ML20052F271

Testimony of Johns & Ja Bier on Contentions C-M 8 & John III E-2.Certificate of Svc Encl

ML20052F271
Person / Time
Site:	Big Rock Point File:Consumers Energy icon.png
Issue date:	05/10/1982
From:	Bier J, Johns S BIER, MILLS, CHRISTA-MARIA, ET AL
To:
Shared Package
ML20052F261	List: ML20052F260 ML20052F264 ML20052F265 ML20052F271
References
NUDOCS 8205120303
Download: ML20052F271 (19)

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TESTIMONY of Shirley Johns and Jo Anne Bier t

t Testimony #or contention C MP and John IIIE-2; The fuel pool cooling system -curge tank, filter sock tanit, fuel pit l

rumpc, fuel pit heat exchanger, ha-c never been tested in a high temu, high humidity environment following a LOCA.

Even if these proved enerable in a high temp, high humidity snvironment, they would become submerged and fail as the comtainment fills un with ntbr intentionally or unintentionally, CPC neglected to reveal their plan to use the ECCS for SFP make up water until af ter the period of discovery.

CPC s clan waa first refealed in it's motion for summary disposition of Cris' ta' Maria's contention 8 and John O'Neill contention IIIE-2, thareby conveniently avoiding their responsibility to reply to interrogatories. I would like to remind the NRC that 6PC has received at least one life-time axemntion to the single failure criteria of the ECCS, And, if an additional' hurden of the SPF make-up water is addad to an already burdened system, the entire system, exemptions an d modifications must te reevaluated to assure that the health and safety of the workers and the public will be protected. T o understand this we must go tack.

l The A.E.C. requested that all operating reactors comply with ECCS accettance criteria by August 1974. CPC requested an extencion in June,1974

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and was granted until March 31, 1075 to meet these requirements.Then Cpc was granted a further extension until July 1975. When July came, CPC asked for a life time exemption from the ECCS To further thicken the plot.cn Dec. 31,1975the NRC granted CPC two limited exrmptions for I

the 3009. The first permitted reactor operation until March 1,1976 to i

complete the R.D.S. (roactor depressurization system). The second allowed than more time to providd information sunporting why they deserve a life-time exemption. In January io"6, Big Rock shut down to install the P.D.S. and to refuel. They were still shut down on March 10, when the NRC wrote stating that the Mach 1 deadline had expired and Big Rock l

was in non-comuliance with the ECCS acceptance criteria and would not be allowed to start-up unless the commission granted a further appropriate extension. On May 18, 1976, a Mr. C.R. Bilby, then vice-president of i

CPC sent a sworn affidavit to the NRC stating, " Delay of start-up of Big Rock Point beyond May 31, 1976 will require generation of reulacement emergy involving the consumption of additional fossil fuel at the rate of 177,000 cal, of oil or 550 tons of ccal per day at the incraased a

l costs to CPC or it's ratepayers of approximatrly 30,000 dollars per l

day." The NRC wrote lack stating that an exemption to the ECCS could j

be granted only if reasonable assurance is given that oneration of the plant will meet an acceptable level of safety. They went on to say that a plant like Big Rock, which is rela tively small, need not nrcessarilly comply with all the requirements applicable to a larger plant.

1. On May 26, 1076, CPC was grantad a tiant life exemption from the ECCS single failure criteria subject to the following conditions,

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a. Provide evidece satisfactorally d'emonstrating adequate spray distrilution of the nozzle, or

b. Provide an analysis of the ECCS performane, which properly demonstre tes tha t in the event of a break in the core ring l

spray line, the feedwater syctem and the flow thruogh the core sprey nozzle will reliably trovide sufficient core i

cooling water; and enhance the reliability of the core ring suray system by augmented surveillance of the valves and valve actuating circuits, or by othe:r modifications or 8205120303 820510 PDR ADOCK 05000155 M

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b) nrocedural changes which provide reasonable assurance that the core ring spray system can, by itdelf, provide adequate core cooling for a LOCA at a location wher reflooding does not provide such cooling.

