ML17325B194
| ML17325B194 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 05/31/1989 |
| From: | Alexich M INDIANA MICHIGAN POWER CO. (FORMERLY INDIANA & MICHIG |
| To: | Murley T NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| AEP:NRC:1086A, IEB-88-011, IEB-88-11, NUDOCS 8906070242 | |
| Download: ML17325B194 (17) | |
Text
gc Cp~ gATED D3FTKBU'EON DEMONSTRAT105 SYFI'EM REGULATOZeZNEORMATION DISTRIBUTION +TEM (RIDE)
ACCESSION NBR:8906070242 DOC.DATE: 89/05/31 NOTARIZED:
YES DOCKET FAClL:50-315 Donald C.
Cook Nuclear Power Plant, Unit 1, Indiana 05000315 50-316 Donald C.
Cook Nuclear Power, Plant, Unit 2, Indiana a
05000316 AUTH.NAME AUTHOR AFFILIATION ALEXICH,M.P.
Indiana Michigan Power Co. (formerly Indiana a.Michigan Ele RECIP.NAME RECIPIENT AFFILIATION Document Control Branch (Document Control Desk)
SUBJECT:
Submits justification for continued operation re pressurizer surge line stratification.
DISTRIBUTION CODE:
AOOID COPIES RECEIVED:LTR I
ENCL I
SIZE:
TITLE: OR Submittal: General Distribution NOTES:
RECIP1ENT ID CODE/NAME PD3-1 LA STANG,J
~ INTERNAL: NRR/DEST/ADS 7E NRR/DEST/ESB 8D NRR/DEST/MTB 9H NRR/DOEA/TSB 11 OC/LFMB EXTERNAL: LPDR NSIC COPIES LTTR ENCL 1
1 5
5 1
1 1
1 1
1 1
1 1
0 1
1 1
1 1
1 RECIPIENT ID CODE/NAME PD3-1 PD NRR/DEST/CEB 8H NRR/DEST/ICSB NRR/DEST/RSB 8E NUDOCS-ABSTRACT OGC/HDS1 RES/DSIR/EIB NRC PDR COP1ES LTTR ENCL 1
1 1
1 NOZE 'IO ALL RIZG" REXXPIEFIS:
PZEPBE HELP US 'I0 REDUCE KEBAB.
CXRZACZ 'IHE DOCUM1&Z CXÃZBOL DESKi RXR Pl-37 (EXT. 20079)
KO ELIMINATERKR SAME HKH DISTRIBUTXCN LISTS FOR DOCUMENZS KRJ DOHIT NEEDf TOTAL NUMBER OF COPIES REQUIRED:
LTTR 22 ENCL 20
'h
'I l
indiana Michigan Power Company P.O. Box 16631 Columbus, OH 43216 AEP:NRC:1086A Donald C.
Cook Nuclear Plant Units 1 and 2
Docket Nos.
50-315 and 50-316 License Nos.
DPR-58 and DPR-74 JUSTIFICATION FOR CONTINUED OPERATION REGARDING PRESSURIZER SURGE LINE STRATIFICATION U.
S. Nuclear Regulatory Commission Attn:
Document Control Desk Washington, D.
C.
20555 Attn:
T. E. Murley May 31, 1989
Dear Dr. Murley:
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).
0 0
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:
o Global bending of the pipe is different than that predicted in the original design.
o 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," dated 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 890b07024Z 890531 PDR ADOCK 050003' FNu
~ C<<w
~
t '
k~
~
~
Dr. T.
E. Murley AEP:NRC'1086A 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, Indiana Michigan Power Company (I&M) 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
'he 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 overall 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 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 Pro ram Status As part of the current WOG program, surge line physical and operating data have 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 conjunction 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 11, 1989.
The evaluation is being formalized into a Westinghouse 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.
Dr. T.
E. Murley AEP:NRC:1086A JUSTIFICATION FOR CONTINUED OPERATION A.
Stratification Severit Thermal stratification (delta T ) 100 F) has been measured on 0
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 conserva-tively 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 fudged to be relatively significant were identified.
1.
Pipe inside diameter 2.
Piping slope (average) 3.
Entrance angle of hot leg nozzle 4.
Presence of mid-line vertical riser These parameters were used in a grouping evaluation which resulted in the definition of 10 monitoring groups corresponding to various combinations of these parameters at Westinghouse PWRs.
Approximately 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 the other nine additional groups.
