ML20247Q729
| ML20247Q729 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 05/31/1989 |
| From: | Fay C WISCONSIN ELECTRIC POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| CON-NRC-89-064, CON-NRC-89-64 VPNPD-89-314, NUDOCS 8906070024 | |
| Download: ML20247Q729 (10) | |
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h!S 6L IMsconsin Electnc poecoum 231 W. M!CHIG AN, P.o. BOX 2046, MILWAUKEE. WI 53201 ou) 221-2345 VPNPD-89-314 NRC-89-064 Y
May 31,.J1989 1
P iU.S.JNUCLEAR REGULATORY COMMISSION V
Document' Control Desk E
Mail-Station P1-137 f
Washington, D.C. 20555 tGentlemen:-
DOCKET NOS. 50-266 AND'50-301 LSUBMITTAL OF JUSTIFICATION FOR CONTINUED OPERATION REGARDING PRESSURIZED SURGE LINE IMERMAL STRATIFICATION POINT BEACH NUCLEAR PLANT UNITS <1 AND 2 m
. Please find attached Wisconsin Electric (WE) Power Company's L,
- cubmittal'of a Justification of Continued Operation (JCO) 4 regardingl thermal-stratification in the pressurizer surge line for Point Beach Nuclear Plant (PBNP) Units 1 and 2, as required by! Item 1.b_of IE Bulletin 88-11.
I WE has' elected to satisfy the requirements set forth in the bulletin by-participating in a program set up by the Westinghouse owners Group' (WOG) to address tus problem on a generic baels.
The-details of this generic detailed analysis were discussed in.
a meeting between the NRC and the owners Group on April 11, 1989.
It was decided at thatLtime that the generic detailed analysis would be' submitted to the NRC under Westinghouse Topical Reports (WCAP)'12277 and 12278 on June.15, 1989, and the member utilities would submit the required JCos' (per Item 1.b.of the bulletin) on May 31, 1989 (referencing the forthcoming WCAPs).
The' attached JC0 for PBNP' demonstrates the' applicability of the-WOG generic detailed analysis and, therefore, continued full-4 U
power operation including at least; ten (10)' additional heatup 7;
-cooldown cycles.
WE commits to addressing the requirements set
.forth.in Item 1.d of IE Bulletin 88-11 by January 1991.
lo This completes Section Item 1.b requiren.ents of IEB 88-11 for J$@!
PBNP.
If you have any questions or' comments, please contact us.
10 0y very truly yours, G
d s
88 6{&
ll R8 j
C.
W.- Fay or
-Vice President
$@o; Nuclear Power
)j8'l Attachment A subsidhtr ofMiscomin Enew Corpmtha Copy to'NRC Regional: Administrator, Regiop III.8 Resident Inspector
_______-1-___-_
JUSTIFICATION FOR CONTINUED OPERATION REGARDING PRESSURIZER SURGE LINE STRATIFICATION POINT BEACE NUCLEAR PLANT
. BACKGROUND It was first reported in INPO SER 25-87 that temperature measurements at a German PWR indicated thermal transients different than design.
Recent measurements at several domestic PWRs have indicated that the temperature difference between the pressurizer and the hot leg results in stratified flow in the surge line, with the top of the flow stream being hot (pressurizer temperature) and the bottom being colder (hot leg temperature).
The top-to-bottom temperature difference can reach 250 F to 300 F in certain modes of operation, particularly Modes 3, 4, or 5 during heatup and cooldown.
Surge line stratification causes two effects:
Global bending of the pipe is different than that predicted in the original design.
Fatigue life of the piping could be reduced due to the global and-local stresses from stratification and striping.
-More recently, the NRC has issued Bulletin 88-11 " Pressurizer Surge Line Thermal Stratification," December 20, 1988, identifying actions to be taken by licensees.
a)
Conduct visual inspection - walkdown b)
Update stress and fatigue analysis to account for stratification and striping c)
Obtain monitoring data, as necessary The bulletin encourages licensees to perform actions b) and c) above through collective efforts with other plants.
In October 1988, Wisconsin Electric (WE) and other members of the Westinghouse Owners Group (WOG) authorized a program to perform a generic evaluation of surge line stratification in Westinghouse PWRs that will address portions of Bulletin 88-11.
The WOG program is designed to benefit from the experience gained in the performance of several plant specific analyses on westinghouse PWR surge lines.
