ML19309B065
| ML19309B065 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco, Davis Besse |
| Issue date: | 12/17/1976 |
| From: | Mattimoe J SACRAMENTO MUNICIPAL UTILITY DISTRICT |
| To: | Reid R Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML19309B066 | List: |
| References | |
| 13028, NUDOCS 8004020671 | |
| Download: ML19309B065 (20) | |
Text
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NRCAOmu,1SS U.S. NUCLEAR RESULATCRY C3' 3SION DOC
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50-312 NRC DISTRIBUTION FOR PART 50 DOCKET MATERIAL TO: Mr. R. W. Reid FROM: SMUD DATE OF DOCUMENT Sacarmento, Calif.
12-17-76 J.J. M&ttimoe DATE RECEIVED 12-28-76
& ETTER MOTORIZEO PROP INPUT FORM NUMBER QF COPIES RECEIV5D
[3 ORIGIN AL hpNCLAS$1FIED OCOPy 3 signed 35 CC DESCRIPTION Ltr notarized 12-17-76 trans the ENCLOSU RE Respohses to NRC Questions on l'
following:
Reactor Vessel Surveillance Program for RanelI Seco Sta. Unit 1.....
(40 cys enc 1 ree'd)
PLANT NAME: Rancho, Seco Unit 1 DO NOT nm'MT]
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DISTRIBUTION FOR REACTOR VESSEL SUPPORT INFO FOR OPERATING REACTORS PER MR. TRAICIELL 7-12-76 SAFETY FOR ACTION /INFORMATION DHL 12-29-76 ASSIGNED ADr BRANCH CMTEF?
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INTERNAL DISTRIBUTION
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NRC PDR KNIGHT ROSZTOCZY CHECK C. TRAMMELL SHA0 P. BARAN0WSKY P. NORIAN R. BOSNAK V. NOONAN EISENHUT SHA0 BAER BUTLER GRIMES J. COLLINS EXTERNAL DISTRIBUTION CONTROL NUMBER LPDR:
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SACRAMENTO MUNICIPAL UTILITY DISTRICT s.,
December 17, 1976 Director of Nuclear Reactor Regulation At'ention:
Mr. Robert W. Reid, Chief Operating Reactors, Branch 4 U. S. Nuclear Regulatory Commission Wa',hington, D. C.
20555 Docket No. 50-312 Rancho Seco Nuclear Generating Station, Unit No. 1
Dear Mr. Reid:
In your letter of November 12, 1976 you requested additional information with respect to the proposed integrated reactor surveillance program to be conducted at Davis Besse No. 1.
The answers to your specific questions are provided in the attachment.
Consistent with these answers, a revision to Proposed Technical Specification No. 43 submitted with our letter of July 9,1976 is also attached.
Your letter also indicated that a related concern is the program the District intends to employ to meet the requirements of 10CFR50, Appendix G, Paragraph V.C. when required. The responses to these questions address addi-tional surveillance capsules and test reactor data as part of a B&W Users Group integrated materials irradiation program which is designed to obtain the data required to conform to Appendix G. Paragraph V.B.
Paragraph V.C. was also considered in development of this program.
g3 Qf Respectively submitted M
SACRAMENTO MUNICIPAL UTILITY DISTRICT
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6 J. J. Mattimoe p#<'/g,$ f Assistant General Manager
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JB ty Mattier Notary Public 1h and for e County of(Sacramento, State of
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9
.i RESPONSES TO NRC QUESTIONS ON REACTOR VESSEL SURVEILLANCE PROGRAM RANCHO SEC0 NUCLEAR GENERATING STATION, UNIT N0. 1 1.
QUESTION:
Provide your contingency plans for assuring that your surveillance 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 host
- reactor (s) for a given outage and justify 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 irradiation stays ahead of the corres-ponding reactor vessel beltline region. This, in itself, will allow for somewhat other than normal outages at the host reactor. Also, there is redundancy incorporated in the operating reactor 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 reactcr.
The operations of this plant will be monitored as discussed in the response to Question No. 6 at the frequencies specified in the response to Question No. 7.
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 will be made as discussed in the response to Question No. 5.
The appropriate corrective action will be taken following review by the NRC.
The time that the host reactor can remain out of service is, of course, a function of the prior service and after a few cycles of operation it would essentially have to be retired from service to seriously jeopardize the program.
