ML20062H665
Text
o e
(_\\)
/
t UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of S
-S HOUSTON LIGHTING & POWER COMPANY S
Dock,et No. 50-466 5
(Allens Creek Nuclear Generating Station, Unit 1)
.S s
v Statement of Material Facts As To Which There Is No Genuine Issue To Be Heard for McCorkle Centention 17 (1)
The Allens Creek containment design does not allow 20 percent of the containment leakace to bypass the filtra-tion systems.
(Affidavit p. 7)
(2)
A complete list of all potential leakage paths through containment penetrations was compiled (Exhibit A).
From this list, six penetrations were identified which e
constitute potential unfiltered leakage paths (Exhibit B).
(3)
Using the list of potential unfiltered leakage paths, the current best estimate of the maximum total un-filtered bypass leakage under LOCA accident conditions is
.0195 percent per day of the containment volume.
(Affidavit p.
4)
The containme - v ti Se designed in any event to limit leakage to t.
3 mt by weight of the conrainment atmosphere per day at calculated peak pressure.
(Affidavit p.
6 )_.
0355 g00819 n
c00
s kw i
(4)
Applicant will perform extensive pre-operational tests in accordance with 10 CFR Part 50, Appendix J, to assure that the containment will maintain its expected level of leak-tightness.
(Affidavit p.
4-6).
e 4
l 7
1
,m l
s i
l, i
1 l
l i
l.
531
d-?.72*
COST $
7 lI:98 McCorkle Contention ^No. 17/
PALO BY PLF. DEF.
Filtration System Leakage
\\.
r 3
UNITED STATES OF A!!?nICA MUCLEAR REGULATORY CO?!!!!S S ION BEFORE THE ATO!!IC SAFETY & LICENSING BOARD i
n I
IN THE MATTER OF:
)
)
i HOUSTOM LIGIITING & POWER COMPA1Y )
DOCKET NO. 50-466 l
(ALLENS CREEK UUCLEAR GENERATING )
STATION, UNIT 1)
)
j i
I i
i t
i i
t DEPOSITION OF:
DRENDA !!cCORKLE Q
)
50?
gg
/
rW at 3 2ters a
a 191/ Bank of the SoJtnwest Suiteing. Houston, Texas 77CO2. (7 3) 552 531 Y
24 1
that, then.
2 Okay.
7urning to the last area 3
that has to do with excessive leakage hypassing 9
4 filtration systens, and I have again an 5
introductory very broad question that I an 6
forced to ash reall"..
Mhat leakage hynassing filtration systen ar you talkinc about?
~
a 8
Me can't pinpoint the structure or systens that s
9 you have reference to.
r-Mhich interrogator;r are you talk ng about?
{
10 A
11 0
7 hat would he veur.interrocatorv, the ua"-
12 I have it nunhered, 17 I will read it the 13 "av I ha"e it recordea.
It saus the containnent i
i 14 as designed will allou excessive leakage to i
15 hvpass the filtration syston, potter conpany t
16 admits that 20 nercent of the leakage would i
17 not even be filtered and also the filter 18 absorher, I think neant adsorher, may start 19 a fire by auto ignition, if there is no i
20 wator sunnlied by such auto ignition as recuired i
21 by the imC regulatien guide 1.52.
That in the l
t 22 contention.
1 i
23 A
I have that one de'in as ny nunber lo.
l 24 0
" ell, we'll straichten the nunberinc out later.
25 But could you describe for ns n o'.'
uh a t leahace i
1 I
l
~g i
h l
23
.i fron uhere to whera 'ypassing what filtration 2
systens ara v u talking about?
^
I d
"'t *"'"
- 2'*"h***
I h^"'
" t 1 h*d 3
f 4
at these since I urote then or tried to 5
answer your interrogatories.
6 9
UO YOU h"0" whether or not you intended to reference normal onerations or were vou I
3 talking about the energency conditions'when you
^
talked about leakage bypassing filtration g
i i
10 373t0NS?
s 11 A
I don't renenher that either.
.t 12 0
Do you have A
It sounds like it's talking about nornal 13
(
14 conditions.
I think that I renenher just t
i 15 discussing or rather rtr. copeland asking ques-16 tions on this and that they had, or HLr,P had contained the 20 percent or they had gotten 17 18 it to nuch less than 20 narcent of the leakage 19 On the filterin4-o0 O
"v next cuestion was going to he where does 21 the annlicant adnit that 20 cercent of 'thatever 22 leakage it is vould not he filtered, that's 23 another uav of naking entrv into the docunents 24 so ue can identify the structures or systens o-that you are talkina about.
no 'rou racall uhere--
-2
~
a34
~
i i
26 1
A Page 629 of the safety report.
