ML20042F787

Proposed Tech Specs,Increasing Max Fuel Enrichment Allowed at Plant to 5 Weight Percent U-235 & Incorporating Changes Described in Generic Ltr 90-02

ML20042F787
Person / Time
Site:	Sequoyah
Issue date:	05/04/1990
From:	TENNESSEE VALLEY AUTHORITY
To:
Shared Package
ML20042F784	List: ML20042F783 ML20042F787
References
GL-90-02, GL-90-2, NUDOCS 9005090387
Download: ML20042F787 (15)

Text

{{#Wiki_filter::;.. p ~ 2: S.,. !- - ...L. : r...- o' 1. ENCLOSURE 1 PROPOSED TECHNICAL SPECIFICATION CHANGE SEQUOYAH NUCLEAR PLANT UNITS 1 AND 2 DOCKET NOS. 50-327 AND 50-328 ~(TVA-SQN-TS-90-12) LIST OF AFFECTED PAGES Unit 1 3/4 5-4 '3/4 5-5 3/4 9-1 3/4 9-la Unit 2 3/4 5 3/4 5-5 3/4 9-2 .i ~l 9005090387 900504 ADOCK05000,g7 1 DR ~

~ 3/4. 9 REFUEllNG OPERATIONS W 9.1 BORON CONCENTRA110N LIMlllNG CONDITION FOR OPERATION 3.9.1 With the reactor vessel head closure bolts less than fully tensioned or with the head removed, the boron concentration of all filled portions of the Reactor Coolant System and the refueling canal shall be maintained uniform and ,, sufficient to ensure that the more rettrictive of the following reactivity conditions is met: Either a K,ff of 0.95 or less, which includes a 1% delta k/k conser-a. vative allowance for uncertainties, or b. A boron concentration of greater than or equal to 2000 ppm, which includes a 50 ppm conservat.ive allowance for uncertainties. APPLICABILITY: MODE 6* ACTION: With the requitements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS or positive reactivity changes.and initiate and continue boration at greater than or equal to 10 gpm of a solution containing greater than or equal to 20,000 p'U.95 or the boron pm boron or its equivalent until K,ff is reduced to less than or equal to 'f-concentration is restored to greater than or equal to 2000 ppm, whichever is ( the more restrictive. The provisions of Specification 3.0.3 are not applicable. SURVEILLANCE REQUIREMENT 5 4.9.1.1 The more restrictive of the above two reactivity conditions shall be determined prior to: a, Removing or unbolting the reactor vessel head, and b. Withdrawal of any full length control rod in excess of 3 feet from its fully inserted position within the reactor pressure vessel. 4.9.1.2 The boron concentration of the reactor coolant system and the refueling canal shall be determined by chemical analysis at least once per 72 hours. 4.9.1.3 One of the fo110 win 0 valve combinations shall be verified closed under administrative control at least once per 72 hours: R16 Combination A Combination B Combination C' Combination 0 a. 1-81-536 a. 1-81-536 a. 1-81-536

a. 81-536 b.

1-62-922 b. 1-62-922 b. 1-62-907 b. 1-62-907 c. 1-62-916 c. 1-62-916 c. 1-62-914 c. 1-62-914 d. 1-62-933 d. 1-62-940 d. 1-62-921 d. 1-62-921 c. 1-62-696 c. 1-62-933 e. 1-62-940 ' p 3 67 d,['q,j,t[ f. 1-62-929 f. 1-62-929 g. 1-62-932 g. 1-62-932 g7pW3[t/7.k h. 1-fCV-62-128 h. 1-62-696 t 'i. 1-fCV-62-128 .DereactorshallbemaintainedinMODE6wheneverfuelisinthereactor vesselwiththevesselheadclosureboltslessthanfu]ytepsioned r with the head removed, mkN 2D id SEQUOYAH - UNIT 1 3/4 9-1 Amendment No.12