These detailed analyses included definition of revised thermal transients (including stratification) and evaluations of pipe stress, fatigue usage factor, thermal striping, fatigue crack growth, leak-before-break, and support loads.
The ove;all analytical approach used in all of these analyses has been consistent and has been reviewed, in detail, by the NRC staff.
As of March 1989, plant specific analyses have been performed on..
l.
l five domestic Westinghouse PWRs.
In addition, twelve Westinghouse plants have completed or are currently performing an interim evaluation of surge line stratification which includes finite element structural analysis of their specific configuration under stratified loading conditions.
WOG Program Status As part of the current WOG Program, surge line physical and i
operating data has been collected and summarized for all domestic Westinghouse PWRs (55 units).
Information relating to piping layout, supports and restraints, components, size, material, operating history, etc., has been obtained.
This data has been evaluated in con] unction with available monitoring data and plant specific analyses performed by Westinghouse.
The results of this evaluation were presented to the NRC in a meeting on April ll, 1989.
The evaluation is being formalized into a Westinghouse i
topical report (WCAP-12277, Proprietary and WCAP-12278, Non-proprietary version) scheduled for submittal to the NRC on June 15, 1989.
^
This topical report forms the basis for the following justification for continued operation.
JUSTIFICATION FOR CONE dUED OPERATION A.
Stratification Severity Thermal stratification (AT > 100 F) has been measured on all surge lines for which monitoring has been performed and which have been reviewed by the WOG to date (eight surge lines).
The amount of stratification measured and its variation with time (cycling) varies.
This variation has been conservatively enveloped and applicability of these enveloping transients has been demonstrated for plant specific analyses.
Various surge line design parameters were tabulated for each plant.
From this, four parameters judged to be relatively significant were identified.
A.
Pipe inside diameter B.
Piping slope (average)
C.
Entrance angle of hot leg nozzle D.
Presence of mid-line vertical riser These parameters were usec in a grouping evaluation which resulted in the definition of 10 monitoring groups corresponding to various combinations of these parameters at Westinghouse PWRs.
Approxi-mately 40% of the plants fall into one group for which a large amount of monitoring data has already been received and for which the enveloping thermal transients, discussed above, are applicable.
The remaining 60% of Westinghouse PWRs are divided among tae _
nine other additional groups.
Although representative monitoring data has not yet been received from all these groups, in general, the combination of significant parameters of these nine groups is expected to decrease the severity of stratification below that of the enveloping transients.
This conclusion is a]so supported by a comparison of available monitoring data.
B.
Structural Effects Significant parameters which can influence the structural effects of stratification are:
a.
Location and design of rigid supports and pipe whip restraints b.
Pipe layout geometry and size i
i c.
Type and location of piping components Although the material and fabrication techniques for Westinghouse surge lines are reasonably consistent and of high quality, the design parameters listed above vary among Westinghouse PWRs.
This variation in design is primarily a result of plant specific routing requirements.
A preliminary evaluation, comparing the ranges of these parameters to those of plants for which plant-specific analyses and interim evaluations are available (approximately 20% of Westinghouse PWRs),
has been performed.
This comparison indicates a high degree of confidence that, from a combined transient severity and structural effects stand 7oint, the worst configuration has most likely been evaluated.
This conclusion is supported by plant-specific analyses covering five plants and interim evaluations of six additional plants (interim evaluation is in progress on six more plants as of March 1989).
These analyses and evaluations have included various piping layouts, piae sizes, support and restraint designs and piping components.
Altaough the full range of vari-ation in these parameters has not been evaluated, experience gained from these evaluations indicates that further evaluations will not result in a more limiting configuration than those already evaluated.
C.
Operating Procedures The WOG currently has available the surveys of operating pro-cedures performed in support of existing plant-specific analyses.
Experience indicates that heatup and cooldown procedures have a significant effect on stratification in the surge line.
All conclusions reached by the WOG to date have assumed a steam bubble mode heatup and cooldown procedure which may result in a temperature difference between the pressurizer and reactor coolant hot leg of more than 300 F.
In many 'ases, individual system (RCS)ing procedures and technical specifica 'ons provide plant operat limits on this value.
WE uses a solid water heatup procedure i
that heats the RCS and pressurizer at the same temperature until approximately 400 F when the steam bubble is finally drawn.
1 _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
1 It is also known that some procedures utilize nitrogen, during at least part of the heatu?/cooldown cycle, as a means of providing a pressure absorbing space in the pressurizer.