'Y O
of the original SER?
2
^
I thi'h 8 3
" hat has sone reference, 21 percent g
of the total containnent -- I think that !tr.
D 6
Coneland said that this had been scaled down, in the last denosition.
e I
g O
So vour concern in expressing this contention is based on the infornation recorded on page i
g 10 629 of the SER?
- 11
^
Ri"ht-g 12 Q
And those are tho typical specification 13 linit and the structures and systens you are r
14 concerned about, whatever is referenced on l
15 that nace?
16 A
Yes.
I didn't cone crepared to ansuer questions 1
j 17 on that one.
t la o
Mell, if you find that you need to anend er i
19 sunnienent sonethina af ter'1ards, give ne a 20 call and we'll tr" to work that out.
2.3 A
O . a ".
o Afterwards.
Let ne ash "ou next, is it your 22 23 conc rn that t"is unfiltered leakage :hich you t
24 have identified here will exceed nart 100 or 25 nart 20 radiatien dose linits?
?to",
it
~75 '
k'
27 1
doesn't really sav that in your contention, but 2
sonet'aing that I think is fairly easily 3
inferred, it ticuld heln us i# we could 4
identify if that is the source of your 5
concern?
6 A
It could he, I don't renenher anything about 7
it.
8 q
Mell, you see the thing that is missing is the j
9 vardstich against erbich to ceasure excessive.
10 nnce tre have. id en t,i fi e d what leakage you 11 are talking abo u t,,. th e n tie need an apnropriate
.3 12 henchnark.
13 A
That's true, but I have not done anything 14 with this since I wrote these contentions, 15 which tras in 'fovenher of
'78, which was 'ihat, 16 18 nonths ago.
17 0
So, you don't kno'r if it's part 100 limits vou 13 were talking abcut?
19 A
tro.
20 m
Moula you agree.that that t 'o u ld he an acceptable 21 hench nark to neasure this leahage?
22 A
Yes.
23 m
If the annlicant " LAP instituted redesign er 24
'rhatever so that its relaases ware a lt'av n h e l c'. '
25 nart 100 or nart 20 linitations as is annlicabi.,
I N036
.4 e
iL 2R I
1 1
would that rencve your source of concern-on k-this contention on leakage?
2 i
A If IIL&P cones within the !!RC guidelines,
3 1
f n a
that's fine.
t 5
Q S
1 ng as NRC nests guidelines on this
{
c-P t
.' t 6
l e ah ac. e, then you would he satisfied, is
,r I
l' that correct?
t 7
I l
3 A
Yes.
l 1
r o
O I do need to clarify one other natter.
It's j
10 sort of a separate cart to this contention,
[
e 11 related, but different.than the leakage, the i
,s 12 concerns you oxpress over charcoal adsorbers.
I f
An I t assume -- is it fair to assune that 13 r
14 the charcoal adsorbers you are talking about would l
-r Fl i
j 15 be in that systen which is referenced on page r
.i 16 620 of the SER?
In other words, there is-i s
e t:
17 a direct link hetueen leahage cast the 18 filtration sisten and the filtraticn systen
. (
I 19 that han the filter adsorher?
20 A
I don't know.
i:
l 0
Do vou hava any idea which filter adsorber i
e 1
22 or is it just any filter adsorber?
23
^
I ha"e no idea.
It'n heen too lanc.
2a O
Mell, the ether clue is that you nahe reference 1
4 25 to regulation guide 1.52, "hich coverns q7 b#
l J
-v
,.,...--,,-m..,
,_,,._,.u, w.-.
29 1
certain filtration systens and the charccal adsorbers in those filtration systans.
2 could it he that that is the clue that 3
4 tells us uhich charccal adsorber you ar2 t31hin" 3h "t?
5 6
A It could he, I just do not renenber.
It's 7
heSn too lon?-
8 O
Mell, do --
9 A
I don't ramenher anything, ever getting i
f f
10 interrogatories on that particular contention.
j 11 0
I'll find then for,you, j
12 A
I'VO Got then-13 0
Nell "e
asked y u in the second set of inter-i 14 rogatories, under Interrogatory C,
Inter-
.I 15 rogatory 1.C, specific cuestions about unfiltered i
16 leakage and then you respended with answers 17 regarding this leahage in part 100 dosec.
