ML20202F456

Proposed Tech Specs,Deleting Lower Flow Rate Requirement Associated W/Nonaccessible Area Exhaust Filter Plenum & Fuel Handling Bldg Ventilation Sys

ML20202F456
Person / Time
Site:	Byron, Braidwood
Issue date:	10/10/1997
From:	COMMONWEALTH EDISON CO.
To:
Shared Package
ML20202F452	List: ML20202F448 ML20202F456
References
NUDOCS 9712090087
Download: ML20202F456 (254)

Text

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-Attachment 1 Improved Technical Specifications (ITS)

Affected Page List and Instructions 4

Enclosure 1: The Nonaccessible Area Exhaust Filter Plenum and the Fuel Handling Building Ventilation Systems License Amendment Request (LAR).

SECTION/ TAB REMOVE REV. A (PAGE) INSERT REV. B (PAGE) _.

3.7 BYRON ITS 3.7-32 3.7-32 i- 3.7 BYRON ITS 3.7-36 3.7 1 3.7 BYRON ITS B 3.7-77 8 3.7-77 3.7 BYRON ITS B 3.7-78 8 3.7-78 3.7 BYRON ITS B 3.7-84 8 3.7-84 3.7 BYRON ITS B 3.7-86 B 3.7-86 3.7 BYRON ITS B 3.7-93 B 3.7-93 3.7 BRWD ITS 3.7-28 3.7-28

3.7 BRWD ITS 3.7-32 3.7-32 l 3.7 BRWD ITS B 3.7-68 B 3.7-68 3.7 BRWD ITS B 3.7-69 B 3.7-69 I 3.7 BRWD ITS B 3.7-75 B 3.7-75 3.7 BRWD ITS B 3.7-77 B 3.7-77 3.7 BRWD ITS B 3.7-84 B 3.7-84 3.7 BYRON CTS MARKUPS 3/4 7-20 3/4 7-20 3.7 BYRON CTS MARKUPS 3/4 9-16 3/4 9-16 3.7-BRWD CTS MARKUPS 3/4 7-18 3/4 -18 3.7-BRWD CT' ARKUPS 3/4 9-16 3/4 9-16 3.7 CTS DOL 3.7-12 3.7-12 3.7 CTS DOCS 3.7-17 3.7-17 3.7 CTS DOCS X 3.7-46a 1:\ shared \its\rev_\rev_b\ attach 1.wpf 1 9712090087 971010 PDR ADOCK 05000454 P PDR , , ,

Attachment 1 improved Technical Specifications (ITS)

Affected Page List and Instructions SECTION/ TAB REMOVE REV. A (PAGE) INSERT REV. B (PAGE)  ;]7 3.7 LC0 MARKUPS 3.7-29 3.7-c9 3.7 LCO MARKUPS 3.7-31 3.7-31 3.7 LC0 JFDs 3.7-2 3. 7- 2 3.7 LCO JFDs 3.7-3 3.7-3 3.7 BASES MARKUPS INSERT B 3.7-61A INSERT B 3.7-61A 3.7 BASES MARKUPS B 3.7-65 B 3.7-65 3.7 BASES MARKUPS INSERT B 3.7-65A INSERT B 3.7-65A/B 3.7 BASES MARKUPS B 3.7-66 B 3.7-66 3.7 BASES MARKUPS B 3.7-70 B 3.7-70 3.7 BASES JFDs 3.7-2 3.7-2 3.7 BASES JFDs 3.7-3 3.7-3 3.7 NSHC X 3.7-66a 3.7-66b 1:\ shared \its\rev_\rev_b\ attach 1.wpf 2

I l

Attachment 1 Improved Technical Specifications (ITS)

Affected Page List and Instructions Enclosure 2: The ITS version of the proposed CTS LAR for " Boron Credit in the Spent Fuel Pool SECTION/ TAB REMOVE (PAGE) INSERT (PAGE) 1.0 BYRON CTS MARKUPS 1-2 1-2 1.0 BRWD CTS MARKUPS 1-2 1-2 1.0 CTS DOCS 1.0-3 1.0-3 1.0 CTS DOCS 1.0-7 1.0-7 __

3.7 BYRON ITS 3.7-37 3.7-37 3.7 BYRON ITS 3.7-38 3.7-38 gm 3.7 BYRON ITS 3.7-39 3.7-39

> 'd 3.7 BYRON ITS 3.7-40 3.7-40 3.7 BYRON ITS 3.7-41 3.7-41 3.7-41a 3.7-41b 3.7 BYRON ITS B 3.7-94 8 3.7-94 l 3.7 BYRON ITS B 3.7-95 B 3./-95 3.7 BYRON ITS B 3.7-96 8 3.7-96 3.7 BYRON ITS B 3.7-97 B 3 7-97 3.7 BYRON ITS B 3.7-98 8 3.7-98 3.7 BYRON ITS B 3.7-99 B 3.7-99 3.7 BYRON ITS B 3.7-100 B 3.7-100 3.7 BYRON ITS B 3.7-101 B 3.7-101 3.7 BYRON ITS B 3.7-102 B 3.7-102 3.7 BYRON ITS B 3.7-103 8 3.7-103 p

Cl 1:\ shared \its\rev_\rev_b\attachl.wpf 3

Attachment 1 Improved Technical Specifications (ITS)

Affected Page List and Instructions P

4 SECTION/ TAB REMOVE (PAGE) INSERT (PAGE)-

3.7 BYRON ITS B 3.7-104 8 3.7-104 3,7 BYRON ITS B 3.7-1N-' B 3.7-105 i' B 3.7-105a B 3.7-105b

B 3,7-105c

B 3.7-105d I B 3.7-105e

3.7 BRWD ITS 3.7-33 3.7-33 l L.7 BRWD ITS 3.7-34 3.7-34 3.7 BRWD ITS 3.7-35 3.7-35 l 3.7 BRWD ITS 3.7-36 3.7-36 3.7 BRWD ITS 3.7-37 3.7 i 3.7-37a 3.7-37b.

1 3.7 BRWD ITS B 3.7-85 B 3.7-85 3.7 BRWD ITS B 3.7-86 B 3.7-86

, 3.7 BRWD ITS B 3.7-87 8 3.7-87 4, _3.7 BRWD ITS B 3.7-88 8 3.7-88

) 3.7 BRWD ITS B 3.7-89 B'3.7-89

!- 3.7 BRWD ITS B 3.7-90 B 3.7-90 3.7 BRWD ITS B 3.7-91 B 3.7-91 l 3.7 BRWD ITS B 3.7-92 B 3.7-92 3.7 BRWD-ITS B 3.7-93 B 3.7-93

[ 3.7 BRWD ITS B 3.7-94 B 3.7-94 3.7 BRWD ITS B 3.7-95 B 3.7-95 i

! 1:\ shared \its\rev_\rev_b\ attach 1.wpf 4 1

n Attachment 1 V

Improved Technical Specifications (ITS)

Affected Page List and Instructions SECTION/ TAB REMOVE (PAGE) INSERT (PAGE) 3.7 BRWD ITS B 3.7-96 B 3.7-96 8 3.7-96a B 3.7-96b B 3.7-96c B 3.7-96d B 3.7-96e 3.7 BYRON CTS MARKUPS 3/4 9-13 3/4 9-13 (LCO 3.7.14)

INSERT 3.9-13A/B 3/4 9-13 (LCO 3.7.15) 3.7 BYRON CTS MARKUPS X 5-5 3.7 BYRON CTS MARKUPS X INSERT 5-5B 3.7 BYRON CTS MARKUPS X INSERT 5-5A (2 PAGES)

(]

'd 3.7 BYRON CTS MARKUPS X 5-Sa 3.7 BYRON CTS MARKUPS X 5-Sb 3.7 BYRON CTS MARKUPS X CTS INSERT C-1 3.7 BYRON CTS MARKUPS X INSERT 5-5bA (Figure 3.7.16-1) 3.7 BYRON CTS MARKUPS X CTS INSERT C-2 3.7 BYRON CTS MARKUPS X INSERT 5-5bA (Figure 3.7.16-2) 3.7 BYRON CTS MARKUPS X CTS INSERT C-3 3.7 BYRON CTS MARKUPS X INSERT 5-5bA (Figure 3.7.16-3) 3.7 BRWD CTS MARKUPS 3/4 9-13 3/4 9-13 (LCO 3.7.14)

INSERT 3.9-13A/B 3/49-13(LCO3.7.151 3.7 BRWD CTS MARKUPS X 5-5 3.7 BRWD CTS MARKUPS X INSERT 5-5B 3.7 BRWD CTS MARKUPS X INSERT 5-5A (2 PA7S) 3.7 BRWD CTS MARKUPS X 5-Sa O

O 1:\ shared \its\rev_\rev_b\ attach 1.wpf 5 f

Attachment 1 (q>

Improved Technical Specifications (ITS)

Affected Page List and Instructions l

SECTION/ TAB REMOVE (PAGEi INSERT (PAGE) .

l 3.7 BRWD CTS MARKUPS X 5-5b 3.7 BRWD CTS MARKUPS X CTS INSERT C-1 3.7 BRWD CTS MARKUPS X INSERT 5-5bA (Figure 3.7.16-1) 3.7 BRWD CTS MARKUPS X CTS INSERT C-2 3.7 BRWD CTS MARKUPS X INSERT 5-5bA (Figure l

3.7.16-2) 3.7 BRWD CTS MARKUPS X CTS I~ ERT C-3 3.7 BRWD CTS MARKUPS X INSERT 5-5bA (Figure 3.7.16-3)

/O. 3.7 CTS DOCS 3.7-7 3.7-7 3.7 CTS DOCS X 3.7-12a 3.7 CTS DOCS 3.7-17 3.7-17 3.7-17a 3.7 CTS DOCS X 3.7-32a 3.7 CTS DOCS X 3.7-46a 3.7-46b 3.7 LC0 MARKUPS 3.7-35 3.7-35 3.7 LCO MARKUPS 3.7-36 3.7-36 3.7 LC0 MARKUPS 3.7-37 3.7-37 3.7 LCO MARKUPS 3.7-38 3.7-38 INSERT 3.7-38A INSERT 3.7-38B 3.7 LCO MARKUPS 3.7-39 3.7-39 INSERT 3.7-39A (3 PAGES) 3.7 LC0 JFDs 3.7-6 3.7-6 OV 1:\ shared \its\rev_\rev_b\ attach 1.wpf 6

7- _

Attachment 1 t >

i-s '

Improved Technical Specifications (ITS)

Affected Page List and Instructions SECTION/ TAB REMOVE (PAGE) INSERT (PAGE) 3.7 LC0 JFDs 3.7-9 3.7-9 3.7-9a 3.7-9b 3.7 LCO JFDs 3.7-11 3.7-11 3.7-11a 3.7 BASES MARKUPS B 3.7-78 B 3.7-78 3.7 BASES MARKUPS B 3.7-79 B 3.7-79 3.7 BASES MARKUPS B 3.7-80 B 3.7-80 3.7 BASES MARKUPS B 3.7-81 E 3.7-81 3.7 BASES MARKUPS B 3.7-82 B 3.7-82

() 3.7 BASES MARKUPS 3.7 BASES MARKUPS B 3.7-83 B 3.7-84 B 3.7-83 8 3.7-84 3.7 BASES MARKUPS B 3.7-84 INSERT B 3.7-81A (7 PAGES) 3.7 BASES MARKUPS B 3.7-85 B 3.7-85 3.7 BASES MARKUPS B 3.7-86 B 3.7-86 3.7 BASES MARKUPS B 3.7-87 B 3.7-87 3.7 BASES MARKUPS B 3.7-87 INSERT B 3.7-85A (7 PAGES) 3.7 BASES JFDs 3.7-7 3.7-7 3.7-7a 3.7 BASES JFDs 3.7-10 3.7-10 3.7 BASES JFDs 3.7-11 3.7-11 3.7-11a 3.7 BASES JFDs 3.7-14 3.7-14 3.7-14a

/)

\- / 1:\ shared \its\rev_\rev_b\ attach 1.wpf 7

Attachment 1 improved Technical Specifications (ITS)

Affected Page List and Instructions SECTION/ TAB REMOVE (PAGE) INSERT (PAGE) l 3.7 NSHC X 3.7-66c 3.7-66d 3.7-66e 3.7-66f 3.9 BYRON CTS MARKUPS 3/4 9-13 3/4 9-13 3.9 BRWD CTS MARKUPS 3/4 9-13 3/4 9-13 4.0 BYRON ITS 4.0-2 4.0-2 4.0 BRWD ITS 4.0-2 4.0-2 4.0 BYRON CTS MARKUPS 5-4 5-4 4,0 BYRON CTS MARkdPS 5-5 5-5 INSERT 5-5A INSERT 5-5A 4.0 BYRON CTS MARKUPS 5-5a 5-Sa 4.0 BYRON CTS MARKUPS 5-5b 5-5b 4.0 BYRON CTS MARKUPS X CTS INSERT C-1 4.0 BYRON CTS MARKUPS X CTS INSERT C-2 4.0 BYRON CTS MARKUPS X CTS INSERT C-3 4.0 BRWD CTS MARKUPS 5-4 5-4 4.0 BRWD CTS MARKUPS 5-5 5-5 INSERT 5-5A INSERT 5-5A 4.0 BRWD CTS MARKUPS 5-Sa 5-Sa 4.0 BRWD CTS MARKUPS 5-5b 5-5b 4.0 BRWD CTS MARKUPS X CTS INSERT C-1 4.0 BRWD CTS MARKUPS X CTS INSERT C-2 4.0 BRWD CTS MARKUPS X CTS INSERT C-3 4.0 CTS DOCS 4.0-1 4.0-1 4.0 CTS DOCS 4.0-2 4.0-2 1:\ shared \its\rev_\rev_b\ attach 1.wpf 8

Attachment 1 Improved Technical Specifications (ITS)

Affected Page List and Instructions SECTION/ TAB REMOVE (PAGE) INSERT (PAGE) 4.0 LC0 MARKUPS 4.0-1 4.0-1 4.0 LCO MARKUPS 4.0-2 4.0-2 INSERT 4.0-2A 4.0 LCO JFDs 4.0-1 4.0-1 >

4.0 LCO JFDs 4.0-2 4.0-2 5.0 BYRON CTS MARKUPS 5-Sa 5-Sa 5.0 BYRON CTS MARKUPS 6-23 6-23 j INSERT 6-23A INSERT 6-23A 5.0 BRWD CTS MARKUPS 5-Sa 5-Sa 5.0 BRWD CTS MARKUPS 6-23 6-23 INSERT 6-23A INSERT 6-23A 5.0 CTS DOCS 5.0-3 5.0-3 1:\ shared \its\rev_\rev_b\ attach 1.wpf 9

ENC _0SUlt 1 Section 3.7

- - - - - , - - - - - - - , . - - - - - - - , - - - - - - - , , - - - - - - ---,,,----,--------,w_--- -w .

