ML17309A314
| ML17309A314 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 11/22/1983 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML17309A313 | List: |
| References | |
| NUDOCS 8311230099 | |
| Download: ML17309A314 (22) | |
Text
Mr. John E. Maier November 22, 1983 CC Harry H. Voigt, Esquire
- LeBoeuf, Lamb, Leiby and MacRae 1333 New Hampshire
- Avenue, N.W.
Suite 1100 Washington, D.C.
20036 Mr. Michael Slade 12 Trailwood Circle Rochester, New York 14618 Ezra Bialik Assistant Attorney General Environmental Protection Bureau New York State Department of Law 2 World Trade Center Nev York New York 10047 Resident Inspector R.E.
Ginna Plant
'c/o U.S.
NRC 1503 Lake Road
- Ontario, New York 14519 Stanley B. Klimberg, Esquire General Counsel New York State Energy Office Agency Building 2 Empire State Plaza
- Albany, New York 12223 Dr.
Emmeth A. Luebke Atqmic Safety and Licensing Board U.S. Nucl'ear Regulatory Commission Washington, D.C.
20555 Dr. Richard F. Cole Atomic Safety and Licensing Board U.S. Nuclear Regulatory Commission Washington, 9.C.
20555 Dr.
Thomas E. Hurley Regional Administration Nuclear Regulatory Commission Region I Office 631 Park Avenue King of Prussia, Pennsylvania 19406 U.S. Environmental Protection Agency Region II Office ATTN:
Regional Radiation Representative 26 Federal Plaza New York, New York 10007 Herbert Grossman, Esq.,
Chairman Atomic Safety and Licensing Board U.S.
Nuc1ear Regulatory Commission Washington, D.C.
20555 Supervisor of the Town of Ontario 107 Ridge Road West
- Ontario, New York 14519 Jay Dunkleberger New York State Energy Office Agency Building 2 Empire State Plaza
- Albany, New York 12223
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t UNITED STATES t
NUCLEAR REGULATORY COMMISSION WASHINGTON, b. C. 20555 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION EVALUATION OF RCS THERMAL SLEEVE PROBLEMS IN I'JESTINGHOUSE PLANTS ROCHESTER GAS AND ELECTRIC CORPORATION R.
E.
GINNA NUCLEAR POIJER PLANT DOCKET N0.50-244 INTRODUCTION For Westinghouse PWR plants, thermal sleeves have been used in the past at hose reactor coolant system branch line nozzles where a rapid thermal
,luctu-ation exists due to mixing of flows wi h different temperatures.
The intent was to
", educe or eliminate thermal stress of the nozzle wall, since the tempera-ture fluctuations produce rapid varying high stresses in addition to the normal 1
stresses present in the coIIIponent.
Thermal sleeves have been traditionally installed at'he following 1'ocations:
For a
3 loop plynt, a total of 8 sleeves were installed:
(1) one at the nozzle of 14" pressurizer surge line, (2) three at the nozzles of 5" cold leg safe.y injection lines, (3) three at the nozzles of 12" accumulator
- lines, and (4) one at the nozzle of. 3" charging line.
For a 4 loop p]ant, a total of 7 sleeves were installed:
(1) one at the nozzle of 14" pressurizer surge line (2) four at the nozzles of 10" accumulator lines with safety injections, and (3) two at the nozzles of 3" charging lines.
Design evolution of the thermal sleeves has resulted in a total of five generations, which were designated by Westinghouse as generation 0, 1, 2, 3 and 4 (Figures 1, 2, 3, 4, and 5).
Sleeves of different generations were used in different Westing-house plants built at different times.
The evolution in sleeve design from generation to generation followed the general state-of-art capability of more sopl isticated analysis.
II.
THERMAL SLEEVE PROBLEMS Since mid-1982, loose therma')
sleeves or sleeves with cracked attachment welds were found successively in several operating Mestinghouse plants, namely
- Trojan, McGuire, Hor h Anna 1 and 2.