13 This is your.ebruarv ist annuers.
r T o, A
Okay.
~
20 0
It has to do u!.th this cententien and our 21 interroaateries on thin.
22 A
All right.
o
" hat's US re I got the part inn in the filter 23 24 leahace.
I think nuch earler ue asked sone 25 cuastions akout t' e charcoal adsorber but net
-D3 1
30 1
necessarily in that last go round.
2 A
All right.
3 0
Let ne a s t-yu this, until advised otherwise 4
ey yourself, can "LtP work on the nresunption 5
that the charcoal adscrher you were concerned 6
about is the one referenced in regulation 7
guide 1.52?
8 A
Yes.
9 Q
And in the sane thing, if HL&P complies with f
__ 10 regulation guide 1.52, uould that re ove 11 the source of "our concern?
1 12 A
- Yes, i
13.o one last qu=stion on this one.
Are you 9
14 f aniliar with 7ex Pir(d's positien on an i
15 identical or similar position on charcoal
,6 adsorbers?
17 A
No, I haven't read any of then.
18 0
So there's no connection.
19 f tR. BInnLE:
chat is all I have.
20 Thanh vou very nuck.
21 22 Drenda 'icCorkle 23 24 s )-
4 qg NJ l
i l
UNITED STATES OF AMERICA NUCLEAR REGULATORY CCMMISSION BEFORE THE AICMIC SAFETY AND LICENSING BOARD In the Matter of
)
)
HOUSTON LIGHTING & PCi4ER
)
Docket No. 50-466 COMP.LTf
)
)
(Allens Creek Nuclear
)
Generating Station, Unit
)
No.
1)
)
AFFIDAVIT OF GUY MARTIN, JR.
State of New Jersey County of Bergen I, Guy Martin, Jr., Supervising Radiological Assessment Engineer, Allens Creek Project, for Ebasco Services Incorporated, of lawful age, being first duly sworn, upon my oath certify that I have reviewed and as thoroughly familiar with the statements contained in the attached affidavit addressing intervenor 3renda r
McCorkle's Contention 17 regarding filtration system leakage. All statements t
contained therein,which relate to Ebasco Services Incorporated scope of supply for the Allens Creek Nuclear Generating Station, are true and correct to the best of my knowledge and belief.
/
/
/.
',Y,.,
( -
t,_
l f l
,7 s
I day,o'f
,-1980.
Subscribed and sworn to before me this J4
^
/
['
/
i
/
s.
CAROL A. OPITENOK Nap:1Y PUBUC 08 NEW : PSP W ccuvlSS:p W 9 3 EPL 13 1983 540
UNITED STATES OF A.._RICA NUCLEAR REGULATORY COMMISSICN BEFORE THE ATOMIC SAFETY A'iD LICENSING 30AIG In the Matter of
)
)
HOUSTON LIGHTING & PCWER
)
Docket No.30-466 C09ANY
)
)
(Allens Creek Nuclear
)
Generating Station, Uni:
)
No. 1)
)
AFFIDAVIT OF WALTER F MALEC State of New Jersey County of Bergan I, *4 alter F Malec, Supervising Mechanical Nuclear Engineer, Allens Creek Proj ec t, for Ebasco Services Incorporated, of lawful age, being first duly i
sworn, upon my oath certify that I have reviewed and as thoroughly familiar with the statements conta*.ned in the attached affidavic addressing intervenor Brenda McCorkle's Contention 17 regarding filtration system leakage and that all statements contained therein are true and correct to the best of my knowledge and belief.
5
/
l
/
,1980.
Subscribed and sworn to before me this
-?
day of
' /. ',
.{*
7
'\\
CARCL A. CPITENCK YCIMY PUSUC (F,v,y jcovv MY 00'."';SSI N E.<F.;ts npy, ;3, 933 l
1 541
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATCMIC SAFETY AND LICENSING BOARD In the Matter of 5
S HOUSTON LIGHTING & POWER S
COMPANY S
Docket No. 50-466 S
(Allens Creek Nuclear S
Generating Station, Unit S
No. 1) 5 AFFIDAVIT OF GUT MARTIN, JR.
AND WALTEE F.
MALEC My name is Guy Martin, Jr.
My business address is Two World Trade Center, New York, N. Y.
