ML20129H840
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UNITED STATES jf"s,eree JUCLEAR REGULATORY COMMISSION CEGICN 84
['
g 101 MARIETTA STREET.NN.
e ATLANT A. GEORGI A 30323 j
JAN 19 390 L
RI TION -
OT FOR LIC g
Request No. RII-90-02 i
T0:
James Y. Vorse, Director j
Office of Investigations Atlanta Field Office, Region II FROM:
Stewart D. Ebneter Regional Administrator t'
i REQUEST FOR INVESTIGATION J
Georgia Power Company 50-424, 425 Licensee Docket No.
t l
Vogtle Electric Generating Plant RII-90-A-0005 l
Tacility Allegation No.
BACKGROUND The following information, which is paraphrased, was in an anonymous letter received by the Resident Inspector Vogtle Electric Generating Plant (VEGP) on January 9,1990:
I On October 13, 1988, at approximately 10:30 a.m., and again at 4:40 p.m.,
VEGP was willfully and intentionally placed in a condition prohibited by Technical Specifications (TS).
At the time of the incident, Unit I was j
shut down for a refueling outage and the Reactor Coolant System (RCS) had i
been drained to midloop at'188 feet. Valves 1-1208-U4-176 and 177, which are required to be locked closed, were opened while the RCS was at midloop, a condition prohibited by TS 3.4.1.4.2.
By opening the above identified valves, a flow path was created which resulted in unborated water flowing from the Reactor Makeup Water Storage Tank (RMST) into the RCS.
The above action placed the unit in an unanalyzed condition and constituted an unreviewed safety question in that the path--RWST to RCS, had not been l
analyzed for a boron dilution accident by Westinghouse for " Mode 5 with l
. reactor coolant loops not filled" conditions.
The unit was placed in this condition by Mr. Kitchens, Plant Operations Manager, who on at least one of these occasions personally opened the valves because other licensed personnel had refused to do so. Valves 176 i
i and 177 were opened so that the chemical hydrogen peroxide could be added for chemical cleaning.
This had been planned to occur in the outage i
schedule prior to the unit reaching midloop, and, in fact, two additions were made prior to reaching midloop.
Due to coordination problems, the Operations Manager was faced with the need to add the chemical at midloop.
In the interest of scheduling, the Operctions Manager decided to order tu
[y addition.
h U51.1C DISCLui M PROVAL
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fojiogi 960827 i
KOHN95-211 P DR - -.
1 7
LIMt1f03STR 0
JAN 191990 I
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Sometime around September 12, 1989, other plant personnel discovered the referenced event and began looking into the matter.
The Operations l
Manager subsequently wrote a letter to the General Manager (NOV-00385) j explaining his understanding that an Imediate Action Statement can be voluntarily entered and that the unit was not really at midloop during the 4
chemical additions made on October 12 and 13,1988. The latter contention is false based on the Unit 1 Control Log.
The former interpretation is contrary to the position taken by NRR which has interpreted "imediately" l
to mean "without delay" or "as your next action." It is also contrary to the position stated in NUREG/BR-0110. Issue No. 87-2, August 1987, regard-ing voluntary entry in TS 3.0.3.
As a result of the issue of voluntary entry into Immediate Action State-A ments, this matter was referred to SONOPC0 corporate for review.
)
position (ELE-00919) was issued and stated "... voluntary entry into an LC0 which expressly prohibits a given condition and requires inmediate correc-
}
tive action should that condition exist, should not be made...."
Based on i
that policy, the event has never been reported to the NRC under 10 CFR 50.72 j
or 50.73.
In the case under discussion, the violation of the LC0 goes far beyond the l
issue of " voluntary entry" and the interpretation of "imediately."
Unborated Valves required to be locked closed were intentionally opened.
i water was intentionally flowed into the RCS for the purpose of adding l
chemicals causing the dilution of the RCS boron concentration.
This created the possibility of uncontrolled boron dilution of the filled portion of the RCS and a subsequent inadvertent criticality accident--the I
very condition stated in the TS bases to be guarded against. Fortunately i
the above scenario did not develop, but because no analysis or testing had been perfonned, the safety of the plant was placed in the hands of luck.
l It would have taken only one single failure such as a wrong size orifice or leaving the valves open too long to escalate the event into an accident Finally, this unnecessary risk was with serious safety consequences.
i taken in the interest of schedule without requesting engineering evalua-tion, without af equate monitoring and controls, and without contacting the l
NRC for consideration or the advisability of discretionary enforcement.
On November 11, 1989, Georgia Power Company (GPC) submitted a TS change f
The request (ELV-01077)toallowopeningvalves176'and177atmidloop.
stated purpose was to allow for chemical additions at midloop in Mode 5 Since GPC has now deemed a TS change necessary for subsequent 4
and Mode 6.
opening of the valves, it has tacitly admitted the TS violation which l
occurred on October 13, 1988.
Notwithstanding the $50,000 to analyze this previously unanalyzed condi-l tion, no disciplinary action was ever taken against the Operations Manager for his cavalier approach to the requirements of regulations and the safe operation of a nuclear power plant. In fact.he was promoted to Assistant General Manager.
The details of this event are known to all line manage-j ment up to and including the Senior Vice President of SONOPCO.
l 4CTISCLOSURE W/0JLAITROVAL LIM I.
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6 01(TRIB# M M ISCt057RE x
3 JAN 191990 i
This event, as well as numerous other events such as: failure to perfom l
1 i
adequate PORY surveillance; failure to perfom adequate steam generator blowdown valve surveillance; failure to perfom adequate E-bar surveil-lance; installation of micro'11tration test unit designed in violation of J
Regulatory Guide 1.140; failure to perfom adequate shutdown margin calculations; failure to report missing fire penetration seals; and other matters involving interpretation of regulations and reporting of events have all served to signal to plant personnel that management concurs with i
this cavalier approach to regulations. A quick review of recent trends at VEGP-indicate a return to the conditions that existed in early 1987 when l
Under 50NOPCO, they have been allowed, j
the same people were "in charge."
if not encouraged, to return to the operating philosophy of that troubled
}
If this is not kept in check and reversed, the quality of opera-period.
tions will not improve and there could be a serious event at the plant.
j' The original of the above letter is being retained in Regional files.
l A.
Request What is the matter that is being requested for investigation?
3 Based on the infomation detailed above, an investigation is urgently needed to determine if Mr. R. Kitchens, Assistant General Manager - Plant Operations, intentionally violated VEGP Technical Specifications on when he allegedly opened locked valves in the RCS.
October 13, 1988, Although the actual safety significance of the' alleged event is minor, the j
alleged willful disregard for Technical Specifications is viewed as a i
concern of utmost significance, particularly in light of the individual's senior management position at VEGP.
Such willful disregard by a senior manager, if true, raises serious questions regarding the individual's suitability for such a position.
h B.
Purpose of Investication What is the basis for the belief that the violation of a regulatory l
1.
requirement is more likely to have been intentional or to have resulted from careless disregard or reckless indifference than from i
error or oversight?
Although the above infomation was received from an anonymous alleg-2 er, the content, tone, and quality of the infomation is such that a high degree of credibility appears warranted. The staff has verified through submittals from the licensee that the above described inci-
]
It would be a prudent course of action to pursue and dent did occur.
develop the extent of Mr. Kitchen's involvement in this matter in an attempt to either prove or disprove the allegation of wrongdoing.
a' What are the potential regulatory requirements that may have been l
2.
violated?
10 CFR 50.50(a)(1),10 CFR 50.54(x),10 CFR 50.72(b)(ii)(A), and TS 3.4.1.4.2.
l
,Am M W
1 O TED D MTRIBli 12LI 3
4 JAN 191990
-3.
If no violation is suspected, what is the specific regulatory con-
~
cern?
N/A 4.
Why is an investigation needed for regulatory action and what is the regulatory impact of this matter, if true?
If the allegation is true, significant regulatory action would be initiated and could include removal of the individual from licensed activities, i
C.
Requester's Priority 1.
Is the priority of the investigation high, normal, or low?
HIGH i
2.
What is the estimated date when the results of the investigation are i
needed? February 9, 1990 3.
What is the basis for the date and the impact of not meeting this date?
~
This invest.igation should be initiated innediately and considered a matter of utmost urgency.
If this allegation were substantiated, based upon the high level of manager involved, a serious question would be raised its to site mangement's capability to safely operate VEGP.
[
D.
Contact 1.
Staff members:
L. Reyes, Ext. 15179 K. Brockman, Ext.16299
- 8. Urye, Ext 14192 i
f 2.
