ML20085G231

Application for Interim Amend to License DPR-4 & Change Request 33,authorizing Low Power Operation

ML20085G231
Person / Time
Site:	Saxton File:GPU Nuclear icon.png
Issue date:	04/29/1969
From:	Neidig R SAXTON NUCLEAR EXPERIMENTAL CORP.
To:
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References
FOIA-91-17 NUDOCS 9110240199
Download: ML20085G231 (52)

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{{#Wiki_filter:. l '/ SA JN NUCLEAR EXPERIMENTAL CORPOR,...ON O. Docket No. 50-146 b[2 License DPR-4 / ,, r ' {*/. +[ l Application f or interin Amenument of Operating Licence \\ and \\*,. [, Change Request No. 33 1. On August 27, 1968, Applicant submitted Change Request No. 32 to the Saxton Technical Specifications, and a request for amendment of Applicant's Operating License No. DPR-4, to allow operation of the Saxton reactor with Core III at a max.imum power level of 28 megawatts thermal. Applicant subsequently submit ted an Addendun to Change Request Kv. 32 dated August 30, 1968, and Amendments Nos. 1, 2 and 3 to Change Request No. 32 dated March 12, April 1, and April 1,1969, respectively. 2. Pending the review and approval by the AEC of the above mentioned app 19.ation of August 27, 1968, and supplements thereto, Applicant submitted a request dated March 19, 1969, and a revision thereof dated April 1, 1969, for loading of Core III fuel into the Saxton reactor. This request was granted by Amendment No. 5 of License DPR-4 dated April 11, 1969, and the accompanying Change No. 32 to the Technical Specifications. 3. Applicant has completed loading the core componen'ts authorized by Amendment No. 5. Pending the review and approval by the AEC of the above mentioned August 27, 1968, application and supplements thereto for operation of the Saxton reactor with Core III c*. a maximna power Acvel of 28 megawatts thermal, Applicant hereby requests that, as a sacond intermediate step, Operating License No. DPR-4 be amended to allow operation of the reactor at steady state power level not e.:cceding one megawatt thermal. 4. In support of this request, Applicant hereby cubmits proposed revisions of Sections F, G.3., and N.4.e. of the Technical Specifications and a proposed addition to Section K. These proposed changes are designated as Change Requent No. 33. 5. Activities to be conducted prior to operation at 1 MWt under the authority. hereby requested will include loading the five romovabic subassemblics, hot functional tects on the newly installed recirculation system and the pressurizer loop-seal installation, pressurizing the main coolant system and heating to 500*F, and zero power physics tests, l I 9110240199 910424 l PDR F01A DEKOK91-17 PDR

.2-- s. e I ( SAXTON NUCLEAR EXPERIMENTAL CORPO iTION l I By ___ /s/ R. E. Neidic,, President 4 ) i i i 1 .2 (S E A L) Attest: a i /s/ R. B. Ecist Secretary i i J ? l Sworn and subscribed to before me this 29th day of April 1969. l (S E A 1.) i F /s/ Charles J. Ansel Notary Public Muhlenberg Township, Berks County Py Cornission Ltpires October lb,1970 ) ) r i I 4 i

. ~ ........ ~... ~,. DPR-4 i Technical Specifications i Change Request No. 33 y Page 1 of 14 Pages i I l 1. DESCRIPTION OF CllANGE During the operation of the Saxton Reactor with Core III, at a l steady state power IcVel not to exceed 1 megawatt (thermal), the Technical Specifications shall be changed as indicated below. Except as changed, all remaining provisions of the Technical Specifications shall remain in effect. Make the folJoning changes to the Saxton Technical Specifications. Change Section F. in its entirety

• F.

RFACTOR COPE A reactor core having the following features shall be provided: 2. The main coolant shall be light water, and shall serve as the moderator and reflector. The designed effective re- [ flector thickness shall be 10 inches. 2. Mixed natural uranium and plutonium' dioxide' enriched l initially to a nominal 6.6 w/o Pu02 and previously irradiated in Saxton to a maximum fuel burnup of 33,500 MWD /MTM shall be in seven of the nine central fuel assemblics. The remaining 2 central assemblies will be unirtadiated enriched UO2 fueled, having enrichements which vary from a nominal 5.7 w/o to a nominal 12.5 w/o U-235. Eleven of the enriched twelve peripheral assemblies, except for the test assemblies described in F.3.f. Will-be irradiated UO assem-2 blies (originally 5.7 v/o U-235 enriched) from Saxton Core 1. The twelfth peripheral assembly will be unirradiated and con-- tain 5.7 v/o U-235, Zircaloy-4 clad fuel. l l 3. The fuel assemblies shall be supplied as follows: i a. General

Description:

Plutonium fuel assemblies Each main plutonium fueled assembly shall have a total } overall length of 50.23 inches with a nominal fuel length of 36.6 inches and shall approximate a 5.386 l inch square in cross section. The fuel rods shall be composed of Zircaloy-4 clad ceramic pellets or vibrationally compacted fuel. _ The-rods shall be arranged in a square lattice with an initial 0.580 inch ce,nter-to-center distance.

• Section F.J. through F.3.c. herein are identical to the existing Section F.1. through F.3.c. designated in Change No. 32 approved April 11, 1969.

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DPR-4 Technical Specifications Change Request No. 33 Page 2 of lJg Pages F. REACTOR CORE (Continued) Every alternate rod will be a water filled tube so that the center-to-center distance of the fuel rods will be 0.820. The ceramic pellets shall have a diameter of 0.3374 inch (nominal) and a length of 0.3660 inch (nominal). One end of each pellet shall be initial 3y dished. The total pellet column tolerance shall be 0.183 inches initially. The maximum initial moisture content of the pellet column shall not exceed 30 ppm on a weight basis. The maximum initial nitrogen content of the pellet column shall not exceed 100 ppm on a weight basis. The vibrationally compacted loose oxide shall have a total column tolerance of 0.188 inches initially. The maximum initf11 moisture centent of the loose oxide fuel column ', 300 ppm on c weight basis. The maximum initia) n' sgen content of the loose oxide fuel columr 00 ppm on a weight basis. The initial clad diameter shall be 0.3445 inches (nominal). The initial diametral clearance for the pelletized fuel -shall be 0.0071 inches -(nominal). The cladding shall have a wall thickness initially of 0.0233 inches (nominal). The gap between the pellet column and the internal plug end shall contain sintered aluminum oxide (A10 ) discs 23 j to provide a minimum end gaps initially of 0.609-inches, j The fuel rods shall,be_ initially hermetically sealed with end plugs welded to the tubing. The end plugs shall be Zircaloy. l The top nozzle of _ these assemblies is removable. L l b. General

Description:

