0 65 17 OCI-F
!!as been withdre um for testing 8.3 x 10 (2a)
OCI-E Has been withdrawn for testing 1.5 x 10 a (2a) 120 56 18 OCI-A To be withdrawn at the time l.2 x 10 (2b) 220 4.0 when the capsule's accumulated neutron fluence (E > 1 Mev) corre-spond to that at 1/4 of ONS-1 reactor vessel wall location at approximately the end of vessel's design service life.
OCI-C To be withdrawn at the time when 2.2 x 10 s (2b) 280 38 l
1 the capsule's accumulated neutron fluence (E > 1 Mev) corresponds to that of ONS-1 reactor vessel inner wall location at approximately the end of vessel's design service life 18 OCI-B S tandby
>2.2 x 10 (2b)
No weld metal No weld metal 18 OCI-D Standby
>2.2 x 10 (2b)
No weld metal No weld metal (1)Uithdrawal schedules may be modified to coincide with those refueling outages or plant shutdown of crystal River 3 most closely approaching the above withdrawal schedule.
The schedule may also be modified, if necessary, af ter the evaluation of each capsule.
. (2a) Measured value using dosimeter data.
(2b) Predicted neutron fluence values for ONS-l's vessel. They are measured values extrapolated based on predicted power distribution leakage flux, and fuel handling procedures. Values contain a 1.2 tafety factor.
Table 3b SCHEDULE FOR WIIIIDRAWAL OF OCONEE II's REACTOR VESSEL t
SURVEILLANCE CAPSULES FROM THE CRYSTAL RIVER UNIT 3 I
Predicted Impact Properties
" ^
}
Approximate Neutron Fluence r_,
g to be Accumulated by Capsule RTNDT Cv-USE Capsule Time of Withdrawal (3)
(
(E > 1 Mev, n/cm )
(F)
(Ft-Lbs)'
2 Unirradiated 0
+10 68 OCII-C
.I 17 l
Has been withdrawn for testing 9.4 x 10 (2a) 100 54 l ' OdIIJd' '
Following the 2nd cycle at l
3.1 x 10 s (2b) 240 44 1
Crystal River 3 1
OCII-B
,,To be withdrawn at the time when the (
1.2 x 10 s (2c)
- i capsule's accumulated neutron fluence (E > 1 Mov) correspond to that at 1/4' iof.DNS-2 reactor vessel wall location'
.g at approxima,tely the end of vessel's
, design service life.
OCII-E To be withdrawn at the time snen 2.2 x 10 ' -(2c) 340 36 1
'the capsule's accumulated neutron fluence (E > 1 Mev) corresponds to
'thae of ONS-2 reactor vessel inner wall location at approxiuately the end of vessel's design service life.
OCII-D Standby
>2.2 x 10 ' (2c)
1 18 OCII-F Standby
>2.2 x 10 (2c)
No weld metal No weld metal (1) Withdrawal schedules may be modified to coincide with those refueling outages or plant shutdown of Crystal River-3 most closely approaching the above withdrawal schedule.
The schedule may also be modified, if necessary, after the evaluation of each capsule.
)
(2a) Measured value using dosimeter data.
(2b) Predicted neutron fluence value for the capsule in the identified location of Table 2.
The assumption made on,
predicting the fluence value are given in Table 2.
(2c) Predicted neutron fluence values for 20NS-2's vessel. They are measured values extrapolated based on predicted 1
power distribution leakage flux, and fuel handling procedures.
Values contain 1.2 safety factor.
e t
Table 3c SCllEDULE FOR WIT!!DRAWAL OF OCONEE III's REACTOR VESSEL SURVEILLANCE CAPSULES FROM THE CRYSTAL RIVER UNIT 3 i"
1 j
Predicted Impact Properties l
of Surveillance Veld Metal Approximate Neutron Fluence i
to be Accumulated by Capsule RT1DT Cv-USE Capsup Time of 'n'ltbdrawal,(1)
(E > 1 Mev, n/cm )
(F)
(Ft-Lbs) 2 i
I Unirradiated 0
60 66 i
.l OCIII-A Has been withdrawn for testing 7.4 x 10 (2a) 100 56 l
27 OCIII-B Following the 2nd cycle at 3.1 x 10 e (2b) 240 44 l
1 Crystal River 3 1
OCIII-C To be withdrawn at the time when 1.2 x 10 ' (2c) 340 38 I
the capsule's accumulated neutron fluence (E > 1 Mev) correspond i
to that at 1/4 of ONS-3 reactor e
vessel wall location at s
approximately the ead of vessel's design service life.