The director of tha URC also imposed additional conditions to be met before operation to resume af ter the 1977 refueling outage:

1 Modify the fire protection system such that long term cooling can be accomulished without relying on nortions of it's underground piping.

Wi tnann 's Damarks: Temporarily, CPC got a deferral and installed a fire hose to act as a back-un water supply. Unfortunately, if the back-up water were needed, radiation levels would preclude entry to containment and the hose could not be hooked up. (see attachment from CPC to NRC re: TMI-2 UEREG N0.1 and also NO. 3.) CPC is placing an additional burden of cooling of the SFP on an already overburdened system. If this system in to be used for SFP cooling, it must be re-assessed in all it's asnects, including its lifetime exemption.

2.

Provide test data showing the adwquacy of the nozzle spray system to provide adequate spray distribution during expected use conditions or modify the nozzle spray system to provide adequate spray distribution.

Witness Remarks

• I believe CPC has modified this system to be acceptable with interrenors.

3 Modify the emergency deisel generator and deisel driven fire pump to bypass protective trips during accident conditions except for retention of engine overspeed and generator differential trips, unless additional trips are apuroved by the Director.

Witness remark: SEP TOPIC VIII-3 3 Evaluation: The Big Rock Point control room has no indication 6f battery current, charger output current, bus voltage, battery breaker / fuse status, or charger breaker / fuse status for any of the Class lE DC power systems; it also has no indication for the battery high discharge rate of bus ground for the two deisel battery systems. Therefore, Big Rock DC nower systems monitoring ie not in comoliance with current licensing criteria.

SUMMARY

Of the six parameters currently required to be indicated or alarmed in the control room, only two are alarmed for the 125 V DC and UPS systems and none are indicated or alarmed for the two deisel battery systems. Therefore, the Big Rock DC power systems are not ronitored in compliance with current licensing criteria. See attachment 2.

4 Provide complete on-line testability the ECCS, including the actuation system.

Witness Remark: The valves and pumps associated with the core spray system ar e periodically tested, as is the instrumentation associated with the system. However, there is no systms integrated test required by the Tech. Spec. to deternine the aparability of the system as a whole.

Therefore, the system does not comply to the General Design Criteria 37.

The dasicn of the core snray and the core spray resireulation systems make testing of the systems impractical during reactor operation.

There are no testing requirements for the valves in the flow pathe of the core spray recirculation system nor is there a recuirement for a periodic system intograted test to determinr the operability of the systm as a whole. The system does not comply to the current reactor licesing critein. The point I'm trying to make is that the core spray and the core spray recirculation systems are patr of the

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ECCS. The fire protection system is part of the ECCS, and is also the make-un water system to the SPF. The Systematic Evalual; ion Progrr:m is not evaluating one individual system, but looking at the system sa a WHOLE. I think it is important to stress that. See S.E.P. TOPIC VI-7-A3 ECCS actuation systems Feb. 22, 1982.

CPC says they only need to make un 2gpm of water if the SPF wre to boil. The NRC staff says that a Tech. Spec. change is required to provide an ligpm of make-up water. ' hen, Paulson says that a maximum of 9gpm would boil away. It is these types t f inconsistencies with e-aluations and Risk Assessments that are of cone'rn to the ability of CPC to adequately maintain an acceptable level er aafety with an expanded SPR.

The ony voy applicant can tell how much wter is in the SFP is by visual inspectuon or actuation of an alarm above the SFP If there were a LOCA, or an inadvertant release of radiation in containment, preventing entry into cintainment, there would be no way of knowing the SFP water level.