Uranium fuel assemblies from Core I. Each Core I uranium fueled assembly shall have a total over-all length of 50.25 inches with a nominal fuel length of 36.6 inches and-shall approximate a 5.386 inch square in cross section. l l ~ w--,,r. ,e-,, y v ,.,.v- -~ =,.~r, m --,-...,.,, ~,,, - -, ,a S. vc w= wv.-

I ~ j Technical Specificationc Change Request No. 33 Page 3 of 14_ Pages C ,F. REACTOR CORE (Continued) s The fuel rods shall be composed of stainless steel tubes which contain uranium dioxide fuel in the form of cylindrical ceramic pellets. The rods shall be arranged in a square lattice with an initial 0.580 inch center-to-center distance. The pellets shall have the following initial dimensions: Diameter (nominal) 0.357 inches Length (nominal) 0.732 inches The ends of each pellet shall be dished initially. The total pellet column tolerance shall be 0.366 inches initially. The initial clad inside diameter shall be 0.361 inches. The diametral clearance between clad 1.D. and pellet 0.D. shall be initially 0.004 inches. The gap between pellet stack and internal plug end shall contain sintered aluminum oxide (A1 03) circular 2 hollow discs, to provide a minimum of 0.174 inch end gap. The initial moistura content of the pellet stack shall not exceed 75 ppm on a weight basis. The fuel rod ends shall be initially hermetically sealed with end plugs welded to the tubing.,Those fuel rods which require no further welding shall be clad with 0.015 inch wall of Type 304 welded stainless steel 10% cold-worked with a 400 ppm maximum cobalt content. The end plugs shall be Type 304 L or 308 stainless steel. Those fuel rods which require subsequent brazing shall be composed of 0.028 inch wall of Type 348 modified carbon, annealed stainless steel with a 500 ppm maximum cobalt content. The end plugs shall be Type 304 or 304 L stainless steel. c. General

Description:

Unitradiated Uranium Fuel Assemblics Two of the central nine assemblies will be unirradiated uranium fueled and shall have a total overall length of 50.23 inches with a nominal fuel dength of 36.6 inches and shall approximate a 5.386 inch square in cross section. The fuel rods shall be composed of ceramic pellets clad in Zircaloy-4 and stainless steel and the rods shall be arranged in a square lattice with an initial 0.580 inch center-to-center distance. The dimensions and enrichment of the ceramic pellets.are given in the addendum. (1) (1) WCAP-7219 Addendum to Saxton Core III Licensing Application (Westinghouse Confidential) Julv 1968. e __m-_- _._____m__

l 1cchnical SpecificatJons. l Change nequest 1;o. 33 Page _f._ of 14 Pages i l F. REACTOR Cone (Continued) The top nozzle of these assemblies is removable and held in place with three tie rods. Two of the tie rods in the UO2 fueled assemblies will be filled with Inconel and one of the tie rods will be filled with stainless steel. One of twelve peripheral assemblies will contain unirradiated 5.7 w/o U-235 enriched Zircaloy-4 clad uranium fuel. This assembly has a reinforced assembly can and shall have a total overall length of 50.23 inches with a nominal fuel length of 36.0 inches and shall approximate a 5.386 inch square in cross section. The fuel rods shall be composed of Zircaloy-4 tubes which contain uranium dioxide fuel in the form of cylindrical ceramic pellets. The rods shall be arranged in a square lattice with an initial 0.580 inch center-to-center distance. The pellets shall have the folloving initial dimension: Diameter (noninal) 0.338 inches Length (nominal) 0.600 inches The ends of each pellet shall be' initially dished. The total pellet column tolerance shall be 0.656 inches initially. The initial clad inside diameter shall be 0.3445 inches (nominal). The diametral clearance between clad I.D. and pellet 0.D. shall'be initially 0.0065 inches. The cladding shall have a thickness of 0.0245 inches (nominal). The gap between pellet stack and internal-plug end shall contain sintered aluminum oxide (A10 ) circular hollow 23 discs, to provido a minimum of 1.473 inch end gap. The-initial moisture content of_the pellet stack shall not exceed 30 ppm on a weight basis. The fuel rod enda shall be initially hermetically sealed with end-plugs welded to the tubing. The end plugs shall be Zircaloy-4 l \\ s .e .,_mm.. . _ _ _.. ~, _ _,.. - _ _,. _.., _ _ _,. _ _... _ _,.,. ~. -,. _.

TechnicaJ Spec 111 cations Change Request No. 33-Page J of E Pages ' F. REACTOR CORE (Continued) d. Special Removable Subassemblics The' center plutonium-oxide fueled 32 rod " loose-lattice" assembly contains the 3-1 test subassembly position which will be occupied by a removable subassembly. The sub-assembly shall contain four previously irradiated fuel rods on the " loose-lattice" pitch; i.e.,.820 inches and are described in Section F.3.a. The four corner positions vill be occupied by Zircaloy-4 vater-filled hollow tubes. The center position will contain a flux thimble. The four rods may be replaced by other previously irradiated plu-tonium oxide fueled rods described in Section F.3.a. Subacscibly No 503-4-33 shall be availabic as an al' ternate for the N-1 position. The center position of the subassembly will contain a flux thimble-and the four corner positions will contain four water-filled hollow tubes..The four re-maining positions will be cccupied by Core II'Pu02-UO2 fueled rods. l c. Special L-Shaped Subassemblies l The L-shaped subassemblies shall be composed of ninc fuci rods spaced on an initial 0.580 inch pitch. The pitch shall be maintained along the length of the a*ssembly by five parallel row-. of formed tubular ferrules brazed to edjacent fuel rods. f. Test Fuel Subassemblies Tes't Fuel Acsembly No. 503-4-25 One 9-rod subassembly shall have the four cernor rods and the center rod clad with Type 304 stainless-steel having a nominal thickness of 15 mils and shall centain uranium dioxide (UO2) Pellets enriched to 5.7% U-235, having a pre - vious irradiation exposure of approximately 10,000 MWD /MTM.