A11eger's identification with address and telephone number is not confidential. N/A 3
E.
Other Relevant Information A review of the original document indicates that the anonymous alleger appears credible. Although the details provided in the Background section above are paraphrased from the original document, they do portray the tone and content of that document.
Paraphrasing was used to provide some degree of protection to the alleger.
There is no indication of any vindictive motive on the part of the alleger. This method of communica-tion with the NRC may indicate other problems not within the scope of this request.
Evidently the alleger is a person who occupies a relatively important position at VEGP which is indicated by the detail of the informa-tion provided.
STRIB W T C 5
- N 191990 You are requested, af ter your review of the original document, to fully develop all logical investigative leads and expand the scope of the i
The staff is available i
investigation as required to address those leads.
to provide any assistance you may require.
</d j
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Stewart D. Ebn ter j
i 1
cc:
H. Thompson, OEDS B. Hayes, OI J. Partlow, NRR
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J. Goldberg, OGC
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J. Lieberman, OE 8
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&TR ON -- NOT C DISCLOSUR APPROY
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) wwwe Resident Inspector Nuclear Regulatory Cosutission P.O. Box 572 Waynesboro, Georgia 30830 9
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To: The Nucle'ar Regulatory Commission, persuant to l
Employees' potential Sa f e t.y Tasues j
On 10-13-88 at app r<e. i ma t.e 13 10:30 asd again at i6:40. CST Plant Vogtle Unit 1: 't operated by the Georgia power Compa'ny) l was willfully and intentionally placed in a condition prohibited by'it's Technical Specifications. ALLthe. time Unit 1 had just. been shut down for a refucJing. outage and the RCS had been drained.Lo 188'-0"(midloop).
Specifically valves 1-1208-U4-176 anc. 177 which are J
required locked closed were opened while the RCS was at midloop, a condition prohibited under Technical Specification ( 3.4.1.4.2-).
.Uy opening valves 3
1-1208-U4-176 and 177 a. flow path was created resulting in unborated water 1 flowing'from'the RMWST into the RCS.
In addition, the above action placed the plant in an unanalyzed condition and-cons t i t.u ted an unreviewed safety
^
question since this path (RMWST to the RCS) had not been analyzed.for a boron dilution accident by Westinghouse for HODE 5 with reactor. coolant loops not filled conditions.
j The plant ~was'placed in this conditir.n by the Operations Manager, Skip.kitc. hens, who holds a ERO license. ON at least one of these occasions he personally opened these valves l
because other licensed personnel.had refused.
Valves 1-1208-U4-176'and 177 were opened to add the
- chemical hydrogen peroxide for chemical cleaning.This had been planned by the outage. schedule-to occurr prior to i
reaching midloop and in fact.-2 additions were made prior to reaching midloop.Due to coordination problems however, the
- Operations Manager was faced with t.he need t,o add at midloop.In the int e res t. o f s h. du le he decided t.o order the
- addition, j
On about.9-12-89 other plant personnel discovered the
]
. reference event and began investigating The Operations.
Manager wrote a letter to 1.he General Manager (NOV-00385) j explaining his int erpretation ' that as, immediate Action statement can be voluntarily entered and that the plant was not really at midloop'during the additions on' 10-12-88 and 10-13-88.The'latter contention is false based on the
- Unit 1 Control Log.The former interpretation is contrary to the position taken by NRR of'the NRC which has interpreted immediately as "without delay" or "aa your next action" and the position t.aken by the NRC (NUREG\\BR-0110, ISSUE NO.87-2, AUGUST-1967) on vt.lunt.ary entry into Specification
- 3.0.3.
.The' issue of volut.tary entry into Im.nediate Action statements ~was' referred to SONOpCO corporate within about a; week.A positiot. ( El.E-00 919 ) was 15. sued which stated,
" voluntary entry into an LCO which e>pressly prohibits a
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given condition and requires immediate corrective action should that condition exist should not be made", but came prohibited. Based on that short of conclu~ ding that it was policy, this event has never been reported to the NRC under 10CFR50.72 or 10CFR50.73.
In the case at hand, the violation of the LCO goes far beyond the issue of " voluntary entry" and the intrepretation of "ir. mediately". Valves required locked closed were intentionally opened.Unborated water was then intentionally 4
flowed into the RCS (for the purpose of adding chemicals) diluting the RCS boron concentration.This created the possibility of uncontrolled boron dilution of the filled portion.of the Reactor Coolant System and a subsuquent inadvertent criticallity accident the very condition stated in the bases of Technical Specifications that is to be 4
protected against.Fortunatly the above accident did not but because no analysis or testing had been done the
- occurr, l
safety of the plant was placed in the hands of luck.It would have taken only one single failure (such as a wrong size orifice or leaving the valve open too long ) to escalate this event into an accident with serious safety significance. Finally this risk was taken in the interest of schedule, without requesting engineering evaluation,without adequate monitoring and controls,and without contacting the NRC for conoideration or the advisability of discressionary enforcement.
ON 11-21-89 Georgia Power submitted a change (ELV-01077) to Technical Specifications to allow opening valves 1-1208-U4-176 and 177 at midloop.The stated purpose is to allow for chemical additions at midloop in MODE 5 and MODE 6.
Since Georgia Power has now deemed a change to Technical Specifications necessary for subsequent opening of these valves it has tasitly admitted the violation of j
Technical Specifications which occurred on 10-13-88.
Despite costing approximately $50,000, to analyze, this previously unanalyzed condition,no disciplinary action was ever taken with the Operations Manager for his cavilier approach to the requirments of regulations and the safe operation of a nuclear power plant. In fact he was subsuquently promoted to Assistant General Manager.
i The details of this event are known to all line management up to and including the Senior Vice President SONOpCO.
This event as well as numerous other events:
Failure.to perform adequate PORV surveillance Failure to perform adequcte Steam Generator blowdown valve surveillance Failure to perform adequate E-bar surveillance Installation of Microfiltration test unit designed in violation of Reg. Guide 1.140 Failure to perform adequate shutdown margin
e calculations Failure to report missing fire penetration seals involving "intrepret.ation" or regulations and reporting of events has signaled to plant personnel, management's concurrance with this cavilier approach.to regulations.
A quick review of recent trends at Plant Vogtle indicates a return to the kind of conditions that existed in early 1987 when the same people were "in charge".Under SONOPCO they have been allowed if not encouraged to return to the operating philosophy of.that troubled period.If this is not kept in check and reversed the quality of operations will not improve and there will be a serious event occurr at the plant.
i A Concerned Employee 1
e i
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TO:
Tim Reid Projact Manager ( Acting)
Plant Vogtle Project Directorate II-3 FRCti:
Ken E Brccken, Chief Projects Section 3b Fegion II Questions poced concerning the opening of valves 176 & 177'at Plant Vogtle on October 12 & 13, 1988.
1.
Conceming the definition of mintenance - Is the injection of H202 into the BCS through the RMST while in Mode 5. Loops Not Filled status, concide md to be a mintenance activity?
2.
Does the injection of H202, in the conditions descri, bed in (1) above, initiate or contribute to a crud burst?
3.
Ie a crud burst, as described in (1) above, conside md maintenance?
4.
Technically, can you do maintenance in Mode 57 5.
In general, is it in some cases, permissible to knowingly create a condition contrary to the LCO, as long as you complete the action statement?
6.
While in the ptreess of perfoming " maintenance", in germ'. is it p.twimmibl.u its seen emm+m te.krtaringly cav, ate a condition contrary to the LCO, se long as you complete the action statement?
7.
Specifically, in Octoter 1988, was it pemissible while in a maintenance situation as described in (1) above, to open Valves 176 & 177, as long ac the action statements wem complied with?
8.
Given the same ecoditions as in question 7, would it h3ve teen remiesible if the eituation was not a mintenance situation?
9.
Reference TS 3.0.2 - In general, in order to be in nora-complisace with a Tech Spec, do you have to NCTI' meet both the LCO AND the action statenent?
- 10. Has anyone seen, or approved, the VEGP position statenent concernire
" Voluntary Entry into Action Statements"?
- 11. What is the NBC definition of IttEDIATE, with respect to the condition of Action Statement "C" of TS 3.4.1.4.27 O
d
- 12. Dws the terrn " voluntary entry into an actior, staterent" definitely
. establish that a condition hac been created that requirec the action statenent to be entered?
- 13. Concerning the ** Note at the tottom of TS 3.4.1.4.2 - Ic that statenent absolute, or is it only applicable concerning the operation of th+ FHF eyetem?