Staff agrees that CPC must install SFP water level monitors outside of containment. Also, all related equipment is operated manually. The SFP cooling pumps and SFP make-up water modifications described in Licensee's responses to Inter. 8-10, 8-11 can be remotely manually operated actuated from outside containment. The rest of the SFP equipment is actuated or operated within containment. - CPC answer to Inter. B-8b (What equiument must be operated manually)

Maximum containment pressure is 23psiq at a temp. of 224'F. Containment sprays are necessary to condensate the steam and return the containment building pre *sure to normal following a LOCA. Motor-operated valves MO-7064, MO 7068 have never been tested in a high temp, high humidity environment and may fail. See David Blanchard Sumarry Disposition.

The criticallity analysis performed by Dr. Kim is based on water temp in SFP at 212'F, assaming boiling of the SFP, with the containment at atmospheric prensure. Even assuming the containment is at atmospheric pressure, the pressure at the bottom of the SFP,9 due to hydro' static load is 28.14psiq. The boiling temp. at that pressure is 247'F. Since the effective activity coefficient K is not permitted to exceed 0.95and since Dr. Kim's calculations reached this maximum, assuming 212 F, 'it is questionable if the calculations can be considered conservative.

-ASLB Order Feb. 1982 e

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Testimony for C-2, O'NeillIIA I will be using the B.R.P. Health Physics Appraisal Document dated June 13,19P1 and Institute of Huclear Power operations " Evaluation of B.R.P." dated Aug.1981 for most of my testimony.

In addition I want to emphasize that the radiation levels above tha SFP at B.R.P. are 10x the level at most reactors, (2mr/hr). B.R. 's radiation level is between 14mr/hr-24mr/hr plus the radiation from the fuel pool sock filter adds up to a radiation level of around 38mr/hr.

Licansee Answers to Oral Inter. og. 156-157. B.R. has higher dose rates than any other power plants. Newar plants have a radwaste domineralizer set aside to process only spent fuel.

Pg.15*, Axtell states that Licensee doas notnplan on using the demineralizer with any amount of frequency, even though radiation levels above SFP went down considerably from 14-24mr/hr t" P-15 mr/ht. Nor does he think it is cost-effective to spand 1/2 million dollars on a dcmineralizer to be used only for the SFP.

Pg. 130 Axtell; Dueto shielding and capacity tha liould radwaste system would be of limited usefullness in accidents whoch generated large quantities of high activity in water.

Mackie Do you foresee any situations where you would have a generation of large quantities of high activity water that would tax the system?

Axtoll-No, I don't, because I don' t ever expect an accident of the magni tud e of TMI-2 to occur at the Big Rock Plant. It's a well known fact that this nlant is not equipped to immediately handle an accident where large quantities of water are generated. So in which case there would have to be designed and built special ntorage tanks.

So fnr we hava been lucky. But with the spread of industrialization, and with a greater nunber of peonle moneying around with a thing they do not comnletely understand, sooner or later a fool will prova greater than the proof, in a fool proof system.

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taking into account shielding afforded by structures within and outside the sphere. Site maps showing approximate isodose lines for the first half hour are included herein.

Actions which must be taken af ter an accident have been reviewed to determine if modifications are required in order to limit personnel radiation exposures f

to 25 Rem in the 30 days following an accident. This review considered actions which necessarily must be performed (such as manual switchover from injection to recirculation mode of core cooling) and actions which might be required in the event of certain failures. The modifications considered necessary to ensure the desired limitation of personnel exposures using the assumptions specified by NUREG 0578 are:

I

. Backup Emerrency Diesel Generator The backup emergency diesel generator is currently mounted in a semitrailer located on site.

If needed, it is moved to the required location and connected to the plant emergency power bus.

Evaluation ei this operation under the above specified post-accident conditions indicates personnel radiation exposures above 25 Rem could be received performing these actions.

The backup. emergency diesel generator will be relocated to an area near the primary diesel generator.

Modifications will be made to enable the e(

backup to be placed in service remotely from the control room.