• This subassembly has not previously;been suemitted for approval 1

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DPK-4 Technical Specifications Change Request No. 33 Page 6 ef 14 Pages T. REACTOR CORE (Continued) The four remaining test fuel rods may be chosen in any combination from those described in F.3.h. This subassembly may be operated in any peripheral 3 x 3 location. ^ Test Fuel Assembly No. 503-4-28 (3) ~ One 9-rod subassembly shall have the four corner rods clad with type 304 stainless steel having a nominal thick-ness of 15 mils, and contain UO2 pellets enricherl to 5.7% U-235. Two of the remaining rods shall be clad with type 304 stainless steel having a nominal thickness of 15 mils, and contain UO2 pellets enriched to 5.7% U-235, and bc in-strumented w3th a thermocouple located in the Thel near the top of each rod. The remaining two rods sh'01 contain-092-347 stainicas steel cermet fuc1, and be instrumented wit h two theruoccuples located in the fuel rmar the center t.nd bottom of each rod. Germet fuel compositaon shall be 347 stainless steel (72 w/o), U02 - (28 w/o), enriched to 44%- U-235, clad with type 347 stainicss steel (0.351 inch 0.D. x 0.17 inch nominal for the outer wall and 0.102 inch 1.D. x 0.017 inch nominal for the inner'wcll). A' standard Saxton flux thimble will be provided in the center of the subassen;bly. The thermocouples for the UO2 fuel rods are positioned in center holes drilled into the fuel pellets, and are scaled at the top end of the fuel rod.- The thermocouple sheath is 347-stainless steel having a nominal 0.D. of 1/8 inch. The thermocouples for the cernet fuel are positioned in the center hole and are sealed at the top end of the fue)-rod. The thernoccupJes sheath is 304 stainless steel having a nominal 0.D. of 3/32 inch.- Two thermocouples to record water temperature are provided in diagonal corners of the subassembly, and are held in place by clips spot welded to the subassembly enclosure. 1 This cubassembly smay beloperated in any peripheraly3 x 3 i location. 1 n ~. _ _ _. _ -,, -.,.,. - _. - _ _. _..., _,, -. _. _,, _,

Docket No. 50-146 l DPR-4 + Technical Specifications Change Request No. 33 Page 7 of 14 Pages l F. P.EACTOR Conli(Continued) Test luel Assembly No. 503-4-32 One nine (9) rod subassembly shall have an 0.580 inch pitch. The center location of the subassembly will contain a flux thimbic and the four corner rods of the subassembly sball be clad with Zr-4 having a nominal thickness of 23 mils and contain uranium dioxide (UO2) pellets enriched to 12.5 w/o U-235. The remainingibur test fuel rods may be chosen in any combination from those described in F.3.h. This subassembly may bt operated in any peripheral 3 x 3 location. Test Fuel Ascembly No. 503-4-33 One nine (9) rod subassembly shall have a center-to-center distance between rod positions of 0.580 inches. The center position of the subassembly will contain a flux thimble and the four corner positions will contain four Zircaloy-4 water-filled hollow tubes so that the center-to-center distance of the fuel rods will be 0.820 inches. The four remaining l test fuel rods may be chosen in any combination from those described in F.3.h. This subassembly may be operated in any peripheral 3 x 3 l location. i Test Fuel Assembly No. _503-4-34 This subassembly may be either the primary or the alternato design #503-4-34A as described in Change Report No. 17.. I The primary design will contain a modified grid and end plate to accommodate two materials compatibility test rods. These test rods will ba Zr-4 to which are attached 10 stainless steel sleeves 2.425 inches long. The sleeves are mechanically attached to the Zr-4 tube by bulges. The primary design will l also contain two test. fuel rods which may be c.hosen in any combination from those described in F.3.h.

• This sebassembly has not previously been submitted for approval.

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DPR-4 Tec1mical Spectfications Change Request No. 33 Page _8_ of,}i Pages F. REACTOR CORE (Continued) ')he alternate design vill contain a modif' icd end plate and grid as described in Change Report No. 17 Two stainicss steel s3ceves are attached to the Zr-4 tube by bulges. The test sections are notched and held between ' grids by the notches. This subassembly vill contain no fuel rods. This subassembly may be operated in any peripheral 3 x'3 location. '~ ~ Reactivity Osci11ator Tests The reactivity os u11ator subassembly described in Chance Report No. 11 and that described in Change Report No. 13 may be inserted in any peripheral 3 x 3 location. Both oscillators contain a mov-abic tube inside a stationary tube. Reactivity change is caused by hafnium sleeves in the movable tube moving and and out of the stationary hafnium sleeves. E. Test Capsu3es Test. capsules containing non-fuel material may be inserted in any of the eleven dummy fuel locatsons adjacent to the t reactor core region or in any of the eight. irradiation sampic tubes on the peripher/ of the core. l Test capsules containing devices may be inserted in irradia-tion sample tubes in the three peripher'a] removable assem-blies. All the capsules described above will require that: 1. Prior to irrat.iation, the design of the test capsule has been evaluated b, the SNEC Safety Committee and found-acceptabic with regard to physical, thermal and hydrau-lic performance, and effect en core reactivity, neutron flux, and reactavicy coefficients. 2. No foreseeable failure of a test capsule could result in mechanical damage to any core component or in any manner alter the ability of t.hc control system to function. h.- Test Fuel Rods 1 1. Rods identical to standard stainless steel clad UO2 fueled rods except that they are internally pressurized to produce tensile stresses up to approximately 41,500 psi in the clad at reactor operating conditions. (4) These rods may occupy l-any peripheral 3 x 3 location. G - + = = * 'y. -v m n r er *- m-ei-u,-----e-r---- -+y.- g v ,.c w -r +=n-,-u- , -+

va s.s s .w. mv 4 DPR-4 7echnien) Specitjectjons Change netuent No. 33 1 Page 9 of & Pages .T. Kl:AC104 CORE (Continned) 2. Rods containing carefully controlled fluoride, troisture and clad bore imperfections as desctibed in Change ) Report No. 18. The ten rods will be pressurized to 2000 psi or less. The experin.cntal parameters are two s longitudinal notches.00 D inches deep x.004 inches wide x.187 inches lonC, both located on the clad I.D. surf ace, snoisture content of 2E-150 ppm and local fluoride contan.ination at the notched areas. These rods niay occupy any peripheral 3 x 3 location. a 3. Rods clad with Zircoloy-4 having a nominal thickness of 23.0 mils and internally pret.surized. (5) 'Ibese rods way occupy any peripheral 3 x 3 location. 4 Rods clad with Zr-4 having a nominal thicknen of 23.5 n:11s and internally pressurized to produce tensile stresses of up to 22,000 psi. (6) These rods may occupy any peripheral 3 x 3 Ivention. other rods will be chosen from anong the following: y,t;caloy-4CladdingJ12 I Clad thickness 20.0 to 27.0 inils Clad 0.D. 0.391 inches Pellet diaceter range 0.325 to 0.347 inches Enrichment range 0,20 to 12.5 w/o U-235 Peak power range up to 16.0 kv/ft Pressurization peak tensile stress hot beginning-of-11fe vill be less than 22,000 psi Pellet Jensity 80 to 100%-theoretical Stainless Steel Cladding Clad thickness 10.0 to 20.0 mils Clad 0.D. 0.391 inches Pellat diameter range 0.341 to 0.371 inches Enrichment range 0.20 to 12.5 v/o U-235 Peak power range up to 16.0 kv/ft Pressu-ization peak tensil. stress he,t beginning-of-life vill be less than 31,000 psi Pellet density 60 to 100% theoretical These rods inay occupy any peripheral 3 x 3 Jocation. (1) NCAIC72i9 Addendum to Saxton Core III 1.icensing Application (Westinghouse Confidential) July 1968