I am the Supervising Radio-logical Assessment Engineer for the Allens Creek Project employed by Ebasco Services Incorporated.
The statement of my background and qualifications is attached as Exhibit I to this testimony.
My name is Walter F. Malec.
My business address is 160 Chubb Avenue, Lyndhurst, N. J.
I am the Supervising Mechanical Nuclear Engineer for the Allens Creek Project employed by Ebasco Services Incorporated.
The statement of my background and qualifications is attached as Exhibit II to this testimony.
This affidavit addresses the issues raised in McCorkle Contention No. 17.
The contention states that the Allens Creek containment as designed will allow 20 percent of the containment leakage to bypass the filtration systems.
542
I.
Introduction The Allens Creek containment consists of a free-standing steel shell 1 1/2 to 1 3/4 inches thick which encloses the reactor vessel holding the reactor fuel.
The containment is designed to protect the public from the release of radioactive
{
fission products by providing a leak-tight bar*^ ter.
However, for r
practical purposes, the containment must be penetrated by piping and other openings.
Although these penetrations are sealed by some means such as redundant valving, a certain quantity of leakage is inevitable.
NRC regulations (10 CFR, Part 50, Appendix J) limit the quantity of leakage allowed.
j II.
Containment Leakage Expected for Allens Creek The Containment vessel is a seismic Category I steel shell designed to confine the radioactive materials, gases under pressures and temperatures associated with a loss-of-coolant accident and all other abnormal operating conditions.
The design leak rate will be 0.5 percent by weight of the contained atmosphere per day at calculated peak pressure.
The Containment vessel will be designed to contain any leakage from the drywell and the nuncondensable gases from reactor vessel blowdown by the safety / relief valves or from the ruptura of the largest pipe inside the drywell.
To determine the type of leakage which can be expected, a list of all potential leakage paths through containment penetrations was compiled (Table 6.2-12a of the Preliminary Safety Analysis 5d3
.:a I
Report).
This list is reproduced as Exhibit A.
From-this list, only six penetrations constitute potential unfiltered leakage paths.
These six penetrations are listed in Table 6.2-13 of the PSAR and the table is reproduced as Exhibit B.
In arriving at the list contained in Exhibit B, an f
evaluation was made of all lines which penetrate the containment to determine the number and types of barriers to bypass leakage provided for each line.
The types of bypass leakage barriers considered were as follows:
[
(a)
Isolation valve outside containment.
(b)
Isolation valve inside containment.
(c)
Closed Category I piping system inside containment.
(d)
Closed Category I piping system outside
~
containment.
4 (e)
Water seal in line.
c (f)
Line beyond isolation valve outside contain-ment vented to annulus for filtration by the Standby fi Gas Treatment System (SGTS).
(g)
Line terminates outside containment in filtered ECCS Area of Auxiliary Building.
Leakage barriers of types (c) through (g) effectively eliminate any bypass leakage.
Leakage barriers of types (a) or (b) limit but do not eliminate bypass leakage.
Therefore, lines
_3_
i l
l i
544
containing any of the bypass leakage barriers (c) through (g) were not considered as potential byyass leakage paths.
Lines containing only types (a) or (b) were included in Exhibit 3 as potential unfiltered leakage paths.
III.
Unfiltered Leakage _
The amount of containment leakage allowed in the Technical Specifications will be significantly less than that which would produce total off-site doses equal to the 10 CFR 100 limits.
The contributors to this total leakage include the Standby Gas Treatment System releases, leakage to the con-trolled ventilation ECCS area of the Auxiliary Building and all unfiltered bypass leakage.
The actual value of the bypass leakage technical specification will be determined as a result of LCCA f
dose calculations performed when the FSAR is prepared for subnittal.
However, a value of.0195 percent / day of the containment volume j
is the present best estimate of the maximum total unfiltered bypass leakage based on preliminary LOCA dose calculations.
)
r These dose calculations are provided in detail in Section 15 and Appendix 15.A of the PSAR. -1/
IV.
Tests and Inspections In order to assure that the containment will maintain its expected level of leak-tightness, Applicant will conduct a leak testing program in accordance with 1/
The fraction of total containment leak rate technical s'pecification which will be released via potential bypass leakage lines is quoted at PSAR, p. 15.A-4b as 2.9 x 10-2, This number is a typographical error.
The correct value is 3.9 x 10-2,
_4_
5d5 4
i Appendix J of 10 CFR 50.