These are the questione as I have transcribed them from today's telephone conversation. Hopefully, they will provide you with the information that you need for tomortw's phone call. Leigh will get together with you concernirs the logistice of the phone hook-up. If you have any questione, plesee call her.
Thanks for the ascietance.
~ Ken e - mm
/
i An H ' ' t/ L + m'tuj The esplicent.dtscents the heren dilutten event during refueling aface. by estatstretive prweure. the ACS Ie iselated free unberated water eeurces by l'
In F$AA Amendment 13. the applicant sensitted te teclude in its fechnical Spoofficattens a requirement te lack closed valves 175, 1ecking cleted hay valves.
/
The staff will confirm that this roepstre-j IM.177, and las during refueling.
ment is Iactuded in the Technical Specf fIcatiens.
i I
hoevita of the applicant's analysis show that the tfee available to the aperater to tehe attigettve steps for a heren dilutten event during het staney. hetDu shutdeun, an4 told shutdeun 16 adeguate.
condittens, the operater has et least as minutes (fe11eving salttation of aleres)
During sold ta attigste the beren diletten ovent (less of se.AJoun margin).sh l
l These tiens for operater action are acceptable.
fres J. A. Bailey, GPC. to E. 8. Adonsas, lebt. the
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- 14. 1944, app 11 cent confleued that for modes 3. 4. and 5. tuo safety-grade searce range
% letter dated June The sourcs tacters are ovatleble to actmete an alors on the aala contre) board.
range instreonts constet of tus redundant channels with seperste Class If poner A bistable for each channel provides lahut to the high fins alares et l
sappites.
In additten, the source range instrumentation provides eneleg feput shotdeus.
Upon the generetten of a hip flun at shutdem eless to t'ho proteus cesputer.
The stare message sipal, an alors message will appear en the camputer screen.1s writt plant precedures j
will feestre that the alare setpoint progressed into the camputer be ad>nted to The staff finds this esceptable.
l aceaant for encey in activity during shutdown.
l The aspitsamt ses asked te describe the potential for boren dil etten because ofthe chemical additten portion of the CVC t
l
, the appitcant has stated that either these other f
then the CVC$. Ia they umuld require more then ene failure te bring sources are recluded j
ehest N$ di utfen er they are bounded by the d11stien source assweed to the FSAs. The staff considers this response acceptable.
Inadvertant Leading of a Fuel Asseely to an taproper positten 15.4.7 Strict adotaistrettve contrele in the fers of previously approved established are feitawed durlag fuel needt te present l
precedures and startup testi in an feproper location er a et leaded burnable i
operation with a fuel assattNevertheless, the conseguences of a leading error have been poisen assembly.
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analysed.
- ampartsens of peuer distributlens calculated for the f
The licent resen notas fuel leading Lattern and these calculated for five 1eedtags with ste-The eclecte heed fuel assembifes er buenable pelsen assentites.
teadinge 747--d the spectrue of potenttel Inedrettant fm1 staplace Calculottene factuded ing with incere detectors.
provistens for eenitor l
As part of the required startup testing the in core detectee systes is used to The analysts dettethod h detect efeleeded fuel before operatine et power.shows that all but o l
In the oncepted case, en laterchange of resten 1 and i essash)1es meer i
test.
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x'.:m rrr Amend. 20 12/85 Amend. 38 10/88 1x400116 2 REV.14 voGTLs CHEMICAL AND VOLUME CONTROL ELECTRIC GENERATING PLANT SYSTEM
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.- UNIT 1 AND UNIT 2 la FIGURE 9.3.4-1. (SHEET 4 OF '6)
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Ax40s1232 REV.S
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Qa voottE BORON RECYCLE SYSTEM j
ELECMC GENERAMG PLANT
@gg a power UNIT 1 AND UNIT 2 FIGURE 9. 3.4-2 (SHEET 2 OF 2)
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- h CHARGING PUMP rih n
SECTION HEADER V.J/ '
0606 COMP 0NENf k^
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l0618 COOLING WATER J
gyg t_J LA
{
-eme-FV 0618 e
5=
VENT 600#
T PSy ACCUMULATOR U
SEAL I
8856A Y
8811A]M I'
Qn COOLER RHT e-REFUELING WATER
= LETOOWN HX CONTAINMENTh H7 STORAGE TANK SUMP 8812A 7~
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% Hy 8701A N(
CROSSOVER TO phy
{-- CS AND SI INTERLOCKS HS 8809A mR TRAIN 8
@ [HVP716 FROM TY M HV 8701 SIS,
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SAFETY [NJECTION LE
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OLO LEG LOOP 1 HOT LEG
$[HV 87168 d
P2 CROSSOVER TO COLD LEG RHR TRAIN 8 l
l RHR Cooldown Lo -TP-pe-c-coe
COLD SHUTDOWN - LOOPS NOT FILLED LIMITING CONDITION FOR OPERATION 1
3.4.1.4.2 Two residual heat removal (RHR) trains shall be OPERABLE
- and at
' least one RHR train shall be in operation.** Reactor Makeup Water Storage Tank (RMWST) discharge valves (1208-U4-175, 1208-U4-176#, 1208-04-177# and l
1208-04-183) shall be closed and secured in position.
I APPLICABILITY: MODE 5 with reactor coolant loops not filled.
ACTION:
With less than the above required RHR trains OPERABLE, immediately a.
initiate corrective action to return the required RHR trains to OPERABLE status as soon as possible.
b.
With no RHR train in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required RHR train to operation.
With the Reactor Makeup Water Storage Tank (RMWST) discharge valves (1208-U4-175, 1208-U4-176#, 1208-U4-177#, and 1208-04-183) not closed l
c.
and secured in position, immediately close and secure in position the RMWST discharge valves.
S'URVEILLANCE REOUIREMENTS 4.4.1.4.2.1 At least one RHR train shall be determined to be in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
j 4
4.4.1.4.2.2 Valves 1208-U4-175, 1208-04-176#, 1208-U4-177#, and 1208-U4-183 l
shall be verified closed and secured in position by mechanical stops at least once per 31 days.
- 0ne RHR train may be inoperable for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for surveillance testing provided the other RHR train is OPERABLE and in operation.
1
- The RHR pump may be deenergized for up to I hour provided:
(1) no operations are permitted that would cause dilution of the Reactor Coolant System boron concentration, and (2) core outlet temperature is maintained at least 10*F below saturation temperature.
- RMWST discharge valves 1208-U4-176 and 1208-U4-177 may be open under administrative control provided the Reactor Coolant System is in compliance with the SHUTDOWN MARGIN requirements of Specification 3.1.1.2 and the high flux at shutdown alarm is OPERABLE with a setpoint of 2.30 times background in accordance with Note 9 of Table 4.3-1.
V0GTLE UNITS - 1 & 2 3/4 4-6 Amendment No. 28 (Unit 1)
Amendment No. 9 (Unit 2) a
~
Procedere No.
Vogtle Electric Generating Plant 10019-C i NUCt. EAR OPERAn0NS w%
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(/
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i gg,ff unn COMMON GeorgiaPower 3,,,
t i
n CONTROL OF SAFETY RELATED LOCKED VALVES f
MAhUAL SET N0*
/<-
1.0
_FURPOSE i
This procedure identifies the administrative controls i
for valves which are important for Safety Related i
Systems that shall be locked in a specified position.-
2.0 DEFINITIONS 1
2.1 LOCKED VALVI A valve whose operation is prevented by 'a chain and 1
j1 padlock arrangement or other positive locking device.
i 2.2 KEY CONTROL I
Keys for safety Related Locked Valves should not be
~
issued without approval from the Unit Shift Supervisor
(
(U88).
i 3.0 RE8PON8181LITIES l
The US$ shall maintain administrative' control of the keys used for locking of Safety Related System Valves.
(The Support Shift supervisor normally implements this i
procedure for the U88s.)
4.0 PRECAUTIONS The status of locked valves shall not be changed unless directed by the Uss.
i i
4 l
s VECp 10019-C 5
2 of 3 5.0 INSTRUCTIONS j
(
5.1 BASIC CONTROL OF SAFETY RELATED LOCKED VALVE 8 The initial status of valve positions and lockiNat are 5.1.1 devices is established by system valve lineups j
performed following an outage.
5.1.2 The valves listed in 11867-1/2, " safety Related Locked i
Valve verification Checklist" shall be locked in the specified position with the specified padlocks using i
lengths of chain or other positive locking devices.
