ML20005B991
| ML20005B991 | |
| Person / Time | |
|---|---|
| Site: | Millstone, Calvert Cliffs, Oconee, Palisades, Saint Lucie, Crystal River, Midland, Fort Calhoun, Crane  |
| Issue date: | 07/16/1981 |
| From: | Knight J Office of Nuclear Reactor Regulation |
| To: | Novak T, Tedesco R Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML19291D364 | List: |
| References | |
| NUDOCS 8109160294 | |
| Download: ML20005B991 (5) | |
Text
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UNITED STATES 4
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g NUCLEAR REGULATORY COMMISSION W ASHING TON. D. C. 20555 Q..v /
.A 1 6 1981 MEM0'RANDUM FOR: Robert L. Tedesco, Assistant Director for Licensing, DL Thomas.H. Novak, Assistant Director for Operating Reacfors, DL FROM:
James P. Knight, Assistant Director for Components & Structures Engineering, DE
SUBJECT:
REPORTING OF UNSATISFACTORY EPRI/PWR TEST RESULTS FOR CONTROL COMPONENTS, INC.. POWER OPERATE.D RELIEF VALVE AND DRESSER MODEL 31739A SAFETY VALVE The attached memorandum from EPRI for the week of July 2,1981 discusses the results of both steam and water tests perfomed at Wyle-Norco on the Control Components. Inc. PORV and the results of steam tests at the CE-Windsor facility on the Dresser 31739A Safety Valve.
Note that this is not the same Dresser safety valve discussed in our June 16, 1981 memorandum. As described in the EPRI memorandum, each of these valves failed an EPRI " screening criterion" in one or more of these tests.
For each " failure", the applicable criterion is as stated in the memorandum.
It is our understanding that '.e Licensees and Construction Permit Holders that utilize or plan to utili e one or both of these valves and the NSSS vendors have been notificd r these test results and have the responsibility for assessing the safety significance of the observed valve behavior for their plants.
Our information from EPRI indicates that the Control Components, Inc. PORY is being used or will be used on the following plants:
McGuire 1 and 2 Catawba 1 and 2 The Dresser 31739A Safety Valve is bein'g used hr will be used on the following plants :
Calvert Cliffs 1 and 2 Crystal River 3 Palisades TMI-1 Midland 1 and 2 Millstone 2 Oconee 1, 2 and 3 G
8109160294 810914 PDR ADOCK 05000289 0
Robert L. Tedesco.
Although the specific safety significance of these test results is still being evaluated, this information may be relevant for licensing board notification.
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' night,': Assistant Director f /for Components & Structures Engineering
' Division cf Engineering cc:
'R. Vollmer F. Cherny E. Hemminger H. Gregg M. Stolzenberg
- 2. Rosztoczy R. Kiessel E. Jordan E. Brown D. Chaney R. Clark S. Varga W. Johnston R. Bosnak D. DiIanni S
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Ju1y 2,1981 TO:
DISTRIBUTION FRCH:.
John J. Carey
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SUBJECT:
S/RV TEST ACTIVITIES The EPRI/PWR Safety and Relief Valve Test Program testing activities for the period of June 29 through July 2 were as follows:
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Da Tuestey, June 30, two tests were performed on the CCI valve utilizing air pressure to open and close the valve.
The first was a 2750 psia steam test, the second a transition, steam to 650, water, at 2515 psia.
During both tests the valve opened and closed on demand.
On Wednesday, July 1, two tests were performed. The first vit e 2750 psia steam test with increased preload utilizing spring force only for valve closure. 'Dering this test the valve opened on demand.
Upon signaling the valve for closure the valve remained open for approximately 3 seconds prior to closure.
Valve closure occurred at an inlet pressure of 2220 psia.
For this test the EPRI screening criteria was not met (i.e., failure to ciose immediately ou demsnd).
The second test was a 2540 psia, water seal sim.rlation test utilizing air pressure to open and close the valve.
