ML20004C019
| ML20004C019 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 05/26/1981 |
| From: | Hukill H METROPOLITAN EDISON CO. |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-2.K.2.13, TASK-TM GL-81-19, LLL-164, NUDOCS 8106010428 | |
| Download: ML20004C019 (2) | |
Text
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Metropolitan Edison Company Post (Mfice Box 480 ll L
Middletown, Pennsylvania 17057 Writer's Direct Dial Numtwr May 26 1981 LlL 164 Office of Nuclear Reactor Regulation Attn: Harold Denton, Director U. S. Nuclear Regulatory Commission Washington, D.C.
20555
Dear Sir:
Three Mile Island Nuclear Station, Unit 1 (TMI-1)
Operating License No. DPR-50 Docket No. 50-289 Reactor Vessel Thermal Shock This letter is in response to your Generic Letter 81-19 issued on April 20, 1981 which required Metropolitan Edison to address the subject concerns and identify proposed specific actions for TMI-1.
Metropolitan Edison Company (Met-Ed) is an active member of the B&W 177 Fuel Assembly Owners Group.
This group is presently involved with resolving, in a generic manner,the Reactor Vessel (RV) brittle fracture concerns in B&W Cperating Plants. A report of the group's effort on this matter was submitted to the NRC on May 15, 1981 by a letter dated May 12, 1981 (Attachment 1).
Met-Ed is familiar with the group's program and will continue to actively participate in all future deliberations concerning the RV thermal shock Conce rns.
The ef fects of high pressure safety injection on RV integrity has been analyzed previously in response to NUREG-0737 (Item II.K.2.13).
The evaluation considered small break LOCA with extended loss of feedwater. Met-Ed submitted the Thermal Mechanical reports pertaining to these evaluations on February 23, 1981 (L1L 023) and May 12, 1981 (LlL 144).
Specific programs on thermal shock concerns which Met-Ed has completed, planned or is presently pursuing to further assure operation of TMI Unit 1 are ennumerated below.
1.
Revision of the small break LOCA operating guidelines regarding thermal shock have been incorporated in the TMI-l Operating Procedures.
These guidelines provide operator instructions for high pressure injection (HPI) throttling on subcooling when there is neither forced nor natural circulation RCS flow during the HPI Cooling Mode.
2.
Upgrading of the Emergency Feedwater System as described in enclosure 3 to our letter dated January 23, 1981 (TLL 680).
d(
.h5(
8106010DT Metrocontan Ecson Comcany is a Memcer of tre General Puchc Utines System
T H. Denton LlL 164 3.
Conformance to RV Material Surveillance Program per requirements of Appendix H of 10 CFR 50.
4.
Participation in the B&W Owners Group rer; tor vessel materials program.
The program objective is to demonstrate RV structural integrity throughout design life considering effects of irradiation embrittlement per item 3 above. The Group's effort include:
a.
development of less conservative fracture analysis techniques such as elastic-plastic considerations, b.
investigation of improved dosimetry and fluence calculation, c.
investigation of enhanced inservice inspection (ISI) techniques to reliably detect smaller flaw sizes and, d.
development of more sophisticated RV heat transfer and fluid mixing calculations to reduce conservatism associated with the simplified models in the generic analysis.
5.
Plant specific evaluation of TMI-l HP1 Cooling Mode to realistically include features that scre conservatively neglected in the generic analyses of BAW-1628 and BAW-1648.
0 6.
Maintenance of BWST te=peratures higher than 40 F which is the Technical Specification minimum.
7.
Participation in the current and future industry research effort through EPRI in the area of RV Thermal Shock, as appropriate.
8.
Final Abnormal Transient Operator Guidelines (ATOG) will incorporate item 1 above.
We have also reviewed the Board Notification on this subject (BN-81-06).
The information provided there is consistent with our views that while continued long term attention to this subject is necessary, there is no short term concern on IMI-l which would warrant extraordinary action.