5.1.3 Locks should be placed on the remote operator for those manual valves that have remote operators such as reach rods.
l a
/QE 5.1.4/
In the cases where it is not feasible to physically i
j EV
/
lock the apparatus, a Rold Tag may be used, i
I 5.1.5 purposes,thelockonentisunlockedforoperational When a locked e and chain should, if possible be i
i locked to adjacent components so as to preclude loss..
5.1.6 If the locking device cannot be affixed at the component, it should be returned to the Uss for disposition.
I
(
5.1.7 status changes in the positions of locked valves shall be documented by use of 11888-C, " Safety Related Locked 4
Valve Manipulation Log".
4 5.1.8 The position and lock status of each' locked valve will i
be verified quarterly and recorded per 11867-1/2 i
" Safety Related Locked Valve Verification Check 11st".
i 5.1.9 Fadlocks and chains should not normally be removed to i
verify position of locked valves.
If locks must be removed, then re-installation must be independently a
{
verified.
i i
l 3
I j-k 4
4
~
w
.-s
,a n,,.,,
enocaouns No.
mensioN MOR NO.
VEGP 10019-C 5
3 of 3 i
J 5.2 MISPOSITIONED VALVES / INOPERABLE LOCKING DEVICES
(> /
i NOTE i
Valves in position other than 1
the required position and j
documented per 11888-C, " safety i
Related Locked Valve Manipulation i
Log" are not considered mispositioned.
l 5.2.1 Ifanylockedvalveisdiscoveredina!ositionother than the required position or a valve 1 cking device is
~
i found inoperable, the operator shall NOTIFY the U88.
5.2.2 The Uss shall:
l PERFORM en evaluation to determine if the valves a.
current position has resulted in any adverse j
system conditions, b.
PERFORM an evaluation to determine whether i
repositioning the valve to its correct configuration will result in any adverse system conditions,
(
c.
Based on an acceptable evaluation, DIRECT the repositioning and locking of the affected valve or l
if unacceptable, shall IRITIATE placing the component / systems affected in a position where the valve can be restored to its correct configuration, i
d.
ENSURE a Deficiency Card per 00150-C, " Deficiency Control" has been initiated.
3 I
5.0 REFERENCES
[
5.1 PROCEDURES
[
5.1.1 00150-C,
" Deficiency control" 4
5.1.2 00308-C,
" Independent Verification Policy" 5.1.3 00304-C,
" Equipment Clearance And Tagging" i
5.1.4 11888-C,
" Safety Related Locked Valve l
Manipulation Log" 5.1.5 11867-1/2,
" Safety Related Locked Valve l
Verification Checklist"
~
(.
END OF PROCEDURE TEXT
i bi
~
j l
s j
j CONTAINMENT SY$TEMS s
l BASE 5
'3/4.6.1.5 AIR TEMPERATURE
~
[
h limitations on containment average air temperature ensure that the L
overall containment average air temperature does not exceed the initial temperature condition assumed in the safety analysis for a steam line break I
accident. Measurements shall be made at all listed locations, whether by fixed l
or portable instruments, prior to determining the average air temperature.
3/4.6.1.6 CONTAINMENT STRUCTURAL INTEGRITY y
This liettation ensures that the structural integrity of the containment will be maintained comparable to the original design standards for the life of i
i the facility. Structura,1. Integrity is required to ensure that the containment
~ will withstand the maximps pressure of 41.9 psig in the event of a steam line break accident. The mespurement of containment tendon lift-off force, the ten-l
. ' r, sile tests of the tendon strands for Unit 1, the visual examination of tendons, anchorages and exposed interior and exterior surfaces of the containment and i
the Type A leakage test 'for both units are sufficient to demonstrate this capa-bility.
(The tendon strand samples will also be subjected to stress cycling i
tests and to accelerated corrosion tests to simulate the tendon's operating conditions and environment.) Lift-off testing on Unit 2 will be accompanied by detensioning of one tendon on Unit 1.
This tendon will alternate between the hoop and inverted -U tendons. With regard'!to 0-cracking, the acceptance criteria.for the visual jnspection of the containment concrete is that the l
area comprising 0-crackt.1g should not exceed 25 sq. ft.
abnormal containment deg,renced by Action state &ent 3.6.1.6.b do not defi N conditions Pete radation.
L
' abnormal degradation and their existence requires an appropriate en
> evaluation and a Special Report in accordance with Specification 6.gineering l
4 s.2.
l h required Specia'l Re ment abnormalities shall:inc$ orts from aht$on 'of the tendon condition, thengi ude a descr i
condition of the concrete (especially at tendon anchorages), the inspection procedures, the tolerances on cracking, the results of the engineering evalua-2-
i tion, and the corrective actions taken, or proposed.
l 3/4.6.1.7 CONTAINMENT VENTILATION SYSTEM M es.or c,~r. wng(
l N 24-inch contain.nent purge supply and exhaust isolation valves are
. required to be sealed closed during plant operations since these valves' have not' been demonstrated capable of closing during a 1,0CA or steam line break accident.
Maintaining these valves sealed closed durInc plant operation ensures that exces-l sive quantities of radioactive materials will det be released via the Containment P
system. To provide assurance that these' containment valves cannot be inad-Sealed close)ves are sealed closed in accordance wit the va
~
var tly opened i
d isolation valves are isolation valves under admini-plan 6.2.4.
strative control to assbre that they cannot beitnadvertently opened. Admini-strative control includes mechanical' devices tc seal er lock the valve claned.
i.
l ',
I i
i REACTOR COOLANT SYSTEM COLD SHUTDOWN - LOOPS NOT FILLED i
LIMITING CONDITION FOR OPERATION 4
\\
j 3.4.1.4.2 Two residual heat removal (RHR) trains shall be OPERA 8LE* and at least one RHR train shall be in operation.** Reactor Makeup Water Storage Tank (RWST) discharge valves (1208-U4-175,1208-04-176,1208-U4-177 and 1208-04-183) shall be closed and secured in position.
APPLICA81LITY: MODE 5 with reactor coolant loops not filled.
j ACTION:
i a.
With less than the above required RHR trains OPERA 8LE, immediately initiate corrective action to return the required RHR trains to OPERABLE status as soon as possible.
I b.
With no RHR train in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant Systes and immediately initiate corrective action to return the required RHR l
train to operation.
i.
c.
With the Reactor Makeup Water Storage Tank (RWST) discharge valves l
(1208-U4-175,1208-U4-176,1208-U4-177,and1208-04-183) not closed i
and secured in position, immediately close and secure in position the RMWST discharge valves.
i l
SURVEILLANCE REQUIREMENTS
]
l 4.4.1.4.2.1 At least one RHR train shall be determined to be in operation and l
circulating reactor coolant at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
4.4.1.4.2.2 Valves 1208-U4-175,1208-U4-176,1208-U4-177 and 1208-U4-183 shall be verifled closed and secured in position by mechanical stops at least once per 31 days.
i
- 0ne RHR train say be inoperable for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for surveillance testing provided the other RHR train is 0PERA8LE and in operation.
l
- The RHR pump may be deenergized for up to I hour provided: (1) no operations are pemitted that would cause dilution of the Reactor Coolant System boron i
concentration, and (2) core outlet temperature is maintained at least 10*F below saturation temperature.
/;
t'
,n NUCl. EAR REGULATORY COMMisslON g
CEo10Nll 4
3
,/
101 MARIETTA STREET, N.W.
ATI.ANTA,GEORQl A 30323 l
\\,,,,,,/
ne n m i
Docket Hos. 50-424, 50-425 j
License Nos. NPF-68, NPF-81 EA 89-227 4,
j Georgia Power Company ATTN: Mr. W. G. Hairston, III Senior Vice President -
Nuclear Operations Post Office Box 1295 Bimingham, Alabama 35201 i
e i
Gentlemen:
SUBJECT:
NOTICE OF VIOLATION AND PROPOSED IMPOSITION OF CIVIL PENALTY - $7,500
~
This refers to the Nuclear Regulatory Comission (NRC) security inspection conducted by Messrs. D. Thompson and A. Tillman on October 23 - 27,1989, at the Yogtle Nuclear Power Plant which identified repeat violations in the areas of Safeguards Information document control and protected area perimeter assessment. The report documenting this inspection was sent to you on
}
November 22, 1989. Subsequently, an Enforcement Conference was held with you and members of your staff on December 11, 1989, at which time the violations, their causes and repetitiveness, and your corrective actions were discussed.