For this test the valve opened and closed.on demand.
On Thursday, July 2, three additional tests were perforced on the CCI valve.
The first test was a 2750 psia, steam test utilizing spring force only for valve closure.
The valve opened on derand, Upon siguling the valve for closure, the valve re.sined open for approxirately 3 se: ends prior to valve cicsure.
Valve closure occurred at an inlet pressure of 2210 psia.
Tne EPRI screening criteria was not net.
The second and third tests were 2750 psia, in:reased valve prelcad, with and without air pressure for valve closure, respe:tively.
For both tests the valve opened and c1csed on demand.
Upon re-evaluation of all the high pressure..prel$ad. tests performed on the CCI valve, it was noted that for the preload test performed on June 18, the valve opened on demand; however, upon signaling the valve for closure, the valve renained open for approximately 2 seconds, cicsing at a pressure of
- 2225 psia. The weekly activity report issued on June 19 did not note this delay in closure.
~ This completes the matrix of testing on the Control Components relief valve.
The valve will be disassert. bled'and inspected on Tuesday. July 7.
Current plans call for installation of the Masonelian relief valve on Tuesday, July 7, with resumption of testing on Wednesday, July 8.
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John J. Carey S/RV TEST ACTIVITIES Page 2
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C0;G'JSTION ENGINEERIh*G On Saturday, June 27, two 2500 psig, high ran'p rate,1cw back pressure, stea n tests were perfomed on the Dresser safety valve. (31739A).
For these tests the ring settings were gradually changed to increase the initial lift cf the valve.
For the first test the valve opened within 13% of the valve desigi set pressure.
A maximum stem position of 82% of rated lift was e.chte,ed at SE of the valve design set pressure.
Rated flow was achieved.*
Tne valve closed at a pressure greater than 2250 psig.
l For the second test parfomed on Saturday (with adjusted ring setting),
the valve opened at a pressure of 2595 ps.ig, which is greater than 13%
of the valve design set pressure. A stem position of 100% rated lift was echieved at 62. of the valve design set pressure. Rated flow was achieved.*
The valve closed at a pressure of 2185 psig, which is_btlow the_EPRLhl.ow-down pressure of 2250 psig.
The EPRI screening _ criteria. were not hat.
On Honday, June 29, a repeat test of Saturday's second test was performed.
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l For this test the valve opened within i 3% of the valve design set pressure.
A stem position of 100% rated lift wat achieved at 6% of the valve design set pressure.
Rated flow was achieved.* The valve close1 ataupressur_s.
oL2150_psig, which is below the EPRI blondowfi~p~risiiJie~ of 2250 psig. The EPRI screening criterlon was not met.
Dr. Tuesday, June 30, with the ring settings in the same position as the last two tests, a high ramp rate, high back pressure, steam test was per-forned.
Tne valve opened within + 3% of the valve design set pressure. A stem position of 1007 rated lift 67as achieved at 6%-of the valve design set pressure.
Rated flow was achieved.'" The valve closed at a presture gr,etter
.tnar. 2250 psig.
Peak back pressure was 855 psig.
On Wednesdsy. July 1, a repeat cf Tuesday's test was perforne'd with a l
slig'.tiy redated back pressure setting.
For this test the valve opened at a pressure within + 3% of the valve design set pressure. A sten position cf 100% rated lift was ach,iaved at 6% of valve de's:Ign set pressure.
Rated flow was achieved.* Tne valve closed at a. pressure of 2220 psig, which is beled the EPRI blowdown pressure of 2250 psig. The EPRI screening criterion was not net.
Peak back pressure achieve 8 during this test was 605 psig.
On Wednesday, a second high ramp rate, high back pressure test was performed.
with adjusted ring settings to further er. tend the time the valve remained in a full lift position.
For this test the valve opened within + 3% of the valve design set pressure.
A stem position of 100% rated lift was achieved at 6T of the valve design set pressure.