2.
QUESTION:
Provide your program and schedule for installing the redesigned surveillance capsule holders in your reactor in the event this action becomes necessary.
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- 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 tcoling and its current status is given in Table 1.
B&W estimates that 26 months will be required to complete 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 irradiated plants.
Once such a requirement is identified, a span time of approximately 29 months must transpire before a plant can be outfitted with installed holder tubes.
3.
QUESTION:
What is the schedule for withdrawal of your capsules from the host reactor (s)? Relate the schedule to predicted trends in adjusted reference temperature and Charpy upper shelf energy.
What arrange-ments have been made with the owners of the host reactors to assure that this wi'hdrawal 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 *he surveillance weld metal.
The District is in the process of formalizing an agreement with the Toledo Edison Company which would provide for the irradiation of Rancho Seco surveillance capsules in the Davis Besse reactor in accordance with the schedule shown in Table 2.
The agreement would provide for withdrawal of the capsules at normal refueling outages on a non-interference basis. As indicated in the response to Question No.1, there are several alternatives available to provide for improbable deviations from the proposed withdrawal schedule.
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, 4.
QUESTION:
Specify the minimum and maximum radiation lead times for:
(a) sur-veillance 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 1
time will be available to use in reviewing the heatup and cooldown pressurization limits as required by 10CFR50, Appendix H, Section IV.
i Such data will assure the requirrments of 10CFR50, Appendix G, Section V.B. are met for the future service period.
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 j
discussed in response to Question No. 3 will provide the required surveillance data at the proper intervals of service life regardless of the operatiomi status of the guest reactor.
Should the guest reactor's service be interrupted for an extended period of time with continued operation of the host reactor, the effect would be the ability to verify reactor 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 alterntative means of obtaining the required surveillance data'should the host reactor experience an extended outage.
The longest 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 instal-lation of surveillance specimen holder tubes should this be necessary (29 months for installation, 3 months for NRC review, and 5 months for development of alterntatives).
Implementation of the 2.5 EFPY lead time is not necessary until the beginning of Cycle 4 or later since the heatup and cooldown pressuri-zation limits can be conservatively reviewed for adequccy based on presently available surveillance data and conservative estimates on materials similar to those of the guest reactor vessel exposed to this range of fluence. This review is presently planned prior to the completion of 2 EFPY in accordance with Technical Specification 3.1.2.3.
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5.
QUESTION:
Indicate the corrective action to be undertaken-at the guest reactor if the limits specified in response to Question No. 4, 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 t
involve reactor shutdown. As discussed in the response to Question No.1, B&W has developed a synergistic surveillance program in which several welds and base metals will 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 and the test reactor programs are described in the
.L responses to the questions on " Traceability of Welds." The syner-gistic program _ assures that applicable data for the 177 fuel assembly B&W design reactor vessels, will be available through the design service life of the vessels.
s In the event that the host reactor has an extended outage which is of sufficient duration to endanger the timeliness of the data avail-ability, several possibilities exist that would minimize the impact of such an outage.
Such possibilities or alternatives are:
4 1
1.
An evaluation of the applicability of the available data (from that reactor or other reactors, including test reactors) to the guest reactor could be made.
Such evaluation may indicate the i
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.
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 outage occurs, since all-the above options require evalu-ation of the data which is or will be available in a timely manner.
The monitoring program described in the response to Question Nos.
[
6 and 7 provides assurance that the potential for untimely data will be identified in time so none of the alternatives, including eventual installation of SSHT's on an irradiated plant, is precluded.
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. 6.
QUESTION:
Describe how the operating 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 4, above.
RESPONSE
A procedure has been developed to determine fluence lead time based on conservatively) estimated fluence values versus 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 sub-tracted 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.
The basic equation for determining the lead factor (in EFPY) is:
Where: TE = Actual cumulative EFPY on the guest reactor based on licensed core power T,1 = Calculated equivalent EFPY the guest reactor E
would have to operate so that the fast fluence on the vessel wall equals the fast fluence of the capsules in the host reactor T,1 is calculated from the following equation:
E T,1 = C,1 + C,1 TH E
2 3
l Where:
C,i = A constant dependant on the initial equivalent 2
fluence on the capsules C,1 = Equivalent fluence factor from the host 3
reactor capsule location to the guest reactor vessel wall TH
= EFPY of the host reactor, cumulative For Rancho Seco, the applicable factors are as follows:
C,1 =.7, C,1 = 4.0 2
3
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. These factors should yield conservative lead times since capsule fluence is based on the lower limit expected in the host reactor and pressure vessel fluence is basedon the upper limit expected in the guest reactor.