4 BYRON I S l

ce

i Nonaccessible Area Exhaust Filter Plenum Ventilation System  !

3.7.12 )

SURVEILLANCE RE00IREMENTS SURVEILLANCE FREQUENCY SR 3.7.12.1 Operate each Nonaccessible Area Exhaust 31 days Filter Plenum Ventilation System train fer

= 15 minutes.

SR 3.7.12.2 Perform required Nonaccessib'le Area Exhaust in accordance Filter Plenum Ventilation System filter with the VFTP testing in accordance with the Ventilation Filter Testing Program (VFTP).

SR 3.7.12.3 Verify each Nonaccessible Area Exhaust 18 months Filter Plenun Ventilation System train actuates on a manual, an actual, or a simulated actuation signal.

SR 3.7.12.4 Verify two Nonaccessible Area Exhaust 18 months on a Filter Plenum Ventilation System trains can STAGGERED TEST maintain a pressure s -0.25 inches water BASIS O gauge relative to atmospheric pressure

> during the emergency mode of operation at a El flow rate of s 68.200 cfm per train.

BYRON - UNITS 1. & ? 3.7-32 10/10/97 Revision B

FHB Ventilation System 3.7.13 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.7.13.2 Perform required FHB Ventilation System in accordance filter testing in ?ccordance with tb? with the VFTc Ventilation Filter Testing Program (VFTP).

SR 3.7.13.3 ----

NOTE- - -

Only required during movement of irradiated fuel assemblies or CORE ALTERATIONS with the equipment hatch not intact.

Verify one FHB Ventilation System train can 7 days on a mcintain a pressure s -0.25 inches water STAGGERED TEST gauge relative to atmospheric pressure BASIS during the emergency mode of operation.

SR 3.7.13.4 Verify each FHB Ventilation System train 1B months

~ actuates on an actual or simulated actuation signal.

SR 3.7.13.5 --

-NOTE -

Only required during movcinent of irradiated fuel assemblies in the fuel handling building with the equipment hatch intact.

Verify one FHB Ventilation System train can 18 months on a maintain a pressure s -0.25 inches water STAGGERED TEST q) gauge relative to atmospheric pressure BASIS 3 during the emergency mode of operation at a yl flow rate s 23.100 cfm.

e BYRON - UNITS 1 & 2 3.7-36 10/10/97 Revision B

Nonaccessible' Area Exhaust Filter Plenum ventilation System B 3.7.12 B 3.7- PLANT, SYSTEMS:

B 3.7.12 Nonaccessible Area Exhaust Filter Plenum Ventilation System BASES-BACKGROUND.

The Nonaccessible Area Exhaust Filter Plenum Ventilation l System filters air from the area'of the active Emergency -

Core Cooling System (ECCS) comaonents during the recirculation )hase of a Loss Of Coolant Accident-(LOCA). ,

The Nonaccessi)le Area-Exhaust-Filter Plenum Ventilation 1

-System. in conjunction with other normally operating systems, also provides environmental control of temperature in the ECCS pump room area and the lower reaches of-the

-auxiliary building.

The Nonaccessible Area Exhaust Filter Plenum Ventilation System is a subsystem of the common auxiliary building g heating. ventilation and air conditioning system-(VA). Each unit has two VA supply and two VA exhaust fans. The VA

[B

' . supply and-exhaust fans-are not required for Nonaccessible Area Exhaust Filter Plenum-Ventilation System OPERABILITY.

The Nonaccessible Area Exhaust Filter Plenum Ventilation' System consists of three 50% trains. Each train consists of :1 prefilters. High Efficiency Particulate Air (HEPA) filters, activated charcoal adsorber sections for removal of gaseous-activity (principally iodines).- and two~100% capacity fans.

Ductwork dampers, and instrumentation also form part of the system. A seccnd bank of HEPA filters follows the'adsorber.

sections to collect carbon fines and provide backup-in case -;

the main HEPA filter bank fails. The prefilters remove any; _

large particles in the air to prevent excessive -loading of the HEPA-filters and charcoal adsorbers. Each fan.in a train is powered from a different ESF bus. Train A fans are powered by Unit.1 buses-131 and 132: train B fans are powered by Unit 2 buses"231 and 232: and train C fans are powered by Unit--I bus 132-and Unit 2 bus 231.

(continued)

-BYRON - UNITS 1 &-2 B 3.7-77 10/10/97 Revision B

Nonaccessible Area Exhaust Filter Plenum Ventilation System B 3.7.12 BASES BACKGROUND The system is normally aligned with two inlet dampers open (continued) and the third train's inlet damper closed. The air passes through the HEPA filters and is routed to the auxiliary builaing exhaust plenum. The system initiates following receipt of a Safety injection (SI) signal from either unit.

O During the emergency mode operation, the auxiliary building normal supply and exhaust- fans associated with the unit D- generating the SI signal dre tripped (if o)erating and there e is a concurrent loss of offsite power to tlat unit), the Nonaccessible Area Exhaust Filter Plenum Ventilation System dampers realign, and a fan in each train with an open inlet damper starts to begin filtration. Interlocks are provided; to start the second hn after a time delay in a train if the first fan does not stert: to prevent start of a fan in a train with a closed it.let damper; and to prevent start of a fan with a closed discharge damper, The train with the closed inlet damper can be realigned manually from the control room, if required. The Nonaccessible Area Exhaust Filter Plenum Ventilation System emergency mode of operation can also be initiated manually by starting a fan in each train that is aligned for operation. A manual fan start signal will realign the associated dampers to begin i

filtration.

!~

The Nonaccessible Area Exhaust Filter Plenutt Ventilation System is discussed in the UFSAR, Sections 6.5.1. 9.4.5, ,

and 15.6.5 (Refs. 1. 2, and 3. respectively)

(continued)

BYRON - UNITS 1 & 2 B 3.7-7B 10/10/97 Revision B

_ =

Nonaccessible Area Exhaust Filter Plenum V:ntilation System B 3.7.12 BASES SURVEILLANCE SR .3.7.12.3 REQUIREMENTS (continued) This SR verifies that each Nonaccessible Area Exhaust Filter Plenum Vertilation System train aligns, starts, and operates on a manual, an actual, or a simulated actuation signal.

The 18 month Frequency is consistent with that specified in Reference 6.

SR 3.7 12.4 ,

This SR verifies the integrity of the ECCS pump room arecs.

The ability of the ECCS nump room areas to maintain a negative pressure, with iespect to potentially uncontaminated adjacent areas. is periodically tested to verify aroper functioning of the Nonaccessible Area Exhaust Filter )lenum Ventilation System. During the emergency mode g of operation, the Nonaccessible Area Exhaust Filter Plenum

> Ventilation System is designed to maintain a slight negative el pressure in the ECCS pump rooms, with respect to adjacent areas. to prevent unfiltered LEAKAGE, The Nonaccessible Area Exhaust Filter Plenum Ventilation System is designed to maintain a s -0.25 inches water gauge relative to atmospheric pressure with two trains operating, each at a flow rate s 68.200 cubic feet per minute (cfm).

t Nonaccessible Area Exhaust Filter Plenum Ventilatinn System function must be maintained ccasidering the design basis scenarios of an SI signal only on one unit or an SI signal concurrent with a loss of offsite power to a unit. This SR U should be performed with the postulated number of VA supply and exhaust fans running considering the SI signal only t scencrio. Performance of the SR in this manner produces the tc least negative pressure in the ECCS pump room areas (i.e.,

the least margin to 5 -0.25 inches water gauge). The Frequency of 18 months is consistent with the guidance.

provided in NUREG-0800. Section 6.5.1 (Ref. 7).

The testing of two of the three trains on an 18 month e Frequency on a STAGGERED TEST BASIS. requires that the

> combination of trains be varied, such that all 30ssible

@ combinations of trains be tested over a 54 monti period.

(continued)

BYRON - UNITS 1 & 2 B 3.7-84 10/10/97 Re.

FHB-Ventilation System B:3.7.13 B 3.7- PLANT. SYSTEMS B 3.7.13 Fuel Handling Building-Exhaust Filter Plenum (FHB) Ventilation-System BASES BACKGROUND The FHB Ventilation System filters airborne radioactive aarticulates from the area of the fuel pool following a fuel landling accident. The FHB Ventilation System, in conjunction with other normally operating systems, also_

provides environmental control of temperature-in the fuel pool area.

The FHB Vantilation System is a subsystem of the comon 0) auxiliary building heating. ventilation and air.

l conditioning system (VA). Each unit has two VA supply and

.?

e two VA-exhaust fans. The VA' supply and exhaust fans are'not l required for FHB Ventilation System OPERABILITY.

l The FHB Ventilation System consists of two inaependent and redundant trains. . Each train consists -of a prefilter, a i High Efficiency Particulate Air-(HEPA) filter,. an activated charcoal.adsorber section for removal.of gaseous activity (principally iodinesh and a fan, Ductwork, valves or dampers, and instrumentation also form part of the system.

A second bank of HEPA filters follows-the adsorber section=

to collect carbon-fines and provide backup in case the main HEPA filter bank fails. The downstream HEPA filter is not credited in the analysis, but serves to collect charcoal L

fines, and.to back up the upstream HEPA filter should it develop'a leak. The system initiates- filtered ventilation of the fuel handling building following receipt of a high radiation signal or a Safety Injection (SI) on either unit.

The FHB Ventilation System start on an SI signal is not

' credited in any accident analysis.

1 (continued)

BYRON - UNITS 1 &.2 B 3.7-B6 10/10/97 Revision B

I FHB Ventilation System B 3.7.13 BASES SURVEILLANCE SR 3.7.13.5 -

REQUIREMENTS (continued) This SR verifies the integrity of the fuel handling building enclosure. The ability of the fuel handling building to maintain negative pressure with respect to potentially uncontaminated adjacent areas is periodically tested to verify proper function of the FHB Ventilation System.

During the emergency mode of operation the FHB Ventilation System 1s designed to maintain a slight negative pressure in g the fuel handling building, to prevent unfiltered LEAKAGE.

The FHB Ventilation System is designed to maintain a t s -0.25 inches water gauge with respect to atmospheric tr l 3ressure at a flow rate s 23.100 cfm to the fuel handling auilding. The Frequency of 18 months is consistent with the guidance provided in NUREG-0800. Section 6.5.1 (Ref. 7),

An 18 month Frequency (on a STAGGERED TEST BASIS) is consistent with Reference 6.

This SR is edified by a Note that requires this SR only during movement of irradiated fuel assemblies in the fuel handling building when the equipment hatch is intact.

REFERENCES 1. UFSAR. Section 6.5.1.

2. UFSAR Section 9.4.5.

3. UFSAR. Section 15.7.4.

4. Regulatory Guide 1.25.

5. 10 CFR 100,

6. Regulatory Guide 1.52 (Rev. 2).

7. NUREG-0800. Section 6.5.1. Rev. 2. July 1981.

. +

BYRON - UNITS 1 & 2 B 3.7-93 10/10/97 Revision B

BRAI) WOO) ITS 9

Nonaccessible Area Exhaust Filter Plenum _ Ventilation System 3.7.12 SURVEILLANCE-REQUIREMENTS SURVEILLANCE FREQUr.NCY SR 3.7.12.1 Operate each Nonaccessible Area Exhaust 31 days F

Filter Plenum Ventilation System train for

= 15 minutes.

SR 3.7.12.2 Perform required Nonaccessible Area Exhaust In accordance 4

Filter Plenum Ventilation System filter with the VFTP testing in accordance with the Ventilation Filter Testing Program (VFTP).

SR 3.7.12.3 Verify each Nonaccessible Area lxhaust 18 months Filter Plenum Ventilation Systen. train actuates on a manual, an actual, or a simulated actuation signal.

. SR .3.7.12.4 Verify two Nonaccessible Area Exhaust 18 months.on a Filter Plenum Ventilation System trains can -

STAGGERED TEST maintain a pressure 5 -0.25 inches water BASIS O gauge relative to atmospheric pressure

?. during the emergency mode of operation at a ccl flow rate of s 73.590 cfm per train, f

n i

1. ,

J BRAIDWOOD - UNITS 1 & 2 3.7-28 10/10/97 Revision B

. . - . . - . . ~. . . . - . - - - . . . - . - - . - . - - . . - . . _ . - - - - - . - ~ . - . . . ,

L FHB Ventilation System-j- -3.7.13

[ L SURVEILLANCE' REQUIREMENTS (continued)-

[ SURVEILLANCE .FREQUENCi E

i- - --

L SR 3.7.13,2 Perform required FHB Ventilation System In accordance i filter testing in accordance with the with the VFTP-i_ . Ventilation Filter Testing Program (VFTP).

. SR 3.7.13.3 --- -

NOTE -

2-F Only required during movement of-irradiated

~ fuel assemblies or CORE' ALTERATIONS with l + the equipment hatch not intact.

l l'

, Verify one FH'B Ventilation System train can 7 days on a o maintain a pressure 5 -0.25 inches water STAGGERED TEST-

. -gauge relative to atmospheric pressure BASIS j during the emergency mode of operation.

2

SR 3.7.13.4 Verify each FHB Ventilation System train' 18 months-

'.i' 2

actuates on an actual or simulated-actuation signal.

i-

-SR 3.7.13.5_ --- -

- _ NOTE . .

i- Only-required during movement of irradiated fuel assemblies in the fuel handling building with the equipment hatch _ intact.

s L

Verify one FHB Ventilation System train cai, l'8 months on a F maintain a pressure s -0.25 inches water STAGGERED TEST

'_ n) ' gauge relative-to atmospheric ~ pressure BASIS

> during the emergency mode of operation at a el flow rate s 23.100 cfm.

