ML20086E729

Proposed Tech Spec 3/4.4.4 Re Relief Valves & 3/4.4.8 Re Pressure/Temp Limits

ML20086E729
Person / Time
Site:	Comanche Peak
Issue date:	11/27/1991
From:	TEXAS UTILITIES ELECTRIC CO. (TU ELECTRIC)
To:
Shared Package
ML20086E728	List: ML20086E729 TXX-9142, License Amend Request 91-010 to License NPF-87,changing TS to Incorporate Provisions of Generic Ltr 90-06, Resolution of Generic Issue 70, 'Porv & Block Valve Reliability' & Generic Issue 94, 'Addl Low Temp Overpressure....'
References
NUDOCS 9112020283
Download: ML20086E729 (11)

Text

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'NOU 21 '91 9:01 past,ega i

REAO'OR : Lab' Id'E" 3/4 4 a RELIEr VALVES Liv!?'NG 00NDI?!ON FCA CPERA?!ON ,

bc0

(

3.4.4

.ai.e5[sbat'// te :DERABLE,  ::.er :ersted relie' <alves (DORVs) are tSeir asso a:DLIA8!LI M MOCE5 1. 2 ane 3.

w% pew rrentamedtoihe.

AC'! 3: dock. YdVE.(M BOO

a. With one cc //// DORV/s/ inoperatie, tecause of excessive seat leau-age, witnin 1 nove either restore the PORV(s) to OPERABLE status or close tre associateo clock valve (s), otherwise, ce in at least W 5TAh0BY witnin tre next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHJT00WN witnsn t*e 'ol-lowing 6 nours.

R t. With one 20RV imoceratie due to causes other than excessive seat e, ithin 1 mour estner restore the PORV te CPERABLE stat'.5 Or g.b 'eakag/// associated block valve and remove power from tre ol:cs

iose'

<alve- restcre tre PCRV to ODERABLE status within the 'oli:wieg

, 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> er ce in HDT STANDBY *ittin tre next 6 tours ans in 07 58UTDOWN witntn the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

ok\taskowe c.

With toth PC.tV/s/ inocerable cue to causes s er than excessive sest gg leakate. wi-nin_1 rour either restore M/ [/ /// PORV // to

.PERAELE status or close%///// associated D .;k valve // and e* '.e c er from tr.e ol:ck valvef// arc ce in H3T ~ANCBY witnin tre tat i Pours and 19 n0T SHUTDOWN .itnin ne folicairg 6 h:ves, hxsth

c. aitn ene or //// elock valve /s/ inoperatie, .'tnin 1 nour // rest: e tae 31o_c3 valve (s) to OPERABLE status p / // ff( / //// ///

/9/ $9 / ll l

e. The provisions of Specification 3.0.4 are not acclicaole.

SURVEILLANCE REQUIREMENTS

4. 4 ** 1 In addition to the requirements of Spe:ification 4.0.5, eac- c Rv snail :e demonstrated OPERABLE at least once per 18 montas Oy:

C;erating t*e val.e .Prough one complete cycle o' full travel 1c a.

0. Performing'a CHANNEL CALIBRATION of the actuation instrueentat'Or place its associated PORV(s) in manual control. Restore at least one block valve to OtutAtti status within the next hour if both block valves are inoperable; restore any remaining inoperable block valve to operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />; othewise he in at least NOT STAfG8Y within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in ICT 5161T96l01 within_tha.followino-6 heum_ -

COMANCHE : ear ,.LNIT 1 3/4 4-11 9112O20283 911127 PDR f4 DOCK 05000445 P PDR

NOV 21 '91 9:03 PA3E.015

&Ei:*:t #:ct W 5<3 EM REL:!' VALVES 5utvi!LLANCE REOUIREMENTS 4.4.4.2 !ach block salve sh$11 be cem:nstratec OPERABLE at 'esst 0 ce ce-92 cays by ;ceratin; :ne vahe thrcu;n one cociete cycle of full travei unless the ticck valve is closec in orcer to fteet tPe requirements of ACT!ON/

g cr / in Scecification 3.4.4 c

i 1

CCMANCHE DEAK - UNIT 1 3/ 4-1;