Sincerely, Director, IMI-l Attachment cc:
J. F. Stolz R. Jacobs H. Silver
W
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SACRAMENTO MUNICIPAL UTILITY DISTRICT C 6201 S Street. Box 15830. Sacramento. Califemia 95813; G16) 452-3211 May 12, 1981 DR HAROLD DENTON DIRECTOR OFFICE OF NUCLEAR REACTOR REGULATION US NUCLEAR REGULATORY COMMISSION WASHINGTON DC 20555 DOCKET 50-312, RANCHO SECO NUCLEAR GENERATING STATION, UNIT NUMBER ONE REACTOR VESSEL BRITTLE FRACTURE At your request, we met with the NRC on March 31, 1981.
At that meeting, your staf f presented information on overcooling transients with repressurization.
As a result of that meeting, we agreed to present a letter report by May 15, 1981 to put the thermal shock issue into perspective.
Again, on April 29, 1981 your staf f requested another meeting in which additional information was requested to be presented in the May 15th report.
This additional request included justification for continued operation.
The attached letter report provides the requested in fo rma t io n.
This report has been discussed with representa-tives of the utilities with Babcock 6 Wilcox operating plants, and they concur with the conclusions.
Each Babcock 6 Wilcox licensee plans to make a submittal by May 22, 1981 which will identify the planned plant specific actions for thermal shock.
At the March 31, 1981 meeting, we agreed to furnish you with a review of the current industry research ef forts in the area of reactor vessel thermal shock.
Attachment two is a summary of I
the ongoing programs by the Electric Power Research Institute.
l d
i John
. Mattimoe Chairman of the Babcock 6 Wilcox Regulatory Response Group Attachments:
- 1) Letter report on Reactor Vessel Brittle Fracture concerns in Babcock 6 Wilcox l
l Operating Plants
- 2) Summary of the Electric Power Research Institute Programs pertaining to brittic l
fracture l
cc: G. P. Beatty, Jr. (FPC); K. S. Canady (DPCO); D. C. Trimble, (AP6L); R. F. Wilson (GPU); and W. C. Rowles (TECO)
I dup-8 mostrovX AN ELECTRIC S YSTE M 3 E R vlN G MORE THAN 600.000 IN THE HEART C'
C ALI F 0 R N I A
W
.t EPRI RESEARCH ON THE PROPERTIES OF IRRADIATED MATERIALS PERTINENT TO THE OVERCOOLING TRANSIENTS by T. U. Marston April 1981
--.;g gm.n DUPLICATE DOCUMENT Entire document previously i
entered into system under:
ANO8;G5)8e3'lJL of 'pages:
[a N o.
l B
)
1 LETTER REPORT ON REACTOR VESSEL BRITTLE FRACTURE CONCERNS IN B&W OPERATING PLANTS 1
f B
D l
l Bocument Identifier l-77-1125756 May 15, 1981 i'
Prepared by:
Babcock & Wilcox Company for the j
Owners Group of Babcock and Wilcox l
177 Fuel Assembly NSS Systems
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Table of Contents PAGE 1
I.
ABSTRACT 1
II.
GENERAL 1
Reactor Vessel Brittle Fracture during Design Basis LOCA A.
2 Reactor Vessel Brittle Fracture during Small Break LOCA B.
3 Small Break LOCA - Specific 3
Generic Assumptions 1.
4 Bounding Assumotions (No Vent Valve Flow Mixing) 2.
4 3.
Mixing Assumptions (Vent Valve Mixing) 5 4.
Operating Yessels Fluences (EF?Y) 6 5.
Generic Analysis Conclusions 7
111.
NON-LOCA OVERC00 LING EVENTS f
7 Comparisons of Overcooling Event to SBLOCA Analysis A.
CONCERNS EXPRESSED IN BASDEKAS' LETTER TO UDALL (4/10/81) 8 IV.
8
" Overcooling Transient Cools Vessel to 150*F" A.
