ML17194A439
| ML17194A439 | |
| Person / Time | |
|---|---|
| Site: | Dresden, Quad Cities, 05000000 |
| Issue date: | 01/07/1982 |
| From: | Rausch T COMMONWEALTH EDISON CO. |
| To: | James Keppler NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| Shared Package | |
| ML17194A438 | List: |
| References | |
| NUDOCS 8202010197 | |
| Download: ML17194A439 (8) | |
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....,ea Commonwe.Edison One First National Plaza, Chicago, Illinois Address Reply to: Post Office Box 767 Chicago. Illinois 60690 January 7,c 198 2 Mr. James G: Keppler, Regional Administrator Directorate of Inspection and Enforcement - Region III U.S. Nuclear Regulatory commission
.799 Roosevelt Road Glen Ellyn, IL
- 60137,
Subject:
Dresden Station Units 2 and 3 Quad Cities Station Units l and 2 Final Corrective Actions Associated with HPCI Steam Line Problems.
Inspecfion Report No~.50~237/81~25 NRC* Doeket *Nos~* 50.;_237 /249* an*d
- 50.;_254/265 References (a):
c: E: Norelius letter to Cordell Reed dated 0 c t 0 be r 8 ' l 9 81 (b):
L: O. DelGeorge letter to J: G. Keppler dated October 23, 1981 (c):
c:,E: Norelius letter to Cordell Reed dated December 2; 1981
Dear Mr. Keppler:
As committed i*n Reference (b); Commonwealth Edison has performed the analyses described in our response to items 2, 3 and 4 of Reference (a).
Concerning item 2, Sargent & Lundy:has reviewed the damaged high energy line break restraints and recommended modifications to Dresden Unit 2 which were implemented by the station:
Sargent &
Lundy determined that these hangers were located in a direction not deis~ned for in the original analysis critefia during the event.
There is no reason to re-evaluate the high _energy line break analysis.
The original design criteria and hanger calculations are available for NRC review in the Sargent & Lundy off ices.
The item 3 fatigue assessment evaluating the long-term operability of the HPCI system has been completed and is provided in Attachment A to this letter:
The overall conclusion of this study is that the pipe fatigue damage induced by the two water hammer events is insignificant and that the long-term operability of the HPCI System has not been degraded.
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JAN 1 t 198e
Regarding item 4, the !EB 79~14 documentation is available for review in the EDS Nuclear Inc: offices in Walnut Creek, calfifornia:
We understand that items 1, 3, and *4 w~re reviewed at the EDS. offices by Region III during the week of January 3, 1982:
Should copies of the documentation reviewed,be required for your use~ please contact this office:
In.Reference (b), we further indicated that the Quad Cities Unit 2 HPCI steam line would be inspected within 30 day~ of unit start-up:
This inspection was completed on December 28, 1981, with the reactor at pressure and the HPCI steam line isolation valves open~
th~ line was found to have a continual downward slope from the isolation valve to the drain pot, and the pipe* run from turtiine steam inlet valve M0~2~2301-3 to the drain pot was also sloped downward.
No further inspections are planned.
In summary, Commonwealth Edison has now completed all actions requested in Reference (a):
Please direct any questions you may have concerning this m~tter to this office:
Attachment very truly yours,
- J~(/(? /f." *.L_
Thomas J: Rausch Nuclear Licensing Administrator*
Boiling Water Reactors cc:
RIII Inspector - Dresden w/att.
RIII Inspector - Quad Cities w/o.att.
3186N
- 1.
INTRODUCTION ATTACHMENT A DRESDEN 2 HIGH PRESSURE COOLANT INJECTION SYSTEM FATIGUE ASSESSMENT The scope of work presented herein has been performed to evaluate the long-term operability of the Dresden 2 High Pressure Coolant Injection (HPCI) System. In particular, it addresses the NRC Con-cern on the effect of the recent water hammer event and the prior water hammer loading on the pipe fatigue life.
An equivalent static approach, similar to the one recommended by the NRC for the fatigue evaluation of the Dresden 2 Containment
- Cooling Service Water (CCSW) System, was adopted in this fatigue evaluation.
The intent of this approach is to provide a conservative estimate of the f ~tigue damage induced by the two water hammer.events in the region where failure of supports have been observed.
- 2.
EVALUATION OF.CAPACITY LOADS FOR FAILED SUPPORTS The supports observed.to have fa*iled were.identified to be support 2305-G-219, Sway Braces 2305-M-213 and 2305-M-220.
Capacity loads based on the observed failure modes of these supports were developed.
The methodology and" assumption used are described below:
- 1.
Support 2305-G-219 During the walkdown inspection, two of the four anchor bolts were found on the floor and the other two bolts were loose on the base plate.
However, no damage on the concrete was identified.
As illustrated by Figure 1, failure of this support may have been caused by either a vertical or a horizontal pipe reac~ion load.
Both cases were considered.
The capacity load based on the direct pullout of four anchor bolts, using the ultimate tension load, was determined to be 25,240 lbs, while that due to a horizontal pipe reaction load was found to be 9200 lbs.
2.
- 3.
Support 2305-M-213 During the walkdown inspection, no evidence of failure in the support was reported; later, however, the sway brace was found to be broken.
