ML20092M144
| ML20092M144 | |
| Person / Time | |
|---|---|
| Site: | 05000172 |
| Issue date: | 01/29/1971 |
| From: | Poore H LOCKHEED AIRCRAFT CORP. |
| To: | US ATOMIC ENERGY COMMISSION (AEC) |
| Shared Package | |
| ML20092L085 | List: |
| References | |
| FOIA-91-233 NUDOCS 9202270183 | |
| Download: ML20092M144 (29) | |
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/ cn LGD/303992 United States Atomic Energy Cormission Division of Reactor Licensing ,, -
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Washington, D.C. 20545
Subject:
Technical Specifications for Rodlotion Effects Reactor, Docket 50-172 J
Gentlemen:
Our lett'er of August, 14,1970 (LGD/299741, Subject, Radiation Effects Reactor, Docket 50-172), advised the Commission that operation of the Radiation Effects Reactor (RER) had been terminated. We also enclosed o set of proposed technical specifications for the RER facility pending develop-me.nt of a reactor decommissioning plon. During the intervening period, representatives of the Division of Reactor Licensing ond_ of Lockheed have had several conferences regarding the odequacy of the proposed technical specifications. Lockheed has, as'a consequence, modified its proposed technico! specifications in several arcos to incorporate several additional restrictions which the Division of Reactor Licensing considers necessary to enhance the safety of the RER prior to the stort of decommissioning.
Accordingly, Lockheed now requests that the proposed technical specifi-cotions contained with our letter of August 14, 1970, be withdrawn, ond-
- that the proposed technical specifications and supporting analyses enclosed
-- with this letter be substituted. Addendum l-to our letter lof August 14, 1970, listing the incumbent membership of the Procedure Review Committee at that time, is still valid and should be considered o part of the new proposed
- technical specifications.
We wish to advise that Lockheed is also continuing its plans for decom-missioning the reoctor facility, and expects to subtrif a proposed decom-missioning pion to the Commission in the very near future.
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9202270183 910725 g PDR SWEENEY91-233 FOIA PDR uk8
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- United States Atomic Energy Commission LGD/303992 If you have any questions about the above or ottoched information, please ,
contact M. A. Dewar, Department 72-14, Zone 401, Lockheed-Georgio Company, Marietto, Georgio 30060, phone Area Code 404, 424-8367.
Your early consideration of this request will be oppreciated.
Very truly yours, LOCKHEED RCPAFT CORPORATION
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/j . . wd H. L. Poore >
Vice President HLP:ph Enclosures H . L . Poor e , under oath, states that the clove and ottoched statements are true to ths best of his knowledge and belief.
Subscribed and sworn to before me this 29th day of January ,1971, of Marietto, Georgia.
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Notory Public My Commission expires: March 1,1971 l
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,' u.umS nunShs ggulatcr, NY In conjunction with its applicction for permission to own but not operate the Radiation Effects Reactor under Facility License Number R-86 Inckheco is enclosing a set of revised technical specification. Changes are identified in the specifications as -r- (revicions), -d- (deletions), and -a- (adlitions).
Following are the basis for the indicated changes:
- 1. SITE A-2. As indicated in our annual report dated September 25, 1968, security of LGNL is subcontracted to a professional security agency. The sub-contractor security personnel operate under the same rules and regulations as lockheed Plant Protection, and there is no degradation in the level of security afforded.
A-3 Inasmuch as there will be no further operation of the reactor, those portions of A-3 which refer to access control during reactor operation are no longer necessary, Appropriate access control requirements have been moved to Section L.2.1. and r:11 be ccvered in the discussion cf Sect;cn L B, BUILDD:GS B-2. Inasmuch as there will be no further operation of the reactor, those portions of B-2 whien refer to conditions during reacter operations are no longer necessary. None of the reactor system controls will be operated.
Certain portions of the primary cooling systcm valving may be altered, but only at the direction of and with the cognizance cf, the Reactor Supervisor.
The reactor lift controls may be cperated, but only at the direction of and with the cognizance of, the Resctor Superviscr. The reactor lift position indication system is only operable when the reactor upper closure and equipnent tanks are assembled en the present vessel. Hence in present shutdown coniiguraticn there is no reactor lif t position indication system.
When the reactor lift is in the fully raised position and there is no shielding water in the shield tanks, the dose rate outside the reactor building is not greater than 100 tr/nr. Hence we visualize no circumstance by which operation of the lift controls would be detrimental to the health and sef~w/ of the general public. The electrically operated coors controlled from.the operations building prevented inadvertent departure of perscnnel fron the operationc building daring reactor operation. Since the reactor will no longer be operated, the requirement for electrically locking the doors no longer exists. The ventilation system kept the operations building at a positive pressare with respect to the reactor building during reactor operation to pr- ..; reactor-activated ar6cn 41 and Airborne particulate centamination from entering the operations building.
Ecace opcration of the ventilacion sys.em is no longer required.
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RADIATION MONITORUKi C-1, opccifications are pertinent only during reactor opera are no longer necessary.
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the current technical specifications constitutes only. nal requirez,ent an operaw fuel is ne longer stored in the primary coolingeactor
, operation of the systemSince t fission products monitor would not reveal a fuel elment cladding rupture.
Analysis of the epecifications it grab samples as specified in the pmposed technical method prescribed in the last sentence of Paragraph to the RER Technical Specifications. 7 4 of C been removed from the f acility, the requirement for monitoring fFurthe
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or inadvercent release of fission products ceases to exist propcsed reduction of sempling frequency. , justifying the C-).
conit:rs is required only during periods of reactor operatio .