8 j
Vessel fracture......cause core meltdown 9
B.
j Rancho Seco Transient, 3/20/78 9
C.
D.
Maine Yankee vessel Fluence 10 E.
Operator Instructions j
11 V.
OTHER ACTIONS 11 h
A.
Completed 11 1.
ICS/NNI & EFW Systems Upgrades 11 Revised SBLOCA Operating Guidelines t
11 l
2.
Recomendation to Maintain SWST Temperatures Greater 3.
Than Technical Specification Minimums 11 ATOG Considerations of Problem 3
11 4.
Owners Group document (BAW-1511P) on Reactor Vessel 5.
Materials 11 8.
Currently Underway 11 1.
Owners Group Reactor Vessel Materials Program 11
'{
Reactor Vessel Material Surveillance Programs in
.)
2.
Accordance with Appendix H of 10CFR50 12 C.
Imediate Future Plans w
12 1.
Plant Specific Evaluations 12 2.
2-0 vessei Heat Conduction Evaluations 12 3
3.
NON-LOCA events Evaluations 1
+
}
Table of Contents (Continued)
PAGE
.e
~
.a D.
Long Term Plans 12 1.
Consideration of Thermal. Mix Test (discussion with 12
.)
EPRI,CREARE,etc.)
- 2. ' Consideration of Enhanced Inservice Inspection 12 Techniques o
3.
Evaluation of In-place Reactor Vessel Thermal Annealing 12 d
4.
Investigation of Improved Dosimetry and Fluence 12 Calculations i
l VI.
SUMMARY
- JUSTIFICATION FOR CONTINUED OPERATION 12 1
VII. ATTACH.'.ENTS
.}
Table 1 A-1 Primary System Response During Overcooling Transients
.i Figure 1 A-2 Allowable and Actual Pressures vs. Time for Rancho Seco.
1 40F BWST, Mixing s
References A-3 a
.I l an' 4
e J.
9 Ji M
4 9
4
'1
' jI J
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Reactor Vessel Brittle Fracture I
i I.
Abstract:
o This letter report summarizes the evaluations made to date regarding a
possible brittle fracture of B&W operating plant reactor vessels during M
transients that result in severe overcooling with potential repressurization of the reactor vessel.
It was prepared in response to an e
NRC request during a March 31, 1981 meeting between the NRC and various The basis for concluding that there is no irnnediate
- 1 industry gN ups.
I brittle frecture concern (into 1983) for B&W operating units resulting from thermal shocking of the reacter vessel during small break LOCA transients is presented. A comparison of the small break LOCA event with other overcooling events is made to demonstrate the small break analysis bounds the overcooling transient.
Long term plans to resolve the concern are
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surmiarized.
II. General:
A.
Reactor Vessel Brittle Fracture during Design Basis LOCA
.j l
Babcock & Wilcox evaluated the capability of its pressurized water h
reactor vessels to withstand thermal shock caused by the double-ended rupture of a 36-inch-diameter hot leg pipe as early as 1969.(1) d
' At that time, the hot leg rupture was ascertained to represent the most severe LOCA condition (i.e. frcm the standpoint of a brittle fracture l
lm failure).
Based on this early analysis of the hot leg rupture. it was
~"
concluded that "The reactor vessel will not lose its integrity due to crack propagation as a result of thermal shock caused by actuation of f]
the ECCS following a LOCA even if this transient occurs at the end of 40 years of irradiation and the vessel wall contains a flaw of critical l
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size".
l3 l
L 9
Reactor Vessel Brittle Fracture during Small Break LOCA B.
i As a result of the TMI-2 transient, new operating guidelines wer issued which included operation of the HPI system in a once-thru cooling mode as a means of core cooling until the plant could be cooled I
This mode and depressurized and then placed on the decay heat system.
5 of operation raised new questions concerning the thermal shocking of the reactor vessel due to the cold HPI flow being injected into the vessel with no RCS flow.