Based on information currently available to the WOG, a high confidence exists that the steam bubble mode heatup, assumed to date, is conservative with respect to Westinghouse PWRs.
D.
Pipe Stress and Remaining Life The design codes for surge line piping have requirements for checking pipe stress limits and the effects of fatigue loadings.
These stress limits provide a means of controlling stress from primary loads such as pressure, deadweight, and design mechanical loading, as well as stress from secondary loads such as thermal and anchor motion effects.
Stratification in the surge line is a secondary load which will only affect the qualification of secondary stresses.
The qualification of primary stresses is not affected by this loading.
Secondary stresses are controlled to prevent excessive displace-ments and gross plasticity and to prevent excessive fatigue loadings in the pipe.
The basic characteristic of a secondary thus, a failure from a single stress is that it is self limiting;is not expected.
application of a secondary loading For the stratification issue, the potential effects of excessive displacements have been investigated through a detailed visual observation of the surge line during the walkdown required per Bulletin 88-11 action 1.a.
Prior to the issuance of Bulletin 88-11, WE visually inspected the Point Beach Nuclear Plant (PBNP)
Unit 2 surge line piping and support geometry during the fall 1988 refueling outage, and instrumented the surge line at several locations in order to assess the degree of thermal stratification, if any.
Although the inspection was not a VT-3 inspection, it was conducted by individuals experienced in piping stress analysis and inservice inspection.
No signs of piping or support distress were observed during the inspection.
The piping is supported by spring hangers only and is provided with pipe whip restraints.
The continuous insulation jackets between the pipe whip restraint hardware and the piping showed no signs of either scuffing or denting, indicating that no contact between the piping and the whip restraints has occurred.
Sufficient gap is available to permit the piping to expand routing.y without encountering restraint at any point on the thermall
. Temperature data has been collected during heatup and at steady-state full power operation.
Preliminary evaluation of the data indicates that stratification occurs in the Unit 2 surge line, however, the magnitude of the temperature data collected is in question.
Even though it is believed that the thermocouple in - _ - _ _ _ _ _ - _ - _ - _
place on Unit 2 are not providing accurate temperature readings and that the magnitudes are not correct, the data shows that stratification occurs to some degree.
To alleviate the concern about the questionable magnitude of the thermocouple readings on Unit 2, WE decided to instrument the Unit 1 pressurize surge line for temperature monitoring.
WE is currently collecting data on Unit 1 at steady state full power and is seeing stratification on the order of 20 F to 40 F.
This is consistent with plant specific data collected by Westinghouse to date.
WE also completed the inspection of the PBNP Unit 1 pressurizer surge line in accordance with the requirements of IE Bulletin 68-11.
A VT-3 inspection was performed on April 3, 1989 with the system in a hot condition (approximately 500 F) and on April 28, 1989 with the system at ambient conditions.
The following were observed:
1.
During the hot inspection, it was observed that the surge line pipe collar (Reference Attachment 1, Drawing M-400, Sh. 5, Detail 17) was in apparent contact with the pipe whip restraint R-4 at the 270 point indicating a conventional uniform thermal expansion problem unrelated to thermal stratification.
No visual deformation of the restraint structure, collar, lugs, or piping was observed.
2.
During the cold inspection, two small cracks were observed in the fillet welds between the pipe collar and the top of the lug (position 0 ) at the north end of whip restraint R-5.
One crack, approximately 3/8" long and 3/16" deep, was located on the west side of the lug.
The other crack was located on the east side of the lug and was approximately 1/8" long and 1/8" deep.
Both indications were ground out and verified as removed by liquid penetrant examination.
The design of these attachents welds was specified as a 1/4" fillet for 1" in length (see Attachment 1).
- However, the inspection revealed that a 1/4" fillet weld was placed along the full length of the lug to the pipe collar Joint (8").
The integrity of the joint has not been compromised by the removal of these indications and the restraints were operable with the cracks considering the excess amount of weld material.
3.
The piping and all other support installations did not show any discernable distress or damage.
In order to eliminate any potential contact between the surge line collars and the whip restraint structural hardware due to thermal expansion during future operating cycles, WE has removed the portion of the collars at locations R-4 and R-5 on the side opposite the steam generator (SG) during the outage.
This is acceptable since, according to the original design, only the side _ _ _ _ - _ _ _ _ _ _ _ _ _ _ -
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facing the SG is necessary to protect the.SG from pipe whip.
