ML20151L878
| ML20151L878 | |
| Person / Time | |
|---|---|
| Site: | Trojan File:Portland General Electric icon.png |
| Issue date: | 04/15/1988 |
| From: | Cockfield D PORTLAND GENERAL ELECTRIC CO. |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| References | |
| TAC-65471, NUDOCS 8804220173 | |
| Download: ML20151L878 (8) | |
Text
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~ Portland General Bectric Company
_E David W. Cockfield Vice President. Nuclear April 15, 1988 Trojan Nuclear Plant Docket 50-344 License NPF-1 U.S. Nuclear Regulatory Commission ATTN:
Document Control Desk Washington DC 20555
Dear Sir:
Main Feedwater Restraint Failure Safety Evaluation Open Items On July 27, 1987, Portland Cenoral Electric Company (PGE) submitted a report describing the evaluation and root cause for the failure of Main Feedwater System Restraint SR-8.
On November 11, 1987, the Nuclear Regulatory Commission (NRC) issued a safety evaluation to PGE for this issue. The safety evaluation identified six open items / recommendations.
During the NRC audit of our long-term Pipo Support Design Verification Program in January 1988, the reviewers discussed the SR-8 open items /
recommendations with us.
As a result of those discussions, PGE committed to provide additional information by letter to the NRC.
The requested information is attached.
Sinecroly, Attachments c:
Mr. John B. Martin Regional Administrator, Region V U.S. Nuclear Regulatory Commission Mr. William Dixon Stato of Oregon Department of Energy Mr. R. C. Bare NRC Resident inspector Trojan Nuclear Plant h0 8804220173 880415
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Trojan Nuclear Plant Document Control Desk Docket 50-344 April 15. 1988 License NPF-1 Attachment A Page 1 of 3 MAIN FEEDWATER RESTRAINT FAILURE SAFETY EVALUATION REPORT OPEN ITEMS / RECOMMENDATIONS The Nuclear Regulatory Commission (NRC) letter of November 11, 1987 contained a safety evaluation for the f ailure of main feedwater restraint SR-8.
The report contained six open items / recommendations. PGE's response to these open items / recommendations is as follows:
1.
NRC Open Item The licensee should confirm that the failure of restraint SR-8 has not reduced the fatigue life of the feedwater piping to unacceptable icvels for continuing service.
PGE Response A fatigue review / estimate of the feedwater piping allowable cycles was performed based on stresses induced during a water hammer event caused by a 500 psi pressure pulse. The calculation performed for this evaluation was reviewed by the NRC during the January 19-21, 1988 audit and was found to be acceptable. The calculation demonstrated the fatigue life of the piping was not adversely affected.
2.
NRC Open Item The licensee should confirm consistency between the estimated failure load of SR-8 and the extent of damage observed.
PGE Response The NRC reviewed Attachment A of our June 12, 1987 letter to the NRC during the January 19-21 audit of the Pipe Support Design Verification Program.
The information in that attachment addressed this NRC concern and was satisfactory to the auditors for resolving this issue.
3.
NRC Open Item The licensco should confirm the proposed feedwater check valve maintenance and testing program is not inconsistent with the IST program for all valves in the Trojan Plant.
PGR Response On June 16, 1987, PGE forwarded a report to the NRC documenting the evaluation of the SR-8 failure performed by the Nuclear Plant Engineering Department. This report contained recommended corrective actions, including proposed naintenance testing of feedwater check valves.
The recommended
J Trojan Nucicar Plant Document Control Desk Docket 50-344 April 15, 1988 License NPF-1 Attachment A Page 2 of 3 testing consisted of leak testing the main feedwater check valves for each steam generator and auxiliary feedwater check valves each year and inspec-tion by disassembly of these same valves once overy two years. The leak testing was recommended to be performed during Plant startup from refueling outages.
In response to these recommendations, test and/or maintenance instructions are being developed for eight valves. The valves include the four main feedwater check valves in the branch lines to the four steam generators (FW 2017, 2018, 2019, and 2020) and the four first-off auxiliary foodwater check valves from the main feedwater lines (FW 2013, 2014, 2015, and 2016).
The periodicity of the testing recommended in the June 16, 1987 report, however, is not consistent with testing prescribed for similar valves by the Inservice Testing (IST) Program.