NOV 21 '91 9:03 PA3E,016 l

REACTOR C00 TANT 5'5 TEM 3/4.4.8 PRESSURE? TEMPERATURE .!MIT3 REACTOR COOLANT SYSTEM LIMITING CONDITION FCR CPERAT!gN 3.4.S.1 The React:r Coo'. ant System (except tae pressuri:e") temperature anc pressure shall ce limited in a:carcance aith the limit lines snown en rigures 3.4-2 and 3.4-3 curing heatus, coolco n, criticality, and inservice lea ( and hydrostatic testing with;

a. A maximum heatup of 100'F in any 1-hour period,

b. A maxirrum cooldown of 100'F in any 1-hour period, and

c. A maximum temperat;re change of less than or equal to 10*F in any 1-nour period duch'; Mservice nycrostatic anc leak testing c;erations acove the neatup ano cocicown limit curves.

APPLICAEILITY: At all times.

ACTION:

With any of the abcVe litnits exceeded, estore the ter eeature and/or ; essure to within the limit within 30 minutes; perform a, eng eering evaluatics to oetermine the effects of t9e out-cf-limit concition o the structural i .tegrity of the Reactor Ccclant Systet.; 1etermine that the Reac.or Coolant System remains acceptable for centinued cceration or De in at least -:' STANOBY within t e next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and recuce tne RCS T,,g and pressure to ass than 200*F anc 500 psig, respective'y, within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURNEILLANCE REOLIREMENT!

4.4.8.1.1 The Reactor Coolant System temperature and pressure snail be caternincd to be within the limits at least once per 30 minutes during system heatup, cooldown, and inservice leak and hydrostatic testing operat ens. i 4.4.8.1.2 The reactor vessel traterial irradiation surveillance specia' ens stall te removed and examined, to cetermine changes in traterial proce-ties, as required by 10 CFR Part 50, Appendix H, in accordance with che scheca!e in Taele 4.4-2. The results of these eneminations shall be used to eccate Figures 3.4-2j CCMANCHE CEAK s L. NIT 1 3/a :-23

NOV El '91- 9:04 PA3E.017

^

REAC*:2 ::0; W 3 5 Eu OVER:RESSURE DROTE:*:3N 5*$~ EMS LIMIT:NG CONDITION FOR OPERAT!CN 3roups eMWo device 4

3. 4. 5. 3 At least cre of the #0l bwir erer tsare Protection //// edl

c. , cq te OPEUE4 he.n%e,3eadeo Coo \ of POLet(RC.S)

Yen i n Vgn-poge s.

T-o oo er yh o$ 198 S$4 aft. in operated _*elie' walves (20RVs) ith lif t settings unich vary with RCS temperature and which 00 not exceec the limits established ir.-Figure 3.4-4, or D. Two Resicual Heat Removal (RHR) suction relief valves ea:n with a ONM sdNMd M've."a.nd o ne.90RV w'A Sd.toids w pest.Aedd>ove,

e. HtffAllt! 11119111/11pl\ f171/ Itprribyt1755 /1/111 FU tsti- 11 ilt$pW 11M M dpA/19 J/19 !?/ats Jtt/W Apot!Castt:TY: MODE 4",

400E'5 and MODE 6 / )( the(reactor vesse 4// //.

AC*!ONi OhYhe kwo fo' tea % o W.sE% e. Ce%]

in MODE 4 a With7nArequired 7/ # #H ##//ed M (Att/M N//// /A//t V.nocerabl9 fitfff restore tu f7911 it IM-1H f/tOtt it1W /4/14f ithin 7 days or cepress>. 'ze and vent the RC5 to CPERABLE status "

throug at feast a 2.9S scuare inch we

%e tu80 Ovu pessuYe. c'.c%r the next 3 not.rs.

n 6evices

j. Aith t h re;Jired\ffl555 112 5 W fenktni 1 R nr.tJ2i /Alh / M1h /

C, inoperable.tcepressuriz the RCS tr sugt at least a 2.35 scuate 4ch kent witrin/ ar.d ven 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />; gb oQ.