OCIII-D To be withdrawn at the time when 2.1 x 10 ' (2c) 380 34 1
the capsule's accumulated neutron l
l fluence (E > 1 Mev) corresponds l
to that of ONS-3 reactor vessel j
inner wall location at approxi -
g mately the end of vessel's design
,,,,,,,g service life.
j OCY$I-E Standby
>2.1 x 10 ' (2c) 380 34 1
OCIII--F,,
Standby
>2.1 x 10 ' (2c) 380 34 1
(1) Withdrawal schedules may be modified to coincide with those refueling outages or plant shutdown of Crystal River 3 most. closely approaching the above withdrawal schedule.
The schedule may also be modified, if necessary, af ter the evaluation of each capsule.
(2a) Preliminary measured value using dosimeter data.
l (2b) Predicted neutron fluence value for tha capsule in the identified location of Table 2.
The assumption made on predicting the fluence value are given in Table 2.
(2c) Predicted neutron fluence values for ONS-3's vessel.
They are measured values extrapolated based on predicted power distribution leakage flux, and fuel handling procedures.
Values contain a 1.2 safety factor.
'i
Table 4
[
Comparison of Oconee -1,
-2, -3, and Crystal River-3 Crystal Parameter Oconee-1 Oconee-2 Oconce-3 River-3 1
- 'I Design lleat Output (Core), FMt 2568 2568 2568 2452
I Design overpower, % Design Power'
' 112 112 112 114*
System Pressure, Nominal, psia 2200 2200 2200 2200 Coolant Flow Rate, Ib/hr x 10 '/GPM 131.3/
131.3/
131.3/
131.3/
~ ' ' ' ' "
352,000 352,000 352,000 352,000 Coolant Temperatures ( F)
Nominal Inlet 554 554 554 555
' Avg. Rise in Vessel i 50 50 50 48 Avg. in Vessel 579 579 579 579 Fuel Assemblics, No.
177 177 177 177 Fuel Assemblics, Type MKB(15x15) 1EB(15x15)
MKB(15x15) 1KB(15x15)
Core Barrel, ID/0D, in.
141/145 141/145 141/145 141/145 Thermal Shield, ID/0D, in.
147/151 147/151 147/151 147/151 Core Structural Characteristics
' Core Diameter, in. (equivalent) 128.9' 128.9 128.9 128.9 8
Core IIcight, in. (active fuel) 144 144 144 144
'l
Reflector Thicknesses and Composition Top (11ater plus Steel), in.
12 12 12 12 Bottom (tJater plus Steel), in.
12 12 12 12 Side (11ater plus Steel), in.
18 18 18 18 i
f A
Table 4 (Cont'd.)
Comparison of Oconce -1,
-2,
-3, and Crystal River-3 Crys tal Parar_eter Oconee-1 Oconee-2 Oconee-3 Rive r-3 Reactor Vessel Design Ptrameters Principal Material SA-302, SA-508, SA-508, SA-533, Grade B, Grade B, Grade B, Grade B, Class 1, Class 2 Class 2 Class 1 as-modi-fled by Code Case 1339 Design Pressure, psig 2500 2500 2500 2500 Design Temperature, F
650 650 650 650 ID of Shell, in.
171 171 171 171 OD Across Nozzles, in.
249 249 249 249 Overall Height of Vessel
~40-8-3/4 40-8-3/4 40-8-3/4 40-8-7/8 and Closure IIcad (over Cid and Inst. Nozzles),
ft/in.
J heactor Internals, Principal Material 304SS 304SS 304SS 304SS D
0 e
d.
.. ~.
- ~
I Table 5a
(
WELD METAL INFORMATION AND DATA (Oconee I) 1/4T EOL Location Unirradiated Iepact Chemistry Composition Neut. Fluence Shift in Adjusted,
USE Adjusted Weld Metal in reactor Data. Transverse E > 1 Mev RTNDT RTNDT Reduction USE Ident.
Wire Flux vessel 2
RT nT USE Cu P
S N1 n/cm (1)
AF F
Ft-lbs N
87 Vf9 72445 8632 C3 (20)
. (66)
.17
.015.012
.60
<1.5 x 10
< 20
< 40
<10
>59 WF25 299L44 8650 C1
-15 81
.34
.015.013
.71 1.2 x 10 '
290 275 44 45 8
SA1073 IP0962 8445 L1 (20)
(66)
.21
.025
.01'7
.64 7.6 x 10 '
198 218 33 44 8
8 SA1135 61782 8457 CO (20)
(66)
.17
.015.013
.50 2.6 x 10 '
115 135 23 51 SA1229 71249 8492 C1 (20)
(66)..20
.021.012
.57 1.2 x 10 '
246 266 36 42 8
SA1430 8T1762 8553 L3 (20)
(66)
.16
.017.015
.60 1.0 x 10 '
165 185 30 46 8
88' SA1493 8T1762 8578 L2 (20)
(66)
.22
.017.010
.43 7.5 x 10 195 215 34 44 8
SA1585 72445 8597 C2 (20)
(66)
.25
.016.011
.51 1.2 x 10 '
274 294 40 40 WF112 406L44 8688 SW 0
65
.22
.024.006
.58 1.2 x 10 '
285 285 37 41 8
(1) Measured values extrapolated based on predicted power distribution flux leakage, and fuel handling procedures. Values contain a 1.2 safety factor.