1)ocl.et No. 50-14 6 1)PR-4 Technical Specificotions Chai.ce lleguest No. 33 i Page _1j of 14 Pages i 1 i Chanpc Section C.3 to read: G. _CONU;0L AND SATETY SYSTE.M_S l 3. The reactor shall be automatically tripped under the following conditions: Condition? Set Point i i Tast startup rate 2 decades / min. 1ow eain coolant pressure 1800 psia i liigh main coolant temperature (hot leg) 511*r Low water Icyc1 in pressuriecr 0.3% Lor.s of main coolant pump power Contact on breakers, failure of power supply, or loss of variable frequency set clutch quency set is cupplying power excitation when variabic frc-for main coolant pump operation ' liigh power level at pover: I The overpower trip set point will be set 'for a thermal power no greater than 5 MWt. 7 l l 1 I 1,, r arw-,r-r--s--s-ir-rww-:-- =g-aswr e arvreisi Wwww e p e r -t e-s*ff*fe==+,,w,%=Hveite+..- M m.e egy-w e4pw-e w p,q v yw ypw r our y*- ew' avvti- - %r* -qs M w pe g-r e-r w eg wa-+cyq v y, a,9 y vv a

- ~- 1 Dochet h'o. $0-146 Dpit-4 Technical Specifications Chance 1:cque.; i0. 33 Page 11,,, of 14 Pages { l i l Add lo Section K K. EMTRGEI?CY COULI!;G SYSTE!!S r 5. In the event the initial refueling water inventory is injected following a main coolant systems rupture, a fully automatic spilled coolant and injection water recirculation system ufil be provided to naintain the core flooded and insure lonD term decay heat removal. This systert shall conist of two 60 gpm submersible canned motor pumps piped to deliver spilled coolant into the reactor vessel via connections to the safety inicction system lince. A recirculation initiation signal derived frorn. redundant low water icvel signals in the S0,000 gallon borated water storage tank will automatically open the isolation valve-in each recirculation path, start both recirculation punips, clote the safety injection valves, and stop the safety injection l pumps. 8 4 I i I l l l i e l A. --- - L --.-,,.,,,..---w,--,-,-- ~ ~ - ,,w._--.. --,,..v.~.-,- ,,~,.,-~~---~,n,---_..

Docket Nu. 50-146 DP},-4 l Technical Sperificatitinc Change neque ^4o. 33 Page J of j Pages _ Change Section N.4.e to readt i N. AD.'11NISTRATIYB A"D PROCEDURAL SATEGUARDS 4 During "licactor Operation" (as defined in Section N.3 above) the i following operating limitations shall applyt e. Reactor Core 1. The steady state reactor power level shall not exceed 1 int i 2 Maximum number of fuel assemblies in Ccre 21 3. Maximum fuel burnup (fuel assemblien) 60,000 IND/ NTH Maximum fuel burnup (control t od fellowers and 1. asr.coblies) 60,000 FND/MrH 2 4 Maximum heat fibx (cLeady state) 26,743 BTU /hr-ft 5. Average heat flux 6,647 BTV/hr-ft

6. _ Minimum DNB ratios (W-3 correlation)

Power - 1 int 22* i Overpower - 5 }NL >5* 7. Maximen fuel clad surface temperature (at nominal system pressure) 520'F 8 Maxinum fuel speelfic power <1 kw/ft 9. Average power density ~ 3 kw/ liter 30 Overpower trip setpoint 5 }Nt

11. The design maximum void coefficient of reactivity.at operating-temperature.

-0.0015/% void

12. The design rnaximum temperature reactivity defect (cold cican to hot clean) 0.05

13. -The moderator temperature coefficient of reactivity shall not be more posftive than a) Borated at 495'F, 'l FNt

-0.2 x 10~ /*F ~

• No nucleate boilint f

DpH-4 Technical Specifications Change Request No. 33 I page 13 of 14 Pa ter. 11. pt'EPOSE OF CilANCE ) To operate the Saxton reactor with Core 111 at lou power in order to conduct rero power physice measurements and operator training. t III. SAFETY CONSIDERATIONS { Tabulated below for comparison are some of the Core III design perameters reported in Safeguards _Pepj1rt for Saxton Core 311 und t the value of the pararneter for 1 INI operation. Core III Design Value at 1 FNL Value Operation Maximum steady reactor power 28 ?!Wt 1 FNL 2 !!aximum heat flux (steady 799,800 ETU/hr-ft 26,743 BTU /hr-ft state) Hinimum DNB ratio 100% power 1.75 22 nominal conditions Maximum fuel clad surface 657 F 520 F i i temperature Maximum fuci specific power 24.0 kw/It <1 kw/ft l As shown in the above table, the temperature, DNS ratioc, heat flux and specific powcr of Saxton Core III operating at 1 int are extremely j conservative co.npared to the Core III. design values. l l The 1 FNt power level will be determined by measuring the core at j with platinum resistance thermometers and the primary flow with a l venturi. The accuracy of the measurementL will be approximately 3%. l Operation at 1 }Nt gives a wide safety raargin of over 20 FNt. Further, the fission product inventory of Core III at beginning-of-life will be at least 30% less than that of Core II at the end-of-life and the fissile inventory of Core III at the beginning-of-life-is 18% lower than that of Core 11. Based on-the above tabic, the reduced fission product and fissile inventory of Core 111 relative to Core II, and the Safeguards Report for Saxton Core III, it is our conclusion that operation of Core III at 1 Mut results in no danger to the public or operating personnel. l .-,-e.Ew,-.,w e, r.. ym- ,.--,,,wmi.,.... my,,m m m ,n.--wy. , em ~ e 1 v.---,, ..r w..e. - -3...n.ew. p.m m wJ

pot s.e t hv. bu-A s v I - DPit-4 7echnien) Specifications Change licquest No. 33 l'a ge 14 of 14 l' ages IV. Ill:A1.Til A"D SAP!.TY j i j It is our conclusion that the health and sciety of the Dublic will not be endangered by this change. i i i i l t ) (2) I'reviously approved in Chann No. 18, 22, 23, 24, 25 and 30 as Test l Fuel Assoir.bly XI, XIII, XIV, XV and XXI. L l (3) l'reviously approved in Change No. 27 as Test fuel Assembly XVI. l l 22, 24, 25 and 28. l (4) Thece rods have previously been approved in Change No. i I (5) These rods have previously been approved in Change No. 29. 1 (6) These rods have previously been_ approved in Change No. 30. l l I i l l e 9 T 4 , - -,w- ,,, ~++ w,.-, -w,, ,% r ~r,.,m *,em.-er .ww.e-.- - =,. -. 2,--, .,.-wwe-...,-,-*---n-., ,---v-w-,,,, -e- -+e, -~~~e., + - - -