Because of these new considerations and in response to NUREG-0737(2), analyses were performed in 1980 for the small Reports break LOCA transients with extended loss of feedwater.
documenting these analyses were submitted to the NRC by the Licensees 1981.(3),(4) in January.
Recently, the issue has been raised by the NRC as to whether or not the small break loss-of-coolant transient with extended total l of feedwater indeed represents the worst overcooling transient which should be considered with regards to reactor vessel brittle fracture.
This report addresses this concern and concludes that the small break LOCA transient (as analized in BAW-1648) is the limiting transient for
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This limiting event is, therefore, treated in the B&W NSSS designs.
some detail in the following section. followed by sections discussing the Non-LOCA events, other activities (ongoing and pl'anned) related to the brittle fracture concern and finally a sumary presenting q
justification for centinued plant operation.
I o
il
' 3 L1
.=
l Small Break LOCA - Scecific The snail break LOCA transient with extended loss of feedwater has been thoroughly analyzed with regard to reactor vessel brittle fracture (3),(4)
(A description of the transient scenario 'is provided in Section 1 of Reference 3.)
The analyses envelope all of the B&W operat'.ng units, (i.e., worst-case inputs are combined).
Some of the salient conservative assumptions used in these generic analyses are as follows:
1.
All feedwater is lost for an extended period of time.
2.
All reactor coolant flow is lost for an extended period of time.
3.
Core flow into the downcomer is assumed to pass through four vent valves rather than the eight valves existing on all but one plant.
This reduces the amount of warm water entering the downcemer.
A hypothetical maximum HPI flow capacity is assumed over the entire RCS 4.
pressure range analyzed. No single plant can achieve this hypothetical
)
capacity over the entire pressure range.
This assumption affects all the' analyses, including those which assume operator action to throttle HPI, since the initial reactor vessel cooldown prior to achieving 100 F subcooled conditions at the core outlet is maximized, resulting 0
in increased thermal stress during the transient.
0 5.
A worst-case HP1 fluid temperature of 40 F was assume '.
i Linear elastic fracture mechanics (LEFM) methods were used in the 6.
1 brittle fracture analysis.
No credit was taken for warm prestressing.
,1 7.
Materials information was taken from Regulatory Guide 1.99.
Reactor vessel most limiting welds were assumed to be located directly 8.
/
beneath the cold leg inlet nozzles.
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L Reactor vessel cooldown was calculated based on a one-dimensional heat 9.
conduction analysis.
lI
- 10. Mixing in the cold leg piping was not modeled.
ll
- l1 x
o
- - - ~
m 4
The major uncertainty associated with the analyses is the degree of heatup of the high pressure injection water due to
- Upstream mixing in the cold leg piping
- Heating by the reactor vessel walls J
- HPI pump energy (minimal)
- Heating by the cold leg piping (minimal)
- Mixing with vent valve fluid
~o The last item, the preheating of the incoming HPI by mixing with vent valve fluid, represents the most significant contributor to reducing the 1
brittle fracture concern.
J In order to evaluate the thermal shock concern, various thermal hydraulic assumptic is were made.
The major thermal hydraulic assumptions
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were:
1.
Boundino Assumotions Analyses were performed asstming no heatup of HPI due to any of the above" effects.
When natural circulation was assumed to be inhibited at approximately 10 min. into the transient, the downcomer fluid temperature at reactor vessel wall was ramped to the SWST temperature 4
(400F or 900F) in approximately 60 seconds. This case is essentially a zero mixing case after 10 minutes into the transient.
J 2.
Mix Assumptions Analyses were also performed assuming HPI fluid enters the downcomer, 3
mixes with the warmer vent valve flow, which is assumed to be circumferentially distributed, and then streams down the reactor vessel wall.