The following provides a brief descrip-tion of such events:
A.
Trojan Plant:
(Ref.
1 to 3)
During the June 1982 refueling outage,.
an underwater TV inspection of reactor vessel internals revealed several loose metal objects beneath the lower core plate.
Subsequent investigation by Mes inghouse and the Utili-ty concluded that the loose parts were four thermal sleeves from the noz-zles of 10" accumulator lines in all four loops.
All loose sleeves had migrated through the cold leg into the reactor vessel.
One of the loose sleeves was broken intp several pieces but all four sleeves were recovered.
Radiographic examinations of other similarly designed sleeves on the unit revealed one broken weld and a slight movement of the 14" pressurizer surge line nozzle sleeve as well as a cracked weld in one of the two 3" charging lines.
Failed sleeves were all removed.
Justification for sleeve removal was reviewed by the staff.
The plant was permitted to con-tinue operations until next extended outage pending generic resolution of the sleeve'roblems.
B.
McGuir~ PIant:
(Ref.
4 to 7)
At nearly the same time as in the Trojan plant, a similar 10" accumulator nozzle ffiermal sleeve was found missing during the June 1982 outage for steam generator maintenance.
Based on signals from loose part monitoring, the licensee concluded that the missing sleeve was located in he lower reactor internals.
The welds of the remaining sleeves were shown to be in act by radiographic inspections.
All sleeves remaining in place were removed and the missing sleeve is expected to be recovered at first re-fueling outage or the first extended outage period.
Justification for operation until next extended outage with the unrecovered sleeve was
accepted by the staff.
Operation beyond that point requires the generic resolution of the sleeve problems.
In light of the occurrences in Trojan and McGuire Unit 1, McGuire Unit 2 started operations without any thermal sleeves installed in the reactor coolant system nozzles.
C.
North Anna Plant:
(Ref.
8 to 11)
In the summer of 1982, radiographic examination was conducted of RCS pipe welds d ring.he plant outage.
On Unit 1, one sleeve at-the 6"
safety'njection nozzle was found missing and was later recovered from the bottom of reac'or vessel.
In addition, the sleeve at 3" charging line nozzle was found to have cracked attachment welds and was subsequently removed.
Examination in Unit 2 also revealed that four thermal sleeves had cracked welds and were removed.
The plant resumed operation with six intact sleeves in-plaice in Unit 1 and with four intact sleeves in-place in Unit 2.
Justification to continue operations until the next outage was reviewed and accepted by the staff. Operations beyond that point will requires generic resolution of the sleeve problems.
D.
V.C.
Summer Plant:
(Ref.
12 E 13) ig Secause of the foregoing occurrences in other Mestinghouse plants, NRC was informed in July 1982 that all eight thermal sleeves in the reactor cool-ant system were to be removed p'rior to plant startup.
Justification for operation until the next extended outage was provided and was reviewed and accepted. by the staff.
Similarly, Summer will implement the generic reso-lution when it becomes available.
III.
PROSLEM CAUSES Mestinghouse investigation concluded that these occurrences have been confined to hose thermal sleeves of "Generation 3" design (Figure 4), which utilize
2 fillet welds 180 apart at the upsteam attachmen, instead of a continuous 360 weld used in other design.
The failures were attributed to attachment weld cracking, which was caused by high cycle fatigue due to flow induced vibrations.
The failure mechanism was substantiated by metallurgical evidence
,ound at the failed welds.
Initiation of cracks occurred on the ID and failure occurred through transgranular cracks propagated by the failure mechanism.
Mestinghouse has also indicated that sleeves of earlier design (i.e. Generation O, 1, 2) remain intact due to the fact that earlier design codes required the use of larger design margins.
IV.
SUMYJIRY OF CORRECTIVE MEASURES Further investigation conducted by Mestinghouse using the latest analytical capability has concluded that thermal sleeves are not really needed to maint in
,he stluc ural integrity of nozzles under thermal transients induced by injec-tion flows.