,y BRAIDWOOD - UNITS 1 & 2 3.7-32 10/10/97 Revision B

Nonaccessible Area Exhaust Filter Plenum Ventilation System B 3.7.12 B 3.7 PLANT SYSTEMS l

B 3.7.12 Nonaccessible Area Exhaust Filter Plenum Ventilation System BASES BACKGROUND Th? Nonaccessible Area Exhaust Filter Plenum Ventilation System filters air from the area of the active Emergency Core Cooling System (ECCS) components during the recirculation phase of a Loss Of Coolant Accident (LOCA).

The Nonacceulble Area Exhaust Filter Plenum Ventilation System, in conjunction with other normally operating systems. also provides environmental control of temperature in the ECCS pump room area and the lower reaches of the auxiliary building.

The Nonaccessible Area Exhaust Filter Plenum Ventilation System is a subsystem of the common auxiliary building heating, ventilation 6nd air conditioning system (VA). Each ho unit has two VA supnly and tm VA exhaust fans. The VA M supply and exhaust fans are not required for Nonaccessible Area Exhaust Filter Plenum Ventilation System OPERABILITY.

The Nonaccessible Area Exhaust Filter Plenum Ventilation System consists of three 50% trains. Each train consists of prefilters. High Efficiency Particulate Air (HEPA) filters, at.Livated charcoal adsorber sections for removal of gaseous activity (principally iodines), and two 100% capacity fans.

Ductwork, dampers, and instrumentation also form part of the system. A second bank- of HEPA filters follows the adsorber sections to collect carbon fines and prov1<1e backup in case the main HEPA filter bank fails. The prefilters remove any large ) articles in the air to prevent excessive loading of the HEDA filters and charcoal adsorbers. Each fan in a train is powered from a different ESF bus. Train A fans are powered by Unit 1 buses 131 and 132: train B fans are powered by Unit 2 buses 231 and 232: and train C fans are powered by Unit 1 bus 132 and Unit 2 bus 231.

t (continued)

BRAIDWOOD - UNITS 1 & 2 B 3.7-68 10/10/97 Revision B

Nonaccessible Area Exhaust Filter Plenum Ventilation System B 3.7.12 BASES BACKGROUND The system is normally aligned with two inlet dampers open (continued) and the third train's inlet damper closed. The air passes through the HEPA filters and is routed to the auxiliary building exhaust plenum. The system initiates following receipt of a Safety Injection (SI) signal from either unit, O Ouring the emergency mode operation, the auxiliary building

> normal supply and exhaust faas associated with the unit

.E generating the SI signal are tripped (if o>erating and there is a concurrent loss of offsite power to t1at unit). the Nonaccessible Area Exhaust Filter Plenum Ventilation System dampers realign, and a fan in each train with an open inlet damper starts to begin filtration. Interlocks are provided; i to start the second fan after a time delay in a train if the first fan does not start: to prevent start of a fan in a train with a closed inlet damper: and to prevent start of a faa with a closed discharge damper. The train with the closed inlet damper can be realigned manually from the control room, if required. The Nonaccessible Area Exhaust Filter Plenum Ventilation System emergency mode of operation i

L can also be initiated manually by starting a fan in each train that is aligned for operation. A manual fan start signal will realign the associated dampers to begin filtration.

The Nonaccessible Area Exhaust Filter Plenum Ventilation System is discussed in the UFSAR. Sections 6.5.1. 9.4.5.

and 15.6.5 (Refs. 1, 2, and 3 respectively).

(continued)

BRAlbWOOD-UNITS 1&2 B 3.7-69 10/10/97 Revision B l

r _ _ _ _ _ _ _

-Nonaccessible Area Exhaust Filter Plenum Ventilation System B 3.7.12

= BASES 4

-SURVEILLANCE SR 3.7.12.3- -

REQUIREMENTS (continued) This SR verifies that each Nonaccessible Area Exhaust Filter Plenum Ventilation System train aligns, starts, and operates on a manuC. an actual, or a simulated actuation signal.

The 18 month Frequency..is consistent with that specified ir Reference 6.

SR -3.7.12.4-This SR verifies the integrity of the ECCS pump room areas.

The ability of the ECCS pump room areas to maintain a negative pressure, with respect to potentially uncontaminated adjacent areas, is periodically tested-to-i verify proper functioning of the Nonaccessible Area Exhaust Filter Plenum Ventilation System. .During the emergency mode

In of operation, the Nonaccessible Area Exhaust Filter Plenum

> Ventilation System is designed to maintain.a slight negative

@l. pressure in the ECCS pump rooms, with respect to adjacent areas. to prevent unfiltered LEAKAGE. The Nonaccessible Area Exhaust Filter Plenum Ventilation' System is designed to maintain a 5 -0.25 inches water. gauge relative to atmospheric pressure with two trains operating, each at:a '

flow rate s 73.590 cubic feet per minute (cfm).

Nonaccessible Area Exhaust Filter Plenum Ventilation System-function must be maintained considering the desi_gn basis scenarios of an SI signal only on one unit or an SI signal concurrent with a loss of offs 1ta power to a unit. This SR

, e should be performed with the posculated number of VA supply

> and exhaust fans running considering the SI signal only tE scenario. Performance of the SR in this manner produces the least negative pressure in the ECCS pump room areas (i.e.,

the least margin to 5 -0.25 inches water gauge). _The Frequency of 18 months is consistent with the guidance provided in NUREG-0800. Section 6.5.1 (Ref. 7).

The testing of two of the three trains on an 18 month -

e Frequency on a STAGGERED TEST BASIS. requires that the a combination of trains be varied, such that all- Jossible -

3 combinations of trains be tested over a 54 monti period.

(continued)

BRAIDWOOD - UNITS 1 & 2 B 3.7-75 10/10/97 Revision B

. . . _ . . ~ - ._.__m._ .. _ _ ._ _ . ._ _ - . - - ..

FHB Ventilation System.

t B 3.7.13 x B;3.7 PLANT SYSTEMS B 3.7;13 Fuel Handling Building Exhaust Filter Plenum (FHB) Ventilation System- .

BASES-BACKGROUND:'

The FHB Ventilation System filters airborne radioactive 3 articulates from the area of the fuel pool following a- fuel landling accident. The FHB Ventilation System, in conjunction with other normally operating systems, also provides environmental control of temperature in the fuel pool area.

The FHB-Ventilation System is a subsystem of the common auxiliary building heating, ventilation. and air-(n conditioning system (VA). Each unit has two VA supply and two VA exhaust fans. The VA supply and exhaust fans are not-d required for FHB Ventilation System OPERABILITY.

The FHB Ventilation System consists of two independent and redundant trains. Eacl. train consists of a prefilte. , a High Efficiency Particulate Air (HEPA) filter, an activated charcoal adsorber section for removal of gaseous' activity-4 (principally iodines). 'and a fan. L Ductwork, valves'or dampers, and instrumentation also form ) art of the systemc

~

' A second bank of HEPA filters follows tle adsorber-section

-to collect carbon fines and provide backup in case the main HEPA filter bank fails. The downstream HEPA filter is not -

credited.in the analysis, but serves to collect-charcoal: ,

• -fines.:and to back up the upstream HEPA filter should:it develop a leak. The system initiates filtered ventilation of the fuel handling building following 'eceipt of a high radiation signal.or a Safety Injection (SI) on:either unit.

The FHB Ventilation System start on an SI signal is not credited-in any accident analysis.

(continued)

BRAIDWOOD - UNITS 1 & 2 8 3.7-77 10/10/97 Revision B

FHB Ventilation System B 3.7.10

BASES SURVEILLANCE SR 3.7.13.5 -

REQUIREMENTS (continued) This SR verifies the integrity of the fuel handling building enclosure. The ability of the fuel handling building to maintain negative pressure with respect to potentially uncontaminated adjacent areas is periodically tested to verify proper function of the FHB Ventil6 tion System.

During the emergency mode of operation the FHB ventilation System is designed to maintain a slight negative pressure in 1

the fuel handling building, to prevent unfiltered LEAKAGE.

g The FHB Ventilation System is designed to maintain a s -0.25 inches water gauge with respect to atmo.;pheric yl pressure at a flow rate s 23.100 cfm to the fuel handling i, building. The Frequency of 18 months is consistent with the guidance provided in NUREG-0800. Section 6.5.1 (Ref 7). ,

An 18 month Frequency (on a STAGGERED TEST BASIS) is consistent with Reference 6.

This SR is modified by a Note that requires this SR only during movement of irradiated fuel assemblies in the fuel handling building when the equipment hatch is intact.

REFERENCES 1. UFSAR, Section 6.5.1.

2. UFSAR. Section 9.4.5.

3. UFSAR. Section 15,7.4

4. Regulatory Guide 1.25. >

5. 10 CFR 100.

6. Regulatory Guicte 1.52 (Rev. 2).

7. NUREG-0800. Section 6.5.1 Rev. 2. July 1981.

4 i

BRAIDWOOD - UNITS 1 & 2 B 3.7-84 10/10/97 Revision B

D BYRON CTS MAR (UPS l

1 e

4

1 .

. . '-

• Y W 8 i EElllEfBW h 51 M ?l.f

3) Verifying a system flow rete between 55,669 cfm and 68,200 cfm through the exhaust filter plenum during operation when tested in accordance with ANSI N510-1980; and

4) Verifying that with the system operating at a flow rate between 55,669 cfm and 68,200 eft throug1 the train and exhausting through the HEPA filter and charcoal adsorbers, the total bypass flow of the system and the damper leaka e is less than or equal ll 1

to15whenthesystemistestedbyadmifting00Patthesystem intake and the osaper leakage rate is determined by either '

l direct measurements or pressure decay measurements at a test pressure of 2 inches of water and the auxiliary building exhaust.

fans are operating at their rated flow.

c. After every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of charcoal adsorber operation, by verifying, rithin 31 days sfter removal, that a laboratory analysis of a repre-sentative carbon sample obtained from each bank of adsorbers of the train in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52,-Revision 2. March 1978, meets the laboratory testing cri-teria of Regulatory Position C.6.a of Riqulatory Guide 1.52, Revision

2. March 1978, when the average for a e.ihyl iodide penetration of less than 15 when tested at a temperat re of 30*C and a relative

, humidity of 705.

i

d. At least onu per 18 months by:

1) Verifying for each filter bank of the train'that the pressure i drop across the combined HEPA filters and charcoal adsorber i

banks of less thar 6.0 inches Water Gauge while operating the ,

exhaust filter plenum at a flow rate between 55,66g cfm and 68,200 cfm through the train;

w h M n s e e,.< m n r i 2) Verifying than the _ _ _. ... c r m tarts on manual i M M# 3 initiation on hf;t ht:t';; t::t gpf : =f 1 I an sa..o oe s. a u s.1 a. 4._.. s.v. s. ;

g '"8 , 3) Verifying thatr_

n:t = maintains,the ECCS equipment rooms at l a negative pressure of greater than or equal to 1/4 in. Water

_ Sl1 3 '2 4 Gauge relative to the outside atmosphere during system operation C while operating at a flow ratelhtwoonmL46Hefs-*nd368,200 cfm L

through the trainh on 31.g,,..l b.nQ.-(MNSTE) l C:

,- up x -

l e . After each complete or partial replacement of a HEPA filter bank, by verifying that the exhaust filter plenum satisfies the in-place penetration testing acceptance criteria of less than 1% in accordance with ANSI N510-1980 for a DOP test aerosol while operating at a flow rate between 55,669 cfm and 68,200 cfm through the train; and. l l Manssed ;n' Sube 5.0 l Se e Dor.s (.r Sad.on S.o NAi BYRON - UNITS 1 & 2 3/4 7-20 AMENDMENT NO. 56 Rev.B L_ _ __,_ - _.

33,3;g -

Sodtok C.5 ll

_REr'UELIl4G OPERATIONS

.sg 3 7.83.5 DTE Cerf 7 9-d t SURVEILLANCE RE0VIREMENTS (Continued) '

[on on is ,. n+6 5746f Af D TEST B ASIS / pyg q,A,44,%  !

7

3) Verifying ^that the Tv.1-::.i.diir,; 0.iidir.; E d =:t P lt:r ." M r-5 8. 3.7.13. 5 maintains the fuel building at a negative pressure of greater than or equal to 1/4 inch water cauge relative to the outside atmosphere during operatioj,e+ afdw reti -

6

" Zen s e n e,4... .}

3 r

e. -

After each complete or partial replacement of a HEPA filter bank, by verifying that the Fuel Handling But1 ding Exhaust Filter Plenum satis-fies the in place penetration testing acceptance criteria of less than 1% in accordance with ANSI N510-1980 for a DOP test aerosol while operating the system at a flow rate of 21,000 cfm i 10%; and

f. ' After each complete or partial replacement of a charcoal adsorber bank, by verifying that the Fuel Handling Suilding Exhaust Filter Plenum satisfies the in place penetration testing acceptance criteria af less than 1% in accordance with ANSI N510-1980 for a halogenated hydrocarbon refrigerant test gas while operating the system at a flow

, rate of 21.000 cfm i 10%.

mm y , '

hdd rased in .S a cd.'a n 5.5. Il

.5 < < Oc c. (o, S e e f. o n 5.0 -

e e

BYRON - UNITS 1 & 2 3/4 9-16 Rev. E

mamm 4 X

a B hD CTS VAR (U3S i

l

l2,0 5,1, IX PLANT SYSTEMS SURVEILLANCE RE0UIREMENTS (Continued)

3) Verifying a system flow rate of 66,900 cfm ! 10% through the train and 22,300 cfm 110% per bank through the exhaust filter jg y ,

rsm. plenum during operation when tested in accordance with ANSI n g,3, N510-1980; and c,e coc r Ee6-i. r.o 4) Verifying that with the system operating at a flow rate of 66,900 cfm 110% through the train and 22,300 cfm 210% per bank and exhausting through the HEPA filter and charcoal adsorbars, the total bypass flow of the system and the damper leakage is less than or equal to 1% when the system is tested by admitting cold 00P at the system intake and the damper leakage rate is determined by either direct measurements or pressure decay measurements at a test pressure of 2 irches of water and the auxiliary building exhaust fans are operating at their rated flow.

