ML20092N716
| ML20092N716 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 06/29/1984 |
| From: | Croneberger D, Slear D, Wilson R GENERAL PUBLIC UTILITIES CORP. |
| To: | |
| Shared Package | |
| ML20092N717 | List: |
| References | |
| 83-491-04-OLA, 83-491-4-OLA, OLA, NUDOCS 8407030472 | |
| Download: ML20092N716 (20) | |
Text
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RELAED c ASF0:4DENCQ t${EO UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION
'8f h~
- ki,.3 b i2 Before the Atomic Safety and Licensing-Boar'd~,
In'the Matter of
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- t. -
)
1 METROPOLITAN EDISON COMPANY, ET AL.)
Docket No. 50-289-OLA
)
ASLBP 83-491-04-OLA (Three Mile Island Nuclear
)
(Steam Generator Repair)
~ Station, Unit No. 1)
)
-LICENSEE'S TESTIMONY OF RICHARD F. WILSON, DAVID G.
SLEAR AND DON K. CRONEBERGER ON ISSUE 1.a (CONTENTION 1.a) j
-To Mr.- Wilson:
Q1.
Please state your name and address, and describe your involvement with the TMI-1 steam generator tube repair program.
A1.
My name is Richard F. Wilson.
I am employed by GPU Nuclear Corporation, 100 Interpace Parkway, Parsippany, New
, Jersey 07054.
As the Vice President of Technical Functions, I was responsible for the overall project and technical manage-ment of the TMI-1 steam generator tube repair program.
A' statement of my professional qualifications is attached.
4 To Mr. Slear:
Q2.
Please state your name and address and describe your involvement with the TMI-1 steam generator tube repair program.
8 A2.
My name is David G.
Slear.
I am employed by GPU !!u-4 clear Corporation, 100 Interpace Parkway, ?arsippany, New Jersey 07054.
As the Manager of Engineering Projects for t4 3sca-
f.'
f TMI-1, I was the overall task manager for the TMI-1 steam gen-erator tube repair program, reporting directly to the Vice President of Technical Functions.
My responsibilities included y
all activities associated with the evaluation and r, air of the steam generators.
A statement of my professional qualifications is attached.
To Mr. Croneberger:
Q3.
Please state your name and address and describe your involvement with the TMI-l steam generator tube repair program.
A3.
My name is Don K. Cronoberger.
I am employed by GPU Nuclear Corporation, 100 Interpace Parkway, Parsippany, New Jersey 07054.
As the Director of Engineering and Design, I provided technical management oversight of the failure analysis and repair activities with special emphasin on evaluation of the steem generator's mechanical design and the impact of the repair on the response of the components.
My department also j
provided engineering support in the areas of Materials Engi-neering/ Failure Analysis, Chemical Engineering and Chemistry, Mechanical Engineering and Engineering Mechanics.
A statement of my professional qualifications is attached.
To all witnesses:
Q4.
What is the purpose of your testimony?
A4.
The purpose of this testimony is to address Issue 1.a of Contention 1.a as enumsrated at page 23 of the Board's Memo-randum and Order (Rulings on Motions for Summary Disposition, dated Jurie 1, 1984) in which the Licensing Board stated r
1; The rationale underlying certain proposed license conditions should be addressed, with attention tos-a.
' Reliability of leak. rate measurements.
QS.
Describe the TMI-1 license conditions for leak testing the steam generators.
A5.
The existing license conditions *related to primary-tG-secondary (P-S) leakage through the TMI-1 once-through steam generator (OTSG) tubes are Technical Specifications (T.S.)
3.1.6.3 and 4.1.
Technical Specification 3.1.6.3 reads as follows:
If primary-to-secondary leakage through the steam generator tubes exceeds 1 gpm total for both steam generators, the reactor shall be placed in cold shutdown within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> of de-tection.
' Technical Specification 4.1., requires that leakage be evaluated daily.
In addition, the following proposed license condition dealing with leakage will be imposed:
Repaired Steam Generators In order to confirm the leak-tight integri-ty of the Reacter Coolant System, including the steam generators, operation of the facility i
shall be in sccordance with the followings j
2.
GPU Nuclear Corporation shall confirm the baseline primary-to-secondary leakage rate eatablished during the steam generator hot test program.