Rated flow was achieved.* The valve closed at a pressure of 2185 psig, which is below the EPRI blowdown pressure of 2250 psig. The EPRI screening criterion was not met.
Peak back pressu e achieved for this test was 655 psig.
- Eased on preliminary venturf flow data.
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,e 3 On Thursday, July 2, a high ramp rate. Iow back pressure, steam test.
with the same ring setting used for the previous test, was parformed.
The valve oper.ed at a pressure.with + 3% of the valve design set pressure.
A stem position of 100% rated lift was achieved at 6% of the valve de:ign set pressgra.
Rated flow was achieved.* The valve closed at a pressure of 2070 psig. which is below the EPRI blowdown pressure of 2250 psig. The EPRI screer.ing criterion was not met.
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DISTRIDUTIDS:
D. Ho f fca n Telecopy 4 517-788-0134
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J. Scott Telecepy # 201-430-5734*'
F. Cherny (NRC)
Telecopy # 301 492-4994' Panafax set at 5 R. G. Clisham Telecopy # 301-234-6716 Baltimore G & E)
K. Berry Telecopy A 704-373-B107 Duke Power)
G. A. Becker Telecopy # 904-795-6C86 Florida Power Corp.)
J. Correa Telecopy # 201-263-6500 Met. Ed.)
C. D. Maxson Telecopy # 203-566-6911, etx. 5896 (Northeast Utilities)
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Technical Contects - Participating Utilities '
W. B. Loevenstein J. J. Taylcr F. J. Arrotta G. Will icosor.
S/RV Staff
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T. Vandeventer (Philadelphia Electric Co.)
K. Beskin. Cha irman - CE Owne B. Gill. Cha irman
.B&W Gwner,rs Group s Group R. Jurgen:en, Chairman - W Owners Group
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E NUCLEAR REGULATORY COMMISSION j, ;s,;.. f. p W ASHINGTON, D. C. 20555
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AUG 6 1981
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MEMORAf;DUM FOR: Robert L. Tedesco, Assistant Director for Licensing Division of Licensing Thomas H. Novak, As'sistant Director for Operating Reactors Division of Licensing FE6M:
James P. Knight, Assistant Director for Components & Structures Engineering Division of Engineering.
SUBJECT:
REPORTING OF UNSATISFACTORY EPRI/PWR TEST RESULTS FOR DRESSER MODEL 31739A AND CROSBY MODEL 3K6 SAFETY VA'VE The attached memoranda from EPRI for the weeks of July 6, July 13 and Jul'y 18 discuss the results of steam and water tests. performed at Wyle-Norco on the Masoneilan and Copes Vulcan (17-4 PH plug and cage) PORV's.
These PORV's p,assed the EPRI " screening criteria" for all of the tests.
Both Safety Valves, however failed a screening criteria in one or more tests as described in the memoranda. For each " failure", the applicable l
criterion is as stated in the memorandum. Note that all of these tests were.ptrformed on steam with a short inlet piping configuration. These two valves have not been tested as yet with the long loop seal inlet piping configuration that is normally used on Westinghouse plants.
It is our understanding that the Licensees and Construction Permit Holders that utilize or plan to utilize these safety valves and the I SSS vendor have been notified of these test results and have the responsibility for assessing the safety significance of the observed valve behavior for their plants.
Our.information from EPRI.indicetes that the Dresser 31739A Safety Valve is being used or will be used on the following plants:
Calvert Cliffs 1 and 2, Crystal River 3 Palisades TMI - 1 Midland I and 2 Millstone 2 Oconee 1, 2 and 3 The Crosby 3K6 is-being'used or will be used on the following plants:
St..Lucie 1 and 2 Fort Calhoun 1 319 T 9g9220274
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Robert L. Tedesco _
Although the specific safety significance of these test results is still being evaluated, this infor6ation may be relevant for Board notification.'
