ML20117F311
| ML20117F311 | |
| Person / Time | |
|---|---|
| Site: | South Texas  |
| Issue date: | 05/14/1996 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20117F306 | List: |
| References | |
| NUDOCS 9605170332 | |
| Download: ML20117F311 (7) | |
Text
"
p no ge 4
UNITED STATES g
j NUCLEAR REGULATORY COMMISSION o
~t WASHINGTON, D.C. 20eeM001 49.....,o SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED T0 fMENDMENT NO. 82 TO FACILITY OPERATING LICENSE NO. NPF-76 i
HOUSTON LIGHTING & POWER COMPANY CITY PUBLIC SERVICE BOARD OF SAN ANTONIO CENTRAL POWER AND LIGHT COMPANY CITY OF AUSTIN. TEXAS DOCKET NO. 50-498 SOUTHTF.Wi. PROJECT. UNIT 1 i
l
1.0 INTRODUCTION
By application dated January 22, 1996, as supplemented by letter dated April 18, 1996 Houston Lighting & Power Company, et.al., (the licensee) l requested changes to the Technical Specifications (Appendix A to Facility l
Operating License No. NPF-76) for the South Texas Project, Unit 1.
The proposed changes would modify the steam generator tube plugging criteria in Technical Specification (TS) 3/4.4.5, Steam Generators, ara the associated i
Bues, to allow the implementation of alternate steam generator tube plugging criteria for the tube-to-tubesheet joints (known in the industry as F*) for Unit 1.
The April 18, 1996, supplement provided clarifying information and did not change 1.he initial no significant hazards consideration determination.
l
2.0 BACKGROUND
The licensee proposed an alternative repair criteria for defects found in the W as generator tube expansion region within the tubesheet. Under the ao ndment request, steam generator tubes with defects in excess of the current plugging limits could remain in service without repair provided the defects exist below a specified distance (F* distance). The F* distance is measured from the secondary face of the tubesheet or the top of the last hardroll, I
whichever is lower. The licensee presented a qualification test program, which was performed by Babcock and Wilcox Nuclear Technologies, to show that the proposed F* distance satisfies the necessary structural and leakage integrity requirements of Appendix A to 10 CFR Part 50 and the existing plant TSs.
The test program was documented in Topical Report, BAW-10203P, Revision 0, which was included in the submittal.
9605170332 960514
?
PDR ADOCK 05000498 p
. The proposed F* criteria apply to Unit 1 only. Unit I uses Westinghouse E-Series steam generators. The tubes are fabricated from low temperature mill annealed Alloy 600 material. The tube-to-tubesheet joints in the Unit I steam generators 7.re fully expanded by a hardroll process. The tube-to-tubesheet joints in the Unit 2 steam generators are expanded by a hydraulic process, which is not covered by this proposed amendment.
3.0 DISCUSSION General Design Criterion (GDC) 14 of Appendix A to 10 CFR Pan 50 requires that the reactor coolant pressure boundary be designed, fabricated, erected, and tested so as to have an extremely low probability of abnormal leakage of rapidly propagating failure, and of gross rupture. Regulatory Guide (RG) 1.121 provides guidance on an acceptable method for establishing the limiting safe conditions of tube degradation of steam generator tubing.
Surveillance reqtfirements in the STP TSs require a periodic inspection of steam generator tubes.
If a tube is found to have defects in excess of the TSs limits (i.e., 40 percent through-wall), it is required to be repaired or removed from service.
Steam generator tubes comprise a significant portion of the reactor coolant pressure boundary. Maintenance of this boundary is provided by the integrity of the steam generator tube wall and tube-to-tubesheet joint. The joint between the tube and tubesheet is an interference fit constructed by roll expanding the tube into the bore of the tubesheet. The tubes were inserted into the bore of the tubesheet followed by seal welding at the primary face of the tubesheet.
Step rolls were then performed on the tube to fully expand the tube in the bore of the tubesheet. The tubes are restrained in the bore of the tubesheet by the elastic springback of the tubesheet. The tube-to-tubesheet joint provides sufficient strength to maintain adequate structural and pressure boundary integrity.
The industry experience has shown that defects have been developed in the tube-to-tubesheet joints by various degradation processes. The staff believes that tubes having degradation in the joint may remain in service provided that the degradation is below a specified distance and that the undegraded portion of the tube in the joint can maintain adequate structural and leakage integrity under loadings from normal operation, anticipated operational occurrences, and,.ostulated accident conditions.
RG 1.121 recommends that the margin of safety against tube rupture under normal operating conditions should not be less than three at any tube location where defects have been detected.
For postulated accidents, RG 1.121 recommends that the margin of safety against tube rupture be consistent with the margin of safety determined by the stress limits specified in NB-3225 of Section III of the Boiler and Pressure Vessel Code of the American Society of Mechanical Engineers (ASME).