If leakage exceeds the baseline leak-age rate by more than 0.1 GPM (6 GPHl, the j
facility shall be shut down and leak tested.
If any increased leakage above baseline is due to i
dtracts in the tube free span, the leaking i
tuba (c) shall be removed from service.
The,
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s 4
baseline leakage shall be re-established, pro-1 vided that the leakage limit of' Tech. Spec.
j 3.1.6.3 is not exceeded.
The key points from this proposed condition are that:
- 1) Licensee was to establish its baseline leakage from the leak rate data obtained during the post repair OTSG hot test pro-gram; 2) an increase of more than 0.1 GPM (6 CPH) above th,is
~
baseline at steady _ state operating conditions requires facility
' shutdown and leak' tenting; 3) if leakage is due to defects in
~the tube free span, the leaking tubes are to be removed from services.4) leakage not identified as originating in the tube free' span during this testing is deemed acceptable if it does l
not exceed the 1 GPM (60 GPH) limit of TMI-1 Technical Specifi-cation 3.1.6.3; 5) the baseline is re-established following shutdown
- and leak testing (possibly at a higher leak rate tnan the initial baseline); and 6) operation can then continue until the increase in leakage exceeds the new baseline by v 1 GPM (6 GPH).
Licensee determined the baseline primary-to-secondary leakage to be 0.02 GPM (1 GPH) during the steam generator hot test program.
This means that the facility is to be shut down if the leak rate reaches 7 GPH total for both steam generators, as compared to the existing limit of 60 GPH in Technical Speci-fication 3.1.6.3. -
c l
'Q6.
How does this compare with the leak rate license con-l ditions for other nuclear plants?
A6.
The TMI-1 leakage limitations in Technical Specifica-tion 3.1.6.3 are comparable to those at most other pressurized water reactors (PWRs) in the United States.
A recent survey by Licensee of approximately 30 PWRs showed that the vast majority i
of the plants have limits similar to TMI-l's current 1 GPM f
limit.
One plant has a limit three times the current TMI-1 i
limit.
A few of the more recently licensed plants have limits
- lower.than T.S. 3.1.6.3.
However,-the proposed TMI-1 license L
condition of 0.1 GPM is more stringent than that for any other operating PWR in the United States.
Q7.: What is the purpose of measuring primary to secondary leakage?
~
A7.
Primary-to-secondary leak rate measurements are made periodically for'all operating PWRs in the United States in order to confirm that the steam generators are performing as anticipated.
TMI-l is no different than other operating PWRs in this respect.
These measurements are one aspect of an over-all defense.in depth approach to maintaining OTSG integrity.
.Tha program includes leak rate monitoring during operation, and periodic eddy current testing, and leak tests while shut down at cold conditions.
The leakage measurements during operation are made both to document the absolute value of leskage and to document any l
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trends which may be cause for concern.
The absolute value is 1 !
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,.i required to,both assess'the performance of the steam generators Land to. ensure that technical specification limits are not ex-coeded.
Trends are monitored because increasing leakage may indicate ongoing' chemical or mechanical degradation of the Ltube.
Increasing leak rates are investigated further to iden-tify leak locations and take appropriate corrective action.
i
.The i'ntent offthe overall defense in depth program is to correct defects in tubes in order.to ensure that the steam gen-erator tubes satisfy the licansing basis specified in General Design Criterion 14, 10 C.F.R Part 50, Appendix A, i
e.,
"to have an extremely low probability of abnormal leakage, or rap-idly propagating-failure, and of gross rupture".
Q8.
How were the leakage limits in the proposed license condition for TMI-1 established?
A8.
The-proposed license condition is based upon strin-gent administrative limits imposed by Licensee as part of its own-program.
Licensee included a number of considerations in
-establishing the absolute value of the leak rate increase dur-ing-steady state operating conditions which would dictate fur-
.ther action.
These considerations are summarized as follows:
1.
Establish a leak rate monitoring capability sensitive enough to detect a leak rate as low as 0.5 GFH (about 1% of the Technical Specification 3.1.6.3 limit) during power operations.
2.
Establish a baseline leakage rate to take into ac-count the anticipated, low level leakage from the mechanical plugs and the kinetically expanded joint.
