ML032820022
| ML032820022 | |
| Person / Time | |
|---|---|
| Site: | Mcguire, McGuire  |
| Issue date: | 09/29/2003 |
| From: | Gordon Peterson Duke Power Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| Download: ML032820022 (156) | |
Text
{{#Wiki_filter:Duke GARY R. PETERSON cPowere Vice President McGuire Nuclear Station A Duke Energy Company Duke Power MGO1 VP / 12700 Hagers Ferry Road Huntersville, NC 28078-9340 704 875 5333 704 875 4809 fax grpe tersoduke-energy. corn September 29, 2003 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555-0001
SUBJECT:
Duke Energy Corporation (Duke) McGuire Nuclear Station Units 1 and 2 Docket Nos. 50-369 / 50-370 Proposed Technical Specification (TS) Amendments TS 3.7.15 - Spent Fuel Assembly Storage, and TS 4.3 - Fuel Storage
Reference:
NRC letter to Duke dated January 31, 2003 Pursuant to 10 CFR 50.90 and 10 CFR 50.4, this letter submits a license amendment request (LAR) for the McGuire Nuclear Station Facility Operating Licenses (FOL) and TSs. Duke met with the NRC in White Flint on December 10, 2002, March 6, 2003 and May 20, 2003 to discuss the basis for this LAR, and also Duke's corrective actions to address the spent fuel storage issues at McGuire. This LAR will change the McGuire TS 3.7.15 to provide revised spent fuel pool storage criteria based upon fuel type, fuel enrichment, burnup, cooling time and partial credit for soluble boron in the spent fuel pool water. This LAR also allows for the safe storage of fuel assemblies with a nominal enrichment of U-235 up to 5.00 weight percent. In addition, this LAR decreases the required soluble boron credit from 850 ppm boron to 800 ppm boron in McGuire TS 4.3.1, which continues to provide an acceptable margin of subcriticality in the McGuire spent fuel storage pools. This proposed amendment is applicable to Facility Operating Licenses NPF-9 and NPF-17 for the McGuire Nuclear Station. As discussed previously with the NRC Staff, Region 1 of both units has been reracked with new racks of equivalent design from Holtec. These new Region 1 racks utilize Boral neutron poison www. duke-energy. corn 4coD
U.S. Nuclear Regulatory Commission September 29, 2003 Page 2 material instead of Boraflex. Region 1 was reracked without prior NRC approval in accordance with the stipulations as set forth in 10 CFR 50.59. Region 2 will continue to use the Westinghouse racks that utilizes Boraflex neutron poison material. This submittal assumes full credit for the Boral neutron poison material for Region 1 and no credit for any remaining Boraflex in Region 2. Upon approval of this LAR, the commitment to perform "blackness testing" of the Boraflex panels as stipulated in Selected Licensees Commitment 16.9.24, "Spent Fuel Pool Storage Rack Poison Material', will no longer be performed. The letter referenced above issued a temporary exemption to 10 CFR 70.24, "Criticality Accident Requirements', which expires December 31, 2005. However, in accordance with 10 CFR 50.68, Duke will comply with the requirements of 50.68(b) in lieu of maintaining a monitoring system capable of detecting a criticality event as described in 10 CFR 70.24. Upon approval of this submittal an exemption to 10 CFR 70.24 will no longer be necessary. provides marked up pages of the existing McGuire TSs showing the proposed changes. Attachment 2 contains the new McGuire TS pages. The Description of Proposed Changes and Technical Justification is provided in Attachment 3. Pursuant to 10 CFR 50.92, Attachment 4 documents the determination that this proposed amendment contains no significant hazards considerations. Pursuant to 10 CFR 51.22 (c)(9), Attachment 5 provides the basis for the categorical exclusion from performing an Environmental Assessment or Impact Statement. A summary of the McGuire Spent Fuel Pool Criticality Analysis is shown in Attachment 6. Attachments 7 and 8 show the proposed and revised BASES for TS 3.7.14 and 3.7.15. Implementation of this amendment to the McGuire FOLs and TSs will impact the station's UFSAR. Consequently, upon approval of this LAR, the applicable revisions will be included in a McGuire UFSAR update. In accordance with Duke internal procedures and the Quality Assurance Program Topical Report, this proposed amendment has been previously reviewed and approved by the McGuire Station's Plant Operations Review Committee and the Duke Nuclear Safety Review Board. Pursuant to 10 CFR 50.91, a copy of this LAR is being forwarded to the appropriate North Carolina state officials.
U.S. Nuclear Regulatory Commission September 29, 2003 Page 3 Consequently, Duke requests approval of this LAR by October 1, 2004. As indicated in the attached "No Significant Hazards Consideration Evaluation" the proposed changes in this LAR will not result in a significant reduction in the facility's margin of safety. Please contact Norman T. Simms of Regulatory Compliance at 704-875-4685 with any questions regarding this LAR. Very truly yours, Gary R. Peterson Attachments xc: (w/attachments) L.A. Reyes Administrator, Region II U.S. Nuclear Regulatory Commission Atlanta Federal Center 61 Forsyth Street, SW, Suite 23T85 Atlanta, GA. 30303 J.B. Brady NRC Senior Resident Inspector McGuire Nuclear Station R.E. Martin, Project Manager (addressee only) Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission One White Flint North, Mail Stop O-8G9 11555 Rockville Pike Rockville, MD 20852-2738 Beverly 0. Hall, Section Chief Radiation Protection Section 1645 Mail Service Center Raleigh, N.C. 27699-1645 r
U.S. Nuclear Regulatory Commission September 29, 2003 Page 4 G.R. Peterson, being duly sworn, states that he is Vice President of McGuire Nuclear Station; that he is authorized on the part of Duke Energy Corporation to sign and file with the U.S. Nuclear Regulatory Commission these revisions to the McGuire Nuclear Station Facility Operating Licenses Nos. NPF-9 and NPF-17; and, that all statements and matters set forth therein are true and correct to the best of his knowledge. G.R. Peterson, Vice President McGuire Nuclear Station Duke Energy Corporation Subscribed and sworn to before me on je 2003. Notary Public My Commission Expires: 4 i Z.
U.S. Nuclear Regulatory Commission September 29, 2003 Page 5 bxc: (w/ attachments) S.W. Moser (MG05EE) K.L. Crane (MGO1RC) G.D. Gilbert (CNO1RC) L.E. Nicholson (ON03RC) P.F. Guill (MG05EE) J.I. Glenn (MG05EE) D.C. Jones (EC08F) J.P. Coletta (EC08F) M.R. Nichol (EC08F) G.A. Copp (EC09A) C.J. Thomas (MGO1RC) N.T. Simms (MGO1RC) ELL (ECO50) NSRB Support Staff (EC05N) Masterfile 1.3.2.9
ATTACHMENT 1 MARKUPS TO THE MCGUIRE TECHNICAL SPECIFICATIONS
Spent Fuel Assembly Storage 3.7.15 3.7 PLANT SYSTEMS 3.7.15 Spent Fuel Assembly Storage LCO 3.7.15 The combination of initial enrichment, burnup and l Nirnahle Absnrber (IFRA) rmitd of each new or spent fuel assembly stored in the spent fuel pool storage racks shall be within the following configurations: a.
%e- e°
- b. ' in Reaion plool- in -:e 49- these oilits./
- 3. -s .e Fia. 7.15-3.of fuel Vhieh dnomot-tho - IE or Is ._
~~eorirdao fuelinhbhhscldpry-dat.ht ps~16 dga'.' max'g he se O8.7. 1 :7; er-.
-9 Rerstriclted store in~~rdnanc ~ Rrpt rp 'A7115-4. of f1 iA-I whirh meets the criteriq of ghipI~ il.5~ -8: or E3. Ci-he rbeard St'^age in eccerdatec with Fir 3.7.15 57 Q-4 a bc des th i nuotuierthll rito of Teble 3.7.15 8.
McGuire Units I and 2 3.7.15-1 Amendment Nos. -494q8-
Spent Fuel Assembly Storage 3.7.15 stred in Region 9P nf th pen fue inav~ Ioe ance with thlese 1 nI ztrictc"d ctorao of fuo! meetingI .ea crlreria of l aite 3.7.15t-O;-or
- 2. flztrie ctragz in ccordanc with Figur3 3.7.16 6, of fuet which meets the criteria of Tahle 3 7 15-1 1 r
-V. ~ J ~I~ UJIU0.J I '.A.J Q
-.-. - -I fu~el vvc does notA rneet hezeOrita of Table 3.7.15 11 APPUCABILITY:- Whenever any fuel assembly is stored in the spent fuel pool.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Requirements of the A.1 -------- NOTE------ LCO not met. LCO 3.0.3 is not applicable. Initiate action to move the Immediately noncomplying fuel assembly to the correct location. SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.15.1 Verify by administrative means the planned spent fuel Prior to storing the pool location is acceptable for the fuel assembly being fuel assembly in stored. the spent fuel pool McGuire Units 1 and 2 3.7.15-2 Amendment Nos. 99+I8-
(an New or irradiated fuel may be allowed for unrestricted storage in Region 1 of the spent fuel pool provided the maximum initial U-235 enrichment of the fuel is < 5.0 weight percent; or New or irradiated fuel which has decayed at least 16 days may be stored in Region 2 of the spent fuel pool in accordance with these limits:
- 1. Unrestricted storage of fuel meeting the criteria of Table 3.7.15-1; or
- 2. Restricted storage in accordance with Figure 3.7.15-1 of fuel meeting the criteria of Table 3.7.15-2 (Restricted Fuel assembly) and Table 3.7.15-3 (Filler Fuel assembly); or
- 3. Checkerboard storage in accordance with Figure 3.7.15-2 of fuel meeting the criteria of Table 3.7.15-4.
Delete all existing Tables 3.7.15-1 through 3.7.15-12 and replace with new Tables 3.7.15-1 through 3.7.15-4 (see Attachment 2) Delete all existing Figures 3.7.15-1 through 3.7.15-7 and replace with new Figures 3.7.15-1 through 3.7.15-2 (see Attachment 2)
e -I I r Spent Fuel Assembly Storage DQe eP Thks Table 3.7.15-1 (page 1 of 1) Th419e 3.7.15 Minimum Qualifying Number of IFBA Rods Versus Initial Enrichment for Unrestricted Region 1A Storage of New Fuel Initial Nominal Enrichment (% U-235) Number of IFBA Rods 3.78 (or less) 0 4.22 16
\ 4.56 32 4.75 48 50 45 CO) 40 a
o 35
< 30 For LL 25 0
a 20 M 15 z in PTABLE Im 5 0 3.75 4.00 4.25 4.! 4.75 INITIAL NOMINAL ENRICHMENT, %U-235 NOTES: Fuel which iffers from those designs used to determine the requirements of Table 3. .5-1 may be alified for Unrestricted Region IA storage by means of an analysis using NRC approv methodology to assure that kens is less than 1.0 with no boron and less than or eqIal to 0.9 with credit for soluble boron. McGuire Units 1 and 2 3.7.15-3 Amendment Nos. 4l9, ,1.
Spent Fuel Assembly Storage
-e e 1- em; SUs 6 3.7.15
(>
.- It Table 3.7.15-2 (page 1 of 1) / 11 r.
I i Minimum Qualifying Burnup Versus Initial Enrichment for Unrestricted Region 1A Storage Initial Nominal Enrichment Asse mbly Burnup (% U-235) (G WD/MTU) 3.78 (or less) 0 4.00 1.58 4.50 4.92 4.75 1U 9-7- 6-age 3-I 5. C,) 4-U) 3-2- UNACCEPTABLE For Unrestricted Storage I1-n I 3.50 3.75 4.00 4.25 \ 4.50 4.75 INITIAL NOMINAL ENRICHMENT, % -235 NOTES: Fuel which diff rs from those designs used to determine the requirements olTable 3.7.15-2 may be qualifed for Unrestricted Region lA storage by means of an analysis ing NRC approved thodology to assure that kent is less than 1.0 with no boron and less an or equal to 0.95 wi credit for soluble boron. Likewise, previously unanalyzed fuel up to a minal 4.75 weightt/ZJ-235 may be qualified for Restricted Region 1A storage by means of an a lysis using 0C approved methodology to assure that ken is less than 1.0 with no boron an ess than equal to 0.95 with credit for soluble boron. cGuire Units 1 and 2 3.7.15-4 Amendment Nos. 1-9',4
Spent Fuel Assembly Storage a r l 3.7.15
'Tabfe 3.7.15-3(page1 of 1)
Minimum Qualifying Burnup Versus Initial Enrichment for Region 1A Filler Assemblies Initial Nominal Enrichment Assembly Bui (% U-235) 1.76 (or less) 2.00 2.50 3.00 3.50 4.00 50 45 5-I- 40 For Use As Filler A emb 0 35 CD CL 30u Z a: 25 m 20 C 15 UNACC For Use As V)10 5 0 4-2.00 2.50 3.00 3.50 4.00 INITIAL NOMINAL ENRICHMENT, %U-235 NOTE' I: Fuel wI -Iic differs from those designs used to determine the requirements of Table 3.7.15-3\ may bEL.Kalified for use as a Region 1A Filler Assembly by means of an analysis using NRC approv d methodology to assure that ken is less than 1.0 with no boron and less than or equal with credit for soluble boron. 3.7.15-5 Amendment Nos. 4-947,18-
ipeit rucl MbbumIiypi orage 3.7.15 Table 3.7.15-4 (page 1 of 1) Minimum Qualifying Bumup Versus Initial Enrichment for Unrestricted Region 1B Storage Initial Nominal Enrichment Assemblyurnup (% U-235) GW TU 1.78 (or less) 0 2.00 3.96 2.50 11.35 3.00 17.61 3.50 23.35 4.00 28.86 4.50 34.10
\.75 /36-6 450 45
~40-Bto 35 B ACC<EAto o7srce 35 < For Un stSo ge 730 Z 25 20 20 W15-ACCEPTABLE 10 For rsicted Storage 5-2.00 2.50 3.00 3.50 4.00 4.50 INITIAL NOMINAL ENRICHMENT, %U-235 NOTES:
Fuel w ch differs from those designs used to determine the requirements of Table 7.15-4 may b qualified for Unrestricted Region 1B storage by means of an analysis using N C appr ed methodology to assure that ket is less than 1.0 with no boron and less than or qual to 0/5 with credit for soluble boron. iuire Units 1 and 2 3.7.1 5-6 Amendment Nos.
Spent Fuel Assembly Storage A lI 1 3.7.15 Table 3.7.15-5 (pagi e1 of 1) Minimui mQualifying Burnup Vers ,us Initial Enrichment for Restricted Region 1B Sto rage with Fillers Initial N ominal Enrichment Assembly Burnup (% U-235) (GWD/MTU) 2.20 (or less) 0 2.50 3.91 3.00 9.65 3.50 15.0 4.00 1 7 4.50 .68 4.75 27.01 en u3.
/ ? 25 ACCEPT 3 20 For Restrictec a-1 a:
Z 15 co L; 10- UNA( UN( CEPTABLE Fr Re stricted Storage in 5) CO) 5 . n _- ,
- I
. .*~~~~~~~~ . . . . . . . . . I 2.00 2.50 3.00 3.50 4. 4.50 INITIAL NOMINAL ENRICHMENT, %U-NOTES:
Fuel whic differs from those designs used to determine the requirements of Table .7.15-5 may be ualified for Restricted Region 1B Storage by means of an analysis using NR appro d methodology to assure that ket is less than 1.0 with no boron and less than ox qual to 0. owith credit for soluble boron. uire Units 1 and 2 3.7.15-7 Amendment Nos.
I- SDent Fuel Assembly Storage 3.7.15
)Te 3..5 Table 3.7.15-6 (page of 1) f 1) Qle Minimum Qual itying Bumup Vers us Initial Enrichment for IRegion 1B Filler Asssemblies Initial Nominal Enrichment Assembly Burnup
(% U-2 (35) (GWD/MTU) 1.444(or less) 0 2.OC 12.68 2.5C 20.17 3.OC 27.03 3.5C 33.35 4.OC 39.33 4.5C 45.0 4.7 47. 9
*1 Du
/
45 - ACCEP ITABL\ E 40- For Use As Filler Assemt 0L 350 Z 25-m 20- UNAC TABLE For Use As Filler ssembly im 15 15 c) 10 - 5-
^ I v - - IX. . , v . .
2.00 .50 3.00 3.50 4.00 4 INITIAL NOMINAL ENRICHMENT, %U-235 NOTES: Fuel which 1ffers from those designs used to determine the requirements of Table 3.7.15-may be q lified for use as a Region 1B Filler Assembly by means of an analysis using NRC approve methodology to assure that keff is less than 1.0 with no boron and less than or equal to 0.95/vith credit for soluble boron. Mc ire Units 1 and 2 3.7.15-8 Amendment Nos. 497+16-
Spent Fuel Assembly Storage e, t ' S I""- cble 3.7.15 Table 3.7.15-7 (page 1 of 1) Minimum Qualifying Bumup Versus Initial Enrichment for Unrestricted Region 2A Storage Initial Nominal Enrichment Assembly Bumup (% U-235) fGWD/MTU) 1.50 (or less) 0.00 2.00 10.50 2.50 17.97 3.00 3.50 4.00 4.50 .71 4.75 60 1- 50 A CD 40 For Unrestricted torage 0-Q Z 30 i 20
?2O a: Ul w
CO l 0 4-2.00 2.50 3.00 3.50 4.00 INITIAL NOMINAL ENRICHMENT, %U-235 NOTES: Fuel which Hfers from those designs used to determine the requirements of Table 3.7.15 lified for Unrestricted Region 2A storage by means of an analysis using NRC \ ethodology to assure that ketf is less than 1.0 with no boron and less than or equal credit for soluble boron. ire Units 1 and 2 3.7.15-9 Amendment Nos. 4
Spent Fuel Assembly Storage
.e- 1yvs l0 )3.7.15 Table 3.7.15-8 (page 1 of 1)
Minimum Qualifying Burndu ) Verstis Initial Enrichment for Restricted Region v2A Stoi rage with Fillers Initial Nominal Enrichment Assembly Bumup (% U-235) (GWD/MTU) 1.80 (or less) 0.00 2.00 3.70 2.50 10.30 3.00 16.10 3.50 21.70 4.00 27.00 4.50 32.1 4.75 34 AX 1
'4U 35 30 ACCEPTABLE For Restricted Storage 0~ 25 z 20 cc 15 UNA EPTABLE -J 102 For Restn d Storage 10 C,)
CO) 5 n 1-2.00 2.50 3.00 3.50 4.00 .50 I INITIAL NOMINAL ENRICHMENT, %U-235 NOTES: Fuel whic iffers from those designs used to determine the requirements of Table 3.7.15-8 may be .alified for Restricted Region 2A Storage by means of an analysis using NRC approv d methodology to assure that ke is less than 1.0 with no boron and less than or equal to 0. with credit for soluble boron. Guire Units 1 and 2 3.7.15-10 Amendment Nos. @ LO ' i94
Spent Fuel Assembly Storage X, -el lliS tvt IV t< 3.7.15 Tabli l 3.7.15-9 (pagea1 of 1) Minimum Qualifyi ng Bumup Versi us Initial Enrichment for Re! gion 2A Filler As,semblies Initial Nominal Enirichment Assembly Bumup (% U-235 (GWD/MTU) 1.15 (Cor less) 0.00 2.00 20.00
\ 2.50 27.80
\3.00 34.60 3.50 41.10 4.00 47.20 4.50 53.10 4 75 55.90 Du N% I I- ACCEPTABLE \
0 . For Use As Filler AssemblyN 0~ 40. z 30. cc UL PTABLE
-J 20- For Use As r Assembly CO, 10-CO
^ I
{J *_ V- - -. -.o - . .I I, I, . .1 .I I.T . . 2.00 p.50 3.00 3.50 4.00 .50 INITIAL NOMINAL ENRICHMENT, %U-235 NOTES: Fuel which differs from those designs used to determine the requirements of Table 3.7.15-9 may be g~alified for use as a Region 2A Filler Assembly by means of an analysis using NRC approved methodology to assure that kets is less than 1.0 with no boron and less than or equal to 0.9, with credit for soluble boron. M uire Units 1 and 2 3.7.15-11 Amendment Nos. -219 9 191
Spent Fuel / kssembly Storage 3.7.15 toil( (, 7A-,- I-, En-/
\'I) I at s l - A Tabl e 3.7.15-10 (page 1 of 1)/
Minimum Qualif) yng Burnup' Versus Initial Enrichment for Unrn3stricted Rel gion 2B Storage Initial Nominal Elnnnchment Assembly Burnup (% U-23 5 ) (GWD/MTU) 1.11 ('for less) 0 2.00 21.58 2.50 29.00 3.00 35.6 3.50 41 7 4.00 .90 4.50 3.57 4.75 56.33 "IX-f vu 0- ACCEPTABLE For Unrestricted Stora a-I Z 5 0-0- 1v I- UN PTABLE 0 For Unrestri Storage 0 0 4- . . . . . . . . . . . . . . . . . . . . 2.00 2.50 3.00 3.50 4.00 4.50 INITIAL NOMINAL ENRICHMENT, %U-235 NOTES: Fuel whic differs from those designs used to determine the requirements of Table 3.7.110 may be ualified for Unrestricted Region 2B storage by means of an analysis using NRC approv d methodology to assure that keys is less than 1.0 with no boron and less than or equa to 0.9 with credit for soluble boron. M uire Units 1 and 2 3.7.15-12 Amendment Nos. 4948-'
Spent Fuel Assembly Storage r IN- IK-f fN -t. ~-+- I J.1.1 5
\j-eVer Table 3.7.15-1 1tj(pagi S lablk V 1 of 1) Iw-Minimum Qualifying Bumul 3Versi isInitial Enrichment for Restricted Region ;2B Stoirage with Fillers Initial Nominal Enrichment Assembly Burnup
(% U-235) (GWD/MTU) 1.22 (or less) 0 2.00 17.55 2.50 24.73 3.00 31.31 3.50 37.40 4.00 43.15 4.50 48.6 4.75 513 Dul 50 - ACCEPTABLE 0D For Restricted Storage z 40 - cr
-J3. 30 -
C,) a) UNA CEPTABLE 20 - For Re ed Storage 10 0 , . . . . . 2.00 2.50 3.00 3.50 4.00 4.50 INITIAL NOMINAL ENRICHMENT, %U-23 NOTES: Fuel which iffers from those designs used to determine the requirements of Table 3. .15-11 may be q lified for Restricted Region 2B Storage by means of an analysis using NR approvedmethodology to assure that keff is less than 1 0 with no boron and less than or ual to 0.95 ith credit for soluble boron. Mc ire Units 1 and 2 3.7.15-13 Amendment Nos.
Spent Fuel Assembly Storage 3.7.15 Table 3.7.15-12 (page 1 of 1) Minimum Qualifying Burnup Versus Initial Enrichment for Re gion 2B Filler Assemblies Initial Nominal Er irichment Assembly Burnup (% U-2355) (GWD/MTU) 1.08 (cor less) 0 2.00 23.14 2.50 30.59 3.00 37. 3.50 4.00 9.72 4.50 55.49 4.75 58.33 Du t 50 ACCEPTABILE For Use As Filler Assen D 40 EL D OF Z 30 UNA CEPTABLE For Use AsHier Assembly
"-i 20 I *m ci20 -1 w)l C21o 04 2.00 2.50 3.00 3.50 4.00 4.50 INITIAL NOMINAL ENRICHMENT, %U-235 NOTES:
Fuel ich differs from those designs used to determine the requirements of Table 3.7. -12 may e qualified for use as a Region 2B Filler Assembly by means of an analysis using C ap oved methodology to assure that keil is less than 1.0 with no boron and less than or equ t 0.95 with credit for soluble boron. cGuire Units 1 and 2 3.7.15-14 Amendment Nos.
Spent Fuel Assembly Storage 3.7.15 Lk~Lbe fk(-5 R-6u(-e-FILLER LOCAllON Restricted Fuel: Fuel ich does not meet the minimu umup requirements of either Ta he 3.7.15-1 or Table 3.7.15.2. (Fuelich does meet the r uirements of Table 3.7.15-1 or Table 3. 5:2, or non-fuel omponents, or an empty location may be pla d in restricted fuel locations as needed). Filler Location: Either fuel which meets the minimum bumup require ents of Table
- 3.7.15-3, or an empty cell.
Boundary Any Restricted Region 1A Storage Area row bounded by a other storage area shall contain a combination of restricted fuel ass rnblies and filler locations arranged such that no restricted fuel assembrn are adjacent to each other. Example: In the figure above, row 1 or c mn 1 can not be adjacent to another storage area, but row 4 or column can be. Figure 3.7.15-1 (page 1 of 1) Require(d 3 out of 4 Loading Pattern for Restricted Region 1A Storage Units 1 and. 2 3.7.15-15 Amendment Nos. 499 '/1, I \.
Spent Fuel Assemt bly Storage 3.7.15 P TW TSeF4\ L7(c'%e RESTRICTED l FUEL
- Il., t!2f8l l.hi1 iiRESTRICTED I FUE~Li Restricted Fuel: i meets the minimum bumup components, or an empty lo4 Filler Location: otither fuel which meets the minimum bumup of Table 3.7.15-6, or an empty cell.
Boundary Any Restricted Region 1B Storage Area must be s other storage area by at least one row of 1B Filler or empty cells, at all boundaries between storage regions. Figure 3.7.15-2 (page 1 of 1) Required 2 out of 4 Loading Pattern for Restricted Region 1B Storage a'uire Units 1 and 2 3.7.15-16 Amendment Nos. -4lO-&-49+
Spent Fuel Assembly Storage t-P2 ef - _# . V\ N IP
==== IN F-EMPTY CELL
,ffCKERBOARI EMPTY FUEL CELL l~~
- ' i , I .ii~
g ti -' EMPTY HECKERBOAR CELL FUEL I'li zitl~gw ga '.1j Checkerboard Fuel: Fy6I which does not meet the minimum bmup requirements of Table
.7.15-5. (Fuel which does meet the requ irments of Table 3.7.15-5, For non-fuel components, or an empty locatio may be placed in checkerboard fuel locations as needed)
Boundary Any Checkerboard Region 1B Storage Area n separated from any other storage area by at least one row of pells, at all boundaries between storage regions. I Figure 3,7.15-3 (page 1 of 1) Required 2 out of 4 Loading Pattern for Checkerboard Region 1B Storage Units 1 and 2 3.7. 15-17 Amendment Nos.
Spent Fuel Assembly Storage 3.7.15, Qe1 (-5 I FILLER FlLiER LOCAViON LOCATIOk RESTRICTE FUEL
- flZi,l-'"'Fi, RETICTED FUELJ Restricted Fuel: uel which meets the minimum bumt non-fuel components, or an empty Filler Location: I Either fuel which meets the minimum of Table
- 3.7.15-9, or an empty cell.
Boundary At least three of the four faces of each 2A Restrict must be adjacent to a 2A Filler Location, an empty wall, at all boundaries between storage regions. Figure 3.7.15-4 (page 1 of 1) Required 2 out of 4 Loading Pattern for Restricted Region 2A Storage McGuire Units 1 and 2 3.7.15-18 Amendment Nos.
Spent Fuel Assembly Storage 3.7.15 t Ue% r( S Utr- _ EMPTY CELL EMPTY CELL EMPTY CELL Checkerboard Fuel: fiuel which does not meet the minimum burp requirements of Table
/3.7.15-8. (Fuel which does meet the requirervnts of Table 3.7.15-8, or non-fuel components, or an-empty location nit be placed in
- checkerboard fuel locations as needed) \
Boundary At least three of the four faces of each Checkerboard Assembly must be adjacent to an empty cell or the pool wall, at between storage regions. Figure 3.7.15-5 (page 1 of 1) Required 2 out of 4 Loading Pattern for Checkerboard Region 2A Storage Units 1 and 2 3.7.15-19 Amendment Nos.
Spent Fuel Assembly Storage 3.7.15 ILER FIULER FILLER FILLER
-OCATION OCATION LOCATION OCATION r~~~~i~~~~
RESTRICTEDl FUE Restricted Fuel Fuel which meets the minimum bum quirements of Table 3.7.15-11, or non-fuel components, or an empty I tion. Filler Location: Either fuel which meets the minimum bumup quirements of Table 3.7.15-12, or an empty cell. Boundary lition: Any Restricted Region 2B Storage Area row boun d by any other storage area shall contain only filler locations arrang such that no Restricted Fuel assemblies are adjacent to any other el except Region 2B Filler Locations. Example: In the figure abod, row 1 or column 1 can not be adjacent to another storage area, bu row 4 or column 4 can be. Figure 3.7.15-6 (page 1 of 1) REzquired 1 out of 4 Loading Pattern for Restricted Region 2B Storage 'McGuire Units 1I and 2 3.7.15-20 Amendment Nos.
Spent Fuel Assembly Storage 3.7.15 ri~ef 1 IS _^ . EMPTY CELL EMPTY EMPTY CELL CELL
' F.."i
_,q
'a, HECKERBOARDI EMPTY FUEL CELL EMPTY EMPTY CELL CELL Checkerboard Fuel: Fu which does not meet the minim m burnup requirements of Table 3 .15-11. (Fuel which does meet thee equirements of Table 3.7.15-1, or non-fuel components, or an emp location may be placed in checkerboard fuel locations as needed)
Boundary Any Checkerboard Region 2B Storage Area r w bounded by any other
-- storage area shall contain only empty cells arra ed such that no Checkerboard Fuel assemblies are adjacent to a fuel. Example: In the figure above, row 1 or column 1 can not be adj cent to another storage area, but row 4 or column 4 can be.
I Figure 3.7.15-7 (page 1 of 1) Required 1 out of 4 Loading Pattern for Checkerboard IcGuire Units 1 and 2 3.7.15-21 Amendment Nos.
Design Features 4.0 4.0 DESIGN FEATURES 4.1 Site Location The McGuire Nuclear Station site is located at latitude 35 degrees, 25 minutes, 59 seconds north and longitude 80 degrees, 56 minutes, 55 seconds west. The Universal Transverse Mercator Grid Coordinates are E 504, 669, 256, and N 3, 920, 870, 471. The site is in northwestern Mecklenburg County, North Carolina, 17 miles north-northwest of Charlotte, North Carolina. 4.2 Reactor Core 4.2.1 Fuel Assemblies The reactor shall contain 193 fuel assemblies. Each assembly shall consist of a matrix of Zircalloy fuel rods with an initial composition of natural or slightly enriched uranium dioxide (UO2) as fuel material. Limited substitutions of zirconium alloy or stainless steel filler rods for fuel rods, in accordance with approved applications of fuel rod configurations, may be used. Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff approved codes and methods and shown by tests or analyses to comply with all fuel safety design bases. A limited number of lead test assemblies that have not completed representative testing may be placed in nonlimiting core regions. 4.2.2 Control Rod Assemblies The reactor core shall contain 53 control rod assemblies. The control material shall be silver indium cadmium (Unit 1) silver indium cadmium and boron carbide (Unit 2) as approved by the NRC. 4.3 Fuel Storage 4.3.1 Criticality 4.3.1.1 The spent fuel storage racks are designed and shall be maintained with:
- a. Fuel assemblies having a maximum nominal U-235 enrichment of 49 weight percent; S.oo
- b. keff < 1.0 if fully flooded with unborated water, which includes an allowance for uncertainties as described in Section 9.1 of the UFSAR; VO0
- c. kef < 0.95 if fully flooded with water borated tongO ppm, which includes an allowance for uncertainties as described in Section 9.1 of the UFSAR; McGuire Units 1 and 2 4.0-1 Amendment Nos. -24&-+191
Design Features 4.0 4.0 DESIGN FEATURES 4.3 Fuel Storage (continued)
- d. A nominal 10.4 inch center to center distance between fuel assemblies placed in Regionj 1A-ard-49; and
- e. A nominal 9.125 inch center to center distance between fuel assemblies placed in Region4 2A-ei42.
4.3.1.2 The new fuel storage racks are designed and shall be maintained with:
- a. Fuel assemblies having a maximum nominal U-235 enrichment of X-Qweight percent; 5.Doo
- b. krff
- 0.95 if fully flooded with unborated water, which includes an allowance for uncertainties as described in Section 9.1 of the UFSAR;
- c. kef
- 0.98 if moderated by aqueous foam, which includes an allowance for uncertainties as described in Section 9.1 of the UFSAR; and
- d. A nominal 21 inch center to center distance between fuel assemblies placed in the storage racks.
4.3.2 Drainage The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 745 ft.-7 in. 4.3.3 Capacitv The spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than 1463 fuel assemblies (286 total spaces in Region4 1A emH4Band 1177 total spaces in Region$ 2A-egd-222). McGuire Units I and 2 4.0-2 Amendment Nos. Q97448-
ATTACHMENT 2 REVISED MCGUIRE TECHNICAL SPECIFICATIONS
Spent Fuel Assembly Storage 3.7.15 3.7 PLANT SYSTEMS 3.7.15 Spent Fuel Assembly Storage LCO 3.7.15 The combination of initial enrichment, bumup and cooling time for each new or spent fuel assembly stored in the spent fuel pool storage racks shall be within the following configurations:
- a. New or irradiated fuel may be allowed for unrestricted storage in Region I of the spent fuel pool provided the maximum initial U-235 enrichment of the fuel is < 5.0 weight percent; or
- b. New or irradiated fuel which has decayed at least 16 days may be stored in Region 2 of the spent fuel pool in accordance with these limits:
- 1. Unrestricted storage of fuel meeting the criteria of Table 3.7.15-1; or
- 2. Restricted storage in accordance with Figure 3.7.15-1 of fuel meeting the criteria of Table 3.7.15-2 (Restricted Fuel assembly) and Table 3.7.15-3 (Filler Fuel assembly); or
- 3. Checkerboard storage in accordance with Figure 3.7.15-2 of fuel meeting the criteria of Table 3.7.15-4.
APPLICABILITY: Whenever any fuel assembly is stored in the spent fuel pool. ACTIONS _ CONDITION REQUIRED ACTION COMPLETION TIME A. Requirements of the A.1 ----------- NOTE-------------- LCO not met. LCO 3.0.3 is not applicable. Initiate action to move the Immediately noncomplying fuel assembly to the correct location. McGuire Units 1 and 2 3.7.1 5-1 Amendment Nos.
Spent Fuel Assembly Storage 3.7.15 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.15.1 Verify by administrative means the planned spent fuel Prior to storing the pool location is acceptable for the fuel assembly being fuel assembly in stored. the spent fuel pool McGuire Units 1 and 2 3.7.1 5-2 Amendment Nos.