] -a: INSERT SQN Unit 1 TS - Page 3/4 9-la 4.9.1.4 The boron concentration in the spent fuel pool shall be determined by chemical analysis to be greater than or equal to 2.000 parts per million (ppm) at least once per 72 hours during fuel movement and until'the configuration of the assemblies in the storage racks is verified to comply with the criticality-loading critesia specified in Design Feature 5.6.1.1.c. O e* t g e 4 s t i i

4 l , 5. 3 REACTOR CORE l FUEL ASSEMBLIES 5.3.1 The reactor core shall contain 193 fuel assemblies with each have a nominal active fuel length of 144 inches. assembly co Each fuel rod shall ' (.. d The shall have a maximum enrie.hment of 3.15 wei ht percent U-235. core loading l It49 shall be similar in physical design to the Reload fuel g: g;.a maximum enrichment of 4+ weight percent U-235.nitial core loading and shall have $.o w "M.'1 CONTROL ROD ASSEMBLIES ' 'rt. 'IEv$d ' gN. ...s.. ;l/.h I .m... .>.4 ,.9

5. 3. 2 The reactor core shall contain'53 full length and no part length control 4

rod assemblies. The full len th control rod assemblies shall contain a nom ' "iW i .142 inches of absorber materi 1. shall be 80 percent silver, 15 percent indium and 5 percent cadmium.T .. Wik

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' M '" control. rods shall be clad with stainless steel tubing. All f[#, 't.

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, l.f. .e.. ' ' y. ~ .. :,.. W. y - o. l 5.4 REACTOR COOLANT SYSTEMbi " p". '.. g.. ; a. ' N j;r[L ' g j DESIGNPRESSUREANDTEMPERATURE..,,,,.,...3;.l/ 1. .f,. e .yq'g,.-lj 5.4.1 The reactor coolant system is designed and shall be maintained: 15:3@v..~ a'. "y - i In accordance with the code requirements.speqified in Section 5.2 of the FSAR with allowance for normal de applicable Surveillance Requirements, gradation pursuant to the 4{HMM.. " e..r.a.g .. ' V)'" For a pressure of 2485 psig, and.,y.,<,.

* *3 n ' MF..,.

b. ~ r,. Aj' .;s r - c. For a temperature of 650*F, except for the pressurizer which is ~k c. u z. : 680'F. M k $p' VOLUME "'Y i' 5.4.2 The total water and steam volume of the reactor coolant system is 12,612 1 100 cubic feet at a nominal T of 525'F. 'i avg .s..r- +@g ' U 're : ) q. 5.5 METEOROLOGICAL TOWER LOCATION ~ lT .c " a

5. 5.1 The meteorological tower shall be located as shown on Figure 5.1-1, 1

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? .Q f.- DE,$IGN FEATURES 5.6 FUEL STORAGE- '{ CRITICALITY - SPENT FUEL 'b/ 5.0 EU. ~ 5.i.1.1 The spent fuel storage racks are designed for fuel enriched to 4 R64 r 2c. Sim.. weight percent U-235 and shall be maintained with-c'.?7. a. Ak0fI equivalent to less than 0.95 when flooded with unborated - )4,,.s-R17 .. e - water, which includes a conservative allowance of-t-49% delta k/k 4 for uncertainties.". 3.o6 Y, t " - b.- A nominal 10.375 inch center-to-center distance between fuel 31W5G/Ur assemblies placed in the storage racks. . Ire es d. CRITICALITY - NEW FUEL 5.6.1.2 The new fuel pit storage racks are designed and shall be maintained with a nominal 21.0 inch center-to-center distance between new fuel assemblies R118 such that keff will not exceed 0.98 when fuel having an enrichment of 4.5 weight percent U-235 is in place and optimum achievable moderation is assumed. 44,r-sc ,., p '=! :r":'unt i; -fted i: 5 0.::!;;h t,,: r ;nt :.; n;t;d in 0. 2.1 ;nd 5. 0.1.1. ORAINAGE ( 5.6.2 The spent fuel pit is designed and shall be maintained to prevent inadvertent draining of the pool'below elevation 722 ft. CAPACITY 5.6.3 The spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than 1386 fuel assemblies, ..s-5.7 COMPONENT CYCLIC OR TRANSIENT LIMIT n'w