The current baseline L i e
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A leak' rate of 0.02 GPM'(1-GPH) was based on monitored leakage during OTSG hot (pre-critical) testing.
3.
- Establish a snutdown limit sufficiently above the
' pre-established baseline so that we can have confidence that the change'is significant as compared with the anticipated
-variation'in the nominal monitored leak rate.
The OTSG hot testing results indicate that the monitored leak rate statisti-cal variation (twice the standard deviation from the mean value) of approximately + 0.01 GPM (+ 0.SGPH) can be expected during steady state operation.
4.
. Establish a shutdown limit low enough to ensure con-formance with the off-site exposure limits of 10 CFR Part 50, Appendix I.'
These limits are based on off-site exposure to
- various. body organs over a one year period.
Licensee has eval-uated off-site releases.
These evaluations are based on 0.03%
failed fuel.
This is the failed fuel percentage prior to the 11ast refueling, so we anticipate the actual failed fuel per-centage to be less when we restart.
We have determined that the gaseous -release mode results :ht the limiting off-site expo-sure dose closest to an Appendix I limit.
This limit is 15
[
mr/ year exposure to the thyroid due to iodine releases.
A con-
'tinuous 0.1 GPM primary-to-secondary leak rate contributes about 5 mr/ year to the off-site thyroid dose rate.
5.
Recognize the probability of multiple lackpaths with-in the OTSG contributing to the aggregate leakage.
The baseline leak rate value was determined at operating conditions
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- following an OTSG. inspection and leak testing with a drip and bubble test. tThese cold leak tests conducted before the hot u
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test program demonstrate that no single tube is causing all of the. current 0.02 GPM (1 GPH) leakage.
The results from these sensitive cold leak tests'showed that the baseline leak rate t
l value'isiand will be in the future the sum of multiple minor i
?leakpaths which would not'be expected to individually jeopar-I
- dize the integrity.of any OTSG tube.
Based'on these considerations, a nominal leak rate of 0.1 i
GPM, aboveLa baseline value, was established as the limit at
- which the plant is to initiate an orderly shutdown for OTSG inspection and identification of the leak source.
t i
Q9. - Can leakage commensurate with the license condition limit-be realiably measured during plant operation?
- A9.
Yes.
Primary-to-secondary leakage is indicated by several diverse methods at-TMI-1.
These methods include meas-
. uring radionoble gas concentrations on the secondary side, and measuring chemistry'dnd radio-chemistry in secondary side OTSG i
water.- The radionoble gas concentration measurement is the most sensitive method of quantifying the primary-to-secondary leakage rate.
The leakage rate is calculated periodically by
- utilizing data from on-line continuous monitors and grab sam-pies analysis.
The following describes the measurement tech-nique and-our evaluation of the sensitivity of this measure-ment.
The purpose of this description is to demonstrate that
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- the: leak rate value obtained by this measurement technique is t i
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sufficiently sensitive relative to the proposed license condi-tion limit.
Primary aide. activity is transported to the secondary sys-tem via the OTSO leakage pathway, and then carried over into the main steam system.
The main steam, condensate and I
feedwater sh'$tems distribute the primary leakage throughout the secondary side of the plant. ' Non-c'ondensable gases entrained in the steam and condensate are concentrated and removed from
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the system via the condenser air removal system.
A measurement of radionoble gas activity in the discharge of the air removal syst'em can be correlated with primary to cocondary side OTSO
>1eakago.
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The measurement of gaseous activity 1s accomplished by instrumenting the vacuum pump diccharge stream and providing a direct' readout of condenser-air removal system rate and activi-ty. concentration, and/ok by taking local'namples and then de-termining the OTSO P-Scloak rate via calculation.
At TMI-1 the radiation monitoring instrument.provided to determine the ac-tivity measurement is a heta scintilation detector designated RM-A5L.
The instrument is located,in the main condenser air removal system' discharge, common eight-inch' diameter header.
5 The me.nitor la manuf actured by, Victoreen, Inc., and includes a detector assembly consisting of a het.1 sensitive plastic crys-tal, optically coupled t.o a photomultiplier tube.
The readout J
associated with the monitor is located in the control room.
Based u. son the control room rearlout and condenwer air remo"ul
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ff-system flow rate, leakage can be calculated as a function of PM-A5L offier.cy and reactor coolant system activity.
