ML20062K806
| ML20062K806 | |
| Person / Time | |
|---|---|
| Site: | Byron, Braidwood, McGuire  |
| Issue date: | 12/17/1993 |
| From: | Bauer J COMMONWEALTH EDISON CO. |
| To: | Murley T NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM), Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML19311B272 | List: |
| References | |
| NUDOCS 9312270263 | |
| Download: ML20062K806 (9) | |
Text
. - _
['
' Comm:nw: lth Edison f^N
/
1400 Opus Place Downers Grove. Ilknois 60515 December 17,1993 Dr. Thomas E. Murley, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Attention:
Document Control Desk
Subject:
Response to the Request for Additional Information Regarding the Byron /Braidwood Steam Generator Tube Sleeving Proposed License Amendment Byron Station Units 1 and 2 (NPF-37/66; NRC Docket Nos. 50-454/455)
Braidwood Station Units I and 2 (NPF-72/77; NRC Docket Nos. 50-456/457)
References:
See Attachment A l
Dear Dr. Murky:
R. R. Assa's letter dated December 2,1993 (Reference 1) transmitted a request for additianal information related to the Byron /Braidwood proposed license amendment regarding steam generator tube sleeving (Reference 2). The enclosed attachments provide CECO's response to this request for i
additional information.
l Attachment B addresses the questions pertaining to Topical Report BAW-2045PA, Revision 1,
-j
" Recirculating Steam Generators Kinetic Sleeve Qualification for 3/4 Inch Tubes". Attachment B also j
contains a copy of J.11. Taylor's letter to H. F. Conrad dated December 10.1993, accompanying affidavit, and B&W Nuclear Technologies (BWNT) Document #51-1228682-00, " Evaluation of BWNT's Kinetic Sleeving Process" This report has been classified as proprietary by BWNT.
Accordingly, i'. is respectfully requested that the subject document be withheld from public disclosure in accordancc with 10 CFR Section 2.790 of the Commission's regulations.
Attachment C addresses the questions pertaining to WCAP-13698, Revision 1. " Laser Welded Sleeves for 3/4 inch Diameter Tube Feedring-Type and Westinghouse Preheater Steam Generators", dated May 1993 (Proprietary). Attachment C also contains a Westinghouse authorization letter (CAVi-93-554), accompanying affidavit, Proprietary Information Notice, and Copyright Notice.
l h(
200006 m
93122702C3 031217 PDR ADOCK 05000454
[
gg
.p PDR
(
u k- _ _ _._
~
.~.-
~
i
?
E a
Dr. T.E. Murley December 17,1993-f i
Please note that the information presented in Attachment C contains information proprietary to Westinghouse Electric Corporation and is supported by an affidavit signed by Westinghouse, the owner of the information. The affidavit sets forth the basis on which the information may be withheld l
from public disclosure by the Commission and addresses with specificity the considerations listed in I
paragraph (b)(4) of Section 2.790 of the Cormnission's regulations. Accordingly, it is respectfully requested that the information which is proprietary to Westinghouse be withheld from public disclorure in accordance with 10 CFR Section 2.790 of the Commission's regulations.
Correspondence with respect to the proprietary aspects of the items in Attachment C or the supporting Westinghouse Affidavit should reference CAW-93-554 and should be addressed to N. J. Liparulo, Manager of Nuclear Safety &. Regulatory Activities, Westinghouse Electric Corporation, P.O. Box 355, Pittsburgh, Pennsylvania 15230-0355.
l To the best of my knowledge and belief, the statements contained in this document are true and I
correct. In some respects these statements are not based on my personal kn >wledge, but on I
information furnished by other CECO employees, contractor employees, and/or consultants. Such l
information has been reviewed in accordance with company practice, and I believe it to be reliable.
Please address any comments or questions regarding this matter to this office.
i Respectfully, c*f k k a
Joseph A. Bauer I
Nuclear Licensing Administ;ator
-i i
JAll/gp t.
Attachments cc:
R. R. Assa, Braidwood Project Manager - NRR II. Peterson, SRI - Byron S. G. Dupont, SRI - Braidwood a
- 11. Clayton, liranch Chief - Region III Office of Nuclear Facility Safety - IDNS l
r f
i b
1 i
ATTACIIMENT A REFERENCES 1.