Thus the corrective measures investigated by Mestinghouse and I
implemented by the utilitihs consist of the following actions:
A.
For operating plants experiencing failed sleeves; l.
Generation 3 Sleeves with cracked welds found were removed in all cases, (i.e. Trojan, McGuire 1 and North Anna 1 & 2).
li 2.
Remaining Generation 3 sleeves with intact welds were either removed (i.e.
McGuire 1) or retained (i.e. Trojan and North Anna 1 8 2).
I 3.
Detached sleeves were either recovered and removed (i.e. Trojan and Hor.th Anna 1) or remain inside the reactor vessel until the next refueling
- outage, or next outage of extended period (McGuire 1) at which time they are planned to be removed.
The above measures were jus ified on the basis that (1) nozzle integrity will be maintained without sleeves, (2) intact sleeves are unlikely to have rapid degradation by judging the pace of events, and (3) the loose parts
0 0
of detached sleeves inside the reactor vessel will not cause safety concerns.
B, Four plants ready to start initial operation with "Generation 3" sleeves have had them completely removed (i.e.
V.
C.
- Summer, McGuire 2).
C.
For plants still under OL review, decisions have been made either not to install thermal sleeves in the reactor coolant system (i.e.
- Catawba, South Texas, etc.)
or to use sleeves of improved design, the "Generation 4"
sleeves (i.e. Vogtle and Seabrook).
V.
STAFF EVALUATION Sta ff r evi ew of the thermal sl eeve probl ems whi ch occurred i n Tr ojan, McGui re and t/orth Anna plants were conducted in two phases.
The initial phase was to pelform a plant specific review to justify continued operation for a specified 1
period.
(Ref. 3, 7, 8, 11', 13).
The final phase was to obtain generic resolu-'.
tion of all Westinghouse plants (Ref.
14 to 17).
The following is a summary of staff findings and conclusions of the final phase:
l.
By comparing the configuration of the various generations thermal sleeves (Figure 1 to 4), the "Generation 3" sleeve is the only design utilizing two brief welds 180'part at the upstream attachment, instead of a continuous 360'eld as used in other designs.
Conse-quently Generation 3 design sleeves are (a) subjected to more severe excitation due to bypass flow through the unwelded gap, and (b) the sleeves are more flexible and responsive to flow excitation due to less rigid constraint at the weld attachment.
Thus high cycle stresses due to flow-induced vibration cause fatigue failure at the at achment welds.
We concur with the Westinghouse finding on fail-ure mechanism as described in Paragraph III above, and our evaluation also conclude that the 360'eld for "Generation 4" sleeves is justifiable and acceptable.
Based on the effec s of the unique design of "Generation 3" sleeves as stated in Item 1 above and the fact that sleeves of former design (i.e. Generation 0, 2, 2) had reported no failure after much longer periods of in-plant service (Table 2),
we concur that the thermal sleeve problems are confined to the "Generation 3" sleeves only.
Thus the sleeve problem is a generic issue only "to those plants using "Generation 3" sleeves.
3.
Me have evaluated the methods, procedures, results and acceptance criteria of detailed stress analyses per formed by Mestinghouse to ensure structure integrity of various size nozzles without thermal sleeves.
Finite element techniques were used to calculate thermal stress effects under the most adverse operating transients, including their contributions to cumulative fatigue damage.
In each analysis, the whole nozzle structure and welds to the connecting pipe were included.
In addition to the operating transients, all other mechan-1 ical loads were 'included.
The analytical results indicated that the stress and fatigue damage are within the allowable limits set by Subsection NB of Sec ion III of the ASHE Code.
Me conclude that such analyses and results are acceptable to verify nozzle integrity without thermal sleeves.
The impact and wedging effects of a loose thermal sleeve on reactor internals, steam generators, and primary systems piping have been evaluated by Mestinghouse.