Licensee has evaluated the sensitivity of the RM-ASL moni-4 tor to determine its suitability for measuring primary to sec-ondary leakage.
For the expected ranges of condenser offgas flow, reactor power and failed fuel, we have concluded that the sensitivity is at least 0.001 OPM (0.07 OPH) during steady state operation (power operation) and 0.003 CPM (0.2 OPil) dur-ing plant cooldown (sub-critical conditions).
The higher sen-sitivity during power operation is due to higher concentration of short half life radioisotopes in the reactor coolant system when the reactor is in operation.
Thus, the measurement tech-nique being utilized at TMI-1 is sufficiently nonsitive to sup-port the 0.1 OPM licensing condition.
Q10.
What cold leak tests are utilized to determine the location of leaks and what is their sensitivity?
A10.
There are two cold leak tests used to locato leaking tubes, the bubble test and the drip test.
The bubble test is conducted by pressurizing the secondary side of the OTSO with nitrogen at about 135 psig.
During this test the secondary side is partially drained and primary nide water is maintained a few inches above the upper tubesheet.
The inspector then looks for gas bubbles at the upper tubenheet bubbling through primary side water which in bainq maintained several inches above the upper tubosheet.
Licensee has evaluated bubbio test i
sanoitivity and determined it in the most sensitive cold leak l i
b.,
y vi test.. Based on bubble, test experience, an 80 mil diameter bub-
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ble originating' once every five, seconds can be located during w
sw the bubble test.
This correiatei to"a leak rate sensitivity of
'O.000005-GPM for any individual. leak.
The bubble test was used s
- e to test--about'the top 18 feeh'of the 56., foot long OTSG tubes.
.. Testing this upper portion of the OTSG tubes results in testing
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r 100%.of.thelnew kinetic expansion joints.
The. entire CSSG tube length is leak tested by the drip E
s test. -Theldrip test is conducted by-pressurizing the' secondary s
' side to approximately 150 $$ig.
During this test, the OTSG is
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full'of water onlthe'secondapy, side and drained on the primary
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side. : The inspect'or looks "for ' drope of water coming from indi-viddal' tubes on the primary side of the lower tubesheet.
Based
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on the' ability to, locate'one drop every three seconds, the sen--
'sitivity,of the bt'ip test.is as-low as 0.0002 GPM for any indi-vidual leak < located at or near the lower tubesheet.
For leak locations' higher-in the OTSG, the drop has further to. travel 3-
!before itacan be observed at(theJlower tubesheet.
This allows-W-
A more. time ?for. evaporation. of'the 1 akage water before' the -water f y-can drip:3down~and out. the bottom of the tube.' This-evaporation 1
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Even so, the
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,p mains quite good, a,nd'is estimated,to-bedibout 0.002 GPM (three
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Q11'.
What is-the relationship or relevance.of the
. leakrate measurements to the repairs made~on the TMI-1 OTSG tubes?
All.
The. leak' rate measurements made at TMI-1 measure T
total P-S leakage from the OTSGs.
This would include the con-tribution from: leakage through the joints.
As previously de-scribed,11f the nominal leak rate increases by 0.1 GPM, the plant will be shut down and the individual tubes, plugs and/or joints will be identified by the nitrogen bubble test and drip
- tests which we discussed earlier.
Q12.
Could leaks be self-sealing?
-A12.
Yes, in certain limited circumstances.
We believe
' there may be a. tendency for some leaks to be self-sealing, but only for leakage pathways between the expanded portion of the
' joint and the,tubesheet.
The joint is formed between the
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Inconel tube and the carbon steel tubesheet.
Since carbon steel has aLpropensity for general corrosion in a normal RCS chemistry environment, corrosion products are formed in the
~ tube-to-tubesheet joint.
Industry experience indicates that-these corrosion products tend to plug up leakage paths in the tight tube-to-tubesheet crevice and to stop.or slow (i.e.,
-self-seal)-leakage.
A~ trend of decreasing leakage with time for joints tested in the qualification program further con-firmed this industry experience.
To-be self-sealing, a leak past the joint would have to A
have a.very small flow through a pathway sufficiently tight to r,
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7 Jenable the build-up of corrosion products adequate to seal the leak.~
A leak of this size would not adversely affect the load bearing _ capability of the joint, or increase the probability of rupture within the joint.