Backup Cooling Water Supply to the Core Spray Heat Excher g 4

Long-term core cooling af ter an accident is accomplished by recirculatin; water from the containment building sump through a heat exchanger and back NO to the core via the core sprays.

Cooling water for the heat exchanger is ggA supplied by the fire water sysym.

Plant Technical Specifications require a backup hose to provide cooling water to the core ay heat exchanger in D. M #

the event of a failure of the buried fire main.

ation of tais syst W i

g ioM has shown that personnel exposures above 25 Rem could be~ received 7

g to installing the backup hose under the above specified post-accident l.

conditions.

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Q geh d A permanent core spray heat exchanger water supply line, separate frtm the current buried fire main and fulfilling the function of the backup hose, g [, will be installed.

This line will be capable of being placed in service remotely from the control room.

Emergency Diesel Generator Fuel Supply The emergency diesel generator currently has a fuel supply capable of

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sustaining operation at full load for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Evaluation of the refueling operation under the above spec,ified post-accident conditions indicates personnel radiation exposures above 25 Rem could be received if refueling occurred 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after an accident.

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SYSTEMATIC EVAL.UATION PROGRAM TOPIC VIII-3.B DC POWER SYSTEM BUS VOLTAGE MONITORING AND ANNUNCIATION BIG ROCK POINT i

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1.0, INTRODUCTION

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The objective of this review is to determine if the DC power system

-bus voltage monitoring and annunciation are in compliance with current licensing criteria for Class lE DC power systems.

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The specific requirements for DC power system monitoring derive from j

the general requiremen,ts embodied {n Sections 5.3.2(4),5.3.3(5),gnd 5.3.4(5) of IEEE Standard 308-1974, and in Regulatory Guide 1.47.

In summary, these general requirements simply) state that the DC system (bat-teries, distribution systems, and chargers shall be monitored to the extent that it is shown to be ready to perform its intended function.

2.0 CRITERIA i

powersystem(s)statusshallbeprovidedinthecontrolroom:gClasslEDC As a minimum, the following indications and alarms of th o Battery current (arxneter-charge / discharge)

.1 Battery charger output current (ammeter) o o DC bus voltage (voltmeter) o DC bus ground alarm (for ungrounded system) o Battery breaker (s) or fuse (s) open alarm o Battery charger output' breaker (s) or fuse (5) open alarm i'

3.0 DISCUSSION.AND EVALUATION 3.1 Discussion. Big Rock Point utilizes seven Class 1E DC power sys-tems: one 125 V DC system comprised of one battery and two chargers; four Uninterruptible Power Supplies (UPS), each comprised of a battery and

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batteryandcharger.geratorbatteryandcharger;andadieselfirepump charger; a diesel ge Control room monitoring of,the 125 V DC system consists of a "125 V D-C System Trouble" alarm which actuates on

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battery / battery charger overcurrent, positive or negative bus ground, loss of charger input supply voltage, or 125V DC bus undervoltage; local indica-l tion consists of charger output current and bus voltage, current, and ground. Control room monitoring of the UPS consists of a "UPS Abnormal" alarm; local indication consists of battery output current, charger output current and voltage, inverter input current, and invs ter output current, l

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qr Control room monitoring of the Diesel Generator and voltage and frequency.

Diesel Fire Pump systems consists of an " Emergency -Generator Engine Trouble" alarm and a " Diesel Fire Pump Trouble" alarm, respectively.

3.2 Evaluation. The Big Rock Point co'ntrol room has no indication of battery current, charger output current, bus voltage, battery breaker / fuse status, or charger breaker / fuse status for any of the Class lE DC power systems;. it also has no indication for battery high discharge rate or bus Therefore, Big Rock Point DC ground for the two diesel battery systems.

power systems monitoring is not in canpliance with current licensing

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4. 0. _S UMM ARY Of six parameters currently required to be indicated or alarmed in the control room, only two are alarmed for the 125 V DC and UPS systems and none are indicated or alarmed for the two diesel battery systers. There-fore, the_ Big Rock Point DC p_owerisystems are_not monitored in compliance i

with current Tlcensing-criteria.