As required by Appendix J, three types of tests will be performed; Type A - This test will measure the primary reactor i
containment overall integrated leakage rate.
It will be conducted after the containment is completed and ready for operation and again about once every three and one-third years thereafter.
In addition, any major modification or replacement of components of the primary reactor containment performed after the initial leak rate test i
shall be followed by either a Type A test or a Type B test of the area affected by the modification.
Type B - Appendix J defines these tests as those:
intended to detect local leaks and to measure leakage across each pressure-containing or leakage-limiting boundary for the following primary reactor containment penetrations:
1.
Containment penetrations whose design incorporates resilient seals, gaskets, or sealant compounds, piping penetrations fitted with expansion bellows, and electrical penetrations fitted with flexible metal seal assemblies.
2.
Air lock door seals, including door operating mechanism penetrations which are part of the containment pressure boundary.
3.
Doors with resilient seals or gaskets except for seal-welded doors.
4.
Components other than those listed above which must meet the acceptance criteria in III.B.3 of Appendix J. ~
54S
Except for containment air locks, Type B tests will be conducted during each reactor shutdown for major fuel reloading but in no case at intervals greater than two years.
The seals of the personnel air locks will.be-tested after each opening or, if left unopened, at an interval not to exceed one year.
Type C - Type C tests are those intended to measure containment isolation valve leakage rates.
The contain-ment isolation valves included are those that:
1.
Provide a direct connection between the inside and outside atmospheres of the primary reactor containment under normal operation, such as purge and ventila-tion, vacuum, relief, and instrument valves; 2.
Are required to close auto-matically upon receipt of a containment isolation signal in response to controls intended to effect containment isolation; 3.
Are required to operate intermit-tently under post-accident conditions; and 4.
Arc in main steam and feedwater piping and other systems which penetrate containment of direct-cycle boiling water power reactors.
Type C tests shall be performed for isolation valves during each reactor shutdown for major refueling.
V.
Conclusion The Allens Creek containment will be designed to limit leakage to 0.5 percent by weight of the containment atmosphere per day at calculated peak pressure.
Applicant has l
- r1 CV /
i l
l L
1 l
J calculated that, under loss of coolant accident conditions, a i
maximum of.0195 percent per day of containment volume may escape via the potential bypass leakage lines and that the resulting doses will not exceed the limits of 10 CFR Part 100.
- Hence, Intervenor's claim that 20 percent of the containment leakage i
will bypass filtration systems does not reflect the present l
I plant design and the updated bypass leakage fraction calculations contained in PSAR, Section 15 and Appendix 15.A.
Finally, the
}
I projected containment integrity will be assured by performing the leak-rate tests called for by 10 CFR, Appendix J.
l i
T s
5 p
1 i
l 1 54S t
EXHIBIT A N
EVALUATION OF POTENTI.AL 3Y? ASS LEAKAGE FOR CCNTAINMEhT PENITRATIONS Line 3ypass Cons,idered Size Leakage Potential System Se rvice (in.)
Barriers
- Bynass Path Main Stea= Lines 25 A,
3, H No A,
3, C, and D Feedwater A and 3 20 A, 3, E No RHR Pu=p A, 3, and 24 A,D,E,G No C Suction from Sup-pression ? col 21R Shutdevn Suction 20 A,3,D,E,G No Free Recirculation Loop RHR Return A and 3 12 A,3,D,E,G No I'
to Recirculation
\\-
Loop RHR A, 3, and C 12 A,3,D,E,G No LPCI RHR A, 3, and C 13 A,D,E,G No Pu=p Test Lines to Suppression Pool HPCS Pu=p Suction 24 A,D,E,G No frca Suppressica Pool HPCS Pu=p Discharge 12 A,3,D,E,G No HPCS Tes: Line to 12 A, D, E, G No Suppression Pool HPCS Mini =u= Flag 4
A, D,E,G No Line i
1 L?CS Pu=p Suction free' 24 A,D,E,G No Suppression ? col LPCS Punp Discharge 12 A,
3, D, E, G No to Pressure Vessel LFCS Test Line A
D, E, G No 549
EXHIBIT A l
Line 3ypass Considered Size-Leakage Potential System Service (in.)