SQ13.
Would the loss of pretension affect the usefulness of. leak testing of the repaired joint?
i
.A13.
No.. Leakage past a repaired joint is independent of the loss of pretension.
Pretension, or preload,'was originally placed on the tubes during-the manufacturing of the steam generators.
The tubes were. heated,'which elongated them-slightly by thermal expan-sion, and were-then attached-at each tubesheet.
When the tubes cooled, the metal'would-have tried to contract back to the
-original length at ambient temperature, but because the ends
-remained fixed,, contraction was prevented.
This produced a
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. tensile load on--the tubes.
At TMI-1, some tubes with complete circumferential cracks were freed from the original joint which
. fixed the, tube in the upper tubesheet.
These tubes contracted a
g a small-fraction,of an> inch, relieving all or part of the pre-r tension. LWhen_the kinetic expansion was performed on these Ltubes,'the tubes were again fixed at each end, but w.th the ab-lsence_of part or all of'the original-pretension.
This " loss of pretension" resulted in a-reductioniof axial tube load of only j
p~
several hundred pounds.
The kinetic process relies on horizontal forces to expand the. tubes, while pretension is an axial load (i.e., vertical in L
((, T n
F direction).
Since these load components are perpendicular with respect tofeach other, the loss of pretension does not affect the-ability:to expand the tube and form the new-joint.
- Thus, kineticly expanded joints formed-in tubes with loss of preten-sion are as tight, and therefore are no more prone to leakage, thanl tubes with preload.
Even if.there is leakage past'the repair joint, it will be
'through the tight crevice between the tube and tubesheet.
The loss of. pretension does not affect.the tightness of this joint and'thus can not affect the potential leakage flow path once tfixed.
Monitoring of leakage through such a joint is thus un-affected by a loss of pretension.
i
_Q14.
Would loss of pretension cause IGSAC cracks to be masked due to decreased leakage?
A14.
In theory, a tube without pretension wculd exhibit a lower leak rate than a tube with pretension for a'circumferen-tial through-wall crack of a given size.
In practice,.however, i
this phenomenon is unlikely to' mask the detection of a critical f
size crack at TMI-1.
.The rigorous testing already conducted on each tube--
'special eddy current testing, bubble testing and leak-testing--
show that such cracks.do not exist in the tube pressure bound-ary.
While the-conditions which caused the circumferential intergranular stress-assisted cracking-in TMI-1 have been elim-
'inated, if such a crack were to exist, it would propagate only (during' conditions when the tube was placed in axial tension; this will tend to offset che effect of pretension loss.
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_ t-C LTubes without a pretension load are placed in axial ten-sion-under some operating.-conditions, just as tubes with 1.
'preloadiare sometimes in axial compression.
During the steam
- -generator hot testing program, transients placed axial tensile loads of at least several hundred pounds on every tube in the
'~ steam generators--even those which_'had lost preload.
Measured Ll'eak rates were assumed to come entirely from one crack, and wereccompared with benchmark calculations of estimated leakage Jthrough cracks of a significant size under the transient load.
1 Tubes both with'and without preload were considered.
These re-
!sults confirmed the conclusion reached after eddy current, drip 4
Land bubble tests--that no large cracks remain undetected in
' tub'ingfin the TMI-1 steam. generators.
If' future cracks are-hypothetically assumed to be l propagating;duelto-IGSAC at' normal operating conditions,;the principa13 direction;ofqpropagation will1be axial along the 92 tube..IGSAC propagat on-is-principally perpendicular,to the i
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direction o'f highest 1 stress.
The highest tube stress is in the hoop;directionjat these conditions.
-A loss ~of pretension will s
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".not;cause reduced leakage from atial tube cracks because there tare' nolforces associatediwith loss of pretension trying to keep
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the: crack. closed.
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PROFESSIONAL QUALIFICATIONS Richard F. Wilson Vice President, fechnical Functions-GPU Nuclear Corporation j
r
'GPU Experience:
s Technical responsibility for the Engineering, Design, i
Licensing'and Technical _ Support of all nuclear _ generating stations for the GPU System.
The position manages'the technical resources of GPU Nuclear including day-to-day support for plant. operations.