i Oocket No. 50-146 DPR-4 Change Report No. 17 Page 1 of 3 Pages Change Report No. 17 + t MATERIALS COMPATIBILITY TEST 1 DESCRIPTION OF CHANGE f A subassembly designated Test Assembly No. 503-4-34 will contain either a primary or an alternate materials compatibility test in which case it will be designated 503-4-34A. The subassembly containing the primary materials compatibility test is illustrated in Figure 1 and the prinary test rod in Figure 2. The subassembly containing the alternate materials compatibility test is illustrated in Figure 3 and the alternate test sections are illustrated in Figure 4 The subassembly for the princry test l will contain two fuel rods in addition to the two materials compatibility test rods: whereas, the subassembly for the alternate test will contain no fuci rods. The rods which will be with the primary materials com-patibility test will be clad with Zircaloy-4 having a nominal wall thick-ness of 23 mils and contain uranium dioxide (UO2) pellets, uniformly en-riched to 12.5 w/o U-235 and 89.5% to 94% the'oretical density, a. Primary Test Design The primary design illustrated in Figures 1 and'2 consists of a Zirculoy-4 tube to which are attached 10 stainless steel sleeves, 2.425 inches long, equally spaced over the tube length. The sleeves are mechanically attached to the Zircaloy tube by bulges. Each stainless steel sleeve in the subassembly grid area has velded to it four Inconel spring fingers which hold the tubo in place by pressing against a stainless steel cylinder which is welded into each grid of the test subassembly, b. Alternate Test Design The alternate design is illustrated in Figures 3 and 4 In this design the Zircaloy tube test sections containing stainless steel l l l l 4 +.,. -.--v ,,,-m., mw-,,v,- ,,-o,.> ,-+,~,~-,r ,c,-r - -.s w n,n-,,w r ,e,, y a. , + m----..n-vmm,-- r- -r,,-w,-,.,,,-rn ~

P Docket No. 50-146 i i DPR-4 Change Report No. 17 Page 2 of 3 Pages I sleeves are notched and held between the subassembly grids by i the notches. As in the primary test design, the sleeves are mechanically attached to the Zircaloy tube by bulges. Section B-B of Figure 3 is an illustration of the cross section of the alternate design subassembly showing the arrangement of the test tections in the assembly. The outer two cytindrical cross ecctions are part of the grid attucture while the inner two are test sections. 2. PURPOSE OF CllANGE I The purpose of the change is to allow substantiation of the expectation of negligible crevice corrosion between stainless steel and Zirealoy-4 1r an operating Pk'R environment. I 3. SAFETY CONSIDLRATIONS The parameters pertinent to safety considerations for the primary ma-l terials compatibility test design are presented in Table I for rAe us-sembly containing the tests and two 12.5 w/o U-235 enriched fuel rcJs. i The location of the fuel rods and materials compatibility tests in the i E assembly are shown in section B-B of Figure 1. I It can be seen in Table 1 that the calculated values of specific power, heat flux. DNB ratios, and fuel rod and rod surface temperatures are well within safe limits. Further, it is highly unlikely that the ma-terials compatibility test should fail in such a way as to interfere I with the heat transfer from the two fueled rods. It is_ highly unlikely in the extreme that a failure of an alternate test section could cause an unsafe condition since the subassembly containing the, backup alternate test will contain no fuel rods, 4 -.~~..,.--.sm.. -,~.nn,----,..r- -,,w--m-s~,--+-A,~,4,,.,a-..,--,,,- -En - w n - e ---,m --,, w n,., m a we n, m v e-- w.r ne-w,--

-. ---. ~ - - -- - - - - - - - - l Docket No. $0-146 l DPR-4 ChanFe Report No. 17 Page 3 of 3 Pages j J i 1 I 4 HialTH AND SAFETY It is our conclusion that the health and safety of the public will not be endangered by this change. b I .a r 1 l e i a 4 4 6 4 4 I l l I i I * ,.-,,.-.e,-. ..,.. ~,,. _

- - ~ -.. 1 i I I i l TABLE 1 THikMAL AND HYDR.AULIC PARAMETERS FOR H M ERIALS COMPATIBILITY I TESTS _IN ASSEMBLY COSTAINING TWO 12.5 w/o U-2D ENRICHED ROD 5 5 i T I b N## i inlet Maximum Linear Power Density 16.1 kw/fe Maximum Surface Heat Flux 540,000 BTU /Hr. ft.2 i Hot Channel Factors Fq 2.43 FH 4.01 u Minimum DND ration 100% Power (2250 psia T 480'F) 2,. O i inlet 1127. Power (2075 psia T 490'F) 1.93 inlet l l Mean Clad Temperature at maximum surface heat flux 1 100% Power 711*F 112% Power 719'T Clad Stresses

• 10,000 psi Maximum i

i Clad Strain l Maximum <0.5% l l l Fuel Center Temperature 100% Power (16.1 kw/ft) 3750*F 112% Power (18.1 kv/f t) - 4120'F 1

• Pressure tensile stress hot, operating, end-of-life 1

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pcket Noe 50-146 OPR-4 Change Report No. 18 Page 1 of 2 Pages 4 Change Report No. 18 TEST FL'EL RODS - HYDRIDING ETTECTS 1 DESCRIPTION OF CHANGE Tour test fuel rods will be inserted in Saxton Core III peript,.a1 test subassemblics. Two of the roc's will contain fuel which has an everage D20 content of 25 ppm and the other two rods vill contain f uel that has an average D 0 content of 120 ppm. The four rods are clad with titcalo)-4 2 having a nominal thickness of 23 mils and contain uranium dioxide (UO2) pellets, uniformly enriched to 12.5 w/o U-23$, 89.$% theoretical density, and contain less than 10 ppm H O in the fuel. Further, each rod contains 2 two fluoride contaminated notches as illustrated in Figure 1. One notch is wcated 1.5 inches from the top of the rod and the other 13 inches from the bottom. Each notch is.0025 inches deep x,004 ir e.hes wide x.187 inches long (nominal) and is located on the surface of the clad bore. 2. PURPOSE OF CHANGE The purpose of the change is to assess the effects f certain combinations of fuel pellet moisture, clad bore imperfections, and fluoride contamination and to determine if these combinations promote hydriding. 3. SAFETY CONSIDERATI03 Parameters pertinent to safety are tabulated in Table.I. The rods will be operated only in those test subassemblies for which the following limits apply.