If a decision is made to forward this information to one or more Boards, we recommend that a copy of the July 22, 1981 memorandum from Frank Cherny to Robert J. Bosnak be forwarded to the Board also. A copy of the memorandum was previoL91y distributed to you.
It contains the minutes of the July 17, 1981 meeting between the staff and EPRI and the PWR Owners Group at which the status of all of the Safety Valve and PORY testing in the EPRI program was discussed.
8##
t trector Js Knight, Assist 6
for Components & Str tures Engineering Division of Engineering cc:
R. Vollmer F. Cherny E. Hemminger H. Grer; R. LaGrange
- 2. Rusztoczy R. Kiessel l
E. Jordan i
E. Brown D. Chaney R. Clark
- 5. Varga W. Johnston R. Bosnak D. Dilanni l
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ELECTRIC POWER Rh8cAnOH INSiliUIh p
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July 24,1981 TO:
D15TF15tT10N FROM:
J:hn J. Carey o
SUEJECT:
S/RV TEST ACTIVITIES Tne EPRI/FWR safety and Relief Valve Test Program testing activities for
.the period of July 1.S 24 were as follows:
L'YLE DJring the pariod Saturday, July 18 through Friday. July 24 eiaht tests were perfoimed on the Copes Vulcan relief valve utilizing the 17-4 P.H.
Flug end Cage.
The tests were performed under steam, prelead, water, steam to water transition and water seal sirulation conditions. During all tests the' valve opened and clcsed on demand.
The valve was disassembled and i
inspa:.ted and no danage was observed-that would affect future valve per-I forca nce.
l The new,t valve to be tested is the Copes Vulcan relief valve utilizing the j
316 w/steliite plug.
CO'$'J*.TI 0's El.'GI HEE RING l
Six tests were perfonied on the'. Crosby 3K6 safety valve this week.
1 Th-es high rar; rate, tfgh tackDressu'e, stean tests were perforr.ed on l
Tuesdey.
Tte first test was re-for.ed with r.n adjested ring setting to irprove vah e b' o,,d:wn. The valve c;ered at a pressure witMn + 3% of the valve
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design set pres'sure. A stem position greater than 100% of rated lift was achieved at a pressure of 6% above the valve design' set pressure.
During the test the valve rtarted to oscillate (chatter), subsequently the test was terminated.
Blowdown.date was not obtained.
Feak back pressure was 855 psia.
The second test was performed with an intercediate ring setting and a slightly reduced backpressure. All EPRI screening criteria were satisfied.
Peak backpressure was 71,5 psia.
The third test was per'fomed with a third ring setting adjustment.
Due to instrumentation problems encountered during the test. valve stem posi-tion was not recorded. The test tlas repeated on Wednesday.
Hso$hutilers 34%' H.lview Avenue. Pett ONee Box 10412. Pro Aho.CA D'4303 (415) 855 2033 Wrhury:m 0]htt>~ thDQ %w.husells kenn. NY,{.9ultre 700. H'eShington. OC 200:46 (202) a'72N22
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O S/RV TEST ACTIVITIES Page 2 Fcr its rep'est tes't the valve cycled three times.
For the first cycle, ail E;F.I screening criteria were satisfied. Feak backpressere was 700 psia.
For the second cycle, the valve opened at a pressure of 3.3%
belev the valve design set pre 3rure w6.ich is below the ECRI coe..ing pressure criterion.
The velve cle:ed at a pressure within 10% of the valve design set pressure.
The data for the third cycle has not yet been reduced.
The second test performed on k'ednescaf, was a high ramp rate low back pressure, steam test with the intermediate rir.g settir.g adjustment.
This ring setting was the sar.e as that used for the sacer.d test per-forced on Tuesday.
This ring setting was selected for the re:iainder of the tests on tne Crosby 3P.6 safety valve.
For this test the EPRI screening criteria were satisfied.
Peak backpressure was 230 psia.
On FridEy, a steam to water transition, low flow test was perform f s For this test the valve cycled twa times.