/comq,\e.Ne. -aMon

~ /- In th' 'v'n: 'ith'r tn' '0RV5- r th' RHR Su: tion ref 3iv'5 cr

c. the RCS vent (s) are used to-mitigate an RCS pressure transient, a 5pecial Recort shall be preparec and submittec te the Ccmmissier oursuant to Specification 6.9.2 within 30 days. The report rai' cescribe the circumstances initiating-the transient, the effect cf the.DORVs, the RHR suction relief valves, or RCS vent (s) en tre transient, and any corrective action necessary to prever.t rect.rvce.

' h. - e. twvist-)nS OS Epc.iMIcok*on LO.4 0.4e. noY 0.pc.ahh.,

"Scecification 3.4;8.3 is not applicaole if all RCS cold legs _are greater than 320'F and the following conditions are mer.:

1. __ At least one reactor coolant pump is in operation

2. Pressurizer level is less than or eaual to 92%

3. The plant heatup rate shall te limited to 60'F in any one hou :er cc.

J , Wt+ 4ke,RC. vaded pc AC.TIOMS oy ; or C V8CLh the "^i NW0i b 3

w ormim b.WiO 000 04 *%nc we 4,u.u .

b o Y e. tre O r omve. ?rdedion dev'ites \nogetoMe..m MDT)ES 5 or jet.*er( uswte m ow cbon de ices 4o OPERASLE

i. Odcu w'#nk 2.+ bomb > or (2.) temQdte, k teu uve. s Ppssuv 'Ngg o t % e_.

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NOJ 21 '9$ 9:05 FA3E 018

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REACTCR :CCLANT SvsTEM BASES SMETv VALVES (Continued)

Daring eceration, all pressurizer Cece safety valves must be ODERABLE t:

prevent the RCS fren being pres:vrized aoove its Safety Limit of 2735 psig.

The coroined relief capacity of all of these valves is greater than the maximum surge rate resulting from a comolete loss-of-load assuming no Reacter tri until the first Reactor Trip System Trip setpoint is reached (i.e., no credit is taken for a direct Reactor trip on the loss-of-load) and also assuming no cperation of the power-ocerated relief velves or steam dump valves.

Demonstration cf the safety valves' lif t settings will occur only during shutdown and will be performed in ac:ordance with the provisions of Secticn C cf the ASME Boiler ano eressure Code.

3/4.4.3 PRESSURIZER The maximum wate- eclume also ensures that a steam butble is forrec ano thus the RCS is not a hydraulically solid system. The 12-hour periot : surveil-lance is sufficient to ensure that the parameter is restered to within its limit following expected transient c eration. The requirement that a minimum rumeer of pressurizer heaters be OPERABLE enhances the capability of the plant to control Reactor Coolant Systen pressure and estaclish natural circulation.

Pressurizer heater groups are powered from sources that reet the re;uirements cf Item II.E.3.1 of NUREG-0737.

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3/4.4.4 RELIEF VA UES 06 The power-operated relief valves (PORVs) anddteam tv0ble f;nction to relieve OC5 pressure during all design transients up to and inclacing the cesign step load decrease with steam dump. Operation of the PORVs minimizes the undesit aDie opening of the spring-leaded pressurizer Ccce safety valves.

ach PORV ' as a remotely operated block valve to provide a positive shutoff capability sneuld a relief valve become inoperable _

f 1MSERTA 3/4.4.5 REACTOR COOLANT SYSTEN LEAKAGE 3/4.a.5.1 tEAKAGE DETECTION SYSTEMS The RCS Leakage Detection Systems required by this specification a#e provided to monitor and detect leakage from the reactor cociant pressure boundary. These Detection Systems are consistent with the recomme, cations of Regulatory Guide 1.45, " Reactor Coolant Pressure Boundary Leakage Detecti:n Systems ," May 1973.

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COMANCFE PEAK + UNIT 1 B 3/4 4-2

tiO'J 21 '

91 9:06 PA3E.019 i

ph ' - g4woh L j.----- .s The POWS are equipped with automatte actuation circuitry and manual f -

control capability. Becault he Credit forrPOW operation is taken in the FSAR. --

the POWS are considered OPERA 8LE in

- analyses either for the Mode annual1, or 2automatic

& 3 transients,It mode. should be noted that the avtamatic --

sede is the preferred configuration, as this provides pressure relieving capability without reliance.on operator action. _ __ __ . . _ .