-w After every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of charcoal adsorber operation, by verifying, within 31 days af ter removal, that a laboratory analysis of a repre-sentative carbon sample obtained from each bank of adsorbers of the train in accordance with Regulatory Position C.6.b of Regulatory

' Guide 1.52, Revision 2, March 1978, meets the laboratory testing cri-turis :f Ac;ul:tery P:siti:r. 0.6.4 et u.golatory Guide 1.52, Revisioni 2, March 1978, when the average for a methyl iodide penetration of l

~ less than 1% when tested at a temperature of 30*C and a relative

! humidity of 70%.

-A At least once per 18 months by: ,

1) Verifying for each filter bank of the train that the pressure drop across the combined HEPA filters and charcoal adsorber banks of less than 6.0 inches Water Gauge while operating the exhaust L filter plenum at a flow rate of 66,900 cfm + 10% through the

~

'~

train and 22,300 cfm 110% per bank; or> Mtut., or :malateu 4) A operation. The CTS words

• removed" are modified to "not intact." to dl l encompass the postulated scenarios (e.g., both air lock doors opened).

Ha -Not used.

Mn Consistent with NUREG-1431. LC0 3.7.3 "Feedwater (FW) Isolation Valves" and LC0 3.7.4. " Steam Generator (SG) Power Operated Relief Valves (PORVs)" are added to the ITS. Changes to CTS containment isolation valve (CIV) requirements are individually annotated and any technic 61 changes are justified separately.

CTS CIV Actions do not include, nor is it appropriate in the CIV TS, Required Actions for the condition of two feedwater valves -in the flow path inoperable, because the second containment isolation boundary is the secondary side piping. From the standpoint of-ITS LCO 3.7.3. a second isolation valve is required. Therefore. LC0 3.7.3 Condition B is g provided. These changes represent an additional restriction on plant

, operation.

m g Hu Consistent with NUREG-1431. ITS SR 3.7.13.S adds an upper flow rate t- limit to CTS SR 4.9.12.d.3. This SR verifies the ability of the FHB

. Ventilation System to maintain the enclosure at a negative pressure. If y the system were to run at a flow rate greater than design, the negative a pressure may be met, but the larger flow rate could be indicative of fl system, degradation. This represents an additional restriction on plant operation.

5H u 2

CTS S.6.1.1.e. g. and h describe conditions for new and spent fuel as well as interfacing requirements for their storage in various regions of d$

the spent fuel pool. These conditions have been relocated to ITS LCO 3.7.16.a. b, and c res)ectively. In order to make the ITS LC0 M complete. an introductory .C0 sentence. APPLICABILITY, ACTIONS and t{, Surveillance Requirements have been added. The additional Actions and Surveillance Requirements constitute a more restrictive change.

BYRON /BRAIDWOOD UNITS 1 & 2 3.7 17 10/10/97 Revision B

DISCUSSION OF CHANGES TO CTS ITS SECTION 3.7 PLANT SYSTEMS La CTS SR 4.7.7.d.3) requires that the Nonaccessible Area Exhaust Filter Plenum Ventilation System maintains, with two trains operating. the ECCS equipment rooms at < -0.25 inches with respect to atmosphere with each train operating within a flow rate band (between 55.669 cfm and 66.820 cfm B 66.900 cfm i 10% Braidwood). Consistent with NUREG-1431 SR 3.yron:7.12.4. the requirements associated with a lower flow rate are deleted from this surveillance. The design bases analyses (offsite and control room dose assessments) assume the ECCS leakage outside containment is instantaneously released from the volume containing the leakage source (i.e., no credit is taken for holdup) and that the release is filtered. As such, the system flow rate is not a parameter in the dose analyses. As described in the Bases for SR 3.7.12.4. the purpose of this surveillance is to verify the integrity of the ECCS pump y room areas. If the surveillance can be satisfactorily performed at a flow rate less than the stated flow rate, the intent of the surveillance I

4 is still aet. The VFTP Surveillances associated with the Nonaccessible Area Exhaust Filter Plenum Ventilation System (SR 3.7.12.2) test the y HEPA and charcoal filters in conjunction with individual trains. These o Surveillances include specific flow rate bands for the train. The te remaining LCO 3.7.12 SRs verify that the system will start when called upon.

L3 CTS Action 3.9.11.a states. * ... restore the water level to within its limit within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />." ITS 3.7.14. Required Action A.1 revises this to state. " Suspend movement of irradiated fuel assemblies in the spent fuel pool immediately." The ITS is less restrictive since it does not require the restoration of the water level in order to place the plant in a safe condition. The purpose of the ITS Specification is to ensure x that there is sufficient water level above the top of the stored fuel to

.2 maintain the initial conditions assumed in the design basis fuel

- handling accident. This accident assumes an irradiated fuel assemble is 5 dropped onto irradiated fuel assemblies seated in the storage racks, o resulting in ruptured fuel rods. Suspending fuel movement precludes 6 D

fucl handling accident from occurring. With the initiation event-for-f the design basis accident removed, restoration of the initial conditions

• is not required, This-change does not reduce the level of safety as described in the applicable Bases. This change is consistent with O NUREG-1431.

BYRON /BRAIDWOOD UNITS 1 & 2' 3.7 46a 10/10/97 Revision B

LCO VARKUPS t  ;

-guwfb /kan. E=LW Rl4*eRh vestikum 5%}_ g g,gg 3.7.12 SURVEILLANCE REOUIREMENTS (continued)

SURVEILLANCE FREQUENCY v

SR 3.7.12.2 Perform required filter testing In accordance in accordance with th tilation Filter Testing Program (VFTP) withthe7VFTPh 3.7.12.3 Verify each :::: Ti;/,;;trainactuateson kl8 months hSR . m.nv.I, an actual or simulated actuation signal. g, SR 3.7.12.4 Verify D ,0';; trai an maintain a months on

@e,ressuresh[-[0.$5 inches w er gauge a STAGGERED relative to atmosp eric pr=dsure during the TEST BASIS cim,,r.,,,g $ 2:?; .a'd:St4. modelf operation at a flow rate o p : r ? r_l ,

f ..

SR 3.7.12.5 Verify each EC '"*

P ;. 6.n ne closed.

-. u1 pass ~[18] months g 4

4 L

By ren : [5 5, 6 ' ^ O ' " 6 6' 8 0 C (#4

-J pre.,lvod' 1 L C,2 '^ @ 3 7 3,5 9 oc% g3 c.

WOG STS 3.7-29* Rev 1, 04/07/95 Psev. 8

t m - va m us.c m = p q '

ir

~SD 45-3.7.13 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME 1

Ch. Two trains k.1 Suspend movement of Immediately inoperable _Icurin / irradiated fuel t of tated assemblies in the P' fuel in the fuel building. '

uildino. .

hadliy

• CTnser4 3138NM1.A em -

SURVEILLAPCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.13.1 O,per..ateeachFB[6& train..for 31 days

. ___m_.._... <_..__.t .<_ t,..__.

.r.. ." . '. ," " '.v...

. .' I ' ' Z _.." :.',"_ ! _ : -'".. .J.". .' ". ' . . . .M. ._. _, s 215 minutes).

SR 3.7.13.2 Perfonn required T"' filter testing in .In accor(ince f ace rdance with thel [lYentilati n Filter withtheQVFTP OBl TestingProgram(VFTPQ. '

h Csg 3. 7. # s ofJ ,se,+ 3.71# in u , ,

-s SR3.7.13.), Verify each HMGG train actuates on an kl8, Jmonths h

actual or simulated actuation signal.

SR 3.7.13. QZns<e4 M.s}t, Verify on i" = 6ra n can maintain a

\l8!monthson Q B. pressure 5 -0.X25 nches water gauge w.ith a STAGGERED cc r.-7 epeepect to atmosph ric pressure during the TEST BASIS

'jart 20:16rt1 mode of operation at a flow -

ce,,,,,, ,,g %g rate ; [?O,l000[cfm.

- a

L n uest m

i; = c.

, g (continued) f WOG STS 3.7-31 Rev 1, 04/07/95 Rev. B

_ _ . , - - . , . - - y_ , - - . - -,

. .a

_ C0 JF)s s

s 4

1

JUSTIFICATION FOR DIFFERENCES TO NUREG 1431 LCOS SECTION 3.7 PLANT SYSTEMS Ba The Byron and Braidwood Control Room Ventilation (VC) Filtration System design contains two different filter un'its, one with heaters (makeup filter unit) and one without (recirculation filter unit). ITS SR 3.7.10.1 is written to reflect the appropriate requirements for how long the units must be run. The Byron and Braidwood fuel Handling Building Exhaust Filter Plenum (FHB) Ventilation System and Nonaccessible Area Exhaust Filter Plenum Ventilation System do not contain heaters, therefore the bracketed information in the SRs and information in the corresponding Bases regarding heater operation is deleted.

Bu CTS 4.7.6.e.5 requires verification of a 0.020 inch water gauge positive pressure for the upper cable spreading room, while CTS 4.7,6.e.3 requires verification of a 0.125 inch water gauge positive pressure for the rest of the control room enclosure. ITS 3.7.10 replaces the NUREG bracketed information with this CTS requirement.

Bu CTS SR 4.7.6.e.3 and SR 4.7.6 e.5 require verification of the VC tr Filtration System train's ability to maintain the area at a specified positive pressure. The allowed makeup flow rate is specified in terms of a range of flow rather than a maximum flow shown in NUREG SR 3.7.10.4

el u bracketed information. The CTS allowance is reflectM in ITS SR 3,7.10.4.

n The Byron and Braidwood Nonaccessible Area Exhaust Filter Plenum ml r

B Ventilation System design does not contain heaters, therefore the bracketed information in the SRs and information in the corresponding Bases regarding heater operation is deleted.

Ba The Nonaccessible Area Exhaust Filter Plenum Ventilation System operation is not referred to as normal and post accident mode of oaeration. Flow from the ECCS aump rooms is normally routed through the HEPA filters'and then through tie auxiliary building plenuni and exhaust fans.

Plant nomenclature is such that Reference to Nonaccessible Area Exhaust Filter Plenum Ventilation System implies the charcoal filter units and the charcoal booster fans. Therefnre the bracketed " post accident" has been changed to " emergency."

Ba The plant specific number of SG PORVs is provided. The analysis described of in UFSAR one inoperable Section 15.6.3 does not assume an initial condition SG PORV.

Ba Consistent with the CT3 ClV requirements, the Frequency is specified as in accordance with the IST Program.

BYRON /BRAIDWOOD UNITS 1 & 2 3.7 2 10/10/97 Revision B

_ .. _ . _ . _. _ - _ _ _ _ _ _ . _ _ _ _ _ _ _ _. _ _ .. _ _ _ _ _ _ .m j

1 JUSTIFICATION FOR DIFFERENCES TO NUREG 1431 LCOS  !

SECTION 3.7 PLANT SYSTEMS  !

Bn ITS LCO 3.7.13 bracketed information relative to FHB Ventilation System requirements during Modes 1. 2. 3. and<4 are deleted. CTS and design i bases analysis'de,not take credit for the FHB Ventilation System during  !

Modes 1. 2. 3 and 4. Correspondin '

editorially corrected as necessary.g Bases information is deleted and  !

Ba The FHB Ventilation System does not have a normal and post accident mode =

of operation >er se. Flow from the fuel handling building is normally  !

routed throug1 the HEPA filters and then through the auxiliary building

)lenum and exhaust fans. Plant nomenclature is such that reference to -

HB Ventilation System emergency mode of operation implies the charcoal  !

filter units and the FHB-Ventilation System exhaust' fans. Therefore the '

s bracketed " post accident" has been revised to " emergency " In addition-t the CTS requirements to perfom the "1/4 inch" test does not include a Il

" ficN rate. -Consistent with NUREG-1431 an upper flow rate has been added.

P Ba Consistent with LC0 3.0.3 the time to be in Mode 4 for ITS LCO 3.7.4 Condition C is-reflected as 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

B n (Byron Only) The Byron cooling tower fans do not receive any automatic signals. Therefore. "2 REG SR 3.7.9.4 is deleted. NUREG SR 3.7.9.3.

regarding periodic fan operation is modified to reflect that the Byron design requires only 6 of the 8 available fans to be Operable.

Bn. (Byron Only) Bracketed information with res)ect to LC0 3.7.9 Actions is ,

modified to reflect the Byron UHS oesign. T1e UHS is common to both -

units and has CTS makeup capability requirements.. In addition, the CTS allowed outage time for an inoperable required fan is 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> versus the NUREG allowed Completion Time of 7 days. See also Pn for a c discussion of the plant specific requirements. '

Bn The Generic Change described in 3C included bracketed information in the

' Action and fable 3.7.1-I with respect to appropriate Actions and power -

levels when the moderator temperature coefficient (HTC) is zero or negai.1ve in addition to when the MTC is positive. Byron and Braidwood is licensed for operation with a positive MTC. Throughout the cycle'it is not practical to determine if the MTC is positive. negative or zero.

Therefore, the information associated with zero or negative MTC is deleted.

Inaddition,thegenericchanges)ecifiesthe"High" Power Range Neutron Flux trip setpoint.-. The Hig1" is not included based on-the premise that reduction of the " Low" Power Range Neutron Flux Trip setpoint at lower powers will provide the same protection function.

i 3.73 BYRON /BRAIDWOOD UNITS 1 & 2~ 10/10/97 Revision B &

n l

E 6

A : .- .;. - - . - -- -' . - - - ~ . _ . - - - - - -.-_.-_,-,-_-.-.,,.------.n-

l BASES (ARKPS 6

BASES INSERT (S)

SECTION 3.7

, Bases 3.7.12 INSERT B 3.7 61A (Pn )

The Nonaccessible Area Exhaust Filter Plenum Ventilation System is a subsystem of the common auxiliary building heating, ventilation and air conditioning (D system (VA). Each unit has two VA supply and two VA exhaust fans. The VA

? supply and exhaust fans are not required for Nonaccessible Area Exhaust Filter ic Plenum Ventilation System OPERABILITY.