Barriers
- Sypass Path l
1 Steam Supply the RCIC 10 A,B D
No Turbine and RRR j
Heat Exchanger RCIC and RER to 6
A,3,D,E No Head Spr'ay i
'RCIC Pu=p Suction from 6
A,D,E No Suppression Pool RCIC Turbine Exhaust 12 A, D No to Suppression Pool RCIC Pu=p Discharge 2
A,D,E No Mini =um Flav Bypass RCIC Vacuu= Pu=p 2
A, G No Discharge
.)
CRD Pu=p Discharge 2
A, 3, E No Station Air Supply 2
A, 3
.Yes Instrument Air 2
A, 3 Yes Supply Reactor Building 14 A,3,E No Closed Cooling Water Supply Reactor Building 14 A,3,E No Closed Cooling Water Return Reactor Water Clean-4 A, 3, E No up to Condenser and Radvaste Reactor Water Clean-4 A,3,E No up 3ackvash Transfer Pump Discharge -
Main Steam Drains 3
A,3,E No to Condenser 550
r h
EXHIBIT A Line Hypas s Considered Size Leakage Po ten tial System Service (in.)
3arriers
- Sypass Fath LPCS Minimus 71cv 4
A,D.E,G No Line RHR Pu=p Minimum 4
A, D, E, G No
)
Ficw Line (Typ ' 3) 1 Chilled Water 4
A, 3 E No i
Sys ten Supply 1
Chilled Water System Return 4
A,3,E No Contain=ent Purge Supply 4
A, 3, ?
Yes Hydrogen Purge 4
A,3,D No Exhaust j
Contain=ent Vacuum 18 A, 3, 7 No Relief A and 3 Fuel Transf er Tube 32 A, 3, E No De=inerali:ed Water 4
A,3,E No Supply to Contain-1
=ent Discharge from Fuel 6
A, 3, I No Pool Cooling and Cleanup to Contain-cent Pool Inlet :o Fuel Fool 10 A, 3, E No Cooling and Clean-up frc: Contain-
=ent Pool Condensate Makeup 2
A, 3. E No Supply Drywell Floor Drain 3
A,3 E
No Discharge Header Containnent Floor 3
A,3,E No Drain Discharge 5D1
EXHIBIT A i
-I Line Bypass Considered Siae Leakage Poten:ial Sys ten Service (in.)
3arri e rs
- Evoass Path i
Containnent 'v'entilation 36 A, 3, y
so l
Air Supply and Exhaus:
i, i
Drywell Containment 3
A, 3, E No i
Equipnen: Drains j
i
- Possible Hypass Leakage 3arrier Designation :
A.
Isolation valve outside containeen 3.
Isolatica valve inside containeen:
C.
Closed Category I piping system inside containeen:
D.
Cicsed Category I piping system outside containment E.
Water seal in line F.
Line beyond isolation valve outside contain=en: vented :o annulus G.
Line terninates outside containnent in filtered ICCS area of I
auxiliary building
)
4 1
- i ia I
l 552
EXHIBIT 3 c.
POTENTIAL UNFILTERED CONTAINMENT SYPASS LEAKAGE PATHS Line Description Size (in)
+
Staticn Air Supply 2
Ins:rument Air Supply 2
Contain=en: Purge Supply (2) 4 Main Steam Line Guard Pipe Feedvater Line Guard Pipe Personnel Air Lock
- d W
I EXHIBIT I j
GUY !!ARTIN, JR Supervising Engineer Radiological Assessment l
SCC!ARY OF EXPERIENCE (Since 1963)
Total Experience - Fifteen years participation in Safety 1
Analysis Reports, Environmental Reports, SAR amendments, 6
licensing documents, and cost analysis for insurance i
premita determination.
t Professional Affiliations - !cerican Scciety of nchrital Eng.neers e
Health.?hysics' Society A=erican Nuclear Society.
Intern Engineer in New York State,-
Certificate No. 022127 i
4 Education -
15, Polytechnic Institute of New York, 1976 Nuclear Engineering BE, City College of the City of New York, School of Harvard University School of ?ublic-3 Health, 1977 - Radiological Surveillance Course.
REPRESENTATIVE E3ASCO ?ROJECT EXPERIENCE (Since 1973)
(
i Supervising Engineer j
Participate in,the coordination, technical review and pre-i paration or Sa ety Analysis Repor:s (SAR), Environmentai Reports (ER), SAR amendments and other licensing docu=ents i]
(e.g., Appendix I to 10 C7R 50 studies) for submit:al to the Nuclear Regulatory Commission as part of'the acclication
- or Construction Permit and Operating License of nh' lear i
c power plants.
i Areas of complete responsibility include sections of the SAR l
dealing with the radiological dose assessment work associated l
with normal and hypothetical accident conditions.