For the first cycle the FPRI screening criterion were satisfied.
For the second cycle the valve opened at a pressure of 4.7% below the valve design set pressure which is below the EPRI blowdown pressure criterion.
The valve closed at a pressure within 10% cf the valve de:1gn set pressure.
OJ:/WJi/mw D:iTT.:5 UTION 1
D. Hoffman
- Telecepy # 517 788-0134 J. Sectt
- Telecopy # 201-430-6734 -
'E. Cherny (NRC)
- Telecopy f 301-492-4994 Panefax set at 6 J. R senberger
- Telecepy # 305-552-4192 (Florida Power & Light)
T. McIvor
- Telecopy # 402-535-4466 (Omaha Public Fower District)
T. Var.derventer (PhilaNelphia Electric Co.)
K. Beskin. Chaircan - CE Dwr.ars Group '
B. Gili. Chairman - 5$W Dwners Group R. Jurgensen. Chairman '- E Owners Group Technical Contacts - Participating Utilities W. B.' Leewenstein J. J. Taylor F. J. Arrntta
P. C e4 E E A LEL_y
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dSmOr8ndum July 17,1981 TO:
DISTRIEUTION ROM:
John J. Carey-SUS *ECT:
5/M *EST ACTI'ATIES Tht EFRI/PWR Safety and. Relief; Valve Test Progre, testing activities for the pe-iod of July 1317 were as follows:
WYl.E During the period f' rom Friday, July 10 through Wednesday, July 16 testing was Tne tests were perfor.t.ed under steam, performed on the Masone11an relief valve.
One additional preload, water, transition and water seal simulation conditions.
The two repeat tests had a slightly increased air supply pressure to Tne valve was dis-For all tests the. valves opened and closed on demand.
lio camage was asserbied and inspected by the Hssoneilen valve representative.The cage to body gas k observed that would affect future valve performance.
had w:s$sd out daring testing.
Tne C: pes Vulven relief valve utilizing the 17 4 ph plug and tage was installed Testing is scheduled to st, art tor.orrow, July 18.
today.
CY.5')~IiON E'GII.EE:lilG This Darir.g this week four tests were perforc.ed on the Crosby 3XE safety valve.T valve tes a design set pressure of 2500 psia, Dae to cer; uter and ra p ra:e, short d ration, hig'. ba:L;ressc*e, stent tests.insteare.tstio not recorded.
For this test, the valve The third test was performed on Thursdsf. July 1'6.
A maximum l
opened at a pressure within + 37, of the valve design set pressure. Rated flow was a:hieved.*
l stem position of 98% of rated lift was achieved.
i EPRI blow-The valve closed at a pressure of 2245 psia, which is.2% belowPeak backpre l
dewr. pressure criterion of 2PSD psia.
l On Friday, July 17 the fourth test on the Crosby 3K6 safety valve was performe The valve This test was a high' ramp' rate, high backpressure, steam test.
A naximum opened at a pressitre within + 3% of the valve design set pres
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Reted flow was achieved.* The valve closed at a l
pressure of 2225 psia Which is 1% below the EPRI blowdown pressure criteri valve design set pressure.
Peak backpressure for this test was 620 psia.
of 2250 psia.
The next test on the Crosby 3K6 safety valve is. scheduled for Hondey, July 20.
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July lo,1951, TO:
D!STEIEUTION.
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John J. Carey' r FROM:
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SUBJECT:
S/RV TEST ACTIV IES Tne EPR1/Ph'R.;afety and Relief Valve Test Frogram test,ing activities for the parfed of July 6-10 v;tre as follows:
WYLE Installation of the H soneila'n relief valve took place this week.
Testing was celeyed dee to facility boiler probler.ts encountered during systen heatup.
The probler.s have beer resolved.