A letter sumarizing this conference was sent to you on January 2,1990.
l Violation A described in the enclosed Notice of Violation and Proposed Imposition of Civil Penalty (Notice) involving the failure to properly protect l
and account for documents containing Safeguards Information is of particular l
l concern to the NRC because of the number of times this type of violation has 4
occurred. This violation is similar to four previous violations identified during four security inspections conducted since February 1988. Although we recognize that each of these violations was identified by you, it is evident i
i from the recurring nature of the violations that you have not adequately addressed the root causes of the failure and that your past corrective actions i
have been ineffective. Specifically, the lack of personnel sensitivity to doctanent control procedures has directly contributed to the repetitiveness of 4
i this problem. Prior to this new violation being identified, these problems had been discussed at a management meeting between your organization and the NRC in August 1989, and again in correspondence in September 1989 which j
highlighted additional examples of previous violations in this area.
To emphasize the need for effective corrective action in order to avoid repetitive violations, I have been authorized, after consultation with the Director, Office of Enforcement,.and the Deputy Executive Director for Nuclear Materials Safety, Safeguards, and Operations Support, to issue the enclosed
==mmmmm-.
g ACC9!50W 399-
2 FE8 021990 i,
Georgio Power Company 1
i i
Notice of Violation and Proposed Imposition of Civil Penalty in the amount of In accordance with
$7.500 for Violation A described in the enclosed Notice.
the " General Statement of Policy and Procedure for NRC Enforcement Actions,"
10 CFR Part 2, (1989) (Enforcement Policy), Supplement III, Violation A described in the enclosed Notice has been categorized at Severity Level IV.
j The base value of a civil penalty for a Severity Level IV violation is The escalation and mitigation factors in the Enforcement Policy were
$15,000.
considered and the penalty was mitigated by 50 percent due to your identifica-tion of the violation. The other escalation and mitigation factors were considered and no further adjustment is appropriate.
l The NRC very rarely issues civil penalties for Severity Level IV violations.
However, the Enforcement Policy provides the discretion to do so when the 4
facts of the case suggest that the licensee has not taken adequate corrective In this case, action for past similar violations to prevent their recurrence.
it is evident from the recurring nature of the violations that you have not adequately addressed the root causes of the failure to take appropriate Furthermore, if steps measures and, therefore, a civil penalty is appropriate.
are not taken promptly to remedy this problem, future violations of this type may result in more stringent enforcement actions.
Violation B involving the failure to provide adequate assessment capability for the Unit 1 protected area perimeter is also of concern to us because of This violation is similar to a previous violation i
its repetitive nature.
identified during a security inspection and is documented in NRC Inspection l
Report Nos. 50-424/88-29 and 50-425/88-39. The root cause of these incidents was determined to be a combination of equipment limitations and certain i
This violation was also discussed at the December 11 barrier interfaces.
1989, enforcement conference at which time you discussed the planned corrective actions, the specific details of which are considered security
(
l infonnation and are exempt from public disclosure under 10 CFR 73.21.
However, it should be recognized that previous ineffective corrective actions in this area directly contributed to its recurrence. Consequently, NRC l
strongly encourages the testing of all related equipment interfaces as well as evaluation of procedural testing methods and consideration of increased l
testing frequencies to ensure "above marginal" operability of the assessment This violation has been categorized a Severity capability at all times.
l Level IV violation and is applicable to Unit I only.
You are required to respond to this letter and should follow the instructions In your specified in the enclosed Notice when preparing your response.
response, you should document the specific actions taken and any additional l
After reviewing your response to this actions you plan to prevent recurrence.
i Notice, including your proposed corrective actions and the results of future inspections, the NRC will determine whether further NRC enforcement action is necessary to ensure compliance with NRC regulatory requirements.
l i
NE 3
Georgia Power Company l
In accordance with 10 CFR 2.790(d) and 10 CFR 73.21, safeguards activities and Therefore, the enclosure security measures are exempt from public disclosure.
to this letter will not be placed in the NRC Public Document Room.
The responses directed by this letter and enclosed Notice are not subject to 1
the clearance procedures of the Office of Management and Budget as required by i
the Paperwork Reduction Act of 1980, Public Law No.96-511.
i Should you have any questions concerning this letter, please contact us.
Sincerely, ut+L[
01 Stewart D. Ebneter Regional Administrator i
Enclosure:
Notice of Violation and Proposed Imposition of Civil Penalty (SafeguardsInformation) f f
cc w/ enc 1:
R. P. Mcdonald Executive Vice President-Nuclear Operations j
Georgia Power Company i
P. O. Box 1295 Birmingham, AL 35201 i
i C. K. McCoy Vice President-Nuclear Georgia Power Company P. O. Box 1295 j
Birmingham, AL 35201 G. Bockhold, Jr.
General Manager, Huclear Operations Georgia Power Company P. O. 1600 Waynesboro, GA 30830 J. A. Bailey Manager-Licensing J
Georgia Power Company Bi in am 35201
===mmamer
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UNITE 3 57ATea
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NUCLEAR RE2ULATGRY COMMISSION wAmamotow.e.c.ssues l
i February 20, 1990
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j Dockets Nos. 50-424 and 50-425 i
Mr. W. G. Mairston, III Senior Vice Presfdont -
Nuclear Operations Georgia Power Company P.O. Box 1295 Birmingham, Alabase 35201
Dear Mr. Nairston:
)
SUBJECT:
ISSUANCE OF AMENDMENT N0.28 TO FACILITY OPERATING LICENSE NPF-68 AND AMENDMENT N0. 9 TO FACILITY OPERATING LICENSE NPF V0GTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2 (TACs 75320/75321)
The Nuclear Regulatory Cosmission has issued the enclosed Amendment No. 28 to 1
i Facility Operating License No. NPF-68 and Amendment No. 9 to Facility Operating License NPF-81 for the Vogtle Electric Generating Plant, Units I and 2.
These amendments consist of changes to the Technical Specifications (TSs) in response to your application dated Noyes6er 21,1989.
- The amendments enable. non-borated chemical additions to be made to the Reactor -
i Coolant System-(RCS) under achinistrative control during Moda Sb (cold shutdown, l
loops not filled) and Mode 6 (refueling) using a flow path via the Reactor
- Nkaup Wster Storage Tank (RMkST).
t A cupy of the related Safety Evaluation is also en' closed. Notice of issuance of
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the avendments will be included in the Commission's biweekly Federal Register notice.
Sincerely,
[
l r
Timotly A. Reed, Project Manager Project Directorate 11-3 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Enclosures:
- 1. Amendment No. 28 to MPF-68
- 2. Amendment No. 9 to NPF-81 i
- 3. Safety Evaluation cc w/ enclosures:
s See next page 7
/
3
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Mr. W. G. Hairston, III Secrgia Power Company Vogtle Electric Generating Plant a
cc:
Mr. J. A. Bailey Resident Inspector i
Manager - Licensing Nuclear Regulatory Commission Georgia Power Company P.O. Box 572 F.O. Box 1295 Waynesboro, Georgia 30830 l
Birmingham, Alabama 35201 James E. Joiner, Esq.
Bruce W. Churchill, Esq.-
Troutsen, Sanders, Lockerman, Shaw, Pittman, Potts and Trowbridge
& Ashmore 2300 N Street, N.W.
1400 Candler Building Washington, D.C. 20037 127 Peachtree Street, N.E.
i
)
Atlanta, Georgia 30303 Mr. G. Bockhold, Jr.
General Manager, Vogtle Electric Mr. R. P. Mcdonald Generating Plant Executive Vice President -
P.O. Box 1600 Nuclear Operations Waynesboro, Georgia 30830 Georgia Power Company P.O. Box 1295 Regional Administrator, Region II Birmingham, Alabama 35201 U.S. Nuclear Regulatory Commission 101 Marietta Street, N.W., Suite 2900
'Mr. J. Leonard Ledbetter, Director Atlanta, Georgia 30323 Environmental Protection Division Department of Natural Resources Office of the County Cossiissioner 205 Butler Street, S.E., Suite 1252 1
i Burke County Commission Atlanta, Georgia 30334 i
Waynesboro, Georgia 30830 Office of Planning and Budget Attorney General Room 615B Law Department i
270 Washington Street, S.W.
132 Judicial Building l
Atlanta, Georgia 30334 Atlanta, Georgia 30334 Mr. C. K. McCoy Mr. Alan R. Herdt, Chief Vice Presient - Nuclear, Vogtle Project Project Branch #3 i
Georgia Power Company U.S. Nuclear Regulatory Commission P.O. Box 1295 101 Marietta Street, NW, Suite 2900 Birmingham, Alabama 35201 Atlanta, Georgia 30323 l
4 l
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UNITED STATES NUCLEAR REQULATORY COMMISSION a
lI WA8848880 TON,0. C.20505 T4 e....