In this i
regard, conduct safety reviews of systems, specifications and operatica from a nuclear safety viewpoint and check their compliance with established nuclear safety criteria.
Furnish technical succort in the orecaration of testimonies for safety hearings a' hd ACRS pres'ent' tion.
Study, develop, a
maintain and use appropriate methods, including computer programs for evaluating radiological exposures.
qb )
GUY F).RTIN, JR (Continued)
]
4 PRIOR EXPERIENCE (3 vea: 3) i Equitable Life Assurance Society of the US i
Cost Analys:
1 I
Work involved calculating and analycing cost of various activi ies performed throughou: :he company; assisting j
departmental managers in their budge: prepara: ion work.
Made s tatis tical s tudies for determination of activity costs and providing company's actuaries support informa--
tion for premium determination.
1 Dividend Soecialis:
Reviewed and analyzed dividend and clain reserve cal-culations.
Prepared disbursement authorizations and.
dividend information reports for policy holders.
Parti-cipated in training programs for new employees.
Publications Martin, G and J Thomas _1973.
Meeting the dose recuirements of 10 CFR 100 for site suitability and general des'ign criteria 19 for control room habitability:
a parametric approach.
Transactions of American Nuclear Society 2Lth Annual Meeting, Vol. 28.
Martin, G, D Michlewic: and J Thomas 1973.
Fission 2120:
a program for assessing the need for engineered safety feature grade air cleaning systems in post accident environment.
Proceedings of 15th DOE Nuclear Air Cleaning Con erence.
Leticia, A P, G Martin and J F Silvey 1979. - Implications for nuclear facilities of changes being initiated in the NRC standard atmospheric diffusion model.
Proceeding of the 413:
Annual Meeting of the American Power Conference.
Bhatia, R K, Mauro, J, Martin, G.
Effects of Containment Purge cn the Consecuences of a Loss-of-Coolant Accident.
Transac: ions of American Nuclear Society 1980 Annual Meeting.
555
,3
- nu t u.s 5 unmoo E B 15 C 0 SERVICES I
"E* "'" 3 "S'
E!GIIBIT II 4 Years With E3ASCO nenneaners,
.i Born Philadelphia, Pennsylvania Education Polytechnic Institute of Technology, degree of Engineer in Nuclear Engineering - 1973 lbssachusetts Institute of Technology,13 in Nuclear Engineering - 1970 U.S. Coast Guard Academy, 35 - 1963 Member American Nuclear Society Licensed Registered Professional Engineer in the State of New York (No. 56673)
Experienca:
19S0 Ebasco Services Incorporated, Lyndhurst (NJ) Office; Supervising Engineer, Mechanical-Kuclear Engineering Depart = eat:
Houston Lighting & Power Co - Allens Creek NCS - Unit No. 1-1200 MW(e) EUR Technical and administrative responsibility for =echanical, fire protection, plumbing, HVAC, stress analysis, hangers and supports, and inservice inspection activities. Includes schedules, budgets, and client relations.
1978-1980 Ebasco Services Incorporated, Lyndhurst (NJ) Office; Principal Engineer, Mechanical-Nuclear Engineering Deparencat Houston Lighting & Power Co - Allens Creek NCS - Unit No. 1-12001M(c) BWR, Lead NSSS Engineer Respcasible for preparation and maintenance of ECCS and 30P flow diagrams, piping layours, system design descriptions, inservice inspection provisions, Nuclear Island building general arrangements, PSAR and FSAR preparation, equipment I
sizing and specification, NSSS vendor interface for corre-spondence, drawing review, and contract administration.
1976-1973 Ebasco Services Incorporated, New York Office; Senior Engineer, Mechanical-Nuclear Engineering Department including:
Houston Lighting & Power Co - Allens Creek NCS - Unit No. 1-f200 >M(e) BUR, Lead NSSS Engineer Louisiana Pcuer & Light Co - Waterford SES Unit No. 3-1165 MW(e) PWR, Lead NSSS Engineer (Same responsibilities as listed for 1978-1930 above.)
556 L
EDASCO SI:Bl' ICES 84CSAPS24Tta 1976-1978 Responsible for preparation and =aintenance of ECCS and (Cont'd)
B0p flow diagrams, piping layouts, system design descrip-tions, inservice inspection provisions,' Nuclear Island building general arrangements, PSAR and ?SAR preparation, equipment sizing and specification, NSSS vendor interface for correspondence, drawing review, and contract ad=inis-tration.