1) The maximum calculated operating stress in the notched areas is 1 22,000 psi, ii) The rod pes.k specific power is 1 16.1 kw/ft.

iii) The maximum pressure tensile stress at' areas other than a' notch is 1 7.000 psi,-hot, operating, end-of-life. ,7

Docket Wo. 50-146 hPR-4 Change Repo7t No. 18 Page 2 of 2 Pages e Approval to operate rods similar to these at a maximum calculated operat-ing stress of 22,000 psi has been previously given in Change No. 30. Fuel rods containing 30-7$ ppm moisture have previously been successfully I o~erated at Saxton, it is our opinion that this change does net represent any unreviewed safety question. 4 HEALTH AND SAFETY It is our conclusion that the health and safety of the public will not be endangered by this change. i e J I 7 6 P P i -r w -ww -w ww w-ww v<m-ar % -v-v y-mer-e-w e awy* wa-w w=w ww.-7e-- e' e W-eg srea m-w e em erwuet -- = whe+-ey**Tmmt< w e v *st-e+ -w* r w

• w^m --*W7-W-~r

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t I L i i j-TAETI I [ j. 1 t SAFETY PARAK6TERS FOR EXPER. MENTAL HYDRfDE RODS i 4 Roo Operatint

• l Peak Initial Fill Average Stress l

Power Pressure (cold) Gas D2O Ler21 kv/ft psia ppm psi t i f $16.1 815 80% He-20% Xe 25 <7000 $16.1 815 80% t'e-20% 7e 25 <7000 i r <16.1 915 He 120 <7000 l t <16.1 915 He 120 <7000 r ? l t i i i [

• Pressure tensile stress, hot, operating, end-of-life i

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octet No. $0-146

• 0PR-4 Change Report No. 19 Page 1 of 2 Pages Change Report No. 19 TEST TUEL RODS - CIAD CREEP BEHAVIOR 1.

DESCRIPTION or CHAN,GE Four test fuel rods will be inserted in Saxton Core 111 peripheral test subassemblies. Two of f.he test rods will be high pressure creep rods and two low pressure creep rods. The four toda are clad with Zircaloy-4 having a nominal thickness of 23 mils and contain uranium dioxide (U02) r pellets, unifarmly enriched to 12.5 w/o U-2'$, having less than 30 ppm a H O and are helium filled. 'Ihe high pressure crsep rods contain 94% 2 theoretical density fuel 1,ellets and have a 10 inch gas plenum where as the low pressure creep rods contain 89.5% theoretical density fuel pel-lets and have a 2 inch gat plenum. 2 PURPOSE OT CHANGE The purpoco of the change is to assess the effect of irradiation on clad creep behavior. 3. SAFETY CONSIDERATIONS Parameters pertinent to safety are given in Table 1. The maximum calculated i operating stress in the rods is 16,000 psi which is much less than pre-viously approved stress levels. The stress level in the low pressure creep rods was not determinea since it will be much less than that in the high j pressure creep rods, which is well within acceptable limits. Rods similar to the low pressure creep rods were previously approved in Change No. 22, 24 and 29. The high pressure creep rods will be similar to rods previously approved in Change No. 30, the principle difference being that the high i j pressure creep rods will operate at a 6,000 psi lower stress level than those of Change No. 30. l It is our opinion that this change does not represent any unreviewed safety question. ..,.,f-p--...,._.. ,3_..,,. r,- ya--, n .,yy._.,. .,..,..,,._y. ,,.v._,

.. ~.. .. ~ --.. _ _ - -. _ - _.. -..... -. - _...... - l ocket No. 50-146 DPR-4 I Change Report No. 19 l Page 2 of 2 Pages i I l l 4, llEA1.Til AND SAI'ETY It is our conclusion that the health and safety of the public will not l be endangered by this change. i s-l ) l 1 l i r I O I l t i o l l l l ' r I I i l 5 r \\, s k i l,.... _ _. -..,, - _ _ .. - -,...,. -.. ~.... _.. -,.. - -. ~.. - _ _ -. _ _ -.... _ _... - -, -..... - -,. -...,..,. -.,,,,. - _.... - _. -., _,,...

l TABLE I t SAFETT PARAMETERS FOR CREEP TEST RODS I Test Rod Type Peal Power Initial Pressure cold Operating Stress Level

• kw/ft psia psi l

I 1 i l 111gh Pressure Creep <16.1 1915 <1(,000 l i Low Pressure Cr-ep <16.1 290 negligible compressive stress 1 e i i. 4 l - i i

• Pressure tensile stress, hot, operating, end-of-life y

I 1 l

- - -. _. ~. - Docket No. 50-146 DPR-4 Change Report No. 20 Page 1 of 2 Pages Change Report No. 20 CORROSION TEST OF ZIRCALOY-4 SPOT WELDS AND BI-METALLIC SANDWICH 1. DESCRIPTION OF CHANCE Test sections containing sandwich spot velds as shown in Figure I will be suspended vertically in four removabic water filled tubes as shown in Tigure 2. The tubes containing the spot veld tests will replace four standard wa r tubes in two of the central nine Core 111 assemblies. r The location of the four removable water tubes which contain the tests are shown in the following table. Assembly Core Location Tube Location in Assembly * $03-17-8 C-2 E-1 and A-5 $03-17-9 E-4 E-1 and A-5 The following tabi.e is a tabulation of the parts shown in Figures 1 and 2. Part No. Title Material Figure 1 1 Top Eng Plug

• Zircaloy-4 2

Retainer Zircaloy-4 l 3 Pin Zircaloy-4 1 4 Retainer Zircaloy-4 5 Retainer Zircaloy-4 6 Retainer Zircaloy-4 7 Clip Stainless Steel 304L 8 Clip Inconel 718 Figure 2 9 Water Tube Zircaloy-4 10 Bottom End Plug Zircaloy-4

• See Figure 3 for explanation of assembly location designation.

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~___-_____ ___ _ _ _ ____.~ ____. Docket No. 50-146 4 DPR-4 j Change Report No. 20 Page 2 of 2 Pages e 2. PURPOSE OF CRANGE t The purpose of the change is to permit the evaluation of corrosion and hydriding effects on spot welded joints ia an operating pressurized water reactor environment. 3. SAFr.rt CONSIDERATIONS The tout sections containing bl metallic sandwich and spot welds will be suspended ve.rtically in the water tube by being welded to the upper end plug. Coolant access will be through 0.125 inch and 0.062 inch diameter bleed holes in the upper and lower sections of the tube respectively. The bleed holes are such that it is highly improbable that material large enough to be detrimental to the reactor operation could get from inside the tube into the coolant. It is our opinion that this change does not represent any unreviewed safety question. 4 HEALTil AND SAFEQ It is our conclusion that the health and safety of the public will not be endangered by this change. i i ) I ~ 1;... - -. =.