Previously was Acting Director for TMI-2 from September, 1979,Jto about March, 1980, and before that was Director of the Engineering and Quality Assurance Departments within the GPU
' Service Corporation.
Between 1975 and 1977, was Manager of Quality Assurance for the GPU Service Corporation with responsibilities for design and construction Quality Assurance.
l other Experience:
Prior work. experience. included two years (1973-1975) as l
Manager of Manufacturing Engineering for Offshore Power
- Systems, Jacksonville, Florida.
Responsibilities included activities associated with manufacturing planning, tooling,
?
industrial engineering, manufacturing engineering, and
. technical support to.the planned manufacturing facility.
Prior to' joining offshore Power Systems, held a number of positions at:the Atomic International Division ~of Rockwell International, 1954 to 1973.
Some of these positio'ns included Engineering i
Supervisor, Department Manager, Chief Project Engineer, Program j
Manager, and Chief Program Engineer on a wide variety of Atomic International programs.
The lastl position was Program Manager for the Atomic International work on the fast breeder program.
l
. Performed ~and supervised work in almost every facet of reactor L
engineering, physics, facility design, safety, reactor operations, etc.
l c
Committee affiliations have included the EEI QA Task
}
Force, the AIF Committee on Power Plant Design,-Construction and Operation, B&W Plant Owners and BWR Owners Groups, EPRI i
i Nuclear Divisional Committee, etc.
Outside the utility j
' industry has~ served on a number of company and
. company / government advisory groups as related to specific o
p
- programs.
Education-and training includes a B.S. degree in Mechanical Engineering, University of California at Berkeley, 1951; an M.S. degree in Mechanical Engineering, University of j
Michigan, 1953; and one year attendance at the former Oak Ridge School of Reactor Technology in 1954.
Has attended a j
large number of management and other courses, including the University of Michigan Public Utility Executive Program.
i l
h
_._,._ _ _. _ __ i
i PROFESSIONAL QURLIFICATIONS l
t DAVID G. SLEAR a
WORK EXPERIENCE
(
i Company:
GPU Nuclear Corporation
'~
Title:
.TNI-l Manager Engineering Projects i
Responsibilities:
Management of TMI-l modification, which i
entails:
Management of the $25 million annual budget allocated for plant modifi-i cation; prioritization.of the various phases of plant modification; oversight of the technical adequacy of plant modifi-
?
cation and.of the components involved in plant modification; consultation regarding
. problem resolution with, respect to matters concerning plant modification; and direct supervision of 16 GPU employees.
This position demands constant attention to
~
long term and daily plant modification concerns.and,an extremely firm grasp of both the technical aspects of.TMI-Unit 1 and of the various modes and components of i
modification available for implementation i
4 at TMI-Unit 1.
-Dates:
1983 - Present j
Company:'
GPU Nuclear Corporation
Title:
OTSG Repair Project Manager Responsibilities:
Management (in conjunction with individual I
task managers) of all aspects of the OTSG Recovery program at TMI-l including f ailure f
-analysis, eddy current testing, corrosion testing, RCS. examination, RCS sulfer cleanups, and plant performance analysis.
This position involved direct management of the OTSG repair process and personal involvement in the decision making process with respect to the repair program.. This position also entailed the definition and implementation of the overall project, and required a broad 7
overview and analysis of the OTSG Recovery
- program.
In his capacity as CTSG Repair Project Manager, Mr. Slear was also called L
i i
e
.m g
-m--
---.,..--..,-.v.----.-..+.--...,-------.-%--
i ec
. David G. Sloar i
. Professional Qualifications l
~Page-Two c'.?on to deliver numerous presentations j
- ,oncerning project details before the NRC, ACRS, TPR, and the GPU Nuclear Corp.
[
management.
l
' Dates:
-December-1981 - November 1983 i
Company:-
GPU Service Corporation
Title:
.TMI-l Manager Engineering Projects Responsibilities: _
Similar to those listed for Mr. Slear's present position including management of i
'a.520 million budget and of project engineer-ing for modifications.
i Dates:
1979
'1981 A
Company:L GPU Service Corporation g
t
Title:
Preliminary Engineering Manager t
. Responsibilities:
This position entailed: the. analysis.and preliminary design of 400 Megawatt-l i
combustion turbines and of a 600 Megawatt L
coal-fired power plant; extensive analysis of the reliability and availability of the components to be installed in the prospec-L tive power plant; and the establishment of a baseline criteria document for the designated plants including the technical documentation and presentation of the plant design for management review.