7.
QUESTION:
Submit amended proposed Technical Specifications that reflect the appropriate portions of your responses to Questions 3, 4, 5 and 6 above.
RESPONSE
The proposed Technical Specification is attached.
SIMILARITY OF GUEST AND HOST REACTORS 1.
QUESTION:
Provide a comprehensive tabulation for the guest reactor and each host reactor, of the values of all parameters, including construction and operating characteristics, tFat may affect the fracture toughness of the reactor vessel material as it is irradiated. Discuss how all differences in these parameters are' accomodated in the integrated surveillance program.
RESPONSE
The reactor parameters which could possible affect the material properties as the vessel is irradiated are:
(1) The neutron flux energy spectrum, (2) the irradiation rate, (3) the irradiation temperature, and (4) the material type and initial properties.
Each of these is addressed below:
1.
Energy Spectrum:
As discussed in the responses to the questions under " Fluence Estimates," 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 accomodated.
2.
Irradiation Rate:
Any significant difference in dose rate obtained at the guest and host reactors would be due to the variation in power level and power distribution.
Since the licensed power levels are compatible, the only difference is the variation in load swings as the plant maneuvers.
When time averaged over multiple fuel cycles the
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. variation in power ' level and power distribution 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.
3.
Irradiation Temperature:
There are two differences in irradiation temperature considered.
The guest reactor vessel beltline inner surface and the surveillance specimens in the host reactor are exposed to reactor coolant at essentially inlet conditions. The temperature distribution in the surveillance specimens and capsules is controlled primarily by the temperature of the reactor coclant. This is due to the good heat transfer characteristics of the specimen / capsule configuration.
Thus, the variation in reactor coolant inlet temperatures due both to design differences and the variation as the plant is maneuvered must be considered.
The variation due to design differences between the host and guest reactors is insignificant as shown on Table 4.
Between partial (%15%) and full load conditions, the inlet temperature will vary by about 20*F as an inverse function of power level.
Figure 4.2-9 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 temperatures are considered too low to cause significant annealing.
The inlet temperature will also vary about 40 F 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 tenn.
4.
I4terial 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 acconinodated.
FLUENCE ESTIMATES 1.
QUESTION:
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?
1 RESPONSE:
Energy dependent neutron fluxes are determined by a discrete ordinate solution of the Boltzmann transport equation.
Specifically, ANISN, a one-dimensional code, and D0T,.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 cistribution, core liner, barrel, thermal shield, pressure vessel, and water regions.
Inclusion of the internal components is necessary to account for the distortions of the required energy spectrum by attenuation in these components.
The ANISN code uses the CASK 22-group neutron cross section set with S6 order of anguiar quadrature and a P3 expansion 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-eigth 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 calcu-lation uses S6 quadrature and a Pj cross section set derived from CASK.
Fluxes calculated with this D0T 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 raidal-location dependent whereas the latter is axial location dependent. A P /Pj correction factor is obtained by comparing two 3
ANISN 1-D model calculations in which only the order of scat.tering 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 applicable 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 pCi/gm) is given as follows:
Di = Ai 3 x 109 fi (E)&(E) $
F (1-e) 4 tj, 4M) j j
j j
E n j=1 Where: C
= normalizing constant, ratio of measure to calculated flux N
= Avagadro's number
n
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= atomic weight of target material i j
fj
= either weight fraction of target isotope in nth material of fission yield of desired isotope o (E) = group-averaged cross sections for material n n
$(E) = group-average fluxes calculated by DOT analysis Fj
= fraction of full power during jth time interval, tj Ai
= decay constant of ith material tj
= interval of power history T
= sum of total irradiation time, i.e., residual time in reactor, and wait time between reactor shutdown and counting Tj
= cumulative time from reactor startup to end of jth j
time period, i.e., Tj t-
=I k
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
4(E > 1.0 MEV) = C I 4'(E) I Fjtj E=1 j=1 Where: M = the number of irradiation time intervals; the other values are defined above.