This is believed to be a more realistic assumption since some degree of HPI mixing and heatup is expected.
li 4
h t
]
1 l
l Also, the reactor vessel fluences were obtained from the Effective Full Power Years (EFPY) determined frcm core follow and the methodology as
- 12. 1981.
outlined in BAW-1511P which was submitted to the NRC on March This document represents a significant effort as part of the B&W Owners f
Group since 1976.(9)
The EFPY on B&W operating plants as of 4/27/81 is as follows:
Rancho Seco 3.45 EFPY
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Oconee I 4.90 EFPY f
Oconee II 4.36 EFPY Oconee III 4.21 EFPY Crystal River III 2.19 EFPY TMI-1 3.52 EFPY Davis Besse-I 1.25 EFPY Arkansas Nuclear One 3.91 EFPY i
q Unit 1 BAW-1511P also contains information on Quality Assurance of Reactor This includes weld number, vessel in which i
Vessel weld properties.
located, type of filler wire, type of weldment and varicus other 2
surveillance capsule measured and predicted information.
]
The analyses in BAW-1648 assumed operator action to.hrottle high pressure injection such that core outlet conditions wou:d be maintained
]
less than 100 F subcooled.
Appropriate revisions to the Small Sreak 0
In Operating Guidelines have been issued to the affected Utilities.
addition, B&W nas recomended to the operating plants that BWST temperatures be maintained greater than the Technical Specification minimum of 40*F.
]
3 The conservative bouna;ng assumptions were used in the 1980 generic analyses (3),(4) with the intent being to define the extent of the brittle fracture problem.
With these conservatisms, the following conclusions resulted frem the analyses:
1.
Rancho Seco and Oconee I reactor vessels represent the most and the 4
second-most limiting operating B&W units respectively at this point in
]
time. The limiting welds, as analyzed, with respect to brittla fracture in these reactor vessels are longitudinal welds.
These
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vessels have limiting longitudinal welds near the cold leg nozzles.
Hence, the analysis of these operating vessels currently bounds all 1
others.
Using the conservative bounding thermal-hydraulic assumption (thermal 2.
hydraulic assumption #1 on page 4) pl"s ccmbining worst case inputs in o
the generic analyses showed no immed: ste hPittle fracture concern 1
exists for the operating plants.
The analyses show that operator H
action.to throttle HPI flow will preclude brittle fracture.
3.
Using the more realistic mix assumption (thermal-hydraulic assumption 1
- 2 on page 4) indicates the most limiting reactor vessel has more than f'
one additional effective full power year beyond the present bounding analysis (i.e. into 1983) before any concern is approached, even considering worst-case SWST temperatures.
Inis is illustrated in
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Figure 1, which shows allowable and actual pressures during the transient for the generic analysis using Rancho Seco weld material
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properties at 4.8 EFPY, assuming worst-case 40*F BWST water.(3)
The actual Rancho Seco EFPY as of April 27,1981 was 3.45 EFPY.
l.
w 4.
I) 1
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Therefore. given operator action to throttle HPI there is no immediate brittle fracture concern for B&W operating units resulting from thermai shocking of the reactor vessel during small break LOCA transients.
Ul hon-LOCA Overcooling Events NUREG-0737, Item II.K.2.13, required that small break LOCA with extended loss of feedwater events be analyzed for reactor vessel brittle fracture.
Recently, the ACRS and the NRC have expressed the concern that perhaps other transients, such as steam line breaks. which have the potential for overcooling and subsequent system repressurization, may be more limiting transients with respect to the reactor vessel brittle fracture concern.
As a result of the NRC's request in 1975 (Referenca 5), our position regarding these repressurization events has been that operator action to mitigate system repressurization (by throttling HPI and utilizing atmospheric dump or turbine bypass valves) is adequate to keep reactor coolant p'ressure and temperature within technical specification limits over the service life of the reactor vessel.(5)
Table 1 compares primary system response during various overcooling events. As can be seen, the sma'll break LOCA cases (case 1 and 2) already considered in BAW-1648 result in more overcooling (to approximately 900F downccmer temperature) of the reactor vessel than unmitigated large steam line breaks.(7) Also, case 1, Table 1. clearly bounds all overcooling transients presented in Table 1 (with respect to the temperature transient). Based on these considerations, plus reliance upon the operator to mitigate the repressurization, the previous SBLOCA analyses are limiting, with respect to the brittle fracture concern. Assessment of the non-LOCA overcooling events (including subsequent repressurization) has confirmed this for operation into 1983. __
i
D IV.