1. f" 5.7.1 The components identified in Table 5.7-1 are designed and shall be d

maintained within the cyclic or transient limits of Table 5.7-1. ^For some accident conditions, the presence of dissolved boron in the pool water may be taken.into account by applying the double contingency princi.ple which requires two unlikely, independent, concurrent events to produce a criticality gg7 accident. SEQUOYAH - UNIT 1 5-5 Amendment No, 13, 60, 114 May 5, 1989

.REF0ELING'PERATIONS O i SURVEILLANCE REQUIREMENTS (Continued)

4. 9.1. 2 The boron concentration of the reactor coolant system and the refueling canal shall be determined by chemical analysis at least once per 72 hours.

4. 9.1. 3 One of the following valve combinations shall be verified closed urider administrative control at least once per 72 hours:

Combination A Combination B Combination C Combination D a. 2-81-536 a. 2-81-536 a. 2-81-535 6. 2-81-536 b. 2-62-922 b. 2-62-922 b. 2-52 907 b. 2-62-907 c. 2-62-916 c. 2-62-916 c. 2-62-914 c. 2-62-914 4. 2-62-933 d. 2-62-940 d. 2-62-921 d. 2-62-921 N-e. 2-62-696 e. 2-62-933 e. 2-62-940 f. 2-62-929 f. 2-62-929 g. 2-62-932 g; 2-62-932 h, 2-FCV-62-128 h, 2-62-696 i. 2-FCV-62-128 0 hY fue joieoM dopteMpearcon }N bM s/imr puez. Psoz. JHAu. BG Dd7m261fWe5 2y 8t/GuteAL paLy.s.Cs to ( 8 e sp2enrmL mnW og enanc ro accoppa, a ricosr ONCE PG12 9 A ffcut.S buACd6 fuel. vocG ML'HT~ A dh MrCL Nd doHF.CGutar.EoM sf ntf Asse!"ALW 5M 77/8.S7*b2A66 12ACK.5 r.s YetLEF e WE771772 C457BCMEr ta t)is,can knrues jr Lon6sMA eAsizoLrk h#CD'M3 .L 4. /. /. d. 9 SEQUOYAH - UNIT 2 3/4 9-2

BESIGN' FEATURES - 5.3 REACTOR CORE ...) FUEL ASSEMBLIES '5.3.1 -The reactor core shall contain 193 fuel assemblies with each fuel '. c.m9j ~ '""" assembly containing 264 fuel rods clad with Zircaloy Each fuel rod N.?f:? M~ M.n. shall have a nominal active fuel length of 144 inches. The initial core lm loading shall have a maximum enrichment of 3.15 weight t-

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p,6 % fuel shall be similar in physical design to the initial ercent U-235. Reload

1. fAl?

i go!Z shall' have a maximum enrichment of-4r* weight percent U-e leading and .s,..w n,s, ...u. +. :... ... ~ .2 L;cp.... > ;Q. ';q 7<, p ".-], , CONTROL ROD ASSEMBLIES ...g4(,f, p..cp q g:,,,,,7 x .-lQ .. ~, :. -}. u 5.3.2 The reactor core shall contain 53 full length and no part length control rod assemblies. .; a nominal 142 inches of absorber material.The full length control rod assemblies shall co j;

..a The nominal values of absorber

'. " material shall be 80 percent silver, 15 percent indium and 5 percent cadmium. 'All control rods shall be clad with stainless steel tubing. Y. w.w%,, 5.4 REACTOR COOLANT' SYSTEM )3 DESIGN PRESSURE AND TEMPERATURE f.14EGtf.t .W* Th{ reactor coolant syst m is de igned and shall be maintained: tM$$$k

• 5.4.1 B

. j. In accordance with the code requirements specified in Section 5.2 a. '?,h.V W* A of the FSAR, with allowance for normal degradation pursuant to the i ch appitcable Surveillance Requirements, y.

b..