The analytical model described above, for calculating fast fluence at the surveillance capsule includes the pressure vessel region.
- Thus, each calculation 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 sur-veillance capsules from five reactors to date, calculated data at the nearby weld positior.s should have similar reliability.
Dosimeter data comparisons from surveillance capsules irradiated at the host reactor will provide further comparisons with the analytical model.
Because of the similarity of the host and guest reactors, these comparisons will also be applicable to the pressure 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 i30%).
This documentation will be available following completion
.s
- of the available surveillance capsule tests and submittal to NRC in the form of a Topical Report is expected by June 1977.
2.
QUESTION:
Describe any dosimetry checks that you plan to make on the analytical 4
results.
RESPONSE
Dosimeter measurements from Oconee I (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 t 15% was noted in the fast flux (E>l MEV). Multiple dosimeters in surveillance capsules will be 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 provide data for further verification comparisons with the analytical technique 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 the response to Question No. 1.
3.
QUESTION:
What differences in neutron energy spectra and dose rate to 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 damage 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 andmaterials). The internal components design is the same for both guest and host reactors as discussed in the response to the question under " Similarity of Guest and Host Reactors." Thus, the relative energy spectrum at the same spatial location should not vary between reactors.
(And consequently, dose rate will vary directly as the fast flux.) The analytical model is a multigroup calculation with the same internals arrangenent using plant specific core parameters as discussed in the response to Question No. 1.
Consequently, no correction is required between plants since the significant variables are already l
accounted for in the calculation.
The use of mixed oxide fuel would l
harden the spectrum somewhat but any effect on dose rates should be within the analysis uncertainty limits.
Possible corrections in
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4. using data from test reactors will depend on the design of the test reactor program which is not final.
Since impact, tensile and 3
fracture data on many of the same materials will be obtained both from test reactors and the surveilance programs, a basis for com-parison will be available. Such comparison will determine if correction would be needed.
5 TRACEABILITY OF WELDS I
1.
QUESTION:
?
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, 4
including the surveillance weld, and their specific locations are j
given in Table 5.
2.
QUESTION:
State which weld or welds is expected to be controlling with regard to i
radiation damage and why, i.e., give expected neutron flux, initial RTNDT, Charpy upper shelf energy, and chemical composition for the
+
controlling welds.
RESPONSE
l Table 5 also lists the unirradiated RTND" and Charpy upper shelf energy (Cv-USE),theweightpercentofthepertnentelements,theexpected i
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 5, Weld WF70 has the highest adjusted RTNDT and the lowest adjusted Cv-USE of all the welds.
Note that the unirradiated impact properties of the beltline region materials are estimated. Welds WF154 and WF233 also have the potential of being controlling welds.
That is because their cop slightly lower (0.07 and 0.05%, respectively)per content is only than WF70, and they all have the same estimated initial impact properties.
The curveillance weld is considered representative of the beltline region welds because the copper content of WF193 is about the same (within 0.03% Cu) as for WF154 and WF233. The predicted end of service properties-for WF193, WF154, and WF233 are also about the same (within 30F for the. adjusted RTNDT and 2 ft-lbs for the adjusted Cv-USE).
i 3.
QUESTION:
Which. welds are represented in the surveillance capsules irradiated in your reactor?
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. RESPONSE:
See the response to Question 2.
4.
QUESTION:
j 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 177 FA B&W design power plants.
The welds-of Table 6 are con-sidered representative of the beltline region welds.
5.
QUESTION:
1 I
List any test reactor programs on radiation damage in which your weld metals are represented.
RESPONSE
Presently they are two test reactor programs in which representative welds will be studied. These programs are:
1.
HSST Irradiation Studies Program.
2.
NRC-NRL Inplace Annealing Studies Program.
j Data from these programs is, of course, readily available to the NRC.
6.
QUESTION:
List any other test reactor and surveillance programs in which welds that are expected to be in the same category as yours from the stand-point of radiation sensitivity are represented, which you intend to utilize.
RESPONSE
Other than the programs outlined in response to Questions 4 and 5, B&W is investigating the possibility.of irradiating similar weld metals in an EPRI program to be conducted in the latter half of 1977.
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(n STATUh0FREQUIRED TABLE 1 SSHT IRRADIATED PLANT INST'LLATION TOOLING A
Tool Status Comments 1.
Boring Mill Complete with Backup 2.