Concerns Exoressed in Basdekas' letter to Udall 4/10/81 1
The Basdekas' letter of 4/10/81 has been reviewed and clarifications of 7
several items for B&W designed plants are provided below.
The quoted e
sentences have been extracted from the letter q
d A.
"Such transients can cause the reactor vessel to cool-down to about 150*F in about 15 minutes, while the ECCS repressurizes it to about v
2400 PSI."
i In response to the IE Bulletin 79-05C. and as indicated in Section III,
}
a large steam line break was analyzed.
The analysis assumed both OTSG's blowdown, no Main Steam Isolation Valve (MSIV) closure and Emergency Feedwater at full capacity.
The results indicate a minimum Reactor Coolant System (RCS) temperature of 230*F will be reached
]
approximately 14 minutes into the transient.(7) Operator actions to throttle HPI flow will prevent repressurization of the RCS to 2400 PSIG.
B.
"A reactor vessel fracture is one of the most serious accidents a r
reactor may experience.
Depending en its location and mode, it is almost certain that it will cause a core meltdown with all its public
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health and safety ramifications, on which, I am sure, I need not elaborate for you."
lI
'l l
l It is very unlikely that a reactor vessel fracture, at a location and j ';
mode which results in a core meltdown. will occur.
This is
!W demonstrated by the positive margins resulting from analyses previously performed.(1.3,4) l e n
.-_r
C.
"This is supported by analyses performed for the NRC indicating that v
the overcooling transient tha', took place at Rancho Seco on March 20.
)
1978 would have caused such a vessel to rupture, had it been in operation for about 10 FPYE."
We are not aware of the information that Mr. Basdekas,has, but the
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Rancho Seco vessel en March 20, 1978 had only 1.55 EFPY of irradiation and therefore appreciable margin for Brittle Fracture at that time.
In an analysis prepared for the NRC by Oak Ridge National Laboratory (ORNL 7
to Mr. Milton Vagins (NRC) dated March 3,1981) a different analysis (Warm Prestressing) than that the one used in BAW-1648 indicates that 9
the Rancho Seco Vessel has a useful Full-Power Life greater than 14 EFPY.
m J'
D.
"Furthermore. a recent discovery of a discrepancy existing between the estimated vs. the measured values of neutron fluence for the Maine Yankee reactor vessel indicates a generic problem that makes things worse.
The results of dosimetry measurements indicate the actual neutron fluence to be some 2.3 times higher than that estimated in the Maine Yankee Final Analysis Report."
The fluence discrepancy at Main Yankee was apparently due to lack of azimuthal flux variation in their calculational model and/or the use
'1 of cycle 1 ext:apolated data.
Azimuthal variations in a B&W reactor are it on the order of a f actor of 2 frem maximun to minimum.
Core escape flux "l
is generally lover during cycle 1 (ccmpared to subsequent cycles), and, l-therefore, ex-core fluences would be low.
The fluence analysis procedure used at B&W accounts for azimuthal flux variation by using the two-dimensional transport code 00T to model reactor and surveillance j
capsules: and predicted fluences for extrapolated burnups are based on
]
core escape flux from fuel management studies (P0Q criticality calculations) of future fuel cycles.
B&W has always used the l]
two-dimensional modeling approach whereas the initial Maine Yankee data were from a one-dimensional model.
0-
The B&W procedure has been used to calculate the fluence exposure of capsules from five 177 FA reactors, four after cycle 1 and one after cycle 2 Comparisons to measured activities frem capsule contained dosimeters have been +15%.