Spent Fuel Assembly Storage 3.7.15 Table 3.7.15-1 (Page 1 of 7) Minimum Qualifying Burnup Versus Initial Enrichment and Cooldown Time For Unrestricted Region 2 Storage For Fuel Assembly Type MkBW Burnup (GWDJMTU) versus Initial Nominal Enrichment and Cooldown Time Cooldown Time Initia INominalErchmenth( )l (years) _ _2 32.50 SO-"2.001D 3.00 3.50 4 00 L 4.0 0 22.20 30.01 36.67 43.61 50.47 57.18 63.72 5 19.42 26.06 32.23 38.64 44.70 50.80 56.77 10 17.76 24.07 30.01 36.02 41.76 47.56 53.24 15 16.74 22.90 28.95 34.45 40.01 45.64 51.15 20 16.07 22.13 28.05 33.44 39.08 44.38 49.78 70.00 65.00 ACCEPTABLE 65.00 For Unrestricted Storage 60.00 45.00 _-= 4.00% =
- ~~ ~ 400 *-._3.50%/ U-235 -- --- ---- -----
3 00% U-235 2.50% U-235 X 25.00 _ 105.00 UNACCEPTABLE . .2.00% U-235 10.00 20.00 - , . . Unrestricted
. .Storage
. -. 3.5011/6 U-235For 5.00 0 5 10 15 20 COOLDOWN TIME (YEARS)
NOTES: Fuel which differs from those designs used to determine the requirements of Table 3.7.15-1 may be qualified for use as a Region 2 Unrestricted Assembly by means of an analysis using NRC approved methodology to assure that keff is less than 1.0 with no boron and less than or equal to 0.95 with credit for soluble boron. McGuire Units 1 and 2 3.7.15-3 Amendment Nos.
Spent Fuel Assembly Storage 3.7.15 Table 3.7.15-1 (Page 2 of 7) Minimum Qualifying Burnup Versus Initial Enrichment and Cooldown Time For Unrestricted Region 2 Storage For Fuel Assembly Type MkBWb1 Bumup (GWD/MTU) versus Initial Nominal Enrichment and Cooldown Time Cooldown Time Initial Nomina En ichment (% U-35 z.LW_ xi (years) 2.0-.0 350 40 450 .0 0 30.01 36.05 42.52 48.57 54.24 59.74 5 27.27 31.69 37.20 42.92 48.03 53.01 10 25.15 30.01 34.63 40.01 44.85 49.58 15 23.89 29.37 33.09 38.21 43.00 47.57 20 23.09 28.43 32.09 37.13 41.78 46.26 65.00: ACCEPTABLE 60.00 - For Unrestricted Storage 55.00 - D 50.00 45.00: 40.00: -. . 3.50% 4.00%/ U-235
. 3.0%U-235 --------
z 35.00 - 3.00% U-235 30.00 - EC M 25.00 - w 20.00 en 15.00 - Un 10.00 UNACCEPTABLE For Unrestricted Storage 5.00 0.00* 0 5 10 15 20 COOLDOWN TIME (YEARS) NOTES: Fuel which differs from those designs used to determine the requirements of Table 3.7.15-1 may be qualified for use as a Region 2 Unrestricted Assembly by means of an analysis using NRC approved methodology to assure that keff is less than 1.0 with no boron and less than or equal to 0.95 with credit for soluble boron. McGuire Units 1 and 2 3.7.15-4 Amendment Nos.
Spent Fuel Assembly Storage 3.7.15 Table 3.7.15-1 (Page 3 of 7) Minimum Qualifying Bumup Versus Initial Enrichment and Cooldown Time For Unrestricted Region 2 Storage For Fuel Assembly Type MkBWb2 Burnup (GWDIMTU) versus Initial Nominal Enrichment and Cooldown Time Cooldown Time . Ini;-a Nominal Enrh t (% 5) i (years) 3.00 3.50. 4.00 4.50 .. 0 37.51 44.11 49.95 55.50 60.90 5 33.02 38.34 43.97 48.98 53.87 10 30.72 35.73 40.95 45.67 50.30 15 30.01 34.15 38.99 43.72 48.22 20 29.62 33.12 37.88 42.47 46.85 65.00: ACCEPTABLE 60.00 ~~~~~~~~~~~~~For Unrestricted Storage
- 55.00 1--
50.00 0 3¢ 45.00
. 40.00
--~ _,_ 3.50%
~ U-235 -- -- -- -- _-- -- ..----------------
.00% U-235___ ___ ___
z 35.00
=
cc --..-..--.---. _._.__._3.00% U-235
= 30.00 >; 25.00 M 20.00 W
n i< 10.00 - UNACCEPTABLE For Unrestricted Storage 5.00 - 0.00 - 0 5 10 15 20 COOLDOWN TIME (YEARS) NOTES: Fuel which differs from those designs used to determine the requirements of Table 3.7.15-1 may be qualified for use as a Region 2 Unrestricted Assembly by means of an analysis using NRC approved methodology to assure that kff is less than 1.0 with no boron and less than or equal to 0.95 with credit for soluble boron. McGuire Units 1 and 2 3.7.15-5 Amendment Nos.
Spent Fuel Assembly Storage 3.7.15 Table 3.7.15-1 (Page 4 of 7) Minimum Qualifying Burnup Versus Initial Enrichment and Cooldown Time For Unrestricted Region 2 Storage For Fuel Assembly Type W-STD Burnup (GWD/MTU) versus Initial Nominal Enrichment and Cooldown Time Cooldown Time n (years) 2.00 -250 33.00' .50 .0. 4.50 5.0 0 20.02 28.59 35.83 43.37 50.67 57.75 64.63 5 18.50 25.10 31.56 38.35 44.97 51.39 57.66 10 17.14 23.29 30.01 35.78 42.03 48.12 54.08 15 16.32 22.21 28.83 34.24 40.29 46.19 51.97 20 15.79 21.51 27.96 33.24 39.16 44.94 50.61 70.00 65.00 - ACCEPTABLE 65.00 - For Unrestricted Storage 60.00-55.00
;! U-23s - - _ 5.00% U-235 5-__4.50%
_ 50500 - _ '3500% U-235 3.0- ---- _..O.................. U-235 4._ ._300% 30.00 0 ti25.00 _ 2.50%% U-235 2 20.00 ~~~~~~~~---- - 2.oohU-235 3015.00
< 00UNACCEPTABLE 10 10.00 ~~~For Unrestricted Storage 5.00 0 5 10 15 20 COOLDOWN TIME (YEARS)
NOTES: Fuel which differs from those designs used to determine the requirements of Table 3.7.15-1 may be qualified for use as a Region 2 Unrestricted Assembly by means of an analysis using NRC approved methodology to assure that kuf is less than 1.0 with no boron and less than or equal to 0.95 with credit for soluble boron. McGuire Units 1 and 2 3.7.15-6 Amendment Nos.
Spent Fuel Assembly Storage 3.7.15 Table 3.7.15-1 (Page 5 of 7) Minimum Qualifying Burnup Versus Initial Enrichment and Cooldown Time For Unrestricted Region 2 Storage For Fuel Assembly Type MkBI Bumup (GWD/MTU) versus Initial Nominal Enrichment and Cooldown Time Cooldown Time -Initl No inalEn ent (%l U-2:5 (years) 2.00 2.50 3.00 3.50 0 20.21 28.01 34.47 40.82 5 17.71 24.76 30.66 36.92 10 16.35 23.04 29.42 34.60 15 15.53 22.01 28.16 33.19 20 15.00 21.33 27.34 32.27 45.00 - ACCEPTABLE For Unrestricted Storage 40.00 - 0S. -. _._.50%
-- U-235 35.00 -
I B 30.00 -
. ~.
-_ _ _ . . __ - -. - 3.00% U-235
%0.
tL D 25.00: D 20.00 :
----- __--_ _-_-_--_-2.00%
-- 2.00% U-235 235--- --
M 15.00 - Cn 10.00-S UNACCEPTABLE 5.00 - For Unrestricted Storage 0.00 0 5 10 15 20 COOLDOWN TIME (YEARS) NOTES: Fuel which differs from those designs used to determine the requirements of Table 3.7.15-1 may be qualified for use as a Region 2 Unrestricted Assembly by means of an analysis using NRC approved methodology to assure that keff is less than 1.0 with no boron and less than or equal to 0.95 with credit for soluble boron. McGuire Units 1 and 2 3.7.15-7 Amendment Nos.
Spent Fuel Assembly Storage 3.7.15 Table 3.7.15-1 (Page 6 of 7) Minimum Qualifying Bumup Versus Initial Enrichment and Cooldown Time For Unrestricted Region 2 Storage For Fuel Assembly Type W-OFA Burnup (GWD/MTU) versus Initial Nominal Enrichment and Cooldown Time Cooldown Time Initial Nominal Enrichmt( U-23) ______ (years) 00 2.50 3.00 3.50 4.00 4.50 5.0 0 18.55 26.08 33.28 40.01 46.83 53.25 59.71 5 16.53 23.30 30.01 36.27 42.01 48.05 53.98 10 15.43 21.83 28.25 34.10 40.01 45.34 50.99 15 14.75 20.92 27.12 32.78 38.60 43.68 49.19 20 14.32 20.33 26.40 31.91 37.62 42.62 48.02 65.00 60.00 ACCEPTABLE 60.00 For Unrestricted Storage 55.00 _ 40.00 . -. - 4.00% U-235 50.00 E 4.50% U-235 345.00 _ (5 3 0.00 ------ -- 4.00% 3._ U-235 CL 250 20.00
.50 % U-235 2--
-2.00-25 ...... _._.3.00.% U.235 Mi 30-.00 - 3.00 % U-235 D 15.00 A; 0----_250%--3 --
CD1
'> 0 10.00 UNACCEPTABLE For Unrestricted Storage CD0.00 . - - - - - - - - - - - - - - - - - - -
5.00 0 5 10 15 20 COOLDOWN TIME (YEARS) NOTES: Fuel which differs from those designs used to determine the requirements of Table 3.7.15-1 may be qualified for use as a Region 2 Unrestricted Assembly by means of an analysis using NRC approved methodology to assure that keff is less than 1.0 with no boron and less than or equal to 0.95 with credit for soluble boron. McGuire Units 1 and 2 3.7.15-8 Amendment Nos.
Spent Fuel Assembly Storage 3.7.15 Table 3.7.15-1 (Page 7 of 7) Minimum Qualifying Burnup Versus Initial Enrichment and Cooldown Time For Unrestricted Region 2 Storage For Fuel Assembly Type W-RFA Burnup (GWD/MTU) versus Initial Nominal Enrichment and Cooldown Time Cooldown Time -;itial mi h (years) _ _ _ _ __ _ _ _ _ _ _ _ _
.5
.00 .00 4.50 .00 0 35.46 42.04 47.88 53.50 58.94 5 31.19 36.62 42.23 47.30 52.23 10 30.01 34.11 39.05 44.15 48.85 15 28.85 32.63 37.41 42.31 46.87 20 27.93 31.67 36.35 41.12 45.57 65.00 -
ACCEPTABLE 60.00 For Unrestricted Storage S. 55.00 I- 50.00 -
- U-235 5 -__4.50Y0 00% U-235 45.00 5
?-1 40.00 5 - _.- . - .3. 0% - - - -- 4.00 z ~~
- - - - - 3.50% -----
~~.... ............... .. . .
- 4. 00% U-235 35.00 5 9L 30.00 5 25.00 5 IC (0 20.00 -
15.00: 10.00 - UNACCEPTABLE For Unrestricted Storage 5.00: 0.00: . . . . . . 0 5 10 15 20 COOLDOWN TIME (YEARS) NOTES: Fuel which differs from those designs used to determine the requirements of Table 3.7.15-1 may be qualified for use as a Region 2 Unrestricted Assembly by means of an analysis using NRC approved methodology to assure that keff is less than 1.0 with no boron and less than or equal to 0.95 with credit for soluble boron. McGuire Units 1 and 2 3.7.15-9 Amendment Nos.
Spent Fuel Assembly Storage 3.7.15 Table 3.7.15-2 (Page 1 of 7) Minimum Qualifying Burnup Versus Initial Enrichment and Cooldown Time For Restricted Region 2 Storage For Fuel Assembly Type MkBW Burnup (GWDIMTU) versus Initial Nominal Enrichment and Cooldown Time Cooldown Time__ __ntl NominalEnri chment i ( 235) (years) 2.00 2.50 300 35 4.00 456 5 0 18.26 25.32 31.73 38.39 44.73 51.04 57.20 5 15.69 22.29 28.56 34.16 40.01 45.66 51.27 10 14.36 20.68 26.60 31.95 37.54 42.83 48.19 15 13.59 20.01 25.42 30.62 36.04 41.16 46.36 20 13.10 19.29 24.66 30.01 35.05 40.07 45.16 65.00 65.00 - ~~~~ACCEPTABLE 60.00 For Restricted Storage 55.00 BE50.00 - -_ .0 -3 D ~ ~ -_ =4.50% U-235 5.00% U-235 45.00 - - T- -- ,.40 -3 40.00
.0__
to 35.00 --.~~ -~ 3.50% _ _ _ _ _3_ _ U_-_ _2.00%6 U235
>. 325.00 0.00 - -- - - - - - - - -2.50%
- -UT235B M .00% U-235 3
Ml 15.00 - 10.00 UNACCEPTABLE For Restricted Storage 5.00 0.00 - , . , . , . . . . .. . . . . . . . 0 5 10 15 20 COOLDOWN TIME (YEARS) NOTES: Fuel which differs from those designs used to determine the requirements of Table 3.7.15-2 may be qualified for use as a Region 2 Restricted Assembly by means of an analysis using NRC approved methodology to assure that kff is less than 1.0 with no boron and less than or equal to 0.95 with credit for soluble boron. McGuire Units 1 and 2 3.7.15-10 Amendment Nos.
Spent Fuel Assembly Storage 3.7.15 Table 3.7.15-2 (Page 2 of 7) Minimum Qualifying Bumup Versus Initial Enrichment and Cooldown Time For Restricted Region 2 Storage For Fuel Assembly Type MkBWb1 Burnup (GWD/MTU) versus Initial Nominal Enrichment and Cooldown Time I Cooldown Time Initial NominalhEnrich; m (% U-235)_ (years) 250 3.00 3500 400, 4.50 5.00 0 26.29 31.14 37.02 43.12 48.55 53.83 5 23.07 28.91 32.89 38.20 43.29 48.07 10 21.37 26.88 30.73 35.78 40.54 45.07 15 20.35 25.66 30.01 34.30 38.82 43.31 20 19.66 24.86 29.76 33.36 37.77 42.16 60.00 55.00 ACCEPTABLE 55.00 - For Restricted Storage S 50.00 - 45.00 - --_450°h U-235 - -500 % U-235 3: 40.00 - - - - - - _- - - - - - - - - - - - - - - - -_ _ 4......U-235 4.00% CL 35.00 - - _-. __. 3.50% U-235 --..--.---.----
~
30.00 _ _ *-~--__._._._._.._ .......... m 25.00 _ _2350%U-235 .00 im 20.00 w 15.00 en
< 10.00 UNACCEPTABLE For Restricted Storage 5.00 0.00 , I 0 5 10 15 20 COOLDOWN TIME (YEARS)
NOTES: Fuel which differs from those designs used to determine the requirements of Table 3.7.15-2 may be qualified for use as a Region 2 Restricted Assembly by means of an analysis using NRC approved methodology to assure that kff is less than 1.0 with no boron and less than or equal to 0.95 with credit for soluble boron. McGuire Units 1and 2 3.7.15-11 Amendment Nos.
Spent Fuel Assembly Storage 3.7.15 Table 3.7.15-2 (Page 3 of 7) Minimum Qualifying Burnup Versus Initial Enrichment and Cooldown Time For Restricted Region 2 Storage For Fuel Assembly Type MkBWb2 Burnup (GWD/MTU) versus Initial Nominal Enrichment and Cooldown Time Cooldown Time NominaEnrment (% U235) -_.____ MIniial (years) 3.0 3i.5 4.00 4 5.00 0 32.33 38.06 44.25 49.61 54.82 5 29.93 33.86 38.94 44.09 48.80 10 27.89 31.65 36.48 41.24 45.69 15 26.66 30.32 34.99 39.40 43.85 20 25.85 30.01 34.01 38.34 42.67 60.00: ACCEPTABLE 55.00 For Restricted Storage n 50.00 R 45.00: ., ~~~~-__.50Syo U-235 0%U-3 40.00 EL 35.00
~ -- _
3.50%
.__O.5%
U-235 zc 30.00 _ _ _ ~~~~3.00%U-235 m 25.00 m 20.00 W 15.00
< 10.00 UNACCEPTABLE For Restricted Storage 5.00 0.00 0 5 10 15 20 COOLDOWN TIME (YEARS)
NOTES: Fuel which differs from those designs used to determine the requirements of Table 3.7.15-2 may be qualified for use as a Region 2 Restricted Assembly by means of an analysis using NRC approved methodology to assure that keff is less than 1.0 with no boron and less than or equal to 0.95 with credit for soluble boron. McGuire Units 1and 2 3.7.15-12 Amendment Nos.
Spent Fuel Assembly Storage 3.7.15 Table 3.7.15-2 (Page 4 of 7) Minimum Qualifying Bumup Versus Initial Enrichment and Cooldown Time For Restricted Region 2 Storage For Fuel Assembly Type W-STD Burnup (GWDIMTU) versus Initial Nominal Enrichment and Cooldown Time Cooldown Time 1-k In tia mina n ent (years) 2.0 2.50 .00 3.50 4.00 4.5 50 0 16.34 23.70 30.62 37.69 44.55 51.21 57.70 5 14.55 21.04 27.88 33.62 39.84 45.90 51.80 10 13.58 20.42 26.08 31.49 37.37 43.11 48.73 15 12.99 20.01 24.99 30.21 35.89 41.45 46.90 20 12.63 19.56 24.28 30.01 34.93 40.37 45.70 65.00 65.00 X ~~~~ACCEPTABLE 60.00 - For Restricted Storage
.. 55.00 50.00 ---- __-5.00%
U-235
- _4.50% U-235 40.00 -'-......... 235 3 - -- - 3.50% U-235 4.00% U-235 300
---- ----- ---- ---.- -- ------ ~ 3.00% U-235 m 20.00 co 15.00 2.00% U-235
< 10.00 UNACCEPTABLE For Restricted Storage 5.00 0.00 - . . . . . . . . . . . . . . . . . . .
0 5 10 15 20 COOLDOWN TIME (YEARS) NOTES: Fuel which differs from those designs used to determine the requirements of Table 3.7.15-2 may be qualified for use as a Region 2 Restricted Assembly by means of an analysis using NRC approved methodology to assure that kff is less than 1.0 with no boron and less than or equal to 0.95 with credit for soluble boron. McGuire Units 1and 2 3.7.15-13 Amendment Nos.
Spent Fuel Assembly Storage 3.7.15 Table 3.7.15-2 (Page 5 of 7) Minimum Qualifying Burnup Versus Initial Enrichment and Cooldown Time For Restricted Region 2 Storage For Fuel Assembly Type MkBI Burnup (GWD/MTU) versus Initial Nominal Enrichment and Cooldown I Time Cooldown Time Initial Nominal Enichment(% NS-235)E (years) 2.00 2.50 3.00 3.50 0 16.13 23.62 30.01 36.37 5 14.26 21.08 27.43 32.71 10 13.27 19.71 25.72 30.74 15 12.67 18.87 24.67 30.01 20 12.30 18.33 23.99 29.52 40.00 - ACCEPTABLE For Restricted Storage 35.00 S. -. _ _ =.- 3.50% U-235 30.00 ---. ------ -------------.--. 3¢ ~ ~~~~~~~~~~~3.00% -- - --- U-235 25.00 Do~~~~~~~~25% U-235 z Z 20.00
>; 15.00 - -- -- -- -- - - - - -- - - - - - - - - - - - - - -2.00% U-235 w 10.00 -
UNACCEPTABLE 5.00- For Restricted Storage 0.00 0 5 10 15 20 COOLDOWN TIME (YEARS) NOTES: Fuel which differs from those designs used to determine the requirements of Table 3.7.15-2 may be qualified for use as a Region 2 Restricted Assembly by means of an analysis using NRC approved methodology to assure that kff is less than 1.0 with no boron and less than or equal to 0.95 with credit for soluble boron. McGuire Units 1 and 2 3.7.15-14 Amendment Nos.
Spent Fuel Assembly Storage 3.7.15 Table 3.7.15-2 (Page 6 of 7) Minimum Qualifying Burnup Versus Initial Enrichment and Cooldown Time For Restricted Region 2 Storage For Fuel Assembly Type W-OFA Burnup (GWD/MTU) versus Initial Nominal Enrichment and Cooldown Time Cooldown Time I niti omina al Enrihmet (% U-2 _____ (years) 2C)0 2.50 .3.00 3.5Q 40 .50 5.00 0 14.85 22.04 29.10 35.62 41.63 47.88 54.01 5 13.38 20.01 26.26 32.15 38.13 43.42 49.04 10 12.53 19.03 24.74 30.33 36.01 41.04 46.41 15 12.00 18.29 23.81 29.54 34.72 40.01 44.82 20 11.67 17.82 23.20 28.80 33.87 39.17 43.77 60.00: ACCEPTABLE 55.00 For Restricted Storage 5 50.00 45.00 = ~-_
~_4350% U-235 40.00
-. _ 0% U-235 ' 4.00% U-235
- 0. 35.00 3.00% U-235 zc 30.00
- ------- ~---- --. _.__ 3.00% U-235 m 25.00
~~
- - - ~2.50% U-235 m 20.00 w 15.00
~~~~~- 2.00% U-235 i 10.00 UNACCEPTABLE For Restricted Storage 5.00 0.00 . . . .
I 0 5 10 15 20 COOLDOWN TIME (YEARS) NOTES: Fuel which differs from those designs used to determine the requirements of Table 3.7.15-2 may be qualified for use as a Region 2 Restricted Assembly by means of an analysis using NRC approved methodology to assure that kcff is less than 1.0 with no boron and less than or equal to 0.95 with credit for soluble boron. McGuire Units 1and 2 3.7.15-15 Amendment Nos.
Spent Fuel Assembly Storage 3.7.15 Table 3.7.15-2 (Page 7 of 7) Minimum Qualifying Burnup Versus Initial Enrichment and Cooldown Time For Restricted Region 2 Storage For Fuel Assembly Type W-RFA Burnup (GWD/MTU) versus Initial Nominal Enrichment and Cooldown Time Cooldown Time Initi NominaIEnrichment m n) % (years) 300 '3.50 4.00 450 5.00 0 30.73 36.55 42.59 47.99 53.22 5 28.49 32.49 37.54 42.73 47.47 10 26.49 30.38 35.15 40.01 44.50 15 25.30 30.01 33.71 38.18 42.75 20 24.53 29.33 32.78 37.16 41.61 60.00: ACCEPTABLE 55.00 For Restricted Storage 50.00 - (5 45.00 -
-. - .=. _ -_5090% U-235 4.00% U-23 40.00 -
35.00 -
. _ * - -* -. .* 4.00% U-235
~~~~~~~~.._.. ...........-- -- --*.-._._._._..
z 30.00 - j.?%-- ._U 3003523 cc 25.00 - CD 20.00 - 15.00 - 10.00 - UNACCEPTABLE For Restricted Storage 5.00 - 0.00 0 5 10 15 20 COOLDOWN TIME (YEARS) NOTES: Fuel which differs from those designs used to determine the requirements of Table 3.7.15-2 may be qualified for use as a Region 2 Restricted Assembly by means of an analysis using NRC approved methodology to assure that keff is less than 1.0 with no boron and less than or equal to 0.95 with credit for soluble boron. McGuire Units 1and 2 3.7.15-16 Amendment Nos.
Spent Fuel Assembly Storage 3.7.15 Table 3.7.15-3 (Page 1 of 7) Minimum Qualifying Burnup Versus Initial Enrichment and Cooldown Time For Filler Region 2 Storage For Fuel Assembly Type MkBW Bumup (GWD/MTU) versus Initial Nominal Enrichment and Cooldown Time I Cooldown Time Nominaifl n____Init.:al Enrment(O/U.,35) (years) 2.00 2.50 3.00 3.50 4.00 40 5.00 0 27.34 34.90 42.58 50.08 57.40 64.52 71.46 5 23.28 30.12 37.14 43.78 50.40 56.89 63.22 10 21.24 28.12 34.35 40.65 46.94 53.10 59.12 15 20.02 26.67 32.70 38.98 44.88 50.86 56.72 20 19.50 25.73 31.65 37.77 43.55 49.42 55.18 80.00 ACCEPTABLE 75.00 For Use as Filler Assembly 70.00-65.00 - 70.00 - 00% U-235 4.50% U-235 5-0--- - - 5 5.00 ..- C 7-----. -40.00
>~~~~~~~~
_ ~.0 30-00 E 25.o00 - _
......................-- 4.00% U-235 415.00 UNACCEPTAB2.00 10.00 For3UseUas illerAemb 5.00 0.00 -
0 5 10 15 20 COOLDOWN TIME (YEARS) NOTES: Fuel which differs from those designs used to determine the requirements of Table 3.7.15-3 may be qualified for use as a Region 2 Filler Assembly by means of an analysis using NRC approved methodology to assure that k 0f is less than 1.0 with no boron and less than or equal to 0.95 with credit for soluble boron. McGuire Units 1 and 2 3.7.15-17 Amendment Nos.
Spent Fuel Assembly Storage 3.7.15 Table 3.7.15-3 (Page 2 of 7) Minimum Qualifying Burnup Versus Initial Enrichment and Cooldown Time For Filler Region 2 Storage For Fuel Assembly Type MkBWb1 Burnup (GWDIMTU) versus Initial Nominal Enrichment and Cooldown Time Cooldown Time lwInil omina Enr6ficmn(%U25 (years) 2.50 3. 3.50 4.00 45 5.00 0 35.45 42.48 48.89 55.13 61.02 66.75 5 30.69 36.65 42.58 48.29 53.61 58.82 10 29.76 33.85 39.24 44.86 49.90 54.84 15 28.18 32.24 37.45 42.88 47.75 52.53 20 27.20 31.19 36.27 41.58 46.35 51.02 75.00 70.00 ACCEPTABLE 70.00 For Use as Filler Assembly _ 65.00 F 60.00 50.00 45.00 30.00 UNACEPABL3.50% U-235 2-_.50% U-235U23 ....
- 35.00 - -------- ~~._....._. _.3..00%. ............
40.00 2.0 '--'-'-'---------a-------- -- - ---- ---- 35.00300U Ed0.00 10.00 For Use as Filler Assembly 5.00 0.00 - , 0 5 10 15 20 COOLDOWN TIME (YEARS) NOTES: Fuel which differs from those designs used to determine the requirements of Table 3.7.15-3 may be qualified for use as a Region 2 Filler Assembly by means of an analysis using NRC approved methodology to assure deg that is less than 1.0 with no boron and less than or equal to 0.95 with credit for soluble boron. McGuire Units 1and 2 3.7.15-18 Amendment Nos.
Spent Fuel Assembly Storage 3.7.15 Table 3.7.15-3 (Page 3 of 7) Minimum Qualifying Burnup Versus Initial Enrichment and Cooldown Time For Filler Region 2 Storage For Fuel Assembly Type MkBWb2 Burnup (GWDImTU) versus Initial Nominal Enrichment and Cooldown Time Cooldown Time Initial Nominal Enrchment (% U-235). _____t (years) 3.00 3.50 40 4.50 5.00 0 44.74 50.89 56.84 62.57 68.14 5 38.31 44.17 49.58 54.76 59.81 10 35.48 40.89 45.99 50.89 55.68 15 33.78 38.71 43.90 48.63 53.27 20 32.70 37.52 42.56 47.17 51.72 75.00 - ACCEPTABLE 70.00 For Use as Filler Assembly 65.00 60.00 I-55.00 50.00 0~ 45.00 _ -.__
- 3. 5 0%23
---- 5 ~~~~ -- -- ._
---------- - .U-235 5---.00 35-- - - -- -- --
z 40.00 50% U- - 3.00% U-235 cc 35.00
-j 30.00 w 25.00 cN 20.00 U, 15.00 UNACCEPTABLE For Use as Filler Assembly 10.00 5.00 0.00 I 0 5 10 15 20 COOLDOWN TIME (YEARS)
NOTES: Fuel which differs from those designs used to determine the requirements of Table 3.7.15-3 may be qualified for use as a Region 2 Filler Assembly by means of an analysis using NRC approved methodology to assure that keff is less than 1.0 with no boron and less than or equal to 0.95 with credit for soluble boron. McGuire Units 1and 2 3.7.15-19 Amendment Nos.
Spent Fuel Assembly Storage 3.7.15 Table 3.7.15-3 (Page 4 of 7) Minimum Qualifying Burnup Versus Initial Enrichment and Cooldown Time For Filler Region 2 Storage For Fuel Assembly Type W-STD Burnup (GWD/MrU) versus Initial Nominal Enrichment and Cooldown Time I Cooldown Time .nitialNominl Enrme ( U--35) (years) 20 25 3.00 3.50 4.00 4.50 500 0 25.55 33.83 42.22 50.27 58.03 65.54 72.89 5 21.90 30.01 36.78 44.01 51.04 57.87 64.53 10 20.06 27.68 34.03 40.87 47.52 54.01 60.34 15 20.01 26.32 32.41 39.02 45.46 51.75 57.91 20 19.68 25.44 31.39 37.84 44.13 50.29 56.33 80.00 75.00 75.00 ACCEPTABLE
~~~~~~~~~~~~For Use as Filler Assembly 70.00 65.00 50.00 . .- -235 __ 4.00% U-235
_ 50.00 2 -._. - - - - - 4.00% 3--- U-235 45.00 - - U-235
.50% -----.........................
id _-..-_--- -------.- --- --------- _. _ cc 40.00 3 .00% U-235 - c3 25.00 -- _ _ W- 20.00 % U-235 CD 15.00 UNACCEPTABLE 10.00 For Use as F-iler Assembly 5.00 0.00 0 5 10 15 20 COOLDOWN TIME (YEARS) NOTES: Fuel which differs from those designs used to determine the requirements of Table 3.7.15-3 may be qualified for use as a Region 2 Filler Assembly by means of an analysis using NRC approved methodology to assure that keff is less than 1.0 with no boron and less than or equal to 0.95 with credit for soluble boron. McGuire Units 1and 2 3.7.15-20 Amendment Nos.
Spent Fuel Assembly Storage 3.7.15 Table 3.7.15-3 (Page 5 of 7) Minimum Qualifying Burnup Versus Initial Enrichment and Cooldown Time For Filler Region 2 Storage For Fuel Assembly Type MkBI Burnup (GWDIMTU) versus Initial Nominal Enrichment and Cooldown Time Cooldown Time (years) Initial NominalE ( U25 2.00 ____i_____ 2..50x 3.00 3.50: 0 25.14 32.48 40.01 46.65 5 21.76 29.20 35.28 41.37 10 19.98 27.03 32.82 39.13 15 18.94 25.73 31.34 37.45 20 18.27 24.89 30.40 36.39 50.00-ACCEPTABLE For Use as Filler Assembly 45.00 _. S. - - - _.- - 3.50°h-. U 235 I 40.00 _ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~-
.,~ - - _ . _ . _ . _ . _ .-_ . -_ . _ . _.-
35.00 .-..... ._
- ..-.-.- . 3.00% U-235
--- ~~~~~~~------------------------
30.00 zr 25.00 X 20.00 --- _ _-_--_-_--_-2.00%
-- 2.00% U-235 235 - ----
M 15.00 (U 10.00 UNACCEPTABLE For Use as Filler Assembly 5.00 0.00 0 5 10 15 20 COOLDOWN TIME (YEARS) NOTES: Fuel which differs from those designs used to determine the requirements of Table 3.7.15-3 may be qualified for use as a Region 2 Filler Assembly by means of an analysis using NRC approved methodology to assure that kff is less than 1.0 with no boron and less than or equal to 0.95 with credit for soluble boron. McGuire Units 1and 2 3.7.15-21 Amendment Nos.
Spent Fuel Assembly Storage 3.7.15 Table 3.7.15-3 (Page 6 of 7) Minimum Qualifying Burnup Versus Initial Enrichment and Cooldown Time For Filler Region 2 Storage For Fuel Assembly Type W-OFA Burnup (GWDJMTU) versus Initial Nominal Enrichment and Cooldown Time I Cooldown Time X_1lini Nominal Enchment .- ( 235) (years) i.00 3 2520.00. 4.00 4.50 5.00 0 22.71 30.79 38.56 45.46 52.60 59.57 66.38 5 20.01 27.42 34.25 40.55 47.08 53.46 59.71 10 18.87 25.56 32.01 38.51 44.22 50.31 56.27 15 18.00 24.44 30.67 36.96 42.51 48.42 54.24 20 17.43 23.71 30.01 35.96 41.41 47.20 52.92 70.00 70.00 ~~~~~ACCEPTABLE 65.00 - For Use as Filler Assembly 60.00 5.00% U-235 E 55.00 j-__3.50% U-235 5L0.00 . ., - - -_ -- - -- -- - - .- 035.00 .......-- _.........._._... 300%U-235 m 30.00 2.50% U-235 Co 25.00 15.00 5 20.00-~~--_______2.00% U~235 0 5 UNACCEPTABLE 20.00 10.00
<l , .2 For Use as Filler Assembly . . . . . . .00 U-235 . .
5.00 0 5 10 15 20 COOLDOWN TIME (YEARS) NOTES: Fuel which differs from those designs used to determine the requirements of Table 3.7.15-3 may be qualified for use as a Region 2 Filler Assembly by means of an analysis using NRC approved methodology to assure that keffis less than 1.0 with no boron and less than or equal to 0.95 with credit for soluble boron. McGuire Units 1and 2 3.7.15-22 Amendment Nos.
Spent Fuel Assembly Storage 3.7.15 Table 3.7.15-3 (Page 7 of 7) Minimum Qualifying Bumup Versus Initial Enrichment and Cooldown Time For Filler Region 2 Storage For Fuel Assembly Type W-RFA I Burnup (GWD/MTU) versus Initial Nominal Enrichment and Cooldown Time Cooldown Time Enrichmnt (% U-35)
.niialNmnal (years) 3.50 400 4.50 i- 5.00 0 41.90 48.19 54.22 60.04 65.72 5 35.92 41.96 47.39 52.66 57.81 10 33.22 38.50 44.00 49.01 53.90 15 31.66 36.77 42.03 46.89 51.62 20 30.66 35.65 40.76 45.49 50.14 70.00 70.00 ~~~~~ACCEPTABLE 65.00 For Use as Filler Assembly 60.00 5 5 00 ,,,_4ab-_5.0
-3 = 4.50% *U* - 5.00% U-235 05 45.00 .. W ~~~~~3.50% U-235
................................ 4O%-3 g40.00 -....---- _._.. ._ _ . ._ __._. .... _._. _._. _._......