Pintle Removal Tool Complete, no Backup No backup Necessar) 3.
Drill and Tapper a) Basic tool 98% com-plete.
b) Backup drills and taps must be sealed water-tight and test-ed.
c) Drilling and_ tapping at other than pintle locations has not been developed.
4.
Thread Inspection Concept only Tool 5.
Spot Face Tool Basic tool 20% complete 6.
Spot Face Inspec-Concept only tion Tool 7.
.S.S.H.T.
Installa-Concept only tion Tool 8.
Verification of Concept only Bracket Contact Inspection Tool 9.
Crimping Tool Concept only 10.
Free Path Inspec-Complete, no Backup.
tion Tool
TABLE.2 INSERT AND WI'IllDRANAL SCilEDULE OF INTEGRATED PROGRAM AT DAVIS BESSE I Lead FUEL CYCLES loldar Factor Tuba at hT Location Capsule 0
1 2
3 4
5 6
7 8
9 10 11 12 13 58 WZ 9.6 Upper ANI-B x------6 53 DB-L1 x----- ---d 2*4 ANI-D x----
-- - - o 53 Lower RSI-B x------o 2 45
--- 6 RSI-E x------ - - - -
135 XY 6.9 Upper TEI-F x------6
---S2 1.
DB-L2 x-----
73
6 Lower TEI-B x- - - - - - -
2.5 TEI-E x----
- - -- --+ 16
-- 1 I 6
YZ 9.6 Upper ANI-A x-------
2.2 ANI-F x----
o 22
6 Lower ANI-C x-------
80 l
YX 9.6 Upper IISI-D x------- -----6 1.4
o RSI-A x----
2.63 Lower TEI-C x- - - - - - -
o 0
XW 6.9 Upper TEI-1; x-------
-- 2 5
o Lower RSI-C x------- ---- -
L) 1.55 WX 9.6 Upper TEI-A x-------
o
- b-[>
Lower RSI-F x-------
The essumed EFPD per cycle are 450 days for the first cycle and 250 days for the others.
'Ihe vclues to the right of o(Identgicatiog of Withdrawal) is the predicted "best estimate" accumulated neutron fluence x 10 (n/cm E>1Mev) at the. capsule location of the host reactor.
x
. Capsule Insertion ANI-Surveillance Capsules of ANO-1 o - Capsule Withdrawal RSI-Surveillance Capsules of Rancho Seco TEI-Surveillance Capsules of D.B-1 DB-L "New" surveillance capsules
TABLE 3 SCHEDULE' FOR WITHDRAWAL OF' RANCHO SEC0's' REACTOR VESSEL SURVEILLANCE CAPSULES FROM DAVIS BESSE UNIT 1 Predicted Impact Pmparties -
of Surveillance Weld Metal 4
Approximate Neutron Fluence to be Accumulated by Capsule RT Cv-USE (NDT (ft-Lbs) 2 Capsule Time of Withdrawal (E>1MeV, N/CM )
F)
Unirradiated 0
15 66 RSI-B.
Following the l'st cycle at 5.3 x 1018 (2a) 153 (3)-
48
' Davis Besse 1 18 (2a) 185 (3)
'46 -
RSI-D Following the 2nd cycle at 8.0 x 10 Days Besse 1 RSI-A To be withdrawn at the time 1.2 x 1019 (2b) 205 (3) 44 when the capsule's accumu-m l'
lated neutron fluence (E>l Mev) correspond to that at 1/4 of.ncho Seco reactar vessel wall location at approx-imately the end of vessel's design service life.
RSI-F To be withdrawn at the time 2.2 x 1019 (2b) 282 (3) 40 when the capsula's accumu-lated neutron fluence (E>l Mev) corresponds to that of Rancho Seco reactor vessel inner wall location at approximately the end of vessel's design service life.
RSI-C Standby
>2.2 x 1019
>282 (3)
<40
,j RS I-E Standby
>2.2 x 1019
>282 (3)
<40-(I) Withdrawal schedules may be modified to coincide with those refueling outages or plant shutdown of Davis Besse 1 most closely approaching the above withdrawal schedule. The schedule may also be modified, if necessary, after the evaluation of each capsule.
(2a) Predicted neutron flu'ence values for the capsules in the identified location of Table 2.
