ML20030D425

Safety Considerations for 24 Element Graphite-Reflected Core. Modified Ntr Core Loading Diagram & Fr Pages Encl

ML20030D425
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Site:	05000087
Issue date:	12/03/1980
From:	WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
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ML20030D417	List: ML20030D416 ML20030D418 ML20030D422 ML20030D423 ML20030D425
References
0078C, 78C, NUDOCS 8109010374
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{{#Wiki_filter:. f~sb) WESTINGHOUSE NUCLEAR TRAINING CENTER NUCLEAR TRAINING REACTOR Safety Considerations for the 24 Element Graphite-Reflected Core December 3, 1980 1. As defined by the Technical Specifications, Section 4.0, operation of the 24 element graphite-reflected core (Figure 1) is an " experiment". Therefore, orarations to evaluate the 24 element reflected core are authorized by the current facility operating license. 2. However, if the 24 element graphite reflected core is to be considered the normal configuration, then the reactor, as described in the Safety Analysis Report, will be changed. 10CFR50.59 states that this Og may be completed without prior NRC approval provided that there is no unreviewed safety question involved. The identification of an unreviewed safety question will necessitate an amendment to the operatinag license. Establishing that there is no unreviewed safety question involves demonstrating that: (i) the probability or the consequences of any analyzed accident are not increned, (ii) there is no possibility of an unanalyzed accident occuring, and (iii) no margin for safety in any Technical Specification basis is reduced. 1 i i O nniOP 8109010374 810825 PDR ADOCK 05000087 s PDR

3. Probability and Consequences of Analyzed Accidents The following accidents are analyzed in the Safety Analysis Report: Flood Earthquake Windstorm Fire Loss of electrical power Uncontrolled rod withdrawal Uncontrolled moderator insertion Failure of experiments Reactor loadings Temperature coefficient measurements Void coefficient measurements importance function measurements Flux distribution measurements and power calibration Reactor physics parameter measurements Irradiation experiments Operational demonstrations and experiments Mechanical damage and failure Reactor Maintenance Radiaticn conditions and analysis N0rmal operating levels Loss of shielcing water Radioactive material storage areas Fuel storage Radioactive sample and source storage Fuel element handling and irradiated material handling Maxim m creditable accident Most of these analyses would be unaltered by the 24 element graphite reflected core. Exceptions are discussed below. 3.1 r i-e N e graphite comprising the reflector rods is not normally considered a combust 101e material. Graphite does undergo oxidation in air in a reaction the rate c# which is increased with temperature. Normally the graphite rods would be lx atea underwater. Even when oJt of the water the scarcity of other flammable ::,5terials inside the building makes it unlikely that the graphite could ever -each the threshold temperature of about 800*F for the oxidation reaction. ) G 0078C

o O Graphite has a very small neutron absorption cross-section (on the order of.003 barns for thermal neutrons), therefore, the reflector rods will not become activated during reactor operation and consequently would not contribute to a radiological hazard in the unlikely event of a fire. 3.2 Uncontrolled Moderator Insertion Inserting the reflector rods has the same effect on the reactor as a moderator intertion. The rods are inserted by hand, which means that the reactor must ae shut down to make the insertion. Since the fully reflected core is still substantially subtritical with all control rods inserted, the insertion of reflector rods does not represent a reactivity addition hazard. The Safety Analysis Report considers it incredible that a core be loaded without control rods present. 3.3 Temperature Coefficient Measurements 3 The reflector rods will each displace approximately 260 in of 3 3 moderator-shield water for a total of about 5200 in -3ft - 22.5 gal-lons, this represents less than 1 percent of the approximately 2800 gallons of water in the reactor tank with the level at about the top of the upper reflector region of the core. Since graphite has a specific heat on the same order of magnitude as that of aluminum, it will have a negligible effect on the rate of temperature change during temperature coefficien'. measurements. 3.4 Void Coefficient and Importance Function Measurements The void tube and absorber material specimens could conceivably be worth more than 0.8% of reactivity near the center of the smaller core. None of these materials, however, is subject to a catastrophic failure which could insert 0.8$ of reactivity in a positive step. 3.5 Irradiation fxperiments Reduction of the fuel inventory in the core from 37 to 24 fuel ele-ments would raise the peak thermal neutron flux at full power from about 10 2 8 x 10 n/cm /s to about 37/24 x 8 x 1010 = 1.2 x 10ll 2 n/cm /s, slightly greater than the 10' n/cm /s assumed in the Safety Analysis 2 v 3 0078C

Report (page 7-7). There is, however, no limit on the flux, only on the induced activity, and no problem is forseen in adnering to the limit, es-pecially since only small gold and indium foils are ever routinely irradiated, producing activities on the order of only microcuries. 3.6 Mechanical Damage and Failure The support structure of the core bears less total weight in that the gain in weight from displacing water by graphite is more than made up by the reduction in the numoer of fuel elements. This should more than make up for any displacement in loading caused by the asymetrical placement of the core in the reactor tank. The control rods weigh on the order of 80 pounds apiece. Any change in the deflection of the rod drive platform and support beams brought about by removing four control rods from the south side of the platform should be negligible. The graphite of the reflector rods is fairly strong, but soft. Care must be taken to avoid marring their surfaces. Under extremely rough hand-ling, pieces of graphite could conceivably be dislodged. At worst, these might lodge in the seats of the fast or fine dump valves, preventing their complete closure, which is a failure in the safe direction of moderator removal. i l Graphite tends to change its physical properties to a remarkable extend under neutron irradiation, becoming harder, more brittle and swollen in size. The onset of these changes, however, requires exposure to an integrated l9 neutron flux on the order of 10 nyt. Even if the thermal neutron flux at 10 the reflector rods were as high as 10 nv at full power, the safety iimit of 200 kilowatt hours per year would mean that it would still take ren the order of I9 10 4 - 1.4 x 10 years for these effects to manifest themselves, 10 200 10 x 7 x 3600 ) so that they are of no concern in the NTR. 4 0078C