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Docket No. 50-146 DPR-4 Chonge Report No. 21 Page 1 of 2 Pages i Change Report No. 21 IRRADIATION Or llYDRIDED WATER TUBESS 1. DESCRIPTION OF CilANCE Forty-two. Saxton Loose Lattice, Zr-4, water filled tubes hydrided to various IcVels will be placed in fuel assemblies at core locations i D-2 and D-4 Seventeen of the tubes will be placed in the aosembly at D-4 and twenty-five in the assembly at D-2. Tables 1 and 2 give hydrogen content and location of the tubes within each assembly. 2. PURPOSE OF C} LANCE The irradiated, hydrided tubing will be used in a program to investi-gate the effects of high hydrogen content and irradiation on the l mechanical properties of Zirealoy-4 l 3. SAFETY CONSIDERATI0"1 The water tubes being inserted in the loose lattice assembly will have known hydrogen levels of 125-600 ppm at start-of-life. The hydrogen uptake during service is estimated at 25-35 ppm and may be ignored as a relevant factor. There is information that irradiation increases the yield and tensile strength of Zircaloy and reduces the ductility and impact strength. There is also information that hydrogen increases the strength of Zir-caloy-4 slightly with increasing levels up to 600-800 ppm at the ex-pense of some reduction in tensile ductility and (at low temperatures only) a decrease in impact strength.( ).There is little data on the cumulative effects of hydrogen and irradiation on mechanical properties but IrvinI ) has tested irradiated Zircaloy-2 containing up to 250 ppm hydrogen and found little additive effect as a result of the hydro-Wood, et al(3) found that hydrogen and irradiation.tt.ded to be l gen.

_ _ = _. - Docket No. $0-146 DPR-4 Change Report No. 21 i 1 Page 2 of 2 Pages additive in reducing impact strength at low hydrogen levels (<kOO ppm) but that the cumulative effect was less marked at levels up to 250 ppm.I3) Although the tubes may have relatively low ductility at the operat-ing temperatures and hydride levels of the experiment, their high strength makes it unlikely that failure will c at the level of stress encountered. The tubes will have particularly low impact strength at room tempera-ture; however, rod handling problems will be no worse than for a full irradiated fuel element. It is our opinion that this change does not represent any unreviewed safety question. 4 HEALTH AND SAFETY lt is our conclusion that the health and safety of the public will l not be endangered by this change. REFERENCES 1. Love, A. L., Jr. and Johnson, C. R., " Corrosion and Hydriding of Zir-caloy Task 2 Final Report", BAW-3765-6, March 1968. 2. Irvin, J. E., " Effects of Irradiation and Environment on the Mechani-cal Properties and Hydrogen Pickup of Zircaloy", Electrochemical Tech-nology, 4, May-June, 1966. 3. Wood, D. S., Winton, J., Watkins, B., " Effects of Irradiation on the Impact Properties of Hydrided Zircaloy-2 and Zirconium-Niobium Alloy", 4, May-June, 1966. . -. - - =

t l l l TABLE 1 l LOCATION OF HYDRIDED WATER FILLED TUBES IN SAXTON ASSEMBLY. 503 12 2 AT CORE LOCATION D-2, I HYDROGl.N CONTENT (ppm) ASSEMBLY LOCATION,* TUBE IDENTITY W24 125 200 B-1 i W76 125 - 200 F-1 l W27 400 - 500 E-2 W28 400 - 500 G-2 I W29 400 - 500 12 l W30 350 - 500 B-3 l W31 350 - 500 D.3 i W33 125 - 200 C-4 W34 400 - 500 E-4 W35 125 - 200 G-4 i W36 500 650 B-5 l W37 500 650 D-5 l W38 300 - 350 F-5 W39 450 - 600 C-6 W40 350 - 500 E-6 ( W41 450 600 G-6 W42 400 - 500 B-7 W43 400 - 500 D-7 W44 125 - 200 F-7 W45 501 - 650 H-7 W46 400 - 500 A-8 W47 300 - 350 C-8 W48 350 --500 E-8 W49 450 - 600 G-8 W50* 350 - 500 B-9 1 1

• 3ee Figure.'. for explanation of assembly location-designation-l l

i l l l

~ - - ~... - -. ___~..-.. TABLE 2 LOCATION OF HYDRIDED WATER FILLED TUBES IN f tXTON ASSEMBLY 503 17 6 AT COR.E LOCATION D-4 I TUBE IDENTITY HYDROGEN CONTENT (ppm) LSSEMBLY LOCATION

• l l

S21 400 500 D_1 i 522 400 - 500 F-1 S23 125 - 200 H-1 l S24 500 - 650 C-2 S25 500 - 650 E-2 S26 300 350 G-2 r,6 400 - 600 D-3 T. 8 350 - 500 F-3 S29 450 - 600 H-3 530 400 - 500 C4 S31 400 - 500 E-4. S32 125 - 200 G_4 I S33 500 - 650 D-5 S34 300 - 350 F-5 S35 300 - 350 ti-5 S36 350 - 500 C-6 S38 350 - 500 G-6 9

• See Figure 1 for explanation of assembly location designation l

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t Docket Nc. 50-146 DPR-4 Change Report No. 22 Page 1 of 2 Pages i i ( Change Report No. 22 l IRRADIATION OF ZlRCALOY-4 CLADDING TEST SPECIMENS I 1. DESCRIPTION OF CHANGE i Three specimen assemblies as shown in Figure 1, each containing six end capped cladding test specimens will be inserted in the central l nine Core III fuel assemblies. The specimen etl cap is shown in Figure 2. Two of the specimen assemblies will replace water filled tubes in two loose lattice fuel assemblies and one will replace a water filled tube in a load follow fuel assembly. The locations of the specimen assem-blies are shown in the table below. Specimen Assembly Location Fuel Assembly Core Location in Fuel Assembly

• 503-17-5 C-4 A-5 (Loose Lattice Assembly)

A-5 503-17-3 E-2 (Loose Lattice Assembly) 503-18-3 E-3 E-1 (Load Follow Assembly) The table below is a list of the parta and materials in Figure 1. ,Part No. Title Material 1 Cladding Test Specimen Zircaloy-4 -2 Top End Plug Zitcaloy-4 3 Bottom End Plug .Zircaloy-4 4 Specimen End Cap Zircaloy-4 l 5 Nickel Flux Wire Nickel l t l 2. PURPOSE OF CHANGE The purpose of the change is to permit irradiation of Zircaloy-4 l cladding for mechanical testing. l

• See Figure 3 for explanation of assembly location designation V

. -... ~ -.. -..