Although monitoring data has not yet been received representative of 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 also supported by a comparison of available monitoring data.
B.
Structural Effects Significant parameters which can influence the structural effects of stratification are:
1.
Location and design of rigid supports and pipe whip restraints
>> ~
)
Dr. T.
E. Murley AEP:NRC:1086A 2.
Pipe layout geometry and size 3.
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' preliminary evaluation, comparing the ranges of these parameters to those of plants for which plant-specific analysis,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 standpoint, 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, pipe sizes, support and restraint designs and piping components.
Although the full range of variation 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.
0 eratin Procedures The WOG currently has available the surveys of operating procedures 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 system (RCS) hot leg of more 0
than 300 F.
In many cases, individual plant operating procedures and technical specifications provide limits on this value. It is also known that some procedures utilize nitrogen, during at least part of the heatup/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.
l 1+
r
Dr. T. E. Murley AEP:NRC:1086A D.
Pi e Stress and Remainin 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 displacements and gross plasticity and to prevent excessive fatigue loadings in the pipe, The basic characteristic of a secondary stress is that it is self limiting; thus, a failure from a single application of a secondary loading is not expected.
For the stratification issue, the potential effects of excessive displacements have been investigated through detailed visual observations of the surge line during the walkdowns required per Bulletin 88-11 action l.a.
I&M performed a visual inspection (ASME,Section XI, VT-3) of the Donald C.
Cook Nuclear Plant Unit 2 pressurizer surge line on January 14, 1989.
An identical inspection was performed on the Unit 1 surge line on March 28, 1989.
The results of these inspections did not indicate any gross discernible distress or structural damage in either surge line.
These results are available for inspection.
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 have demonstrated compliance with applicable ASME codes and a surge line fatigue life in excess of a 40-year plant life.
Review of plant specific fatigue calculations indicates 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
1 1'g V
5 t t
a J
i
>,a
~
4
Dr. T.
E. Hurley AEP:NRC:1086A 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.
For a design life considering 200 heatup-cooldown cycles (used in plant specific analyses),
this would indicate approximately 100 remaining cycles.
This number of remaining cycles far exceeds the postulated worst case number for the two-year 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 permit a safe and orderly shutdown.
F.
Ins ection Histor The NDE inspection history at Donald C.
Cook Nuclear Plant, 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.
Summar of Conclusions From WOG Pro ram Based on information assembled on surge lines for all domestic Westinghouse
- PWRs, and evaluation of that information in conjunction with plant-specific and other interim evaluation results, the WOG concludes that:
a.
A high degree of confidence exists that further evaluation will confirm that the worst combination has already been evaluated for stratification severity, structural effects and operating procedures.
4 G
1
Dr. T.
E. Murley AEP:NRC:1086A b.
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 remaining Westinghouse
- PWRs, the worst case remaining life is approximately 100. heatup-cooldown cycles.
c.
Through wall crack propagation is highly unlikely,
- however, "leak-before-break" would permit a safe and orderly shutdown if a through wall leak should develop.
d.
NDE inspection history demonstrates the present day integrity of Westinghouse PWR pressurizer surge lines.
e.
While additional monitoring, analyses, and surveys of operating procedures are expected to further substantiate 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, I&M believes it is acceptable for Donald C.
Cook Nuclear Plant to continue power operation for at least ten additional heatup-cooldown cycles.
I&M has committed to address the requirements of Bulletin 88-11 by January 1991.
This letter is being submitted under oath pursuant to the provisions of Section
- 182a, Atomic Energy Act of 1954, as amended.
As such, an oath is being submitted with this letter.
Sincerely, M. P. Alexich Vice President MPA/eh cc:
D.
H. Williams, Jr.
W.
G. Smith, Jr.
- Bridgman R.
C. Callen G. Charnoff G.
Bruchmann A. B. Davis - Region III NRC Resident Inspector
- Bridgman
4 Pg 5
~'
COUNTY OF FRANKLIN Milton P. Alexich, being duly sworn, deposes and says that he is the Vice President of licensee Indiana Michigan Power Company, that he has read the forgoing Response to NRC Bulletin 88-10:
Pressurizer Surge Line Thermal Stratification and knows the contents thereof; and that said contents are true to the best of his knowledge and belief.
Subscribed and sworn to before me this~i~
day of 198j.
NOTARY LI ffiunai~ir~ 3-P- 9l
'Ve If H
f f
tlf