p d) Fuel element insert adapters will be used to hold the reflector rods in locations 6-3 and 6-9. The reflector rods will weigh more than the fuel elements for which the adapters were designed, by a ratio of about 20 pounds to 14 pounds = 1.4. It is conceivable that the insert adapters could fail under this extra weight, particularly if a reflector rod were accidentally dropped into the adapter tube. Tne worst consequences would be for adapter and rod to break on impact, an event having no safety implications. A re-flector rod would be removed from its location, adding negative reactivity to the system, which is failure in the safe direction. 3.7 Radiation Conditions and Analysis The graphite reflector rods represent poorer gamma and neutron shields than the water they displace. Consequently, radiation levels at the side of the reactor tank may be expected to increase slightly, especially at the north side where the core itself is closer to the tank wall. Radiation levels outside the reactor tank are nominal in any event, and it is doubtful th&t any increase woula he measureable. O \\") Since the neutron absorption cross-section of graphite is so low (on the order of.003 barns for thermal neutrons), the reflector rods will not become activated and therefore will not represent a radioactive hazard in themselves nor contribute to the radiation hazard during a loss of shielding water accident. Because the critical water height will be higher in the 24 element core, the reactor will shut itself down sooner on a loss of shielding water accident. 3.8 Maximum Credible Accident Under the postulated conditions of the maximum credible accident, the j graphite reflector rods would be subject to a steam and water environment. Of concern under these conditions are the possibilities of oxidation of the graphite with consequent release of hydrogen and carbon monoxide gases and of annealing the graphite with the consequent release of Wigner energy as heat. l m 5 0078C .~,-, -,. ~ - - - ~, r_-_,

The temperature for the onset of the oxidation reaction i,s on the I order of 600*C. Since the hottest temperature in the reactor is estimated (SAR, page 7-17) to be about 360*C (slightly higher in the slightly smaller 24 element core), the graphite rods will not increase the consequences of the accident. The integrated neutron flux required for the buildup of Wigner energy I9 due to radiation damage is on the order of 10 nyt. As explained ir. sec-tion 3.6 of this analysis, integrated tiuxes of this magnitude can not cred-ibly be achieved. Therefore, the graphite rods will not increase the con-sequences of the accident. 4. Probability of an Unanalyzed Accident The change in core configuration and addition of graphite reflector rods does not admit the possibility of any unanalyzed accident occurring. Graphite is chemically inert and insoluble and would neither degrade the moderator-shield water quality nor react with the aluminum core structure. References on graphite as a reactor material are included in the appendix to this analysis. Purchasing documents must specify an adequate grade of graphite purity. 5. Technical Specification Margins for Safety The principle items pertaining to safety oiscussed in the Technical Speci-fication bases are listed in this section. Most require no cormlent since they would essentially be unaffected by the proposed change in core loading and reflection. Comments, where made, are underlined. The applicable section of the Technic.al Specifications are parenthized. 5.1 The average moderator temperature increase at 20 kw is less than 10*F. (2.1) 5.2 The radiation level irmiediately outside the reactor room at 20 kw is less than 20 mrem / hour. As discussed in section 3.7 of this analysis, any possible increase in the radiation levels would likely be so small as to be negligible. (2.1) 6 0078C

OV 5.3 The estimated error in absolute neutron flux measurement by acti-vation methods is 20 percent. The hot channel factor and fuel mass constants used in the flux determination will be different but will not effect the accuracy of the determination. (2.2) 5.4 The maximum error.in the nonlinearity of the neutron flux monitoring instruments is 35 percent. The characteristics of the instrument channels will be unchanged. The neutron flux at the detectors will increase for Channel A and probably decrease for Channels E and F, due to the Geometry of the core and detectors (Figure 1). Analysis of the core must provide this assurance before the proposed configuration is operated. The prediction must be verified early in the testing programs. (2.2) 5.5 The minimum flux level specified will prevent a source-out startup. (2.2) 5.6 A startup transient would be terminated in less than 200 milliseconds ("]) after a period trip. (2.2) Q, 5.7 The gamma level channels will assure that increasing radiation levels will be detected before they become excessive when the reactor is operated at moderator-shield water heights other than the normal level. (2.2) 5.8 There is adequate shutdown capability even for the stuck control rod condition. Analysis of the core must provide this assurance before the pro-posed configuration is operated. The prediction must be verified early in the evaluation program. (3.1.1) 5.9 The maximum excess reactivity for the reactor is 155. The same comments apply as for 5.8, above, although it may be noted that the excess reactivity available in the N-37-5 core is considerably less than 155, and will be still less in the 24 element reflected core. (3.1 Bases, Paragraph 2) 5.10 The maximum control rod withdrawal and moderator-shield water reac-tivity addition rates, far from and near criticality, assure that the reat-3 tivity addition rate is less than that analyzed in the maximum credible acci-dent. The same comments apply as for 5.9, above. (3.1.2) 7 0078C t