1 GEORGIA POWER COMPANY
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OGLETHORPE POWER CORPORATION i
i MUNICIPAL ELECTRIC AUTHORITY OF GEORGIA CITY OF DALTON GEORGIA
}
V0GTLE ELECTRIC GENERATING PLANT. UNIT 1 j
AMENDMENT TO FACILITY OPERATING LICENSE l
Amendment No. 28 License No. NPF-68 i_
l 1.
The Nuclear Regulatory Cosmiission (the Commission) has found that:
The app (lication for amendment to the Vogtle Electric Generating Plan 1
A.
Unit 1 the facility), Facility Operating License No. NPF-68 filed by the Georgia Power Company, acting for itself Oglethorpe Power Corpo-ration, Municipal Electric Authority of Georgia, and City of Dalton, Georgia (thelicensees),datedNovember 21, 1989, complies with the standards and req)uirements of the Atomic Energy Act of 1954, as amended (the Act, and the Commission's rules and regulations set l
forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the l
provisions of the Act, and the rules and regulations of the Comission; l.
C.
Thereisreasonableassurance(1)thattheactivitiesauthorizedby l
this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations set forth in 10 CFR Chapter I;
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I D.
The issuance of this license amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comunission's regulations and all applicable requirements have been satisfied.
n t
L l
- i 2.
Accordingly, the license is hereby amended by page changes to the Technical Specifications as indicated in the attachment to this license amendment i
i and paragraph 2.C.(2) of Facility Operating License No. NPF-68 is hereby amended to read as follows:
Technical Specifications and Environmental Protection Plan The Technical Specifications cor,tained in Appendix A, as revised i
through Amendment No. 28, and the Environmental Protection Plan j
contained in Appendix 8, both of which are attached hereto, are hereby incorporated into this license. GPC shall operate the facility in accordance with the Tect.nical Specifications and the Environmental Protection Plan.
3.
This license amendment is effective as of its date of issuance and shall be implemented within 30 days of issuance.
l FOR THE NUCLEAR REGULATORY COMMISSION i
6W David B. Matthews, Director Project Directorate II-3 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Attachment:
Technical Specification Changes Date of Issuance: February 20, 1990 i
i i
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t..;..vaos m as NUCLEAR REGULATORY COMMISSION g
wasasworow.o. c. noses l'
4 f
GEORGIA POWER COMPANY f
OGLETHORPE POWER CORPORATION l
MUNICIPAL ELECTRIC AUTHORITY OF GEORGIA j
CITY OF DALTON. GEORGIA V0GTLE ELECTRIC GENERATING PLANT. UNIT 2 l
AMENDMENT TO FACILITY OPERATING LICENSE t
i Amendment No. 9 i
License No. NPF-81
}
L1.
The Nuclear Regulatory Commission (the Commission) has found that:
l A.
The app (lication for amendment to the V'ogtle Electric Generating Plant, Unit 2 the facility), Facility Operating License No. NPF-81 filed by l
the Georgia Power Company, acting for itself, Oglethorpe Power Corpo-ration, Municipal Electric Authority of Georgia, and City of Dalton, Georgia (thelicensees),datedNovember 21, 1989, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the l
provisions of the Act, and the rules and regulations of the j
Commission; C.
Thereisreasonableassurance(1)thattheactivitiesauthorizedby i
this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted j
in compliance with the Comnission's regulations set forth in 10 CFR Chapter I;
- D.
The issuance of this license amendment will not be inimical to the consen defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
l
I 2.
Accordingly, the license is hereby asunded by page changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 2.C.(2) of Facility Operating License No. NPF-81 is hereby i
amended to read as follows:
i Technical Specifications and Environmental Protection Plan J
The Technical Specifications contained in Appendix A, as revised through Amendment No. 9 and the Environmental Protection Plan contained in Appendix 8, both of which are attached hereta, are hereby incorporated into this license. GPC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
This license amendmeat is effective as of its date of issuance and shall j
be implemented within 30 days of issuance.
FOR THE NUCLEAR REGULATORY Comt!SSION d
4 David B. Matthews, Director Project Directorate II-3 Division of Reactor Projects - I/II t
Office of Nuclear Reactor Regulation
Attachment:
l Technical Specification Changes I
Date of Issuance: February 20, 1990 i
i j
4 4
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ATTACHMMT 10 LICENSE AMENDNENT NO. 28 FACILITY OPERATING LICENSE NO. NPF-68 AND LICENSE AMENDNENT NO. 9 I
FACILITY OPERATING LICENSE NO. NPF-81 u
DOCKETS NOS. 50-424 AE 50-425 4
l Replace the following pages of the Appendix 'A' Technical Specifications with l
the enclosed pages. The revised pages are identified by Amendment num6er and i
contain vertical lines indicating the areas of change. The corresponding overleaf pages are also provided to maintain document coupleteness.
5 J
1 i
Amended Page Overleaf Page
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3/4 4-6 3/44-5
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3/4 9-1 B3/4 4-1 83/4 4-2 i
B3/4 9-1 33/4 9-2 l
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6 9
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i E
i REACTOR C00LANT SYSTEM COLD SHUTD0WN - LOOPS FILLED LIMITING CONDITION FOR OPERATION j
i 3.4.1.4.1 At least one residual heat removal (RHR) train shall be OPERABLE and j
in operation *, and either:
J l
i a.
One additional RHR train shall be OPERABLE **, or l
b.
The secondary side water level of at least two steam generators shall be greater than 17% of wide range (LI-0501, LI-0502, LI-0503, i
LI-0504).
1 l
APPLICABILITY: MODE 5 with reactor coolant loops filled ***.
i, ACTION:
With one of the RHR trains inoperable or wi e less than the required i
s.
i steam generator water level..immediately initiate corrective action to return the inoperable RHR train to OPERABLE status or restore the l
required steam generator water level as soon as possible.
b.
With no RHR train in operation, suspend all operations involving a reduction in baron concentration of the Reactor Coolant System and i
immediately initiate corrective action to return the required RHR train to operation.
I j
SURVEILLANCE REQUIREMENTS 4.4.1.4.1.1 The secondary side water level of at least two steam generators when required shall be determined to be wi',idn limits at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
4.4.1.4.1.2 At least one RHR train shall be determined to be in operation and l
circulating reactor coolant at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
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i
- The RHR pump may be deenergized for up to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> provided:
(1) no operations are permitted that would cause dilution of the Reactor Coolant System boron concentration, and (2) core outlet temperature is maintained at least 10*F j
below saturation temperature.
- 0ne RHR train may be inoperable for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for surveillance testing provided the other RHR train is OPERABLE and in operation.
l
- A reactor coolant pump shall not be started unless the secondary water temperature of each steam generator is less than 50*F above each of the i
Reactor Coolant System cold leg temperatures.
V0GTLE UNITS - 1 & 2 3/4 4-5
1 COLD SHUTDOWN - LOOPS NOT FILLED LIMITING CONDITION FOR OPERATION i
j 3.4.1.4.2 Two residual heat removal (RHR) trains shall be OPERABLE
- and at j
least one RHR train shall be in operation.** Reactor Makeup Water Storage Tank (RMWST) discharge valves (1208-U4-175, 1208-U4-176#, 1208-U4-177# and j
i 1208-04-183) shall be closed and secured in position.
1 i
APPLICA8ILITY: MODE 5 with reactor coolant loops not filled.
ACTION:
1 a.
With less than the above required RNR trains OPERABLE, immediately i
initiate corrective action to return the required RHR trains to OPERABLE status as soon as possible.
4; l
t;.
With no RHR train in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required RHR i
train to operation.
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j c.
With the Reactor Makeup Water Storage Tank (RmfST) discharge valves (1208-U4-175, 1208-U4-1768, 1208-U4-177#, and 1208-U4-183) not closed l
l and secured in position, immediately close and secure in nosition the l
RMWST discharge valves.
SURVEILLANCE REQtlIREMENTS 4.4.1.4.2.1 At least one RHR train shall be determined to be in operation and l
circulating reactor coolant at least once.per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
4.4.1.4.2.2 Valves 1208-U4-175, 1208-U4-176#, 1208-U4-177f, and 1208-04-183 l
shall be verifi'ed closed and secured in position by mechanical stops at least once per 31 days.
i l
l
- 0ne RHR train may be inoperable for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for surveillance testing l
provided the other RHR train is OPERABLE and in operation.
- The RHR pump may be deenergized for up to I hour provided:
(1) no operations are permitted that would cause dilution of the Reactor Coolant System boron i
concentration, and (2) core outlet tagerature is maintained at leert 10*F below saturation temperature.