C-4 system is required only during periods of reactor operation .
C-5 clarify the mode of cperation of the samplers.The paracraph de The river samplers are the sarple is homogenized. continuously drawing a small mater c cample into at the indicated sampling frequency.A sample is withdrawn from the large container The number of air sexpling locations has been reducedcated to the indi level inasmuch as to document a fissien product a sa .pling program of this tagnitude is emed adequate de release.
The change in werding in the paragraph en the soil aand program is for clarification only.
veget ti on sampling C-6.
i Specificatior of a criticality alam system is deemed necess ary to being stored in a reactor. satisfy the monitoring requirements for special nuc fixed by the Code cf Federal Regulations.Alann set points for the monitoring device a the criticality alam system are intended toof recognize thse e fact that a system by can Lockheed. break dcwn, and to encourage expeditious repair s em of the sy t
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E. DElt3DCY SYSTDG
-The cooling requirements for the fuel have been ca'.culated, and it has been ascertained that using air convection cooling only, the ma.ximum fo;el fa ste temperature will not exceed 413'P. The calculation assumes Spo,cenduction of heat to the fuel element storage rack, which weighs 363 lbs, and which represents a very significant heat sink by conduction.
The current power generation in the two RR cores has been calculated based on the Way-Wigner equations p=0.0622P,[.2 -(ieT7**
Where P = Present power Pg = Operational power level t = Shut /. awn time, seccnds T = Operating time, seconds The first core was cperated for total of 3820.28 IGH between December 14, 1953, and February 2s,1964 The following table gives the number of megawatt hours generated in each year, the last date of operation in each year, and the caxicua power of operation during that years Date of last- Maximum operation in Power Vegawatt hours Year that year level in year 1959 Dec. 30 10 la 793 80 1960 Dec. 1 10 hr# 2655 55 1961 June 27 10 !G 33 32 1962 Dec. 23 1 13 45 91 1963 Dec. 31 1 106 287.65 1964 Feb. 23 1 MW 11 97 For conscrvatist, all operation was assumed to be continuous at the maximum power level, ending on the last indicated date of operation. In actuality, operation occurred in almost every month of each year and at varying power levels up to the authorized level. Hence lueping each year's operation into one continuous period at the end of the year increases the conservatis:c of the calculation. 'The power generation from the first core, based on these assu:ptions, was 2 503 KW on December 31, 1970.
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r A sittilar set of assumptions was cade for the second core based on the followi.ng data:
Date of last Maximum operation in Power Megawatt hours Year _that year Level in year 1964 Dec. 23 3W 802.66 1965 Dec. 30 3n 953 82 1966 Dec. 28 3W 1092 92 1967 Dec. 19 3W 457 53 1968 Dee, tv 3a 422.22 1969 Dec. 11 3s 157.88 1970 Jul. 16 3H 115 10 The power generation from the second core on December 31, 1970, is 9.497 hT.
The fuel elements are assumed to be cooled by natural convection o' air.
A fuel element storage rack with 20 elements, rather than the actual core loading of 26 elements, was conservatively estimated to have a total shut-down power of 9 497 KW. The hottest coolant channel is estimated to dissipate about 120 UTU/hr, or about 1/3 more power than the average channel. The temperature rise of the air in passing through achannel ist bT S ii ! p where q is the heat flow into the channel, c is the specific heat (a value of .243 BTU /lb'P is assumed) and i pis the mass flow rate. 6 = /i 0 where \' is the velocity of the air, A is the cross sectional area of the channel (.0018 2ft )and 9 is the density of the air in lb/ft). /0 ; A. (p t.
where the zero subscript refers to initial conditions (that is, upon entering the channel); T is the absolute temperature: [,,=.0711lb/ft)foranassumed To of 560'l. The velocity which the air ;eaches in traversing this channel is thus kb (1)
V = T-560/ see The acceleration due to the bucyant force resulting from heating the air is given b, a = g ;.ST where g is t' e acceleration due to gravity (32.17 ft/ U ;* ); 8 is the volumetric expansion coefficient (0.002033/*?);
/L T- is the increase in temperature in 'F of the air. The average acceleration of the air in passing through the fuel element may be taken as a/2. Hence:
(=(32.17) (.002033) o T = ( .0327 A T) Lt2 2 see With an initial velocity of zero, the distance the air stream traverses in time { is given by z = 1 K t 2; with z = 2 ft, t = (4/a)I.
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The velocity after time T is: 1/ = at = .362 (AT)2 ft/sec (2) p __g g , e 'W
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Cocbining equations (1) and (2) provides an exit temperature for the air of 627'R (367'F). The mas. mum temperature differential between the air and the fuel plates is estimated by tsT = i A where q is the heat-flow rate (a value of 90 BTU /hr or about twice tid'avsrage value is assumed to approximate the maximum conditions) from one face of the fuel plate; A is the surface area of'ene face of the fuel plate (.2 ft x 2 ft) y is the distance from the fuel plate surface to mid channel (0.0045 ft) k is the conductivity of air (.022 BTU /ft br'F). Hence 4 T = 46'F. The maximu heat flow rate will occur near the midplate of the fuel. It is thus-conservative to consider the maximum tc=perature differential between the air and the fuel plate at the top of the fuel. The maximum fuel 71 ate temperature would thus be 413'F which is far below the melting poin; of about 1220*F.
The te=perature differential across the .02 inch aluminum cladding and across the fuel itself ( 02 inch thickness) was found to be on the order of milli degrees.