5.11 The control rod insertior, time from fully withdrawn assures that the assumed time for establishing the minimum period LSSS is satisfied. (3.1.3) 5.12 Cnntrol rod withdrawal prior to adding moderator-shield water as-sures that reactor trip will have the capability of adding negative reactivity during reactor startup. (3.1.4) 5.13 The auxiliary reactor trip assures tha' 'here is a secondary mode of shutdown available during reactor operations The minimum time for it to add negative reactivity limits the consequences of a potential power transient. Negative reactivity will be added sooner in the 24 element reflected core because of its higher minimum critical water level. (3.1.5) 5.14 The normal moderator-shield water level assures an adequate shield during maximum power operation. At lower moderator levels, reduced trip settings are required to further reduce the possibility of operating with a high neutron and gamma radiation field. Controlling the reactor by moderator level near criticality only after ther reactor is first made critical by control rod movement assures that the control rod is the primary mode of reactivity control in the critical reactor. (3.1.6) 5.15 The manual reactor trip assures that the trip may be activated readily by either the operator or individuals in the reactor room. (3.1.7) 5.16 Electrical power interruption provides an additional mode to manually trip the reactor. (3.1.8) 5.17 The minimum safety system channels provide a high degree of redun-dancy to assure that human or mechanical failures will not endanger the reac-tor facility or the general public. (3.1.9) 5.18 The interlock system assures that only authorized personnel can ~ operate the reactor, that the proper sequence of operations is performed, that no one can accidentally enter the reactor room and that the reactor room is entered witn proper conditions prevailing when the master key is on. (3.1.10) O 8 0078C

O) 5.19 The minimum absolute value of the temperature coefficient of reac-t1vity assures that an adequate inherent negative reactivity effect takes place when the reactor temperature increases above the salue where the coef-ficient become negative. (3.2.1) 5.20 The minimum absolute value of the void coefficient of reactivity assures that the negative reactivity insertion due to void formation is greater than that which was calculated to occur in the SAR. (3.2.2) 5.21 The moderator-shield water quality assures adequate corrosion con-trol in the reactor environment. Graphite will not reduce the water quality l as discussed in section 4 of this analysis. (3.2.3) 5.22 Area radiation monitors assure warning of the existence of any abnormally high radiation levels. The availability of instruments to measure air and water activity assists in monitoring fuel clad integrity and assures continued compliance with the requirements of 10CFR20. The availability of / \\ I t portable monitors provides assurance that personnel will be able to monitor C,/ potential radiation fields before an area is entered. (3.3) 5.23 To assure that experiments are well olanned and evaluated prior to being performed, detailed written procedures or all new experiments must be prepared, reviewed by RSC and approved by the Facility Manager. This requirement will be met for the 24 element graphite reflected core. (4.1) 5.24 Since the control rods enter the core + favity and are required by other Technical Specifications to be operable, no experiment should be allowed to interfere with their functions. To assure that specified power limits are not exceeded, the nuclear instrumentation muit be capable of accurately moni-toring core parameters. The 24 element graphite reflected core will alter the flux seen by the nuclear instrument detectors but will in no way interfere with their operation or function. See section 5.4 of this report. (4.2) d 9 0078C

5.25 A maximum rear.tivity change is established for the remote posi-tioning of experimental samples and devices during reactor operations to assure that the reactor controls are readily capable of controlling the reactor. The e.periment will not include remote positioning of experimental samples and devices _. (4.3) 5.26 All experimental apparatus placed in the reactor must be properly fabricated and made physically secure in the reactor. In consideration of potential accidents, the reactivity effect must be limited to the maximum accidental step reactivity insertion analyzed in the SAR. In actual practice, no single unit of experimental apparatus will be placed in the reactor which has a reactivity worth greater than 0.803 which is less than the reactivity addition accident analyzed in the SAR. Refer to section 3.4 and 3.6 of this analysis. (4.4) 5.27 Restrictions on irradiations of explosive::; and highly flammable materials are imposed to minimize the possibility of explosions or fires in the vicinity of the reactor. To minimize the possibility of exposing facility personnel or the public to radioactive materials, no experiments will be performed with materials that could result in a violent chemical reaction and/or produce airborne radioactivity. Refer to sections 3.1 and 3.5 of this analysis. (4.5) 6. Conclusion Since it has been established that (i) the probability or the consequences of any analyzed accident are not increased, (ii) there is no possibility of an unanalyzed accident occurring, and (iii) no margin for safety in any Technical Specification basis is 'duced, it is very unlikely that there are any unreviewed safety questions involved in the operation of the 24 element graphite reflected core. Consequently, there is no need for NRC approval prior to operation and no need to amend the NTR Faci"ty operating license. 10 nn7ar

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REFERENCES ON GRAPHITE AS A REACTOR MATERIAL

5. Glosstone anc A. Sesonske, "lUclear Reactor Engineering", von t.ostrand Reinhold, tiea York, 1967, pages 438-442 H. Etherington (ed.), "haclear Engineering Handbook" McCraa-Hill, 1958 (First Edition) pages 12-/0 to 12-74, 10-55 to 10-59, 10-109 to 10-115, 13-180 C. 4. Tioton, Jr. (ed), "Reactcr H3ncacok Volume I - Materials", Interscienca, Ne a-York, 1960 (2nd Edition)