35.00 .......... 3.00% U-235 m 30.00 25.00
- 20.00 in 15.00 en UNACCEPTABLE
< 10.00 For Use as Filler Assembly 5.00 0.00 0 5 10 15 20 COOLDOWN TIME (YEARS)
NOTES: Fuel which differs from those designs used to determine the requirements of Table 3.7.15-3 may be qualified for use as a Region 2 Filler Assembly by means of an analysis using NRC approved methodology to assure that kf is less than 1.0 with no boron and less than or equal to 0.95 with credit for soluble boron. McGuire Units 1and 2 3.7.15-23 Amendment Nos.
Spent Fuel Assembly Storage 3.7.15 Table 3.7.15-4 (Page 1 of 7) Minimum Qualifying Burnup Versus Initial Enrichment and Cooldown Time For Restricted with Empty Checkerboard Region 2 Storage For Fuel Assembly Type MkBW Burnup (GWD/MrU) versus Initial Nominal Enrichment and Cooldown Time I Cooldown Time I Et ch______%U-235) (years) o2.00 2.50; .00 3 00 .0s 0 8.12 16.50 22.94 29.15 34.67 40.43 46.20 5 7.49 14.77 20.81 26.50 31.60 37.03 42.18 10 7.07 13.77 19.79 24.96 30.01 34.99 40.01 15 6.81 13.16 18.98 24.00 29.10 33.73 38.67 20 6.64 12.78 18.45 23.37 28.35 32.90 37.75 50.00 ACCEPTABLE For Restricted wffh Empty Checkerboard 45.00 40.00 5.00%U-235 i 25.00 - _ _ _ _ _ _ _ 4.50% U-235 35.00_____________ __. CL 320.00
°-° - ------ 4.00%
3 U-235 D 15.00 2.50% U-235 X 10.00 5.00 UNACCEPTABLE For Restricted with Empty Checkerboard 0 .00 - , . . . . . . . . . . . . . . . . . . 0 5 10 15 20 COOLDOWN TIME (YEARS) NOTES: Fuel which differs from those designs used to determine the requirements of Table 3.7.15-4 may be qualified for use as a Region 2 Restricted with Empty Checkerboard Assembly by means of an analysis using NRC approved methodology to assure that keff is less than 1.0 with no boron and less than or equal to 0.95 with credit for soluble boron. McGuire Units 1 and 2 3.7.15-24 Amendment Nos
Spent Fuel Assembly Storage 3.7.15 Table 3.7.15-4 (Page 2 of 7) Minimum Qualifying Burnup Versus Initial Enrichment and Cooldown Time For Restricted with Empty Checkerboard Region 2 Storage For Fuel Assembly Type MkBWbl Burnup (GWD/MTU) versus Initial Nominal Enrichment and Cooldown Time Cooldown Time IiilNmnlErcmn %U25 (years) 250 3.0 3.50 4.00 4.0 5.00 0 16.23 23.10 28.95 33.14 38.33 43.68 5 14.53 20.89 26.24 30.19 34.95 39.57 10 13.55 19.52 24.65 29.63 32.99 37.40 15 12.95 18.65 23.66 28.48 31.78 36.06 20 12.58 18.08 23.01 27.73 30.99 35.17 50.00 - ACCEPTABLE For Restricted with Empty Checkerboard 45.00 - I 40.00-aN 35.00- 4.00% U-235 5 a.. 30.00- -----
- - %U
................... ... . . 4.00%.U--235 zW 25.00- ~----~-~~ -- ------.. _ _
30% --3S
, 20.00- -. --.. -.. .. - - -. -. - ~3.00% U-235 M 15.00 -
C,, Ca 10.00-UNACCEPTABLE For Restricted with Empty Checkerboard 5.00 - 0.00 . . . . . . . . . . . . . . 0 5 10 15 20 COOLDOWN TIME (YEARS) NOTES: Fuel which differs from those designs used to determine the requirements of Table 3.7.15-4 may be qualified for use as a Region 2 Restricted with Empty Checkerboard Assembly by means of an analysis using NRC approved methodology to assure that k., is less than 1.0 with no boron and less than or equal to 0.95 with credit for soluble boron. McGuire Units 1 and 2 3.7.15-25 Amendment Nos
Spent Fuel Assembly Storage 3.7.15 Table 3.7.15-4 (Page 3 of 7) Minimum Qualifying Bumup Versus Initial Enrichment and Cooldown Time For Restricted with Empty Checkerboard Region 2 Storage For Fuel Assembly Type MkBWb2 Burnup (GWD/ImTU) versus Initial Nominal Enrichment and Cooldown Time Cooldown Time Iniial Enrichment) _______ (years) -. 00.3.50 : 4.00 4.50. 00. 0 23.77 29.59 33.83 38.97 44.48 5 21.44 26.88 30.81 35.52 40.41 10 20.03 25.29 30.01 33.52 37.89 15 18.88 24.29 29.01 32.29 36.53 20 18.35 23.64 28.25 31.48 35.62 50.00: ACCEPTABLE For Restricted with Empty Checkerboard 45.00-I 40.00 _o_U__ _-_0% .- U-235-235 R 35.00 sL CL 30.00 - ._ _ 3.50% 5'. - - - - -- - -- -- - - - - - - 235
---3.5%
_. U-235 zW 25.00 M 20.00 ~~~~- -- ~ - --- ~3.00% U-235
- 15.00-
) 1 UNACCEPTABLE
- 0) 10.00 For Restricted with Empty Checkerboard 5.00 0.00 . . . . . . . . . . .
0 5 10 15 20 COOLDOWN TIME (YEARS) NOTES: Fuel which differs from those designs used to determine the requirements of Table 3.7.15-4 may be qualified for use as a Region 2 Restricted with Empty Checkerboard Assembly by means of an analysis using NRC approved methodology to assure that kff is less than 1.0 with no boron and less than or equal to 0.95 with credit for soluble boron. McGuire Units 1 and 2 3.7.15-26 Amendment Nos
Spent Fuel Assembly Storage 3.7.15 Table 3.7.15-4 (Page 4 of 7) Minimum Qualifying Burnup Versus Initial Enrichment and Cooldown Time For Restricted with Empty Checkerboard Region 2 Storage For Fuel Assembly Type W-STD Burnup (GWD/MTU) versus Initial Nominal Enrichment and Cooldown Time I Cooldown Time Initia ina; Enrichment U25 __i____ (years) 2.00 .50 ........ K5.00 0 7.24 14.97 21.42 28.18 34.01 40.25 46.34 5 6.79 13.83 20.01 25.78 31.08 36.80 42.40 10 6.46 13.11 19.69 24.37 30.01 34.81 40.14 15 6.24 12.65 18.99 23.48 29.01 33.57 38.74 20 6.11 12.37 18.54 22.91 28.31 32.76 37.83 50.00 ACCEPTABLE 45.00 For Restrcted with Empty Checkerboard S. 1 40.00 - 5.00% U-23
- _- _4.50% U-235 Ad2.0.00.-...... 4.00% U-235 30.00 zCC 25.00
. -. --. _. 3.50% U-235 D~
15.00 2.50%U-235 C210.00
< ___________- 2.00% U-235 5.00 UNACCEPTABLE For Restricted with Empty Checkerboard 0.00 .
0 5 10 15 20 COOLDOWN TIME (YEARS) NOTES: Fuel which differs from those designs used to determine the requirements of Table 3.7.15-4 may be qualified for use as a Region 2 Restricted with Empty Checkerboard Assembly by means of an analysis using NRC approved methodology to assure that keff is less than 1.0 with no boron and less than or equal to 0.95 with credit for soluble boron. McGuire Units 1 and 2 3.7.15-27 Amendment Nos
Spent Fuel Assembly Storage 3.7.15 Table 3.7.15-4 (Page 5 of 7) Minimum Qualifying Burnup Versus Initial Enrichment and Cooldown Time For Restricted with Empty Checkerboard Region 2 Storage For Fuel Assembly Type MkBI Bumup (GWD/MTU) versus Initial Nominal Enrichment and Cooldown Time Cooldown Time . (years) FT;:> Ini> Etal Nominal En^:hmentI 235) ____________ H -2.00 < 2.502 0 3;.00 ;4 3.500 0 7.67 15.06 21.81 28.16 5 7.20 13.82 20.01 25.83 10 6.86 13.05 18.92 24.44 15 6.66 12.57 18.23 23.56 20 6.53 12.26 17.78 22.99 30.00 - ACCEPTABLE _. . . -. For Restricted with Empty Checkerboard
' --. 3.50% U-235 m 25.00 - ~~~--
- 20.00 -
... _ .. 3.00% U-235 %2.w - -.... -
- a. ... ..... .. ..
ze 15.00-B CD 10.00-w ---- __________ 2.00% U-235 M 5.00- UNACCEPTABLE For Restricted with Empty Checkerboard 0.00 - I ~ ~ 1 0 5 10 15 20 COOLDOWN TIME (YEARS) NOTES: Fuel which differs from those designs used to determine the requirements of Table 3.7.15-4 may be qualified for use as a Region 2 Restricted with Empty Checkerboard Assembly by means of an analysis using NRC approved methodology to assure that keff is less than 1.0 with no boron and less than or equal to 0.95 with credit for soluble boron. McGuire Units 1 and 2 3.7.15-28 Amendment Nos
Spent Fuel Assembly Storage 3.7.15 Table 3.7.15-4 (Page 6 of 7) Minimum Qualifying Burnup Versus Initial Enrichment and Cooldown Time For Restricted with Empty Checkerboard Region 2 Storage For Fuel Assembly Type W-OFA Burnup (GWDImTU) versus Initial Nominal Enrichment and Cooldown Time Cooldown Time____ IiilNmnlErcmn(%.125 (years) 2.00 2.50 300 4.00i3.50v4. 5.004 0 6.69 14.08 20.44 26.78 32.70 38.68 44.03 5 6.32 13.06 19.21 24.72 30.15 35.68 40.65 10 6.07 12.40 18.26 23.50 28.94 33.93 39.12 15 5.91 11.98 17.66 22.73 28.00 32.83 37.87 20 5.82 11.71 17.27 22.22 27.38 32.10 37.05 50.00 50.00 - ~~~~~ACCEPTABLE 45.00 - 45.00 ~~~~~~~~~~~For Restricted with Empty Checkerboard 1- 40.00~ S 35.00~~A~-0 -------------------------------------- o 3 4.50
- U-235
-- 4 35.00 30.00 ................... _._._._ .4.00% U-235 215.00 --
- -- 3.50% U-235 M 20.00
~ ~ ~ ~ ~ *-. -. 3.00% U-235 15.00 U-235
~~~~~2.50%
w C 10.00 2.00% U-235 5.00 UNACCEPTABLE For Restricted with Empty Checkerboard 0.00 I I I I 0 5 10 15 20 COOLDOWN TIME (YEARS) NOTES: Fuel which differs from those designs used to determine the requirements of Table 3.7.15-4 may be qualified for use as a Region 2 Restricted with Empty Checkerboard Assembly by means of an analysis using NRC approved methodology to assure that kef is less than 1.0 with no boron and less than or equal to 0.95 with credit for soluble boron. McGuire Units 1 and 2 3.7.15-29 Amendment Nos
Spent Fuel Assembly Storage 3.7.15 Table 3.7.15-4 (Page 7 of 7) Minimum Qualifying Burnup Versus Initial Enrichment and Cooldown Time For Restricted with Empty Checkerboard Region 2 Storage For Fuel Assembly Type W-RFA Burnup (GWDjMTU) versus Initial Nominal Enrichment and Cooldown Time Cooldown Time Iit-ial Nominal Enrichment (% U-235 (years) 300 . 0 4.50 5.00 0 22.87 28.80 32.88 38.06 43.49 5 20.56 26.06 30.01 34.66 39.27 10 18.99 24.48 29.27 32.73 37.11 15 18.23 23.47 28.12 31.52 35.77 20 17.74 22.81 27.37 30.73 34.89 50.00-ACCEPTABLE For Restricted with Empty Checkerboard 45.00 E 40.00 1 aN 35.00 ----. _450% U-235 aI 30.00 1 -.. ..... ..... ... .. .......... ,. 4.00X U-235
'. --- -.- . _% _3 /. U-235 zM 25.00 A 20.00 --.--.- --- -- ._ _35 3.00. U-235 M 15.00 X.1 0.00 UNACCEPTABLE For Restricted with Empty Checkerboard 5.00 0.00 4 ~ . . . . .
0 5 10 15 20 COOLDOWN TIME (YEARS) NOTES: Fuel which differs from those designs used to determine the requirements of Table 3.7.15-4 may be qualified for use as a Region 2 Restricted with Empty Checkerboard Assembly by means of an analysis using NRC approved methodology to assure that kef is less than 1.0 with no boron and less than or equal to 0.95 with credit for soluble boron McGuire Units 1 and 2 3.7.15-30 Amendment Nos
Spent Fuel Assembly Storage 3.7.15 II r I11 RESTRICTED FUEL U Restricted Fuel: Fuel which meets the minimum burnup requirements of Table 3.7.15-2, or non-fuel components, or an empty cell. Filler Location: Either fuel which meets the minimum burnup requirements of Table 3.7.15-3, or an empty cell. Boundary Condition: None. Figure 3.7.15-1 (page 1 of 1) Required 2 out of 4 Loading Pattern for Restricted Region 2 Storage McGuire Units 1 and 2 3.7. 15-31 Amendment Nos
Spent Fuel Assembly Storage 3.7.15 EZTh r-r10 ro 0 CHECKERBOARD CHECKERBOARD CHECKERBOARD CHECKERBOARD FUEL FUEL FUEL FUEL 0 h 0 CHECKERBOARD EMPTY CHECKERBOARD EMPTY FUEL CELL FUEL CELL 0
-7__W 0
CHECKERBOARD CHECKERBOARD CHECKERBOARD FUEL FUEL FUEL 0 0 I2 izz CHECKERBOARD EMPTY CHECKERBOARD EMPTY FUEL CELL FUEL CELL 0>Srmu~^C Checkerboard Fuel: Fuel which meets the minimum burnup requirements of Table 3.7.15-4, or non-fuel components, or an empty cell. Boundary Condition: Row or Column of only Checkerboard Fuel (Example: Row 1 or Column 1) shall be bounded by either: a) Alternating pattern of Checkerboard Fuel and empty cell, b) String of empty cells, or c) Spent fuel pool wall. No boundary conditions for a row or column of alternating pattern of Checkerboard Fuel and empty cell (Example: Row 4 or Column 4) Figure 3.7.15-2 (page 1 of 1) Required 3 out of 4 Loading Pattern for Checkerboard Region 2 Storage McGuire Units 1 and 2 3.7.1 5-32 Amendment Nos
Design Features 4.0 4.0 DESIGN FEATURES 4.1 Site Location The McGuire Nuclear Station site is located at latitude 35 degrees, 25 minutes, 59 seconds north and longitude 80 degrees, 56 minutes, 55 seconds west. The Universal Transverse Mercator Grid Coordinates are E 504, 669, 256, and N 3, 920, 870, 471. The site is in northwestern Mecklenburg County, North Carolina, 17 miles north-northwest of Charlotte, North Carolina. 4.2 Reactor Core 4.2.1 Fuel Assemblies The reactor shall contain 193 fuel assemblies. Each assembly shall consist of a matrix of Zircalloy fuel rods with an initial composition of natural or slightly enriched uranium dioxide (UO2 ) as fuel material. Limited substitutions of zirconium alloy or stainless steel filler rods for fuel rods, in accordance with approved applications of fuel rod configurations, may be used. Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff approved codes and methods and shown by tests or analyses to comply with all fuel safety design bases. A limited number of lead test assemblies that have not completed representative testing may be placed in nonlimiting core regions. 4.2.2 Control Rod Assemblies The reactor core shall contain 53 control rod assemblies. The control material shall be silver indium cadmium (Unit 1) silver indium cadmium and boron carbide (Unit 2) as approved by the NRC. 4.3 Fuel Storage 4.3.1 Criticality 4.3.1.1 The spent fuel storage racks are designed and shall be maintained with:
- a. Fuel assemblies having a maximum nominal U-235 enrichment of 5.00 weight percent; I
- b. kff < 1.0 if fully flooded with unborated water, which includes an allowance for uncertainties as described in Section 9.1 of the UFSAR;
- c. keff < 0.95 if fully flooded with water borated to 800 ppm, which I includes an allowance for uncertainties as described in Section 9.1 of the UFSAR; McGuire Units 1 and 2 4.0-1 Amendment Nos.
Design Features 4.0 4.0 DESIGN FEATURES 4.3 Fuel Storage (contiinued)
- d. A nominal 10.4 inch center to center distance between fuel assemblies placed in Region 1 and I
- e. A nominal 9.125 inch center to center distance between fuel assemblies placed in Region 2. I 4.3.1.2 The new fuel storage racks are designed and shall be maintained with:
- a. Fuel assemblies having a maximum nominal U-235 enrichment of 5.00 weight percent; I
- b. keff < 0.95 if fully flooded with unborated water, which includes an allowance for uncertainties as described in Section 9.1 of the UFSAR;
- c. kff
- 0.98 if moderated by aqueous foam, which includes an allowance for uncertainties as described in Section 9.1 of the UFSAR; and
- d. A nominal 21 inch center to center distance between fuel assemblies placed in the storage racks.
4.3.2 Drainage The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 745 ft.-7 in. 4.3.3 Capacitv The spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than 1463 fuel assemblies (286 total spaces in Region 1 and 1177 total spaces in Region 2). McGuire Units 1 and 2 4.0-2 Amendment Nos.
ATTACHMENT 3 DESCRIPTION OF PROPOSED CHANGES AND TECHNICAL JUSTIFICATION
Attachment 3 Page lof 10 Description of Proposed Changes Duke Power Company proposes to modify the McGuire Nuclear Station (MNS) Technical Specifications (TS) Sections 3.7.15, Spent Fuel Assembly Storage, and 4.3, Design Features - Fuel Storage. A markup of the specific changes is shown in Attachment 1. This License Amendment Request (LAR) provides revised spent fuel storage criteria based upon fuel type, fuel enrichment, burnup, cooling time and partial credit for soluble boron. In addition, this amendment allows for the safe storage of fuel assemblies with a nominal enrichment of U-235 up to 5.0 weight percent. Finally, this LAR reduces the required soluble boron credit from 850 ppm to 800 ppm. The proposed TS changes are based upon the new McGuire Fuel Storage Criticality Analysis (Attachment 6). The criticality analysis was performed in accordance with the regulatory criteria of 10 CFR 50.68(b). The TS changes in this LAR include the following: a) LCO 3.7.15: This LCO is modified by deleting the reference to Integral Fuel Burnable Absorber (IFBA) rods and replacing it with a reference to cooling time. The criticality analysis, as discussed in Attachment 6, is performed in accordance with the requirements of 10 CFR 50.68(b). The evaluation takes credit for Boral in the new Region 1 spent fuel storage racks and no longer takes credit for any remaining Boraflex in the Region 2 spent fuel storage racks or for iFBA rods that may be present. Credit is taken for burnup, and cooling time for the Region 2 spent fuel storage racks. In addition, the analysis utilizes partial credit for the soluble boron in the spent fuel pool water. b) LCO 3.7.15a: This LCO Section defined the fuel limits and acceptable storage configurations for new or irradiated fuel to be stored within Region IA of the spent fuel pool. The entire LCO Section is to be deleted and replaced with a new LCO Section that defines the requirements for the safe storage of new or irradiated fuel within Region 1 of the spent fuel pool. This LCO Section will specify that the maximum initial U-235 enrichment of fuel stored in Region 1 be equal to or less than 5.00 weight percent. There will be no other restrictions for the safe storage of new or irradiated fuel within Region 1. c) LCO 3.7.15b: This LCO Section defined the fuel limits and acceptable storage configurations for new or irradiated fuel to be located in Region lB of the spent fuel pool. The entire LCO Section is to be deleted and replaced with a new LCO Section that defines the requirements for the safe storage of new or irradiated fuel within Region 2 of the spent fuel pool. This LCO Section establishes three fuel storage configurations for Region 2; 1) unrestricted storage; 2) restricted storage or 3) checkerboard storage. The new TS Tables define the fuel limits for the three storage configurations and are referred within this LCO Section. d) LCO 3.7.15c: This LCO Section defined the fuel limits and acceptable storage configurations for new or irradiated fuel to be located in Region 2A of the spent fuel pool. This entire LCO Section is to be deleted, since the criticality analysis no longer takes credit for Boraflex as a neutron absorbing material.
Attachment 3 Page 2 of 10 e) LCO 3.7.15d: This LCO Section defined the fuel limits and acceptable storage configurations for new or irradiated fuel to be located in Region 2B of the spent fuel pool. This entire LCO Section is to be deleted since the criticality analysis no longer takes credit for Boraflex as a neutron absorbing material. f) Tables 3.7.15-1 through 3.7.15-12: These TS Tables specify the burnup and enrichment limits for fuel stored in the spent fuel racks. The data for these tables were derived from criticality analysis that relied on partial credit for Boraflex. These TS Tables are deleted since credit for Boraflex as a neutron absorber has been eliminated. g) Figures 3.7.15-1 through 3.7.15-7: These TS Figures illustrate acceptable loading patterns and define the boundary conditions for various storage configurations. The loading patterns were based on the criticality analysis that relied on partial credit for Boraflex. These TS Figures are deleted since credit for Boraflex as a neutron absorber has been eliminated. h) New Table 3.7.15-1: This TS Table specifies the minimum burnup requirements as a function of initial enrichment, fuel assembly design type and post-irradiation cooling time to be stored as unrestricted fuel in Region 2 of the spent fuel racks. The data for this table is based on the criticality analysis discussed in Attachment 6. The analysis no longer takes credit for any remaining Boraflex in the Region 2 spent fuel storage racks. For each fuel assembly design type, credit is taken for burnup, cooling time and partial credit for the soluble boron in the spent fuel pool water. i) New Table 3.7.15-2: This TS Table specifies the minimum burnup requirements as a function of initial enrichment, fuel assembly design type and post-irradiation cooling time to be stored in Region 2 of the spent fuel racks as a restricted fuel assembly for the 2 out of 4 restricted/filler storage configuration. The data for this table is based on the criticality analysis discussed in Attachment 6. The analysis no longer takes credit for any remaining Boraflex in the Region 2 spent fuel storage racks. For each fuel assembly design type, credit is taken for burnup, cooling time and partial credit for the soluble boron in the spent fuel pool water. j) New Table 3.7.15-3: This TS Table specifies the minimum burnup requirements as a function of initial enrichment, fuel assembly design type and post-irradiation cooling time to be stored in Region 2 of the spent fuel racks as a filler fuel assembly for the 2 out of 4 restricted/filler storage configuration. The data for this table is based on the criticality analysis discussed in Attachment 6. The analysis no longer takes credit for any remaining Boraflex in the Region 2 spent fuel storage racks. For each fuel assembly design type, credit is taken for burnup, cooling time and partial credit for the soluble boron in the spent fuel pool water. k) New Table 3.7.15-4: This TS Table specifies the minimum burnup requirements as a function of initial enrichment, fuel assembly design type and post-irradiation cooling time to be stored in Region 2 of the spent fuel racks as a checkerboard fuel assembly for the 3 out of 4 checkerboard/empty storage configuration. The data for this table is based on the
Attachment 3 Page 3 of 10 criticality analysis discussed in Attachment 6. The analysis no longer takes credit for any remaining Boraflex in the Region 2 spent fuel storage racks. For each fuel assembly design type, credit is taken for burnup, cooling time and partial credit for the soluble boron in the spent fuel pool water.
- 1) New Figure 3.7.15-1: This TS Figure illustrates the loading pattern to be employed in the Region 2 spent fuel storage racks for the 2 out of 4 restricted/filler storage configuration.
There are no boundary conditions specified for this storage configuration. The loading pattern illustrated by this figure is based on the criticality analysis discussed in Attachment 6. The analysis no longer takes credit for any remaining Boraflex in the Region 2 spent fuel storage racks. For each fuel assembly design type, credit is taken for burnup, cooling time and partial credit for the soluble boron in the spent fuel pool water. m) New Figure 3.7.15-2: This TS Figure illustrates the loading pattern to be employed in the Region 2 spent fuel storage racks for the 3 out of 4 checkerboard/empty cell storage configuration. Boundary conditions for this storage configuration are specified. For this configuration, a string of checkerboard fuel is to be bounded by either; 1) an alternating pattern of checkerboard fuel and empty cell, 2) string of empty cells, or 3) spent fuel pool wall. The loading pattern illustrated by this figure is based on the criticality analysis discussed in Attachment 6. The analysis no longer takes credit for any remaining Boraflex in the Region 2 spent fuel storage racks. For each fuel assembly design type, credit is taken for burnup, cooling time and partial credit for the soluble boron in the spent fuel pool water. n) TS 4.3.1.1a: The maximum nominal U-235 enrichment of fuel to be stored in the spent fuel storage racks is increased from 4.75 weight percent to 5.00 weight percent. The criticality analysis discussed in Attachment 6 is performed in accordance with the requirements of 10 CFR 50.68(b). The analysis takes credit for the Boral in the new Region 1 spent fuel storage racks and no longer takes credit for any remaining Boraflex in the Region 2 spent fuel storage racks. For each fuel assembly design type, credit is taken for burnup, cooling time and partial credit for the soluble boron in the spent fuel pool water. In addition, the analysis utilizes partial credit for the soluble boron in the spent fuel pool water. o) TS 4.3.1.1c: The required soluble boron concentration necessary to maintain keff less than 0.95 is reduced from 850 ppm to 800 ppm. The criticality analysis confirms that 800 ppm of partial soluble boron credit is sufficient to maintain keff less than 0.95. The criticality analysis as discussed in Attachment 6 is performed in accordance with the requirements of 10 CFR 50.68(b). The analysis takes credit for the Boral in the new Region 1 spent fuel storage racks and no longer takes credit for any remaining Boraflex in the Region 2 spent fuel storage racks. For each fuel assembly design type, credit is taken for burnup, cooling time and partial credit for the soluble boron in the spent fuel pool water. p) TS 4.3.1.1d: The TS specifies the nominal center to center spacing between fuel assemblies stored within Regions lA and lB. The subregion designations of lA and lB are deleted, since the criticality analysis no longer takes credit for Boraflex as a neutron
Attachment 3 Page 4 of 10 absorbing material in Region 1. The new designation stated in the TS is Region 1. q) TS 4.3.1.1e: The TS specifies the nominal center to center spacing between fuel assemblies stored within Regions 2A and 2B. The subregion designations of 2A and 2B are deleted, since the criticality analysis no longer takes credit for Boraflex as a neutron absorbing material in Region 2. The new designation stated in the TS is Region 2. r) TS 4.3.1.2a: The maximum nominal U-235 enrichment of fuel to be stored in the new fuel storage racks is increased from 4.75 weight percent to 5.0 weight percent. The criticality analysis discussed in Attachment 6 is performed in accordance with the requirements of 10 CFR 50.68(b)(2) & 10 CFR 50.68(b)(3). For the new fuel storage racks, all fuel is considered to be unirradiated within the analysis. The analysis takes no credit for spacer grids or other neutron poisons that may be inserted in the fuel assembly. s) TS 4.3.3: The TS specifies the total storage capacity of the spent fuel storage pool. The subregion designations of IA, 1B, 2A and 2B are deleted since the criticality analysis no longer takes credit for Boraflex as a neutron absorbing material. The new designations stated are Region 1 and Region 2. Technical Justification This section provides the technical justification for the proposed modifications to the MNS Technical Specifications. These changes address revised spent fuel storage criteria based upon fuel type, fuel enrichment, burnup, cooling time and partial credit for soluble boron. In addition, the nominal fuel enrichment that can be stored within the spent fuel racks is increased. Finally, this proposed amendment reduces the required soluble boron credit. These changes allow for the storage of fuel without the need to credit Boraflex for reactivity control in the MNS spent fuel pool. These changes, also, increase design and operational flexibility, while at the same time maintaining acceptable criticality safety margins and decay heat removal capabilities. The existing design basis for preventing criticality in the McGuire spent fuel storage pools is that, including uncertainties, there is a 95% probability at a 95% confidence level that keff of the fuel storage assembly array will be less than 1.0 if fully flooded with unborated water, and keff will be equal to or less than 0.95 if fully flooded with water borated to 850 ppm, with credit for the presence of IFBA rods where applicable, and reduced credit for the degraded spent fuel rack Boraflex neutron absorber panels. Each spent fuel pool contains a two region rack design. Region 1 racks (286 storage locations) have a fuel assembly spacing of 10.4 inches, utilizing a neutron absorbing material. These racks (2 modules/pool) are typically reserved for temporary core off loading and storage of non-irradiated fully enriched fuel. The Region 1 racks are composed of individual storage cells made of stainless steel that utilize Boral as the neutron absorbing material. The Region 1 racks had utilized Boraflex as the neutron absorbing material. In July 2003, the Region 1 racks were replaced with a similar designed rack, except for the neutron absorber material being Boral. The replacement of the Region 1 racks was performed per the provisions of 10 CFR 50.59.
Attachment 3 Page 5 of 10 Region 2 (1177 storage locations) has a fuel assembly spacing of 9.125 inches and utilizes Boraflex as a neutron absorbing material. The Region 2 racks provide normal long term storage for irradiated fuel assemblies and can be used for restricted storage of new fuel. Currently, each region is further subdivided into two subregions based on the amount of remaining Boraflex. Placement of fuel into a given subregion without restriction is limited to assemblies meeting a certain minimum required assembly burnup versus enrichment. In the event that fuel assemblies do not meet the minimum requirements for unrestricted storage, a restricted storage configuration must be utilized. In the event that fuel assemblies do not meet the minimum requirements for restricted storage, a checkerboard storage configuration must be utilized. McGuire TS 3.7.15 will be amended to provide revised spent fuel pool storage configurations, and revised spent fuel pool storage criteria, specifying minimum burnup requirements as a function of initial fuel enrichment, post-irradiation cooling time, and fuel assembly design type. With the applicable minimum concentration of soluble boron present in the spent fuel pool, and credit for the Boral neutron absorber panels where applicable, these changes will ensure that the pool storage rack kff is < 0.95 under non-accident conditions, and accident conditions (including the unlikely occurrence of a credible spent fuel pool dilution event with thorough mixing). The applicable minimum concentration of soluble boron is ensured by existing McGuire TS 3.7.14. The new McGuire Fuel Storage Criticality Analysis evaluates the Region 1 and Region 2 storage racks in the McGuire spent fuel pools. These spent fuel storage racks originally contained Boraflex poison panels for reactivity holddown. However; ongoing degradation of the Boraflex material has limited the effectiveness of continuing to rely on this poison material in Region 1 and Region 2. To address the continuing degradation of the Boraflex panels, the McGuire criticality analysis considers "permanent solutions" to this issue. The permanent solutions for Region 1 and Region 2 include the following: Region 1 Re-rack with Boral poison panels. The old Region 1 Boraflex racks were replaced with racks designed, fabricated and supplied by Holtec in mid-2003. The new Region 1 racks have the same dimensions as the old racks, and thus the same storage capacity (286 cells). The new Region 1 racks will allow unrestricted storage of fresh fuel, up to 5.00 weight percent U-235. Region 2 Retain existing racks, but eliminate credit for any remaining Boraflex poison. Take credit for cooling time reactivity reduction (due primarily to Pu-241 decay and Gd-155 buildup following the end of reactor irradiation). Segregate storage burnup requirements by fuel assembly type to take advantage of lower reactivity associated with certain fuel designs (seven different types have been identified). Finally, take credit for burnup in storage arrays containing empty cells, such as 3 assemblies with one empty cell. The Region 2 criticality analysis employs specific 3-D calculations for the fuel storage configurations that take credit for burnup.