Me agree with the results of the evalua-tion that such effects are unlikely either to impair the reactor
'coolant pressure boundary or to cause unacceptable safety concerns due to the limited available impact energy which can be imparted on randomly targeted mechanical components.
I 5
Our justification to permit operation of Trojan, McQuire, and North 4 ~
Anna to continue for the period previously described is based on our evaluation as stated in Item 3 and 4 above.
6
0 Westinghouse has presented the historical background in the design and analyses evaluation of connecting nozzles.
Currently, due to
/
the advancement of analytical technology, two-0 and three-D finite element modeling permits more confidence in analytical results.
Heat transfer, stress and fatigue effects were evaluated under sev'erely defined transients to check whether the requiremen s of the ASNE Code were met.
We conclude that the analytical approach and results presented in both case specific evaluations stated in Item 3 above and the generic evaluation are acceptable.
At our request, Westinghouse has provided
- detailed lis of flow transients with specified thermal variations and anticipated number of occurrences in the nozzle areas (Ref. 16).
We have reviewed the Westinghouse information (Ref.
- 17) and conclude that these temper-a ure variations and number of cycles assumed in their stress analyses are reasonably conservative and are acceptable for their use in the l
evaluation of nozzle integrity.
8.
In our generic review concerning the effect of thermal cycling due r
to flow mixing from branch lines to the "main coolant loop at the nozzle locations, we find that Westinghouse analyses are adequate for handling high cycle and low temperature difference transients.
A steps>function of maximum temperature variation was assumed for the s ress calculation and fatigue evaluation.
It was found that the maximum stress so induced was below the endurance limit with adequate margins.
Thus fatigue fai lure at a nozzle location is 'unlikely to occur due o this type of transients.
We also evaluated the effect of other type transients which have low cycle-and high temperature differences.
The cumulative fatigue usage factor at each nozzle location was reviewed and was within acceptable code limits.
However in some locations, there is less margin available to the allowable limits.
Westinghouse also indi-ca-'d that detailed flow mixing information is not available.
Since flow stratification during mixing is a potential mechanism to initiate
micro crack at the surface of inner pipe wall, we believe that nozzle locations with high fatigue factors may need special attention.
YI.
REGULATORY POSITION A.
The following Mestinghouse plants have installed thermal sleeves using design details different from the Generation 3 series.
Such sleeves may re=ain in place.
Generation 0
Generation 2
R.
E. Ginna Point Beach 1
Indian Point 2 H.
B.
Robinson 2
Turkey Point 3 San Onofre '1 H&dam Heck Point Beach 2
Kewaunee Zion 1 Zion 2 Sal em 2 Diablo Canyon 2
Beaver Valley 1 Prairie Island 2
D.
C.
Cook 1 D.
C.
Cook 2 Sequoyah 1
Sequoyah 2
Generation 1
Generation 4
Salem 1
Indian Point 3 Turkey Point 4 Surry 1 Surry 2 Prairie Island 1
Diablo Canyon 1
Vogtle 1 Vogtle 2 Seabrook 1
Seabrook 2
B.
Fcr the Mestinghouse 3-loop and 4-loop plants choosing not to install thermal s >>eves at the reactor coolant loop branch line nozzles, the Licensee
should revise the Technical Specifications to monitor the injection, low transients which occur at he following nozzles and to evaluate their cumulative fatigue usage factors:
For 3-Loop Plants For 4-Loop Plants 6" Cold leg safety injection 3" Charging 10" Accumulator with Safety injecti on 3" Charging As long as the cumulative fatigue usage factor (CUF) at any nozzle loca-tion listed avove remains below 0.8, no action is required.
When the CUF exceeds 0.8, the Licensee should prepare and submit a specific plan (i.e.
more frequent inservice inspection at such nozzle locations) to ensure early detection of possible nozzle degradation for NRC approval.