. Docket No. 50-146-DPR-4 Change Report No. 22 Page 2 of 2 Pages 3. SAFETY CONSIDERATIONS All welds will be inspected and required to meet the standards used for fuel element plug cap welds. No new materials or fabrication techniques are required. It is our opinion that this change does not represent any unreviewed safety question. 4 HEALTil-AND SAFETY It is our conclusion that the health and safety of the public will not be endangered by this change. 4 l M

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.chet No. 50-146 DPR-4 Change Report No. 23 Page 1 of 3 Pages Change Report No. 23 SOLID ZIRCALOY BARS 1 DESCRIPTION OF CHANGE Solid removable Zircaloy bars as illustrated in Figure I will be inserted in Saxton Core 111 in the fuel assemblies shown in the following table. Fuel Assembly Core Location. 503-18-1 (Load Follow Assembly) C-3 503-18-3 (Load Follow Assembly) E-3 503-10-7 (Unitradiated Peripheral UO2 Assembly) B-2 Theloadfollowassemblieswillconta,inacingleZircaloybarwhike the unirradiated uranium peripheral assembly will contain two Zircaloy bars. The location of the Zircaloy bars within each assem-bly is shown in Figure 2. 2. PURPOSE OF CHMGE The purpose of the change is to allow replacement of four fuel rods by solid Zircalcy bars to maintain thermal hydraulic safety margins ar.d minimize disturbances of power distribution in surrounding fuel rods. 3. SAFETY CONSIDERATIONS Two flow channels in each of the load follow and unitradiated uranium peripheral fuel assemblies have restricted coolant flow caused by full length angles spot welded to the inside of'the enclosure skin. (The angles as explained in Summary Report on Buckling of Saxton Core II Fuel Assemblies and Prevention of Buckling in Core III are used to in-crease the stiffness of the can.) The load follow assembly had a fuel rod at one restricted coolant channel and a water filled tube at the other, while the unirradiated uranium peripheral fuel assembly had fuel rods at both restricted coolant channels. I

Docket No. 50-346 DPR-4 -Change Report No. 23 Page 2 of 3 Pages Fuel rods at restricted coolant channels were replaced by solid Zirca]oy bars to maintain thermal-hydraulic safety margins and to r.in.isian the disturbance of the power distribution in surrounding fuel rods. The following table is a tabulation of the tharmal and hydraulic design parameters which still pertain to Saxton Core 711 operation. Total Core Total Heat Output 28.0 MWt 6 Total Hea? Output 93.56 x 10 Btu /hr Heat Generated in Fuel .97.4% System Pressure - Nominal 2250 paia System Pressure - Minimum - Steady State 2200 psia 6 Total Flow Rate

• 3.21 x 10 lb/hr 6

4tfective Flow Rate for Heat Transfer 2.73 x 10 lb/hr 2 Flov area for Heat Transfer Flow (unit cell) 2.2 ft Average Velocity Along Fuel Rods 6.83 ft/sec l l l Coolant Tecoeratures Neminal inlet _ 480 F. Maximum Inlet Including Instrument Errors and Deadband 485 F Average Rise in Vessel 26.0 F Average Rise in Corc -30.5 F Average in Vessel 493.0 F l l Average in Core 495.2 F l Heat Transfer Active Heat Transfer Surface Area of Fuel. Rods 376.2 ft' 2 Average Heat Flux 220,400 Btu /hr-ft Averaga Thermal Output 6.62 kw/ft Maximum Clad Surface Temperature at Nominal Pressure 657.4 F At 63 cycles t 5 y. ,......... _,.,,,.. _..,, _., ~ ~,..., -.,. - -. -...-..-.m. ~

_.~.-_________ I ket No. 50-146 DPR-4 Change Report No. 23 Page 3 of Pages 4 Loose Load Lattice Follow Assembly Assembly-UO Center Core Region UO -Pu02 2 2 Fq Feat Flux Hot Channel Factor 3.65 3.01 F Nuclear Radial Factor 2.62 2.10 F Enthalpy Rise Hot Channel oH Factor 2.94 2.35 f Nuclear Axial Factor 1.33 1.38 F, Nominal Outlet Enthalpy Hot Channel 528.3 Btu /lb 540.9 Btu /lb Saturation Enthalpy at Minimum Steady State Pressure 695.0 Btu /lb 695.0 Etu/lb 2 Maximum Heat Flux 799,800 Btu /hr-ft 662,300 Bru/hr-ft Maximum Thermal Output 24.0 kw/ft 19.9 kw/ft W-3 DNB Ratio at 100% Power i i Nominal Conditions 2.0 1.75 It is our opinion that this change does not represent any unreviewed safety question. 4.- HEALTH AND SAFETY l-It is our conclusion that the health and-safety of the public will not l be endangered by this change.

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_ _..... ~. _ _ _ _,. _ _ _. _ _ _ SKETCH $HEET tomu 28677 W ELTlHGHOUS E EL ECT nlC CO RP OR AT..ON Location of Zircaloy Bar in Assem-Unitradiated Peripheral UO3 Full length an,gle as shown in igure 5 of Reference'(1) bly and Load Follow AssembIy F Isso 0 6 ; C 00 0 0000 0 O'O O O O O O 1DOO O O O O O C O O 0 Q 0 O O n l D O O O O O D O O O O OO O O O O O O O 00 O O O O C 0 0 0 D0000% z x' Full length angle as shown in Location of Zircaloy Bar in Figure 5 ' of Ref erence (1) Unitradiated Peripheral UO2 Assembly-LOCATIONS'0F SOLID ZIRCALOY BARS WITHIN ASSEM FIGURE 2 l tion (1) Sumary Report on Buckling of Saxton Core 11 Fuel Assemblie of Buckling in Core III. n.-,, ,,,.-.,,.,e,. .,.w .,-,.,.,,-.,,,,..-._,,.,,,,..,n,., ,w. +.,.n.,,, ,.n ,.-.,n .,.,, -,,w, w,,-e

= - -. - i ..~ Lks - 4: 3 gl. / RE: SAXTON FJCLEAT EXPERDENTAL CORFORNZION / R. J. Scheinel, DRL DOCET NO. 50-146 Dt. sed on difficulties that ve. nave experienced with Saxton eeneerming the availability of I minutes of their Safety Review Committee beetings, it is sweemmended that a phrase similar to the fouoving be tacorporated in the Technical Spec-itientions during the next 1.toansing actions

• atinutas of Mafety Deview Committee meetings i

shall be maintained.* cc: $ C. Moseley, 00:I R. W. Waodntff, DHL d Oni;ini Sf J F. O*5'D J. P. O'Heilly, CC 7hal 4/24/69 1 1 l PRS "'1959 ' i c n. I + y y ,,v---v}}