Attachment 3 Page 6 of 10 The criticality evaluation demonstrated that the new Region 1 Boral racks can store fresh McGuire reactor fuel of any type, up to 5.00 weight percent of U-235, with no restrictions. The minimum burnup requirements for Region 2 storage were developed for seven different fuel types, as a function of initial enrichment and post-irradiation cooling time. These burnup requirements were specified for three Region 2 storage configurations: Unrestricted, 2 out of 4 Restricted/Filler, and 3 out of 4 Checkerboard/Empty. For the spent fuel pool storage rack criticality analyses, the maximum 95/95 kff is determined to be less than 1.00 with no boron in the spent fuel pool water for both Region 1 and 2 storage racks. These results meet the no-boron 95/95 kff criterion in 10 CFR 50.68(b)(4). Further, the criticality analysis confirmed that 800 ppm of partial soluble boron credit is sufficient to maintain the maximum 95/95 keff less than 0.95. A minimum boron concentration of 1600 ppm is adequate to maintain the maximum 95/95 keff below 0.95 for a worst-case misleading event in the McGuire spent fuel pool. Finally, for the worst-case weir gate drop on the new Region 1 Boral racks, the maximum achievable 95/95 keff is well below the 0.95 subcriticality criterion, when credit is taken for 2475 ppm soluble boron in the SFP. The new McGuire Fuel Storage Criticality Analysis demonstrates that under non-accident conditions a spent fuel storage pool boron concentration of 800 ppm would be adequate to maintain the spent fuel storage rack keff < 0.95. Existing McGuire TS 3.7.14 states that the spent fuel pool storage boron concentrations shall be maintained within the limits specified in the McGuire Core Operating Limits Report (COLR). The spent fuel pool boron concentration limit currently specified in the COLR is 2675 ppm, which is well above the minimum required boron credit of 800 ppm for non-accident conditions. A possibility does exist that the boron concentration in the spent fuel pool could be lowered below the COLR limit by a pool dilution event. Consequently, an analysis of a dilution event of the spent fuel pool boron concentration is necessary to ensure that acceptable levels of subcriticality are maintained during and following the event. As part of this analysis, calculations were performed to define the dilution time and volumes for the spent fuel pool. The dilution sources available at McGuire were compiled and evaluated against the calculated dilution volume to identify the bounding "continuous flow" dilution event. The McGuire dilution analysis concluded that the bounding event was a pipe break in the non-seismic fire protection system, as this could deliver the largest flow rate (700 gpm) of unborated water into the SFP. For this dilution event, in conjunction with an isolation of the cask loading pit, calculations determined that it would take at least 9.5 hours to dilute the SFP from an initial boron concentration of 2675 ppm to 800 ppm. Such a scenario would involve substantial overflow of the SFP in less than two hours, and it was deemed incredible, because numerous indicators such as level alarms, flooding in the auxiliary building, fire protection pump header flow alarms, etc., would alert Operations long before 9 hours had elapsed (Reference 4). The above post-dilution event is based upon the assumption that all of the unborated water is thoroughly mixed with the water in the pool. Given the spent fuel storage pool cooling water flow and convection from the spent fuel decay heat, it is likely that this thorough mixing would occur. However, if mixing was not adequate, it is possible that a localized pocket of non-borated water could form somewhere in the spent fuel pool. This possibility is addressed by the calculation in which shows that a spent fuel storage pool kff will still be less than 1.0 on a 95/95
Attachment 3 Page 7 of 10 basis with the spent fuel pool filled with unborated water. Thus, in the unlikely event that the worst case dilution event occurred and then a pocket of non-borated water formed in the spent fuel pool due to inadequate mixing, acceptable subcritical conditions would still be maintained in the McGuire spent fuel storage pools. Many of the postulated spent fuel pool accidents at McGuire will not result in an increase in keff of the spent fuel racks. Such accidents are the drop of a fuel assembly on top of a rack, the drop of a fuel assembly between rack modules, and the drop of a fuel assembly between rack modules and the pool wall. At McGuire, the spent fuel assembly rack configuration is such that it precludes the insertion of a fuel assembly between rack modules. The placement of an assembly between the rack and the pool wall would result in a lower keff relative to the criticality analysis due to the increased neutron leakage at the spent fuel pool wall because the criticality analysis assumes an infinite array of fuel assemblies. In the case where a dropped fuel assembly in its most reactive condition is dropped onto the spent fuel racks, it is assumed that the rack structure pertinent for criticality is not excessively deformed. For this event, previous accident analysis with unborated water showed that a dropped fuel assembly resting horizontally on top of the spent fuel rack has sufficient separation from the active fuel height of stored fuel assemblies to preclude neutronic interaction. However, three accidents can be postulated which could result in an increase in reactivity in the spent fuel storage pools. The first is the misloading of a fuel assembly. Another postulated accident to be addressed is a significant change in the spent fuel pool water temperature. The third event is a heavy load drop (limited to Region 1 racks). A fuel assembly misload accident relates to the use of administratively controlled storage locations based on fuel assembly type, initial enrichment, burnup and cooling time. The misloading of a fuel assembly constitutes not meeting the enrichment, burnup or cooling time requirements for that administratively controlled location. The result of the misleading is to add positive reactivity, increasing keff toward 0.95. For Region 1, any type of McGuire reactor fuel, with any enrichment (up to 5.00 weight percent) and burnup, can be stored without restriction in the Region 1 racks. As such, there is no possibility of a misloading accident in Region 1. For Region 2, the worst-case misload event involves placing a fresh 5.00 weight percent W-OFA fuel assembly in an empty cell, within the 3 out of 4 Checkerboard/Empty storage configurations. The analysis of this event demonstrates that 1600 ppm is sufficient to ensure that the SFP Region 2 system kff remains below 0.95. A significant change in the spent fuel pool water temperature can be caused by either the loss of normal cooling to the spent fuel pool water which causes an increase in the temperature of the water passing through the stored fuel assemblies or a large makeup to the pool with cold water which could happen if the spent fuel pool were used an as emergency source of borated water. Loss of spent fuel pool cooling causes water density to decrease, typically increases reactivity in the SFP. A decrease in pool temperature causes water density to increase, typically reduces SFP reactivity. However, this event is bounded by the misloading accident, which is much more severe, from a criticality perspective, than a change in SFP water temperature. As far as loads heavier than a fuel assembly are concerned, the largest loads that may be moved over the Region 1 area of the McGuire SFPs are the weir gates. An analysis of the criticality
Attachment 3 Page 8 of 10 consequences of a worst-case weir gate drop on the new Region 1 Boral racks demonstrates that even with up to 9 fuel assemblies crushed by the weir gate into an optimum-reactivity configuration, the maximum achievable 95/95 keff (0.874) is well below the 0.95 subcriticality criterion, when credit is taken for 2475 ppm soluble boron in the SFP. The heavy load drop accident does not need to be considered for Region 2, because the weir gate is not carried directly over Region 2, and thus an end-drop of the gate onto Region 2 - the only type of weir gate drop capable of deforming the storage racks - is not possible. In summary, for each of the accidents evaluated, these analyses determined that the minimum boron concentration required to maintain keff less than or equal to 0.95 is well below the spent fuel pool storage boron concentrations specified in the McGuire Core Operating Limits Report (COLR). The spent fuel pool boron concentration limit currently specified in the COLR is 2675 ppm. Consequently, under the applicable accident conditions, maintaining spent fuel pool boron concentrations within the COLR limit will ensure that the spent fuel storage rack key is < 0.95 when fuel is stored in accordance with the revised spent fuel pool storage configurations and storage criteria (fuel enrichment limits, specified fuel assembly design types, post-irradiation cooling time and burnup requirements) in the proposed changes to TS 3.7.15. The current TS 3.7.15 specifies the requirements for spent fuel pool storage configurations with fuel pool storage criteria involving fuel enrichment and fuel burnup. Consequently, plant operating procedures already include controls to ensure these existing requirements are satisfied. These procedural controls will be revised and maintained as needed under the revised TS 3.7.15. In addition, new controls necessary to ensure that independent administrative confirmation of fuel type and for determining cooling time achieved will be incorporated into plant operating procedures prior to implementation of the proposed TS changes. Note that existing McGuire spent fuel pool storage systems, spent fuel pool cooling systems, fuel handling systems instrumentation and other supporting systems are not modified as a result of this proposed LAR. McGuire TS 4.3 will be revised to increase the maximum allowable U-235 enrichment from 4.75 to 5.00 weight percent that can be stored in the spent fuel storage racks and in the new fuel storage racks, to decrease the boron concentration required to maintain keff
- 0.95 from 850 ppm to 800 ppm, and to eliminate the sub-region designation within Regions 1 and 2 (replace designation Region IA, lB, 2A, & 2B with Region 1 and Region 2).
The criticality analysis for the New Fuel Vault storage racks is performed in accordance with the requirements of 10 CFR 50.68(b). This analysis determined that the New Fuel Vault storage racks can store unirradiated MkBW (with or without axial blankets), W-RFA, and W-STD fuel up to 5.00 weight percent of U-235, with no location restrictions. Fresh W-OFA fuel up to 4.76 weight percent of U-235 may be stored with no location restrictions. The analysis determined that the maximum 95/95 kff if the New Fuel Vault area is flooded with full-density unborated water would be 0.9498 and if flooded with optimum-moderation unborated water, the maximum 95/95 keff would be 0.9618. These results meet the requirements of 10 CFR 50.68(b)(2). As noted in the criticality analysis for the New Fuel Vault storage racks, fuel design type W-OFA is limited to 4.76 weight percent of U-235. Note that only fresh, un-irradiated fuel can be stored in the New Fuel Vault storage racks. Fuel design type W-OFA was utilized in batches 4 through
Attachment 3 Page 9 of 10 9 for both McGuire Units. Further, the current operating cycles for both units do not contain this fuel design type. As such, all fuel design type W-OFA assemblies have been irradiated and, thus are stored in the spent fuel pools. The current fuel design type in use at McGuire is W-RFA. In addition, there are no plans to utilize the W-OFA fuel design type in future operating cycles at MNS. As such, storage of W-OFA assembly within the New Fuel Vault storage racks is highly unlikely. In addition, the design requirements specified by TS 4.3.1.2b and TS 4.3.1.2c provide the necessary regulatory control regarding the safe storage of fuel assemblies within the New Fuel Vault storage racks. These TS requirements will ensure that the nominal enrichment of a W-OFA assembly that would be stored within the New Fuel vault storage racks is equal to or less than 4.76 weight percent of U-235. Conclusion Revision of the McGuire TS's as proposed in this LAR will provide a level of safety comparable to the conservative criticality analysis methodology required by References 1, 2, and 3 of this attachment. Consequently, the health and safety of the public will not be adversely affected by the proposed Technical Specification changes. The bases for these conclusions are as follows:
- 1. Utilizing the revised spent fuel pool storage configurations and revised spent fuel pool storage criteria (fuel enrichment limits, identified fuel assembly design types, cooling time and burnup requirements) specified in the proposed change to TS 3.7.15, the new McGuire Fuel Storage Criticality Analysis demonstrates that a minimum spent fuel storage pool boron credit of 800 ppm would be adequate to maintain the spent fuel storage rack keff < 0.95. This minimum boron concentration is ensured by existing McGuire TS 3.7.14.
- 2. Utilizing the revised spent fuel pool storage configurations and revised spent fuel pool storage criteria (fuel enrichment limits, identified fuel assembly design types, cooling time and burnup requirements) specified in the proposed change to TS 3.7.15, the new McGuire Fuel Storage Criticality Analysis demonstrates that spent fuel storage rack kff would remain below 1.0 with the spent fuel pool fully flooded with unborated water.
- 3. The new McGuire Spent Fuel Pool Criticality Analysis demonstrates that the amount of soluble boron necessary to ensure that the spent fuel rack keff will be maintained less than or equal to 0.95 following a significant change in spent fuel pool temperature or the misleading of a fuel assembly is well below the spent fuel pool storage boron concentrations specified in TS 3.7.14 and in the McGuire Core Operating Limits Report (COLR). The analysis also demonstrates that for the worst-case weir gate drop on the new Region 1 Boral racks, the maximum achievable 95/95 keff is well below the 0.95 subcriticality criterion, when full credit is taken for the soluble boron concentration in the SFP.
References
- 1. Guidance on the Regulatory Requirements for Criticality Analysis of Fuel Storage at Light-Water Reactor Power Plants, Memorandum from L. Kopp (NRC) to T. Collins (NRC), U.S. Nuclear Regulatory Commission, August 19, 1998.
Attachment 3 Page 10 of 10
- 2. USNRC Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, LWR Edition, NUREG-0800, June 1987.
- 3. Title 10 of the Code of Federal Regulations Part 50 Section 68
- 4. Duke Power letter to the NRC, dated October 30, 2002, Response to NRC Request for Additional Information - Boron Dilution Analyses.
- 5. McGuire Nuclear Station, Units 1 and 2 Re: Issuance of Amendments for Spent Fuel Pool (TAC NOs. MB5014 and MB5015)," Letter from R. Martin (NRC) to D. Jamil (Duke),
February 4, 2003.
- 6. McGuire Nuclear Station Re: Issuance of Exemption to 10 CFR 70.24, Criticality Accident Requirements (TAC NOs. M97863, M97864, MB5014 and MB5015), Letter from R. Martin (NRC) to D. Jamil (Duke), January 31, 2003.
ATTACHMENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION EVALUATION
Attachment 4 Page 1 of 4 No Significant Hazards Consideration Evaluation In accordance with 10 CFR 50.92(c), a proposed amendment to an operating license involves no significant hazards consideration if operation of the facility in accordance with the proposed amendment would not: 1) Involve a significant increase in the probability or consequences of an accident previously evaluated; 2) Create the possibility of a new or different kind of accident from any previously evaluated, or; 3) Involve a significant reduction in a margin of safety. This proposed amendment provides revised spent fuel storage criteria based upon fuel type, fuel enrichment, burnup, cooling time and partial credit for soluble boron. In addition, this amendment also allows for storage of fuel assemblies with a nominal enrichment up to 5.0 weight percent of U-235. Finally, this proposed amendment reduces the required soluble boron credit from 850 ppm to 800 ppm. In accordance with the criteria set forth in 10 CFR 50.91 and 50.92, McGuire Nuclear Station has evaluated the proposed Technical Specification changes and determined they do not represent a significant hazards consideration. The following is provided in support of this conclusion.
- 1. Will the change involve a significant increase in the probability or consequence of an accident previously evaluated?
NO The change in the amount of soluble boron specified by Specification 4.3 has no impact on the likelihood or consequences of any previously evaluated accident. This decrease in the soluble boron specified is not considered to be an initiator of any accidents nor does it influence how previously evaluated accidents are mitigated. There is no significant increase in the probability of a fuel assembly drop accident in the spent fuel pools when allowing for credit to be taken for different fuel types, fuel enrichments, burnup, plutonium decay and soluble boron to maintain an acceptable margin of subcriticality in the spent fuel pool. The increase of the nominal fuel enrichment for storage within the spent fuel pool does not increase the likelihood of a fuel assembly drop accident. The method of handling fuel assemblies in the spent fuel pool is not affected by the changes made to the criticality analysis for the spent fuel pool or by the proposed TS changes. The handling of fuel assemblies during normal operation is unchanged, since the same equipment and procedures will be used. The radiological consequences of a fuel assembly drop accident will not be adversely impacted due to taking credit for different fuel types, fuel enrichments, burnup, plutonium decay and soluble boron for criticality control in the spent fuel pool in the criticality analysis. The fission product inventory of individual fuel assemblies will not change significantly as a result of an increase in the nominal fuel enrichment. The criticality analysis showed that the consequences of a fuel assembly drop accident in the spent fuel pools are not affected when allowing for credit to be taken for different fuel types, fuel enrichments, burnup, plutonium decay and soluble boron to maintain an acceptable margin of subcriticality in the spent fuel pool.
Attachment 4 Page 2 of 4 There is no significant increase in the probability of the accidental misloading of spent fuel assemblies into the spent fuel pool racks when allowing for credit to be taken for different fuel types, fuel enrichments, burnup, cooling time and soluble boron to maintain an acceptable margin of subcriticality in the spent fuel pool. Fuel assembly placement and storage will continue to be controlled pursuant to approved fuel handling procedures and other approved processes to ensure compliance with the Technical Specification requirements. These procedures and processes will be revised as needed to comply with the revised requirements which would be imposed by the proposed Technical Specification changes. The proposed amendment decreases the number of different storage configurations specified by Technical Specification 3.7.15, but the number of criteria to consider increases. However, the revised storage requirements and criteria are considered no more complicated then what is currently specified by Technical Specifications. In some ways, the proposed amendment simplifies the process for identifying the placement of fuel assemblies within appropriate locations in the spent fuel pool storage racks. For instance, boundary conditions between storage configurations are significantly simpler. As such, station procedures and processes for appropriate placement of fuel assemblies in the spent fuel pool storage rack will continue to provide additional assurance that an accidental misloading of a spent fuel assembly will not occur. There is no increase in the consequences of the accidental misleading of spent fuel assemblies into the spent fuel pool racks because criticality analyses demonstrate that the pool will remain subcritical following an accidental misloading if the pool contains an adequate soluble boron concentration. Current Technical Specification 3.7.14 ensures that an adequate spent fuel pool boron concentration is maintained in the McGuire spent fuel storage pools. The probabilities of a loss of spent fuel pool cooling or reduction of pool temperature are not influenced by the proposed amendment changes. Fuel storage requirements, nominal fuel enrichment, or the amount of soluble boron present in the spent fuel pool water are not initiators of a loss of spent fuel pool cooling accident or in events resulting in a decrease in the pool water temperature. The consequences of a loss of Spent Fuel Pool cooling is not affected by this change. The concern with this accident is a reduction of spent fuel pool water inventory from bulk pool boiling resulting in uncovering fuel assemblies. Loss of spent fuel pool cooling at McGuire is mitigated in the usual manner by ensuring that a sufficient time lapse exists between the loss of forced cooling and uncovering fuel. This period of time is compared against a reasonable period to re-establish cooling or supply an alternative water source. The heat up rate in the spent fuel pool is a nearly linear function of the fuel decay heat load. The fuel decay heat load will not increase subsequent to the proposed changes since the number of fuel assemblies and the fuel burnups are unchanged. In the unlikely event that all pool cooling is lost, sufficient time will still be available for the operators to provide alternate means of cooling before the onset of pool boiling. Therefore, the proposed changes represents no increase in the consequences of loss of pool cooling. A decrease in pool water temperature from a large emergency makeup causes an increase in water density, increasing reactivity. However, the additional negative reactivity provided by the current boron concentration limit, above that provided by the
Attachment 4 Page 3 of 4 concentration required to maintain kff less than or equal to 0.95 (800 ppm), will compensate for the increased reactivity which could result from a decrease in spent fuel pool water temperature. Because adequate soluble boron will be maintained in the spent fuel pool water, the consequences of a decrease in pool water temperature will not be increased. Current Technical Specification 3.7.14 ensures that an adequate spent fuel pool boron concentration is maintained in the McGuire spent fuel storage pools.
- 2. Will the change create the possibility of a new or different kind of accident from any previously evaluated?
NO Criticality and other related accidents within the spent fuel pool are not new or different types of accidents. They have been analyzed in the Updated Final Safety Analysis Report and in Criticality Analysis reports associated with specific licensing amendments. Specific accidents considered and evaluated include fuel assembly drop, accidental misloading of spent fuel assemblies into the spent fuel pool racks, and significant changes in spent fuel pool water temperature. The accident analysis in the Updated Final Safety Analysis Report remains bounding. The possibility for creating a new or different kind of accident is not credible. In a previous amendment request, taking credit for the soluble boron in the spent fuel pool water for reactivity control in the spent fuel pool was approved by the NRC. For the
-proposed amendment, the spent fuel pool dilution evaluation demonstrates that a dilution of the boron concentration in the spent fuel pool water which could increase the rack kff to greater than 0.95 continues not to be a credible event. The proposed amendment regarding fuel storage requirements, nominal fuel enrichment, and amount of soluble boron in the spent fuel pool water specified by Specification 4.3 will have no effect on normal pool operations and maintenance. There are no changes in equipment design or in plant configuration. The Technical Specification changes will not result in the installation of any new equipment or modification of any existing equipment. Therefore, the proposed amendment will not result in the possibility of a new or different kind of accident.
- 3. Will the change involve a significant reduction in a margin of safety?
NO The proposed Technical Specification changes and the resulting spent fuel storage operating limits will provide adequate safety margin to ensure that the stored fuel assembly array will always remain subcritical. Those limits are based on a plant specific criticality analysis (Attachment 6). This methodology takes partial credit for soluble boron in the spent fuel pool and requires conformance with the following NRC Acceptance criteria for preventing criticality outside the reactor:
- 1) kff shall be less than 1.0 if fully flooded with unborated water which includes an allowance for uncertainties at a 95% probability, 95% confidence (95/95) level; and
Attachment 4 Page 4 of 4
- 2) kff shall be less than or equal to 0.95 if flooded with borated water, which includes an allowance for uncertainties at a 95/95 level.
The criticality analysis utilized credit for soluble boron to ensure k,. will be less than or equal to 0.95 under normal circumstances, and storage configurations have been defined using a 95/95 ke calculation to ensure that the spent fuel rack ker will be less than 1.0 with no soluble boron. Soluble boron credit is used to provide safety margin by maintaining k, less than or equal to 0.95 including uncertainties, tolerances and accident conditions in the presence of spent fuel pool soluble boron. The loss of substantial amounts of soluble boron from the spent fuel pool which could lead to exceeding a klff of 0.95 has been evaluated and shown to be not credible. Accordingly, the required margin to criticality is not reduced. Therefore the proposed changes in this license amendment will not result in a significant reduction in the facility's margin of safety. References
- 1. USNRC Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, LWR Edition, NUREG-0800, June 1987.
- 2. ANS, Design Requirements for Light Water Reactor Spent Fuel Storage Facilities at Nuclear Power Stations, ANSI/ANS-57.2-1983.
- 3. Guidance on the Regulatory Requirements for Criticality Analysis of Fuel Storage at Light-Water Reactor Power Plants, Memorandum from L. Kopp (NRC) to T. Collins (NRC),
U.S. Nuclear Regulatory Commission, August 19, 1998.
- 4. Attached McGuire Criticality Analysis and other attached documentation (including references therein) forming the basis for this license amendment request.
ATTACHMENT 5 ENVIRONMENTAL IMPACT ASSESSMENT
Attachment 5 Page 1 of I Environmental Impact Assessment: The proposed Technical Specification amendment has been reviewed against the criteria of 10 CFR 51.22 for environmental considerations. The proposed amendment will allow credit to be taken for different fuel types, burnup, cooling time and soluble boron to maintain an acceptable margin of subcriticality in the spent fuel pool. Appropriate controls are in place to monitor the soluble boron concentration in the spent fuel pool water and to monitor the placement of different fuel types in the spent fuel storage cells. Consequently, the proposed amendment does not involve a significant hazards consideration, nor increase the types and amounts of effluents that may be released offsite, nor increase individual or cumulative occupational radiation exposures. Therefore, the proposed amendment meets the criteria given in 10 CFR 51.22(c)(9) for a categorical exclusion from the requirement for an Environmental Impact Assessment.
ATTACHMENT 6 MCGUIRE FUEL STORAGE CRITICALITY ANALYSIS
Attachment 6 Page 1 of 48 1 Introduction This analysis examines the criticality aspects of fuel storage in the McGuire new fuel storage vaults (NFVs) and spent fuel pools (SFPs), to ensure that alLpertinent regulatory subcriticality criteria are satisfied for proposed configurations of fuel stored in these areas. The objective of this criticality evaluation is to demonstrate that:
- Fresh fuel up to 5.0 wt % U-235 may be stored in the NFV.
- Fresh or irradiated fuel up to 5.0 wt % U-235 may be stored in the SFP if specific requirements for minimum burnup, fuel assembly design, cooling time, and storage pattern are met.
The NFV criticality evaluation looks at the most reactive fresh fuel assembly designs used at McGuire, to determine whether these assemblies meet the requirements of 10 CFR 50.68 (b) (2,3) when stored in the normally-dry NFVs. The SFP criticality analysis evaluates the Region 1 (flux trap) and Region 2 (egg-crate) storage racks in the McGuire SFPs. These high-density racks originally contained Boraflex poison panels for reactivity holddown. However, ongoing degradation of the Boraflex material in these racks (see, e.g., Reference 1) has limited the effectiveness of this poison material in Region 1 and Region 2. To address the continuing degradation of the Boraflex panels, the McGuire criticality analysis for the SFPs considers "permanent solutions" to these issues. The permanent solutions for Region 1 and Region 2 include the following: Region 1 Re-rack with new storage racks containing Boral poison panels. The old Region 1 Boraflex racks were removed and replaced with new racks containing Boral, which were supplied and installed by Holtec International in mid-2003. The new Region 1 racks have the same dimensions as the old racks, and thus the same storage capacity (286 cells per SFP). As Section 8.1 demonstrates, the new Region 1 racks will allow unrestricted storage of fresh McGuire reactor fuel up to 5.0 wt % U-235. Region 2 Retain the existing egg-crate storage racks, but eliminate credit for any remaining Boraflex poison. The revised evaluation of fuel storage in the Region 2 racks takes credit for cooling time reactivity reduction (due primarily to Pu-241 decay and Gd-155 buildup following the end of reactor irradiation). The analysis also segregates storage burnup requirements by fuel assembly type, to take advantage of lower reactivity associated with certain fuel designs. In addition, the revised Region 2 evaluation takes credit for burnup in storage arrays containing empty cells, thereby allowing increased fuel storage density - such as 3 assemblies with one empty cell - within these arrays. The SFP Region 2 criticality analysis, documented in Section 8.2, employs specific 3-D calculations for all of the fuel storage configurations considered, and meets a rigorous interpretation of 10 CFR 50.68 (b) (4).
Attachment 6 Page 2 of 48 The revised criticality analyses for both Region 1 and Region 2 continue to take partial credit for soluble boron in the SFPs under normal conditions, in accordance with the criteria of 10 CFR 50.68 (b) (4). The general goal in developing the SFP Region 2 storage requirements in this analysis is to model the fuel isotopic inventory as accurately as possible. In order to do this, it is necessary to base the 3-D burned fuel models on actual core operation data, considering axial profiles for burnup, moderator temperature history, fuel temperature history, boron concentration history, and burnable poison exposure. The methods used in quantifying the reactivity effects of these variables, as well as their uncertainties, are discussed in Section 8.2. The results of the calculations performed to generate minimum burnup requirements for Region 2 storage are also documented in that section.
Attachment 6 Page 3 of 48 2 Fuel Storage Facilities at McGuire Figure 1 shows an overhead view of the pertinent fuel storage areas in one of the McGuire fuel buildings. This layout is typical of the two (2) fuel buildings at McGuire. Fresh fuel is first received in the new fuel receiving area and stored temporarily, prior to being removed from its shipping container. Upon removal from the shipping container fuel assemblies are placed in a new fuel storage vault (NFV) location for inspection and then are either kept in the NFV or transferred to the spent fuel pool (SFP) for storage prior to reactor irradiation. Fresh fuel and irradiated reload fuel assemblies are transported to the reactor via the water-filled Fuel Transfer Area. Discharged fuel assemblies from the reactor are also returned to the SFP through the Fuel Transfer Area. Qualified spent fuel assemblies may be loaded into dry storage casks in the Cask Area. Once the dry storage casks are drained, sealed, and decontaminated, they are taken to the on-site independent spent fuel storage installation (ISFSI) for interim storage. The McGuire SFPs are designed to store fresh and irradiated fuel assemblies in a wet, borated environment. The SFPs are divided into two regions: Region 1 and Region 2. The Region 1 storage racks have a flux trap design, with stainless steel rack cells. Boral poison panels are attached to the outsides of each of the Region 1 rack cell walls (with the exception of the outer perimeter cells adjacent to the SFP walls). Figure 2 depicts the storage of four fuel assemblies in the Region 1 cells. McGuire Region 1 is normally used for storage of fresh fuel and irradiated fuel that will be reloaded into the reactor core. Region 2 in the McGuire SFPs is designed to store fuel assemblies that have been permanently discharged from the reactor. Generally these are high-burnup fuel assemblies with low enough reactivity that they can be stored in the tighter Region 2 configuration. Figure 3 shows the McGuire Region 2 storage layout. This design is called the "cell / off-cell" or "egg-crate" pattern because it consists of a tight checkerboarded cluster of stainless steel rack cells. The holes in this pattern are the off-cells, and fuel assemblies are stored in these off-cells as well. Boraflex poison panels - which are not credited in this criticality analysis - are attached to each of the cell walls in the Region 2 racks (again with the exception of the outer perimeter cells adjacent to the SFP walls). Tables 1 and 2 provide the McGuire NFV and SFP rack data important to the criticality modeling of these storage areas.
Attachment 6 Page 4 of 48 4-N aoo 000 000 00a0 DOD 000 0003 0DD 000 000 000 0 00 9000 1000 ODDD 11 New Fuel Storage Vault To Reactor Building Gates New Fuel Receiving Area Figure 1. Overhead View of the McGuire Fuel Building (Typical of Each Unit)
Attachment 6 Page 5 of 48 [ I Fuel I Assembly v-Fuel Assembly 4 [ Fuel Assembly Figure 2. McGuire SFP Region 1 "Flux Trap" Storage Cell Arrangement Figure 3. McGuire SFP Region 2 "Egg-Crate" Storage Cell Arrangement
Attachment 6 Page 6 of 48 Table 1. General Design Information for the McGuire NFV Storage Racks Design Parameter Value-
# of storage locations in each NFV 96 Storage cell pitch (cm) 53.3 Storage cell ID (cm) 22.9 Concrete center dividing wall thickness 81.9 (cm)
Table 2. General Design Information for the McGuire SFP Storage Racks McGuire McGuire Region 1 Region2 Design Parameter Value Value
# of storage locations in each SFP 286 1177 Storage cell pitch (cm) 26.4 23.2 (avg.)
Boral minimum B-10 Loading (g/cm 2 ) 0.020 -- Storage cell wall thickness (cm) 0.19 0.19 Normal SFP water temperature range 50- 150 50 - 150 (OF) Minimum required SFP boron 2675 2675 concentration (ppm)
Attachment 6 Page 7 of 48 3 Fuel Assembly Designs Considered The following fuel types are considered for the McGuire criticality analyses:
- MkBI - this generic fuel type represents the old Oconee 15x15 MkB2, MkB3, and MkB4 fuel assembly designs, which used Inconel spacer grids in the active fuel area. 300 of these assemblies, which operated in the Oconee reactors up through September 1983, were transshipped to McGuire in the 1980s. Currently, 35 of the MkBI assemblies reside in Region 2 of the McGuire Unit 1 SFP, and 265 reside in Region 2 of the McGuire Unit 2 SFP.
- W-STD - this is the standard 17x17 Westinghouse fuel design which was used in the initial cycles (batches 1-3) of both the McGuire reactors. At that time the W-STD design had Inconel grids.
- W-OFA - this is the 17x17 Westinghouse "Optimized Fuel Assembly" design, which had thin rods, Zircaloy grids, and a low total uranium loading. This design was deployed for batches 4 through 9 in both McGuire units.
- MkBW - this is the standard 17x17 Framatome (B&W) fuel design which was modeled after the standard Westinghouse product. The MkBW design contains Zircaloy grids. This fuel type (without axial blankets) was used for batches 10 through 13 in both McGuire reactors.
- MkBWbl - this is the same design as the standard MkBW, but it employs solid, 6-inch, 2.00 wt % U-235 axial blankets at the top and bottom of the active fuel zone. This fuel type was used in McGuire Unit 1, batches 14 to 16, and McGuire Unit 2, batch 14.
- MkBWb2 - this is also the same design as the standard MkBW, but it employs solid, 6-inch, 2.60 wt % U-235 axial blankets at the top and bottom of the active fuel zone. This fuel type was used in McGuire Unit 2, batch 15.
- W-RFA - this is the advanced 17x17 Westinghouse fuel design. It is similar to the MkBW assembly design, and contains Zircaloy grids, but uses annular, 6-inch, 2.60 wt % U-235 axial blankets at the top and bottom of the active fuel zone. This fuel type has been chosen for McGuire Unit 1, batches 17 to present, and McGuire Unit 2, batches 16 to present.
The reason the MkBW fuel design has been split into non-blanketed, 2.00 wt % U-235 blanketed, and 2.60 wt % U-235 blanketed fuel types is that axial blankets have a profound effect on the axial burnup profiles of irradiated fuel assemblies. Note that it is not necessary to consider the blanketed MkBW fuel types for the SFP Region 1 criticality analysis because, as Section 8.1 shows, burnup credit will not be used for Region 1.
Attachment 6 Page 8 of 48 Likewise, the blanketed MkBW fuel types are not considered in the NFV criticality analysis, which assumes the entire fresh MkBW assembly is enriched to 5.0 wt % U-235. Note also that since the 300 MkBI fuel assemblies that were transshipped from Oconee to McGuire are typically stored only in Region 2 of the McGuire SFPs, and because these old fuel assemblies are irradiated (with a maximum enrichment of just 3.20 wt % U-235), this fuel type is not explicitly evaluated in either the SFP Region 1 or NFV criticality analyses. However, the MkBI fuel assemblies are judged to be sufficiently low in reactivity that they may also be stored without restriction in Region 1 of the SFPs. Pertinent design data for all of these fuel types, and the BPRAs they have contained, are provided in Tables 3 and 4. Note that the "WABA" and "Pyrex" BPRAs detailed in Table 4 have a standard '0 B content. The other BPRA designs that have been used - in the MkBI fuel and the MkBW fuel - can have variable 10B content. For this criticality calculation, it is assumed that the MkBI BPRA contained 1.4 wt % B4 C, and the MkBW BPRA contained 4.0 wt % B4 C. These are at or very near to the highest boron concentrations that have been used in the BPRAs for these fuel types. Reference 2 shows that higher BPRA boron concentrations yield higher keff increases in the fuel assemblies that once contained those BPRAs during irradiation. Table 4 also indicates the numbers of BPRA rodlets that have been employed in their corresponding fuel assembly types. Note that as the number of BPRA rodlets increases, so does the amount of fissile plutonium production in the irradiated fuel assembly, as the BPRA rodlets displace moderator from the fuel assembly lattice, resulting in local spectral hardening. For conservatism in the SFP Region 2 criticality analysis, it is assumed that the maximum number of BPRA rodlets (16 with the MkBI assembly, and 24 for all other fuel designs) were present for the fuel assemblies that underwent irradiation with BPRAs inserted.
Attachment 6 Page 9 of 48 Table 3. Design Data for Fuel Types Considered in the McGuire Criticality Analysis W :-;-. W- XRFA W- MkBW OFA RFA blnkt MkBI iSSTD (blb2) Avg fuel density (gkcc) 10.30 10.34 8.02 10.20 10.29 10.36 Fuel pellet OR (cm) 0.3922 0.4096 0.4096 0.4681 0.4096 0.4058 Cladding IR (cm) 0.4001 0.4178 0.4178 0.4790 0.4178 0.4140 Cladding OR (cm) 0.4572 0.4750 0.4750 0.5460 0.4750 OA750 Pin pitch (cm) 1.26 1.26 1.26 1.442 1.26 1.26 Pin array size 17x17 17x17 17x17 l5xl5 17x17 17x17 Guide tube IR (cm) 0.561 0.561 0.561 0.632 0.572 0.572 Guide tube OR (cm) 0.602 0.612 0.612 0.673 0.612 0.612 Full op pressure (bars) 155 155 155 151.7 155 155 Avg power density (W/gU) 41.73 38.10 49.10 31.30 38.30 38.74 Spacer grid material zirc zirc zirc inconel inconel zirc Grid linear density (g/cm) 19.2 18.1 18.1 11.3 14.8 18.7 Table 4. Design Data for Burnable Poison Rod Assemblies (BPRAs) Considered in the McGuire SFP Region 2 Criticality Analysis WY- W- I-MkBW _ OFA RFA MkBI STD (blb2) BPRA type WABA WABA B4C Pjyrex B4C Poison pellet density (gIcc) 2.577 2.577 3.38 2.23 3.10 Poison pellet IR (cm) 0.3531 0.3531 0 0.2413 0 Poison pellet OR (cm) 0.4039 0.4039 0.432 0.4267 0.401 Blo conc (wt %) 1.9374 1.9374 0.2004 0.7118 0.5740 BI, conc (wt %) 8.6282 8.6282 0.8956 3.1702 2.5565 C conc (wt %) 2.9344 2.9344 0.304 - 0.8695 O conc (wt %) 40.720 40.720 46.416 55.218 45.192 Al conc (wt %) 45.780 45.780 52.184 - 50.808 Si conc (wt %) - - - 40.900 -
# of rodlets (fingers) in BPRA 4 to 16 4 to 24 16 9 to 20 4 to 24
Attachment 6 Page 10 of 48 4 Criticality Computer Code Validation The main neutronics codes employed in the criticality analysis are SCALE 4.4/KENO V.a and CASMO-3/SIMULATE-3. These codes are well-suited to SFP and NFV criticality applications, and have been extensively benchmarked to critical experiments and reactor operational data. KENO V.a is a 3-D Monte Carlo criticality module in the SCALE (Reference 3) package. CASMO-3 (Reference 4) is a 2-D transport code that performs fuel criticality and depletion calculations, using a 70-group cross-section library that is based on ENDF/B-IV. CASMO-3 also produces nodal macro-group cross-sections that can be used by SIMULATE-3 (Reference 5), its counterpart 3-D nodal diffusion code, for applications involving arrays of fuel assemblies with varying enrichments or burnups. SCALE 4.4/KENO V.a is used for the evaluation of fresh fuel storage in the NFVs and in Region 1 of the McGuire SFPs, as well as verification of the Checkerboard/Empty configurations considered in the SFP Region 2 analyses. As discussed in Section 8.2, CASMO-3/SIMULATE-3 cannot properly model a true "empty cell" within a Checkerboard/Empty configuration. Instead, CASMO-3 requires some fissile material in order to generate nodal cross-section data for SIMULATE-3. CASMO-3/SIMULATE-3 is used for all SFP Region 2 irradiated fuel cases because this is the only code system qualified by Duke to perform criticality analyses using burnup credit. Note that KENO V.a is capable of doing calculations for burned fuel, using isotopic data produced via the SAS2H module of SCALE 4.4. However, because SAS2H (which was not originally intended for fuel criticality applications) is a 1-D transport code, it is preferable to use a more explicit 2-D transport code such as CASMO-3 for irradiated fuel evaluations. 2-D calculations should more accurately model fuel assemblies that are not radially uniform, such as the fuel types described in Section 3 that contain BPRAs during initial reactor irradiation. The following subsections discuss the benchmarking validation that has been performed for both SCALE 4.4/KENO V.a and CASMO-3/SJMULATE-3. Given the similar types of critical experiments with which these code systems have been validated, the use of these code packages is appropriate for the McGuire NFV and SEP criticality evaluations. As an additional check on the accuracy of both code systems used in these analyses, comparisons were made between results from CASMO-3/SIMULATE-3 and SCALE 4.4/ KENO V.a for several of the same SFP Region 2 storage configurations. These comparisons are presented in Section 8.2.