Cnapter 43, Graonite, by L. D. Locn, pages 888-896 Chapter 53, Bulk Shielding Data by R. D. Schamber, et. al., pages 1l19 and 1123 Cnapter 41, Review of Mcceratcr hterials by E. M. Simons, pages 835-337 Proceedings of the International Conference on Peaceful Uses of Atomic Energy, United Nations, N.Y., 1956 Voi ce 7, "teclear Cnemistry and tne Effects of Irradiation" Irradiation Damage to Artificial Grapnite by Wooas, et. al., pages 455-471 _The Effects of Irradiation on Gracnite by G. H. Kinchin, pages 472-478 Volume 8, " Production Technology of the Materials Used for Nuclear Energy" The Production and Properties of Graonite for Reactors by Currie, et. al., pa9es 451-473 T. J. Tnompson and J. G. Beckerly (ecs.), "Tne Tecnnology of Nuclear Reactor Safety", The M.I.T. Press, Cambridge, Mass., 1964 blume 1, " Reactor Pnysics anc Control", pages 633-636 Volume 2, " Reactor Materials and Engineering", pages 436, 464, 469, 45 A. R. Kaufman (ed.), "taclear Reactor Fuel Elements Metallurgy and Fabrication", Interscience, New York, 1962, page 251 (NOTE: The following sources aere not consulted in the preparation of the NTR safety anelysis, but are listed for completeness.) J. H. W. Simmons, " Radiation Damage in Graphite", Pergamon Press, New York, 1965 W. G. O'Driscoll and J. C. Bell, Graphite: Its Properties and Behavior, Nuclear Engineering 3: 479-489 (tiov. '56) ana 533-537 (1938) R. F. Nightingale (ed.), " Nuclear Graphite", Academic Press,1962 0 0078C

... _ _. = _ _ _. . _ = _ O i i l I i REFERDICE C O 1 l I I I I I I l I e 1 O

a \\ g PART 50 e DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION FAclLITIES 1 (illnsenice eaaminations of 15 Mil The inservice inspection program foi components. insenice tests to verify a hodms or pressurbed usintooled nuclear operational readiness of pumps and Powere facilny shsH be rmsai by the licensee as necessary, to meet the nah es whose function le required for safety and s) stem pressure testa. 3hu((ments f paragraph (g>(47 of this r cond acted during the Initial 12&rconth ggg, If a revised inservice inspection e: Inspection inten al shall comply with the program for a facility conf!! cts with the }l requirements in the latesi ed tion..nd technical specincadon for the facihty, addendalf the Code incorporated by the licensee ahtll ap9ly to the Commis-slon for amendment of the technical referencein parapaph(blof this section spec 1Scations to con!crm the technical on the date 12 months prior to the date ofissuance of the operating hcense. 8P'C[atio C 8 ,pp a tted a I subject to tbe hmitahone and 6 months before the start of the period moifications hated in paragraph (b) of during shich the provisions become sp-this sectics phenble as determined by paragraph <g-(ii)Insenice easminations of i4' of this section. t111) If the !!censee has determined corcponents. insenice lesta ?o s erify that conformanae sith certain code re-operational resiness of pumps and quirements is impractical for his !ac!hty. sahes whose functionis required for the licensee shall nc!!!y the Commissicn safety and system p essure testa. and submit informauon to support his conducted during successive 120 month determinations. Inspectionintervals shall comply with 'Ivi Where an examination or test the rergirements of the latest cition requirement b3 the code or addenda is and addenda of the Code inco po ated determmed to be impractical by the 11-b) reference in parupeph (b) of this censee and is not included in the revised seetion 12 months pner to the stmrt of insenice inspection program as permit-th e 120 month i:psection inte-s al, ted by paragraph (gi t4, of this secticn. sdject to the hmitations end the basis for this determination ahan be demonstrated to the satisfaction of c?d hea tions hsted in Iaragraph (b) of the Commissien not later than 12 months Ib8 'd af ter the expiration of the initial 120-(in)For a f Ws A r op r..t7 month perrt of operation from start cf 7s 3 / \\ hrr se mas i+ rut d r. r to M mh I facHity commercial operation and each (d., ; of thie section are t.'L cme sfier..ap ; h !;"4] subsequent 120 menth period of opera-3G. the prou.or.. tion during shich the examination or a Sepamber 1.19 6. at the start of the test is determmed to be impractical. f 6 als The Comndssion sin evaluate nett-one third of a 120 month irspection ldetermmatiens under paragraph agi(5s intenal. During that third cf an g of this section that code requirements inspection inten c.l and the remindct o, m are impracts:st The Commission rna) grant the inspection inten el, the InSenice

such rebene and ma) trnpose such alternatne enaminations of components. tests to requuemer.ttt as et determines is authortred s erif) operational readmess of pumps t$ las and s ah es whose function is required and will not endanger life or property for safety, and s) stem pretsure tests, for or the common defense and security and such facihties shall compl) with the is otheru-1se in the public interest giving due consideration to the burden upon requirernents in the latent edition and the licensee that could result if the addenda of the Code incorporated by reference in parepsph (b)of this section