Ioss of power at the RE? will activate the criticality monitoring system alarm. Hence from the standpoint or criticality monitoring, no need exists for emergency power. To inhibit corrosion of the fael cladding, the pool in which the RER fuel is stored is maintained at a high degree of purity by the pool circulating system. It has been determined that the quality of the pool water remained within acceptable limits when the pool circulating system was shut down for a period of two weeks. Hence it is concluded that temporary loss of poser at the REP will not lead to a condition detrimental to the cladding on the reactor fuel, and thus no need for emergency power exists.
On the basis of the above observations, no description of the operational l
emergency systems is considered necessar/ in the Technical Specification.
F. CORE COUOZT STORAGE F.1 The changes in this section have included reidentification cf the title to include the reactor start-up source; description of the current core co:ponent inventory; and establishment of that inventory as the upper limit.
Lockheed has on hand at LGNL (but not at the REP) 33 new fuel elements and ,
5 new control rods which have'not been irradiated and which are still l stored in their original criticality-safe shipping containers. At the 1 time shipping casks arrive at 1425, for removing the irradiated fuel, it may be desirable to ship some of the new elements along with the irradiated fuel. Inasmuch as there are no Lockheed facilities outside the RIF for i opcning the shipping cask after irradiated fuel has been placed in the l cask, it will be necessarf either to 1 cad the new elements into the cast 4
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d before the cask in taken to the REP, or to take the new elements to the RIP and load th(*r into the cask while the cask is in the storage pool.
Kritten procedures for accom;411shing the handling and loading of the new elements till be subject to the approval of the Procedure Review Comi tt e e . The procedures will incorporate the cri,eria that no more than one fuel arscibly will be out of a criticality-safe container at any one time; and that all fuel handling will be conducted under the direct surveillance of the heactor Supervisor.
P.3 A question has been raised by the United States Atomd c Pnergy Commission no to the poenibility of spilling fuel elemento into a critical array during handling of the fuel elcment storage racks. Spilling fuel from the storace rach ic precluded by the storage rack design. The lif ting tail on the rack will not operate unless the rack lid to closed
, and inechanically locked. As long as there is pressare on the lid in either directier., the opring-loaded latch cannot be released. Hence epillinc of the fuel to ast;uted to be incredible.
F4 At the twe the resetor facility wris designed at.d built, the anticipted turnover of reactor fuel was expected to be a core loading approxintely every three cenths. A criticality-safe fuel element chipping ensk meeting Bureau of Explosives requirteents was provided stth the reactor, ar.d an undernator cutting tool sas also provided to remo'/e excests tetal (end boxes l lif ting balls) frem the fuel assen.blies. The shippinc cask v ta rited to acecanodate the fuel assemblics only when the exercs octh! had b<en removed. Lockheed will not attempt to get its shippinc cask t:pproved fer chiptent of the irradiated fuel, but will instead plan shiprent of the fuel in lictnded fuel tihippinc caths. Depending upon the economics involved, bcwever, Imkheed tudy elect to remove tne excetti itetal, using the underv.ater -
cutting tool, be fore the fue) is t!Qped Tat tool ct.rriage contains cuides for posm:ning the fuel assembliert as necessary to assure that no cut n11 Le made into a fuea plate. If it is dttermined that tne excess tctal will be cut cff prior to shiptent, written proc edures for utilizing the cutting tool n11 be prepared, and will be subject to review and approval by the Trocedure Review Constittee. The previously stated conditions on administrative centrol (no more than one fuel acoembly being handled at cne time; all handling under direct surveillt. nee of Reactor Supervisor) will also apply.
O. FINACR SHIELD TA'IS The basic purroces for the reactor shield tar.hs were (1) to serve as a housinc for the reactor oystem nuclear detectors; and (2) to attenuate the neutron flux (tid thus minittize neutron activation) in a]I directione cxcep ton.rd tu tm article beind irra:iit ted. Since the fuel is no lencer in tu reactor and the reacter will not be loaded and o;>erated rmy store, the above functions are 1.c lencer necessary.
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- 11 - 2 . Shutdom procedures exist which provide that, during non. operational
- pe.lods, the reactor lif t may be ineobilited at any desired level by means of hand-operated bleed line valves. Procedures also exist which provide i that a specific storage rack will be used for storing tha upper closure wh(.1ever the upper closure is re noved frce the pressure ,essnl. There is no degradation in the level of safety involved in the changes which have teen made in this section, and the only purpose of the changes is to i document the above practices, t
3-3 The only changes to the descriptien of the internal cceponents are to exclude fuel from btains reloaded into the reactor and to specify the 2
manner of storage of the other components. Neither of the above changes represents a degradation of reactor safety, and both act2ons are permitted by current operating and/or maintenance procedures.
H-4, The changes to subparagraph b. of this section a4= to specify the actual inventory of drive techanisms and to indicate approved areas for storagt. These chances represent approved procedural practices, and do not represent a derradatien of reacter safety. The deletion of the operating limitations (subparagraph c.) applies only to operation of the reheter and hence does not degrade safety in the shutdown facility.
H-5 The change to subpararraph c. of this section deleted authorization to use a plutentu~.-beryllium source and prohibit reactivation of the antimony-124 scurce. The above eht.nges and the changes to subparagraph d. all apply to cperating ccnditions; since there will be no further operation of the RER, none of these chanres will affect reactor safety.
I. NU; LEAR A!C FRO:ESS INSTRU)520'ATION In previeus discussicns it has been hown that the only requirements for continuous instrzentation consistr. cf monitoring for criticality and
- sitorinc the level of water in the s*orage pool. The sections of the opersticnal technical speelfications applied only to conditions during facility operation, and none of the conditions of the operational specificatiens could, in our opinion, be considered mandatory to assure the safety of the facility in its present configuration. .