5.0' REFRENCES 1.

IEEE Standard 308-1974, " Standard Criteria for Class lE Power Systems

, for Nuclear Power Generating Stations."

Regulatory Guide 1.47, " Bypassed and Inoperable Status Indication for 2.

Nuclear Power Plant Safety Systems."

3.

Letter, NRC (Crutchfield) to Consumers Power (Hoffman), "SEP Topic VIII-3.8, DC Power System Bus Voltage Monitoring and Annunciation, Safety Evaluation for Big Rock Point," dated August 15, 1981.

4 Letter, Consumers Power (Hoffman) to NRR (Crutchfield), "SEP Topic VIII-3.B, DC Power System Bus Voltage Monitoring and Annunciation," dated January 7,1982.

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s This system includes an estimated 1000 feet of buried cast iron

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l yard pipe, six fire hydrants, several fire sprinkler systems and

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an automatically actuated deluge valve for fire protection of the Y

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, station transfomers.

Numerous other smaller pipe,sdare connected to the system. CPCo has submitted an analysis which indicates that the fire protection system has the capacity to provide sufficient ECCS t

cooling water flow except when the station transforner deluge valve t.

is open.

For this case, ECCS cooling water flow is marginal.

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The acceptability of the fire protection system asA41nk the ECCS is depend.gnt on three considerations: 1) physical protection N

for the exposed _pipi g in the tutbine building and for the station transfor=er deluge valve to assuie that severe da= age cannot occur due to overhead crane operation or moving of heavy eq0ipment, 2) an interlock,or acceptable physical constraints on a

' manual operation of the deluge valve to ensure that the valve re=ains shut during a LOCA, and 3) a review of the piping specifications, joint t

type, and test schedule and procedures relating to the plant fire

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protection system. Staff review will detemine whether system upgrading i

1 is required to provide the needed reliability for this vital link in ECCS.]Untilfurtherinformationontheseconsiderationsisdeveloped 1

! I and reviewed by the Staff, the continued acceptability of the fire 1 i l

protection syste: for ECCS use cannot be determined.

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em m

4.5L SM

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.0TED 5148-SA I -:

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l,f_ ', j ' '

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FILE G-RA-12

[

. c,,..,

I ADDElDUittio. 1 DATE i

L i a CG

..-t_..,e.. v.N BRP. SFP, CPC0

!. 4

/

BY

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LIEF I

- FILL f!O.

Ch'ched Cy SUCJECT Ne f~'evrW w

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TABLE 12.1 -(fnrfid{

QSC, SU!.tMARY CT CRITICALITY A::AI.YCIS RESULT 3 (Big Rock Point Plant) l l

r j MDid hk k

1.

Nominal ec!! PCC-7 %

0.8956 /

9'x 9 in. pitch, 7 in can ID 0.250 in. S.1 c.in, 68 F Modified G-3 fuel (3.30 w/o) o.a ry 2.

Pool temperature (68 " 212 r) 0.0156-

+0.0036 3.

Tuct enttchment (+2i',)

'O.0192

o. o u er y

~

4.

Tolerances Rac). pitch (-0.12 in.)

0 bl60 i

Can slac ('0.03) 0.0013 /

Can wo!! thickncss (-0.010 in.)

0.0022 /

0.0014 /

Can S3 composition Eccentric fuel !cading 0.0158 v' 0.0052 /

+

' Minimum water qcp 4

0.0233 (r.m.s)(2k

~

ANf6's 5.

Total Ak bia s (2. + 3. + 4.)

0,0425' g*0VL7

6..

Maximum k. PCC-7. 212 T (1. + 5.)

0,938 a.f 3r.3 7.

KENO-IV/PCQ-7 bi.a s (C3 F)

-0. 00 92 /

8.