1974-1976 United States Coast Guard, Marine Inspection Office, New York; Lieutenant - Supervisory Boiler Inspector.
Responsibility for supervision, assignment and training of Marine Inspectors in largest Marine Inspection Of fice in country. Inspection of hull and machinery material condition of U.S. flag and foreign merchant vessels, and pressure vessels under construction. Application of various laws and regulations of the United Sectes, ASSE Code, ANSI, TEMA, NEC and NTPA Standards. Review of engineering plans and alterations, reports from field and residant_ inspectors.
1973-1974 United S tates Coast Guard, USCGC Spencer (WHEC-36),
Lieutenant - Chief Engineer. Responsibility for operation, caintenance and repair of hull and engineering plant of 6200 slip twinscrew steamship. Direct supervision of 40 officers and men.
Duties included preparation of repair specifications and maintenance of vessel records. : Received Coast Guard Achievement Medal for superior performance of du ty.
1970-1973 United States Coast Guard, Marine Inspection Of fice, New York, Lt and Ltjg - Marine Inspector.
Inspection of hull and machinery of U.S. and foreign flag merchant vessels.
1968-1969 United States Coast Guard, USCGC Mellon (WHEC-717), Ensign, Assistant Engineer Officer.
S 9
4 j
m
1 UNITED STATES OF AMERICA NUCLEAR REGULATORY CCMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD S
In the Matter of S
HCUSTON LIGHTING & PCWER S
S Docket No. 50-466 COMPANY S
S (Allens Creek Nuclear Generating Station, Unit S
S No. 1)
CERTIFICATE OF SERVICE I hereby certify that copies of Applicant's Motions 4, 1979 in the above-for Summary Disposition dated August captioned proceeding were served on the following by deposit in the United States mail, postage prepaid, or by hand-1980.
delivery this 4th day oC August, Richard Lowerre, Esq.
Sheldon J. Wolfe, Esq., Chairman Assistant Attorney General Atomic Safety and Licensing for the State of Texas Board Panel P.
O. Box 12548 Nuclear Regulatory Commission Capitol Station U.S.
Washington, D. C.
E. Leonard Cheatum Hon. Charles J. Dusek Route 3, Box 350A Mayor, City of Wallis Watkinsville, Georgia 30677 P.O. Box 312 Wallis, Texas 77485 Mr. Gustave A.
Linenberger Atomic Safety and Licensing Hon. Leroy H. Grete Board Panel County Judge, Austin County Nuclear Regulatory Commission P.O.
Box 99 U.S.
Washington, D.
C.
20555 Bellville, Texas 77418 Mr. Chase R.
Stephens Atomic Safety and Licensing Docketing and Service section Appeal Board Office of the Secretary of the U.S. Nuclear Regulatory Commission Commission U.S. Nuclear Regulatory Commission Washington, D.
C.
20555 Washington, D.'C.
20555 Atomic Safety and Licensing Mr.
F.
H.
Potthoff Scard Panel 7200 Shadyvilla, No. 110 Nuclear Regulatory U.
S.
Houston, Texas 77055 Ccmmission Washington, D.C.
l
t Mr. Bryan L. Baker D.
Marrack 1118 Montrose 420 Mulberry Lane Houston, Texas 77019 Bellaire, Texas 77401 Stephen A.
Doggett, Esq.
Mr.
J. Morgan Bishop P.
O. Box 592 11418 Oak Spring Rosenberg, Texas 77471 Houston, Texas 77043 Mr.
W. Matthew Perrined Mr. John F.
Dcherty 4070 Merrick 4327 Alconbury Houston, Texas 77025 Houston, Texas 77021 Mr. Jar.es M.
Scott Ms. Brenda McCorkle 13935 Ivy t unt 6140 Darnell Sugar Land, Texas 77475 Houston, Texas 77074 Mr. Steve Schinki, Esq.
Mr. Wayne E. Rentfro U.
S.
- iuclear Regulatory P.
O. Box 1335 Ccc:nis sion Rosenberg, Texas 774" wasnington, D.
C.
20555 Mc Carro Hinderstein 1
2
- J Fannin, Suite 521 Houston, Texas 77002 l
fb $
h C.
Thomas Biddle, Jr.
- 4. N
_