J. DJER11GL'TAL PACILITIES
-Inasmuch as there will be no further operatien of the RER, and since the res tor core has been un17aded and stored in the storage pool, the limitations icposed on everinents by this secti:n are no longer pertinent. The one exception is the 3;thium hydride shield which is currently stored in the
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r reactor building bancc)ent, and negotiations are undersay to dislose of I the shield to a licensed waste disposal firm as radioactive waste. A pressure cheek of the shield is performed weekly and the proposed technical specifications request that the interval be changed to monthly.
Tbere has been no indicated change in pressure for over two years.
The restrictions on operatien of the locomotive and on the presence of railway cars were 1eposed because of the potential of damaging the ,
reactor and pessibly crushing the fuel in the reactor. Since the fuel is now located in the storage pool at a horizontal distance of more than 20 feet from car position 3A , it is considered unlikely that a car or locccotive could obtain enough kinetic energy to leave the tracks and ,
impact on the fuel. Cara cr locomotives on tracks i and 2 or on car position 5/6 would have kinetic energy only in a direction 90' free the ,
direction to the stored fuel and at a reinimum horizontal distance of 10 feet from the fuel. Hence it is considered unlikely that the fuel could be damared from those directions. Therefore, it is considered that there i te no need for restricticns on the use of railroad rolling stock. For the same reason, the requirement for the derail at the exclusion fence gate was also icposed to assure that a runaway locccotive could not reach the rea: tor, e.nd this reNirement also no longer existe. Use of the derails in the $ndicated ecnfigurations will as a ratter of course be continued, but we propose that there be no requirement for reporting and logging the derail configuratien. l E. REACTOR C001E T Sv33%
As previously discussed in Section E, the.re it: no requirement, under either nomal er emertency conditicns, to provide a cooling system in the present cenfiguration. feriodically circulating the pool water through the pool and/or bypass domineralizer has been shown to maintain the pool sater purity at a hvel adequate to retard corrosion of the fuel cladding.
Limits for the watcr quality are stated in Section I, and the pool water will be circult.ted through portions of the primary coolant system and the dominera11zers when necessar/' to maintain water quality within these limits.
L. AINU ISTRATIVE Ah'D PRCGEHAL SA5'EGUARDS IM . With the shutdem of 141, the emphasis has shifted fr:m operational safety to that of maintaining a se.fe configuration during the planning and conduct of decommissiening of the reactor f acility. The individual respcnoitle for t.aintaining the safe conficuration has not_ changed, but the :retheds by wh10n we intend to anure this continuity of safety are refle cted in tne pr posed technical specifications. The responsibilities assir.c:i the RM. :r Supmirer are the see responsibilities ansigned to the tenu r entner and h;c tern in the cperational technical t;(;.: . a.-. ,
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The Reactor Safety Consnittee, which was responsible for experimentai ;
safety, is replac$d by the Procedure Review Cournittee, three of whose four incumbent cembers are also members of the Reactor Safety Comittee.
The job titles unich were specified for the membership of the Reactor ,
Safety Co::cittee have been abolished, but the last incumbents in those pesiticns er their alternates are continuing as members of the Reactor ,
Safety Co:::tittee until adopti:n of these proposed technical specifications. !
Ex:ept for the fact that the Procedare Review Comittee does not encounter problems involving safety of reactor experiments, the functions of the Reactor Safety Comittee and of the Procedure Review Cotraittee are essentially identical.
V. The provisions for control of access to the exclusion area are essentially identical to those contained in the operational technical specifications, And procedurally have been moved to this section of the 5 specifications because they are now administrative procedures not offe tins resctor safety.
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A. SITE j l
- 1. PFJSICAL LOOATION The reactor facility is located in Dawson County, Georgia, on a site whicn is nominally described by the parallels }4', 2( .6 r.inutes north '
latitude and 34* 2h minutes north latitude, e.nd the meridians 84* OS l
minutes nest longitude and B4' 12 minutes west lon61tude, l
.r- 2. DESCRIPr10N OF C0!; TROL 1ED AREJ.
The reaeter is located within the Georgia !!uclear Laboratories, a contr:11ed area of rouchly 10,000 acres. The nearest uncontrolled areas are the south periccter fence (8240 feet soutf), the east periteter fenre (9520 feet east) and the v'est perimeter fence 20 feet mest). Tne minimas distance to the north perimeter fenc 4265 feet. All land within that area is controlled by Plant Security, and >
the nearcst rout:nely occupied above-ground work area is about 6345 feet frce the reacter. A chain link exclusion fence surrounds the reactor Ect.erally at a radius of 3600 feet. A ser; cent of the fence fiorth East of the reactor is slightly closer than 3600 feet.
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-d- 3 IIC11.3ICU /JZA ACCESS CONTROL
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i B. LUIIDDIO 1 RF.ACNR BUILDD;0 The reactor building is of conventional construction with steel I-beam colur.as and built-up truss work. Siding and roofing are corrugated alur.inum. The building is not beated; however, during periods when the ambient te:tperature is below freeting, the reactor will not be raised from the pool unless heat is provided to prevent auxiliary piping (e.g. shield tank plumbing system) from freeting. Roof mounted fans are provided to ventilate the building as necessary.
-ro 2. OW. RATIONS BUILDDiO The operations building is an underground concrete structure with ,
approxirrately 2 feet c,f concrete and 5 feet of earth on the roof to provide shielding.