All calculated data are subsequently normalized to dosimeter measurements before pressure vessel fluence is determined.
These data are documented in BAW reports that are sent to the appropriate utility after each capsule is analyzed.
]
The B&W procedure was benchmarked when B&W participated in the " Blind Test" phase of the LWR Pressure Vessel Surveillance Dosimetry Program, an on-going study of surveillance analysis procedures that is operated by HEDL and ORNL for the NRC.
B&W calculated f ast flux as documented in NUREG/CR-1872, " Reactor Calculation Benchmarks - PCA Blind Test Results," January 1981, was within 10% of experimentally derived values at the simulated T/4 pressure vessel location in two experimental configurations.
The " Blind Test" results are being documented in a NUREG report, but data are not identified with respect to participant.
t E.
"Moreover, as you may recall, one of the measures ordered by the NRC q
after the TMI-2 accident was to have all reactor operators not turn off the ECCS once it had been initiated."
1 Revised Small-Break LOCA Operating Guidelines have been issued to
]
affected Utilities by B&W.
The guidelines provide operator instruction on when to throttle the HPI flow to prevent repressurization.
El 3
1
y y
V.
Other Actions - The thermal shock concern has been addressed and programs have been either completed, currently underway or planned to assure safe 7
operation.
)
A.
Completed 1.
ICS/NNI upgrades per IE Bulletin 79-27 and associeted Comission
]
orders.
I 2.
EFW Systems Upgrades x
3.
Revised Small-Break LOCA Operating Guidelines regarding thermal shock have been issued to affected utilities.
m These guidelines are intended to:
]
- Enhance understanding.
- Provide operator instruction for HPI throttling on subcooling when w
in the HPI cooling mode with no RCS flow.
- Emphasize re-establishing RC Loop Flow,
.l 4.
Abnormal Transient Operating Guidelines ( ATOG) procedure under development to address item I.C.1.in NUREG-0737 includc consideration of the brittle fracture doncern.
5.
B&W has recomended that Utilities maintain SWST temperatures higher
[
than Technical Specification minimums.
6.
BAW-1511P (reference 9) has been completed as part of an Owners Group program on Reactor Vessel materials.
l7 B.
Currently Underway p.;
I 1.
The Owners Group reactor vessel materials program is geared toward i
demonstrating adequate structural integrity of the reactor vessel throughout plant design life.
Efforts currently underway include:
a determination of fracture toughness properties which are l
expected to demonstrate higher resistance to fracture than l?
'4 current industry predictions based on Charpy V notch specimens.
the development of less conservative fracture analysis procedures, which include elastic-plastic techniques.
,i 2.
Reactor Vessel Material Surveillance Programs in accordance with
, }-
Appendix H of 10CFR50.
l m'
w C.
Imediate Future Plans 1.
Plant specific evaluations to address the conservatisms associated with generic analyses are ceing investigated 2.
More sophisticated vessel cooldown calculations are being considered to reduce the conservatisms associated with the one-dimensional heat
]
conduction analysis previously employed.
T 3.
Consideration of analysis for Non-LOCA events.
c D.
Long Term Plans 1.
Discussions are in progress with EPRI regarding possible testing to w
obtain a better understanding of the thermal-hydraulic mixing phenomena associated with these overcooling transients.
2.
CREARE, Inc. and other consultants have been contacted and involved in discussions concerning the thermal-hydraulic mixing aspects of the problems.
a 3.
The investigation of enhanced inservice inspections methods with the objective being the reliable detection of smaller fl.aw sizes.
Y 4.
.The evaluation.of the in-place reactor vessel thermal annealing to recover some of the material properties lost through neutron 7
irradiation.
5.
The investigation of improved dosimetry and fluence calculations.
l' VI Sumary - Justification for Continued Oceration
]
l As a result of the NRC's request of March 31, 1981 to put the reactor vessel
]
brittle fracture issue in perspective, the following have been concluded:
A.