.-Dates:-
1978 - 1979 Company:
GPU Service Corporation r
Title:
Canponent Engineer Responsibilities:
This position entailed: the review of design specifications and technical details of products going into TMI-2, including the
~
steam generators, pressurizer, main j
r
e Dtvid G. Sloar Professional' Qualifications
.PageLThree-l E
condensors, cooling towers, reactor vessel, and' internals; technical consultation and analysis of problems; and review of the contractor's design work on new components going into a plant.
UNITED STATES NAVY NUCLEAR SUBMARINE FORCE OFFICER
Title:
Engineer Officer Responsibilities :
This position entailed: essentially
[
primary responsibility and control of the i
. onboard nuclear power plant; control of all engineering sections, command of 4 divisions; i
and supervision of approximately 55 crewmen.
l t
Dates:
1972 - 1974 i
i
Title:
Machinery Division Officer Responsibilities:
As Machinery Division Officer, Mr. Slear was
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responsible for: all mechanical components l
of the primary-and secondary. systems of the i
power plant' including the1 steam generator, i
reactor, and-drive controls; chemistry control.
of the primary.and-secondary systems;'and l
the-. supervision of~15 crewmen.
Mr. Slear also served as an Auxiliary Division Officer in charge of.non-nuclear life support systems,
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and as a Communications Division Officer.
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?- Dates :'
1968'- 1972 O
I Mr. Slear also attended the Nuclear Power Submarine School
- from'1966
.1968,- during'which time he obtained.one year of nuclear l
power plant training (6 months 'classrcom, 6 months actual. plant
- training) in addition to the submarine qualification program.
f EDUCATION t
College:
University of Oklahoma a
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' Degree:
B.S. Mechanical Engineering
? Dates:
1961 - 1966 I
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. College:.'Stevens Institute of Technology L
I Degree:
M.'S. Mechanical Engineering t
Dates:
1974 1978 i
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- V PROFESSIONAL QUALIFICATIONS Don-K. Croneberger Director - Engineering & Design-GPU, Nuclear ^ Corporation i
'GPU Experience:
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Technical responsibility for the Mechanical, Electrical, Civil / Structural, Chemical, Radwastefand Materials Engineering
' support for all' nuclear generating stations for the GPU Systems.
l 1978lto 1980 was. Manager - Design and later Manager -
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Engineering & Design with GPU Service.Co'poration.
Directed r
. design engineering activities for all nuclear and fossil power
. generating facilities.and modifications assigned to GPUSC.
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Other-Experience:
[
Prior work experience. included a number of positions at Gilbert / Commonwealth during the period 1963 to 1978.
The last position was Manager Structural Engineering.
It included
-technical responsibility-for structural engineering mechanics for.all nuclear and fossil generating facilities.
Some of the
.other positions included Project Manager for balance of plant
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studies for a liquid metal fast-breeder reactor demonstration I
plant.
Other positions as Project Structural Engineer included responsibility for-technical supervision of structural engineering and engineering mechanics for a number of domestic nuclear power plants.
Earlier experience with the U.S. Navy included engineering and construction of radio telescope and 1
ancillary experience.
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Industry affiliations have included the EPRI Steam Generator. Owners Group, ASME Section 3 Division 2 (former Chairman) and other industry nuclear standards activities including Nuclear Structures and Plant Design Against Missiles.
' Education and training includes a B.S. degree in Civil Engineering from Pennsylvania State University,-1959.
Other technical training includes courses at U.C.L.A.,.M.I.T. and the University of Michigan.
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I have been' involved in the Steam Generator tube failure issue from the beginning.
I provided technical management l
oversight of failure analysis.and repair activities.
Special
. emphasis was placed on understanding the mechanical design of l
the Steam-Generators and' applying that understanding to the repair program and the understanding of the impact of the repair ton the response of the components.
.My department provided engineering support in the areas of Materials Engineering / Failure Analysis, Chemical Engineering O
and Chemistry, Mechanical Engineering and Engineering Mechanics.
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