'.ments were imposed on the on the date 12 months prior to the start till The Commission may require the of that third of an Inspection inten al. heensee to follow an augmented inservice subject to the limitations and inspectbn program for systems and com-rno6fications Lsted in paragraph (b) of ponents for which the Commission deems this section. that added assurance of structural rell-(tv)Iriservice eltaminations of ab!nty is necessary. ,Poe purposes or this reculatice the pro-tion e rn i e a te a ary 91 afin[anY'[e' ib"5 components, tests of pumps and s alves* ta and system pressure tests.may meet the quirements set forth in editions or revi. "h copies snar be obtained from the Instasute protection systems ahan meet the re-sn ist! became ~ta eher on June 3. len ( requirements sel fo*th in subsequerst edations and addenda that are sions of the Institute of Electrical and of E3*ctrical and siectronics snatn ers. g teria for Protection Systems for Nuclear p,,$n3;,,,,,,,,,,gg,,,,g3,e,,,,,{,T,,,p,,,_ so a e r a saat 4 th Electronics Engineers Standard: "Crl-a u,stn Inmirporated by reference la pararaph y (b).Of this section, subjact to the l lirrdlations and mo1Ccations listed in Pos er ynerating Stations. 1IgEE-279) l Ise parument sten. 1717 se str,et N w., m eftut on the formal docket date of the {wantungscatp.c. par'agraph (b) of this section, and

• II"#'" " ' * # ""'"#" " I" *"

subject to Commission aIIroyal tecteum >> stems may meet the fawhere an sprinetton for a construction Portions of eations or addenda may be requirements set forth in subsequent edt-e. perma a suhmuted m four pets rursuem to used provided that a11related ltions or revisions of IEEE-279 which l the provisions of t 2.sonta a p and Sut rari i requirements of the respective edations become effective. c. of Part 2 of this charter."the formal daart (m at dai$,of the arch ation for sonstrussion r* or addenda are met h snit for the retroses of it+ ses sion shah be y the date of dodefins of the eritornuien te s 1 Amended 41 FR 2s931-querd to t 2.losta 3 )(21 or s 4 m hnheter BE later. 'AmendeJ 42 I R 22ss2. August 1.1980 l

o PART 50 e DOMESTIC LICENSING OF PRODUCTION AND UTIUZATION FACILITIES tit Tracture toughness requ!Nments: 6 50.~>7 Inuance of aperating licen n. [ paragraph tai of this section If no party Pre'sure-reta:rJng ec ponents of the re*

a. cpposes the motion, the pres!dmg ct":cer actcr cW. ant pressure boundary shall fal Pursuant to i 50 56 an opetatmg 2 u111 issue an order pursuant to a 2730.e.

meet the requ:rements set forth an Ap-heense may be issued by the Ccmmission. $ of this chapter, authartring the Durctor of %. , pendnes G and H to this part. up to the fu!! term authcrtzad by 150 51. g dear Reactor Regulation to make appropnate aji Pcser reactors far which a notice upon find.ng that:

• fmd-h of her.rg on an applicaticn for a pro-

'Is Construet;on of the facthty has ;; ings on the rnatters spec'ned in pars-un ni construction permat or a con. teen substant. ally comp;eted. In con-graph (a) of this secti:n an'. to issue s a strr : on f ern.tt has been publ.shed on forrruty w ith the const1uct:en perm:t hcense for the requested opersQon. g and the appheation as amended. the adl IDeleted 40 FR 8774-) -f g or bef ore Dwember *.L 1970. may meet

Iov15.cns of the Act. and the ru

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' he ren.rements of paragtsphs (c)(1), a'l' 'e'd18 and if1815 of this sec. regulat: ens of t;1e Comm.ssion: and 5 so.>a 11earina. and repoei of the a ..on :rJtead of raragraphs d e' 8 2). 'd) '2' The facahty will operate in con. si.ory Cumn ittee on Reactor Sfe-2- +e> 2.and 'f2s of this sect;on, forrnity uith the appheat:cn as ar' ended guards. respecta cly the pronsions of the Act. and the ru;es fat Each apphcation for a construc-and reeu:ations of the Ccmmissten. and t.on permit or an opetating license for 50 55b ' R n. Ad 4 3 i R u,,! ! '3' There is reasontt:e assurance 't) a facthty ahtch is of a type described in that the actn;ttes author: Zed by the op-3 4 50 21'bi or 150 22. or for a testing f a-

erat

nr heense can be cond.:cted w:thcut 6 cthty. sta:1 te referred to the Aduscay 7 ;endan;;etmg the health and safety of the

- Committee en Reactor Safeguards for 2hc and a u' that such acmities v 1. E a renen and report. An apphcation for

oe conducted an com;L

ance utth the

• an ar".endment to such a const uction and R ;elmtt or o terating 1"snse may be re-g; re:J:at: ens in this chapter:

The appheant is technically and ( ferred to t *e Adusory Comrnattee on .'4

• na nc:ahy qual.* led, to eneage in the Reactor Sa.eguards for reuew and te-ai 6t:es a ut hot i... by the c; erat.ng

crt. Any report shall be made part of

. cense in ace:rdance uith the ren;a-the record of the apphcation and avail- !:ene in th:s chapter and , able to the pubhc. except to the extent 'Si The a;phcable pro :s.ces of Patt l 140 of this chapter ha*.e been satisfied. { tPat security classancation presents dis-

• c k sure.

and ~ ebi The Comrr/ssion mil hol! a hear-

6. The issuance cf the hecnse udl rot be anim; cal to the co nm:n defense gr7 af ter at : east 30 Ays' not;ee and pub-I e.catt:r. cnce.n the