As such, the periodicity of testing is being changed to every two years and will consist of either a leak test or valve disassembly for inspection. The leak testing instructions will include provisions for performing the testing during shutdown or startup from the refueling outages.
4.
NRC Recorpendation The staff feels the addition of some accelerometers mounted in locations where peak accelerations would be expected would greatly enhance the monitoring effort.
PGE Responso PGE has evaluated the use of accelerometers in the monitoring program and has concluded their use would not be beneficial.
Our consultant has indi-cated accelerometers were used without success at another Plant. The water hammer events are of such short duration that by the time the accelerometer is triggered to record, the initial pipe movement would be missed. Continuous recording of the accelerometer is not practical.
5.
NRC Recommendation The staff recommends the slug motion analyses table be expanded to include columns listing estimates of pipe displacements at the hanger locations and pipe strenses at key locations.
PGE Response The data to expand the slug motion analyses table are available from the analyses, but there is little benefit in extracting the information at this time. The slug motion analyses table is specific to SR-8 and would have to be significantly revised to incorporate the displacement and stress data.
In the event of another water hammer, the displacement and
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Trojan Nuclear Plant
' Document Control Desk Decket 50-344 April 15, 1988 License NPF-1 Attachment A Page 3 of 3 stress data could be extracted and compared to data collected from the water hammer to corroborate the event. This recommendation will be reevaluated in such a case.
6.
NRC Open Item The staff has reviewed the Impell report and concurs with its findings and recommendations. The staff requires that the licenseo comply with all the Impell proposed actions.
PGE Response All of the Impell-proposed actions have been or are in the process of being complied with.
Specific responses to each of the Impell-proposed actions are provided as Attachment B.
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Trojan Nuclear Plant Document Control Desk Docket 50-344 April 15, 1988 License NPF-1 Attachment B Page 1 of 4 RESPONSE TO IMPELL-PROPOSED ACTIONS 1.
Impell-Proposed Action Finalize the evaluations of water hammer loadings due to fast valve closure events, and include this analysis in the documentation of the resolution to this issue.
Portland General Electric Company (PGE) Response Bechtel Corporation has completed analysis calculating the ef fects on the feedwater system due to various valve closure events
(
Reference:
Letter BP-13096, dated September 18, 1987).
2.
Impell-Proposed Action The equivalent static failure load of 40 kips calculated for SR-8 should be viewed as an upper bound.
PGE Response PGE agrees, no action required.
3a.
Impell-Proposed Action Clarify to Bechtel whether the entire seismic restraint SR-8 or only the anchorages should be qualified for 40 kips.
If only the anchor-ages, identify whether the 40-kip load is to be applied as only a pullout force or some combination of pullout, shear and moment.
PGE Response The design load for SR-8 continues to be based on dead weight, thermal and seismic conditions.
Even though water hammer has occurred in the past, corrective action to minimize the possibility of water hammer hns been taken so that restraint design loads need not include this load case. Thermal stratification could contribute to restraint loading but will only be included as a design load case if data collection confirms that stratification is a significant contributor to restraint loads.
See Resnonses Ac and 6 below for further discussion on this matter.
The previous discussion on SR-8 capacity in the range of 40 kips was presented, not to suggest that this is the "design load" for the support, but rather to demonstrate that significant margin exists to accommodate unusual loads from water hammer and stratification above design loads. The current design load for SR-8 is less than
Trojan Nuclear Plant Document Control Desk Docket 50-344 April 15, 1988 License NPF-1 Attachment B Page 2 of 4 10 kips. The strut currently installed in SR-8 has a maximum recommended design load of 15.7 kips with a significant safety factor to failure. The Level D faulted load for the strut at 350*F is 26.2 kips. The suitability of support SR-8 to withstand incurred loads in the as-built condition is also being reviewed as a part of the current large-bore Pipe Support Design Verification Program.
3b.
Impell-Proposed Action Officially convey to Bechtel additional details of modified support SR-8.
For example, indicate that the 5/8-inch-diameter threaded rods are grouted into the 1-inch-diameter holes through the concrete deck. This information should be incorporated into the t
calculation to close out the support evaluation.
PGE Response As-built details of SR-8 have been forwarded to Bechtel. The design calculations have been completed as part of the large-boro Pipe Support Design Verification Program.
Based on the as-built informa-l tion; the SR-8 configuration is adequate to withstand design loads..