After removal of the collar pieces, a visual examination of the i
collar and support hardware was performed to see if any discernable galling existed on either surface, which would confirm significant contact between the surfaces.
No viuual evidence of galling was present.
Measurements of the actual surge line displacements at locations R-4 and R-5 were made during plant heatup to confirm whether or not actual collar / restraint contact has occurred during plant operation.
The surge line displacements measured with the RCS system at 525 F and the surge line at 630 F indicate that a gap closure problem has existed during plant operation, and may have existed since plant startup.
Therefore an analysis was performed on the Unit 1 pressurizer surge line to include the effects of the gap closure on pipe stress and fatigue.
The results of Sargent and Lundy Welded Attachment analysis (reference Sargent and Lundy Report entitled " Pressurizer Surge Line Thermal Interference Evaluation," accession number EMD-065696, Rev. 00) yield pipe stresses and loads below their respective code allowables, and a cummulative usage factor of less then 0.02.
Consequently, based on the analysis results indicated above and on the fact that there are no other interferences, rigid supports, or visual piping / support distress in the piping system, PBNP Unit 1 falls under the WOG generic detailed analysis results for thermal stratification.
Even with the question 21e magnitude of the data collected on Unit 2, WE believes that the Unit 2 pressurizer surge line falls under the WOG generic detailed analysis because of sufficient pipe whip restraint gaps, being supported only by spring hangers, solid water heatup, and finding no visual evidence of contact on the insulation jacket.
The effects of secondary stresses on the remaining life of the surge line have been evaluated on a generic basis through the WOG program.
The following summarizes the results of this evaluation.
All plant specific analyses performed as of March, 1989 has demonstrated compliance with applicable ASME Codes and a surge l
line fatigue life in excess of a 40 year plant life.
Review of plant specific fatigue calculations indicate that the surge line fatigue life is primarily dependent on the number of heatup and cooldown cycles, rather than years of operation.
Considering the worst case years of operation (28.5 yr.) in combination with the worst case number of heatup-cooldown cycles (75, at a different plant) at any Westinghouse PWR, and assuming a 40 year life for all surge lines, it is estimated that no more than approximately 50% of the fatigue life has been used at any Westinghouse plant to date. 1
3
-For a design life considering 200 heatupicate approximately 100cooldown cycl plant specific analyses), this would ind remaining cycles.
This number of remaining cycles far exceeds the postulated worst case number for the time frame needed to resolve the stratification issue.
E.
Leak Before Break All the plant specific analyses performed to date that have included the loadings due to stratification and striping have validated the " leak-before-break" concept and have substantiated a 40-year plant life.
Fatigue crack growth calculations, performed as part of these plant specific analyses, have demonstrated that any undiscovered crack as large as 10% of the wall thickness would not grow to cause leakage within a 40 year plant life.
Nevertheless, any postulated through wall crack propagation would most likely result in " leak-before-break" and thus permitia safe and orderly shutdown.
F.
Inspection History The NDE inspection history at WE, as well as all other domestic Westinghouse designed PWRs, has not revealed any service induced degradation in the surge line piping that has been attributed to thermal stratification.
Summary of Conclusions From WOG Program Eased on information assembled on surge lines for all domestic Westinghouse PWRs, and evaluation of that information in ceniunction with plant-specific and other interim evaluation results, the WOG concludes that:
A high degree of confidence exists that further evaluation will confirm that the worst combination has already been evaluated for stratification severity, structural effectr, and operating procedures.
All plant specific analyses, to date, have demonstrated a 40 year life of the surge line.
Assuming that further evaluation leads to the same conclusion for the reamining Westinghouse PWRs, the worst case remaining life is approximately 100 heatup-cooldown cycles.
Through wall crack propagation is highly unlikely, however, " leak-before-3reakd would permit a safe and orderly shutdown if a through wall leak should develop.
NDE inspection history demonstrates the present day integrity of Westinghouse PNO pressurizer surge lines.
While additional monitoring, analyses, and surveys of operating procedures are expected to further sub- _ _ _ _ _ _ _ _
stantiate the above conclusions, the presently available information on surge line stratification indicates that Westinghouse PWRs may be safely operated while additional data is.obtained.
Overall Conclusion-Based'on the above discussions, WE believes it is acceptable for PBNP to continue power operation for at least ten additional heatup-cooldown cycles.
WE has committed to address the requirements of Bulletin 88-11 by January, 1991.
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