R. R. Assa Letter to D. L. Farrar dated December 2,1993 transmitting a Request for
]
Additional Information Regarding the Byron /Braidwood Steam Generator Tube Sleeving Proposed License Amendment (TAC NOS. M87229, M87230, M87227, and M87228) 2.
J. A. Bauer Letter to T. E. Murley dated August 13,1993 transmitting a Byron /Braidwood Proposed License Amendment Regarding Steam Ger.erator Tube Sleeving Methodology j
i 3.
WCAP-13698 Revision 1. " Laser Welded Sleeves For 3/4 Inch Diameter Tube Feedring-Type.
And Westinghouse Preheater Steam Generators" l
4.
Topical Report BAW-2045PA Revision 1, " Recirculating Steam Generators Kinetic Sleeve Qualification For 3/4 Inch Tubes
- 5.
B&W Nuclear Technologies (BWNT) Document #51-1228682-00, Evaluation of BWNT's Kinetic Sleeving Procesc" i
'l i
l 1
I 1
ATTACIIMENT B i
1.
RESPONSE TO QUESTIONS PERTAINING TO TOPICAL REPORT BAW 2045PA, REVISION 1, " RECIRCULATING STEAM GENERATORS KINETIC SLEEVE QUALIFICATION FOR 3/4 INCil TULLES" 2.
IlWNT LETTER FROM J. II. TAYLOR TO II. F. CONRAD TRANSMITTING IlWNT DOCUMENT #51 1228682-00, " EVALUATION OF IlWNT'S KINETIC SLEEVING PitOCESS" 3.
IlWNT AFFIDAVIT l
l 4.
IlWNT DOCUMENT #51-1228682-00, " EVALUATION OF llWNT'S KINETIC SLEEVING PROCESS" I
I 4
I r
I 6
i i
l t
RESPONSE TO QUESTIONS PERTAINING TO TOPICAL REPORT BAW 2045PA, REVISION 1,
" RECIRCULATING STEAM GENERATORS KINETIC SLEEVE QUALIFICATION FOR 3/4 INCll TURES" i
t i
NRC QUESTION
)
1.
The repon covers Westinghouse Model "D" steam generators (SG); however, the bounding l
conditions in the Westinghouse repon (WCAP-13698, Revision 1, page 3-22) are different i
than Babcock and Wilcox's (B&W), (page 4-4). Do the conditions stated in the report bound the conditions at Byron and Braidwood.
.{
RESPONSE: The B&W and Westinghouse repons, noted above, specify the same primary l
and secondary design pressures: 2485 psig (2500 psia) - primary and L
1285 psig (1300 psia) - secondary. These values are consistent with the Byron and Braidwood (B/B) UFSAR. Other parameter values specified on the subject pages for both the Westinghouse and B&W reports, conservatively bound the assumed parameter values as noted in B/B UFSAR Table 5.4-3,
" Steam Generator Design Dm", Table 15.6-2, " Input Parameters Used in the ECCS Analysis", and Table 15.0-3, " Nominal Values of Pertinent Plant l'arameters Utilized in the Accident Analysis" (see attached tables).
As noted by the question, Westinghouse and B&W list some dissimilar parameters in the design section of the referenced pages. Westinghouse has used more conservative primary to secondary parameters for it's calculations.
These assumed parameters (such as 235 psig secondary pressure during design i
primary pressure conditions and 465 psig primary pressure during design
{
secondary pressure conditions) are included merely to illustrate the origin of i
the conservative primary to secondary (2250 psig) and secondary to primary 2
(820 psig) differential pressures.
I In summary, CECds seview confirmed that both the Westinghouse and B&W reports are consistent with or bound the applicable parameter values specified in the Byron and Braidwood UFSAR.
2.
There is a change in the postweld heat treatment (PWHT) which may require re-qualification.
l This issue must be addressed.
t
RESPONSE
Babcock and Wilcox is providing information to address this issue in threc separate documents:
a.