Attachment 6 Page 11 of 48 4.1 Validation of Benchmark Critical Experiments for SCALE 4.4/KENO V.a Duke Power performed a SCALE 4.4/KENO V.a benchmark analysis of critical experiments to determine calculational biases and uncertainties for both the 44-group and 238-group cross-section libraries included with the SCALE 4.4 package. For McGuire SFP criticality applications, the SCALE 4.4/KENO V.a biases and uncertainties are based on analysis of 58 critical experiments performed by Pacific Northwest Laboratories (see References 6 to 8). The critical experiments evaluated cover a wide range of enrichment (2.35 and 4.31 wt % U-235), and include both over- and under-moderated lattices. For the NFV criticality analyses, a subset of 41 of the 58 critical experiments described above was employed. Because the NFV analysis models fresh fuel at high (4.76 to 5.00 wt % U-235) enrichments, the 41 critical experiments were all at the highest enrichment (4.31 wt % U-235) used in the PNL experiments. The results from the benchmark analyses indicate that the 238-group cross-section library yields the more consistent results (i.e., smaller variations in reactivity bias) across the ranges of moderation and enrichment considered. Therefore, the 238-group cross-section library is used for all the SCALE 4.4/KENO V.a computations performed in this criticality analysis. The 41 critical experiments used for the NFV analysis yielded a benchmark calculational bias of +0.0061 Ak (average under-prediction of keff) and an uncertainty of +/-0.0071 Ak. The 58 experiments used in the benchmarking for the McGuire SFP criticality analyses resulted in a calculational bias of +0.0064 Ak and an uncertainty of +/-0.0066 Ak. These biases and uncertainties are used in determining the total bounding 95/95 system keffs for each NFV or SFP storage configuration analyzed with SCALE 4.4/KENO V.a. 4.2 Validation of Benchmark Critical Experiments for CASMO-3/SIMIULATE-3 For all of the SFP Region 2 irradiated-fuel criticality evaluations, the CASMO-3/ SIMULATE-3 code set is used. All CASMO-3 calculations will be carried out with the fine-energy-group (70-group) neutron cross-section library available with that code. Duke Power has performed a benchmark analysis of 10 B&W critical experiments with CASMO-3 and SIMULATE-3. These B&W critical experiments (Reference 9) were specifically designed for reactivity benchmarking purposes. Results from these 10 B&W critical benchmark cases yielded a calculational bias of -.0015 Ak (average over-prediction of keff) and an uncertainty of +/-0.0121 Ak. Even though SIMULATE-3 tends to over-predict keff, its negative bias will be conservatively ignored. The uncertainty, however, will still be used in computing the overall 95/95 kffs for the McGuire SFP Region 2 irradiated-fuel storage configurations described in Section 5.
Attachment 6 Page 12 of 48 5 Proposed Storage Configurations for the McGuire NFVs and SFPs Figure 4 shows the various NFV, SFP Region 1, and SFP Region 2 fuel storage configurations that are specifically evaluated in Sections 7 and 8. The minimum burnup limits for SFP Region 2 storage, in accordance with these configurations, are determined in Section 8.2. The symbols in the repeating patterns of Figure 4 correspond to the following storage types: U - Fuel assembly qualified for Unrestricted storage in the NFV, SFP Region 1, or SFP Region 2 R- Fuel assembly qualified for Restricted storage in the SFP Region 2 F- Fuel assembly qualified for Filler storage in the SFP Region 2 C- Fuel assembly qualified for Checkerboard storage in the SFP Region 2 E- Empty storage location Unrestricted Storage - NFV, SFP Region 1, or SFP Region 2 U U U U U U U U U U U U
-U u___
2/4 RestrictedlFiller Storage - SFP Region 2 R F R F F Fi R R F R F F R F R 3/4 Checkerboard/Empty Storage - SFP Region 2 C C C C C EC-E C C C C C CJE Figure 4. McGuire SFP Fuel Storage Configurations Considered in this Analysis
Attachment 6 Page 13 of 48 6 Computation of the Maximum 95/95 kff For every fuel assembly design, fuel enrichment, cooling time, and storage region combination that is considered in the scope of the McGuire SFP and NFV criticality analyses, a nominal keff is calculated. This kff is only the base value, however. A total keff is determined by adding several pertinent reactivity biases and uncertainties, to provide an overall 95 percent probability, at a 95 percent confidence level (95/95), that the true system keff does not exceed the 95/95 keff for that particular storage condition. The total 95/95 keff equation has the following form: keff=knominaj+ E B. + where: kno.,nai is the kff computed for the nominal case being considered. B, is a pertinent bias, as indicated in Table 5. ks, is the pertinent 95/95 independent uncertainty on knnw , as indicated in Table 5. Table 5 lists the various biases and uncertainties that are considered in the McGuire NFV and SFP criticality analyses. Each of these biases and uncertainties is discussed in more detail below:
- Benchmark Method Bias This bias is determined from the benchmarking of the code system used (SCALE 4.4/KENO V.a or CASMO-3/SIMULATE-3), and represents how much the code system is expected to overpredict (negative bias) or underpredict (positive bias) the "true kefe" of the physical system being modeled. The critical experiment benchmarks for these codes are discussed in Sections 4.1 and 4.2. The bias for SCALE 4.4/KENO V.a with its 238-group cross-section library is +0.0061 Ak for NFV applications, and +0.0064 Ak for SFP applications. The bias for CASMO-3/
SIMULATE-3 with its 70-group cross-section library is 0.0015 Ak. Note that negative biases are conservatively ignored in this calculation, per Reference 10.
Attachment 6 Page 14 of 48
- Fixed Poison Self-Shielding Bias This reactivity penalty accounts for the slight self-shielding effects associated with the clustering of boron carbide particles in the SFP Region 1 Boral panels.
The self-shielding bias was conservatively estimated for the Region 1 Boral replacement panels to be +0.0010 Ak.
- Cooling Time / Enrichment Interpolation Error Section 8.2 discusses this reactivity penalty, which accounts for the maximum difference in kff between a minimum burnup limit "estimate" using the interpolation technique specified in that section, and the "true" burnup limit that specific evaluation at that enrichment and cooling time would yield. That section determines a bounding error of +0.00036 Ak for interpolation between the tabulated SFP Region 2 minimum burnup data points (see Tables 18 through 21).
- Benchmark Method Uncertainty This uncertainty is determined from the benchmarking of the code system used (SCALE 4.4/KENO V.a or CASMO-3/SIMULATE-3), and is a measure of the expected variance (95/95 one-sided uncertainty) of predicted reactivity from the "true keff" of the physical system being modeled. The critical experiment benchmarks for these codes are discussed in Sections 4.1 and 4.2. The method uncertainty for SCALE 4.4/KENO V.a with its 238-group cross-section library is
+/-0.0066 Ak for SFP applications and +/-0.0071 Ak for NFV applications. The uncertainty for CASMO-3/SIMULATE-3, with its 70-group cross-section library, is +/-0.01211 Ak.
- Monte Carlo Computational Uncertainty For all the nominal SCALE 4.4/KENO V.a computations performed in this analysis to determine 95/95 keffs, the Monte Carlo computational uncertainty is equal to either L.752*no.mi.A, (if 600 neutron generations are run), or L.7 7 8*GnomI (if 400 neutron generations are run). The acmoming factor is the calculated standard deviation of knh0 0 w (the nominal kff for that particular case).
The 1.752 or 1.778 multiplier is the one-sided 95/95 tolerance factor for 600 or 400 neutron generations, respectively. Each of the SCALE 4.4/KENO V.a cases in the SFP Region 1 and NFV calculations counted 400 neutron generations, and the SFP Region 2 calculations used 600 neutron generations.
Attachment 6 Page 15 of 48
- Mechanical Uncertainties The "mechanical uncertainty" represents the total reactivity uncertainty contributions of various independent fuel manufacturing-related and storage rack-related mechanical uncertainty factors. These factors include reactivity effects for variations in fuel enrichment, fuel pellet diameter, fuel density, cladding dimensions, storage rack dimensions and material tolerances, fixed poison panel width, fuel assembly positioning within the storage cell, etc. The following bounding total mechanical uncertainties have been determined:
NFV (no boron in full-density water): +/-0.0073 Ak NFV (no boron in optimum-density water): +/-0.0079 Ak SFP Region 1 (no boron in SFP water): +/-0.00973 Ak SFP Region 1 (310 ppm boron in SFP water): +/-0.01324 Ak SFP Region 2 (no boron in SFP water): +/-0.01110 Ak SFP Region 2 (800 ppm boron in SFP water): +/-0.01247 Ak Burnup Computational Uncertainty This burnup-related uncertainty quantifies, in a global sense, the ability of the CASMO-3/SIMULATE-3 codes to accurately determine the isotopic content, and hence kff, of a collection of irradiated assemblies in the McGuire reactors, assuming the actual average burnup of the fuel in the reactor core is the same as the average burnup of the SIMULATE model for that reactor core. Duke Power has determined a bounding McGuire CASMO-3/SIMULATE-3 burnup computational reactivity uncertainty of +/-{0.00454
- BU / 50}Ak, where BU is the average burnup of the system modeled, in GWD/MTU.
- Burnup Measurement Uncertainty This uncertainty represents the reactivity penalty associated with difference between the measured burnup and the code-predicted burnup. Measured burnups, which are used for Technical Specification verification, have many sources of instrumentation error that can contribute to overall measurement inaccuracies.
Section 8.2 discusses the method used to calculate a bounding measured burnup reactivity uncertainty for fuel storage in Region 2 of the McGuire SFP. The analysis of predicted and measured core follow data yields a burnup measurement uncertainty of +/-:0.00125 Ak.
Attachment 6 Page 16 of 48
- Axial Profile Uncertainty This uncertainty represents the bounding reactivity penalty associated with differences between the kff calculated using the average "estimated" axial burnup and history profiles for a particular fuel assembly, and the kff calculated using the actual axial burnup and history profiles for that fuel assembly. Section 8.2 discusses the method used to determine average "estimated" profiles, and how to quantify the axial profile uncertainty for McGuire SFP Region 2 irradiated fuel applications. An analysis of the keff differences for a large database of "estimated" and actual axial profiles has determined a bounding axial profile uncertainty of +/-0.00305 Ak.
Table 5. Pertinent 95/95 Biases and Uncertainties to be Considered in the McGuire NFV and SFP Criticality Analysis Include for Include for Include for Biases NFV SIFPTRegion iSFP2Region2 Anialyses? Analyses:? Analyses?# Benchmark Method Bias f 1 _ Fixed Poison Self-Shielding Bias _ Cooling Time I Enrichment Interpolation Error Uncertainties Benchmark Method Uncertainty I I I Monte Carlo Computational Uncertainty Mechanical Uncertainties I I I Burnup Computational Uncertainty _ _ Burnup Measurement Uncertainty Axial Profile Uncertainty _
Attachment 6 Page 17 of 48 7 McGuire New Fuel Storage Vault Criticality Analysis To allow storage of fuel in the normally-dry environment of the NFVs, the following requirements of 10 CFR 50.68 (b) (2) and (3) must be satisfied:
"The estimated ratio of neutron production to neutron absorption and leakage (k-effective) of the fresh fuel in the fresh fuel storage racks shall be calculated assuming the racks are loaded with fuel of the maximum fuel assembly reactivity and flooded with unborated water and must not exceed 0.95, at a 95 percent probability, 95 percent confidence level. ...
If optimum moderation of fresh fuel in the fresh fuel storage racks occurs when the racks are assumed to be loaded with fuel of the maximum fuel assembly reactivity and filled with low-density hydrogenous fluid, the k-effective corresponding to this optimum moderation must not exceed 0.98, at a 95 percent probability, 95 percent confidence level." The McGuire NFVs are described in Section 2. The following assumptions and simplifications are made in performing the criticality analysis of the NFVs:
- 1) All fuel designs that have been or are projected to be used in the McGuire reactors are evaluated. This includes the W-STD, W-OFA, MkBW, and W-RFA fuel assembly types described in Section 3.
- 2) A simplified 3-D axial model of the fuel assembly is employed. Only the active fuel region is modeled - the top and bottom nozzles are ignored.
- 3) All fuel is unirradiated. The W-OFA fuel assembly design is limited to 4.76 wt % U-235. All other fuel assemblies considered are allowed to be enriched up to 5.00 wt % U-235.
- 4) The fuel assemblies are stored without any location restrictions in the NFVs, in accordance with Figure 4 in Section 5.
- 5) No credit is taken for spacer grid material in the active fuel regions of the fuel assemblies.
- 6) No credit is taken for any burnable poison assemblies (BPRAs), control rods, or other neutron poisons that may be inserted in the fuel assemblies.
Using the pertinent reactivity biases and uncertainties described in Section 6, the SCALE 4.4/KENO V.a analyses for fuel storage in the NFVs yield the following maximum 95/95 keffs:
- NFV flooded with full-density unborated water: 0.9498
- NFV flooded with optimum-moderation unborated "water": 0.9618
Attachment 6 Page 18 of 48 Table 6 presents the various biases and uncertainties that comprise the NFV maximum 95/95 kffs. Table 6. Maximum 95/95 kffs for Fuel Storage in the McGuire NFVs (No Boron in "Water" flooding NFV) NFV0looded with -NFV flooded wit 1full-density i o- ptimum-density _-: _;y: : .- _ tE.--_ 40_: e;:i_ : _ :t04 L:-_ t _ .m oderator moderator Maximum Nominal kff 0.9329 0.9446 Biases Benchmark Method Bias 0.0061 0.0061 Fixed Poison Self-Shielding Bias -- -- Cooling Time I Enrichment Interpolation Error Uncertainties Benchmark Method Uncertainty 0.0071 0.0071 Monte Carlo Computational Uncertainty 0.0035 0.0032 Mechanical Uncertainties 0.0073 0.0079 Burnup Computational Uncertainty -- -- Burnup Measurement Uncertainty Axial Profile Uncertainty Maximum 95195 kff 0.9498 0.9618
Attachment 6 Page 19 of 48 8 McGuire Spent Fuel Pool Criticality Analysis For storage of fuel in the McGuire SFPs, the following requirements of 10 CFR 50.68 (b) (4) must be satisfied:
"... If credit is taken for soluble boron, the k-effective of the spent fuel storage racks loaded with fuel of the maximum permissible fuel assembly reactivity must not exceed 0.95, at a 95 percent probability, 95 percent confidence level, if flooded with borated water, and the k-effective must remain below 1.0 (subcritical), at a 95 percent probability, 95 percent confidence level, if flooded with unborated water."
In addition, for evaluations of burned fuel in SFP criticality analyses, Reference, 10 provides the following general criteria:
"A reactivity uncertainty due to uncertainty in the fuel depletion calculations should be developed and combined with other calculational uncertainties."
"A correction for the effect of the axial distribution in burnup should be determined and, if positive, added to the reactivity calculated for uniform axial burnup distribution."
The following assumptions and bases are employed for the McGuire SFP criticality evaluations:
- 1) Partial soluble boron credit is used in both the Region 1 and Region 2 criticality evaluations. These analyses adhere to the regulatory subcriticality criteria defined in 10 CFR 50.68 (b) (4), as well as the guidance provided in Reference 10.
- 2) All McGuire Region 2 criticality calculations are performed in three dimensions, with 24 axial fuel segments analyzed. The 3-D model includes top and bottom axial reflectors containing a mix of water, steel, and Zircaloy. Reference 11 supports the assumption that 24 axial fuel segments are more than sufficient to accurately capture the reactivity effects associated with axial variations in fuel burnup. Extensive historic and projected 3-D burnup, temperature, boron, and burnable poison data are employed to appropriately quantify the isotopic content of the fuel assembly designs considered.
- 3) McGuire Region I calculations are performed in 2-D, with perfect axial reflection. This is acceptable, because only fresh fuel is considered in the criticality evaluation for the Region 1 racks. It is also conservative, because it ignores axial leakage.
Attachment 6 Page 20 of 48
- 4) Credit is taken for the fixed Boral poison material within the new Region 1 SFP racks.
- 5) It is conservatively assumed that no Boraflex remains in the McGuire SFP Region 2 storage racks. In reality, some Boraflex remains in the Region 2 racks, which are not currently being replaced. This assumption is part of the "permanent solution" proposed in this amendment to the licensing basis for fuel storage in the McGuire SFPs.
- 6) For one of the storage configurations defined for the McGuire SFP Region 2, a water hole (empty cell) is used in one out of every four cells. This water hole cannot be modeled directly with CASMO-3, which requires at least a trace amount of fissile material in each unit cell. Thus, a low-enrichment, low-loading "water hole" is modeled with CASMO-3 to allow the overall storage configuration to be evaluated with SIMULATE-3. This is a conservative approach, as comparisons with KENO V.a in Section 8.2 show.
- 7) No credit is taken for any short-lived Xe-135 poisons in the fuel stored in the SFPs, consistent with Reference 10.
- 8) In the McGuire SFP Region 2 analysis, credit is taken for the spacer grids in each fuel assembly design considered. The standard CASMO-3 grid model, which homogenizes the grid material-into the coolant surrounding the fuel assembly, is used to account for the effects of the grids. This is the same model as that used in the McGuire reactor core design and core follow calculations.
- 9) For accident conditions, the McGuire SFP is fully-flooded (full-density water) at the minimum McGuire SFP boron concentration as specified in the Core Operating Limits Report (2675 ppm). Per the double contingency principle (see Reference 10), it is allowable to assume that the minimum boron concentration is present in the event of an accident condition - such as a misloaded fuel assembly - in the McGuire SFP.
- 10) Credit for the reactivity reduction associated with fuel burnup and cooling time is employed for SFP Region 2 storage in this calculation. The reactivity reduction with cooling time is primarily due to Pu-241 decay
(-14.3 yr half-life), and Gd-155 buildup (via Eu-155 decay with - 4.7 yr half-life).
Attachment 6 Page 21 of 48 8.1 SFP Region 1 Criticality Analysis Section 6 documented the biases and uncertainties pertinent to the Region 1 Boral storage racks. Note that the biases and uncertainties related to fuel assembly burnup are not applicable for the Region 1 criticality analysis because the fuel storage requirements for Region 1 do not take credit for burnup. The new Region 1 Boral storage rack design is almost identical to the previous Region 1 Boraflex rack design. The pertinent design information used as input to the Region 1 criticality analyses is provided in Tables 2 and 3. Each of the McGuire SFP Region 1 criticality computations considers the SFP water temperature at both 32 OF and 212 'F. This ensures the maximum-reactivity condition is properly determined for every case. According to the McGuire UFSAR Section 9.1.3.1.1, SFP water temperatures will not exceed 150 'F under "normal" conditions, or 212 'F under "accident" conditions. The normal-condition Region 1 criticality calculations are performed with no boron in the SFP water [to satisfy the 95/95 keff < 1.0 criterion of 10 CFR 50.68 (b) (4)], and with 310 ppm of soluble boron credit (to satisfy the 95/95 kff < 0.95 criterion of the same regulation). Since the Region 1 normal-condition calculations are already performed at the conceivable extremes of SFP water temperature, the only Reference 10 accident conditions that need to be evaluated are the fuel assembly misload and fuel assembly drop events. In addition, per NUREG-0612, the criticality consequences of dropping a load heavier than a fuel assembly on the Region 1 racks are considered. All of these accident conditions are allowed to take full credit for the minimum required boron concentration in the McGuire SFPs. That minimum boron concentration, controlled though the COLR per McGuire TS 3.7.14, is currently 2675 ppm. As discussed in Section 3, specific Region 1 criticality calculations are performed for the W-STD, W-OFA, W-RFA, and MkBW fuel types, using SCALE 4.4/KENO V.a. These cases consider fresh 5.0 wt % U-235 fuel, stored in the Unrestricted Region 1 configuration shown in Figure 4. The maximum nominal key in unborated SFP water is computed to be 0.9631. The maximum Region 1 95/95 kff from this case, as shown in Table 7, is 0.9829. This includes the pertinent biases and uncertainties identified in Section 6. In unborated SFP conditions, then, the maximum 95/95 kff for Region 1 storage remains below 1.0. The SCALE 4.4/KENO V.a calculations also show that if credit is taken for 310 ppm soluble boron in the McGuire SFP, the maximum 95/95 kff for Region 1 fuel storage is reduced below 0.95 for all normal conditions.
Attachment 6 Page 22 of 48 These results demonstrate that, in the new McGuire Region 1 SFP racks, Unrestricted storage of any type of fresh McGuire reactor fuel up to 5.0 wt % U-235 meets the boron credit subcriticality criteria of 10 CFR 50.68 (b) (4) for normal storage conditions. Three Region 1 accident conditions were identified earlier in this section - the fuel assembly misload, assembly drop, and heavy load drop events. Because any type of McGuire reactor fuel, with any enrichment and burnup, can be stored without restriction in the new Region 1 racks, there is no possibility of a misloaded assembly. The fuel assembly drop accident, from a criticality perspective, may be considered in the same category as a single isolated fuel assembly stored in water. That is because a dropped fuel assembly dropped onto the McGuire storage racks will rest far enough above the active fuel zones of the normally stored fuel assemblies that it is effectively isolated. SCALE 4.4/KENO V.a was used to model a single, fresh, 5.0 wt % U-235 assembly of the most reactive type (W-OFA), surrounded by 30 cm of water in all directions. With only 170 ppm of boron credit taken for this "accident" condition, the largest 95/95 keff (at 32 0F) was only 0.916, well below the 0.95 subcriticality criterion. As far as loads heavier than a fuel assembly are concerned, the largest loads that may be moved over the Region 1 area of the McGuire SFPs are the weir gates (see Figure 1). An analysis of the criticality consequences of a worst-case weir gate drop on the new Region 1 Boral racks demonstrates that even with up to 9 fuel assemblies crushed by the weir gate into an optimum-reactivity configuration, the maximum achievable 95/95 kff -(0.874)-is well below the 0.95 subcriticality-criterion, when credit is taken for 2475 ppm boron in the SFP.
Attachment 6 Page 23 of 48 Table 7. Maximum 95/95 kff for Fuel Storage in Region 1 of the McGuire SFPs (No Boron in SFP Water) SFP Region 1 Storage Maximum Nominal kff 0.9631 Biase Benchmark Method Bias 0.0064 Fixed Poison Self-Shielding Bias 0.0010 Cooling Time / Enrichment Interpolation Error Uncertainties Benchmark Method Uncertainty 0.0066 Monte Carlo Computational Uncertainty 0.0038 Mechanical Uncertainties 0.0097 Burnup Computational Uncertainty -- Buroup Measurement Uncertainty Axial Profile Uncertainty Maximum 95195 kff 0.9829
Attachment 6 Page 24 of 48 8.2 SFP Region 2 Criticality Analysis The first step in analyzing the McGuire SFP Region 2 racks is to assess the validity of the CASMO-3/SIMULATE-3 models that are employed to determine fuel burnup requirements for storage in Region 2. The CASMO-3/SIMULATE-3 models use the fuel, burnable poison, and SFP Region 2 rack data summarized in Tables 2, 3, and 4. Figure 3 shows a heterogeneous (actual) representation of the Region 2 egg-crate racks. Note the heterogeneous Region 2 racks have their storage cell walls very close to the centerline between assemblies stored in neighboring "cells" (storage locations with cell walls) and "non-cells" (storage locations without cell walls). To simplify the analysis of the Region 2 racks with the nodal SIMULATE-3 code, it is desirable to use a homogeneous CASMO-3 model of the Region 2 racks. A homogeneous rack model allows all nodal interfaces between adjacent fuel assemblies to look the same. To accomplish this, the homogeneous Region 2 model for this analysis adjusts the Figure 3 cell wall location to be at the midpoint between stored assemblies, making neighboring cells identical to each other. An individual cell within the homogeneous Region 2 rack model would then have a stainless steel wall approximately half the actual cell wall thickness at its outer edge. Table 8 shows the kff results from SCALE 4.4/KENO V.a calculations for heterogeneous and homogeneous Region 2 rack models storing different fuel types and enrichments, and also provides the keffs from the equivalent CASMO-3 homogeneous model. The results in-Table 8Vindicate that the homogeneous-Region-2 rack model is valid, and yields essentially the same keffs as the heterogeneous model. Likewise, the CASMO-3 computations agree very well with the SCALE 4.4/KENO V.a results. Table 8. McGuire SFP Region 2 - Fresh Fuel keff Comparisons between Homogeneous and Heterogeneous KENO V.a Models, and Homogeneous CASMO-3 Models {O ppm boron in SFP water) SFP W-OFA W-OFA W-STD W-STD) water 2.00 5.00 2.0[500 Region 2 Storage Model Itemp wt% wt % it% i%wt % C U-. - 02iXi3-U-235 U-235 U-235,: KENO V.a Heterogeneous model 150 1.1862 1.4423 1.1947 1.4411 KENO V.a Homogeneous model 150 1.1840 1.4441 1.1948 1.4405 CASMO-3 Homogeneous model 150 1.1877 1.4424 1.1960 j 1.4395 Section 1 mentioned that separate Region 2 fuel storage burnup limits would be determined for each fuel assembly type considered. These seven (7) fuel types, as well as the discrete BPRAs they have contained during irradiation, were described in Section 3. Note that the concept of separate Technical Specification storage limits for different fuel types does have a precedent. In Reference 12, the NRC approved separate sets of burnup requirements for storage of MkB10 and MkBI 1 fuel in the Oconee SFPs.
Attachment 6 Page 25 of 48 As first noted in Section 1, it is desired to determine, for each of the seven fuel types described in Section 3, the "average" axial distributions of the following five (5) reactor irradiation environment history variables that affect the isotopic composition and, hence, reactivity, of irradiated fuel:
- exposure (burnup)
- moderator temperature history
- fuel temperature history
- soluble boron history
- burnable poison (BPRA) exposure history Fortunately, McGuire has an extensive repository of reactor core follow data available.
These sources provide the complete SIMULATE-3 irradiation histories for McGuire Units 1 and 2, from their initial cycles to the present. The McGuire core follow information provides a comprehensive database of axial distributions for the W-STD, W-OFA, MkBW, MkBWbl, and MkBWb2 fuel types described in Section 3. For the old Oconee "MkBr' fuel stored in the McGuire SFPs, representative single-cycle Oconee core data are used. Individual fuel assembly fuel temperature histories are not available in these data, so a core-average fuel temperature history profile is used as the "average" for all the MkBI fuel. It is notable that most of the MkBI fuel assemblies in the McGuire SFP have cooling times well over 20 years (average - 25.6 years as of June 2003). However, the minimum burnup requirements documented for all fuel types (later in this section) are tabulated only to a maximum cooling time of 20 years, and no extrapolation is to be performed beyond 20 years. The additional uncaptured reactivity reduction for the actual MkBI fuel assemblies that have cooled significantly longer than 20 years may be considered further conservatism for the Oconee MkBI fuel model. As Reference 2 mentions, the post-irradiation reactivity of fuel assemblies continues to decrease for around 100 years, after which the reactivity begins to increase again very gradually, due to decay of longer-lived poisonous isotopes such as Am-241 and Pu-240. However, calculations show that for McGuire Region 2 storage of a spent fuel assembly, more than 500 years must elapse before the fuel assembly again achieves the reactivity it had after 20 years of cooling. For the current W-RFA fuel that has been recently implemented at McGuire, actual high-burnup core follow data are not yet available. Therefore equilibrium-cycle projections for W-RFA fuel are used to provide the most realistic axial profile data for this fuel type. Use of the projected profiles for W-RFA fuel is judged to be conservative, due to the fact that current and projected W-RFA burnable poison exposure histories are almost all attributable to irradiation of fuel with integral burnable poisons (IFBAs). References 2 and 13 demonstrate that IFBAs have a much smaller effect on Pu isotopic production than discrete BPRAs, primarily because integral poisons do not displace moderator as the discrete BPRAs do. However, as noted in Section 3 and Table 4, the criticality analysis
Attachment 6 Page 26 of 48 of the W-RFA fuel type considers all of its burnable poison exposure histories to be due to discrete, "WABA"-type BPRAs. The final profile data histories for each of the seven fuel types considered in the SFP Region 2 evaluation are compiled and then further segregated into four different burnup "groups." Table 9 shows the burnup groups that are used for this analysis, and the ranges of average burnup data that are used to determine "average" 24-level profiles for those bumup groups. The following procedure for generating the average profiles is used:
- Collect all the 24-level axial profile data (normalized burnup, fuel temperature history, moderator temperature history, boron concentration history, and BPRA exposure history) for each of the seven fuel types. Note that a normalized burnup profile is determined by taking the actual burnup profile of an assembly and dividing each axial level by the average (2-D) burnup of that fuel assembly.
- Sort these profiles by fuel type, and then by average (2-D) burnup.
- Determine an average value, at each axial level, of each history parameter, for the fuel type being considered, using an average of the data that fall within the bumup ranges listed in the rightmost column of Table 9. Note the 2.5 GWD / MTU overlap beyond the boundaries of the final bumup groupings is used to enhance the "smoothness" of the transition between one final burnup group and the next.