KENO-IV bencht :rk 0.0086 /

uncertaintle s @v-)

40,000f /

0.0167 /

9.

Maximum k,1:1 M(' tv212 r Maximum bot!!n:(20.6% void) 0.0044* #,

0,,345 7 0.1VIF (G. + 7. ' ti.)

10.

11.

!.!aximum k ';C:C-IV. 20. 6?', vold (212 F)

I 4iiC2 (9. + 10.)

0,.9500 wa_fj-

?

T Note:

(1)

Pe cult :: 3ro for Medified C-3 fuel (No Zr reds, no watar he!cs) with 1.r?0 w 'c U-235 (with unifer-. enrichecnt :stt: t.ttc ;.

e (2)

Alcetrale sum of tolerances Ak's gives 'O.0419.

Analy:13 indicates that the maximum vcid under the ecmp!cte shutdown cf ccoling systcmsis 20.63.

sfr3&i

<~eq

---#-.,----.-.-----.---3.

-.r--#

-.---gw-r,e-mw---v.-,m,--,-ve-r--,+-,-

.-,v--_r e+

-m-w

---+o

,m,----,w-.

--g-i-*,+w-

---3-m w

....- _1 0,2. 5

c..

rT D' 9' A FIL3 G-R's-10 *.

F e but

~

' ~ '

.L. 'J 9J.]-RE Aniu ::at:': :ro. 1

,/_.,,,,

'~'

_ CCPPCRJrCry

?2,3 C'. 9,, CI'CC DATE

)

O fili fiU BY I

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CLIEtJT T'

SUBJECT Checked By Swwy OF /Temtrs (Ans~c'en a.o

&/>.&/ swimyor7esca 7: da he<?

v;sd i

4.w./ c.- n.%. ;<.:':a ni,i of a/,//:,: A;r R iT Ge<<<cd>.:.

,%,. w i s ey n r.y h 6 s b u d e i u x;n i, is e k /4 g L e w

o, c/So,.<:did.Wh s) ore /b epl'hi/s/ o.2S~

7!c u

inb.)/ 5x : wry c/ /e:at': w:// we30er' c-aus d= ar e s

/

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N 7/c a,,c4,,at cc,,.c5oaa%; i 4 in>:s,cii.i ca& A sf e qs>o & cahAbdJ, fhe eer;/ er ad doeso an

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;;,.,.' yy [e,,sj,. i,pp a',,/g,., Q,3,,,,.,,. Qa 7',),,

p,

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p.ru i.f 7;,,.,,.i ce-:,.,-v d'/r;//nsS<:.

4'c i; cay,n/eeH W2-7 g$ n :. es : l. ~,., s. 2?5/s f/n.z es >j/,

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9 r

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i __i CORPCRAT;CN a

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CLIENT C

fitEtJ0.

BY

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SUBJECT Checked By 7he nv&es'(ei<es/ed> swwy f ww/h da:eo'sn genew(: s,txt;,ikdas a,prewitVm 50e /.7./(d~"be ss 4.ow/p;e.

M i e,d & a /e M J m s xhi, w.,

enricL,edjbi kdy%/ G-2 i= 3. 79& '% (/.2:53

,/ c.

$, f0 Y ".E & /

0

[s.*i : / ~+ % ^ C '

c f,

o 4

/ $,.4 ( k w i d, d ).

resw/6 sie ss/ay/ie.cd/e ii the Mr; des!

Ts L4 a. t >.<ccptYhaane/;ex/pake: aye

-/f C<nk Ak } &;(/b?//f2 9 l8y' Gre ridC W

i

/ l

/

M[ient ca dete.

! s:.dhi s

e i

6 bN

_I

.-m

r, ----. -. - r.

r,. - A NPTED 51:8-SA l..ne

• of f r

{ ^r " [ I \\. 3 t~-./ i.- ;,)

<, ;.a, G,..d. r 1 ~,,

I i

CCSPC.tMTCN

/3 W CM N0- 1 77 j

lillP, SFP, CPCO DATE

)

Ct.lE f1T I-[#/7 TASK NO. O N -

By O

V )I, W!