- RMWST discharge valves 1208-U4-176 and 1208-U4-177 may be open under administrative control provided the Reactor Coolant System is in compliance with the SHUTDOWN MARGIN requirements of Specification 3.1.1.2 and the high
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flux at shutdown alarm is OPERABLE with a setpoint of 2.30 times background in accordance with Note 9 of Table 4.3-1.
f 3/4 4-6 Amendment No. 28 (Unit 1) i V0GTLE UNITS - 1 & 2 Amendment No. 9 (Unit 2) t
,._.,..m.,
3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION LIMITING CONDITION FOR OPERATION 3.9.1 The boron concentration of all filled portions of the Reactor Coolant System and the refueling canal shall be maintained uniform and sufficient to ensure that the more restrictive of the following reactivity conditions are met:
A K,ff of 0.95 or less, or a.
b.
A boren concentration of greater than or equal to 2000 ppe.
Additionally, valves 1208-U4-175, 1208-U4-177f, 1208-U4-183, and 1208-04-176f l
shall be closed and secured in position.
APPLICABILITY: MODE 6.
ACTION:
a.
With the requirements of a. and c. above not satisfied, immediately suspend all operations involving CORE ALTERATIONS or positive reactivity changes and initiate and continue boration at greater than or equal to 30 gpm of.a solution containing greater than or equal to 7000 ppe baron or its equivalent until K,ff is reduced to less than or equal to 0.95 or the boron concentration is restored to greater than or equal to 2000 ppa, whichever is the more restrictive.
b.
With valves 1208-04-275, 1208-U4-177f, 1208-U4-183, and 1208-U4-176f l
not closed and secured in position, immediately close and secure in position.
l l
SURVEILLANCE REQUIREMENTS 4.9.1.1 The boron concentration of the Reactor Coolant System and the refueling canal shall be detemined by chemical analysis at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
4.9.1.2 Valves 1208-U4-175, 1208-U4-177f, 1208-04-183, and 3208-U4-176f shall l
be verified' closed and secured in position by mechanical stops at least once per 31 days.
- RMWST discharge valves 1208-04-176 and 1208-U4-177 may be open under administrative control provided the Reactor Coolant System is in compliance with the requirements of Specification 3.9.1 and the high flux at shutdown alarm is OPERABLE with a setpoint of 2.30 times background.
For the purpose of this Specification, the high flux at shutdown alarm will be demonstrated OPERABLE pursuant to Specification 4.9.2.
V0GTLE UNITS - 1 & 2 3/4 9-1 Amendment No.28 (Unit 1)
Amendment No.9 (Unit 2)
3/4.4 REACTOR C0OLANT SYSTEM BASES 3/4.4.1 REACTOR COOLANT LOOPS AND C0OLANT CIRCULATION Tne plant is designed to operate with all reactor coolant loops in operation and maintain DN8R above 1.30 during all nomal operations and antici-pated transients.
In MODES 1 and 2 with one reactor coolant loop not in operation this specification requires that the plant be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
In MODE 3, two reactor coolant loops provide sufficient heat removal capability for removing core decay heat even in the event of a bank withdrawal accident; however, a single reactor coolant loop provides sufficient heat removal capacity if a bank withdrawal accident can be prevented, i.e., by opening the Reactor Trip System breakers.
l In MODE 4, and in MODE 5 with reactor coolcnt loops filled, a single reactor coolant loop or RHR train provides sufficient heat removal capability i
for removing decay heat; but single failure consioarations require that at i
{
least two trains / loops (either RHR or RCS) be OPERABLE.
1 i
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In MODE 5 with reactor coolant loops not filled, a single RHR train i
provides sufficient heat removal capability for removing decay heat; but single i
failure considerations, and the unavailability of the steam generators as a j
heat removing component, require that at least two RHR trains be OPERABLE. The i
locking closed of the required. valves, except valves 1208-U4-176 and 1208-U4-177 for short periods of time to maintain chemistry control, in Mode 5 i
(with the loops not filled) precludes the possibility of uncontrolled boron dilution of the filled portion of the Reactor Coolant System. These actions 8
prevent flow to the RCS of unborated water in excess of that analyzed. These l
limitations are consistent with the initial conditions assumed for the boron dilution accident in the safety analysis.
The operation of one reactor coolant pump (RCP) or one RHR pump provides adequate flow to ensure mixing, prevent stratification and produce gradual reactivity changes during boron concentration reductions in the Reactor Coolant System.
The reactivity change rate associated with boron reduction will, l
therefore, be within the capability of operator recognition and control.
The restrictions on starting an RCP with one or more RCS cold legs less than or equal to 350*F are provided to prevent RCS pressure transients, caused by energy additions from the Secondary Coolant System, which could exceed the li. nits of Appendix G to 10 CFR Part 50.
The RCS will be protected against overpressure transients and will not exceed the limits of Appendix G by restricting ttarting of the RCPs to when the secondary water temperature of
}
each stew generator % less than 50*F above each of the RCS cold leg i
temperatures.
i l
V0GTLE UNITS 1 & 2 B 3/4 4-1 Amendment No. 28 (Unit 1)
Amendment No.g (Unit 2)
BASES j
3/4.4.2 SAFETY VALVES j
The pressurizer Code safety valves operate to prevent the RCS from being i
pressurized above its Safety Limit of 2735 psig. Each safety valve is designed i
to relieve 420,000 lbs per hour of saturated steam at the valve Setpoint..The l
relief capacity of a single safety valve is adequate to relieve any overpressure i
condition which could occur during shutdown.
In the event that no safety valves are OPERABLE, an operating RHR train, connected to the RCS, provides overpressure j
relief capability and wil' prevent RCS overpressurization.
In addition, the Cold Overpressure Protection System provides a diverse means of protection j
against RCS overpressurization at low temperatures.
l During operation, all pressurizer Code safety valves must be OPERABLE to prevent the RCS from being pressurized above its Safety Limit of 2735 psig.
The combined relief capacity of all of these valves is greater than the maximum surge rate resulting from a complete loss-of-load assuming no Eeactor trip until the first Reactor Trip System Trip Setpoint is reached (i.e., no credit is taken for a direct Reactor trip on the loss-of-load) and also assuming i
1 l
no operation of the power-operated relief valves or steam dump valves.
J During shutdown conditions in Mode 5 only one pressurizer code safety is j
required for overpressure protection.
Ir. lieu of an actual operable code safety i
. valve an unisolated and unsealed vent pathway (i.e. a direct unimpaired opening) of equivalent size can be taken as synonymous with an OPERABLE code safety.
l Demonstration of the safety valves' lift settings will occur only during shutdowr. and will be performed in accordance with the provisions of Section XI t
of the ASME Boiler and Pressure Code.
I J
3/4.4.3 PRESSURIZER i
The 12-hour periodic surveillance is sufficient to ensure that the param-eter is restored to within its limit following expected transient operation.
The maximum water volume ensures that a steam bubble is formed and thus the RCS is not a hydraulically solid system.
The requirement that a minimum number of i
pressurizer heaters be OPERABLE enhances the capability of the plant to control j
Reactor Coolant Systes pressure and establish natural circulation.
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V0GTLE UNITS - 1 & 2 8 3/4 4-2
l l
3/4.9 REFUELING OPERATIONS BASES i
3/4.9.1 BORON CONCENTRATION i
The limitations on reactivity conditions during REFUELING ensure that:
l (1) the reactor will remain subcritical during CORE ALTERATIONS, and (2) a uniform boron concentration is maintained for reactivity control in the water volume having direct access to the reactor vessel. The locking closed of the' l
required valves, except valves 1208-U4-176 and 1200-04-177 for short periods of time to maintain chemistry control, during refueling operations precludes j
the possibility of uncontrolled boron dilution of the filled portions of the Reactor Coolant System. These actions prevent flow to the RCS of unborated water in excess of that analyzed. These limitations are consistent with the j
initial conditions assumed for the Boron Dilution Accident in the safety analysis. The Boron concentration value of 2000 ppe or greater ensures a K,ff of 0.95 or less and includes a conservative allowance for calculational l
uncertainties of 100 ppe of boron.
i 3/4.9.2 INSTRUMENTATION I
l The OPERABILITY of the Source Range Neutron Flux Monitors ensures that redundant monitoring capability is available to detect changes in the reactivity condition of the core.