The Nonaccessible Area Exhaust Filter Plenum Ventilation System consists of three 50% trains. Each train consists of prefilters. High Efficiency Particulate Air (HEPA) filters, activated charcoal adsorber sections for removal of gaseous activity (principally iodines). and two 100% capacity fans.

Ductwork, dampers, and instrumentation also form part of the system. A second bank of HEPA filters follows the adsorber section to collect carbon fines and provide backup in case the main HEPA filter bank fails. The prefilters remove any large particles in the air to prevent excessive loading of the HEPA filters and charcoal adsorbers. Each fan in a train is powered from a different ESF bus. Train A fans are powered by Unit 1 buses 131 and 132:

train B fans are p0wered by Unit 2 buses 231 and 232: and train C fans are powered by Unit 1 bus 132 and Unit 2 bus 231.

The system is normally aligned with two inlet dampers open and the third train's inlet damper closed. The air passes through the HEPA filters and is routed to the auxiliary building exhaust plenum The system initiates following receipt of a Safety injectict (SI) signal from either unit. During g the emergency mode operation, the auxiliary building normal supply and exhaust

, fans associated with the unit generating the Si signal are tri p d (if

'lk operating and there is a concurrent loss of offsite power to t> . unit), the Nonaccessible Area Exhaust Filter Plenum Ventilation System dampers realign, and a fan in cach train with an open inlet damper starts to begin filtration.

Interlocks are provided: to start the second fan after a time delay in a train if the first fan does not start: to prevent start of a fan in a train with a closed inlet damper: and to prevent start of a fan with a closed discharge damper. The train with the closed inlet damper can be realigned manually from d, the control room. if required. The Nonaccessible Area Exhaust Filter Plenum 3 l Ventilation System emergency mode of operation can also be initiated manually by starting a fan in each train that is aligned for operation. A manual fan start signal will realign the associated dampers to begin filtration.

10/10/97 Revision B

( we ' - y CC: ** E?.C S

• LPlanum. V0MMim.44m Syettw j E .

81 7.17 O

SURVEII. LANCE SR 3.7.'12.4 REQUIREMENTS

(:entinued) 'This SR verifies the integrity of the ECCS pump roomferee s)

The ability of the ECCS pump room oa tain a negative pressure, with respect to potentiall a uncontaminated adjacent areas, is periodically tested to

.- verify proper functioning of the-E466-MEA 6SF puring the

,,,..c.f !;;:t 2:0 6-t' mode of operation, the C00: "":A ; is 4 ,

fh" #9 designedtomalntainaslightnegativepressureintheECCS

-!- <WMM 4wrax pumn room; with respect to adjacent areas, to prevent t<A'

• % M A*

• unf\iteredLEAKAGE. The :::: "":A;f is designed to maintain i

• a s K-0.t5)' inches water gauge relative to atmospheric

-e pressurew:t : #1-" r:t: Of '0000] :'- ' :: th: SCC! ;;;p l c saviiWS r.aam4 tThe Frecuency of '18 ' months is consistent with the

! Gee WW guidance provic ed 1 -0 Uf, m%. gns ,t a37-45

, Section 6.5.1 (Ref. g).

anrc+ 637-658 7 g g,yg

0--tr

t he; th::, :,, :18.3month _.. t"St: # r '"ter l

N: :^" '5 "@_ &,

.g,. g,, .w. Frequency on a STAGGERED g g, g ppf g TEST BASIS is consist ith that specified in Reference %g t w .m ~. -

{

i

@ .5 Operating the ECCS P er is necessary to ensure that the sys ions . ' The OPERABILITY of the EC bypass damper is veri '

can be in Reference 4. - -

REFERENCES 1. UFSAR, Sectio 6.5.1h h 2. UFSAR, Section 9.4.5[

3. (JSAR, Sectikl.E.6.5 G%. Regulatory Guide 1.52 (Rev. 2).

D.r, 10 CFR 100.11, uF5 A A ce t h 4.4 3Yy NL' REG-0800, Section 3.5.1, Rev. 2, July 1981.

O WOG STS B 3.7-65 Rev 1, 04/07/95 Rev. B W

BASES INSERT (S)

SECTION 3.7

, Bases 3.7.12 INSERT B 3.7 65A (P3 and P3 )

.I The testing of two of the three trains on an 18 month Frequency on a STAGGEREO

-e t TEST BASIS requires that the combination of trains be varied, such that all possible combinations of trains be tested over a 54 month period...

If a particular pump room is isolated such that there is no potential for post-

. accident fluids to pass.through the room, or that room's ECCS equipment is not required, that room can be excluded from meeting the acceptance criteria of the SR. Performance of this SR with a room excluded. represents a change in the ECCS-pump room area volume that the system is maintaining at a negative

)ressure. Prior to the room bein

>e performed with the new volume,g put back to assure in service that-the systemthis SRmaintain can would have the to entire volume at the reauired negative pressure.

INSERT B 3.7 65B (Pn )

Nonaccessible Area Exhaust Filter Plenum Ventilation System function must be maintained considering the design basis scenarios of an SI signal only on one e unit or an 51 signal concurrent with a-loss of offsite power to a unit. This D

SR should be performed with the postulated number of VA supply and exhaust E fans running considering the S1 cignal only scenario. Performance of the SR in this manner produces the least negative pressure in the ECCS pump room areas (i.e.. the.least margin to s -0.25 inches water gauge).

10/10/97 Revision B

( i .e

rrro v evn w -n my*-)

.g,A;&.

R 'l 7 l 't

@ "'"'88 A 'r * ** %eare VA ot.d t.shaact n e cupIutlted te 4r rae fai"*

B 3.7 PLANT SYSTEMS

• "'# '*% #"" NM W-B 3.7.13 . . . . . . . . . , . . . . . . . . , . , .. . . . .

f hhe FH Ve*Hvs ittWm 6..thsam . .. . . a . *..* .s

• 6,u87.n.Hem eMe Seeen era,ii.,y b vos.,3 6 , . + ,..

• * '" **d" * #^ ""* ") p' "'..4,J.vA) 4' . E. d u n4 h. , he d, BASES BACXGROUND TheN filters airborne radioactive particulates from th area of the fuel pool following a fuel handling accidentpr 1

r :sc;ent ;;; me.i ( = r .; ine 08"I10:: with other normally operating syJtems/=also provides

.:,, in conjunction - .

environmental control of temperatur :=-i h; iilyi in the

, fuel pool area.

The'FBAC-5 consists of two independent and redundant trains.

Each train consists of a a prefilterik d:9 tan a hig.. efficiency particula )(HEPA) r filter,anactivated oc c sr,9 charcoal adsorber section for removal of gaseous activity Lp (42% (principally iodines), and a fan. Ductwork, valves or e.i @ c- w .t, % c. ._ dampers, and instrumentation also forn part of the system, Fily S tW een '

u:11 ::; dec.ister;, Nneuemn; u rece : ricm ed v a i e f.A clari en Agg.idity of the ? nt, amm- 1 A second bank of Mt.PA filters 4

u 4:c *ml b follows the adsorber section to collect carbon fines and wt c'8AM. 4 provide backup in case the main HEPA filter bank fails. The by WL'* downstream HEPA filter is not credited in the analysis, but

• *ly*t*,

serves to collect charcoal fines, and to back up the upstream HEPA filter should it develop a leak. The system initiates filtered ventilation of the fuel handling building following receipt of a high radiation signalf 4

_The TOA : is a standby system.jparts of which may a We% normal p4anI'"5perations. ceipt of Jase< f) the actuating sigTah-sc a arges from the 8 L' 8 R building, the fuel h '

ui ated, and the stream of ion ai/ discharges through to

f M rains.1 The orefiltersi:- d ri n e : remove any large particles in the air.__:nd :ny :ntr&ed etter- trie n !

peh]arsoal ano cn adsorbers.to prevent excessive loading of the HEPA filters 8.

[ The* "' is discusse -in th l SAR, Sections ([6.5.1 ,

9.4 andN15.7.4 (Refs.1. 2. and 3. respectiv lvL G' '

be::::: it m:y be a:Od f r r.amal, :: well :: por,t :::ide % .

3

tm::;heri: cle:nu? functi:n:.

4 (continued)

WOG STS B 3.7-66 Rev 1, 04/07/95 i

Rev. 8 I r .

5 3.7.13 sAsEs SURVEILLANCE (, SR 3.7.13.2 (continued) .

gener.# c.

N ,.r .eeJ REQUIREMENTS A - o 7 Program (VFTP)).- The fimit filter tests e in :.jd  ;;;

withRegulatoryGuide1.52(Ref.6). The VFTP[: includes testing HEPA filter performance, charcoal dsorber i

h eYficiency,Msystem properties of the activated flow and the-physical ratg(general use and charcoal following specific operations (andaddhionalinformationare). discussed Specific test frequencies'.

in detail in the '

TVFTPJ~. ** **~'s~' c.-

neer+ 6 M-7csX$8 Lill D .

3.7.13.Kd l SR L

This 3R verifies that eachh traihs* tart op tes l on an actual or simulated actuation signal. e 18 month

_ Frequency is consistent with Reference 6. _

SR 3.7.13.'t,5 b b[hadhM) . -

This SR verifies the integrity of thfuelPbuilding enclosure. The ability of the fuelfFullding to maintain-negative pressure with respect to potentially uncontaminated adjacent areas is periodically tested to verify-proper function of the#9466. During the {F eeritt' mode of' ,* Ny operationgtheN is designed to maintain a sl ght I negative fressure in the fuel building, to prevent Iba id ynfilteredj.;,AKAGE. The7 0",00 T3 casigned to maintain

" C D ' M -0.T,25 finches water gauge with respect to atmospheric m -

AWOO ; orensure bui' ding. at TheaFrequency flow rate.;f of '"0,["^^] cfm 18 months is to the. fuelwith consistent had.tig the guidance provided in NUREG-0800, Section 6.5.1=(Ref. 7). f-An\l8 month Frequency (on a STAGGERED TEST BASIS) is consistent with Reference 6.

3r (Insec+ 3 3 a.3rGib h 13.5 Operating the FBAC .

b pass s necessary to ensure that the system f erly. The OPERABILITY _

3 of-the FBACS fil ass damper is if it can be closed. month Frequency-is consisten '

nce 6.

(continued)- .

WOG-STS B 3.7-70 Rev 1, 04/07/95 ReN. 8

i ia 4

BASES JF)s I

r l

- w ^ h--.m__.

JUSTIFICATION FOR DIFFERENC7.5 TO NUREG 1431 BASES SECTION 3.7 PLANT SYSTEMS B, The Byron and Braidwood VC Filtration design contains two different filter units. one with heaters (makeup filter unit) and one without (recirculation filter unit). ITS SR 3.7.10.1 and associated Bases are written to reflect the appropriate requiremerts i for how long the units must be run. The Byron and Braidwtod FHB Ventilation System does not contain heaters, therefore the bracketed information in the SRs and information in the corresponding Bases (ITS SR 3.7.13.1) regarding heater operation is celeted.

B, 3 CTS 4.7.6.e.3 and 4.7.6.e.5 require verification of the VC Filtration System train ability to maintain the control room at a specified positive pressure. The allowed makeup flow rate is specified in tr ms of a range of flow rather than a maximum flow shown in NUREG SR 3.7.10.4's bracketed information. The CTS allowance is reflected in ITS SR 3.7.10.4.

Bn The Byron and Braidwood CTS requirements relative to control room air temperature is that the temperature remain s 90'F. This is reflected in replacing the NUREG bracketed information in the Bases Background and Applicable Safety Analysis sections.

00 u The Byron and Braidwood Nonaccessible Area Exhaust Filter Plenum El B Ventilation System design does not contain heaters, therefore the bracketed information in the SRs and information in the corresponding Bases regarding heater operation is deleted. Plant nomenclature refers to "emergen y mode of operation" versus " post accident mode of M operation

@l B u Not used.

83 ITS LCO 3.7.13 bracketed information relative to FHB Ventilation System requirements during Modes 1. 2. 3. and 4 are deleted. CTS and design bases analysis do not take credit for the FHB Ver:t11ation System in these Modes. Corresponding Bases information is deleted and editorially corrected as necessary.

B3 Consistent with LC0 3.0.3 the time to be in Mode 4 for ITS LC0 3.7.4 Condition C is reflected as 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. '9propriate Bases modification are made.

BYRON /BRAIDWOOD UNITS 1 & 2 3.7 2 10/10/97 Revision B

JUSTIFICATION FOR DIFFERENCES TO NUREG 1431 BASES SECTION 3.7 PLANT SYSTEMS B, 3 Not us $d.

8, 3 Not used.-

1 B, 3 The FHB Ventilation System does not have a normal and post accident mode of operation 3er se. Flow from the fuel handling building is normally  ;

routed throug1 the HEPA filters and then through the auxiliary building alenum and exhaust fans. Plant nomenclature is such that reference to

HB Ventilation System in the emergency mode implies the charcoal filter i units and the FHB Ventilation System exhaust fans. Therefore the

@ bracketed ")ost accident" has been revised to emergency mode . In addition, tie CTS requirements to perform the "1/4 inch test does not jl include a flow rate. Consistent with NUREG-1431 an upper flow rate has F .een added.

4 B,3 The plant specific number of SG PORVs is provided. The analysic l described in UFSAR Section 15.6.3 does not assume an initial condition of one inoperable SG PORV.

B3 (Byron Only) NUREG SR 3.7.9.1 (ITS SR 3.7.9.2) Bases are revised to reflect the Byron design. The Byron design credits makeup ca) ability to meet tt.e 30 day requirement. In addition, the cooling tower aasin '

' inventory provides a source for auxiliary feedwater. Therefore the

. SR Bases discussion is revised accordingly, ,

' ~

d NUREG SR 3.7.9.2 (ITS SR 3.7.9.3) Bases are revised to reflect the Byron CTS requirements. The VHS analysis includes two temperature

- assumptions, depending on fan operating status. Therefore the SR Bases discussion is revised accordingly.