NOTE: L1 = Upper longitudinal weld L2 - Middle longitudinal vold L3 = Lower longitudinal weld CO = li!gher-upper circumferential veld C1 - Upper circum. weld C2 = >Uddle circum, weld C3 = Lower circum. weld SW = Surveillance veld
( )= Estimated per BAW-10046P 1
. _... ~.. _ _..
.. _... t.. j "W*
' -t: * ' *
'f.
u 4
r
Table 5b WELD METAL INFORMATION AND DATA (Oconee II) 1/4T EOL Iocation Unitradiated Impact Chemistry Composition Neut. Fluence Shift in Adjusted USE Adjusted Veld Metal in reactor Data. Transverse E > 1 Mev RTNDT RTNDT Reduction USE Idtnt.
Wire Flux vessel RTMDT USE P
S Ni n/cm*(l)
AF F
Ft-lbs 3
WF25 299L44 8650 C2
-15 81
.34
.015.013
.71 1.2 x 10 '
290 275 44 45 37 WF112 406L44 8688 C3 0
65
.22
.024.006
.58
<1.5 x 10
< 32
< 32
<10 58 WF154 406L44 8720 C1 (20)
(66)
.20
.015.021
.59 1.2 x 10 '
214 234 35 41 3
3 WF209-1 72105 8773 SW 10 68
.34
.013.010
.48 1.2 x 10 '
290 300 44 38 I
i l
I (1) Heasured values extrapolated based on predicted power distribution flux leakage, and fuel handling i
procedures. Values contain a 1.2 safety factor.
I I
NOTE: C1 = Upper circum. weld C2 = luddle circum. weld i
C3 - Lower circum, weld SW = Surveillance weld ej'
( ) = Estimated per BAW-10046F I
e
......j s
- : Y '.i-
- ' ~.-l I
._ ;. ; s S
a
=
Table 5e WELD METAL INFORMATION AND DATA (Oconee III)
C. -
1/4T EOL r h.'
- i. :
Location Unirradiated Impact Chemistry Composition Neut. Fluence Shif t in Adjusted USE Adj usted
...l,, Weld M.atal in reactor Data Transverse E > 1 Mov RTNDT RTNDT Reduction USE Id:nt.
Wire Flux vessel h
USE Cu P
S Ni
_n/ca (1)
AF F
Ft-lbs t
WF67 72442 8669 C2 (20)
(66)
.27.
.014.017
.57 1.2 x 10 '
285 305 42 38.
3 l*
UF70 72105 8669 C2 (20)
(66)
.27
.014.011
.46 1.2 x 10
285 305 12 38 WF169-1 8T1554 8754 C3 (20)
(66)
.106.014.013.59
<1.5 x 10 '
20 66 3
UF200. I 80IT44 8773 C1 (20)
(66)..26
.010.015
.64 1.2 x 10
252 270 41 39 WF209-1 72105 8773 SW 60 66
.34
.013.010
.48 1.2 x 10 '
290 350 44 37 8
(1)' Measured' val.ues extrapolated' based'on predicte'd' power' distrib'ution flux leakage, and fuel handling procedures. Values contain a 1.2 safety factor.
NOT6* Ci',- U'pp'er circum. weld C'2 - Middle circus. wcld C3 = Lower circum. wel.i Su = Surveillance weld s
...y..j 2
":I?.*
'.lC $5. " '
~..
s.
j'.
' V.'?si '
'j ~~-.E *
<},.j. e. ;
d 9
TABLE 6 I
MATERIAL PROPERTIES OF REPRESENTATIVE WELDS TO BE IRRADIATED IN SURVEILLANCE PROG!wG OF 177 F. A. B&U DESIGN PIACTOR VESSELS l
Weld C USE(Ft-lbs)
RTNDT ( )
i Designation Cu P
y W1 40
.020 67
+65 W2 22
.024 65 0
W3 24
.016 78
+10 W4 36
.011 74
-20 W5 35
.015 72
+10 3,
W6 24
.022 70
+10 W7 34
.015 81.3
+9 WF25 29
.0 19 81
+9 WF112 22
.024 65 0
WF182-1 18
.014 83
+15 WF193 19
.016 ~
66
+15 WF209-1 30
.020 66
+43
?
9 t
+
M g
1
~. -
~-
T b
- i. -
i I
._ - - - - _ -.,