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C. RADIATlal HalITORUJO
- d. i. GASEDUS AND PARf10ULATE MONI1CR
-r- 2. RER STORAGE 1%L While reactor fuel is stored in the RER storage pool, grab samples of the storage pool water shall be collected at least weekly and analyred for gross beta-gama and gross alpha activity. If the pool pump is not operable at the time the sample is to be taken, the sample shall be taken from within 3 feet of one of the baskets containing stored fuel elements. If a sample has a radioactivity concentration in l excess of 1x10-2 c/ce, an invertiEntion shall be made as to the cause. )
If it is found that a fuel elemer.t is leaking, action shall then be taken to identify the leaking element and seal it in a leak-proof l container.
-d- ). B'JILDD:0 MONITORS
-d- 4. RULOTE AREA MONITORD;G 6fSID4
-r- 5. UNIR0!WDITAL SM511h0 A minimu'n of two continuous water samplers shall be located in the Etowah River, at least one being located 3600 feet or more upstream from the reactor e.nd at least ene 3600 feet or more downstream from the reactor. Samples shall be collected and analyzed at least twice conthly the collection interval shall be approximately two weeks except when veather conditiens make the sample collecting points inaccessible.
Samples shall be analyzed for gross beta-gama and gross alpha activity.
A continuous particulate air sar:mler at the REM demineralizer building entrance shall be cperated. Filter aucples from the particulate air sampler shall be collected and analyted weekly for gross beta-gama and gross alpha activity. Operation of the air sampling system may be discontinued after the fuel is removed from the RER storage pool.
Soil and vegetation samples shall be collected and analyzed for gross beta-cama and gross alpha activity at Icast quarterly.
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-o- 6, CRITICALITY A1>JM While remeter fuel is stored in the RER storate pool, a criticality alam system shall be operable. The location, sensitivity, and alam set point of the monitoring device for the criticality alam system shall cceply with the require: tents set forth in the Code of Federal Regulations, Title 10, Part 70, Paragraph 70.24 (a) (1) as arrended.
The system shall alam in Plant Security Headquarters. Response to alams shall be stade by Pla.nt Security personnel in accordance with '
procedures approved by the Procedure Review Ccenittee. Tne alam
- system shall be checked for operability at least weekly. If the criticality alarm system is non-operable, the radiation level in the RER storage pool shall, during the first 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of such non-operable periods, be monitored at least once every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by an individual i authorized by Health Thysics utilizing an operable currently-calibrated portable survey meter. After 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, the tnonitoring frequency shall !
be increased to at least once every 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> until the criticality alam system is operable. No activity involving fuel handling nor pool sater transfer other than nortal circulation shall be conducted while the criticality alarm system is inoperable.
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D. WASTE DISPOSAL SYSTfM The TIP waste disrcsal system is designed to handle activation products generated by RDt operation. Features incorporated in the design include de:tineraliser resin beds with decontamination factors of 10} to 10 4: two 50X) sallon waste decay tanks for retention of waste water with high radioactivity concentrations l and one 150,000 1 gallon hold and drain tank which may be used for decay and dilution '
of activity prior to release to the seepage basin. .
We activity Icvels of radioactive waste releasert to the seepage basin will not exceed limits specified in 10 CPR 20 for restricted areas. The total annual release will not exceed 1 curie.
Radioactive waste materials will not be permanently disposed of by burial at the site except as provided by 10 CPR 20 304.
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.r. E. IM3tGDICY SYST12 '
There shall be no requirement at the REP for emergency cooling or emergency power. :
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P. CORE CQ/,PMD.T STORAQE ,
- 1. 1 W U CORY The number of irradiated fuel elements on hand at the REF is 56.
The number of irradiated control rod fuel assemblir,s on hand at the REP is 8. There are elsewhere at 14NL }3 new fue) elements and 5 new control rods. When being prepared for shipment off-site, these fuel elements and control rods may be brought to the REP only in accordance with procedures approved by the Procedure Review Contittee. Other identifiable radioactive material (i.e., licensed byproduct material as distinguished from activated reactor components and activated tools and experimental equipment) consists of the antimor./-124 reactor startup source (currently Ices than 10 curies) and about 10,000 curies of encapeulated cobalt-60. All core components and byproduct material may be disposed of by authorized means.
- 2. STORAGE Irradiated fuel elements shall le stored in fuel element storage racks lat the bottom of the storace pool. Irradiated centrol< rods shall be
' stored either in control rod storage racks along the pool wall; or in control rod storage positions, t*o of ubich are available in each fuel elecent storage rack. Antimony-124 shall be stored in the pool or in the pressure vessel. Any cobalt-60 in storage at the REF shall be stored in the pool. Fuel shall not be transferred to, nor stored in, the reactor.
3 PMEL ELDCC Stolb GE RACT.S Fuel element storage racks are constructed of aluminum, and contain layers of aluminum-clad .cadmiua around the oatside of jbr fuel plate resten and between each tier of four fuel elements. Each rack shall hold a maximum cf 20 fuel elements in a 4 x 5 array. The 4 x 5 array has accommodations for ster:nc two control rods in fuel element posit;cns. The calculated U-235 loading required for the fuel elements in- one fuel element storas;e rack to achieve criticality, assuming no leakage, is 211 Cms per element.
When leakage is assumed from the sides only, a full rack of 211 Cm elements would have a calcalated multiplication constant of 0 78.
4 FREPARAT10:1 FOR SHIWm7 When fuel is being prepared for shipment off-site,' fuel element end boxes may be removed from the fuel assemblics before the fuel is loaded into the cask. Tne poison scetions and lifting balls may be removed from the control F-1
rods before the control rod fuel sections are loaded into the cask.