To help ensure conservatism in the overall averaging of these axial profiles, with each fuel type the individual burnup "group" axial profile for BPRA exposure that yields the highest average (2-D) BPRA exposure is applied to all burnup groups for that fuel type. The 24-level "average" axial profiles resulting from the above process are shown in Tables 10 through 16, for each of the fuel assembly designs described in Section 3. Note that the grouping of axial profiles into applications within burnup ranges helps to simplify the overall Region 2 criticality analysis, and is similar to the axial profile burnup-grouping concept documented in Reference 11. Table 9. Grouping of 24-Level Axial Profile Data by Average Burnup Range Average Bunup Range of Aal Profile Data:
"Group" Average Burnup Range Used to Determine the "Average" History Profiles withinthis Group"~
< 20 GWD / MTU 0 to 22.5 GWD / MTU 20 - 30 GWD / MTU 17.5 to 32.5 GWD / MTU 30 - 40 GWD / MTU 27.5 to 42.5 GWD I MTU
> 40 GWD / MTU 37.5 to max GWD / MTU
Attachment 6 Page 27 of 48 Table 10. Average 24-Level Axial Profiles for MkBW Fuel T T - BPRA Normalized Moderator Temperature Fuel Temperature ' . : i:::E' iSoluble Boron11:- (GWD/ Axial lurnnup _IS HStO t1vCC) I Histor t() , Coneentranon History 1 t MIU) axial BU BU BU BU BU nU BU nU BU BU BU BU BU BU nU nU nU level < 20 20.30 30-40 > 40 < 20 20-30 30-40 > 40 < 20 20.30 30-40 >40 < 20 20-30 30-40 >40 >0 1 (top) 0.588 0.592 0.585 0.590 0.641 0.651 0.661 0.660 28.904 28.410 27.765 27.728 358.206 409136 499.441 517.457 0.000 2 0.774 0.780 0.788 0.793 0.644 0.654 0.664 0.663 29.769 29.168 28.438 28.372 362.422 416.774 515.777 533.276 11.655 3 0.900 0.905 0.915 0.919 0.648 0.657 0.667, 0.666 30.278 29.607 28.796 28.711 368.792 424.298 526.971 544.520 15.995 4 0.977 0.982 0.994 0.997 0.651 0.661 0.670- 0.670 30.521 29.823 28.981 28.893 375.551 431i433 535.814 553.255 16.780 5 1.021 1.025 1.036 1.038 0.656 0.664 0.674 0.673 30.619 29.912 29.054 28.963 381.087 437.255 543.131 560.783 17.589 6 1.028 1.031 1.042 1.043 0.660 0.668 0.678 0.677 30.569 29.871 29.024 28.937 385.243 441.680 548.840 566.692 17.751 7 1.039 1.042 1.050 1.051 0.665 0.672 0.681 0.681' 30.551 29.857 29.007 28.916 388.505 445.086 553.228 571.420 17.920 8 1.063 1.065 1.072 1.072 0.670 0.677 0.685 0.685 30.594 29.895 29.037 28.945 391.043 447.711 556.619 575.107. 18.344 9 1.065 1.067 1.073 1.073 0.674 0.681 0.689 0.689 30.559 29.864 29.009 28.917 392.683 449.414 558.894 577.637 18.394 10 1.059 1.060 1.063 1.063 10.679 0.685 0.693 0.693 30.497 29.806 28.945 28.849 393.656 450.358 560.202 579.298 18.298 11 1.070 1.070 1.075 1.074 10.684 0.690 0.697 0.697 30.504 29.814 28.960 28.866 394.308 451.019 561.189 580.381 18.483 12 1.081 1.080 1.083 1.082 10.689 0.694 0.700 0.700 30.519 29.825 28.958 28.860 394.578 451.187 561.460 580.909 18.684 13 1.070 1.069 1.070 1.069 0.694 0.698 0.704 0.704 30.450 29.760 28.893 28.793 394.305 450.817 561.076 580.703 18.497 14 1.082 1.081 1.080 1.078 0.698 0.702 0.707 0.707Q 30.476 29.781 28.903 28.799 393.956 450.312 560.434 580.227 18.716 15 1.099 1.096 1.095 1.092 0.702 0.706 ,0.711 0.711 30.523 29.819 28.923 28.816 393.283 449.430 559.288 579.247 19.007 16 1.095 1.092 1.091 1.088 0.706 0.710 0.714 0.714 30.495 29.791 28.894 28.785 391.967 447.937 557.494 577.533 18.948 17 1.085 1.081 1.077 1.075 0.710 0.713 0.717 0.717 30.446 29.742 28.836 28.722 390.072 445.796 554.882 575.100 18.775 18 1.099 1.096 1.093 1.089 0.715 0.717 0.721 0.721 30.501 29.787 28.873 28.758 387.850 443.316 551.867 572.145 19.017 19 1.099 1.096 1.091 1.087 0.719 0.721 0.724 0.724 30.506 29.787 28.860 28.741 385.030 440.130 547.836 568.294 19.064 20 1.081 1.076 1.069 1.066 0.723 0.725 0.727 0.727 30.447 29.726 28.785 28.658 380.922 435.653 1542.535 563.379 18.697 21 1.072 1.067 1.059 1.056 0.727 0.728 0.731: 0.731 30.432 29.709 28.760 28.628 376427 430.734 536.622 557.974 18.519 22 1.025 1.021 1.010 1.009 0.731 0.732 0.734 0.734 30.262 29.561 28.621 28.487 375.385 428.459 530.754 552.775 18.358 23 0.888 0.887 0.873 0.876 0.734 0.735 0.736 0.736 29.655 29.040 28.188 28.069 373.772 425.303 522.781 545.466 13.653 24 0.639 0.638 0.616 0.622 0.737 0.738 0.739 0.739 28.456 27.988 27.293 27.225 365.316 415.490 508.486 531.469 0.000
Attachment 6 Page 28 of 48 Table 11. Average 24-Level Axial Profiles for MkBWbI Fuel BPRA Normalized Moderator Temperature Fuel Temperature SolubleBoron (GWD/ Axial Burnup Histor Wlce) - Histo (K) Conentration istory PPii MU) axial U n U BU BU fu nu BU nu n EU nBU BU BU BU BU BU BU nu0 level < 20 20-30 30-40 >40 <20 20-30 30-40 >40 <20 20-30 30-40 >40 < 20 20-30 30-40 > 40 >0 1 (top) 0.368 0.375 0.383 0.393 0.645 0.643 0.662 0.655 28.084 28.165 27.211 27.451 474.149 472.070 525.571 515.425 0.000 2 0.718 0.718 0.747 0.750 0.648 0.647 0.669 0.661 30.401 30.336 28.385 28.689 353.641 356.121 490.872 474.225 14.313 3 0.866 0.864 0.894 0.894 0.651 0.650 0.672 0.664. 30.961 30.863 28.764 29.091 367.703 369.921 506.756 487.383 17.219 4 0.988 0.986 1.007 1.004 10.655 0.654 0.675 0.667 31.272 31.156 29.017 29.359 380.263 382.442 520.009 499.131 19.677 5 1.045 1.043 1.057 1.053 0.660 0.659 0.679 0.671 31.317 31.200 29.088 29.437 390.582 392.857 530.737 508.720 20.835 6 1.065 1.063 1.073 1.069 0.664] 0.663 0.682 0.675 31.256 31.142 29.072 29.424 398.789 401.242 539.427 516.620 21.252 7 1.073 1.070 1.079 1.076 10.669 0.668 0.686 0.680 31.177 31.065 29.033 29.392 405.134 4a7.779 546.394 523.149f 21.404 8 1.100 1.098 1.102 1.098 10.674 0.673 0.690 0.684 31.181 31.068 29.055 29.409 410.206 413.032 552.111 528.655 21.969 9 1.106 1.104 1.107 1.103 0.679 0.678 .0.694 0.689. 31.132 31.019 29.028 29.383 413.764 416.787 556.333 532.829 22.097 10 1.083 1.082 1.085 1.084 0.684 0.683 0.698 0.693 31.005 30.897 28.931 29.284 415.95I 419.182 559.052 535.478 21.648 11 1.116 1.114 1.113 1.110 0.689 0.688 0.701 0.697 31.062 30.949 28.980 29.330 418.025- 421.381 561.709 538.3551 22.313 12 1.121 1.119 1.116 1.113 0.694 0.693 0.705 0.701 31.039 30.926 28.961 29.307 418.994 422.489 563.1]15 539.897 22.418 13 1.106 1.104 1.101 1.100 0.698 0.698 0.708 0.704 30.962 30.851 28.896 29.238 419.020 422.666 563.5.13 540.384 22.121 14 1.118 1.116 1.111 1.110 0.702 0.702 0.711 0.708 30.977 30.861 28.902 29.247 418.773 422.525 563.690 540.826 22.357 15 1.131 1.129 1.122 1.120 0.707 0.706 0.714 0.712 31.003 30.885 28.916 29.256 417.910 421.748 563.087 540.420 22.633 16 1.133 1.131 1.122 1.120 0.711 0.710 0.718 0.715 30.999 30.878 28.900 29.236 416.236 420.172 561.583 539.050 22.666 17 1.110 1.108 1.102 1.102 0.715 0.715 0.721 0.719 30.929 30.807 28.827 29.160 413.A18 417.472 558.834 536.352 22.213 18 1.140 1.138 1.127 1.125 0.719 0.719 0.725 0.723 31.036 30.907 28.892 29.221 410.363 414.529 555.889 533.704 22.811 19 1.137 1.135 1.123 1.121 0.723 0.723 0.728 0.726 31.053 30.919 28.879 29.201 405.702 410.094 551.330 529.420 22.756 20 1.097 1.096 1.088 1.087 0.728 0.727 0.731 0.730 30.966 30.827 28.779 29.089 390244 404.040 544.994 523.443 21.973 21 1.084 1.083 1.075 1.074 0.732 0.732 0.734 0.733 30.979 30.831 28.753 29.043 392.089 397.498 538.204 517.698 21.721 22 1.013 1.016 1.006 1.009 0.736 0.735 0.737 0.737 30.796 30.651 28.570 28.821 384.223 390.821: 530.216 511.644 20.398 23 0.877 0.894 0.855 0.868 0.739 0.739 0.740 0.740 30.286 30.192 28.130 28.303 390.176 398.546 521.765 507.049 18.073 24 0.403 0.413 0.406 0.416 0.742 0.742 0.742 0.742 27.514 27.575 26.667 26.821 500.777 505.802 558.191 547.530 0.000
Attachment 6 Page 29 of 48 Table 12. Average 24-Level Axial Profiles for MkBWb2 Fuel BPRA Normalized ;Moderator Temperature Fuel...Temperature : .: i Soluble Boron:: Ii (GWD/
-- fiffn Axial Burnup History (lee) :fi i j History (K)- I__ (Concentrafion History (ppm) MTU) axial BU BU BU BU BU BU BU BU BU BU BU BU BU BU BU BU DU level < 20 20.30 30.40 >40 <20 20.30 30-40 > 40 < 20 20.30 30.40 > 40 < 20 20.30 30-40 >40 >O 1 (top) 0.399 0.429 0.433 0.442 0.659 0.647 0.663 0.650 27.871 28.493 27.384 27.772 301.837 30S.805 481.979 517.706 0.000 2 0.729 0.725 0.757 0.764 0.663 0.650 0.670 0.653 29.723 30.286 28.403 29.032 229.666 229.243 480.300 507.591 16.587 3 0.875 0.863 0.899 0.902 0.666 0.654 0.672 0.657 30.185 30.728 28.782 29.467 245.157 242.180 488.905 516.379 19.753 4 0.991 0.980 1.002 1.002 0.670 `0.658 00.675 0.661 30.432 30.985 29.011 29.716 258.948 254.750 495.593 525.098 22.416 5 1.043 1.033 1.046 1.045 0.674 0.662 0.679 0.665 30.443 31.007 29.066 29.774 270.478 265.633 502.154 532.708 23.647 6 1.060 1.052 1.062 1.061 0.678 0.667 0.682 0.670 30.357 30.925 29.048 29.759 280.177 274.942 508.585 539.434 24.085 7 1.066 1.059 1.072 1.071 0.682 0.672 10.686 0.675 30.253 30.824 29.024 29.754 288.466 282.871 514.903 545.629 24.244 8 1.095 1.088 1.089 1.087 0.687 0.677 0.690 0.680 30.237 30.810 29.014 29.718 295.913 289.918 518.793 550.245 24.905 9 1.101 1.095 1.096 1.094 0.691 0.682 0.694 0.685 30.170 30.744 28.989 29.696 301.772 295.479 522.697 554.150 25.066 10 1.078 1.073 1.079 1.079 0.695 0.686 0.697 0.689 30.029 30.603 28.911 29.620 306.042 299.559 525.523 556.411 24.563 11 1.110 1.106 1.102 1.099 0.699 0.691 0.701 0.694 30.069 30.643 28.932 29.623 310.127 303A07 527.370 558.905 25.306 12 1.115 1.110 1.105 1.102 0.702 0.696 0.705 0.698 30.034 30.607 28.910 29.592 312.902 306.033 528.642 560.165' 25.420 13 1.100 1.096 1.093 1.092 0.706 0.700 0.708 0.702 29.950 30.522 28.854 29.531 314.551 307.580 529.311 560.413 25.092 14 1.111 1.107 1.105 1.103 0.709 0.704 0.711 0.706 29.953 30.524 28.864 29.544 315.867 308.758 530.021 561.140 2 5 .3 4 9 15 1.125 1.121 1.114 1.111 0.713 0.708 0.715 0.710 29.971 30.543 28.864 29.528 316.495 309.240 529.235 560.579 25.664 16 1.126 1.122 1.114 1.111 0.716- 0.712 0.718 0.714 29.959 30.530 28.846 29.500 316.195 308.854 527.777 559.048 25.695 17 1.104 1.100 1.098 1.097 0.720 0.7166 0.721 0.717, 29.884 30.449 28.795 29.455 314.692 307.325 525.750 556.372 25.172 18 1.134 1.129 1.118 1.115 0.7230 0.720 0.725 0.7210 29.974 30.542 28.837 29.474 313.142 -305.789 522.575 553.831 25.843 19 1.132 1.127 1.115 1.113 0.727 0.724 0.728 0.725 29.980 30.543 28.821 29.445 310.015 302.926 518.450 549.809 25.796 20 1.095 1.090 1.084 1.084 0.730 0.728 0.731 0.729 29.893 30.440 28.734 29.347 305.098 298.723 513.278 544.366 24.962 21 1.086 1.082 1.071 1.071 0.734 0.732 0.734 0.733 29.899 30.431 28.689 29.267 299.488 294.560 508.137 540.291 24.762 22 1.018 1.023 1.009 1.010 0.737 0.736 0.737 0.7360 29.726 30.240 28.503 29.010 292.385 290.916 503.115 537.027 23.422 23 0.873 0.915 0.879 0.881 0.740 0.739 0.740 0.740 29.223 29.795 28.125 28.458 289.684 303.042 498.840 535.987 21.011 24 0.435 0.475 0.461 0.465 0.742 0.742, 0742 0.742 27.138 27.704 26.854 27.012 361.672 371.658 510.922 546.591 0.000
Attachment 6 Page 30 of 48 Table 13. Average 24-Level Axial Profiles for W-STD Fuel I - - - - I - BPRA Normalized Moderator Temperature Fuel Temperature A_in_ Soluble
, Boron W;S ^:- (GWD/
. ,,, . -,,,z5 Axial Blurnup I, Histor (Wee) Histor (Kr- _ Concemntraon History pm); :I MTI U) axial RU BU BU BU BU BU BU BU BU BU BU BU BU BU BU BU BU level <20 20-30 30-40 >40 <20 20.30 30A40 >40 <20 20-30 30-40 >40 <20 20-30 30-40 >40 >0 1I(top) 0.459 0.536 0.568 0.568 0.659 0.656 07655 0.655. 28.344 28.737 28.622 28.622 386.217 418.381 42398 423.989 2.445 2 0.665 0.735 0.759 0.759 0.662 0.659 0.658 0.658 29.238 29.441 29.172 29.172 400.502 429.149 434.440 434.440 3.806 3 0.841 0.899 0.914 0.914 0.666 0.663 0.662 0.662 29.933 29.966 29.579 29.579 414.495 439.269 444.133 444.133 5.033 4 0.956 0.993 1.000 1.000 0.669 0.666 0.666 0.666 30.240 30.151 29.719 29.719 427350 447.697 45 1.957 451.957 5.951 5 1.024 1.041 1.041 1.041 0.673 0.671 0.670 0.670 30.304 30.137 29.706 29.706 438.712 454.474 458.061 458.061 6.613 6 1.064 1.063 1.059 1.059 0.678 0.675 0.674 0.674 30.266 30.061 29.650 29.650 448.240 459.642 462.598 462.598] 7.072 7 1.087 1.074 1.068 1.068 0.682 0.679 0.678 0.678 30.206 29.991 29.603 29.603 455.845 463.334 465.766 465.766 7.378 8 1.100 1.080. 1.072 1.072 0.686 0.683 0.682 0.682. 30.141 29.928 29.563 29.563 461.698 465.775 467.811; 467.811 7.583 9 1.108 1.082 1.075 1.075 0.690 0.688 0.687 0.687 30.080 29.874 29.526 29.526 466.036 467.226 468.974 468.974 7.721 10 1.113 1.084 1.076 1.076 0.694 0.691 0.691t 0.691 30.030 29.831 29.496 29.496 469.143 467.978 469.487 469.487 7.814 11 1.116 1.086 1.078 1.078 0.697 0.695 0.695 0.695 29.992 29.799 29.470 29.470 47 1.239 468.220 469.500 469.500 7.878 12 1.120 1.088 1.080 1.080 0.701 0.699 0.698 0.698 29.965 29.774 29.448 29.448 472.473 468.035 469.082 469.082- 7.922 13 1.123 1.090 1.082 1.082 0.704 0.702 0.702 0.702 29.950 29.755 29.428 29.428 472.927 467A92 468.287 1468.287 7.952 14 1.128 1.093 1.084 1.084 0.707 0.706 0.705 0.705 29.946 29.743 29.410 29.410 472.612 466.639 467.157 467.157 7.970 15 1.132 1.096 1.087 1.087 0.710 0.709 0.708 0.708 29.956 29.738 29.396 29.396 471.436 465.433 465.664 465.664 7.974 16 1.136 1.100 1.089 1.089 0.714 0.712 0.712 0.712 29.982 29.744 29.388 29.388 469.208 463.739 463.712 463.712: 7.958 17 1.139 1.103 1.092 1.092 0.717 0.716 0.715 0.715 30.022 29.762 29.386 29.386 465.671 461.359 461.154 461.154 7.909 18 1.139 1.105 1.093 1.093 0.720 0.719 0.719 0.719 30.071 29.790 29.388 29.388 460.508 458.038 457.792 457.792 7.807 19 1.131 1.103 1.092 1.092 0.723 0.722 0.722: 0.722 30.128 29.833 29.402 29.402 453.363 453.482 453.384 453.384 7.614 20 1.110 1.095 1.087 1.087 0.726 0.726 0.725 0.725 30.192 29.904 29.443 29.443 443.864 447372 447.648 447.648 7.275 21 1.058 1.063 1.061 1.061 0.730 0.729 0.729 0.729 30.175 29.933 29.458 29.458 432.110 439.614 440.454 440.454 6.709 22 0.952 0.979 0.985 0.985 0.733 0.732 0,732 0.7320 29.899 29.766 29.327 29.327 418.665 430.339 431.825 431.825 5.804 23 0.763 0.811 0.828 0.828 0.735 0.735 0.735 0.735 29.185 29.246 28.926 28.926 404.116 419.688 421.792 421.792 4.451 24 0.534 0.601 0.629 0.629 0.738 0.738 0.738 0.738 28.238 28.535 28.378 28.378 389374 408.574 411.243 411.243 2.887
Attachment 6 Page 31 of 48 Table 14. Average 24-Level Axial Profiles for MkBI Fuel BPRA Normalized Moderator Temperature Fuel Temperature . 7i IfSoluble Boron --tI . (GWD/ __ . ,-fis X __I_ ._ _ :! - A_, ' _'_ '. ______ ~Axial Burnup) Histor (Afec) IHistor W-y ____ Coneernraun oppm Iutr : MI U)I axial nu DU BU DU BU BU BU DU BU DU BU BUUU BU DU DU BU DU level < 20 20-30 30.40 >40 <20 20.30 30.40 >40 <20 20-30 30-40 >40 <20 20-30 30-40 >40 >0 1 (top) 0.580 0.582 0.613 0.622 0.664 0.667 0.681 0.687 27.539 27.539 27.539 27.539 424.585 465.007 548.130 551.778 9.947 2 0.768 0.770 0.792 0.801 0.667 0.670 0.683 0.689 28.376 28.376 28.376 28.376 440.484 480.869 562.852 565.534 13.073 3 0.920 0.922 0.936 0.943 0.670 0.672 0.686 0.691 28.803 28.803 28.803 28.803 454.796 0495.159 576.115 577.656: 15.600 4 1.001 1.002 1.011 1.014 0.674 0.676 0.688 0.694 28.954 28.954 28.954 28.954 465.511 505.896 586.122 586.355 16.950 5 1.037 1.037 1.042 1.042 0.678 0.680 0.691 0.697 28.966 28.966 28.966 28.966 473.267 513.716 593.510 592.567 17.536 6 1.051 1.051 1.053 1.051 0.682 0.683 0.694 0.699 28.949 28.949 28.949 28.949 478.696 519.253 598.905 597.210 17.768 7 1.057 1.057 1.058 1.055 0.685 0.687 0.697: 0.702 28.925 28.925 28.925 28.925 482.357 523.047 602.747 600.732 17.881 8 1.060 1.060 1.061 1.058 0.689 0.691 0.700 0.704 28.903 28.903 28.903 28.903 484.730 525.543 605.285 603.127 17.940 9 1.062 1.062 1.062 1.059 40.693: 0.694 0.703 0.706 28.867 28.867 28.867 28.867 486.200 527.111 606.817 604.557 17.980 10 1.064 1.064 1.063 1.060 10.696, 0.697 0.705 0.709 28.845 28.845 28.845 28.845 487.060 528.046 607.690 605.367 18.010 11 1.066 1.065 1.064 1.060 0.699 0.700 0.708: 0.711- 28.827 28.827 28.827 28.827 487.580 528.615 608.166 605.794 18.039 12 1.067 1.067 1.064 1.061 0.702 0.7041 0.710 0.713 28.827 28.827 28.827 28.827 488.004 529.057 608.41 1 605.959 18.070 13 1.070 1.069 1.066 1.062 01706 0.707 0,713 0.716 28.816 28.816 28.816 28.816 488.495 529.534 608.519 606.024 18.109 14 1.073 1.072 1.067 1.064 0.709 0.710 10.715 0.718 28.791 28.791 28.791 28.791 489.124 530.122 608.506 606.176 18.162 15 1.076 1.075 1.069 1.067 0.712 0.713 0.718 0.720 28.780 28.780 28.780 28.780 489.712 530.641 608.240 606.291 18.229 16 1.081 1.080 1.072 1.071 0.715 0.716 0721 0.723 28.786 28.786 28.786 28.786 489.830 530.666 607.447 605.921 18.306 17 1.085 1.084 1.074 1.074 0.719 0.719 0.723 0.725 28.796 28.796 28.796 28.796 489.036 529.755 605.799 604.625 18.379 18 1.087 1.086 1.075 1.075 0.722 0.722 0,726 0.727 28.791 28.791 28.791 28.791 486.885 527.457 602.923 601.982 18.423 19 1.087 1.086 1.075 1.074 0.725 0.725 0.728 0.730 28.799 28.799 28.799 28.799 482.773 523.190 598.430 597.651 18.425 20 1.085 1.083 1.072 1.071 0.728 0.729 0.731 0.732 28.807 28.807 28.807 28.807 475.987 516.284 591.934 591.385 18.382 21 1.066 1.064 1.055 1.054 0.732 0.732 0,734 0.734 28.790 28.790 28.790 28.790 466,874 507.062 583.409 583.346 18.062 22 1.002 1.002 0.995 0.995 0.735 0.735 0.736 0.737 28.657 28.657 28.657 28.657 456.846 496.849 573.186 574.103 17.011 23 0.865 0.868 0.864 0.866 0.737 0.737 0.738 0.738 28.287 28.287 28.287 28.287 447.364 487.013 561.611 564.245 14.763 24 0.689 0.696 0.696 0.700 0.740 0.0.740.740 0.740 27.485 27.485 27.485 27.485 438.633 477.832 549.660 554.375 11.892
Attachment 6 Page 32 of 48 Table 15. Average 24-Level Axial Profiles for W-OFA Fuel V ... -- BPRA Normalized Moderator Temperature Fuel Temperature Soluble foron. ..., I I (GWD/ I Axial Biurnup I, - IHistory twieC) _I _ History (K)11 __ Concernrfon History I Ipm) MLJU) axial nU nU DU BU BU BU BU BU RU BU BU nU nU RU BU BU DU level <20 20-30 30-40 >40 <20 20.30 30-40 > 40 < 20 20-30 30-40 > 40 < 20 20-30 30-40 >40 >0 1 (top) 0.562 0.587 0.603 0.615 0.651 0.657 0.655 06'6-57 28.161 27.869 27.866 27.723 360.929 4411011 490.089 515.734 7.702 2 0.753 0.774 0.787 0.800 0.653 0.660 0.658 0.660 28.906 28.475 28.408 28.232 368.982 450.302 500.358 526.950 10.215 3 0.908 0.923 0.931 0.937 10.656 0.662 0.661 0.664 29.483 28.936 28.805 28.576 376.549 458.770 509.448 i536.382 12.160 4 0.993 1.002 1.006 1.009 0.660 0.666 10.665 _0.667 29.738 29.135 28.975 28.724 382.862 465.710 516.575 543.569 13.248 5 1.031 1.036 1.039 1.040 0.664 0.669 0.669 0.671 29.800 29.178 29.010 28.761 387.930 471.248 522.115 549.070 13.737 6 1.048 1.051 1.051 1.050 0.668 0.673 0.673 0.675 29.790 29.162 28.991 28.737 391.764 475.493 526.375 553.334 13.960 7 1.058 1.059 1.059 1.057 0.672 0.677 0.677 0.679 29.774 29.143 28.971 28.718 394.474 478.551 529A69 556.449 14.091 8 1.065 1.065 1.064 1.064 0.676 0.681 0.681 0.683 29.755 29.123 28.953 28.710 396.258 480.565 531.496 558.517, 14.176 9 1.069 1.068 1.066 1.066 0.680 0.685 0.685 0.687 29.734 29.100 28.927 28.682 397.302 481.721 532,679 559.819 14.236 10 1.072 1.070 1.067 1.065 0.685 0.689 0.690 0.691 29.718 29.078 28.899 28.647 397.773 482.210 533.196 560.474 14.287 11 1.076 1.072 1.069 1.068 0.689 0.693 0.694 0.695 29.708 29.063 28.880 28.630 397.796 482.168 533.152 560.502 14.337 12 1.080 1.075 1.072 1.070 0.693 0. 697 0.697 0.699 29.702 29.050 28.862 28.615 397.444 481.680 532.647 560.061 14.386 13 1.083 1.077 1.072 1.068 0.698 0.701 0.701 0.702 29.698 29.038 28.840 28.582 396.765 480.808 531.776 559.330 14.432 14 1.087 1.079 1.074 1.071 10.702 0.704 0.704 0.706 29.698 29.030 28.826 28.571 395.785 479.568 530.490 558.095 14.473 15 1.090 1.082 1.077 1.075 0.705 0.708 0.708 0.709 29.701 29.025 28.816 28.566 394.468 477.933 528.797 556.428 14.514 16 1.094 1.084 1.078 1.075 07097 0.711 0.711 0.712 t 29.711 29.023 28.803 28.545 392.722 475.818 526.663 554.377 14.560 17 1.098 1.086 1.079 1.074 0.713 0.715 0.715 0,716 29.727 29.024 28.791 28.521 390.461 473.122 523.961 551.773 14.604 18 1.100 1.087 1.080 1.076 0.717. 0.718 0.718 0.7199 29.741 29.027 28.785 28.518 387.601 469.739 520.527 548.370 14.624 19 1.099 1.087 1.079 1.074 0.720 0.722 0.722 00.722 29.760 29.032 28.778 28.506 383.849 465.415 516.238 544.210 14.621 20 1.096 1.084 1.076 1.069 0.724 0.725 0.725 0.726 29.793 29.050 28.776 28.487 378.741 459.764 510.889 539.173 14.591 21 1.074 1.065 1.059 1.054 0.728 0.729 0.729 -0.729 29.768 29.026 28.747 28.457 372.486 452.932 504.393 533.062 14.325 22 0.999 0.996 0.995 0.995 0.732 0.732 0.732 0.732] 29.522 28.834 28.579 28.312 365.969 445.613 496.903 526.017 13.368 23 0.837 0.845 0.853 0.860 0.735 0.735 0.735 0.735 28.892 28.341 28.158 27.934 360.101 438.516 488.554 518.022 11.352 24 0.629 0.648 0.661 0.667 0.738 0.738 0.738 0.738 28.055 27.676 27.570 27.355 354.810 431.716 4:79.834 1509.316 8.615
Attachment 6 Page 33 of 48 Table 16. Average 24-Level Axial Profiles for W-RFA Fuel 1 F 1 21 - - BPRA Normalized Moderator Temperature Fuel Temperature ; 1: Id l;SSoluble Boron ] m;:0: (GWD/
*--_._ ,Lq5 He~~~ IdAn_
Axial Il11rnup ___ i Histor 1te) _ History tNy)- iConcentration HsTory ' IpLVM lI U) axial BU BU BU BU BU BU BU BU BU BU BU BU BU BU BU BU BU level <20 20.30 30-40 >40 <20 20-30 30-40 > 40 < 20 20-30 30-40 >40 < 20 20-30 30-40 > 40 >0 1 (top) 0.424 0.464 0.479 0.476 0.667 0.646 0.670 0.664 27.016 27.762 26.829 26.927 526.165 549.175 630.332 657.616 0.000 2 0.731 0.756 0.771 0.771 0.669 0.649 0.677 0.6690 29.244 30.116 28.243 28.453 460.213 491.773 620.620 649.071 0.000 3 0.878 0.879 0.905 0.906 0.672 0.653 0.683 0.6740 29.854 30.735 28.409 28.759 437.632 444.446 607.091 636.009 21.496 4 0.987 0.981 1.000 0.999 0.675 0.658 0.686 0.678- 30.186 31.080 28.618 29.012 446-.630 1452.92 614.668 643.605 23.971 5 1.032 1.025 1.039 1.038 0.679 0.663 0.689 0.681 30.246 31.142 28.664 29.086 455.164 460.634 621.977 651.076 25.044 6 1.050 1.043 1.055 1.053 0.683 0.667 0.693 0.685 30.210 31.104 28.649 29.083 462.489 467,974 628.462 657.715 25.484 7 1.069 1.062 1.070 1.068 0.687 0.672 0.695 0.688 30.191 3 1.081 28.639 29.080 468.702 474.388 634.057 663.655 25.946 8 1.077 1.070 1.075 1.074 0.691 0.677 0.699 0.692 30.149 31.034 28.613 29.057 473.313 479.100 638.154 668.003 26.150 9 1.087 1.079 1.082 1.081 0.694 0.681 0.702 0.696 30.124 31.005 28.596 29.040 476.742 482.647 641.164 671,268 26.384 10 1.075 1.069 1.072 1.072 0.698 0.686 0.705 0.699 30.040 30.920 28.541 28.980 478.435 484.154 642.535 672.580 26.122 11 1.093 1.086 1.085 1.084 0.701 0.690 0.708- 0.703 30.064 30.940 28.551 28.990 480.053 485.986 643.717 674.176 26.548 12 1.096 1.089 1.087 1.086 0.705 0.695 0.7 11 0.706 30.046 30.920 28.535 28.970 480.5'19 486.421 643.763 674.387 26.626 13 1.087 1.081 1.078 1.078 0.708 0.699 0,.714 0.710 29.995 30.869 28.497 28.926 479.882 485.572 642.774 673.360 26.414 14 1.104 1.097 1.090 1.090 0.712- 0.704 0.717 0.713- 30.039 30.908 28.515 28.940 479.288 485.145 641.681 672.665 26.809 15 1.111 1.103 1.094 1.094 0.716- 0.708 0.720 0.717 30.055 30.923 28.516 28.935 477.514 483.361 639.418 670.622 26.956 16 1.111 1.104 1.093 1.093 0.719 0.713 0.723 0.720 130.063 30.932 28.512 28.924 474.586 480.309 636.078 667.354 26.974 17 1.103 1.096 1.086 1.087 0.723 0.717 0.726 0.723 30.054 30.927 28.496 28.900 470.232 475.688 631.348 662.568 26.787 18 1.120 1.113 1.099 1.098 0.726 0.721 0.729 0.727 30.146 31.017 28.540 28.934 465.319 470.960 626.009 657.664 27.191 19 1.118 1.111 1.096 1.096 10.730 0.726 0.732, 0.730 30.191 31.065 28.552 28.934 458.375 463.955 618.867 650.705 27.151 20 1.089 1.083 1.071 1.073 0.733 0.730 0.735 0.734 30.159 31.042 28.516 28.878 449.216 454.474 609.869 641.604 26.470 21 1.067 1.064 1.052 1.053 0.737 0.735 0.738 0.737 30.168 31.050 28.496 28.826 439.586 445.098 600.659: 632.994 26.000 22 0.984 0.987 0.978 0.982 0,740 0.739 0.741 0.740 29.925 30.798 28.322 28.591 429.262 435.910 591.112 624.315 24.123 23 0.810 0.838 0.826 0.830 0.743 0.742 0.743 0.743 29.201 30.051 28.077 28.178 449.321 479.329 602.230 633.709 0.893 24 0.436 0.475 0.477 0.480 0.745 0,745 0,745 0.745 26.516 27.181 26.315 26.371 513.316 534.007 611.839 639.983 0.000
Attachment 6 Page 34 of 48 With the 24-level axial profiles now "averaged" into burnup groups, fuel to be stored in Region 2 is evaluated by the following procedure:
- Determine the fuel assembly type for the assembly to be stored in SFP.
- Determine the average (measured) 2-D burnup of the assembly being analyzed.
- With the fuel type and average burnup, obtain the "average" profiles for the five history variables discussed at the beginning of this section, by selecting the profiles from the appropriate burnup "group" for that fuel type.
- Convert the selected normalized burnup profile into an estimated "real" burnup profile by multiplying the normalized value at each axial level by the user-defined or measured 2-D assembly-average burnup.
- Build a SIMULATE-3 Region 2 model for this fuel assembly with the 24-level estimated burnup profile from the previous step, along with the profiles for the other four history variables obtained in the third step of this procedure.
Having established the average axial profiles to be used in calculating system keffs for fuel assemblies stored in McGuire Region 2, it is necessary to have all the SlMULATE-3 nodal cross-section data available to analyze all the pertinent fuel types in 3-D. Fuel irradiation cases in reactor operating conditions are first needed to determine accurate fuel isotopic content as a function of burnup. In addition to the seven fuel types discussed in Section 3, it is necessary to generate nodal cross-sections for a pseudo-fuel type that approximates a water hole (empty storage cell). This will allow the nodal SIMULATE-3 code to model the 3/4 Checkerboard/Empty storage configuration shown in Figure 4. The CASMO-3/SIMULATE-3 codes require at least a small amount of fissile material to compute nodal cross-section data for any fuel or "water holes" used in the SFP Region 2 rack model. For this calculation the best convergence, for test cases of the 3/4 Checkerboard/Empty model, was observed with a pseudo-fuel type that used a fuel pellet diameter of just 0.20 cm, an enrichment of 0.30 wt % U-235, and a fuel density of 10.00 g/cc. Verification of the accuracy and conservatism of using this "water hole" fuel type for the 3/4 Checkerboard/Empty SIMULATE-3 model is documented in Table 17, which compares SCALE 4.4/KENO V.a and SIMULATE-3 cases that specify unirradiated 2.00 wt % U-235 Checkerboard fuel assemblies mixed with varying ratios of "water holes."
Attachment 6 Page 35 of 48 Table 17. Comparisons between SIMULATE-3 and KENO V.a for Various Checkerboard/"Empty" Storage Configurations {all cases at 150 0F, 0 ppm boron} SIMULATE-3 kj iKENO a k i- KENO V.a kl.f
~using mdel with usingjmode wit using model With NO Region2 -fissile material in ;issilerialin fi:sle manterial in 1-40I Storage Configuration "wilater hole" "water hole" iater hole -
All Checkerboard Assemblies (4/4) 1.1929 1.1901 1.1905 3 Checkerboard / I 'Empty" (3/4) 1.0569 1.0565 1.0482 2 Checkerboard / 2 "Empty" (2/4) 0.8556 0.8464 0.8185 1 Checkerboard / 3 "Empty" (1/4) 0.7507 0.7526 0.7330 All 'Empty" Cells (0/4) 0.1961 0.1970 Once all the nodal cross-section data for the necessary fuel types have been compiled into a master fuel library, actual 3-D models of the Region 2 racks can be constructed, using an automated form of the procedure outlined above. In this manner, minimum burnup requirements are determined for each of the SFP Region 2 storage configurations shown in Figure 4, as a function of fuel type, initial enrichment, and post-irradiation cooling time. These are the 2-D fuel burnups needed to satisfy the pertinent regulatory subcriticality criteria from 10 CFR 50.68 (b) (4). Tables 18 through 21 document, for the different Region 2 storage configurations shown in Figure 4, the minimum bumup requirements calculated by this process. Note that each of the "normal-condition" McGuire SFP Region 2 criticality -computations considers the SFP water temperature at both 32 0F and 150 TF, to ensure the maximum-reactivity SFP temperature condition is determined for every case. Since the master fuel library only has specific nodal cross-section data for enrichment increments of 0.50 wt % U-235 and cooling times at 5-year intervals, these are the only data points provided in Tables 18 through 21. From an implementation standpoint, it is important to define how the end user should determine the burnup requirements for a fuel assembly that has an enrichment and/or cooling time that is outside of or in between the specifically tabulated data points. In evaluating a fuel assembly to determine whether it meets the minimum burnup requirements for the desired storage configuration, no extrapolations are performed. That is, if a fuel assembly type has a lower maximum enrichment than the lowest tabulated enrichment for that fuel type, the lowest tabulated value is used instead of performing an extrapolation to the actual assembly enrichment. Likewise, if a fuel assembly has cooled longer than 20 years, the minimum burnup requirement for a 20-year cooling time is used, rather than an extrapolation of the burnup data beyond 20 years.