NUREG SR 3.7.9.3 (ITS SR 3.7.9.5) Bases are revisec'. to reflect Byron CTS and design. There are a total of 8 fans of which only 6 are " required".

' The fans have two speeds of which the'high speed is credited in the accident anal accordingly. yses. Therefore the SR Bases discussion 1s revised NUREG LC0 3.7.9 SR 3.7.9.4 is not applicable to the Byron design. The' cooling tower fans do not get any automatic start signals. Therefore this SR is deleted.

i

'~ -

BYRON /BRAIDWOOD UNITS 1 & 2 3.7 3 10/10/97 Revision B

-r =

g n----y ~w - + --s+ --

.mw - p -u.- 4..e-s -

7+e- m -- .= r-r-

6 9

NS-C 1

NO SIGNIFICANT HAZARDS EVALUATION ITS SECTION ').7 PLANT SYSTEMS TECHNICAL CHANGE LESS RESTRICTIVE "Soecific~"

("L 3* Labeled Comments / Discussions)

Commonwealth Edison Company (Comed) has evaluated each of the pro Technical Specification changes identified as "Technicai Change posed Less Restrictive (Specific)" in accordance with the criteria set forth in 10 CFR 50.92 and has determined that the proposed changes do not involve a '

, significant hazards consideration.

The bases for the determination that the proposed changes do not involve a significant hazards consideration is an evaluation of these changes against each of the criteria in 10 CFR 50.92, The criteria and the conclusions of the

. evaluation are presented below.

1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?

Q1

> The proposed change deletes the CTS requirement to maintain the train E flow rate greater than a specified value when conducting the maintain negative pressure test of the Nonaccessible Area Exhaust Filter Plenum -

Ventilation System. The up)er flow rate limit is retained, The Nonaccessible Area Exhaust 711ter Plenum Ventilation System is not considered as an initiator for any analyzed accident. Therefore the 3roposed change associated with the Nonaccessible Area Exheust Filter

)lenum Ventilation System will not significantly increase the probability of occurrence for a previously analyzed accident.

The consequences of a previously analyzed event are dependent on the initial conditions assumed for the analysis, and the availability and ,

successful functioning of the equipment assumed to operate in response

~ to the analyzed event, and the setpoints at which these actions are initiated. Deleting the restriction on the lower flow rate during the Nonaccessible Area txhaust Filter Plenum Ventilation System negative

)ressure test does not effect the ability of the Nonaccessible Area Exhaust Filter Plenum Ventilation System to respond to previously analyzed accidents as designed. Other requirements exist which would require reconfirmation of the bank flow rates in the case of structural changes, As a resulte no analyses' assumptions are violated. Based on this evaluation. there is no significant-increase in the consequences of a previously analyzed accident.

BYRON /BRAIDWOOD -UNITS 1 & 2' 3.7 66a 10/10/97 Revision B 4 *.e - s ---

e"4v+- se u- 7 m

NO SIGNIFICANT HAZARDS EVALUATION ITS SECTION 3,7 PLANT SYSTEMS

2. Does the change create the possibil4ty/of a new or different kind of accident irom any accident previously evaluated?

The proposed change does not involve a physical alteration of the plant.

No new equipment is being introduced, and installed equipment is not being operated in a new or different manner. There is no change being made to the parameters within which the plant is operated. There are no setpoints at which protective or mitigative actions are initiated affected by this change. This change will not alter the manner in which equipment operation is initiated, nor will the function demands on credited equipment be changed. No alteration in the procedures which ensure the plant remains within analyzed limits is being proposed, and no change is being made to the procedures relied upon to respond to an off-normal event. As such, no new failure modes are being introduced.

o.) The change does not alter assumptions made 1.1 the safety analysis or u

licensing basis. Therefore, the change does not create the possibility or of a new or different kind of accident from any accident previously evaluated.

3. Does this change involve a significant reduction in a margin of safety?

The margin of safety is established through equipment design, operating parameters, and the setpoints at which automatic actions are initiated.

Sufficient equipment remains available to actuate upon demand for the purpose of mitigating an analyzed event. There is no detrimental impact on any equipment design parameter, and the plant will still be required to operate within prescribed limits. Therefore. the change does not involve a significant reduction in the margin of safety.

BYRON /BRAIDWOOD UNITS 1 & 2 3.7 66b 10/10/97 Revision B l

ENCLOSURE 2 Section 1.0 I

k

6 BYRON CTS MAR (UPS i

I

Deb % 4%s 1.I OEFINITIONS Lco 3. 6.1 9'"b'd'"D E 0 CONTAINMENLINTEGRITY f g c -l. .s.s., 3. t..

1.7 CONTAINMENT-INTEGRITY shall-exist when:

i a.

All arepenetrations either: -

required to be clos 6d' during accident conditions 1)

Capable of being closed by cn OPERABLE containment automatic isolation valve system, or 2)

Closed by manual valves, blind. flanges, or deactivated automatic valves secured in their closed positiv.is, except as provided in Table 3.6-1 of Specification 3.6.3.

b.

All equipment hatches are closed and sealed, c.

Each air lock is in compliance with the requirements of Specification 3.6.1.3, d.

The containment 3.6.1.2, and leakage rates are within the limits of Specification

,n h o) s.o.see T e.

bellows, or 0 rings) is OPERABLE.The sealing mechanismDocs for s.lassociated e.,i s.o with CONT 7 LLEDIIAKAGE/ %

oola t u al CORE ALTERATION Q w, ,<- rMuWy ca:Wuil d2!B CORE ALTERATION shall be the movementer:rnmmM Suspension of CORE ALTERATI01@shall not preclude completion of movem compongnttoasag r_:gition.

MITtSALIK1 ANALYSIS OF BYRON AND BRAIDWOOD STATION FU l.9.a The CRITICALITY F BYRON AND RACKS, is a document that provide D STATION FUEL STORAGE storage. These limits um allowable fuel enrichment for q -

Specification e determined an l

.. . in accordance with (h pecif),a$ ions. Plant operation within these limi ressed 1

%[

'A A A ~

A DIGITAL w/

TIONAL TEST 1.10 A DIGITAL CHANNEL computer hardwar

%ta *^ ata base manipulation and insist of exercising the digital y OPERABILITY of alam and/or trip functions. ' ulated process BYRON - UNIT 1 1-2 AMENDMENT NO. 29 Rev.B

_ _ _ _ l

)

BR O CTS MARKU35

'~

@ inMC" M

-_ - .w er 1 5 5.0 h0EFINITIONS

~

(Addressed S *a b oc. Q S u h.u, 3.(. < Sn%e)e E (s.

TONTA'INNENT~INTEditIW X ~ ' ~ ^ ~

^

1.7 CONTAINNENT INTEGRITY shall exist when:

a.

All penetrations required to be closed during accident conditions are either: ~

1) Capable of being closed by an OPERABLE containment automatic isolation valve systa.a. or

2) Closed by manual valves blind flanges, or deactivated automatic valves secured-in their, closed positions, except as provided in Table 3.6-1 of Specification 3.6.3.

b.

All equipment hatches are closed and sealed, c.

Each air lock is in compliance with the requirements of Specification 3.6.1.3, d.

The containment 3.6.1.2, and leakage rates are within the limits of Specificatio[ s *h, L

e. w0%:

The sealing hallnwa. mechanism nr 0-rings) associated with each penetration (e.g., welds, is OPERABLE. 2Dl ICONTRdLLEDLEAKAGE(

1. 8 CONTROlLEDLEA to antpum)pseals. shallbet[atsealwat4rflow-sugliedtot)le-reacto)

/

CORE ALTERATION l- (5uA5.e.., W V4yunkyl %

QCORE ALTERATION shall be the movementMCeiimmih of anykomponenQ) within the reactor vessel with the vessel head removed and-fuel in the Suspension ofto CORE ALTERATI0 shall not preclude completion of movement of a component - a safiFaununumu@nden positiort .

1. CRITICALITYANALYSI5'0FSYRONANDBRAIDWOODSTATION* FUEL N /TORAGERACK 1.9a Th'CRITICALITYAMLYSISOFBYRON k f RACKS, i a document t 'at provides the 0BRAIDWOODSJTIONFUELST0 E , ce i storag ximus allowab e fuel enrichme for 1 These limit shall be deters ned and subai ed in accordan with .

g.

du S i _ _n .J su -

DIGITALCHAIMELOPERATIONA[ TEST '

1.10' A.D ITAL CNANNEL 0 ERATIONAL TEST hall consist of exercising the igital computer ardware using ata base manip ation and injec ing simulated ocess gtato erify OPERABI TY of alarm an or trip functi s.

DOSE EQUIVALENT I-131 G3) which alone would produce the same thyroid-dose as the quantity a mixture of-I-131,1-132 I-133,1-134, and 1-135 actually present. The thyroid

' dose conversion factors used for this calculation shall be those listed in Table III of TID-14844, " Calculation of Distance Factors for Power and Test Reactor Sites."

BRAIDWOOD UNITS 1 & 2 1-2 AMEN 0 MENT tt0. 18 fPev. B h E o letier claiecl Eept en uber20 M' 1

C~S DOCS

DISCUSSION OF CHANGES TO CTS ITS SECTION 1.0 USE AND APPLICATION A. The CTS CONTROLLED LEAKAGE definition h'as been deleted and exception to IDENTIFIED LEAKAGE and UNIDENTIFIED LEAKAGE dafinitions were added consistent with NUREG-1431. The definition is no longer required for the ITS because the addition of new LCO 3.5.5 will ensure that seal injection flow remains within limits. As such, this ciange is not a technical change and is considered to be administrative.

A, The CTS CORE ALTERATIONS definition has been revised to delete the word

" conservative" to eliminate the confusion associated with the movement of a component to a " safe" position consistent with NUREG-1431. When CORE ALTERATIONS are required to be suspended, it is known that a s)ecific movement may have to be completed. Completing the movement tlat was in progress at the time of the requirement to suspend Core Alterations is required to establish a " safe" configuration (e.g., no fuel bundle suspended from the fuel mast). The requirement to establish a " safe" position is deemed proper and sufficient. Eliminating the requirement to also be a " conservative" position avoids potential confusion and perhaps overly restrictive interpretation. Since there is no reference on which to base the conservative evaluation (i.e,.

conservative with respect to what?) it is assumed that " conservative" i

is intended to reflect the same context as " safe." That is, if it is

" safe" it is also " conservative." Given this understanding, the wording change is editorial. This is acceptable since " safe" adequately controls the allowance to complete the move. This is perceived as the 0 intent of the CTS wording, and therefore, the revised wording more

> accurately reflects this intent and is considered to be administrative.

Ol Ar The CTS DIGITAL CHANNEL OPERATIONAL TEST. MEMBER (S) 0F THE PUBLIC, PROCESS CONTROL PROGRAM, PURGE - PURGING. REPORTABLE EVENT. SITE BOUNDARY. SOURCE CHECK. UNRESTRICTED AREA. VENTILATION EXHAUST TREAT SYSTEM. VENTING. and WASTE GAS HOLDUP SYSTEM definitions have been deleted consistent with NUREG-1431. These definitions are deleted since the Technical Specifications referencing the definitions no longer contain their use, or no longer are retained in the ITS. Discussion of the technical aspects of this change are addressed in each Technical Specification where the definitions are used. The removal of the definitions is considered administrative, with no impact of its own.

A, The CTS FREQUENCY NOTATION definition and CTS Table 1.1 have been deleted since the abbreviations in Table 1.1 are no longer used. All Surveillance Requirement Frequencies in the ITS are directly specified.

The removal of the definition and Table 1.1 is considered administrative, with no impact of its own. This change is consistent with NUREG-1431.

BYRON /BRAIDWOOD UNITS 1 & 2 1.0 3 10/10/97 Revision B J

DISCUSSION OF CHANGES TO CTS ITS SECTION 1.0 USE AND APPLICATION Aa The ' definition of SHUTDOWN MARGIN has been revised to add the phrase "With any RCCA not capable of being fully ir,serted, the reactivity worth of the RCCA must be accounted for in the determination of SDM: and" consistent with NUREG-1431. This requirement to account for RCCAs not

-capable of being fully inserted was simply moved from CTS 4.1.1.1.1.a and 4,1,1.2.a. Additionally, certain wording preferences or English language conventions were adopted which results-in no technical changes (either actual or interpretational) to the TS, As a result, the TS should be more readily readable. and therefore understandable.-by plant operators as well as other users.

An The CTS E-BAR - AVERAGE DISINTEGRATION ENERGY definition has been revised to add the recuirement to assure that at least 95% of the total non-iodine activity bs included consistent with NUREG-1431. The CTS definition requires that E-bar be the weighted average in proportion to l

the concentration of each radionuclide. of the sum of the average beta and gamma energies per disintegration.for the radionuclides in a sample.

The revision explicitly exempts all iodines and requires that at-least 95% of the total non-iodine activity with half-lives greater than 10 minutes in the coolant be included. This requirement is specifically denoted in the CTS Table 4.4-4. Table Notation ***. This revision to l

the definition provides clarification of the current requirements

.O. without any modification of intent and provides a reformatting of existing requirements. The change is a presentation preference which is editorial'in nature and does not involve a technical change (either actual or interpretational) to the TS. The change is consistent with NUREG-1431.

An -NRC letter dattG April 2.1997 issued Amendment 86 for Byron and Amendment 78 for Braidwood for soluble boron in the spent fuel pool (SFP). Since the license amendments were temporary in nature. Comed letter dated June 30. 1997 aro for soluble boron in the SF). posed changesComed Additionally. to permanently respondedtake credit to the NRC's request for additional information in Comed letter dated September 25. 1997. Although not yet approved by the NRC. Comed has E) used the June 30. 1997 and the September 25. 1997 submittal revisions as

,; the CTS markup 3 ages for the ITS conversion. The clouded portions-o reflect these c1anges. Reference to the CRITICALITY ANALYSIS OF BYRON k

AND BRAIDWOOD STATION FUEL STORAGE RACKS is located in ITS Section 4.3.1. " Criticality." and in the Reference section of the Bases for ITS LCOs 3.7.15 and 3.7.16.