Such removal shall be acev:nplished only in accordance with detailed procedures approved by the Procedure Review Co:ntittee. Casks used for recoval and shiscent of fuel shall have been duly licensed for such usaEe in accordance with appropr; ate AEC & DOT reEulations. Procedural control shall be used to pseure that no more than one fuel assembly at a time is out of a storaEe rack or a shipping cask. Individual fuel assemblies shall be handled only under the direct personal surveillance of the Reactor Supervisor.
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The reactor vessel is surrounded by sepented shield tanks. The shield ,
tanks in the quadrant 180' away fror car position 3/4 shall be approximately 20 inches thick. The shield tanks in any of the other three quadrants ,
may be either 8 inches or 20 inches thick. The tanks may be either tilled or drained. The shield tanks may be removed from the reactor.
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- s H. REAC1CR DESIGN 1 LOCATION The vertical center line of the reactor is located approximately 4'6" free the NE end of the reactor pool. The pool is rectangular and is 11-1/2' by 10' by 36-1/2' deep. There is also a storage tool which joins the reactor pool at the SW end. An aluminum gate is previoed for separation of the pools. In plan view, the storage section resembles an un-symmetrical letter T. Rough dimensions are stem, 17' by S'; cross, 24' by 6'. The depth of both of these parts is 21' below the finished reactor building floor. Curbing for both pools is continuous and extendo one foot above the finished floor.
-r- 2. PJEMA'JICAL DESIG?i OF EE RER 1RESSURE VESSEL The RER stainless steel pressure vessel is designed for 150 psig at 200'P.
The tninimwn design and construction rrquire1ents of *he vessel conform to the AS:7. Teiler and Pressure Vessel Code,Section VIII,1956 Edition, and the vessel bears the official code stunp. The pressure vessel is supported by tuo steel bands welded to four equally spaced vertical members which rest on bearing plates bolted to the platfore, which in turn rests on top of the hydraulic lift. The reactor eurport eystem has been designed to support a enc-fourth 0 side load on the reactor and shield tank.
The design loading er the hydraulic lif t is 42,000 rounds, and it is designed for a tots 1 moreat of ',6,500 foot-rcunds free eccentric leading and other causes. The 11 't bas a stroke of 30 feet and is capable of raising the reactor at a naxitum speed of 10 feet per minute. Shoes which slide on T-rails fastened to the pciol walls guide the upper end of the ram through its full travel anc restrain the top of the reactor to within one inch cf its nominal path.
Two rsrallel bleed lines, centro 11ed by individual solenoids which fall open on loss of power or on itpro;+r operation of the safety doors in the operations building, are used to lower the reactor. The bleed lines may be valved off so that the reactor tay be kept in any desired position regard-less of the status of electric power and other automatic interrupt devices.
The reactor upper closure is a flat, circular, forged plate } feet 9 inches in diameter and 5 inches thick. This closure is equipped with holes to accommodate the control rods, the regulating rod, t.nd the fission enamber.
It can be removed to provide access to the internals of the pressure vessel.
The upper closure Aay be located either on the reactor or off the reactor.
A star.d at one end of the stcrace pool may be used for storace of the upper closare.
In additicn to u.c cptning clos 2d by the reactor upper closure, the Ir'ssure vessel his four penetrations above the c:re, cons 13 ting of 6-inch instrament ports, bolcw the tsp of the core, only four renetraticns exist. These are H-1
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four 8-inch pipes shleh serve as tuo primary coolant inlets and two prieary coolant outlets. No new penetration shall be added to the pressure vessel or reactor closure.
-r- 3 PIR VESSEL DtTERNAL mw@NRE Internally the vessel consists principally of the inner tank, the hold-doun plate, and the core support structure. The inner tank, which is open at the bottom and top, and otherwise has no penetration, serves as a flow guide. The hold-down plate, located above the core, covers the entire core section. Since flow through the core is down, the hydraulic loading of the hold donn plate is not a factor. If the reactor should
- inverted, the maximum loading on the hold-down plate would be less than 10,000 psi. The yield stress for the hola-down plate is 32,000 psi at 200'F.
The core support structure consists of a grid plate to position the various core components, a support plate which retains fuel elements, reflector elements airs the start-up source within the core and a control rod shock damper whleh consists of a main cone, individual shock absorber tubes, and associated structural members. Its function is to transmit the shock load which the control rods impart subsequent to scram to the pressure vessel rall. As a backup to the shock damper, a mechanical stop is welded to the botto:s of the pressure vessel.
The entire core structure is supported by a ring selded to the pressure vessel vall. The design criteria on Stress for these structures are a ma.ximum stress of 7200 psi on the core support bracket, 2440 psi mavimum stress on the support plate,14,500 psi maximum stress on the grid plate, and 5600 psi on the cone. The yield stress for all of these components is in excess of 25,000 psi.
Fuel elements and control rods shall not be loaded into the RER vessel.
The hold-down plate, inner tank, and grid and scram damper assembly may be in the PIR vesst , or may be stored in the reactor building or storage pool.
The start-up source, du::r::y fuel elements, reflector assemblies and flow baffles r.ay be left in the RER vessel or may be removed from the vessel.