Testing of the Masoneilan relief valve is scheduled to start teday.
i CCT:!.55 ION Et:0!NEEP.!NG 0i h:nday, July 6. a high rar.p rate, high tack pressure, stea:n test was per-forn.ed on the Dresser 31730A safe'./. valve.
For this test the upper ring cont ollir.g the.effect of back pressure or. valve' perfomsnce was adjusted.
l Tre s a'.e tF.e ed at a presiuFe within + 31 of the valve destgr. set pressure.
A r.t>1 c s'.r. p,sition of lil'! cf ra6-lift was a:. hie.ed at E'A of the valve design. set pressure.
Rated flow 6!st achieved.* The valve closed at a pre!! Lie E243 psig, Which is b!Iois the EEEl blcWdoWn preis01:e of 2253 psig, ins EFFI screer.ing criterie was not tret.
The Dresser 31739A safety valve was :(.noved fr.oi.the test stand on July 8.
'The Crosby 3KS safe,ty valve is presently being installed for testing. -
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UNITED STATES OF A!! ERICA HUCLEAR REGULATORY C0llMISSION BEFORE THE AT0!ilC SAFETY AND LICENSIf G BOARD In the Matt _r of I1ETROPOLITAN EDIS0N CO. ET AL.
Docket No. 50-289 (Three 1111e Island Nuclear (Restart)
Station, Unit 1) i JOINT AFFIDAVIT OF JOHN F. STOLZ AND DOMINIC C. DIIAllNI John F. Stolz and Dominic C. Dilanni state under oath as follows:
1.
I, John F. Stolz am a Branch Chief assigned to Operating Reactors Branch #4, Division of Licensing, Office of Nuclear Reactor Regulation of the U. S.
Nuclear Regulatory Commission.
I am currently responsible for managing the branch activities that include the review associated.with TMI-1.
In addition I review and approve all Safety Evaluations prepared by the st3ff members of the Branch. A copy of my professional qualifi-cations is attached.
2.
I, Dominic C. Dilanni am a Project Manager assigned to Operhting Reactors Branch #3, Division pf Licensing, Office of Nuclear Reactor Regulation of the U. S. fluclear Regulatory Cornission.
I am currently re,onsible for managing all of the review activities and licensing actions associated with the Prairie Island Nuclear Generatirig Station, Units flos.1 and 2.
(
At the time when activities associated with the reviews of unsatisfactory test results of safety valves for Tit!-l, I was responsible for coordinating the reviews of all activities and the preparation of the staff Safety Evaluations for Tl11-1 that were not related to the restart hearing matters.
A copy of my professional qualifications is attached.
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The "NRC Staff's Report on Board's Cocnents Regarding Board Notification of Unsatisfactory Test Results of Safety Valves" was prepared by us'and is true and correct to the best of our knowledge and beliaf.,
b 1 F. Stolz NM cb b
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Dominic C. Dilanni Subscribed,qnd sworn to before me 'this il Wday of September 1981.
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Notary Pubidt
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11y Commission Expires:
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JOHN F. STOLZ PROFESSIONAL QUALIFICATIONS OPERATING REAC' ORS BRANCH N0.'4
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T DIVISION OF LICENSING I am the Peanch Chief of the Operating Reactors Branch No. 4 of the Divis' ion of Licensing, U. S. Nuclear Regulatory Commission. This Branch is responsi-
' ble for the overall safety and environmental project management for assigned licensed operating power reactors that includes the review of technical and precedural aspects of proposed amendments to operating licenses. Operating plants having Babcock and Wilcox reactor systems have been assigned to this Branch.
1 I accepted an appointment with the technical staff of the NRC Regulatory organization in,1969 and was assigned as Senior Proje'ct Manager for safety } ; -,
review of Quad-Cities Station Units 1 and 2 and Mendocino Power Plant Units 1 and 2.