Assessment of overcooling events indicates that the small break LOCA event as analyzed is bounding, i
B.
Generic analyses (including mixing) of the small break LOCA events show no immediate problem (into 1983) given operator action.
C.
Revised operator guidelines have been issued.
Immediate operator action is not required.
Required operator action is straightforward.
D.
Efforts are underway to resolve the long term issue.
] -
(
- 1>
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1 Table 1
.J i
Primary System Resoonse Durino Overcoolino Transients Minimum Cooldown Downcomer Case Descriotion Rate Temcerature Ccmment 1
1 BAW 1648 A unding 460F in 90F No temperature
]H Analysis (
60 Second!
recovery (4600 / min)
No repressurization*
F I
(90F BWST) 2 BAW 1648 445F in Analysis (
x 40 minutes 90F No temperature (11.10 / min) recovery F
No repressurization*
(40F BWST) 3 Unmitigated Large 320F in 230F Temperature recovers e
Steam L 10 Minutes System repressurizes**
0 Rupture (32 F/ min) 4 Rancho Seco Rapid 310F in 285F Scme temperature Cocidown Ing 60 Minutes recovery of3/20/78Wjdent (5.2*F/ min)
Stable pressure
.j between 1400 and 2100 psig
)
Assuming operator action Can be mitigated by operator action L
a 13 a
.3 9
U A-1 1
UUU L a-U-5,,,$
$,.,E U
ns~s m~L_Z~~ J~~ J ~ J G J
L.,L d
u,,1_
2 Figure i ALLOWABLE AND ACTUAL PRESSURES VS TIME, 0.023-fT PRESSURIZER BREAK 3lTH OPERATOR ACIl0N, RANCHO SECO, 40f BWSI, MIX 2 i
NOTE:
BASED ON GENERIC ANALYSES AND ASSOCIATED 2500 CONSERVATISMS DOCUMENTED IN REFERENCE 3, 2000 C.
8.
p 5
1500 e
a_
1000 4.8 EFPY ALLOWABLE PRESSURE
- 1/1/83 BASED ON 100%
CAPACITY FROM 4-27-81
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ACTUAL TRANSIENT PRESSURE I
2 3
Time (hrs)
3-
]
References
]J (1) Analysis of the Structural Integrity of a Reactor Vessel Subjected to Thermal Shock, BAW-10018, Babcock & Wilcox, Lynchburg, Virginia, May 1969,
~
Transmittal letter. J. H. MacMillan (B&W) to Dr. P. A. Morris ( AEC).
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Dated May 22, 1969.
(2) NUREG-0737, Item II.K.2.13
]
(3) Thermal-Mechanical Report - Effect of HPI on Vessel Integrity for Small Break LOCA Event with Extended Loss of Feedwater, BAW-1648. Babcock &
Wilcox, Lynchburg, Virginia, November 1980.
(4)
Reactor Vessel Brittle Fracture Analysis During Small Break LOCA Events N
with Extended Loss of Feedwater, BAW-1623, Babcock & Wilcox. Lyncnburg, Virginia. December 1980.
(5)
F. Schroeder (NRR) to K. E. Suhrke (B&W), Letter dated June 10, 1975.
(6)
K. E. Suh; ke (B&W) to F. Schroeder (NRR), Letter dated August 12, 1975.
(7)
J. J. Mattimoe (SMUD) to R. H. Engelken (NRC), October 24, 1979.
(8)
Coninittee Report on Rancho Seco Unit 1 Transient of March 20, 1978, Dated June 19, 1978, by Sacramento Municipal Utility District.
T J
(9)
Irradiation-Induced Reduction in Charpy Upper-Shelf Energy of Reactor Vessel Welds, BAW-1511P (Prop. ietary), Babcock & Wilcox, Lynchburg, Virginia, October 1980, Transmittal Letter, J. H. Taylor (B&W) to e
.g J. S. Berggen (NRC), Dated March 12, 1981.
]
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