The following plants are -in this category:
Byron 1 Byron 2 Braidwood 1 Bra~ood 2
Marble Hi 1 l 1 Mar bl e Hill 2 Comanche Peak 1
Comanche Peak 2
Catawba 1
Catawba 2
Cal 1 away 1 Wolf Creek 1
Shearon Harris 3
Sh aron Harris 4 South Texas 1
South Texas 2
C.
For operating plants using thermal sleeves of "Generation 3" design:
, ~
Removal of thermal sleeves at reactor coolant loop branch line noz-zles is acceptable provided a program is implemented per revision of the Technical Specifications to monitor the occurrence of injection flow transients and evaluate their fatigue usage factors for nozzles specified in item B above.
Action is not required until the
cumulative fatigue usage factor in a non-sleeved nozzle exceeds 0.8.
Mhen that. occurs, the Licensee should propose a plan for NRC approval to ensure early detection of possible nozzle degradation.
2.
For those nozzles having intact thermal
- sleeves, in lieu of sleeve removal the option to retain sleeves in place during continued plant operation is acceptable, provided a program to inspect attachment welds of these sleeves at each refueling, outage is initiated.
The program should be submitted for NRC approval.
Sleeves should be removed or replaced should any degradation of the attachment welds be found.
Action requirements after removal of sleeves are as specified in items C. 1 above.
The fol'lowing plants are in this category:
Trojan North Anna 1 & 2 Farley 1 8
2 V.
C.
Summer watts Bar 1
- 5. 2 McGuire 1 8
2 D.
For plants that have not received an operating license and that originally planned to use "Generation 3" thermal
- sleeves, the following options'pe acceptable:
1.
Remove thermal sleeves and implement requirements specified in Item B above.
2.
Rep]ace sleeves by those other than the "Generation 3" design and implement requirements specified in Item A above.
~ 1 3.
Retain sleeves of "Generation 3" design in place and implement re-quirements specified in Item C.2 above.
The following plants are in this category:
10
Shearon Harris 1 & 2 Millstone 3
Beaver Valley 2
~
~ ~
~ ~
11
Tab1e 1
OPERATING HISTORY TH""CHAL SLEEVE GENERATION tIUHBER OF OPERATING PLANTS/TOTAL A~~RO YEARS ON LINE
'0 9>
(9-U) 6/7'0 (6-10) 9/11 55 (1-9) 6/10 19 (1-7) 0/a 0
12
Fiaure 1
ORIGINALDESIGN THERMALSLEEVE BUTT N/ELD NOZZLE Mate r ial: SA376, TP304 Nozzle 360o Fillet Weld (2) 5/32" Dla. Vent Holes l
Sleeve R.C.
Pipe Wall j/8" Min.
Lower Collar or (4) Weld Deposits at 90
~
Safety Edge Note: Typical for 3" and larger nozzles 13
Fiaure 2
FlRST GENERATlON THERMALSLEEVE BUTT WELD NOZZLE Material: SA312-TP316 or SA240- TP316 Ci Nozzle
~ 360o Fillet Weld t
5/32" Dia.Vent Holes yt 180o Continuous Weld Bead Reactor Coolant Pipe Wall-"
t/8'g 1/4"
~(4) Weld Deposits at 90o-Grind for Tight Fit to Sleeve Note'ypical for all 3" and larger nozzles
FIGURE 3
S ECON D GENERATION THERMALSLEEVE BUTT WELD NOZZLE Material: SA312-SA240 TP304 or TP316 Nozzle~
360o Fillet Weld (2) 5/32" Dia.Vent Holes at 180o
~Continuous Rolled Bead
, (4) Spot Welds at 90 Grind for Tight Fit to Sleeve Reactor Co'olant Pipe Wall 1/8" Note: Typical for all 3" and larger nozzles
FIGURE 4 TH]FID GENEFIATlON THERMALSLEEVE BUTT WELD NOZZLE Material: SA240 or SA312 TP304 or TP306 1/8" Wide Slots Typ. 6 Places Flow Reactor Coolant Pipe Hall I'/16" 2 Fillet Welds
't iSOo (2) 5/32" Dia.