From April 1972 to April 1980, I have had branch supervisory respon-sibility' for the project management of licensing reviews of BWR 4/5 and 6 plants. PWR plants using Westinghouse, B&W, and Combustion reactors, and l
standard designs from General Electric, Westinghouse, Stone and Webster and l
l Fluor Pioneer for preliminary design approvals. During 1974, I also partici-l l
pated in the staff review of the Reactor Safety Study that was subsequently released as WASH-1400.
From April 1980 to March 1981, I was Branch Chief o
of the Systems Interaction Branch responsible for the development of criteria and methods that can be used to identify and evaluate common cause type of failures that can lead to adverse systems interactions.
I graduated frhm the City College of N5w ' York in 1942 with e Bachelor of Science Degree in Civil Engineering, obtaining a Master Degree in Civil Engineering from the University of Southern California in 1966.
I have also l
r John F. Stolz taken additional graduate level courses in nuclear engineering, structural engineering and mechanical., engineering at ~the University of California and New York University,.
My experience following my undergraduate degree, from 1942 to 1951, included military service in the Air Force, a member of the Civil Engineering staff at the City College of New York, and structural engineering and field construction with several consulting engineering and industrial firms.
From 1951 to 1953 I was employed with the consulting firm of Devenco Inc., where I worked on the structural design and analysis of the fjrst nuclear powered submarines, Sea Wolf and Nautilus.
In 1953, I joined the Atomic Energy Department of North America Aviation which subsequently became Atomics International Division of North American Rockwell Corporation.
My starting position of Research Engineering involved design and analysis of reactor core and system components related to a sodium-graphite reactor development orogram.
I subsequently became supervisor of a unit responsible for the design of supporting facilities for all nuclear power prototype plants and nuclear research facilities.
In 1958, I was ass,igned as Project Engineer for the design of the plant, fuel handling and support syster.s for the Hallam Nuclear Power Facility, a 75 MWe sodium-graphite reactor plant at Hallam, Nebraska.
In 1959, I was assigned as Project Engineer to modify the Organic Mod; ated Reactor Experiment at the National Reactor Test Station in Idaho, which involved redesign of the reactor core.. pressure vessel, fuel handling, instrumentation and control and process systems.
From 1962 to 1965 I held the position of Group Leader directing the work of four supervised units assigned to
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John F. Stolz support the development, design and qualification of compact nuclear reactor systems (SNAP Program) specifically in the areas of testing facilities required to sim'ulate nuclear spaceflight environment, stress analysis, and mechanical and electrical design on the SNAP systems.
From 1965 to 1966, I supervised a process systems unit responsible for systems design and analysis supporting the company's development projects on sodium ccoled reactors, organic-moderated heavy-water cooled reactor and desalinization systems.
in1966.IspentayearasAssistantProject Manager 'or the preliminary design and development of a 500 MWe sodium cooled fast-breeder reactor plant, specifically responsible for developing concepts, testing programs and budgetary plans for the overall plant' and fuel handling.
In 1967, I assumed a project management assignment with the Autonetics Division of North American Rockwell on system analysis studies in water and transportation systems, including manage-nent o' the contract studies for the State of California on the systems analyses of operations and maintenance for the California State Water Project.
I am a member of the American Society of Civil Engineers, ha ! been past Chairman of the Nuclear Structural and !!aterials Committee in the Structural Division of the Society, and am still an active member of that Committee and the Publications Committee.
I am registered as a Civil Engineer in the State of California and a Professional Engineer in the State of New York.
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7 D0!!!NIC C. DIIANNI J
PROFESSIONAL OUALIFICATIONS OPERATING REACTORS BRANCH #3 I am a Project fianager assigned to Operating Reactors oranch #3, Division of Licensing, Office of Nuclear Reactor Regulation of Mie U. S. Nuclear Regulatory Commission.
I am responsible for nanaging all review activi-ties and licensing actions associated with the Prairie Island Nuclear Generating Station Units 1 and 2.
I recently was responsible for co-ordinating the reviews of all activities for Three' Hile Island Nuclear Station, Unit 1 (T!!I-1) not related to the restart hearing matters.