Attachment 6 Page 36 of 48 Table 18. SFP Region 2 Unrestricted Storage -- Minimum Burnup Requirements as a Function of Initial Enrichment, Cooling Time, and Fuel Assembly Type. Cooling Enrichment Time (wt % U-235) (m) 2.00 2.50 3.00 3.50 4.00 4.50 5.00 MkBW 0:; : ;fo '22.20 30.01 36.67 - 43.61' : ',50.47 57.18' -i 63.72
' 5000 -,; 0'19.42 26.06 32.23 ',38.64, ,'44.70 50.80 56.77' 0 ; '-1000 ' 17.76 24.07 30.01 36.02 41.76 47.56 53.24:
if i15 4' ' 16.74 22.90 :'28.95 34.45 40.01 45.64 51.15 _,20 0 16.07 ; - i 22.13' : 2'"; 28.05 t ;i';- :- 33.44 39.08 44.38 "49.78 MkBWbl 0 30.01 36.05 42.52 48.57 54.24 59.74 5 27.27 31.69 37.20 42.92 48.03 53.01 10 ^S'iN-o^"^^6^o425.15 30.01 34.63 40.01 44.85 49.58 15is ^Wg'^6s>6',4 23.89 29.37 33.09 38.21 43.00 47.57 20 23.09 28.43 32.09 37.13 41.78 46.26 MkBWb2 0 5 .S, ~~~~~~~33.020 38.340 ' 43.97 , '48.98 5;'3.87,
- 015° S:;0ljj -0 -- 4 '30.720 ; 335.73 40'3°'995 ; 45.677 ' $'3
- 2
_______ . '^s44 g - 29.62 33.12 :37.88 42.47 46.85 W-STD 0 20.02 28.59 35.83 43.37 50.67 57.75 64.63 5 18.50 25.10 31.56 38.35 44.97 51.39 57.66 10 17.14 23.29 30.01 35.78 42.03 48.12 54.08 15 16.32 22.21 28.83 34.24 40.29 46.19 51.97 20 15.79 21.51 27.96 33.24 39.16 44.94 50.61 MkBI 0O 20.21 28.01 34.47 40.82 I 17.71, 24.76 30.66 36.92 10 16.35 ~~23.04 29.42 34.60 i5 15.53 22.01 28.16 33.19 20 15.00 21.33 27.34 32.27 W-OFA 0 18.55 26.08 33.28 40.01 46.83 53.25 59.71 5 16.53 23.30 30.01 36.27 42.01 48.05 53.98 10 15.43 21.83 28.25 34.10 40.01 45.34 50.99 15 14.75 20.92 27.12 32.78 38.60 43.68 49.19 20 14.32 20.33 26.40 31.91 37.62 42.62 48.02 W-RFA 80 -..: gggg~gs g~00 ^:.sf 2 ;35.46 42.04 47.88 53.50 58.94
-5 -31.19 36.62 42.23 47.30- 52.23 10 wi 30.01 34.11 39.05 44.15 48.85 1s 28.85' 32.63 37.41 42.31' 46.87-
- 10 <K427.93 31.67, 36.35' 41.12: 45.57
Attachment 6 Page 37 of 48 Table 19. SFP Region 2 Restricted Storage -- Minimum Burnup Requirements as a Function of Initial Enrichment, Cooling Time, and Fuel Assembly Type. Cooling Enrichment Time (wt % U-235) (yrs) 2.00 2.50 3.00 3.50 4.00 4.50 5.00 MkBW 0 18.26 25.32 i31.73 38.39 44.73 51.04 - 57.20 0 5 :15.69 22.29 28.56 34.6 '40.01 45.66' 51.27 10 14.36 '20.68 26.60 31.95 37.54 42.83 48.19 15 1359 20.01 25.42 30.62 36.04 41.6 46.36 20 13.10 ' 19.29- 24.66 30.01 35.05 -40.07 45.16-MkBWbl 0 26.29 31.14 37.02 43.12 48.55 53.83 5 23.07 28.91 32.89 38.20 43.29 48.07 10 21.37 26.88 30.73 35.78 40.54 45.07 15 20.35 25.66 30.01 34.30 38.82 43.31 20 19.66 24.86 29.76 33.36 37.77 42.16 MkBWb2 0 32.33 38.06 44.25 49.61 54.82
; l5o > ,,X,>,, 33.86 tX 38.4' 29.93 ; 44.09 48.80 10 27.89 31,65 36.48 41.24 :45.69 15 0 ;' 0 ' ' gS t 5 26.66 30.32 34.99 39.40 -43.85
;20 : ; L 8tW~t,' NggX,,0,2,= - 25.85 30.01 34.01 38.34 42.67 W-SlD 0 16.34 23.70 30.62 37.69 44.55 51.21 57.70 5 14.55 21.04 27.88 33.62 39.84 45.90 51.80 10 13.58 2042 26.08 31.49 37.37 43.11 48.73 15 12.99 20.01 24.99 30.21 35.89 41.45 46.90 20 12.63 19.56 24.28 30.01 34.93 40.37 45.70 MkBI 0 e, .16.13 23.62 30.01 36.37 5 14.26 21.08' 27.43 3.1' 10 13.27 19.71 2572 30.74 15: 12.67 18.87 24.67 30.01 20 ~~12.30 1 8.33 239 2.5 >
W-OFA 0 14.85 22.04 29.10 35.62 41.63 47.88 54.01 5 13.38 20.01 26.26 32.15 38.13 43.42 49.04 10 12.53 19.03 24.74 30.33 36.01 41.04 46.41 15 12.00 18.29 23.81 29.54 34.72 40.01 44.82 20 11.67 17.82 23.20 28.80 33.87 39.17 43.77 W-UFA 0 -;0 ti30.73 - 0 0XR<LS 36.55 42.59 J 47.99 53.22 i' 050 t gt'0W - W',0¢;' ' 28.49 : '32.49 37.54 42.73 47.47 10 . 26.49 30.38 35.15, 40.01 44.50 f15 if E -= ^g ^ .0S00St ^a} - 4 25.30 30.01 33.71 38.18 42.75 20' 24.53 29.33 32.78 37.16 41.61
Attachment 6 Page 38 of 48 Table 20. SFP Region 2 Filler Storage -- Minimum Burnup Requirements as a Function of Initial Enrichment, Cooling Time, and Fuel Assembly Type. Cooling Enrichment Time (wt % U-235) (53) 2.00 2.50 3.00 3.50 4.00 4.50 5.00 MkBW - 27.34 34.90 42.58 50.08 ' 57.40 64.52 '7146 5:0 23.28: . 30.12 37.14' 43.78 50.40 ,56.89 63.22 10 21.24 28.12 34.35 40.65 46.94 53.10 59.12 15 20.02 26.67 32.70,' 38.98 44.88 50.86 56.72 20 19.50 25.73 31.65, - 37.77 - 43.55 :49.42 55.1 MkBWbl o 35.45 42.48 48.89 55.13 61.02 66.75 5 30.69 36.65 42.58 48.29 53.61 58.82 10 29.76 33.85 39.24 44.86 49.90 54.84 15is't' 28.18 32.24 37.45 42.88 47.75 52.53 20 2006 27.20 31.19 36.27 41.58 46.35 51.02 MkBWb2 5~~~~~~ 37 38.71 4 . 90 48 6 53. 27 1520.01 26.-32 32.41 3.44 07 4 6 514.75 57.981 20 33,7S8 f 38.71 t:043.90 ' 48.63 ;153.27 W-STD 0 25.55 33.83 42.22 50.27 58.03 65.54 72.89 5 21.90 30.01 36.78 44.01 51.04 57.87 64.53 10 20.06 27.68 34.03 40.87 47.52 54.01 60.34 15 20.01 26.32 32A1 39.02 45.46 51.75 57.91 20 19.68 25.44 31.39 37.84 44.13 50.29 56.33 MkBI 0 25.14 32.48 0 1.. 46.65 5 21.76 29.20 3.28 41.37
- 15 10 18.94:
19.98: gS 25.73
'27.03: 31.3 32.82 37.45 39.13 20 18.27 24.89 30.40 36.39 _ _ _ _ _ _ _ _ _
W-OFA 0 22.71 30.79 38.56 45A6 52.60 59.57 66.38 5 20.01 27.42 34.25 40.55 47.08 53.46 59.71 10 18.87 25.56 32.01 38.51 44.22 50.31 56.27 15 18.00 24.44 30.67 36.96 42.51 48.42 54.24 20 17.43 23.71 30.01 35.96 41.41 47.20 52.92 W-RFA 0 < 41.90 48 .19 54.22 60.04 65.72 1; 5i'0 ^X :-
' ,,' v ,'4 -35.92 41.96^- 47.39 ;52.66 57.81.
10 33.22, 38.50 44.00 49.01 53.90. 15 4.>31.66 36.7'7 :42.03, 46.89 51.62
'20 30.66 35.65 40.76 '45.49 50.14
Attachment 6 Page 39 of 48 Table 21. SFP Region 2 Checkerboard Storage (for 314 Checkerboard/Empty Configuration) - Minimum Burnup Requirements as a Function of Initial Enrichment, Cooling Time, and Fuel Assembly Type. Cooling Enrichment Time (wt % U-235) (yrs) 2.00 2.50 3.00 3.50 4.00 4.50 5.00 MkBW ,o0--;0-- i 8.12020'0 )16.50- '22.94 29.15 34.67 40-43 46.20 i-0i5 l tlD7.49 14.77 20.81 26.50 3 1.60 37.03 42.18 10 7.07 13.77 A9.7 19.79 24.96 30.0 34.99 40.01'
'15 L-' - '6.81 -13.16 18.98 24.00 29.10 33.73 38.67
-'20 ' i -' ' '0006.64 4" '12.78 23.37
-8.45 28.35 32.90 ' 37.75 MkBWbl 0 16.23 23.10 28.95 33.14 38.33 43.68 5 X=.t,0=g' 14.53 20.89 26.24 30.19 34.95 39.57 10 13.55 19.52 24.65 29.63 32.99 37.40 1s 12.95 18.65 23.66 28.48 31.78 36.06 20 7.24 12.58 18.08 23.01 27.73 30.99 35.17 MkB\%b2 5 600j -023.77 41433.83 20.-29.59 5. 3 1.08 38.970 4 4A80 10 6 1. 19.69 246.88 30.81 35.52 40.41 15 6.24 12.65 24 8 318.991 29.1 33.52 378.9 20 6 1. 18.35 23.64 28.25 32.768 35.62 W-STD 0 7.67 14.97 21.82 28.18 34.01 40.25 46.34 5 6.79 13.83 20.01 25.78 31.08 36.80 42.40 10 6.46 13.11 19.69 24.37 30.01 34.81 40.14 15 6.24 12.65 18.99 23.48 29.01 33.57 38.74 20 6.11 12.37 18.54 22.91 28.31 32.76 37.83 MkBI W-OFA 0 16.69 14.068 20.44 26.78 32.70 38.68 44.03 5 6.32 13.06 19.21 24.72 30.15 35.68 40.65 10 6.07 12.40 18.26 23.50 28.94 33.93 39.12 15 5.91 11.98 17.66 22.73 28.00 32.83 37.87 20 5.82 11.71 17.27 22.22 27.38 32.10 37.05 W-RFA 0 66 1 22.87 28.80 32.88 38.06 43.49 5 13-06 0.56 1921 2 26.067 30.01 34.66 039.27 10 6.. 1 18.99 24.48 29.27 32.73 37.11 15 5 18.23 23.47 28.12 31.52 35.77 20 5.82 .71 17.74 22.81 27.37 30.73 34.890
Attachment 6 Page 40 of 48 On the other hand, when the user wishes to determine a burnup requirement for an enrichment and/or cooling time in between the specific data points listed in Tables 18 through 21, it is acceptable to perform some kind of interpolation procedure. With the assertion that changes in system kff for stored fuel are proportional to changes in concentrations of fissile and poisonous isotopes, one would expect a relatively linear increase in kff with increased initial U-235 enrichment, while kff should decrease in a logarithmic fashion with cooling time, since reactivity changes following reactor irradiation are primarily attributable to the decay of fissile Pu-241 (- 14.3-yr half life) and the buildup of poisonous Gd-155 via Eu-155 decay (- 4.7-yr half life). Given this expected kff behavior, the following interpolation procedure is used:
- Determine the fuel type, maximum design enrichment, and cooling time for the fuel assembly being evaluated. Locate, for this fuel type and the desired SFP Region 2 storage configuration, the minimum burnup requirements tabulated as a function of cooling time and initial enrichment.
- Make fourth-order polynomial fits, as a function of cooling time, to the five cooling time data points at each enrichment. Use these equations to find the minimum burnup requirements for the actual cooling time of the fuel assembly being evaluated, at the two enrichment data points bounding the actual enrichment of the fuel assembly.
- Perform a linear interpolation between the bounding enrichment data points (as determined in the above step) to find the minimum burnup requirement for the actual enrichment of the fuel assembly being evaluated.
Note that there is some error associated with using the interpolation process described above. That is, at interpolated values of cooling time and enrichment between those points specifically calculated with CASMO-3/SIMULATE-3, one would expect the "true" burnup requirement to be slightly greater or less than the estimated value obtained via interpolation. To quantify this error, specific cases at various "in-between" enrichments and/or cooling times are analyzed and compared with the interpolation estimates. These cases show a maximum interpolation error of +0.00036 Ak. This interpolation error is applied as a bias in the total 95/95 kff calculations for McGuire SFP Region 2 storage, as noted in Section 6. To determine the maximum 95/95 Region 2 kff corresponding to the minimum burnup requirements listed in Tables 18 through 21, it is necessary to evaluate the potential reactivity increases associated with variations among fuel assemblies stored within a particular configuration, as well as increases due to boundary effects between adjacent Region 2 storage configurations.
Attachment 6 Page 41 of 48 Within a particular Region 2 storage configuration, reactivity increases are examined by "mixing up" the stored fuel; that is, by randomly matching assemblies of one fuel type / enrichment I cooling time combination with another. This is important to check because many of these combinations use different axial history profiles, and so it is possible for a non-uniform radial assortment of fuel assemblies in a storage configuration to have a slightly higher system keff than a uniform array of such fuel assemblies. Because Region 2 has three defined fuel storage configurations (see Figure 4), it is also important to examine the reactivity effects of storing one storage configuration next to another. To limit the potential reactivity increases associated with storing one type of SFP configuration next to, within, or around another, the following Region 2 storage configuration boundary restrictions are proposed:
- Unrestricted storage - No boundary restrictions.
- 214 Restricted/Filler storage - No boundary restrictions.
- 3/4 Checkerboard/Empty storage - Any row or column of fuel in a 3/4 Checkerboard/Empty storage configuration that borders any other storage configuration must have alternating Checkerboard fuel and empty cells.
That is, it cannot be a row or column of solid Checkerboard fuel. Using the boundary restrictions defined above, several scenarios are considered in which one of these storage configurations is adjacent to or surrounded by another. These cases are evaluated with random variations of fuel type / enrichment / cooling time within the Unrestricted, 2/4 Restricted/Filler, and 3/4 Checkerboard/Empty storage arrays. The results of all these analyzed storage configuration scenarios indicate that, with no boron in the SFP water, the maximum Region 2 95/95 keff associated with the minimum bumup requirements listed in Tables 18 through 21 is 0.99888. As the discussion above demonstrates, this bounding 95/95 kiff accounts for the variations of fuel assembly parameters such as fuel type, enrichment, and cooling time within a particular defined configuration, and it meets the proposed boundary restrictions between different SFP Region 2 storage configurations. Prior to confirming this maximum 95/95 keff for all of the proposed Region 2 storage configurations, it is still necessary to quantify three burnup-related uncertainties discussed in Section 6 and listed in Table 5. These are the burnup computational uncertainty, the burnup measurement uncertainty, and the axial profile uncertainty. Each of these uncertainties can be determined by examining a "global" collection of fuel assemblies, either in the McGuire operating reactor or the SFP racks, and evaluating the maximum system reactivity increases associated with variations of the pertinent parameters for these assemblies from their nominal, or assumed, values. As noted in Section 6, the burnup computational uncertainty quantifies the accuracy of the CASMO-3 / SIMULATE-3 codes in determining the isotopic content, and hence kef, of a collection of irradiated assemblies in the McGuire reactors, assuming the actual average burnup of the fuel in the reactor core is the same as the average bumup of the
Attachment 6 Page 42 of 48 SIMULATE model for that reactor core. Determining the burnup computational uncertainty in this manner is an alternative to performing benchmarking to actual isotopic measurements of irradiated fuel. Several cycles of McGuire reactor operational data were examined to evaluate the differences between measured and SIMULATE-3-predicted core reactivity at various times during the operating cycle. The analysis of these data yielded a bumup-dependent reactivity uncertainty. No definitive bias was observed. The burnup computational reactivity uncertainty is +/-(O.00454
- BU / 50JAk, where BU is the average burnup of the system modeled, in GWD/MTU.
Section 6 stated that the burnup measurement uncertainty represents the reactivity penalty associated with difference between the measured burnup and the code-predicted burnup. Measured burnups are used for Technical Specification verification of, for instance, the minimum burnup requirements listed in Tables 18 through 21. However, these 2-D measured burnups have many sources of instrumentation error that will result in the measurement value being different from the "true" burnup of a specific fuel assembly. For the purposes of this analysis, the measured burnup error for an individual fuel assembly is defined as the difference between the measured burnup and the core follow predicted burnup. In this way, differences between measured and predicted burnups can be evaluated to produce the distribution of burnup measurement errors for a database of McGuire discharge fuel assemblies, and quantify an appropriate measurement uncertainty. This is similar to the -approach used in Reference 14. Measured burnups are available from the master special nuclear material (SNM) database used for Duke Power's reactors. These burnups are obtained from in-core detector measurements taken regularly during power operation. The code-predicted burnup for each of these fuel assemblies is taken from reactor core-follow computations using the SIMULATE-3 code. As expected, the differences between predicted and measured burnup data for the database of all McGuire reactor discharge fuel (from Cycle 1 through the present) form a distribution comparable to a normal distribution. The maximum individual assembly error observed is about 4.0 %. When an array of fuel assemblies large enough to affect system reactivity is evaluated for the McGuire SFP Region 2, and the distribution of predicted-to-measured burnup differences is accounted for, the maximum system reactivity increase observed is - 0.00125 Ak. The burnup measurement uncertainty to be used in the maximum 95/95 keff calculation for Region 2 is thus specified as +/-0.00125 Ak. The axial profile uncertainty, as Section 6 mentioned, represents the bounding reactivity penalty associated with differences between the keff calculated using the average "estimated" axial burnup and history profiles for a particular fuel assembly (see Tables 10 through 16), and the kff calculated using the actual axial burnup and history profiles for that fuel assembly (from core follow computations). Earlier in this section, the discussion of the average axial history profiles noted the large database of McGuire core
Attachment 6 Page 43 of 48 follow profiles available. The axial profile keff error for an individual fuel assembly in this database is defined as the difference between the kff calculated with the actual core follow axial profiles for that fuel assembly and the kff calculated with the average axial profiles (based on fuel type and burnup) for the assembly. As with the measured burnup errors, the distribution of the axial profile keff errors in SFP Region 2 storage compares rather well with a normal distribution. The slightly negative bias observed is conservatively ignored. The largest individual assembly axial profile error calculated is +0.030 Ak. However, the bounding axial profile uncertainty is quantified in the same "global" manner as the burnup measurement uncertainty, considering a group of fuel assemblies large enough to affect system reactivity in Region 2 of the McGuire SFPs, and taking into account the distributions of axial profile kff errors within this group of assemblies. In addition, the determination of the bounding uncertainty allows for the fact that groups of four or eight fuel assemblies are often symmetrically designed for reactor operation, and these fuel assembly groups will have the same axial profile characteristics when those assemblies are ultimately discharged together from the reactor. When all of these factors are analyzed, the resulting bounding axial profile uncertainty is +/-0.00305 Ak. Finally, now that all the pertinent reactivity biases and uncertainties have been determined, the maximum calculated 95/95 k.ff for McGuire Region 2 storage can be confirmed for normal conditions in unborated water, in accordance with the equation presented in Section 6. Table 22 includes all of the biases and uncertainties for Region 2 storage; and shows that-the maximum 95/95 keff in unboratedwater remains less-than 1.0, meeting the requirements of 10 CFR 50.68 (b) (4). If credit is taken for 800 ppm soluble boron in the McGuire SFPs, SIMULATE-3 calculations considering all of the SFP Region 2 normal-condition storage requirements (viz., the minimum burnup limits specified in Tables 18 through 21, and the allowable storage configurations in Figure 4) show that the maximum 95/95 keg for Region 2 fuel storage is reduced well below 0.95. It is worth mentioning that because the cases that analyze 800 ppm of soluble boron credit in Region 2 are actually performed in 3-D with irradiated fuel, the potential non-conservatisms associated with applying fresh fuel reactivity-equivalencing to burned fuel in a borated environment (see Reference 15) are not applicable here. The only remaining task for the Region 2 criticality analysis is to evaluate potential accident conditions. Of the Reference 10 accident scenarios, only the fuel assembly misload and high abnormal water temperature (212 SF) events need to be considered. The fuel assembly drop accident was discussed in Section 8.1 for the SFP Region 1 criticality analysis. The analysis for this accident is valid for Region 2 as well, since it is not rack-dependent. In addition, the heavy load drop accident mentioned in Section 8.1 does not need to be considered for the Region 2 criticality evaluation, because the weir gate is not carried directly over Region 2, and thus an end-drop of the gate onto Region 2 - the only type of weir gate drop capable of deforming the storage racks - is not possible.
Attachment 6 Page 44 of 48 Of the two Region 2 accident conditions that need to be analyzed, the misload accident clearly is much more severe, from a criticality perspective, than an increase in SFP water temperature from 150 'F to 212 TF. The fuel assembly misload event is thus considered the bounding SFP Region 2 accident condition. Reference 10 states that for a fuel assembly misload event, it is acceptable to consider a single misload error to be the worst case, "unless there are circumstances that make multiple loading errors credible." Reference 10 also notes that it is permissible, for accident scenarios, to take credit for the full boron concentration (2675 ppm) required as a minimum in the McGuire SFPs. The worst-case misload event in Region 2 involves placing a fresh 5.0 wt % fuel assembly in an empty cell, within the 3/4 Checkerboard/Empty configuration storage configuration shown in Figure 4. The analysis of this misload event demonstrates that 1600 ppm is sufficient to bring the SFP Region 2 system keff below the 0.95 subcriticality criterion. Table 22. Maximum 95/95 kff for Fuel Storage in Region 2 of the McGuire SFPs (No Boron in SFP Water) SFP Region 2 ______________________________IStorage II Nominal kcff 0.98126 Biases Benchmark Method Bias Fixed Poison Self-Shielding Bias Cooling Time / Enrichment Interpolation 0.00036 Error _ _ _ _ _ _ _ _ _ _ Uncertainties Benchmark Method Uncertainty 0.01211 Monte Carlo Computational Uncertainty - Mechanical Uncertainties 0.01110 Burnup Computational Uncertainty 0.00413 Burnup Measurement Uncertainty 0.00125 Axial Profile Uncertainty 0.00305 Maximum 95/95 kOf 0.99888
Attachment 6 Page 45 of 48 9 Conclusions The criticality analysis for the McGuire NFVs and SFPs has been performed in accordance with the requirements of 10 CFR 50.68 (b). This evaluation takes credit for Boral poison material in the new SFP Region 1 storage racks, but no longer takes credit for any remaining Boraflex in the SFP Region 2 racks. Credit has been taken for burnup and cooling time reactivity reduction in Region 2. In addition, partial credit for soluble boron is employed in the SFPs. This analysis determined that the McGuire NFVs can store unirradiated MkBW (with or without axial blankets), W-RFA, and W-STD fuel up to 5 wt % U-235, with no location restrictions. Fresh W-OFA fuel up to 4.76 wt % U-235 may be stored in the NFVs with no location restrictions. The SFP criticality evaluation demonstrated that the Region 1 Boral racks can store fresh McGuire reactor fuel of any type, up to 5 wt % U-235, with no restrictions. The existing irradiated Oconee "MlkBI" assemblies in the McGuire SFPs may also be stored in the Region 1 racks without restriction. Minimum burnup requirements for SFP Region 2 storage were developed for seven different fuel types, as a function of initial enrichment and post-irradiation cooling time. These burnup requirements were specified for three defined Region 2 storage configurations: Unrestricted, 2/4 Restricted/Filler, and 3/4 Checkerboard/Empty. The following restrictions for adjacent storage of different fuel configurations in Region 2 of the SFPs were determined in this analysis:
- Unrestricted storage - No boundary restrictions.
- 2/4 Restricted/Filler storage - No boundary restrictions.
- 3/4 Checkerboard/Empty storage - Any row or column of fuel in a 3/4 Checkerboard/Empty storage configuration that borders any other storage configuration must have alternating Checkerboard fuel and empty cells. That is, it cannot be a row or column of solid Checkerboard fuel.
The maximum 95/95 kffs for the NFV analysis were calculated to be 0.9498 (NFV flooded with full-density unborated water) and 0.9618 (NFV flooded with optimum-moderation unborated "water"). These results meet the requirements of 10 CFR 50.68 (b) (2,3). For the SFP criticality analyses, the maximum 95/95 kffs with no boron in the SFP were calculated to be 0.9829 (SFP Region 1 storage) and 0.9989 (SFP Region 2 storage). These results meet the no-boron 95/95 keff criterion in 10 CFR 50.68 (b) (4).
Attachment 6 Page 46 of 48 The SFP criticality analysis confirmed that 800 ppm of partial soluble boron credit is sufficient to maintain the maximum 95/95 kff less than 0.95 for all normal conditions. The current minimum boron concentration required in the McGuire SFPs (2675 ppm) is adequate to maintain the maximum 95/95 keff below 0.95 for all credible accident conditions in the McGuire SFPs.
Attachment 6 Page 47 of 48 10 References
- 1. 'McGuire Nuclear Station, Units 1 and 2 Re: Issuance of Amendments for Spent Fuel Pool (TAC NOs MB5014 and MB5 015)," Letter from R.
Martin (NRC) to D. Jamil (Duke), February 4, 2003.
- 2. Parametric Study of the Effect of Burnable Poison Rods for PWR Burnup Credit, NUREG/CR-6761, Oak Ridge National Laboratory, March 2002.
- 3. SCALE 4A - A Modular Code System for Performing Standardized Computer Analyses for Licensing Evaluation, NUREG/CR-0200 (Rev. 5),
CCC-545, Oak Ridge National Laboratory, March 1997.
- 4. CASMO A Fuel Assembly Burnup Program, STUDSVIK/NFA-89/3, June 1993.
- 5. SIMULATE Advanced Three-Dimehsional Two-Group Reactor Analysis Code, STUDSVIKISOA-95/15, October 1995.
- 6. Criticality Experiments with Subcritical Clusters of 2.35 and 4.31 wt %
U-235 Enriched U02 Rods in Water at a Water to Fuel Volume Ratio of 1.6, PNL-3314, July 1980.
- 7. Critical Separation Between Subcritical Clusters of 2.35 wt % U-235 Enriched U02 Rods in Water with Fixed Neutron Poisons, PNL-2438, October 1977.
- 8. Criticality Experiments to Provide Benchmark Data on Neutron Flux Traps, PNL-6205, June 1988.
- 9. Critical Experiments Supporting Close Proximity Water Storage of Power Reactor Fuel, B&W-1484-7, July 1979.
- 10. "Guidance on the Regulatory Requirements for Criticality Analysis of Fuel Storage at Light-Water Reactor Power Plants", Memorandum from L. Kopp (NRC) to T. Collins (NRC), U.S. Nuclear Regulatory Commission, August 19, 1998.
- 11. Review of Axial Burnup Distribution Considerations for Burnup Credit Calculations, ORNL/TM-1999/246, Oak Ridge National Laboratory, March 2000.
- 12. 'Issuance of Amendments - Oconee Nuclear Station Units 1, 2, and 3 (TAC NOs M91043, M91044, and M91045)," Letter from L. Wiens (NRC) to J. Hampton (Duke), May 3, 1995.
Attachment 6 Page 48 of 48
- 13. Study of the Effect of Integral Burnable Absorbers for PWR Burnup Credit, NUREG/CR-6760, Oak Ridge National Laboratory, March 2002.
- 14. Determination of the Accuracy of Utility Spent Fuel Burnup Records, EPRI TR-1 12054, prepared for Electric Power Research Institute, July 1999.
- 15. NUREG/CR-6683, A Critical Review of the Practice of Equating the Reactivity of Spent Fuel to Fresh Fuel in Burnup Credit Criticality Safety Analyses for PWR Spent Fuel Pool Storage, J. Wagner and C.
Parks, Oak Ridge National Laboratory, prepared for the U.S. Nuclear Regulatory Commission, September 2000.
ATTACHMENT 7 MARKUPS TO THE MCGUIRE TECHNICAL SPECIFICATION BASES
Spent Fuel Pool Boron Concentration B 3.7.14 B 3.7 PLANT SYSTEMS B 3.7.14 Spent Fuel Pool Boron Concentration BASES BACKGROUNI ) In the two region poison fuel storage rack (Refs. 1 and 2) design, the spent fuel pool is divided into two separate and distinct regions. Region 1, with 286 storage positions, is designed and generally reserved for temporary storage of new or partially irradiated fuel. Region 2, with 1177 storage positions, Is designed and generally used for normal, long term storage of permanently discharged fuel that has achieved qualifying burnup levels.
;bMcGuire spent fuel storage racks contain Boraflex neutron-absorbng pan btat surround each storage cell on all four sides (except for peripherd ides). The function of these Boraflex panels is to ensure that the reactivif the stored fuel assemblies is maintained within required limits. Borafle, manufactured, is a silicon rubber material that retains a powder of boro arbide (B4C) neutron absorbing material. The Boraflex panels are en sed in a formed stainless steel wrapper sheet that is spot-welded to the rage tube. The wrapper sheet is bent at each end to complete the enc ure of the Boraflex panel. The Boraflex panel is contained in the plenum between the storage tube and the wrapper plate. Since the wrapper p te enclosure is not sealed, spent fuel pool water is free to circulate th gh the plenum. It has been observed that after Boraflex receives a hig amma dose from the stored irradiated fuel (>1010 rads) it can begin to deg de and dissolve in the wet environment. Thus, the B4C poison material n be removed, thereby reducing the poison worth of the Boraflex sheets. This phenomenon is documented in NRC Generic Letter 96-04, 'Bora x Degradation in Spent Fuel Pool Storage Racks".
To address this degradation, each region of the spent fuel po has been divided into two sub-regions; with and without credit for Borafle.For the regions taking credit for Boraflex, a minimum amount of Borafle was assumed that is less than the original design minimum B10 areal dens The McGuire spent fuel storage racks have been analyzed taking credit for soluble boron as allowed In Reference 3. The methodology ensures that the spent fuel rack multiplication factor, kef, is less than or equal to 0.95 as recommended in ANSI/ANS-57.2-1983 (Ref. 4) and NRC guidance (Ref. 5). The spent fuel storage racks are analyzed to allow storage of fuel assemblies with enrichments up to a maximum nominal enrichment of( veight percent Uranium-235 while maintaining krf<
.S. 4,0 McGuire Units 1 and 2 B 3.7.14-1 Revision No. O:F-
Spent Fuel Pool Boron Concentration B 3.7.14 BASES BACKGROUND (continued) 0.95 including uncertainties, tolerances, bias, and credit for soluble boron. Soluble boron credit is used to offset uncertainties, tolerances, and off-normal conditions and to provide subcritical margin such that the spent fuel pool kff is maintained less than or equal to 0.95. The soluble boron concentration required-to maintain kff less than or equal to 0.95 4eO' under normal conditions i§pm. In addition, sub-criticality of the pool (k.ff < 1.0) is assured on a 95/95 basis, without the presence of the soluble boron in the pool. The criticality analysis performed shows that the acceptance criteria for criticality is met for the storage of fuel assemblies when credit is taken for reactivity depletion. u~el bunufih-ereesence of Integral Fuel Burnatip Absorber CB)od due rS o h Borabx netro bor4 ge configurations and enrichment limits pecifie y 3.7.15. Pos~~otr.<vz to'o);,O , e-APPLICABLE Most accident conditions do not result in an increase in reactivity of the SAFETY ANALYSES racks in the spent fuel pool. Examples of these accident conditions are the drop of a fuel assembly on top of a rack, the drop of a fuel assembly between rack modules (rack design precludes this condition), and the drop of a fuel assembly between rack modules and the pool wall. However, three accidents can be postulated which could result in an increase in reactivity in the spent fuel storage pools. The first is a drop or placement of a fuel assembly into the cask loading area. The second is a significant change in the spent fuel pool water temperature (either the loss of normal cooling to the spent fuel pool water which causes an increase in the pool water temperature or a large makeup to the pool with cold water which causes a decrease in the pool water temperature) and the third is the misloading of a fuel assembly into a location which the restrictions on location, enrichment, bumup(anNnumber f IF EArods is not satisfied. A ntsife For an occurrence of these postulated accidents, the double contingency principle discussed in ANSI N-16.1-1975 and the April 1978 NRC letter (Ref. 6) can be applied. This states that one is not required to assume two unlikely, independent, concurrent events to ensure protection against a criticality accident. Thus, for these postulated accident conditions, the pre~senc of additional soluble boron in the spent fuel pool water (above e?'7C7 -iFi~5~)ppm required to maintain kff less than or equal to 0.95 under normal conditions) can be assumed as a realistic initial condition since not assuming its presence would be a second unlikely event. Calculations were performed to determine the amount of soluble boron required to offset the highest reactivity increase caused by either of McGuire Units 1 and 2 B 3.7.14-2 Revision No.-8;
Spent Fuel Pool Boron Concentration B 3.7.14 BASES APPLICABLE SAFETY ANALYSES (continued) these postulated accidents and to maintain kff less than or. u I-toZ It was found that a spent fuel pool boron concentration ofi m was adequate to mitigate these postulated criticality related accidents and to maintain ke~f less than or equal to 0.95. Specification 3.7.14 ensures the spent fuel pool contains adequate dissolved boron to compensate for the increased reactivity caused by these postulated accidents. Specification 4.3.1.1 c. requires that the spent fuel rack.ke be le s than ' or equal to 0.95 when flooded with water borated to@ ;Wr. A spent l fuel pool boron dilution analysis was performed which confirmed that sufficient time is available to detect and mitigate a dilution of the spent fuel pool before the 0.95 kff design basis is exceeded. The spent fuel pool boron dilution analysis concluded that an unplanned or inadvertent event which could esult in the dilution of the spent fuel pool boron concentration to ppm is not a credible event. The concentration of dissolved boron in the spent fuel pool satisfies Criterion 2 of 10 CFR 50.36 (Ref. 5). LCO The spent fuel pool boron concentration is required to be within the limits specified in the COLR. The specified concentration of dissolved boron in the spent fuel pool preserves the assumptions used in the analyses of the potential criticality accident scenarios as described in Reference 4. This concentration of dissolved boron is the minimum required concentration for fuel assembly storage and movement within the spent fuel pool. APPLICABILITY This LCO applies whenever fuel assemblies are stored in the spent fuel pool. ACTIONS A.1 and A.2 The Required Actions are modified by a Note indicating that LCO 3.0.3 does not apply. When the concentration of boron in the fuel storage pool is less than required, immediate action must be taken to preclude the occurrence of an accident or to mitigate the consequences of an accident in progress. This is most efficiently achieved by immediately suspending the movement of fuel assemblies. The concentration of boron is restored simultaneously with suspending movement of fuel assemblies. McGuire Units 1 and 2 B 3.7.14-3 Revision No. Spent Fuel Assembly Storage B 3.7.14 BASES ACTIONS (continued) If the LCO is not met while moving irradiated fuel assemblies in MODE 5 or 6, LCO 3.0.3 would not be applicable. If moving irradiated fuel assemblies while in MODE 1, 2, 3, or 4, the fuel movement is independent of reactor operation. Therefore, inability to suspend movement of fuel assemblies is not sufficient reason to require a reactor shutdown. SURVEILLANCE SR 3.7.14.1 REQUIREMENTS This SR verifies that the concentration of boron in the spent fuel pool is within the required limit. As long as this SR is met, the analyzed accidents are fully addressed. The 7day Frequency is appropriate because no major replenishment of pool water is expected to take place over such a short period of time. REFERENCES 1. UFSAR, Section 9.1.2.