O BYRON /BRAIDWOOD - UNITS 1 & 2 1.0-7 , 10/10/97 Revision B 1

ENCLOSURE 2 Section 3.7 l

4 BYRON ITS i

l

-Spent Fuel Pool Water Level-3.7.14 3.7 PLANT SYSTEMS 3.7.14 Spent Fuel Pool Water Level LCO 3.7.14- The spent fuel pool water level shall be = 23 ft over the top of irradiated fuel assemblies seated in the' storage racks.

= APPLICABILITY: During movement of irradiated fuel assemblies:in the spent fuel pool.

ACTIONS

.....................................N0TE-------------------------------.-----

LC0 3.0.3 is not applicable.

g CONDITION REQUIRED ACTION COMPLETION TIME Jt te A. - Spent fuel pool water A.1 Suspend movement of .Immediately level not within irradiated fuel-limit. -assemblies in the spent fuel pool.

1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY-SR. 3.7.14.1 Verify the spent fuel pool water level is 7-days

= 23 ft above the top of the irradiated fuel assemblies seated in the storage-racks.

BYRON - UNITS 1 & 2 3.7-37 10/10/97 Revision B

Spent Fuel Pool Boron Concentration 3.7.15 3.7 PLANT SYSTEMS 3.7-.15- Spent Fuel Pool Boron Concentration C0 3.7.15 The spent fuel _ pool boron concentration shall be = 2000 ppm.

APPLICABILITY: Whenever fuel assemblies are stored in the spent-fuel pool.

ACTIONS-

.....................................N0TE---------------------------......

LC0 3.0.3-is not applicable.

-m CON 0iTION REQ'JIRED ACTION- COMPLETION TIME A. Spent fuel pool boron A.1 Suspend movement of-- Immediately concentration not fuel assemblies in~

within limit, the spent fuel-pool.

i bE A.2 - ' Initiate action to Immediately restore. spent fuel-pool. boron concentration-to within limit.

-1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FRE0VENCY SR 3;7.15 1 .

Verify the spent fuel pool boron 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> concentration is = 2000' ppm.

BYRON - UNITS 1 & 2 ~ 3.7-38 10/10/97 Revision B

Spent-Fuel Assembly Storage-

-3.7.16 3.7 : PLANT SYSTEMS 3.7.16- Spent Fuel Assembly l Storage ,

LCO' 3.7.16 Each s shall: pent fuel assembly stored in the spent fuel pool

a. Region 1 Have an initial nominal enrichment of s 4.7 weight aercent U-235 or satisfy a minimum number of Integral

uel Burnable Absorbers (IFBAs) for . higher initial l enrichments up to 5.0 weight percent U-235 to permit-storage in~ any cell--location.

_g b. Region 2

[ Have a combination of initial enrichment, burnup, and

-decay time within the Acceptable Burnup Domain of Figure 3.7.16-1, 3.7.16-2 or 3.7.16-3, as applicable for that storage configuration.

c. Interface Requirements Comply with the Interface Requirements within and -

between adjacent racks as described in the " Byron and Braidwood Spent Fuel Rack Criticality Analysis-Using Soluble Baron Credit,"

1 APPLICABILITY:

Whenever fuel assemblies are stored in the spent fuel pool.  !

BYRON - UNITS 1 & 2 3.7-39 10/10/97 Revision B

-~

e Spent Fuel Assembly 3.7.Stora$6 ACTIONS

.....................................N0TE...-.......-...-..-..-....-..........

LC0 3.0.3 is not applicable.

CONDITION REQUIRED ACTION COMPLETION TIME I A. Requirements of the A.1 Initiate action to Immediately LCO not met. .

move the noncomplying fuel assembly into a location which restores compliance.

03 j SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.16.1 Verify by administrative means the initial Prior to nominal enrichment of the fuel assembly is storing the s 4.7 weight percent U.235 or a minimum fuel assembly number of IFBAs is met. in Region 1 SR 3.7.16.2 Verify by administrative means the Prior to combination of initial enrichment, burnup, storing the and decay time of the fuel assembly is fuel assembly within the Acceptable Burnup Domain of in Region 2 Figure 3.7.16-1, 3.7.16-2. or 3.7.16-3.

SR 3.7.16.3 Verify by administrative means the Prior to interface requirements within and between storing the adjacent racks are met, fuel assembly in the spent fuel pool BYRON - UNITS 1 & 2 3.7-40 10/10/97 Revision B

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Figure 3.7.16-1 (page 1 of 1)

Region 2 All Cell Configuration Burnup Credit Requirements BYRON - UNITS 1 & 2 3.7-41 10/10/97 Revision B

Spent Fuel Assembly Storage 3.7.16 45000 ,

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Region 2 3-out-of-4 Checkerboard Configuration Burnup Credit Requiremrits BYRON - UNITS 1 & 2 3.7-41a 10/10/97 Revision B

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Figure 3.7.16-3 (page 1 of 1)

Region 2 2-out-of-4 Checkerboard Configuration Burnup Credit Requiremnts BYRON - UNITS 1 & 2 3.7-41b 10/10/97 Revision B ll (O __-

Spent Fuel Pool Water Level B 3.7.14 B 3.7 PLANT SYSTEMS B 3.7.14 Spent Fuel Pool Water Level BASES BACKGROUND The minimum water level in the spent fuel pool meets the assuin]tions of iodine decontamination factors following a fuel landling accident. The water also provides shielding during the movement of spent fuel.

A general description of the spent fuel pool design is given in the UFSAR Section 9.1.2 (Ref.1). A description of the Spent Fuel Pool Coolin UFSAR, Section 9.1.3 (g and Ref. 2).Cleanup System is given The assumptions of theinfuel the- ,

I handling).

Ref. 3 accident are given in the UFSAR Section 15.7.4 O

i

@ APPLICABLE The minimum water level in the spent fuel pool meets the SAFETY ANALYSES assumptions of the fuel handling accident described in Regulatory Guide 1,25 (Ref. 4). The resultant 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> thyroid cose per person at the exclusion area boundary is a small fraction of the 10 CFR 100 limits (Ref. 5).

According to Reference 4. there is 23 ft of water between the top of the damaged fuel bundle and the fuel pool water surface during a fuel handling accident. With 23 ft of water, the assumptions of Reference 4 can be used directly.

In practice. this LCO preserves the assumption for the bulk of the fuel in the storage racks. In the case of a single bundle droppeo and lying horizontally on top of the spent fuel racks, however, there may be < 23 ft of water above the width of the bundle. To offset this small nonconservatism, the analysis assumes that all fuel rods fail, although analysis shows that only the first few rows fail from a hypothetical maximum drop.

The s)ent fuel pool water level satisfies Criterion 2 of 10 CF:1 50.36(c)(2)(ii).

0 (continued)

BYRON - UNITS 1 & 2 B 3.7-94 10/10/97 Revision B

Spent' Fuel Pool' Water Level-B 3.7.14 BASES (continued) 4 LCO. The_ spent fuel-pool water-Tevel is required to be a-23 ft ov6r the to) of irradiated fuel assembiies seated in the storage racts.. The-specified water-level preserves the=

assumptions of the fuel handling accident analysis-(Ref. 3).

As such. it is the minimum repJired for fuel storage and movement within_the spent fuel pool.

APPLICABILITY- This LCO applies during movement of irradiated fuel assemblies in the spent fuel pool. since the potential for- a-release of fission products exists. -

ACTIONS-The ACTIONS have been modified by a Note indicating that

to - LC0 3.0.3 does not apply.

Y g g When the initial conditions assumed in the accident analysis cannot be met, steps.should be taken-to preclude the accident from occurring. When the spent -fuel pool water level is lower than the required level, the movement of irradiated fuel assemblies in the spent fuel pool is.

immediately suspended-to a safe position. This action.

-effectively 3recludes the occurrence of:a fuel handling:

accident. . T11s does not preclude movement of a fuel assembly.to a safe position.

If moving irradiated fuel assemblies while.in MODE 5 or- 6,

' LCO 3.0.3 would not saecify any action. If moving-irradiated fuel assem) lies while in MODES 1. 2. 3. and 4.

the fuel movement-is independent of reactor operations.

Therefore. inability to suspend' movement of irradiated fuel assemblies is not sufficient reason to require a reactor shutdown.

i.

(continued)

BYRON - UNITS 1 & 2 B 3.7-95 10/10/97 Revision B

Spent Fuel Pool-Water Level B 3.7.14-BASES' (continued)

.SURVElLLANCE SR 3;7.14.1 -

REQUIREMENTS, This SR ' verifies sufficient spent fuel pool- water is available in the event of a fuel handling' accident - The -

water level in the spent fuel pool must= be checked periodically. The -7 day Frequency is appropriate because the volume in the pool is normally stable. Water level changes are controlled by plant procedures and are acceptable based on operating experience.

During refueling operations, the level in the spent 1 fuel pool 1s in equilibrium with the refueling cavity when they are hydraulically coupled, and the level in the refueling cavity is checked daily in accordance with SR 3.9.7.1, a) .

h REFERENCES 1. UFSAR. Section 9.1.2.

Q'

2. UFSAR Section 9.1.3.

3. UFSAR. Section 15.7.4.

4 Regulatory Guide 1.25. May 1972.

3. 10 CFR 100,11, i

i a

y l BYRON - UNITS 1 & 2 B 3.7-96 10/10/97 Revision B

Spent Fuel Pool Boron Concentration B 3.7.15

~B 3.7 PLANT SYSTEMS-B'3.7.15 Spent Fuel Pool Boron Concentration --

l BASES-BACKGROUND- The spent fuel-pool provides for storage of various

-Westinghouse Optimized fuel Assembly (0FA) types of different initial fuel enrichments and exposure histories in two distinct regions. There are 23 separate racks which provide placement locations for a total of.2870~new or used fuel assemblies. Included in this are six. specific storage locations in one of the racks for placement of failed fuel

-assemblies. These locations are identified as the failed -

fuel storage cells. -Of the 23 racks four are designated

" Region 1" with the remaining 19 racks designated as

" Region 2". The anal

(g criticalityanalyses{asbeenreviewedandapprovcdbythe-1 cal meth ,

NRC (Ref. 1).

I Region 1 racks contain 392 cells which are analyzed for Oc storing Westinghouse 0FAs in an "All . Cells" arrangement (that is, the criticality analysis assumes that spent fuel assemblies reside in all available cell locations, with the exception of the boundary requirements). The-stored fuel-1 assemblies may contain an initial nominal enrichment of-s 4.7 weight percent U-235-(without Integral Fuel Burnable >

Absorbers (IFBAs)' installed) up-to an initial nominal enrichment of s 5~.0 weight percent U-235. provided that the requirement for a minimum number of 16;IFBAs is met' (Ref. 2). The IFBAstare required to-have, as a minimum, a boron loading S

of 1.0X.-equal to' an amount of 1.5 mg B / inch. :This is the minimum standard poison material loading offered by Westinghouse.for 17X17.OFAs.

. Region 2 racks contain 2472 cells which are also analyzed for storing Westinghouse 0FAszin a combination'of storage configurations. These patterns are:

1) "All Cells" Storage:

2) "3-out-of-4 Checkerboard" Storage: and

3) "2-out-of-4 Checkerboard" Storage.

.v (continued) t - BYRON - UNITSil & 2 B 3.7-97 10/10/97 Revision B y -~

m Spent Fuel Pool Boron Concentrat' ion-B 3.7.15:

' BASES.

. BACKGROUND For the "All Cells" storage'configur6 tion, the stored fuel (continued) assemblies may contain an initial- nominal enrichment of s 1.14 weight percent U-235. (without taking credit for Wel burnup or radioactive decay of fuel constituents) u) to an initial nominal enrichment of s 5.0 weight percent J-235, when fuel burnup and radioactive decay of fuel constituents are credited.

For the "3-out-of-4 Checkerboard" storage. configuration, the stored fuel assemblies may contain an initial nominal-enrichment of s 1.64 weight percent U-235 (without taking credit for fuel burnup or radioactive decay of f el l

constituents) weight- percent u)J-235, to an initial nominal enrichment of s 5.0.

when fuel burnup and radioactive decay.

of fuel constituents are credited. In this storage pattern.

there can be no more than three stored assemblies in any 2X2 g matrix of cell lattices.

i U For the "2-out-of-4 Checkerboard" storage configuration, the

.l stored fuel assemblies 'nay contain an initial nominal enrichment of s 4.10 wei credit-for fuel buroup) ght up percent U-235 to an initial (without nominal taking enrichment' of s 5.0 weight percent U-235, wher: fuel burnup-is credited.

In this storage pattern, no two fue; assemblies may be stored " face adjacent" (that is, th::re must be an empty cell-

-opposite each face of the fuel assembly).

4

-The water in the spent fuel pool normally contains soluble boron which results-in large subcriticality margins under

-actual operating conditions.

APPLICABLE NRC approved methodolo SAFETY ANALYSES -criticality analyses (gies Ref. 1) .were Theused fuel to develop handling the accident analyses-are provided.in Reference 3. The accident analyses-for criticality and spent fuel pool dilution are provided in References 2 and 4. respectively.

-1 i

(continued)

-BYRON - UNITS 1-& 2 B 3.7-98 10/10/97 Revision B

L Spent Fuel Pool Boron Concentration B 3.7.15 BASES APPLICABLE The criticality analyses for the spent fuel assembly storage SAFETY ANALYSIS racks confirm that k remain (continued) uncertainties and tol,e,rances)sat<a1.0 95%(including probability with a 95% confidence level (95/95 basis), based on the accident condition of the pool being flooded with unborated water.

ihus, the design of both regions assumes the use of unborated water while maintaining stored fuel in a subcritical condition.

However, the presence of soluble boron has been credited to provide adequate safety margin to maintain spent fuel assembly storage rack k , s 0.95 (also on a 95/95 basis) for all postulated accident,, scenarios involving dropped or misloaded fuel assemblies and loss of spent fuel pool temperature control. Crediting the presence of soluble boron'for mitigation of these scenarios is acceptable based (0

. on applying the " double contingency principle" which states that there is no requirment to assume two unlikely,

.g independent, concurrent events to ensure protection against a criticality accident (Refs. 5 and 6).