The minimum load required to fail the sway brace was determined based on the. rated ultimate load of.2400 lbs.
Support 2305-M-220 During the walkdown inspection, it was noted that the rod was bent and the weld on the rod-to-rod coupling was broken.
The ultimate capacity load for the sway brace was rated at 7200 lbs and the capacity load for the 1/8" fillet weld was calculated to be higher than 7200 lbs.
Since the rated capacity provided by the manufacturer represented the best available estimate of the ultimate capacity of the component, the capacity load of 7200 lbs was used for this sway brace.
- 3.
EVALUATION OF PIPE STRESSES The method of determining the pipe stresses which may have been induced in the system during the transierit event was based on an equivalent static approach, as recommended by the NRC for a similar fatigue evaluation on the Dresden 2 CCSW system.
Pipe stresses were determined by statically applying the capacity loads of the failed supports ( Support 2305-G-219, Sway Braces 2305-M-213 and 2305-M-220) at the locations and the direction of participation of the failed supports.
The pipe stresses due to the three failed supports were determined separately and then combined by the SRSS method {see assumption f.in item 4) to o.btain the total pipe stresses which.may have resulted from the transient water hammer loadings.
Since the pipe rupture restraint, Support 2305-G-219, was observed to have failed by pullout of the anchor bolts, it has been assumed that the direction of participation of this restraint is the vertical direction.
However, a closer review of the piping layout and possible water hammer loads in this area shows that the hori-zontal direction is a more probable direction of participation.
This is based on the fact ~hat in the region of interest, both the pipe routing and water hammer loads are contained in the horizontal plane and therefore significant vertical responses for
the pipe are not expected.
Either direction of participation would cause an anchor bolt pullout failure mode for the rupture restraint, as illustrated in Figure 1.
Both directions of partici-pation were considered in this evaluation.
It is intended that the vertical direction would provide a conservative estimate of the pipe stress due to the water hammer loads in the region of interest,_
while the horizontal direction would provide,a more realistic estimate.
Pipe stresses due to the design pressure and temperature of the system weie also calculated.
The total pipe stresses to be ~sed in the fatigue evaluation were obtained as the Combined stresses due to the water hammer, thermal and pressure loadings.
- 4.
FATIGUE ASSESSMENT The following major assumptions were used in the fatigue evaluation of the HPCI system:
- a.
The Code of Record for the IE *79-14 work scope, ANSI B31.1 Code, does not provide explicit rules for performing a fatigue evaluation.
The 1971 ASME Section III Code, Subsection NB was considered an appropriate method.and was.used in the evaluation.
- b.
Since drainage improvements have been made on the system to pre-vent future occurrences, only two occurrences of this event were considered.
Also, the water hammer loads from the recent and prior events are assumed.to be :similar.
- c.
In determining fatigue damage, only water hammer stress cycles which were higher than one-third (33%) of the maximum water hammer stress were considered.
Stress cycles with lower magni-tudes had insignificant fatigue damage contribution.
- d.
The number of significant stress cycles (stresses above 33%
of the maximum stress level) was determined as the number of cycles for the stress level to decay from the maximum stress level to a 33% level of the maximum stress.
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- e.
A 2% damping was used for the piping system. Due to the high pipe stress level induced by the water hammer loads, it was considered that the SSE damping level of 2%, recommended in Regulatory Guide 1.61? would be appropriate.
- f.
The pipe stresses obtained separately for the three support capacity loads were combined by the SRSS method.
We believe the SRSS method is justified based on the following reasons:
.i)
It is very unlikely that the three supports would have failed (or reached their respective capacity loads) at ii) the same time, since the magnitudes of the capacity loads are quite different.
We believe that they failed at different times due to a "domino" effect of failing ~upports.
The rupture restraint has a large gap.and 'therefore did I
not immediately participate as a pipe support until the gap had been closed.
On the other hand, the sway braces participated immediately as pipe supports.
Therefore 1 it may be expected that these supports did not reach their.
maximum loads simultaneously.
- 5.
RESULTS The fatigue assessment p~rformed showed that the pipe fatigue damage induced by the two occurrences o*f the water hammer transient events is as 'follows:
Case I (Vertical direction of participation for Support 2305-G-219) : Usage Factor = 0.120 Case II (Horizontal direction of participation for Support 2305-G-219) : Usage Factor = 0.003
. 6.
. CONCLU STION Piping stresses as a result of the water hammer event were con-servatively approximated by a static analysis of the system, using the support failure loads.
These stresses were then used to conservatively estimate the pipe fatigue damage due to the two water hammer events which have occurred.
The results of this study show that the fatigue damage induced by the water hammer loads in the region where support failure had occurred was insignificant.
While an acceptable fatigue usage factor of 0.120 was provided by the conservative Case I Analysis, a fatigue usage factor of 0.003 was obtained from the more realistic approach used in the Case II Analysis.
Therefore, it is concluded that the long-term op~rability of the HPCI System has not been degraded by these two occurrences of the water hammer loading.
VERTICAL DIRECTION Anchor Bolt Reaction Forces HORIZONTAL DIRECTION FIGURE 1:
FAILURE MODES OF SUPPORT 2305-G-219