For a pressure of 2485 psig, and "M.%.,7[,; For a temperature of 650*F, except for the pressurizer which is c. .., "., ' N 680*F. 9.;j'., - }- ~- 3.y- .:. M v VOLUME g.. g) 's g pp i Q 5.'4.2 The total' water and steam volume of the reactor coolant system is - MfM ' .m 12,612 i 100 cubic feet at a nominal T,yg of 525*F. 'isf!N d( 5.5 METEOROLOGICAL TOWER LOCATION , d... / - yi.' Jh o -5.5.1 The meteorological tower shall be located as shown on Figure 5.1-1. Q^M. AF qg=

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Amendment No. 37 4 .5 rMar uD. ..~.SQ.p. .. r:9;.;. - yf b >. s i <.<e;, I .:$t we. K:Y. {; r Mbka.c.... ~, +-

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DESIGN FEATURES I 5.6 FUEL STORAGE CRITICALITY - SPENT FUEL ), i 5.0.1.1 The spent fuel storage racks are designed for fuel enriched to 4,,. -+4-weight percent U-235 and shall be maintained with: RS2 y ~ A k,7f equivalent to less than 0.95 when flooded with unborated a. U-

e water, which includes a conservative allowance of W delta k/k for uncertainties.*

3.o65 1 k b. A nominal 10.375 inch center-to-center distance between fuel i

,Lgg g7-assemblies placed in the storage racks.

j ITM CRITICALITY - NEW FUEL 5.6.1.2 The new fuel pit storage racks are designed and shall be maintained with a nominal 21.0 inch center-to-center distance between new fuel assemblies such that kef t will n t exceed 0.98 when fuel having an enrichment of 4.5 weight percent U-235 is in place and optimum achievable moderation is assumed, t, 61 R52

r r i

N r t i ; l 'i t :' i; ". 0.. 6... n.

.., ~ ..m. ........w i ORAINAGE 5.6.2_ The spent fuel storage pool is designed and shall be maintained to i prevent inadvertent draining of the pool below elevation 722 ft. -{ CAPACITY 5.6.3 The spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than 1386 fuel assemblies. 5.7 COMPONENT CYCLIC OR TRANSIENT LIMIT 5.7.1 The components identified in Table 5.7-1 are designed and shall be maintained within the cyclic or transient limits of Table 5.7-1, i l i i 1 i c ? AFor some accident conditions, the presence of dissolved boron in the pool water [ may be taken into account by applying the double contingency principle which . A requires two unlikely, independent, concurrent events to produce a criticality R4 accident. SEQUOYAH - UNIT 2 5-5 Amendment No. 4, 52 October 19, 1987 44 4. g

INSERT jg SQN Units 1 and'2 TS Page 5-5 L Fuel assemblies with enrichment greater than 4.0 weight-percent U-235 c. and burnup less than 7,500 megawattday/ metric ton (mwd /mtu) shall be placed in cells in.the spent fuel storage racks that face adjacent cells containing either: 1. Fuel-assemblies with accumulated burnup of at least 22,000 mwd /mtu, or-2. Water t 0 !o J ? i t { s a 3 1 d

u -- W N ? t t' . 3 r ENCLOSURE 2 PROPOSED TECHNICAL SPECIFICATION CHANGE ~ I SEQUOYAH NUCLEAR' PLANT UNITS 1 AND 2 DOCKET NOS. 50-327 /0H) 50-328 (TVA-SQN-TS-90-12) DESCRIPTION AND JUSTIFICATION FOR INCREASING THE MAXIMUM FUEL ENRICHMENT TO 5.0 WEIGHT-PERCENT URANIUM-235 i ( e e f J 6 s