.'t rant Rrctstra en ar.d security or to the heaMh and sa!ety 1 ach a;7.tcan:n f:t a canst:uct:cn per-of the ;.bbc. tb' Each creratmg heetise u d! :n-m.t !;r a ;:cdacten cr utthzat;cn facd-clude appropriate prousions u tth respect .ty sh.ch is c! a type d" cute 1 in ! 50 - to any uncompleted terns of constructicn 1.bi cr 150 00 er uluch is a teat-ng fa-and such hm:tattens or condit: ens as are c;hty. When a c:nsta act:on perm;t has ren; red to assure that cperation durid been issud for such a f aen.ty fci ctrg the p"tod of tre com;;eticn of such the hc!d:ng of a pub;!c hearing and an items u til not endanger pubhc health and ' apphcat;cn is made for an operatmg h- ,; cense or f0r an amtndment to a cen- , safety. .ci An appheant may in a case w be.e 3 struction pe:mit or et crating heense. 6he cc C;mm.ss.on may hcid a h: arms atter at a hearing is held in connection utth a = Icast 30 days not:ce and pubucatacn once rendmg proceeding under inis sectacn. gp.he Ft:grat R[ctsita or, in the ab-inake a motten in srtimg. pur=uant to p sence cf a request thetefor by any ;er-th;s paragtaph 8c'. fcr an cpelating 11-s;n s h0$e interest may be a*!tcted. may cense at. thor; ing low-pow er testmg acp-Issue an cpelattn; hcense er an amend-eratien at not more than I percent of full r-ent to a ccnstructM permit er cpera-power for the purpose of testing the fa-ting 1: cense without a hea-ing, upon U cahty ), and further operations short of da>s' nottce and pabhcation cace in the . full poner operaticn. Action on such a FEt.Laat R cts En of its intertt to do so

• /, meticn by the presiding ef".cer shall be If the C mm:ssion finds that no si:n.n-

taken with due regard to the rights of cant hazards cons;deration is presented

- the parties to the proceeding. including ty an a;; heat:cn for an amendment to 2 tre r:ght of any party to be heard to a c;nst:uct;cn permit or ope:stmg 11- , the ettent that his contentions are rele". cen.se. it may d.spense u tth such not:ce m sant to the activity to be authorized and pub:: cat;On and may issue the P: tor to taking any action on such a r).otion slach any party cpposes, the pre- -amendment, s.cmg o?.cer shall make nndings on the l 50.59 Clunsee, tests and esperiments. - 15056 Co n t e*sio n of co't s t ruc tic a r..atters spectned in paragraph ta) of ran ili The holder of a license author-peraint to facestie, ne amendment o.r h-this section as to shtch there is a con-tzing operation of a production or utt-rense l' pen completion of the con-trosersy, in the form of an inattal de-luatten f acthty may all rnake changes in struction or alteration of a facahty, in c: won with respect to the contested the facahty as described in the safety comphance with the terms and conch-activity sought to be authorized The analym report. (1D make changes in the tions of the construction perrnst and Duettor of % lear Reactor Regulation wilf

• procedures as described in the safety wb.ect to any necessary testing of the snake findings on all other matters specified in analysts report and (tip. conduct tests f actht) for health or saftty purposes the

- or experiments not descritg1 in the Commission mill. in the-absence of sc~J z safety analysts report. withotit prior cause shcun to the cohtrary issue a h-

• Commission approval, un.less the pro-l cense of the etass forNhich the con-R posed change. test or experiment in-sches a change in the technical specia-struction permit was assued or an appro* I ' n.e commission may issue a prousiorist priate amendment of the beense. as the
• rperaung ir ense pursuant to the ce.deuona cations incorporated in the license or an

} an)trus part in e*ect on Maren so.197o. for A proposed change. test or esperi-in unrevtemed safety question. ,,.f ase may be ran.t.y f or u htch a r;.t.ce of tear.n' (2p g, on an arpitcanon for a prountensi cperaurg rnent shall te d:emed to invo an un-mense or a nottte of preposed tm.ance r reglemed safety quest!on eli if...e prod-fM a grous.orat cperating n ense has been p 2 bnes on or t>erare test date atinty of occurrence or the consequences l l 50 22 l August 1,1980

sv} PART 50 e DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION FACILITIES of an accident er* malfunction of equip- .. Inspections. Records Reports. connection sith the b ensed actiuty, as ment.tmpertant to safety previously evaluated in the safety analysis report Netifications may be required b) the concittens cf the hcense or permit cr ty the rules reguia-may be increased, or itD if a possibihty [ 30.70 Inspections. E t.ons. and crders c! the Ccmmiss:.m in for an accident or malfunction tt a dif-(a)Each 11:ensee and each holder of a E eiectuating the purpcses cf the Ac=.. in-ferent type than any evaluated previ-constr !!cn permit shn:] permit inspec-g c uding sect)cn 105 of the Act. ously in he safety analysis report may tion, by du'y author 12'd representatives