SUBJECT Checked By Il TABLE 12. l (Revised)

SUMMAT0' CT CPITIC/d.ITY ANid.YSIS !!CSULTS

(!319 Pock Point Plant)

3,j

__b k k

1.

Ncmin 21 c _Il PCQ -7 %

9 x 9 In. pitch, ' in. ca n ID 0.8954 o

0. 2's 0 :n. :23 ca n, 63 r Modifi-d G-3 ruel (3.30 w/o) 2.

Pool temperature (68 -> 212 F) 0.0151 3.

Tuc! cn ichment (+20',)

+0.0016 0.0194 4

Tolerances Rack pitch (-0.12 in.)

0.0160 Ca n size (+0.03) 0.0013 Can wall thickness (-a.010 in.)

0.0022 Can SS cemrosition 0.0014 Eccentric fue! Icading 0.0158 Minimum water <;ap

+0.0052 0.0233 (r.m.s) 5.

Total Ak bia s (2. + 3. + 4.)

0.0427 6.

Ma ximur-i L. P DQ -7, 212 P (1. & S.)

0.9381 7

K E::C-IVe F N -7 Ha - (6S F)

-0.0092 j

8.

KENC-P/ Imnch-trk 0.0086 un cu t a int p : (';.7) 3 0081 O.0lG7~

1 o

l 9

Maximum 4. E '.'? - P! ~:12 P (6. + 7. + 9. )

0.9456 r

g 10.

Maximum bett in; A' 2 3. M void 0.0044*

5 11.

Maximum *,. ??C -I., 20. 6!!, void (212 F)

(9.

• 10.)

0.9500*

+

Note:

(1) i? cal: ar fer !.!cJiflod C-3 fuct (::o 2r rods, no.ater holes) w t5 3. 3. /o U-225 (witn unticrm carichment distributicn).

(2)

A!, L:lle um of tolerances 4%'s gives *0.0419.

{

I Ar.a!ym. Indicates that the ma::tmum void under the completc

.,h u: t cn of cc. cling system:is 20.6;',.

'j m.,.; w --... -, - -.-,.......

....-.-.i

CERTIFICATE OF SERVICE I certify that on the

[

day of May, 1982, the above Testimony was served on the attached list by United States Mail, first class, postage prepaid.

/

/e HERBERT SEMMEL

\\

f f

I i

i l

m C.

(

ll At:.ic safety and Licensing JerspF Colle, Enquire E:ard Panel Ishe, Linen 3n and Beale 3

1120 Conne::icutt' Ave, N.W.

U.S. Nuclea r Re gulatory Suisc 325 Cor.m:ssien Washingten, D.C.

20555 Washington, D.C.

20036 Peter B. Bloch,. Esq., Chairman Atomic Safety and Licensing Board Panel

, t.,k U.S.' Nuclear Regulatory l

F.ti Commission Washington D.C.

20555

{

.jf

\\

d' f

i Dr. Oscar H. Paris Atomic Safety and Licensing l

Board Panel U.S. Nuclear Regulatory Docketing and Service Section Commission of fice of the Secretary Washington D.C.

20555 U.S.

Nuclear Regulatory Commis sion Mr. Fredrick J. Shon Washingten, D.C.

20555 Atomic Safety and Licensing

~

John O'Neill, II Bodrd Panel Reute 2, Box 44 U.S. Nuclear Regulatory Maple City, MI 49664 Commission Washington D.C.

20555 Janice E.

Moore, Esq.

Counsel for NE Staff U.S. Nuclear Re gulatory Co r.r.i s s i_c n Washington, D.C.

20555 i

.sn>>

'g<W}%.Q.:f*k *l vf.N 'P,.

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