4a.
Impell-Proposed Action The original 1970 stress analysis incorrectly modeled SR-8 as a two-way XY stop located on the horizontal portion of the elbow, while the 1975 stress analysis thermal runs did not include SR-8, even though this is an active support under thermal loads.
PGE Response Bechtel has annotated the 1970 calculation as being superseded by the 1975 calculation and has annotated the 1975 calculation as being superseded by the June 1987 calculation. The 1987 calculation correctly models SR-8.
i Ab.
Impe11-Propcsed Action j
Loop B was qualified by comparison to analyses of the Loop A piping, which is generally a mirror image to Loop B.
Perform separate 4
seismic analyses to account for the different locations of three snubbers in Loop B.
The relative snubber locations differ by up to 5 feet, and Snubber SS-2 of Loop A is located on a horizontal run of piping, while the comparable Snubber, SS-6, of Loop B, is on a riser.
i PGE Response Bechtoi has completed a separate stress analysis calculation for the "B" loop.
Reference:
Bechtel Calculations 2-16 Rev. 2, for "A" loop and 2-21, Rev. O, for "B" loop.
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Trojan Nuclear Plant Document Control Desk Docket 50-344 April 15, 1988 License NPF-1 Attachment B Page 3 of 4 Ac.
Impe11-Proposed Action If the thermal stratification calculation is to become part of the permanent Plant record, correct the discrepancy in which the temperature difference at the two horizontal elbows of thn model is applied in the horizontal, rather than the vertical direction.
PGE Response System monitoring to be performed over the 1987-1988 operating cycle will include collection of data needed to assess the degree, if any, of thermal stratification.
If this data leads PGE to conclude that thermal stratification contributes significantly to the system design basis, the Bechtel calculation will be incorporated (including resolution of the Impell-noted discrepancy) au part of the system design basis documentation for SR-8 design loadings.
4d.
Impe11-Proposed Action In the calculations for requalifying the pipe supports, use the seismic loads from the latest analyses.
Alternatively, note in the calculations that they are conservative, since the original seismic loads were used rather than thn loads from the latest analyses, which are smaller.
Also, note in the SR-8 qualification calculation that the old seismic load, which was 721 pounds larger than the new seismic load, was used. This offsets the fact that the dead weight reaction of 541 pounds at Support SR-8 was not considered.
PGE Response The revised Bechtel calculation t'ur the "A" loop and the new cal-culation for the "B" loop utilize the current seismic response and the dead weight loads for SR-8 (reference Bechtel calculation number from Response Ab. above).
5.
Impe11-Proposed Action The steam bubbio collapse analysis was performed using the dynamic slug model described in NUREG-0291, "An Evaluation of PWR Steam Generator Water llammer". The analysis assumed two symmetrical slugs on each vide of the bubbla moving toward each other and included sensitivity studies for the bubble volume, pressura differenco, and clug lengt. Complete the analysis, including:
a.
Sensitivity of the :alculated loads te variations in the height of the bubble (ie, void fraction at the location of the bubble),
and b.
A study of the situation where a singic slug on the steam generator side will be accelerated toward the bubble.
r Trojan Nuclear. Plant Document Control Desk Docket 50-344 April 15, 1988 License NPF-1 Attachment B Page 4 of 4 PGE Response PGE agrees with the Impell recommendation.
Bechtel has incorporated this recommendation'in the analysis and the results and conclusions did not chango.
Reference:
Bechtel Letter BP-13032, dated August 3, 1987.
6.
Impell-Proposed Action The thermal stratification calculation is based on a unit "load" of a.100'F temperature difference across the piping cross-section.
At this temperature difference, a 5.4-kip force acts at SR-8.
Due to
- the large difference between the main feedwater and auxiliary feed-water temperatures, variations larger than 100*F are possible, how-ever. Under this situation, relatively high loads at SR-8 could occur. It would then not take 1,significant hydro-dynamic-type event. to overload the support.
Examine this issue in more detall over the next fuel cycle to determine the range of temperature differences the piping actually experiences under Plant operation. Depending on the magnitude of the resulting loads, decide whether this condition should be incorporated into the design basis.
PGE Response As noted in our response to Recommendation oc above, the impact of thermal stratification on SR-4 and SR-8 loading will be evaluated as operating data are collected.
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