E i1 ENCLOSURE 2 Description of Change Tennessee Valley Authority proposes to modify the Sequoyah Nuclear Plant (SQN) Units 1 and 2 Technical Specifications (TSs) to revise Section 5.0, r- " Design Features," and add Surveillance Requirement 4.9.1.4 to increase the maximum fuel enrichment from the current 4.0 weight-percent to .5.0 weight-percent uranium (U)-235. This change will also permit the substitution of Zircaloy-4 or stainless steel filler rods or open water channels for. fuel rods in fuel assemblies. Reason for Change _The change to increase the allowable fuel enrichment is necessary to allow the use of higher discharge burnup fuel. Higher discharge burnups are achieved by reloading with smaller fresh batch fractions with higher enrichment. The change to allow the substitution of filler rods or water channels is desirable to permit timely removal of fuel rods that are found to be leaking or are determined to be probable sources of future leakage. Justification for Change Each core reload design used at SQN is confirmed to meet all design cyiteria and to be within the bounds of the accident analysis presented in Chapter 15 of the' Final Safety Analysis Report (FSAR) by performance of a reload s,afety analysis. This analysis considers modifications to the plant design and any changes to fuel design including increases in fuel enrichment. The performance of the reload safety analysis ensures the unit, with its specific core design and fuel enrichment, will operate within the prescribed safety limits. Any restriction on core operation identified through the reload safety analysis process is documented and any changes to the plant license are made at that time. Therefore, operation with revised Design Feature 5.3.1 allowing the use of fuel assemblies with a maximum enrichment of 5.0 weight-percent will be justified for each fuel cycle. TVA has performed a criticality analysia to justify the storage of fuel assemblies with a maximum enrichment of 5.0 weight-percent U-235. This analysis was sent to NRC for review as an enclosure in TVA letter to NRC dated February 14, 1990, "Sequoyah Nuclear Plant (SQN) Units 1 and 2 - Spent Fuel Rack Criticality Analysis for 5.0 Weight-Percent Fuel." This analysis concluded that by administrative 1y controlling the placement of high reactivity fuel in the spent fuel storage racks at SQN, the enrichment limit can be increased from 4.0 weight-percent to 5.0 weight-percent while maintaining required criticality safety margins. Therefore, fuel enriched to greater than 4.0 weight-percent with burnup less than L 7,500 megawattdays/ metric ton uranium (mwd /mtu) shall be placed only in E spent fuel rack locations with face adjacent cells that contain either fuel L assemblies with at 1 cast 22,000 mwd /mtu of burnup or water. Jnis p requirement is being included in Design Feature 5.6.1.1 of the SQN TSs. The analysis discussed above was performed with a conservative allowance of 3.06 percent delta k/k for uncertainties. This change to the allowance for l uncertainties has been included in Design Feature 5.6.1.1.a. W

f .f 1 x, - i s -An additional requirement identified in the criticality analysis is that during fuel movement and until the configuration of-the assemblies in the j s spent-fuel storage rack is verified to comply with the criticality loading j criteria, the spent fuel pool must be borated with greater than. 1 2,000 parts per million of boron in order to ensure that a dropped or misplaced assembly will not be a criticality concern. This requirement.is .] being added as a surveillance requirement of TS 3/4.9.1.. Design Feature 5.6.1.2 has been revised to remove the limit of 4.0 weight-percent for new fuel based on the justification provided above for changes to Design Features 5.3.1 and 5.6.1.1. The maximum enrichment for fuel. assemblies in the new fuel storage rack will remain 4.5 weight-percent U-235. The requirements for fuel assemblies specify the quantity of fuel assemblies and the number of fuel rods per assembly. Flexibility to deviate from the number of fuel rods per assembly is desirable to permit timely removal of fuel rods.that are found to be leaking during a refueling outage or are determined to be probable sources for future leakage.- This improvement in SQN's fuel performance program will provide for reductions:in future occupational radiation exposure and plant radiological releases. As stated in' Generic Letter 90-02, the substitution of filler rods or open i water channels for fuel rods is acceptable when' justified by cycle-specific reload analyses using an NRC-approved methodology. These reload analyses will demonstrate that existing design limits and safety analyses criteria are mat in advance of the next operating cycle. An NRC-approved methodology includes those methodologies acknowledged in the FSAR and applied in support of the issuance of the original operating license. It also includes those subsequent methodologies that have been - cubmitted to and-accepted by NRC's staff as amendments to the operating 1 - license. If the reconstitution of fuel asec.nblics through the use of' filler rods or I open water channels is extensive, this information should be reported to NRC. Therefore, if more than 30 rods in the core, or 10 rods in any assembly, are replaced during a refueling, a special report describing the number of rods replaced will be submitted to NRC in accordance with the provisions of the requested TS change. Environmental Impact Evaluation As stated in the Federal Register, Volume 53, Number 39, dated February 29, 1988, the NRC staff has concluded that the environmental impact summarized in Tables S-3 of 10 CFR 51.51 and in Table S-4 of 10 CFR 51.52 for a burnup level of 33 gigawattdays/ metric ton uranium i (GWd/mtu) are conservative and bound the corresponding impacts for burnup i