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P A 3 E . 0:'O tlOJ 21 '91 9:07 REa;*;4 :::. N 3 5'Ev ELSE5

RE53URE '!MPERAT'.4E L:MIT5 (Contiruec) wee *c0 (;;-ti seo)

The ese cf a ecmposite curve 's recessary to set conservative neatup i1nitations Decause it is possible for conaitions to exist such that ove* the course of the reatap rarp the cartrolling :oncitien switenes from tne inside to the octsice ano 'ne pressure limit must at all t'tes ce cased on ans!jsts of the most critical criterion.

Tne rew 10 CFR 50 Appsncix G rule addresses the metal temcerature of the This rule states that the mini-clorare head flange and vessel flange regions.

num metal temceratu e of the closure flange region should be at teast 120 degrees-F higher than tne limiting RTNDT for tnese regions when the pressy e exceeds 20 percent cf the p-eservice hydrostatic test pressure (521 psig for Westin;ncuse plants). ~e r Comantne Peak (nit 1, the minimum temperature of *he closure

'lange anc sessel fiange regions is 160 degrees-F since the limiting RT g.

is 40 degrees-F (see Table B 3/4.4-1). The Comanche Peak Unit 1 coo 10:wn curves sncan in Figsre 3.4-3 are impacted by this new rule, and therefore *,ne

" notch" in the cocidcan curves.

Finally, the compcsite cur es for the heatup rate data and the c:o co.a r rate data are adjustec fcr^possible errers in the prei+Jre and te?persta e sensing 'n:truments Dy tre values indicated on the ret:ective curves.

Altnough tne pressurizer coerates in temperature -'acges aDove th:se for whicn there is eason for tencern of renductile f ailure, opereting limits are proviced to assure coecatibility of operation with the f atigue analysis

erformeo in ac:orcance with the ASME Code requirementsOf. ComMrAph one,hM ad Mwgg e LOW TEuDERATURE NERPRESSURE PROTECTION The OPERASILITY of two DORVs, two RHR suct on reli or an RCS vent opering of at least 2.98 square irches ensures that the RCS will te protected from pressure transients whien could exceed the limits of 10 CCR 50 Aopendix G when one or more of the RCS cold legs are less than or equal :.o 350'F. Eitner PORV or wither RHR relief valve has adequate. relieving capability to protect the RCS f rom overpressuri:ation unen the transient is limited to either: (1) the start of an idle RCP with the sec:ncary water te'rperat'. ire of the staan generator less than or equal to 50*F above the RCS cold leg temperatures, or (2) the start of two charging pumps and their injection into a water-solid RCS.

The taxillum Nominal Allowed PORV Setooint curve is derived from analyses which model the performance of tne overpressure protection system for a range of mass input and heat inprt transients. Figure 3.4-4 is cased upon this analysis includirg consicerstier of tne maximum pressure overshoot beyonc the DDRV setp: int which can occur as a result of time celays in signel processing CCMANCHE oE4 - gN*7 1 S 3/4 4-13

Attachment 4 t0 TXX-91427 Page 1 of 3 SIGNIFICANT HAZARDS EVALVATION lhtCkgrELn.d The NRC staff recommended, via Generic Letter (GL) 90-06, several changes to the plant Technical Specifications. The intent of these changes was to improve-the availability of the pressurizer power-operated relief valves

-(PORVs)-'for use during plant cooldown, the mitigation of the design basis steam generator tube rupture (SGTR) event, and the mitigation of low temperature overpre3sure (LTOP) transients. In TXX-901053, dated i December 20, 1990, TV Electric committed to revise certain Technical l

. Specifications in respon;e to the recommendations of GL 90-06. The proposed I revisions are described in Attachment 1. The significant hazards consideration analysis, in acco_rdance with the three part test of 10CFR50.92, is provided below.

(1) Does the_ proposed revision involve a significant increase in the )

probability or conseauences of an accident Dreviously evaluated in the i Safety Analyses Report.?