Structural loads imposed on the tube-to-tubesheet joint primarily result from the differential pressure between the primary and secondary sides of the tubes. The peak postulated loading occurs during a main steam line break due to a lowering of the secondary side pressure.
However, normal operating loads, cyclic loading from transients (e.g., startup/ shutdown), and potential
1 4
1 thermal expansion loads can also be significant. The analysis supporting tis i
F* criteria should address all loading conditions necessary to maintain adequate integrity of the tube-to-tubesheet joint.
The elastic preload betwen the tube and tubesheet not only prevents pullout of the tube from the tubesheet, but also provides a leak tight barrier minimizing the potentini for primary to secondary coolant leakage. With 4
sufficient length of haviroll, the tube-to-tubesheet joint will not allow any 1
leakage under normal ar.d faulted conditions.
If a through-wall crack is
- ~
present in the joint, primary coolant may leak to the secondary side. Under the proposed F* criteria, the licensee has to demonstrate that leakage integrity of the joint is maintained under all analyzed conditions.
j 4.0, EVALUATION 1
4.1 Qualification Test Program 1
The licensee focused the qualification analyses and testing program for the F*
criteria on (1) establishing axial tube loads based on operating and faulted conditions, (2) performing mechanical tests necessary to verify the F*
criteria, (3) adjusting results from mechanical tests for actual steam i
generator conditions, and (4) performing eddy current inspection to determine the uncertainties associated with length measurements for final F* distance determination.
4.1.1 Fabrication of Test Specimens 4
The licensee fabricated several mockup blocks from the material with the same l
properties as the tubesheet material used in the Westinghouse E-Series steam 4
generators. The blocks were 4 inches thick having a 4 by 4 array of holes to
}
simulate the tubesheet.
The F* tubes were located in the middle of the test 3
block and surrounded by the undegraded test tubes. The F* tubes were installed as two separate sections in the bores of the blocks. The tube sections were inserted from the top and bottom of the blocks and they were roll expanded in the bore.
The separation between tube sections represented a full 360 degree through-wall crack at the F* distance. Various installation parameters such as tubesheet bore diameter, tubesheet bore surface finish, block temperature, and tubing yield strength were considered to address ranges of possible stean generator conditions. After the tubes were expanded into the blocks, the blocks were heated to simulate the effects of actual steam 4
generator service temperature.
4.1.2 Testing for F* Determination i
The licensee used a combination of internal pressurr and axial loading to simulate the applied loads on the steam generator tubes. The radial preload stress, thermal effect, internal (primary side) pressure effect, and tubesheet bow effect were considered.
The radial preload stress was calculated from the I
tube springback distance.
The springback distance was determined by cutting the block away from the test tube and comparing the difference between the i
L
- l. :
outer diameter of the expanded tube and the bore of the block.
The thermal effect was determined from the differential thermal expansion of the tube and l
tubesheet in the radial direction. The primary-to-secondary differential pressure acts to both increase and decrease the strength of the tube-to-tubesheet joint. The primary pressure in the tube increases the radial stress on the outside surface of the tube and thus increases the joint strength.
However, the primary-to-secondary differential pressure causes tubesheet bow and, therefore, dilates the tubesheet bore. When the bore dilates, the joint strength is reduced. Overall, the primary pressure would slightly reduce the strength of the tube-to-tubesheet joint.
A series of tensile tests were conducted to deterMna strength of the tube-to-l tubesheet joint under the maximum load, locked tuba loed, pressure cycling l
load, and ultimate load. For the maximum lor: test, u hl and pressure loads l
were applied to the joint. The axial loads weluded iiG 1.121 recommended I
safety factors of 1.43 and 3 for the fauMed and normal conditions, respectively.
For the locked tube test, the tube was aswmed to be locked at j
a tube support plate and the loads caus d by the difference in thermal j
l expansion of the tube, tube support plate, and asscciatad tube assembly were l
used.
For the pressure cycling test, the test tubes were subjected to cycling pressure simulating transient conditions w,d,ioint slippage was monitored to determine the structural integrity of the joints.
For the ultimate load test, the test tubes were internally pressurized and subjected to an increasing axial tensile load until failure. Failure was defined as a relative movement of a specific distance between the tube and tubesheet. The tested engagement lengths showed acceptable structural adequacy for all loading conditions.
As part of the test program, the test tubes were subjected to leak rate testing. The tubes were internally pressurized to sN1 ate differential pressures during normal operating and faulted conditions. The acceptance criteria for these tests were baseG w the 1 gallon per minute requirement of the plant TSs.
Low leak rates were observed for the normal operating condition and faulted condition.