-r- 4. C0!(fROL SYSTE1! AND OPERATH 0 LIMITATIONS
- a. Control Rod Design The contn>l rods are fuel and poison sections' enclosed by aluminum tubes approxicately- 3" square by 65" long. Each has a grapple head at the top and a spring-loaded tip PNge/ at the botto:t. The fuel section contains an aluminum strap extension at the top which extends the length of the poison section. The poison secticn, a square aluminum tube, slides onto the strap l
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- i and fits flush a6ainst the top of the fuel element assnmbly. The entire fuel-poison assembly fits into the control rod tube. A mechanical attach-ment on the fuel-poison section prevents assembly of the control rod if the fuel-poison section is inverted. The centrol rod is guided and supported within the core by means of four rollers above the core and four rollers below the core. The lower end of the control rod fits within a scram guide tube, ubich also acts as the hydraulic damper during scram. The roison section is a square cadmium tube, 0.02 inch thick and 32-1/4 inches long.
It is clad with a 0.02-inch layer of aluminum on each side so that all edges are sealed. The length provides approximately 4-1/2 inches overlap at each end of the active fuel plates in the reactor core.
The fuel section, which contains about iii crams of hi8hly enriched U-235, is similar to a standard fuel element; however, it is smaller and contains -
14 plates. A mechanical stop at the bottom, and the affixed poison, position the fuel section within the centrol rod.
The upper end of the control rod tube is fitted with a lif ting knob, with shich the control drive grapple ergaEes by electro-magnet actuttion, for 1lftingthecontrolrod.
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The core regulating rod is located near the peripher7 of the core. The regulating rod poison is a 30-70 cadmium silver alloy material. The tubular poison section, which has a nominal thickness of 0.09 itch, is enclosed in a tubalar aluminum sheath approximately 1-1/4 inches in diameter.
- b. Drive Mechanisms The four control rods are actuated by separate mechanisms, mounted to the top bead of the pressure vessel. Each control rod drive mechanism consists of am electric motor, reduction unit, a rack and pinion, limit switches, an '
electro-magnet and grapple, a spring loaded scram tube which provides an initial 5-g accelerating force to the rods when the grapple is released.
Tne rx.imum drive speed is 4.5 inches per minute.
Magnets in the centrol rod and a limit switch in the hold-down plate indiate by an electric signal the position of the rod wnen fully scrammed as well as engagement of the rod by the mechanism. A selsyn, mounted on the gear reduction casting, gives a continuous position indication of the drive and also an indication of rod position daring normal reactor operations.
The reculating rod drive mechanism, mounted to the top head of the vessel, serves to drive the reculating rod. The mechanism and rod are bolted tecether so that the combination is an integral unit. The ragulating rod drive mechanism is ver/ similar to the control rod drive mechanism, but no scram attachment is provided. The drive motor is designed to operate with a servo centrol system or under manual control.
Including opares, there are 9 control rod drive nachanisms, ) regulating rod drive mechanisms, and 2 fission chamber drive mechanisms. The drive i mechanisms ir.ny be stored either on the upper closure or elsewhere.
-r- 5 CORE OFERATINO LIMITATIONS
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- a. The core, which has an active height of 24 inches, is designed on a 3-inch modulus in a 6 x 7 array with the four corner positions occitted.
The coderator and coolant, are light water. The reflector may be light water, or may be solid or canned aluminum or beryllium designed to conform to the unused spaces in the grid and external to the grid but within the inner tank. '
- b. Puel material is uranium-alumin' alloy. The enrichment of the fuel is nominally 931 U-235 Cladding 2 setallurgically bonded 1100 aluminum.
The fuel elements are flat plate, m, tried WR type, aluminum, uranium assemblies. Each element contains 18 fuel plates having the approximate dimensions 0.060 inch thick, 2.75 inches wide, and 24.5 inches long. Each fuel plate consists of a nominal 0.020 inch thicknees of uranium-aluminum alloyiin a picture fram clad with a nominal 0.020 inch thick layer of 1100 altninu*n. The plates are positioned in the element by aluminum side plates so that a nominal 0.103 inch wide coolant passage is provided between fuel plates. Each element is loaded with approximately 176 cms U-235 The top end of each element has a handling device. The bottom of each element is equipped with.a positiening box about 3 inches square which fits into the grid. Tne overall length of fuel element is nominally 33.5 inches,
- c. The start-up source is an antimony cre:ma emitter, placed in the center of a berylliun dum:::y fuel element, and is positioned in one of the available ppare fuel element positions in the grid plate. For operation, the source provided a neutron flux of at least 15 nv at the fission chamber location.
There n11 be no minimum source stren6th requirement at present. The source shall not be reactivated.
- d. Puel content verification and other core partereter deteminations will nomally be conducted at the CER. The following limitations will apply, howevers (1) The maximum number of fuel elements in the core shall not exceed 33 (2) The maximwn U-235 content of the core shall not exceed 6.2 kg.
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.t. I. NUCIF.LR AND TROCF/>S INSTRUVlNTATION
- 1. hT,1IAR Dl!TTRLM2iTAT10!i The criticality alam system described in Section C.6 shall alam in Plant Security headquarters. Response to alams shall be r.ade by Plant Security personnel in accordance with procedures approved by the Procedure Review Ceanittee. Af ter the reactor fuel has been removed from the pool, use of the criticality alam system r.ay be discontinued.
- 2. PROCF.SS DISTRWRITATICel The reactor storage pool water shall be monitored for til and conductivity at least weekly while reactor fuel is stored in the pool. Conductivity and p3 shall be measured utilizing either portable or fixed instrumentation.