IL n T ' levels up to 60 GWd/mtu and fuel enrichments up to 5.0 weight-percent U-235.- NRC. based this conclusion on the results of a study performed by Northwest Laboratories and documented in the report entitled, " Assessment of the Use of Extended Burnup Fuels in Light Water Power Reactors," -(NUREC/CR-5009. PNL-6258) and the report entitled, "The Environmental Consequences of fligher Fuel Burn-up," (AIF/NESP-032). q; The proposed change request does not involve an unreviewed environmental question because operation of SQN Units 1 and 2 in accordance with this change would nott

1.. Result in a significant increase in any adverse environmental impact previously evaluated in the Final Environmental Statement (FES) as modified by the Staff's testimony to the Atomic Safety and Licensing-Board, supplements to the FES, environmental impact appraisals, or decisions of the Atomic Safety and Licensing Board.

.2. Result in a.significant change in effluents or power 13vols. 3. Result in matters not previously reviewed in the licensing basis for SQN that may have a significant environmental impact. 'l + i

BijM; sac.. it 1^*. v u -'-.J,%'.? 4 = " ); t n h n e' ,e. a,6, / ,... " N$, t p% am .f r e a w,. _c: g e.: a.-. t, @;, W ENCLOSURE:3' d um,b Ic d I E PROPOSED TECHNICAL SPECIFICATION CHANGE ^ 1 s .u ' lI, - -t '( 'SEQUOYAH NUCLEAR PLANT UNITS 1 AND pkm y; [ ~ ' ' DOCKET NOS;'50-327 AND 50-328 4 g'** (TVA-SQN-TS-90-12)- ,f,[ .{/ W DETERMINATION OF-. NO. SIGNIFICANT HAZARDS. CONSIDERATIONS - m 7 } p-' g." 5 g i ? .t, c 'N y v + v t. 1 -..I i ) t 4 # e y 0 0 ,h-gYf< t e 'l-I.t '. e )u U i k t f ;"! l lj, 4 1 l s {. 4 .7 M E ) i i ,'v' qH i i o a i 1

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l ENCLOSURE 3 Significant Hazards Evaluation 4 os TVA has evaluated the proposed technical specification (TS) change and has determined that it does not represent a significant hazards consideration based on criteria established in 10 CFR 50.92(c). Operation of Sequoyah ' Nuclear Plant (SQN) in accordance with the proposed amendment will nott (1) Involve a significant increase in the probability or. consequences of an accident previously evaluated. 'The safety considerations associated with reactor operation with higher enrichment and extended burnup have been evaluated..The { proposed changes _have no adverse affect on the probability of any . accident. The increased burnup may slightly change the mix of tission products that might be released in the event of a serious accident but such small changes would not significantly affect the consequences of serious accidents. The substitution of filler rods or open water channels for f uel rods will be justified by cycle-specific analysis using an NRC-approved methodology. This reload analysis will demonstrate that existing design limits and. safety analyses criteria are met. Therefore,the proposed change does not involve a significant increase in the probability of consequences of an accident previously evaluated. (2) Create the possibility of a new or different kind of accident from any previously analyzed. The proposed change to increase the maximum allowable fuel enrichment or the substitution of filler rods or open water channels' for filler rods does not create any new or different kind of accident from any previously analyzed. (3) Involve a significant reduction in a margin of safety. Based on the discussion provided in Item 1 above and the fact that no changes are being made in the types or amounts of any radiological effluents that may be released offsite, there is no significant reduction in a margin of safety.}}