Applying the F* distance to t*.e tubes in the field would require measuring the tube length by eddy current probes. The licensee stated that any uncertain-ties associated with eddy current measurement beyond an eddy current indication will be considered in the F* distance. To determine the measurement uncertainty, the licensee used various probes with various frequencies and rate of probe pull. A maximum measurement uncertainty was obtained and was included in the final F* distance calculation.
l l
The licensee considered the effects of boric acid corrosion on the steam generator tubes because if primary coolant leaked through the tube wall and t
came in contact with the carbon steel tubesheet, stress corrosion cracking may i
occur in the tubesheet. On the basis of previous industry studies, the i
licensee concluded that any corrosion that might occur is limited only to i
localized degradation in the bore of the tubesheet with an insignificant
{
depth. The licensee concluded that this would not affect the F* tube nor the structural integrity of the tubosheet.
i i
i j
c
e 1
~.
i 1
l l
4.1.3 F* Test Results I
Based on the results of the leakage rate and mechanical testing, the licensee determined a nominal engagement length necessary to ensure adequate margins of structural integrity as reported in BAW-10203P, Revision 0.
The length also i
e included uncertainties that accounts for limited sample size, statistical scatter in the data, and addy current inspection error.
l In addition, the licensee identified a number of tubes having " wavy" roll j
expansion in the Unit I steam generators as listed in Table 3.3.3 of BAW-10203P, Revision 0.
The wavy tubes do not have full 360 degree roll t
contact in the tubesheet because of certain rolling anomalies. The licensee has not qualified these tubes for structural and leakage integrity and has I
requested that the wavy tubes be excluded from the application of the F*
l crit'eria. The staff finds this exclusion acceptable.
i i
4.2 Evaluation of Proposed Technical Specification Changes l
The proposed F* criteria in the applicable TS sections are as follows:
i j
TS 4.4.5.2.d:
i For Unit 1, any tube allowed to remain in service per Acceptance Criterion 10 (of Technical Specification 4.4.5.4) j shall be inspected via the rotating pancake coil (RPC) eddy j
current method over the F* distance.
Such tubes are exempt from eddy current inspection over the portion of the tube below a
the F* distance which is not structurally relevant.
TS 4.4.5.4.a.10:
F* Criteria [For Unit 1 only] Tube degradation below a specified distance from the hard roll contact point at or near the top-of-tubesheet (the F* distance) can be excluded from consideration to the acceptance criteria stated in this section (l.u., plugging of such tubes is not required). The methodology for determination for the F* distance as well as the list of tubes to which the.F* criteria is not applicable is described in detail in Topical Report BAW-10203P, Revision 0.
The licensee also proposed changes to the associated Bases. The proposed Bases describes areas of Unit 1 tubes that are not structurally relevant and therefore are excluded from consideration to the customary plugging criteria, and describes the consideration of priron to secondary leakage from tubes left in service by the application of u P criterion and the associated radiological consequences during postulated accident conditions.
The staff has determined that 1) the licensee's F* criteria were established on the basis of the qualification tests that used specimens simulating the actual tube-to-tubesheet joint configuration of the South Texas steam 1
.._~ _ _ _ _ _.
, l generators, 2) the applied loads for structural and leakage assessment were l
specified in accordance with RG 1.121, and 3) the licensee has committed to reinspect the F* tubes during all future outages to ensure integrity of the F*
tubes. The continued inspection of the F* tubes will minimize the potential for the degradation to occur inside the F* distance of all the F* tubes, j
4.3 Summary J
l The proposed F* criteria wculd permit steam generator tubes to remain in service with degradation in excess of the current plugging limit provided the degradation exists below the F* distance. The proposed F* distance as reported in Topical Report, BAW-10203, Revision 0, is measured down into the tubesheet from the secondary face of the tubesheet or the top of the last hardroll, whichever is lower. The staff concludes that the proposed F*
l criteria is acceptable for Unit I because the licensee has demonstrated I
through an acceptable qualification test program that the F* criteria satisfy GDC 14 and the guidance in RG 1.121. The licensee may incorporate the proposed changes into the TSs and Bases for the South Texas Project Electric l
Generating Station, Unit 1.
5.0 STATE CONSULTATION
f In accordance with the Commission's regulations, the Texas State official was notified of the proposed issuance of the amendment. The State official had no comments.
6.0 ENVIRONMENTAL CONSIDERATION
The amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and changes surveillance requirements. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration, and there has been no public comment on such finding (61 FR 7553). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.
i
7.0 CONCLUSION
The Commission has concluded, based on the considerations discussed above, i
that:
(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, i,
i
.~
3
. and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributor:
J. Tsao Date:
May 14,19%
Y l
l i
f W'