/ I received a B. S. degree in Chenical Engineering from the University of Pittsburgh in 1952.
I have also taken extension courses from the University of Ottawa (' Chalk River, Ontario, Canada) in 1957, in reactor physics and reactor engineering.
From 1953 to 1955 I served in the U. S. Military Service and was given a rating of scientific and professional personnel. While in the s.ervice I was assigned to a U. S. Army Petroleum Laboratory in Oakland California where petroleum products purchased by the armed forces were analyzed.
From April 1955 to llay 1963. I worked as an engineer at Westinghouse Atomic Power Division Bettis Plant.
During this period I was involved in reactor fuel development and design of prinary reactor systems.
In l
addition, I was assigned as a resident engineer at the reactor site in Chalk River,. Ontario, Canada for two years with responsibilities for con-
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ducting reactor fuel irradiafion programs.
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Dominic C. Dilanni '
system (HVAC) inside containment, (4) reducing steam generator sludge levels by increasing steam generator :. low down' capacity (i.e., contributes to a reduc-tion,insteamgeneratortubedenting,etc.). Since backfit engineering.
for nuclear power plants covers a variety of liasks as Project Engineer it was essential that I have thorough knowledge and experience in the design and operation of numerous reactor auxiliary systens nomally associated with nuclear power plants. Furthermore, in this position, I coordinated the effor.ts of instrumentation and controls, electrical and structural engineers.
From October 1975 to the present I worked in the Office of Nuclear Reactor Regulation of the U. S. Nuclear Regulatiory Commission both in the !1echanical Engineering Branch and the Operating Reactors Branches, Division of Licensing.
As a senior staff member in the Engineering Branch typical assignments were to evaluate problems concerned with water hammer developed in feedwater systens, pipe cracks in the PWR chemical volume control systems, decontamination and steam generator chemical cleaning. As a Project Manager assigned to T!11-1. I was responsible for coordinating the reviews of all activities and the preparation of the Staff Safety Evaluations that were not related to the restart hearing matters.
I am a registered Professional Engineer in th~ State of Illinois and a member of the American Nuclear Society.
In sumary, my qualifications span 26 years in the nuclear field during which principles of chemical, mechanical and nuclear engineering have been applied.
In addition, management skills have been developed during the course of handling various nuclear programs.
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..o Dominic C. 011 anni
~ l These programs included the inpile irradiation of defed.tve fuel assemblies, development of primary water chemistry program, development of decontanina-tion programs and naterial testing.
In addition, I was responsible for primary loop system design that included equipment selection (i.e., pumps, valves, heat exchangers, process instrumentation, etc.). Field experience included initial. start up of reactor systems at the ETR site in ARCO, Idaho 1959/1961.
From May 1963 to June 1973 I worked as a Huclear Engineer and Project Manager at tha IIASA Lewis Research Center in the Advance Reactor Division.
I had direct responsibilities in thermionic fuel development for nuclear space power systems. My assignments related to developing U02 a fuel material (clad in tungsten) for thennonic reactors as a space power source. This work involved managing contracts and supervising in house efforts in reactor design, fuel assembly fabrication and irradiation testing.
I have published technical papers at NASA regarding these matters.
From June 1973 to October 1975 I worked as Project Mechanical Engineer at Fluor Corporation, Chicago, Illinois.in the Nuclear Division. As a Project Mechanical Engineer, I supervised highly technical and scientific staff in perfoming backfit engineering on operating nuclear p'ower plants (i.e.,
Kewaunee, Prairie Island, Quac. Cities & Dresden).
In order to perform th,ese duties, specialized knowledge of applicable industry codes and standards was essential. Some of the specialized tasks under my supervision consisted of:
(1) modifying pipe support systems.to handle high energies developed by slug flow at the pressurizer safety valves discharge pipes, (2) upgrading the main steam isolation valves, (3) upgrading air condition and ventilation L
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