The assumption made in predicting the fluence values are given in Table 2.
r (2b) Predicted neutmn fluence values for Rancho Seco's vessel.
They are values extrapolated based on predicted power distribution leakage flux, and fuel handling procedures.
Values contain a 1.2 safety factor.
l (3)The adjustment 'o.f RTNDT is based on the predicted shift at 30 Ft-Lbs.
l.
[
r)
TABLE 4 COMPARISON OF' DAVIS BESSE 1 AND RANCHO SECO UNIT NO. l'
' avis Besse 1 Rancho Seco 1-D Parameter DesignHeatOutput(Core).-MWt
.2772 2772 Design _Overpcwer, % Design Power 112 112' System Pressure, Nominal 2,200 2,200' Coolant Flow Rate, lb/m x 10-6/GPM 131.3/352,000
-137.8/369,000 Coolant Temperature '( F).
Nominal Inlet 555 557 Average Rise in Vessel 53
-51 Average in Vessel 582 582 Fuel Assemblies, No.
177 177 Fuel Assemblies, Type-MKB (15x15)
MKB (15x15)
Core Barrel, ID/00, inches 141/145 141/145 Thermal Shield ID/0D, inches 147/151 147/151 Core l Structural Characteristics Core Diameter, inches (equivalent) 128.9 128.9 Core Height, inches (active fuel) 144 144 Reflector Thicknesses and Composition Top (water plus steel), inches 12 12
' Bottom (water plus '. teel), inches 12 12 Side (water plus steel), inches 18 18 Reactor Vessel Design Parameters Principal Material SA-508, Grade SA-533, Gra,de B B, Class II Design Pressure, Psig 2500 2500 Design Temperature, *F 650 650 ID of Shell, inches 1 71 171 00 Across Nozzles, inches 249 249 Overall Height of Vessel and Closure Head (over clad and instrument nozzles), feet / inches 40/8-3/4 40/8-3/4 Core Barrel and Thermal Shield,
. Principal Material 304 SS 304 SS i
i i
f r~
.,,,,_n._
~
TABLE 5 WELD METAL INFORMATION AND DATA FOR RANCHO SECO UNIT 1 e
1/4T E0L Location Unirradiated Neutron in Impact Data, Fluence Shift Adjusted USE
-Adjusted Weld Metal Reactor Transverse Chemistry Composition E>lMev in RTNDT RTNDT Reduction-USE Ident.
Wire Fl ux Vessel RTNDT USE Cu P
S Ni n/cm2(1)
AF F
Ft-Lbs WF29 72102 8650 L1, L3 (20)
(66)
.16
.017 0.010
.27 1.2E+19 178 198 31 46 WF70 72105 8669 L3 (20)
(66)
.27.014
.011
.46 1.2E+19 281 301 44 37, )
WF154 406L44 8720 C2 (20)
(66)
.20. 01 5
.021
.59
- 1. 2E+19 195 21 5 35
'43 WF233 T29744 8790 Cl, C3 (20)
.(66)
.22.015
.016
.55
- 1. 2E+19 215 235 37 42 WF193 406L44 8773 SW 15 66
.19.016
.008
.59 1.2E+19 190 205 34 44 (1) Estimate based on predicted power distribution, flux leakage, and fuel management procedures.
Values contain a 1.2 safety factor.
NOTE: L1 Upper Longitudinal Weld
=
L3 Lower
=
Cl Upper Circum. W ld
=
e Middle Circum. Weld C2
=
Lower Circum. Weld C3
=
s Surveillance Weld
'si)
=
~
~
.?
TABLE 6 MATERIAL PROPERTIES OF REPRESENTATIVE WELDS TO BE IRRADIATED IN SURVEILLANCE PROGRAMS OF 177 FUEL ASSEELY B&W DESIGN REACTOR VESSELS Weld Designation Cu P
Cy Use (Ft-Lbs)
RTNDT ( F)
W:1
.40
.020 67
+65 W2
.22
.024 65 0
.24
.01 6 78
+10 W4
.36
.011 74
-20 W5
.35
.01 5 72
+10 W6
.24
.022 70
+10 W7
.34
.01 5 81
+9 WF25
.29
.01 9 81
+9 WF112
.22
.024 65 0
WF182-1
.18
.01 4 83
+15 WF193
.19
.01 6 66
+15 WF209-1
.30
.020 66
+43 l
i l
.