• Ib) With yrspect to any product!:n er be created. or fitti if the margin of n of the Conim:ss!cn. of his reccrds.E utih, stion f&cihty of a type described 1::

safely as de*.ned in the basis for any techmes*speedcation ts reduced. ' prersises, a: 1vities, and of heensed ma. ,50 21 tb) cf I 5C. cr a testmg f a:1;:ty. E teraa:s in possessicn or u.se, related to each licenste a-d ea:h h:!dcr Of a ca 3 (bi%e 1;:ensee shall maintain *ec-the IJ:ense or construction per=1t as may St"UCtiCS per:1! shC UPCC C'cd 1s-g ords of char.ges in the faciDty and of be necessary to efectuate the pu! poses suance of its annuc: E cnciai rep: t. in- . changes in prcaedures made pursuant t cf thc act. Including section 105 cf the clading the c:i t::.c d fb.ancial state - g this section, to the extent that such S changes constitute changes in the facil-S g- =ents. fUe a ccpy therecf *.ith tr.e K lty as descnt,ed in the safety analysas Director of Nuclear Reactor Regulation. t'.5. report or ccnnatute changes in proce-Nuclear Regulator) Commusion.Madungton. dures as descr.t4d in the safety analysis '*-{ bjlM Dch hancese and esch holder ,,,D,,.C. 205 5 5 3 report. The license" shall also maintalla of d 4xmstructacrn permit ahtiltipon - records of tem anc experiments carried regaest by the Director.OfLce of (c) Records which are requ: red by the out pursuant to paragraph (a) of this Indrection and Enforcement. provW regulations in this part. by license con-section. These records shall include a reot.4.ree office space for the exclustve dation, cr by te:hnica; spe:2 cation shall sritten safety etaluaticn which pra-use of the Commission inspection be ma:ntamed for tne pened spec 1*ed by vides the base.s for the determmation that the char.ge. test or expenment dc+s Personnel. Hes( air conditioni38 hght. the appropriate regu'at::n. heense con-not invohe an unreviewed safety ques-electrical outlets and janitonal services diticn. or technica; spec;ncaisen If a p J1on. shall be furnished by each heensee and fled. such re:ords sha:: be mamtamed The licersee shal! furnish each holder of a con?truction permit. until the Comm:ssien autherizes *h; - to the ap;rc;r. ate NRC Regici:al Of5cc The cffice chall be convenient to and disposition. shown m Append:x D of Part 2) of this have full access to the fedty and shall (dP(1) Records which rnust be main. , { chapter v:th a copy to the Director of proude the inapector both visual and tamed pursuant to th:s part may be the aInspection and Enforcement. U S Nu' scoustic prisacy. g ongmal or a reprodu ed ccpy or micro-L{ Commassen % ashmg-(2) For a site with a single power . form if such r:produ:ed copy or m1:ro- " clear Re a

• ton. D C am ea sw reactor or fuel facihty beensed pursuant - form is duly autnent:cated by author: red shorter mie-E3 as mas e spe:,,ed n

' the brente a repert ecntamms a brief to Part 50. the gace prouded ehall be E personnel and the m :rcic: 1 is capab;e " M producmg a tiecr and tgible copy (% descn-t2cn c! such chang (. tests and adequate to accommodate a full-time after storage for the pened spe:ified bk h. expenments -~g a summal.s of the inspector, a part-tune secretary and Commission rega:at:cns

• (/)

jafety e.aLat.cn of ea:h transient NRC personnel and will be (2e If there is a confhe between the p Any report generally commensurate with other Commission's regulat:cr in this part . submitted by a licensee pursuant to this office facihties at the site. A space of h:ense condition. cr tc.cnical sM::Sca-3 paragraph sE be made a part of the 250 square feet either within the site's tien, or other written Commissten sp-g pubbe record of the beensing proceeding. office complea or in an office trailer or prosal or authcrizat:cn pertamms to In addition to a signed original. 39 copies a of each report of changes in a facihty 'I other on site space is suggested as a the retention perled !cr the same type " of the type desertbed 10 150 21tb> or. gurde. For sites containmg multipIe cf re:ord. the retenti:n pened spec;*.ed g g g R I 50 22 or a testing f actbt r, end 12 copics. Power reactor umts or fuel facrhties. records shall app:y un:ess the Comm:s-l of each report cf changss in any other., additaonal space may be requested to sion, pursuant to i 5012. has granted a Macihty.sha!! be filed.

• accommodatt edditional full time speciS: exemption frc= the record re-Tne records of changes inspector (s) The office space that la tentien r. autre =ents speciSed m the k"L 'he facihty oha!! te maintained untilProvided shall be subject to the regulations m this part.

Ithe date of termination of the heense, approsal of the Director. Office of E and records of changes in procedures and Inspection and Enforcement. All records of tests and expenments shall be furmrure, supplies and communication (e) Each person licensed to operate a e t will be furnished by the nuclear power reactor pursuant to the u (mamtained fer a penod c'.ive years. provisions of I $0.21 or j $0 22 shall F tct "Itie holder of a heense authortz-(3) The bcensee or cons:ruction persait update periodical!. as prrivided m 3 ing operation of a production or util1+ holder shall afford any NRC resident Paragraphs (e)(3) and (e){4) of this l tion facihty sbo desires (1) a change inspector assigned to that site. or other section. the fmal safety analysis report a change in the facibty or the procedures NRC inspectors identified by the (FSAR) origmally submitted as part of in technical spectScations or (2) to make the apphcation for the operating hcense. h described m the safety analysis report Regional Director as hkely to mopect the 3 to assure that the information included g or to conduct tests or expenments not facihty. immediate unfettered access.

in the FSAR contains the latest material a described in the safety analysis report, equivalect to access provide 4 regular a

• developed This submittal shall contam R shich involve an uru evkaed safety question or a change in technical spec.