Technical Specification 3/4.4.4 Relief Valves and Bases The proposed-changes would revise the Action statements to identify when the PORV block valves may be closed, when power may be removed from the block valves, and when PORVs may be shifted to manual control. As described in '

GL 90-06, these_ changes are intended to improve the availability of the PORVs for use in the mitigation of transients, particularly the SGTR event.

Recognizing that the PORVs may be required for mitigation of the SGTR event, a guidance 'is provided in the Emergency Operating Procedures which directs the 1 reactor operators to insure that power is available to the PORV blMk valves and that at least one block valve is open. The proposed _ changes to the Action statements.will ensure that power is provided to the b_ lock valve when the PORV is inoperable due to seat leakage, but otherwise available for use during the SGTR mitigation. The proposed changes also allow the shifting of a PORV to manual control in lieu of: closing and removing power from the block valve when the block valve is inoperable. The net effect of these changes on the accident analyses is that the availability of_the PORVs is increased for the mitigation of transients. The ability to mitigate the transient from a stuck open PORV.is affected by not closing an inoperable block valve but the ,

likelihood of such an event is minimized and considered insignificant based on i the requirement to place the affected PORV in manual control, i l

l_

The. proposed change would revise the surveillance required when the l

_ operability of the block valves is confirmed by full cycle travel of the valves. The. surveillance was enhanced to assure operability when needed and to prohibit testing when the valves should remain closed. Testing is' allowed when the PORVs-are operable or are inoperable due to seat leakage, and enhances-the ability.to assure mitigation _of accidents as analyzed in the CPSES FSAR by enhancing the availability of the block valve and its associated

PORV. The proposed changes, which would increase the availability of the PORVs for mitigation of transient':, do not significantly increase the ,

j. probability or consequences of any accioent previously evaluated in the CPSES )

FSAR. -i

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l I

.._..._.-.~.-.__.,______,__.;.-_____.., _ ,-m_,.,., .,.,~,.,--m.--r._.. . - , , , . . ,._ --,_m o,., - , - , . - , . . , - - _ - - .

Attachment 4 to TXX-91427

., 'Page 2 of 3 SURVElLLANCE_RJD111REMENT 4.4 ALZ This proposed change would require the Update of the figure which provides the PORV setpoints based on RCS temperature. ThesJ Updates would occur based on existing programs, lhe change merely makes the update a surveillance requirement and therefore, there is no impact on the accident analyses.

TECHN_LCAL SPECIF_lCATION 3/4.4.8d DVLR_P_R[.SIURE PRQJ((lLQN SYST(MS AND__ BAS [S The net affect of the changes to this Technical Specification would be the additional allowance of the use of one PORV (n@ one RHR suction relief valve to satisfy the LTOP protection requirements and the reduced allowed outage time while in MODES S and 6. -

As stated in the current basis for this Technical Specification, any one PORV or RHR suction relief valve has sufficient relieving capacity to prevent the overpressurization of the RCS as a result of a LTOP design transient. The ,

specification that any two valves be available for LTOP mitigation is required to satisfy the single active f ailure criterion, With the removal of the RHR autoclosure interlock (see reference [1]), there is no single active failure which could result in the designated RHR suction relief valve cnd the designed PORV being ur,available for LTOP mitigation. Therefore, the additional allowance for the use one PORV and one RHR suction relief valve for LTOP mitigation would not adversely affect the results of the design LTOP transient analyses. The redundancy of the LTOP mitigation devices has no effect on the probability of occurence of an LTOP design transient.

Reducing the allowed outage time in MODES 5 and 6 would reduce the probability of an overpressure transient when the plant is most vulnerable. This change would not affect the consequences cf any accident but it will reduce the probability of an overpressure event.

Summary The probability of an overpressure transient would be reduced by shortening the allowed outage time for the overpressure protection devices while in MODES 5 and 6. The proposed changes affect the consequences of an accident in the following manner- (1) the ability to mitigate a overpressure transient is improved by the increased availability of the PORVs; (2) the ability to mitigate the ef fects of a stuck open PORV is affected tiy not closing an inoperable block valve; however, shif ting the PORV to manual control minimizes the likeilhood of this event occurring and makes the change insignificant.