BWNT Document #51-1228682-00, " Evaluation of BWNT's Kinetic Sleeving Process" was previously transmitted to the NRC by
_y J. II. Taylor's letter to H. F. Conrad dated December 10,1993. This
~
repon provides the results of the root cause evaluation of the McGuire i
Unit 1 1993 tube failure event as well as guidelines for evaluating in-service sleeves. A copy of this document is enclosul for your convenience, i
4 i
l 1
4 RAl: BAW-2045PA PAGE 2 of 2
-i 1
b.
BWNT will provide CECO and the NRC with a repon that will-establish correlations between post weld heat treatment temperatures j
and tube micro-hardness. It is CECO's understanding that this report will use the McGuire evaluation and independent test results to develop relative tube material corrosion susceptibility thresholds at lower stress relief temperatures. Using the independent test information, B&W will also develop stress relief criteria that is j
independent of material makeup. In conclusion, the report will arrive at specific recommendations regarding post weld heat treatment for the -
'l Byron and Braidwood steam generators, based on tube heat data _
2l provided by CECO. BWNT has indicated that this report will be published on or about December 24,1993.
c.
BWNT will then revise the report (from item b above) to incorporate j
x-ray diffraction information. X-ray diffraction is an indicator of i
residual stresses and is used to make determinations related to the
-l quality of the stress relief process. The revision to this report will be _
i issued on or about January 15,1994. In addition, B&W will provide
.l confirmatory corrosion test data later in the summer ci i994, i
.- l
)
i
~
i P
t P
't f
-l P
6 I
i f
1
-t
~
i l
B/B-FSAR TABLE 5.4-3 STEAM GENERATOR DESIGN DATA Design pressure, reactor coolant side, psig 2485 l
Design pressure, steam side, psig 1185 1
Design temperature, reactor coolant side, *F 650 Design temperature, steam side, *F 600 Total heat transfer surface area, ft 48,300 Maximum moisture carryover, wt percent 0.25 Overall height, ft-in 67-8 l
Number of U-tubes 4578-U-tube nominal diameter, in.
.750 Tube wall nominal thickness, in.
.043 Number of manways 4
Inside diameter of manways, in.
16
.i Number of inspection ports 4
Design fouling factor 0.00005 Preheat section 0.00010 s
i t
i
?
i i
i h
5.4-54
i B/B-FSAR AMENDMENT 43 SEPTEMBER 1983 TABLE 15.6-2 INPUT PARAMETERS USED IN THE ECCS ANALYSIS f
Licensed core power (a), (MWt) 3411 i
Peak linear power, includes 102% f actor (kW/ft) 12.88 Total peaking factor,Ff 2.32 Axial peaking factor, F 1.4968 l
g Power shape Chopped cosine Large break See Figure 15.6-48 Small break i
Fuel assembly array Optimized 17x17 3
Accumulator water volume, nominal (ft / accumulator) 950 3
Accumulator tank volume, nominal (ft / accumulator) 1350 600 Accumulator gas pressure, minimum (psia) t See Figures 15.6-21 Safety injection pumped flow and 15.6-47 See Sec. 6.2 Contain=cnt parameter s Initial loop flow (lb/sec) 9792 Vessel inlet-temperature (OF) 555.4 Vessel outlet temperature (OF) 616.9 Average Reactor coolant pressure (psia) 2280 990 Steam pressure (psia)
Steam generator tube plugging level (%)
0 (a) Two percent is added to this power to account for calorimetric error.
t 15.6 -35
B/B-FSAR AMENDMENT 47 APRIL 1986 TABLE 15.0-3 Nominal Values of Pertinent Plant Parameters Utilized In The Accident Analyses
- Thermal output of NSSS (MWt)a 3425 Core inlet temperature (oF) 559.3
]
Vessel average temperature (or) 588.4 l
Reactor Coolant System pressure (psia) 2250 Reactor coolant flow per loop (gpm) 97,600 Total Reactor Coolant flow (106 lb/hr) 145.1 Steam flow from NSSS (106 lb/hr) 15.13 Steam pressure at steam generator outlet (psia) 990 Maximum steam moisture content (%)
0.25 Assumed feedwater temperature at steam 440 generator inlet (oF)
Avorage core heat flux (Btu /hr-f t2) 197,200
- For accident analyses using the Improved Thermal Design Procedure oSee Table 15.0-2 1
-15.0-30