p; ant amPg, tees, g,gjowmg proper all the changes necessary to reflect ifications, shan subtry an appbcation identification and compliance with information and ana!)ses submitted to rot amendment, of has license pursuant applicable access control mesures for the Commission by the beensee or _tol 60 90. accurity. radiologica1 protection and prepared byihe hcensee pursuant to $ 50.60 [ Deleted 40 FR 87741 -Personal safety. Commissun requirement smce the submission of the original FSAR or. as t 50.65 (Deleted 43 FR 6915.] appropriate. the last updated FSAR. The IsA71 teatntenance of rwores, maadne e.' updated FSAR shall be revised to n report s-include the effects of. all changes made I { tas Each licensee and each holder of in the facihty or procedures as y,/ a cons *ruction permit shall maintain desenbed m the FSAR. all safety su:h records and make such reports. In evaluations performed b) the bcensee % mended 42 i R 2ol3s. e AmenceG es F R3 34 34 T Amended 43 FR 8644s. 50 23 August 1,1960 1

r 3 + m PART 50 e DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION FACILITIES ener m support of regoested hcense . area changed. e g a bold hne sertically (b) With respect to the esents amendmew cr.n support of drawn m the margm adjacent to the reported under subparagraphs (1). (2). I anis 1ons ' hat charges did not msolve s partion actually changed. and a page (3). and (4) of paragraph (a) of this an ;.nreuewed safe'y question, and all d change identification (date of charge or, section each hcensee. in addition to ad ses of rew safety issues performed hhange number or both)

prompt telephone notification. shall also 4 or un beha

f of the bcerasee at estabbsh and maintain an open.

Camm;ss.cn request The updated hM2 Not$cadon of Wgn 0 cent aventa, contmu0us Communication Channel with ,farr ation sham be apprenately (a) Each hcensee of a nuclear power

.he NRC Operations Center. and shAl

aca'ed wi'h:n the FSAR. -

rea tor hcensed under 150 21 or i 50 22

• close this channel only when notifiedtby

,1. Res is.ons containing updated sha!! notify the NRC Operations Center lNRC. M "c-mat:on snau be submitted on a as soon as possible and in all cases ] wpMcement paae basis and shall be within one hour by telephone of the campanied by a hst which identifies occurrence of any of the followmg 3 the carrent pages of the FSAR foliowing significant esents and shallidentify that lpage replacement One signed ongmal esent as being reported pursuant to this ar.d 12 addit.onal cep.es of the required section: , mformanon shan be fded with the (1) Any esent requinng initiation of Director of Nuclear Reactor Regulation. the hcenseis emergency plan or any U S Nuclear Regulatory Commission. section of that plan. \\\\ ashington. D C 20555. (2) The esceed:ng of any Technical a The submittal shall mclude (i) a Specification Safety Lim:t. an ficahon by a duly authonzed officer (3) Any event that results in the vf the hcensee that either the nuclear power plant not bems m a .nfermanon arurately presents changes controlled or espected condition while made since the ::revious submittal. operatmg er sh.t down. "cessary to refiect mformation and (4) Any act that threatens the safety of anais ses subm tied to the Commission the nuclear power plant or site or prepared pursuant to Comm:ssion person,el, or the secunty of special

requ

rement or hat no such changes nuclear material. includmg instances of

were made. and in) an identification of sabe' age or attempted sabotage.

I changes made under the prousions of (5) Any event requinng mit:ation of

i 50 59 but not previously submitted to shutdown of the nuclear power plant m

, the Commission accordance with Technical Specification (3)D) A reusion of the ongmal FSAR Limiting Conditions for Operanon. i contammg those ongmal pages that are (6) Personnel error or procedural soll apphcable plus new replacement

inadequacy which. dunng normal pages shall be filed within 24 months of I cperattens. anticipated operational either July 22,1980. or the date of

occurrences. or accident coaditions.

' ssuance of the operatmg bcense.

presents or could prevent, by itself. the i

wh:cheser is later. and shall bnng the O fulfillment of the safet) function of those FSAR up to da'e as of a masimum of 6 structures. systems, and components months pnor to the date of fihng :he important to safety that are needed to b) revision shut down the reactor safely and (n) Not less than 15 days before mamtain it in a safe shutdown i 50 n(e) becomes effective. the condition, or (u) remose residual heat Director of the Office of Nuclear Reactor fo!!ow ng reactor shutdown. or (ni) hmit Regulabon shall notify by letter the the release of radioactive matenal to hcensees of those nuclear power plants acceptable levels or reduce the potential mitially subject to the NRC s systematic for such release. esaluallon program that they need not (7) Any event resultmg m manual or rwp!) with the prousions of this automatic actuation of Engmeered section wh 'e the program is being Safety Features. including the Reactor conducted at their plant. The Director of Protection System. the Office of Nuclear Reactor Regulation (8) Any accidental. unplanned. or wd1 not;fy by lener the hcensee of eac's uncontrolled radioactive re! case. nuc! ear power plant bemg evaluated (Normal or espected releases from when the systematic evaluation program has been completed Within 24 months rna ntenance or other or erational activities are not included ) after receipt of this notification. the hcensee shall file a complete FSAR (9) Any fatality or serious injury which is up to date as of a masimum of occurnng on the site and requinng 6 months prior to the date of fihng the transport to en offsite medical facility ~ for treatment. (4) Subsequent revisiorg'shall be filed l co(ntamination requinns extensive

10) Any serious personnel radioactive no less frequently than annually and onsite decontammation or outside shall reflect all changes up to a a ssistance.

rrasimum of 6 months pnor to the date (11) Any eserd meetmg the enteria of to CFR 20 403 for notification. (5) Each replacement page shall include both a change indicator for the (12) Stnkes of operstmg employees or secunty guards, or honormg of picket ' lines by these employees. August 1,1980 50-24 ..}}