Therefore, these changes do not significantly increase the probability or consequences of any accident previously evaluated in the Safety Analysis Report.

REFERENCES

[1] NRC letter from Thomas A. Bergman to W. J. Cahill, Jr., dated October 18, 1991, entitled, " Comanche Peak Steam Electric Station, Unit 1

- Amendment 4 to Facility Operating License No. NPF-87 (TAC No. 8D862), t l

Attachment 4 to TXX-91427

,Page'_3 of 3--

(2) Does the Droposed revision create the Dossibility of a new or different kind of accident from any Dreviously analVZed?

= 0f the changes described above, the clarified wording, _the increased availability of the PORVs, operating with the PORVs in manual when the t lock valve is ir. operable, enhancing the block valve surveillance, and reducirig the allowed outage t$me while in MODES 5 and 6, do not create any new failure modes or equipment ccubinations or operating modes, not previously considered to determine potential accidents to be' considered. Only the_new equipment combination of using one RHR suction reilef valve and one PORV for overpressure protection has the potential to create a naw or different kind of accident. However, with the removal of the RHR autoclosure intenlock, there i . no potentui single f ailure that may make both devices inopercble. .Thus, the devices can continue to be assessed separately and the new combination does not create the possibility of a new or different kind of accident.

(3)-Does the Dtoposed revision involve a siunificant reduction in the maroin of safety?

Comnliance with the proposed Technical Specification changes would improve the availability of the overpressure protection devices. The allowance for the use ot.-one PORV and one RHR suction relief-valve for use in LTOP mitigation is

_ entirely within the current bases of the LTOP mitigation systems at CPSES.

-The ability to' operate the PORVs in manual for a period of time when the block valves are inoperable is considered an enhancement due to the increased availability of the'PORV for manual overpressure protection and RCS pressure control even though the duration of a stuck open PORV would be effected..

Hence, the proposed revisions-do not adversely-impact the safety of CPSES and is considered to offer an overall improvement in safety.

The margin of safety is defined as the difference between a regulated acceptance criterion _and the failure point for a particular parameter. The parameters of interest for the proposed revisions are the radiological

~

consequences.for the SGTR event and the RCS pressure for the LTOP events. As described in the'two previous sections, the proposed changes to-the Technical Specifications improve-the availability of the pressure relieving devices for use-in the mitigation of these events. Furthermore, the single active failure of these pressure relieving' devices has been considered in each of the affected safety analyses. Thus, the proposed changes do not adversely affect the results of_the safety analyses and the conclusions of the safety analyses remain unchanged-(i.e., the applicable acceptance-criteria are satisfied),

iThe proposed revisions have no effect on the failure values for the parameters of interest._ Because the current acceptance. criteria are met and the proposed changes do not' affect the corresponding failure values, there is no reduction in the margin of safety.

Attachment 5 to TXX-91427

,Page 1 of 1 Environmental Impact Determination 10 CFR 51.22(b) specifies the criteria for categorical exclusions from the requirement for a specific environmental assessment per 10 CFR 51.21. This amendment request meets the criteria specified in 10 CFR 51.22(c)(9).

Specific criteria contained in this section are discussed below.

(i) the amendment involves no significant hazards consideration As demonstrated in the Significant Hazards Consideration Determination in Attachment 4, the requested license amendment does not involve any significant hazards considerations.

(ii) there is no significant change in the types or significant increase in the amounts of any effluents that may be released offsite.

The requested license amendment involves no change to the facility and does not significantly alter the manner of operation in a way which could cause an increase in the amounts of effluents or create new types of effluents.

(iii) there is no significant increase in individual or cummulative occupational radiation exposure The proposed char,ges do not impact plant design features or operations that affect radiation protection, radioactive effluent processing, radioactive waste handling, or radiological environmental monitoring, The changes do not result in additional exposure by ;mrsonnel nor affect levels of radiation present. The proposed changes do nct result in significant individual or cummulative occupational radiation exposure.

Based on the above, it is concluded that there will be no '.mpact on the environment resulting from this change and the change meets the criteria specified in 10 CFR 51.22 for a categorical exclusion from the requirements of 10 CFR 51.21 relative to a specific environmental impact statement or environmental assessment by the Commission.

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