If the pump is not operating at the time the saeple is to be taken, the sample shall be taken frcm within ) feet of one of the baskets containing stored fuel elements. pool water resistivity shall be maintained greater than 250,000 ohm-em, and pool water pH shall be maintained between 6 and 7 5. A pool level tenitor chall sound an alam in riant security head-quarters if the pool water level drops below to feet. The pool level monitor shall be checked meekly for operability. If the pool level monitoring system is inoperable, the pool level shall be checked visually at least once every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. After the reactor fuel has been removed, operatien of the pool circulation system and monitoring of pH, conductivity and wate. IcVel rey be discontinued.
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-ra J. EXTER1YWfAL PACILITif3
- 1. OD ERAL Except for the lithium hydride shield, no experiments or equipnent having significant kinetic energy shall be operated in the reactor building. No experiments or equiptent involving energetic fluids or staterials having a potential of sudden release of chemical energy in excess of 0.1 lb of Nf shall be stored or used closer than one foot to the reactor pressure vessel. All amounts at other distances may be detennined on the basis of R26eometric attenuation subject to the further limitation that the total potential lateral loading of the pressure vessel support structure shall not exceed the allowable desi6n loading. There shall be no restriction on use of the loccxnotive.
- 2. LITH 1W. HYDRII)E SHIEID A helium environment shall be maintained on the lithium hydride at all times. When installed in the facility and not in use the shield shall be stored at a positive internal pressure. Monthly checks on pressure shall be perfomed in order to detencine the leakaEe rate from the shield.
Gases vented from the chield v.11 be monitored for tritium content during release to the envirorrent. Tritium released to the environment, when averaged monthly, Aha11 not exceed 1% of the applicable limit of the Code of Federal Regulatione, Title 10, fart 20, at the site boundaries. The lithium hydride shield oxy be disposed of in sceordance with the Code of Federal Regulations Title 10 Tart 30.
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. r'. K. RFACZOH C00LWS SYS7f2 There is no further need for a reactor prieAry cooling system. As long as reactor fuel is stored in the storage pool, the prinary cooling system valving shall be set to permit the storage pool water to be circulated through the pool ed/or bypass demineraliter.
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- 1. UNINISTRATIVE ORGANIZATION AND STAFFING
- a. Organization The Reactor Supervisor, appointed by the lockheed-Oeorgia Ccepany Chief Egineer - Experimental and Avionics - shall be responsible for, and shall maintain surveillance over, all activities within the RER emelusion fence. He shall have responsibility for maintenance and removal of equ1 1 :nent, maintaining the integrity of the fuel elements, and precluding the release of radioactivity. A qualified Health Physicist, appointed by the Chief Engineer - Experimental and Avionics -
shall be an advisor to the Reactor Supervisor.
- b. Procedure Review Connaittee A /rocedure Review Comittee, appointed by the Chief Engineer -
Experimental and Avionics - shall monitor activities arising as a result of activities within the RER exclusion fence. As a minimum, the Procedure Review Comittee shall consist of the Reactor Supervisor, the Health j Physicist, and two additional scientists / engineers, one of whom shall be ;
chairman. Each scientist /enEineer on the Procedure Review Comittee shall have at least 5 years of experimental, design, or operational experience with a' test or power reactor. Addendum 1 lists the names and qualifications of the incumbents. % hen any chanEes are made in the Procedure Review Comittee membership, Lockheed shall advise the Co:anission within 30 days of the nature of the chantes and the names and qualifications of new members. All procedures _ pertaining to activities within the RER exclusion fence shall be subject to the approval of the Procedure Review Cecanittee.
Actions of the Comittee sna11 require unanimous agreement. The Concittee r shall meet at least monthly.
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- 2. PROCEWRAL SAFEGWRDS ,
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- a. Exclusion Area Access Control-Plant Security shall centrol access to the RER exclaston fence by controlling keys to the gates. Personnel may enter the area within the 3600-foot fence provided that each such entry is governed by the provisions of administrative procedures approved by the Procedure Review Cocmittee.
~~ Access to the Reactor Ba11 ding shall be under the control of the Reactor Supervisor. The roll doors on the RER building shall be imobiliacd whenever the. facility is not manned. All other doors into areas of the ,
REP where radioactive materials or radiation areas are present shall be kept locked except when authorized personnel are in the area. Areas within the exclusion fence which constitute radiation areas shall be roped off and posted with appropriata ram atinn warning mirr.s in accordance with the Code -
of Pederal Regulations, Tith 10, Part 20. While the reactor fuel, start-ap !
source, and OR% owned Cobalt-60 are in the facility, Plant Security shall 11 e
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., 'i, patrol the exclusion fence at least once per week. After the reactor fuel, start-up source and Cobalt-60 have been removed, Plant Security shall patrol the exclusion fence at least once every six months. The Radiation Effects Facility shall be patrolled at least daily by Plant Security or operating personnel.
- b. Emergency Procedures '" ' ' '
Detailed emergency plans and procedures, covering all classes of potential REP incidents, shall be prepared and published in the GE lhergency Manual.
That portion of the 1GE lhersency hnual pertaining to the REP shall be reviewed and approved by the Procedure Review Cocrnittee,
- c. Puel Element Manipulation All fuel handling operations shall be conducted in accordance with written procedures unaer the direct personal supervision of the Reactor Supervisor.
- d. Health Physics Surveillance Health Thysica personnel shall maintain surveillance over activity involving handling of radioactive materials. The Reactor Supervisor may act as a self monitor '
and may provide routine health physics serviced for daily activities involving routine facility surveillance only.
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!!ealth Physics personnel shall specifically establish the radiation protection requirements involving the handling of radioactive material, and shall be in attende.nce for such activities as deemed appropriate by the Health Physicist.
- e. Maintenance Routine maintenance shall be required only on those items and systems which these specifications state must be maintained in operable condition.
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