Vent Holes at 180o Nozzle Note: Typical ior all 3" and laro r nozzles 4 Weld Deposits at ~0o Grind d'or Tioht Fit to Sleeve
FIGURE 5 FQURTH GENERATION THERMALSLEEVE E,u TT WaCO XOZZLE 2 Notches Typ 4 Places at 90o R.C. Pipe Nozzle Y
360 Fillet.Weld (Between Notches).
~ Upper Collar Shrink Fit
~~
5/32" Dia Vent Hole Typ 2 Places at t 80o l
3/16" I
1/16" 4 Weld Buttons at 90 Grind for Tight Fit to Sleeve Lower Collar Shrink Fit 1/2" 17
REFERENCES 1.
- Letter, B.
D. Mithers of Portland Gener al Electric Co. to NRR Director, attachment:
"Trojan Loose Thermal Sleeve Safety Evaluations",
7/22/82 2.
- Letter, B.
D. Mithers of Portland General Elec ric Co. to NRR Director, attachment:
"Descripti on of Degraded Thermal Sleeves i n Trojan RCS",
7/S/82 3.
Memo, J.
P.
Knight to G.
C.
- Lainas, "MEB Evaluation:
Trojan Thermal Sl eeves",
8/6/82 4.
Letter',
M. 0. Parker, Jr. of Duke Power Co. to H.
R.
- Denton, attachment:
"McGuire Loose Thermal Sleeve Safety Evalua ion", 7/13/82 5.
View-graphs of Mestinghouse presentation, meeting in Bethesda among
- NRC, I
. Duke Power
& Mestinghbuse on McGuire Loose Thermal Sleeves, 7/14/82 6.
- Memo, R. J.
Bosnak to E.
- Ademsan, "McGuire 1 & 2 Thermal Sleeves",
12/27/82
.7..Hemo, J.
P.
Knight to T.
N.
- Novak, "SSER for McGuire 1 & 2", 1/12/83 8.
- Letter, R.
H; Leasburg of Virginia Electric
& Power Co. to NRR Director, attachment:
"North Anna Unit 1 Loose Thermal Sleeve Safety Evaluation",
10/12/82 9.
Hemo, J...P.
Knight to G.
C. Lainas, "HEB Evaluation:
North Anna 1 Thermal Sl eeves",
11/18/82
~
~
10.
- Letter, R.
H. Leasburg of Virginia Electric & Power Co. to NRR Director, attachment:
North Anna Unit 2 Loose Thermal Sleeve Safety Evaluation",
8/4/83 18
,~
Yiemo, J.
P.
Knight to G.
C. Lainas, "NEB Evaluation:
North Anna 2 Thermal Sleeves",
8/16/83 12.
- Letters, 0.
W. Dixon of S. Carolina Electric E
Gas Co. to H.
R. Denton, "Virgil C.
Summer Thermal Sleeves",
7/13/82 and 9/29/82 13, Memo, J.
P.
Knight to G.
C.
- Lainas, "MEB Evaluations:
Summer Plant Thermal Sleeves",
10/15/82 14.
View-graphs of Mestinghouse presentation, meeting in Bethesda between NRC and Mestinghouse on generic thermal sleeve
- problems, 1/27/83 15.
Meeting minutes, S.
Hou to R. J.
- Bosnak, "Generic Discussion with Mestinghouse on Thermal Sleeves",
2/28/83 16.
Letter (Proprietary),
E.
P.
Rabe of Westinghouse to R.
H. Vollmer, "Design
~
Transients Related to,lhermal Sleeve Removal",
NS-EPR-2763, 5/9/83 17.
- chemo, R. M.,Houston to J.
P. Knight, "Thermal Variation in Me'stinghouse Reactor Cool ant Systems",
5/16/83