- 2. Issuance of ene (TAC NOS Jar I S U 3. 441-6-NNA, I 0~ ~ ~ ~ nl Methodol gy, CB . /tea2Wa' 4. American Nuclear Soi ciety, "American National Standard Design Re - -act for LigI it Water Reactor Fuel Storage Facilities at ree<Requirements Nuclear Power Plants,' 'ANSI/ANS-57.2-1983, October 7,1983.
- 5. Nuclear Regulatory Commission, Memorandum to Timothy Collins from Laurence Kopp, "Guidance on the Regulatory Requirements for Criticality Analysis of Fuel Storage at Light Water Reactor Power Plants," August 19, 1998.
- 6. Double contingency principle of ANSI N16.1-1975, as specified in the April 14,1978 NRC letter (Section 1.2) and implied in the proposed revision to Regulatory Guide 1.13 (Section 1.4, Appendix A).
- 7. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).
- 8. UFSAR, Section 15.7.4.
McGuire Units 1 and 2 B 3.7.14-4 Revision No. Spent Fuel Assembly Storage B 3.7.15 B 3.7 PLANT SYSTEMS B 3.7.15 Spent Fuel Assembly Storage BASES BACKGROUND In the two region poison fuel storage rack (Refs. 1 and 2) design, the spent fuel pool is divided into two separate and distinct regions. Region 1, with 286 storage positions, Is designed and generally reserved for temporary storage of new or partially irradiated fuel. Region 2, with 1177 storage positions, is designed and generally used for normal, long term storage of permanently discharged fuel that has achieved qualifying buMuD levels. i cGuitre spent fuel storage racks ontain Boraflex neutron-absorbing pa Is that surround each storage ce on all four sides (except for perip ral sides). The function of these B aflex panels is to ensure that the rea ivity of the stored fuel assemblies maintained within required limits. flex, as manufactured, is a silicon bber material that retains a powder boron carbide (B4C) neutron sorbing material. The Boraflex panes are enclosed in a formed stainl s steel wrapper sheet that is spot-we ed to the storage tube. The wrpper sheet is bent at
-each end to com ete the-enclosure of the Boraflex anel. The Boraflex panel is contained the plenum area between the s rage tube and the wrapper plate. Sinc the wrapper plate enclosure is ot sealed, spent fuel pool water is fre to circulate through the plenum. It has been observed that after Bora receives a high gamma dose rom the stored irradiated fuel (>1010 rads) can begin to degrade and dis lye in the wet nvironment. Thus, the B4 oison material can be rem ed, thereby r ucing the poison worth of t Boraflex sheets. This phe omenon is domented in NRC Generic L tter 96-04, "Boraflex Degr dation in Spe Fuel Pool Storage Racks".
To addre this degradation, each regi of the spent fuel pool ha been divided into o sub-regions; with and wit Ut credit for Boraflex. For the regions takincredit for Boraflex, a mini m amount of Boraflex as assumed that i ess than the original design inimum B 0 areal densi Two storage config ations are defined for each gion; Unrestricted and Restricted storage. nrestricted storage allow storage in all cells without restriction on th storage configuration. Res cted storage allows storage of higher reactivi uel when restricted to a ceyin storage McGuire Units 1 and 2 B 3.7.15-1 Revision No.-&7 -
Spent Fuel Assembly Storage B 3.7.15 BASES BACKGROUND (continued)
'confiu ston with lower Recivity fuel. A tiloa ing pattem, Ceckebadsoae a fined for Regions 9 2A and 2B.
Checkerodsoaealw trg of the highest reacit fuel in each 42^@e$5 { ;5 ) Ache to 1 McGuire spent fuel storage racks have been analyzed taking credit ,e<>,<,,>*,c } 'for soluble boron as allowed in Reference 3. The methodology ensures M3/ that the spent fuel rack multiplication factor, keff, is less than or equal to 0.95 as recommended in ANSIIANS-57.2-1983 (Ref. 4) and NRC guidance (Ref. 5). The spent fuel storage racks are analyzed to allow storage of f el assemblies with enrichments up to a maximum nominal S^°0 T enrichment weight percent Uranium-235 while maintaining kff < 0.95 including uncertainties, tolerances, bias, and credit for soluble boron. Soluble boron credit is used to offset uncertainties, tolerances, and off-normal conditions and to provide subcritical margin such that the spent fuel pool kef is maintained less than or equal to 0.95. The soluble boron concentration re uired o maintain keff less than or equal to 0.95 under normal conditio pm. In addition, sub-criticality of the pool l (kef < 1.0) is assured on a 95/95 basis, without the presence of the soluble boron in the pool. The criticality analysis performed hows the acceptance criteria for criticality is met for the storage of fuel assemblies when credit is taken for reactivity depletion due to fuel burn up,J415i@s-enceo rirlFelB-l boe (IB-A od 4 5* configurations and enrichent limits Pecified by LCO 3.7.15. _) APPLICABLE Most accident conditions do not result in an increase in reactivity of the SAFETY ANALYSES racks in the spent fuel pool. Examples of these accident conditions are the drop of a fuel assembly on top of a rack, the drop of a fuel assembly between rack modules (rack design precludes this condition), and the drop of a fuel assembly between rack modules and the pool wall. However, three accidents can be postulated which could result in an increase in reactivity in the spent fuel storage pools. The first is a drop or placement of a fuel assembly into the cask loading area. The second is a significant change in the spent fuel pool water temperature (either the loss of normal cooling to the spent fuel pool water which causes an increase in the pool water temperature or a large makeup to the pool with cold water which causes a decrease in the pool water temperature) and the third is the misleading of a fuel assembly into a location which the restrictions on location, enrichment, burnup and(u;er of BAldds is not~. < For an occurrence of these postulated accidents, the double contingency principle discussed in ANSI N-16.1-1975 and the April 1978 NRC letter McGuire Units 1 and 2 B 3.7.1 5-2 Revision No.-67*
Spent Fuel Assembly Storage B 3.7.15 BASES APPLICABLE SAFETY ANALYSES (continued) (Ref. 6) can be applied. This states that one is not required to assume two unlikely, independent, concurrent events to ensure protection against a criticality accident. Thus, for these postulated accident conditions, the presence of additional soluble boron in the spent fuel pool water (above s5" ppm required to maintain khf less than or equal to 0.95 under normal conditions) can be assumed as a realistic initial condition since not assuming its presence would be a second unlikely event. Calculations were performed to determine the amount of soluble boron required to offset the highest reactivity increase caused by either of these postulated accidents and to maintain kef less than or eqa to .95. ig was found that a spent fuel pool boron concentration ofDpPm was adequate to mitigate these postulated criticality related accidents and to maintain kff less than or equal to 0.95. Specification 3.7.14 ensures the spent fuel pool contains adequate dissolved boron to compensate for the increased reactivity caused by these postulated accidents. Specification 4.3.1.1 c. requires that the spent fuel raL 8 be less thanj or equal to 0.95 when flooded with water borated tog 'pm.A spent I fuel pool boron dilution analysis was performed which confirmed that sufficient time is available to detect and mitigate a dilution of the spent fuel pool before the 0.95 keff design basis is exceeded. The spent fuel pool boron dilution analysis concluded that an unplanned or inadvertent event which could result in the dilution of the spent fuel pool boron DOC b concentration
= tp ppm is not a credible event.
The configuration of fuel assemblies in the spent fuel pool satisfies Criterion 2 of 10 CFR 50.36 (Ref. 7). McGuire Units 1 and 2 B 3.7.15-3 Revision No.-&7-
Spent Fuel Assembly Storage B 3.7.15 BASES LCO (continued) restrictions on the place nt of fuel assemblies in the Region 1B 0of tspent fuel pool, which hoe accumulated bumu greater than or equal the minimum qualifie bumups in Table 3.15-4 in the accompa ng LCO, ensures the Woof the spent fuel po will always remain < assuming the pool to be ooded with water bo ted to 850 ppm. Fuel ass blies not meeting the cura of Table 3.7.15 shall be stored in accorda e with either Figure 3.. 5-2 and Table 3. 15-5 for Restricted storage, o igure 3.7.15-3 for Chec rboard storage. 2 Th restrictions on the placem of fuel assemblies wi n the Regio' of th spent fuel pool, which ha accumulated burnup reater than r equal Xvthe minimum qualified rnups in Table 3. 15-7 in the accompanl ng LCO, ensures the kff the spent fuel poo will always remain <0. O;sassuming the pool to be f ed with water bor ed to 850 ppm. Fuel ass blies not meeting the crit 'a of Table 3.7.15- hall be stored in accord ce with either Figure 3.7.1 4 and Table 3.7.1 -8 for Restricted storage, Figure 3.7.15-5 for Checke oard storage. Th restrictions on the place nt of fuel assemblies within e Region 2B of th spent fuel pool, which h1e accumulated burnup grear than or equal o the minimum qualifie bumups in Table 3.7.15-1 in the accomp ying LCO, ensures the of the spent fuel pool will ways remain < 95, assuming the pool to b flooded with water borated to 50 ppm. Fuel ssemblies not meeting th criteria of Table 3.7.15-10 sh be stored in cordance with either Figur 3.7.15-6 and Table 3.7.15-11 for Restricted strage, or Figure 3.7.15-7 for heckerboard storage. APPLICABILITY This LCO applies whenever any fuel assembly is stored in the spent fuel pool. ACTIONS A.1 Required Action A.1 is modified by a Note indicating that LCO 3.0.3 does not apply. When the configuration of fuel assemblies stored in the spent fuel pool is not in accordance with the LCO, the immediate action is to initiate action McGuire Units 1 and 2 B 3.7.15-4 Revision No..&7-
Spent Fuel Assembly Storage B 3.7.15 BASES LCO (continued) to make the necessary fuel assembly movement(s) to bring the configuration into compliance.' If unable to move irradiated fuel assemblies while in MODE 5 or 6, LCO 3.0.3 would not be applicable. If unable to move irradiated fuel assemblies while in MODE 1, 2, 3, or 4, the action is independent of reactor operation. Therefore, inability to move fuel assemblies is not sufficient reason to require a reactor shutdown. SURVEILLANCE SR 3.7.15.1 REQUIREMENTS This SR verifies by administrative means that the fuel assembly is in accordance with the configurations specified in the accompanying LCO. REFERENCES 1. UFSAR, Section 9.1.2. 5,tv/ 2. Issuance of Amendments, McGuire Nuclear Station. Units 1 and 2 mg $lI (TAC NOf and MA973 ), November 27, 2000 1 MA=i94' . .._.
-C..........
- 3. P-141 6-NP gho e Spent FuN RackXCriticaliW9 lie
- 4. American Nuclear Society, uAmerican National Standard Design Requirements for Light Water Reactor Fuel Storage Facilities at Nuclear Power Plants," ANSI/ANS-57.2-1983, October 7,1983.
- 5. Nuclear Regulatory Commission, Memorandum to Timothy Collins from Laurence Kopp, "Guidance on the Regulatory Requirements for Criticality Analysis of Fuel Storage at Light Water Reactor Power Plants," August 19,1998.
- 6. Double contingency principle of ANSI N16.1-1975, as specified in the April 14,1978 NRC letter (Section 1.2) and implied in the proposed revision to Regulatory Guide 1.13 (Section 1.4, Appendix A).
- 7. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).
_ri S~.d .. ; n e.c McGuire Units 1 and 2 B 3.7.15-5 Revision No.-ff -
INSERT A The McGuire Region 1 spent fuel storage racks are composed of individual cells made of stainless steel. These racks utilize Boral, a boron carbide aluminum cermet, as the neutron absorber material. The cells within a module are interconnected at six locations along the length of the cell using spacer plates to form an integral structure. Depending on the criticality requirements, some cells have a Boral wrapper on all four sides, some on three sides and some on two sides. The Region 1 racks will store the most reactive fuel (up to 5.00 weight percent Uranium-235 enrichment) without any bumup limitations. Boral is a thermal neutron poison material composed of boron carbide and 1100 alloy aluminum. Boron carbide is a compound having a high boron content in a physical stable and chemically inert form. The 1100 alloy aluminium is a lightweight metal with high tensile strength, which is protected from corrosion by a highly resistant oxide film. Boron carbide and aluminum are chemically compatible and ideally suited for long-term use in a spent fuel pool environment. The McGuire Region 2 spent fuel storage racks contain Boraflex neutron-absorbing panels that surround each storage cell on all four sides (except for peripheral sides). It has been observed that after Boraflex receives a high gamma dose from the stored irradiated fuel (>010 rads) it can begin to degrade and dissolve in the wet environment. Thus, the B4C poison material can be removed, thereby reducing the poison worth of the Boraflex sheets. This phenomenon is documented in NRC Generic Letter 96-04, "Boraflex Degradation in Spent Fuel Pool Storage Racks". INSERT B No credit is taken for the Boraflex neutron absorber panels. The criticality analysis performed for Region 1 shows that the acceptance criteria for criticality is met for unrestricted storage without credit for burnup or plutonium decay of fuel assemblies with enrichments up to a maximum nominal value of 5.00 weight percent Uranium-235. The storage criteria for fuel stored in Region 2 of the spent fuel pool is based upon criticality analysis that was performed in accordance with the criteria of 10 CFR 50.68(b). The fuel storage requirements are defined as a function of enrichment, burnup, cooling time and fuel type. The following are the fuel types considered in the criticality analyses: MkBI - This generic fuel type represents the old Oconee 15x15 MkB2, MkB3, and MkB4 fuel assembly designs, which used Inconel spacer grids in the active fuel area. 300 of these assemblies, which operated in the Oconee reactors, were transshipped to McGuire.
W-STD - This is the standard 17x17 Westinghouse fuel design which was used in the initial cycles (batches 1-3) of both the McGuire reactors. At that time the W-STD design had Inconel grids. W-OFA - This is the 17x17 Westinghouse "Optimized Fuel Assembly" design, which had thin rods, Zircaloy grids, and a low total uranium loading. This design was deployed for batches 4 through 9 in both McGuire units. MkBW - This is the standard 17x1 7 Framatome (B&W) fuel design which was modeled after the standard Westinghouse product. The MkBW design contains Zircaloy grids. This fuel type (without axial blankets) was used for batches 10 through 13 in both McGuire reactors. MkBWb1 - This is the same design as the standard MkBW, but it employs solid, 6-inch, 2.00 wt % U-235 axial blankets at the top and bottom of the active fuel zone. This fuel type was used in McGuire Unit 1, batches 14 to 16, and McGuire Unit 2, batch 14. MkBWb2 - - This is also the same design as the standard MkBW, but it employs solid, 6-inch, 2.60 wt % U-235 axial blankets at the top and bottom of the active fuel zone. This fuel type was used in McGuire Unit 2, batch 15. W-RFA - This is the advanced 17x17 Westinghouse fuel design. It is similar to the MkBW assembly design, and contains Zircaloy grids, but uses annular, 6-inch, 2.60 wt % U-235 axial blankets at the top and bottom of the active fuel zone. This fuel type has been chosen for McGuire Unit 1, batches 17 to present, and McGuire Unit 2, batches 16 to present. INSERT C a Unrestricted storage of fuel assemblies within Region 1 of the spent fuel pool is allowed provided that the maximum nominal Uranium-235 enrichment is equal to or less than 5.00 weight percent. This ensures the kff of the spent fuel pool will always remain < 0.95, assuming the pool is flooded with water borated to 800 ppm. b The restrictions on the placement of fuel assemblies within Region 2 of the spent fuel pool, which have accumulated bumup greater than or equal to the minimum qualified bumups and which have decayed greater than or equal to the minimum qualified cooling time in Table 3.7.15-1 in the accompanying LCO, ensures the keff of the spent fuel pool will always remain < 0.95, assuming the pool to be flooded with water borated to 800 ppm. Fuel assemblies not meeting the criteria of Table 3.7.15-1 shall be stored in accordance with Figure 3.7.15-1 per the initial
enrichment, burnup and decay time criteria specified by Tables 3.7.15-2 and 3.7.15-3 for restricted/filler storage configuration. Another acceptable storage configuration is described by Figure 3.7.15-2 for fuel assemblies that satisfy the initial enrichment, burnup and decay time criteria specified in Table 3.7.15-4 for Checkerboard storage.
ATTACHMENT 8 REVISED TECHNICAL SPECIFICATION BASES
Spent Fuel Pool Boron Concentration B 3.7.14 B 3.7 PLANT SYSTEMS B 3.7.14 Spent Fuel Pool Boron Concentration BASES BACKGROUND In the two region poison fuel storage rack (References. 1 and 2) design, the spent fuel pool is divided into two separate and distinct regions. Region 1, with 286 storage positions, is designed and generally reserved for temporary storage of new or partially irradiated fuel. Region 2, with 1177 storage positions, is designed and generally used for normal, long term storage of permanently discharged fuel that has achieved qualifying burnup levels. The McGuire spent fuel storage racks have been analyzed taking credit for soluble boron as allowed in Reference 3. The methodology ensures that the spent fuel rack multiplication factor, kef, is less than or equal to 0.95 as recommended in ANSI/ANS-57.2-1983 (Reference 4) and NRC guidance (Reference. 5). The spent fuel storage racks are analyzed to allow storage of fuel assemblies with enrichments up to a maximum nominal value of 5.00 weight percent Uranium-235 while maintaining kff< 0.95 including uncertainties, tolerances, biases, and credit for soluble boron. Soluble boron credit is used to offset off-normal conditions and to provide subcritical margin such that the spent fuel pool kff is maintained less than or equal to 0.95. The soluble boron concentration required to maintain kff less than or equal to 0.95 under normal conditions is 800 ppm. In addition, sub-criticality of the pool (kff < 1.0) is assured on a 95/95 basis, without the presence of the soluble boron in the pool. The criticality analysis performed shows that the regulatory subcriticality requirements are met for fuel assembly storage within an allowable storage configuration, when the criteria for fuel assembly type, initial enrichment, bumup, and post-irradiation cooling time, as specified in LCO 3.7.15, are satisfied. APPLICABLE Most accident conditions do not result in an increase in reactivity of the SAFETY ANALYSES racks in the spent fuel pool. Examples of these accident conditions are the drop of a fuel assembly on top of a rack, the drop of a fuel assembly between rack modules (rack design precludes this condition), and the drop of a fuel assembly between rack modules and the pool wall. However, three accidents can be postulated which could result in an increase in reactivity in the spent fuel storage pools. The first is a drop or placement of a fuel assembly into the cask loading area. The second is a significant change in the spent fuel pool water temperature (either the loss of normal cooling to the spent fuel pool water which causes an increase in the pool water temperature or a large makeup to the pool with cold water which causes a decrease in the pool water temperature) and B 3.7.14-1 Revision No. McGuire Units and 2 1 and Units 1 2 B 3.7.14-1 Revision No.
Spent Fuel Pool Boron Concentration B 3.7.14 Bases APPLICABLE SAFETY ANALYSES (continued) the third is the misleading of a fuel assembly into a location which the restrictions on location, enrichment, and burnup are not satisfied. For an occurrence of these postulated accidents, the double contingency principle discussed in ANSI N-16.1-1975 and the April 1978 NRC letter (Reference. 6) can be applied. This states that one is not required to assume two unlikely, independent, concurrent events to ensure protection against a criticality accident. Thus, for these postulated accident conditions, the presence of additional soluble boron in the spent fuel pool water (above the 800 ppm required to maintain kff less than or equal to 0.95 under normal conditions) can be assumed as a realistic initial condition since not assuming its presence would be a second unlikely event. Calculations were performed to determine the amount of soluble boron required to offset the highest reactivity increase caused by either of these postulated accidents and to maintain k.fl less than or equal to 0.95. It was determined that a spent fuel pool boron concentration of 1600 ppm was adequate to mitigate these postulated criticality related accidents and to maintain kff less than or equal to 0.95. Specification 3.7.14 ensures the spent fuel pool contains adequate dissolved boron to compensate for the increased reactivity caused by these postulated accidents. Specification 4.3.1.1 c. requires that the spent fuel rack kff be less than or equal to 0.95 when flooded with water borated to 800 ppm. A spent fuel pool boron dilution analysis was performed which confirmed that sufficient time is available to detect and mitigate a dilution of the spent fuel pool before the 0.95 kff design basis is exceeded. The spent fuel pool boron dilution analysis concluded that an unplanned or inadvertent event which could result in the dilution of the spent fuel pool boron concentration to 800 ppm is not a credible event. The concentration of dissolved boron in the spent fuel pool satisfies Criterion 2 of 10 CFR 50.36 (Reference. 5). LCO The spent fuel pool boron concentration is required to be within the limits specified in the COLR. The specified concentration of dissolved boron in the spent fuel pool preserves the assumptions used in the analyses of the potential criticality accident scenarios as described in Reference 4. This concentration of dissolved boron is the minimum required concentration for fuel assembly storage and movement within the spent fuel pool. B 3.7.14-2 Revision No. McGuire Units and 2 1 and Units 1 2 B 3.7.14-2 Revision No.
Spent Fuel Pool Boron Concentration B 3.7.14 Bases APPLICABILITY This LCO applies whenever fuel assemblies are stored in the spent fuel pool. ACTIONS A.1 and A.2 The Required Actions are modified by a Note indicating that LCO 3.0.3 does not apply. When the concentration of boron in the fuel storage pool is less than required, immediate action must be taken to preclude the occurrence of an accident or to mitigate the consequences of an accident in progress. This is most efficiently achieved by immediately suspending the movement of fuel assemblies. The concentration of boron is restored simultaneously with suspending movement of fuel assemblies. If the LCO is not met while moving irradiated fuel assemblies in MODE 5 or 6, LCO 3.0.3 would not be applicable. If moving irradiated fuel assemblies while in MODE 1, 2, 3, or 4, the fuel movement is independent of reactor operation. Therefore, inability to suspend movement of fuel assemblies is not sufficient reason to require a reactor shutdown. SURVEILLANCE SR 3.7.14.1 REQUIREMENTS This SR verifies that the concentration of boron in the spent fuel pool is within the required limit. As long as this SR is met, the analyzed accidents are fully addressed. The 7 day Frequency is appropriate because no major replenishment of pool water is expected to take place over such a short period of time. REFERENCES 1. UFSAR, Section 9.1.2.
- 2. Issuance of Amendments, McGuire Nuclear Station, Units 1 and 2 (TAC NOS. MB5014 and MB5015), February 4, 2003. I
- 3. 10 CFR 50.68, "Criticality Accident Requirements" I
- 4. American Nuclear Society, "American National Standard Design Requirements for Light Water Reactor Fuel Storage Facilities at Nuclear Power Plants," ANSI/ANS-57.2-1983, October 7, 1983.
- 5. Nuclear Regulatory Commission, Memorandum to Timothy Collins from Laurence Kopp, "Guidance on the Regulatory Requirements for Criticality Analysis of Fuel Storage at Light Water Reactor Power Plants," August 19, 1998.
McGuire Units 1 and 2 B 3.7.14-3 Revision No.
Spent Fuel Pool Boron Concentration B 3.7.14 Bases REFERENCES (continued)
- 6. Double contingency principle of ANSI N16.1-1975, as specified in the April 14,1978 NRC letter (Section 1.2) and implied in the proposed revision to Regulatory Guide 1.13 (Section 1.4, Appendix A).
- 7. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).
- 8. UFSAR, Section 15.7.4.
B 3.7.14-4 Revision No. McGuire Units and 22 I and Units 1 B 3.7.14-4 Revision No.
Spent Fuel Assembly Storage B 3.7.15 B 3.7 PLANT SYSTEMS B 3.7.15 Spent Fuel Assembly Storage BASES BACKGROUND In the two region poison fuel storage rack (Refs. 1 and 2) design, the spent fuel pool is divided into two separate and distinct regions. Region 1, with 286 storage positions, is designed and generally reserved for temporary storage of new or partially irradiated fuel. Region 2, with 1177 storage positions, is designed and generally used for normal, long term storage of permanently discharged fuel that has achieved qualifying burnup levels. The McGuire Region 1 spent fuel storage racks are composed of individual cells made of stainless steel. These racks utilize Boral, a boron carbide aluminum cermet, as the neutron absorber material. The cells within a module are interconnected at six locations along the length of the cell using spacer plates to form an integral structure. Depending on the criticality requirements, some cells have a Boral wrapper on all four sides, some on three sides and some on two sides. The Region 1 racks will store the most reactive fuel (up to 5.00 weight percent Uranium-235 enrichment) without any burnup limitations. Boral is a thermal neutron poison material composed of boron carbide and 1100 alloy aluminum. Boron carbide is a compound having a high boron content in a physical stable and chemically inert form. The 1100 alloy aluminum is a lightweight metal with high tensile strength, which is protected from corrosion by a highly resistant oxide film. Boron carbide and aluminum are chemically compatible and ideally suited for long-term use in a spent fuel pool environment. The McGuire Region 2 spent fuel storage racks contain Boraflex neutron-absorbing panels that surround each storage cell on all four sides (except for peripheral sides). It has been observed that after Boraflex receives a high gamma dose from the stored irradiated fuel (>1010 rads) it can begin to degrade and dissolve in the wet environment. Thus, the B4C poison material can be removed, thereby reducing the poison worth of the Boraflex sheets. This phenomenon is documented in NRC Generic Letter 96-04, "Boraflex Degradation in Spent Fuel Pool Storage Racks". To address this degradation, the McGuire spent fuel storage racks (both Regions) have been analyzed taking credit for soluble boron as allowed in Reference 3. The methodology ensures that the spent fuel rack multiplication factor, kf, is less than or equal to 0.95 as recommended in ANSI/ANS-57.2-1983 (Ref. 4) and NRC guidance (Ref. 5). The spent fuel storage racks are analyzed to allow storage of fuel assemblies with enrichments up to a maximum nominal enrichment of 5.00 weight percent Uranium-235 while maintaining kff < 0.95 including uncertainties, McGuire Units 1 and 2 B 3.7.15-1 Revision No.
Spent Fuel Assembly Storage B 3.7.15 BASES BACKGROUND (continued) tolerances, biases, and credit for soluble boron. Soluble boron credit is used to offset off-normal conditions and to provide subcritical margin such that the spent fuel pool k.f is maintained less than or equal to 0.95. The soluble boron concentration required to maintain keff less than or equal to 0.95 under normal conditions is at least 800 ppm. In addition, sub-criticality of the pool (kef < 1.0) is assured on a 95/95 basis, without the presence of the soluble boron in the pool. The criticality analysis performed for Region 2 shows that the regulatory subcriticality requirements are met for fuel assembly storage within an allowable storage configuration, when the criteria for fuel assembly type, initial enrichment, burnup, and post-irradiation cooling time, as specified in LCO 3.7.15, are satisfied. No credit is taken for the Boraflex neutron absorber panels in Region 2. The criticality analysis performed for Region 1 shows that the acceptance criteria for subcriticality are met for unrestricted storage of unirradiated fuel assemblies with enrichments up to a maximum nominal value of 5.00 weight percent Uranium-235. The storage criteria for fuel stored in Region 2 of the spent fuel pool is based upon criticality analysis that was performed in accordance with the criteria of 10 CFR 50.68(b). The fuel storage requirements are defined as a function of enrichment, burnup, cooling time and fuel type. The following are the fuel types considered in the criticality analyses: MkBI - This generic fuel type represents the old Oconee 15x15 MkB2, MkB3, and MkB4 fuel assembly designs, which used Inconel spacer grids in the active fuel area. 300 of these assemblies, which operated in the Oconee reactors, were transshipped to McGuire. W-STD - This is the standard 17x17 Westinghouse fuel design which was used in the initial cycles (batches 1-3) of both the McGuire reactors. At that time the W-STD design had Inconel grids. W-OFA - This is the 17x1 7 Westinghouse "Optimized Fuel Assembly' design, which had thin rods, Zircaloy grids, and a low total uranium loading. This design was deployed for batches 4 through 9 in both McGuire units. MkBW - This is the standard 17x17 Framatome (B&W) fuel design which was modeled after the standard Westinghouse product. The MkBW design contains Zircaloy grids. This fuel type (without axial blankets) was used for batches 10 through 13 in both McGuire reactors. MkBWb1 - This is the same design as the standard MkBW, but it employs solid, 6-inch, 2.00 wt % U-235 axial blankets at the top and bottom of the active fuel zone. This fuel type was used in McGuire Unit 1, batches 14 to 16, and McGuire Unit 2, batch 14. B 3.7.15-2 Revision No. McGuire Units 1 McGuire Units and 2 1 and 2 B 3.7.15-2 Revision No.
Spent Fuel Assembly Storage B 3.7.15 BASES MkBWb2 - - This is also the same design as the standard MkBW, but it employs solid, 6-inch, 2.60 wt % U-235 axial blankets at the top and bottom of the active fuel zone. This fuel type was used in McGuire Unit 2, batch 15. W-RFA - This is the advanced 17x1 7 Westinghouse fuel design. It is similar to the MkBW assembly design, and contains Zircaloy grids, but uses annular, 6-inch, 2.60 wt % U-235 axial blankets at the top and bottom of the active fuel zone. This fuel type has been chosen for McGuire Unit 1, batches 17 to present, and McGuire Unit 2, batches 16 to present." APPLICABLE Most accident conditions do not result in an increase in reactivity of the SAFETY ANALYSES racks in the spent fuel pool. Examples of these accident conditions are the drop of a fuel assembly on top of a rack, the drop of a fuel assembly between rack modules (rack design precludes this condition), and the drop of a fuel assembly between rack modules and the pool wall. However, three accidents can be postulated which could result in an increase in reactivity in the spent fuel storage pools. The first is a drop or placement of a fuel assembly into the cask loading area. The second is a significant change in the spent fuel pool water temperature (either the loss of normal cooling to the spent fuel pool water which causes an increase in the pool water temperature or a large makeup to the pool with cold water which causes a decrease in the pool water temperature) and the third is the misloading of a fuel assembly into a location which the restrictions on location, enrichment, bumup and decay time is not met. For an occurrence of these postulated accidents, the double contingency principle discussed in ANSI N-16.1-1975 and the April 1978 NRC letter (Ref. 6) can be applied. This states that one is not required to assume two unlikely, independent, concurrent events to ensure protection against a criticality accident. Thus, for these postulated accident conditions, the presence of additional soluble boron in the spent fuel pool water (above 800 ppm required to maintain kff less than or equal to 0.95 under normal conditions) can be assumed as a realistic initial condition since not assuming its presence would be a second unlikely event. Calculations were performed to determine the amount of soluble boron required to offset the highest reactivity increase caused by either of these postulated accidents and to maintain kff less than or equal to 0.95. It was found that a spent fuel pool boron concentration of 1600 ppm was B 3.7.15-3 Revision No. McGuire and 22 Units 11 and McGuire Units B 3.7.15-3 Revision No.
Spent Fuel Assembly Storage B 3.7.15 BASES APPLICABLE SAFETY ANALYSES (continued) adequate to mitigate these postulated criticality related accidents and to maintain keff less than or equal to 0.95. Specification 3.7.14 ensures the spent fuel pool contains adequate dissolved boron to compensate for the increased reactivity caused by these postulated accidents. Specification 4.3.1.1 c. requires that the spent fuel rack kff be less than or equal to 0.95 when flooded with water borated to 800 ppm. A spent fuel pool boron dilution analysis was performed which confirmed that sufficient time is available to detect and mitigate a dilution of the spent fuel pool before the 0.95 kf design basis is exceeded. The spent fuel pool boron dilution analysis concluded that an unplanned or inadvertent event which could result in the dilution of the spent fuel pool boron concentration to 800 ppm is not a credible event. The configuration of fuel assemblies in the spent fuel pool satisfies Criterion 2 of 10 CFR 50.36 (Ref. 7). LCO a Unrestricted storage of fuel assemblies within Region 1 of the spent fuel pool is -allowed provided that the maximum nominal Uranium-235 enrichment is equal to or less than 5.00 weight percent. This ensures the keff of the spent fuel pool will always remain < 0.95, assuming the pool is flooded with water borated to 800 ppm. b The restrictions on the placement of fuel assemblies within Region 2 of the spent fuel pool, which have accumulated burnup greater than or equal to the minimum qualified bumups and which have decayed greater than or equal to the minimum qualified cooling time in Table 3.7.15-1 in the accompanying LCO, ensures the kff of the spent fuel pool will always remain < 0.95, assuming the pool to be flooded with water borated to 800 ppm. Fuel assemblies not meeting the criteria of Table 3.7.15-1 mayl be stored in accordance with Figure 3.7.15-1 per the initial enrichment, burnup and decay time criteria specified by Tables 3.7.15-2 and 3.7.15-3 for restricted/filler storage configuration. Another acceptable storage configuration is described by Figure 3.7.15-2 for fuel assemblies that satisfy the initial enrichment, bumup and decay time criteria specified in Table 3.7.15-4 for Checkerboard storage. APPLICABILITY This LCO aDplies whenever any fuel assembly is stored in the sWent fuel pool.
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Spent Fuel Assembly Storage B 3.7.15 BASES ACTIONS A.1 Required Action A.1 is modified by a Note indicating that LCO 3.0.3 does not apply. When the configuration of fuel assemblies stored in the spent fuel pool is not in accordance with the LCO, the immediate action is to initiate action to make the necessary fuel assembly movement(s) to bring the configuration into compliance. If unable to move irradiated fuel assemblies while in MODE 5 or 6, LCO 3.0.3 would not be applicable. If unable to move irradiated fuel assemblies while in MODE 1, 2, 3, or 4, the action is independent of reactor operation. Therefore, inability to move fuel assemblies is not sufficient reason to require a reactor shutdown. SURVEILLANCE SR 3.7.15.1 REQUIREMENTS This SR verifies by administrative means that the fuel assembly is in accordance with the configurations specified in the accompanying LCO. REFERENCES 1. UFSAR, Section 9.1.2.
- 2. Issuance of Amendments, McGuire Nuclear Station, Units 1 and 2 (TAC NOS. MB5014 and MB5015), February 4, 2003. I
- 3. 10 CFR 50.68, 'Criticality Accident Requirements". I
- 4. American Nuclear Society, "American National Standard Design Requirements for Light Water Reactor Fuel Storage Facilities at Nuclear Power Plants," ANSI/ANS-57.2-1983, October 7, 1983.
- 5. Nuclear Regulatory Commission, Memorandum to Timothy Collins from Laurence Kopp, "Guidance on the Regulatory Requirements for Criticality Analysis of Fuel Storage at Light Water Reactor Power Plants," August 19, 1998.
- 6. Double contingency principle of ANSI N16.1-1975, as specified in the April 14,1978 NRC letter (Section 1.2) and implied in the proposed revision to Regulatory Guide 1.13 (Section 1.4, Appendix A).
- 7. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).
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