3/4.9.3 DECAY TINE l
The minimum requirement for reactor suberiticality prior to movement of j
irradiated fuel assemblies in the reactor vessel ensures that sufficient time has elapsed to allow the radioactive decay of the short-lived fission products. This decay time is consistent with the assumptions used in the safety analyses.
l 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS i
l The requirements on containment building penetration closure and 0PERABILITY l
ensure that a release of radioactive material within containment will be restricted from leakage to the environment. The OPERABILITY and closure I
restrictions are sufficient to remrict radioactive material release from a fuel element rupture based upon ^.he lack of containment pressurization f
potential while in the REFUELING MDDE.
3/4.9.5 ComuNICATIONS The requirement for communications capability ensures that refueling station personnel can be promptly informed of significant changes in the facility status or core reactivity conditions during CORE ALTERATIONS.
j i
J V0GTLE UNITS - 1 & 2 8 3/4 9-1 Amendment No. 28 (Unit 1)
Amendment No. 9 (Unit 2)
J l
)
RERIELING OPERATION,5 j'
BASES l
3/4.9.6 REFUELING MACHINE j
The OPERABILITY requirements of the refueling machine and auxiliary hoist ensure that:
1 (1) The refueling machine will be used for the movement of fuel assemblies j
and/or rod control cluster assemblies (RCCA) or thimble plug assemblies, and the auxiliary hoist will be used for the movement of control rod drive shafts,
)
(2) the refueling machine will have sufficient load capacity to lift a j
fuel assembly and/or a rod control cluster assembly or thimble plug assembly, f
and the auxiliary hoist will have sufficient load capacity to lift a control j
rod drive shaft and attached RCCA, and I
(3) the core internals and reactor vessel are protected from excessive lifting force in the event they are inadvertently engaged during lifting j
operations.
t
. 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE AREAS l
l The restriction on movement of loads in excess of the nominal weight of a l
fuel and control rod assembly and associated handling tool over other fuel assemblies in the storage pool ensures that in the event this load is dropped:
- (1) the activity release wil1~be limited to that contained in a single fuel
(
assembly, and (2) any possible distortion of fuel in the storage racks will not i
result in a critical array. This assumption is consistent with the activity l
release assumed in the safety analyses.
3/4.9.8 RESIDUAL HEAT REMDVAL AND COOLANT CIRCULATION i
j The requirement that at least one residual heat removal (RHR) train be in operation ensures that:
(1) sufficient cooling capacity is available to remove I
decay heat and maintain the water in the reactor vessel below 140*F as required i
during the REFUELING MDDE, and (2) sufficient coolant circulation is maintained l
through the core to minimize the effect of a boron dilution incident and prevent l
boron stratification.
The requirement to have two RHR trains OPERABLE when there is less than i
23 feet of water above the reactor vessel flange ensures that a single failure of the operating RHR train will not result in a complete loss of residual heat removal capability. With the reactor vessel head removed and at least 23 feet i
of water above the reactor pressure vessel flange, a large heat sink is avail-i able for core cooling. Thus, in the event of a failure of the operating RHR train, adequate time is provided to initiate emergency procedures to cool l
the core.
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V0GTLE UNITS - 1 & 2 8 3/4 9-2 i
=.
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'S UNITED sTATas I' i[
NUCLEAR REGULATORY COMMISSION a.!
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3 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION I
RELATED TO AMENDMENT N0. 28 TO FACILITY OPERATING LICENSE NPF-6B AND AMENDMENT NO. 9 TO FACILITY OPERATING LICENSE NPF-81 GEORGIA POWER COMPANY. ET AL.
i i
l DOCKETS N0S. 50-424 AND 50-425 V0GTLE ELECTRIC GENERATING PLANT. UNITS 1 AND 2 I
1.0 INTRODUCTION
By letter dated Novesbar 21, 1989, Georgia Power Company (the licensee) l requested changes to Technical Specifications (TSs) 3/4.4.1.4.2 and 3/4.9.1 for the Vogtla Electric Generating Plant Units 1 and 2.
These changes enable non-borated chemical additions to be made to the Anactor Coolant System (RCS) during Mode 5b (cold shutdown, loops not filled) and Mode 6 (refueling) using a flow path via the Reactor Makeup Water Storage Tank (RMfST). Use of this flow path requires that valves 1208-U4-176 and 1208-U4-177 be opensd periodically under seninistrative control.
The existing TSs require that i
these valves be closed and secured.
b 2.0 EVALUATION Of the accidents and transients addressed in the Vogtle Final Safety Analysis Report (FSAR), the boren dilution event is the only transient that could be l
affeeted by the proposed TS rev1sions..The prolonged and unsenitored addition l
of an unborated chemical solution into the RCS for purpose of contmiling RCS chemistry could lead to a complete loss of shutdown surgin.
l FSAR Sec', ion 15.4.6 presents boron dilution analyses for Modes 3, 4, and Sa j
(loops filled) in accordancer with Standard Review Plan (SRP) Section 15.4.6.
l The analyses verify that adequate operator tium (at least 15 minutas) is available to terminate the dilution flow between the time a "high flux at j
shutdown" alarm is received and when criticality occurs. However, boron 4
dilution analyses for Modes 5b and 6 do not exist because TS 3/4.4.1.4.2 and j
3/4.9.1 assure that possible dilution flow paths are isolated by closing and securing the appropriate valves, thereby administrative 1y precluding a boven i
dilution event.
To permit ch.sical additions to be ande to the RCS during Modes 5b and 6 using a flow path via the RMWST through the chemical mixing tank, valves 1208-U4-176 and 1208-U4-117 sust be opened.
In this regard, the licensee has proposed I
revisions tc, the above referenced TSs and has performed boron dilution analyses for these modes and this particular dilution path in accordance with SRP i
Section 15.4.6.
The SRP acceptance criteria for Modes 5b and 6 are minimum i
. operator action times of 15 minutes and 30 minutes, respectively.
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The licensee's analyses to determine minimum operator action times male use of conservative assumptions regarding boron dilution rate and active reactor coolant volume, as suggested in the SRP. A dilution flow rate of 3.5 gps, representing the maximum rate possible via the proposed flow path under arty operating condition, has beer, assumed. Additionally, the minimum cold drained reactor vessel volume has been utilized in the analyses, and the active RCS volume further minimized by assuming only one residual heat removal train in operation, considering miniflow and typass lines to be emptyflux at shutdown"and neglecting reactor coolant loop volumes. Also, the source range "high alarm is assused to be operable with a setpoint of 2.3 times background, as required by TS Table 4.3-1 Note g.
Shutdown margin rua uirements, as specified ty TS 3.1.1.2 for Mode 5 and TS 3.g.1 for Mode 6, are a'so unchanged. The results of the licensee's analyses indicate that the minimum acceptable operator action times of 15 minutes for Mode 5b and 30 minutes for Mode 6, as specified in the SRP, are met.
We have reviewed the licensee's analyses as provided in the Noves6er 21,1989, submittal and find that conservative assumptions have been used, the SRP acceptance criteria have been set or exceeded, and that the pro>osed TS changes will not have arqy adverse affect on safety. Ary other ;>cron dilution paths will continue to be precluded by the TSs.
On the basis of the above evaluation, the IEC staff concludes that the proposed T5s changes are acceptable.
3.0 EfNIRONMEhTAL CONSIDERATION The amenosents involve changes in requirements with respect to the installation or use of facility components located within the restricted area as defined in 10 CFR Part 20 and changes in surveillance requirements. The staff has determined that the amendments inolve no significant increase in the amounts, l
and no significant change in the types, of ary affluents that may be released l
offsite, and that there is no significant increase in indivishal or cumulative i
occupational radiation exposure. The Commission has previously issued a j
proposed finding that the amenenents involve no significant hazards considers-tion, and there has been no public comment on such finding. Accordingly, the i
asundsents meet the eligibility criteria for cateporical exclusion set forth in 10 CFR 51.22(c)(9).
Pursuant to 10 CFR 51.22(l>), no eiwironmental im>act statement or erwironmental assessment need be prepared in connection wit t the issuance of the amendments.
4.0 CDNCLUSION The Cosmission made a proposed determination that the amendments involve no l
significant hazards consideration which was pd11shed in the Federal Register on Decesher 27,1989 (54 FR 53205), and consulted with the State of Georgia.
No pelic comments were received, and the State of Georgia did not have arty cossents.
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The staff has concluded, based on the considerations disanssed abwe, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (2) such activities will be conducted in compliance with the Commission's regulations, and the issuance of the asundments will not be inimical to the common defense and security or to the health and safety of the public.
Principal Ctntributor:
H. I. Abelson, SRXB/ DST Dated: February 20, 1990 8
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