I The accident analyses address the following five postulated scenarios:

1) fuel assembly drop on top of rack:

2) fuel assembl drop between rack modules:

3) fuel assembl drop between rack modules and spent fuel pool wa 1:

4) change in spent fuel pool water temperature: and

5) fuel assembly loaded contrary to placement restrictions.

Of these only the last two have the capacity to increase reactivity beyond the analyzed condition.

Calculations were performed to. determine the reactivity change caused by a change in spent fuel pool water temperature outside the normal range (50 - 160*F). For the change in spent fuel pool water temperature accident, a temperature range of 32 - 240*F is considered. In all cases, additional reactivity margin is available to the 0.9E k,,, limit to allow for temperature accidents. The temperature change accident can occur at any time during operation of the spent fuel pool.

(continued)

BYRON - UNITS 1 & 2 B 3.7-99 10/10/97 Revision B

Spent Fuel Pool Boron Concentration B 3.7.15 BASES

-APPLICABLE For.the fuel assembly misload accident. calculations were SAFETY ANALYSIS performed to show the largest reactivity increase caused by (continued) a Westinghouse 17X17 0FA fuel assembly misplaced into a storage cell for which the restrictions on location, enrichment. or burnup are not satisfied. The assembly misload accident-can only occur during fuel handling operations in the spent fuel pool.

For the above postulated accident conditions. the double contingency principle can be a) plied. Specifically, the  ;

presence of soluble boron in tie spent fuel pool water can se assumed as a realistic initial condition since not-assuming its presence would be a second unlikely event.

Spent fuel pool soluble boron has been credited in the criticality safety analysis to offset storage rack and fuel assembly tolerances. calculational uncertainties, g- uncertainty associated with burnup credit and the reactivity j increase caused by postulated accident conditions.

L I Based on the above discussion. should a spent fuel pool water temperature change accident or a fuel assembly misload accident occur in the Region 1. Region 2.-or failed fuel storage cells, k s will be maintained s to'0.95 due <to the presence of at least 550 ppm (no fuel handling) or 1650 ppm (during fuel handling) of soluble boron in the spent fuel pool water.

A spent fuel pool dilution analysis (Ref. 4) has been performed as required by Reference 7. The analysis assumes an initial boron concentration of 2000 ppm. The dilution analysis concludes that an unplanned or inadvertent event that would result in the dilution of the spent fuel pool boron concentration from 2000 ppm to 550 ppm (minimum non-accident boron concentration) is not credibid, m

4 (continued)

BYRON - UNITS 1 & 2 B 3.7-100 10/10/97 Revision B

Spet Fuel Pool Boron Concentration -

B 3.7.15 BASES-

-APPLICABLE Interface requirements have'been established to ensure k,,,

SAFETY ANALYSIS is maintained within the appropriate limits. There are (continued)- interface requirements between Region 1 racks, between Region 1-and Region 2 racks, between Region 2 racks, and within racks between different checkerboard configurations.

These requirements are necessary to account- for unique geometries and configurations which exist at.the interfaces.

Interface requirements exist between adjacent-racks to account for the potential reactivity increase in 3-out ;f-4 '

and 2-out-of-4 storage configurations a.long the interface'.

with non-aligned racks.  ;

p The concentration of dissolved boron in the spent fuel pool.

1 satisfies Criterion-2 of 10 CFR 50.36(c)(2)(11).

D LCO The spent. fuel pool boron concentration is required to be s = 2000 ppm. . The specified concentration of dissolved boron g' in the spent fuel pool preserve; the assumptions used-in the analyses. This concentration of dissolved boron is the minimum required concentration to permit fuel assembly:

storage within the spent fuel pool.

APPLICABillTY This LCO applies whenever fuel assemblies are stored'in the spent fuel pool.

-The presence of soluble boron (in various concentrations)-is assumed in the criticality analyses and is credited for-ensuring that spentifuel pool k will be maintained s 0.95 at a 95% confidence level for a9 storage configurations.

The 2000 ppm minimum boron concentration-is also an initial condition in the spent fuel pool dilution analye.is. .

Therefore, the restriction on soluble boron concentration in the spent fuel acol water must be maintained at all times when fuel- assem) lies are stored in the spent fuel pool.

I (continued)

BYRON - UNITS 1 & 2 B 3.7-101 10/10/97 Revision B

Spent Fuel Pool Boron Concentration B 3.7.15 BASES (contin ed)

ACTIONS The ACTIONS have been modified by a Note indicating that LCO 3.0.3 does not apply. '

A.1 and A.2 When the t.oncentration of boron in the spent fuel 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. This does not preclude movement of a fuel assembly to a safe position. Immediate actions are )

also taken to restore spent fuel pool boron concentration to a 2000 ppm.

If moving fuel assemblies while in MODE 5 or 6. LCO 3.0.3 would not s)ecify any action. If moving fuel assemblies O while in M0)ES 1, 2. 3 and 4. the fuel movement is i independent of reactor operations. Therefore. inability to suspend movement of fuel assemblies is not sufficient reason y to require a reactor shutdown.

i SURVEILLANCE SR 3.7.15.1 REQUIREMENTS This SR verifies that the concentration of boron in the s)ent fuel pool is a 2000 ppm. As long as this SR is met, tie analyzed accidents are fully addressed.

The 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> frequency is appropriate based on o>erating experience and because significant changes in t1e boron concentration in the spent fuel pool are difficult to produce without detection, considering the large volume of water contained in the spent fuel pool. An analysis has concluded that a spent fuel pool boron dilution event of sufficient magnitude to reduce boron concentration below the minimum non-accident requirement is not credible (Ref. 4).

(continued)

BYRON - UNITS 1 & 2 B 3.7-102 10/10/97 Revision B i

Spent Fuel Pool Boron Concentration B 3.7.15 BASES (continued)

REFERENCES 1. WCAP-14416-NP-A " Westinghouse Spent Fuel Rack Criticality Analysis Methodology." Rev.1. dated November. 1996.

2. CAC-97-162 " Byron and Braidwood Spent Fuel R6ck Criticality Analysis Using Soluble Boron Credit."

dated May, 1997.

3. UFSAR. Section 15.7.4.

4. " Byron /Braidwood Spent Fuel Pool Dilution Analysis,"

M g Rev. 3. dated June 17. 1997.

O te t 5. 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).

6. ANSl/ANS 8.1 - 1983 "American National Standard for Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors."

7. Safety Evaluation Report (SER) dated October 25. 1996, issued by the Office of Nuclear Reactor Regulation for Topical Report WCAP-14416-NP-A " Westinghouse Spent Fuel Rack Criticality Analysis Methodology."

l BYRON - UNITS 1 & 2 B 3.7-103 10/10/97 Revision B

Spent Fuel Assembly Storage B 3.7.16 i B 3.7 PLANT SYSTEMS B 3.7.16 Spent Fuel Assembly Storage BASES BACKGROUND The spent fuel pool provides for storage nf various Westinghouse Optimized Fuel Assembly (OFA) types of different initial fuel enrichments and exposure histories in two distinct regions. There are 23 separate racks which provide placement locations for a total of 2870 new or used fuel assemblies, included in this are six specific storage locations in one of the racks for placement of failed fuel assemblies. These locations are identified as the failed fuel storage cells. Of the 23 racks four are designate

" Region 1" with the remaining 19 racks designated as

" Region 2". The analytical methodology used to develop the 9 criticality analyses has been reviewed and approved by the

> NRC (Ref. 1).

Region 1 racks contain 392 cells which are analyzed for storing Westinghouse 0FAs in an "All Cells" arrangement

' (that 15. the criticality analysis assumes that spent fuel assemblies reside in all available cell locations, with the

• exception of the tcundary requirements). The stored fuel assemblies may conthln an initial nominal enrichment of s 4.7 weight percent U-235 (without Integral Feel Burnable Absor 2rs (IFBAs) installed) up to an initial nominal enrichment of s 5.0 weight percent U-235. provided that the requirement for a minimum number of 16 IFBAs is met (Ref. 2). The IFBAs are cequired to have, as a minimum. a boron lop' ding of 1.0X. equal to an amount of 1.5 ng B / inch. This is the minimum standard poison material loading offered by Westinghouse for 17X17 0FAs.

Region 2 racks contain 2472 cells which are also analyzed for storing Westinghouse OFA!. in a combination of storage-conitgurations. These patterns are:

1) "All Cells" Storage:

2) '3-out-of-4 Checkerboard" Storage:.and

3) "2-out-of-4 Checkerboard" Storage.

i b

(continued)

BYRON - UNITS 1 & 2 B 3.7-104 10/10/97 Revision B l

l

Spent Fuel Assembly Storage B 3.7.16 BASES BACKGROUND For the "All Cells" storage' configuration. the stored fuel (continued) assemblies may contain an initial nominal enrichment of s 1.14 weight percent U-235 (without taking credit for fuel burnup or radioactive decay of fuel constituents) u) to an initial nominal enrichment of s S.0 wei when fuel burnup and radioactive decay _ ght percent lJ-235.of fuel constitu are credited.

For the "3-out-of-4 Checkerboard" storage configuration. the stored fuel assemblies may contain an initial nominal enrichment of s 1.64 weight percent U-235 (without taking credit for fuel burnup or radioactive decay of fuel l constituents) u) to an initial nominal. enrichment of s 5.0 weight percent 'J-235, when fuel burnup and radioactive decay of fuel constituents are credited. In this storage pattern.

l there can be no more than three stored assemblies in any 2X2 g) matrix of cell lattices.

i T for the "2 out-of-4 Checkerboard" storage configuration, the stored fuel assemblies may contain an initial nominal enrichment of s 4,10 wei credit for fuel burnup) ght percent up to U-235 an initial (without nominal enrichment taking of s 5.0 weight percent U-235, when fuel burnup is credited, in this storage pattern, no two fuel assemblies may be-stored " face adjacent" (that is, there must be an empty cell opposite each face of the fuel assembly).-

The water in the spent fuel pool normally contains soluble boron which results-in large subcriticality margins under actual operating conditions.

APPLICABi.E NRC approved methodologies were used to develop the SAFETY ANALYSES criticality analyses (Ref. 1). The fuel handling accident' analyses are provided in Reference 3. The accident analyses for criticality and spent fuel pool dilution are provided in References 2 and 4. respectively.

(continued)

BYRON - UNITS 1 & 2 B 3.7-105 10/10/97 Revision'B

Spent Fuel Assembly Storage B 3.7.16 BASES ,

1 4

APPLICABLE The. criticality analyses for the spent fuel assembly storage SAFETY ANALYSIS racks confirm that k remain (continued) uncertainties and tol,e,rances)s . at<a1.0 951(including probability with a  !

95% confidence level (95/95 basis), based on the accident '

condition of the pool being flooded with unborated water.-

Thus, the design of both regions assumes the use of unborated water while maintaining stored fuel in a ,

subcritical condition, t However..the presence of soluble boron has been credited to .

provide adequate safety margin to maintain spent fuel t assembly storage rack k,,, s 0.95 (also on a 95/95 basis) for .

all postulated accident scenarios involving dropped or- r misloaded fuel assemblies and loss of spent fuel pool temperature control. Crediting the presence of soluble ,

boron for mitigation of these scenarios is acceptable based on applying the " double contingency principle" which states O that there is no requirement to assume two unlikely, independent, concurrent events to ensure protection against g; a criticality accident (Refs. 5 and 6).

The accident analyses address the following five postulated scenarios:

1) fuel assembly drop on top of rack:

2) fuel assembly drop between rack modules:

3) fuel assembly drop between rack modules and spent fuel pool wall:

4) change in spent fuel pool water temperature; and-

5) fuel assembly loaded contrary to placement i restricticns.

' - Of these, only the last two have the capacity to increase-react 1vity beyond the analyzed condition.

Calculations were performed to determine the reactivity change _ caused by a change in spent fuel pool water temperature outside the normal range (50 - 160'F). For the change in spent . fuel pool water temperature accident.-a  !

temperature range of 32 - 240'F is considered. In all 4

cases, additional reactivity margin is available to the 0.95 k y limit to allow for-temperature accidents. The-temper,ature change accident can occur at any time during operation of the spent fuel-pool.

L L

?

(continued) t BYRON - UNITS 1 & 2 B 3.7-105a - 10/10/97 Revision B

~

..y.'..h' y

. . . , r rr rw- - - - - - e r '4 - w e,n e-. --r- r r--re-we---,e x v -,-,,enw -m,---- m - = ww.e- T- w >- wv -rr ---w t M--

1 Spent fuel Assembly Storage

. B 3.7.16 j

DASES L

APPLICABLE For the fuel assembly misloed accident, calculations were ,

i SAFETY ANALYSIS performed to show the largest reactivity increase caused by. a

. (continued) a Westinghouse 17X17 0FA fuel assembly misplaced into a  ;

storage cell for which the restrictions on location. '

[

enrichment, or burnup are not satisfied. The assembly 1 misload accident can only occur during fuel handling j operations in the spent fuel pool, l#

For the above postulated accident conditions, the double contingency principle can be a) plied. Specifically, the i

aresence of soluble boron in tie spent fuel pool water can

)e assumed as a realistic initial condition since not '

i assuming its presence would be a second unlikely event.

Spent fuel pool soluble boron has been credited in the criticality safety analysis to offset storage rack and fuel assembly tolerances, calculational uncertainties. ,

4 uncertainty associated with burnup credit and the reactivity 9 increase caused by postulated accident conditions, f Based on the above discussion, should a s water temperature change accident or a fu!ent fuel pool-1 assembly misload accident o: cur in the Region 1. Region 2.- or failed, fuel storage cells, k will be maintained s-to 0.95 due to the-presence of at l,,a,st e 550 ppm (no fuel handling) or 1650 ppm (during fuel handling) of soluble boron in the spent fuel

! pool water.

t A spent fuel pool dilution analysis (Ref. 4) has been performed as required by Reference 7. The analysis assumes an initial boron concentration of 2000 ppm. The dilution _

analysis concludes that an unplanned or-inadvertent event that would result in the dilution of the spent fuel pool '

' i boron concentration from 2000 ppm to 550 ppm (minimum l non-accident boron concentration) is not credible, J

t p 1

- 1 (continued)

BYRON - UNITS 1 & 2' B 3.7-105b 10/10/97 Revision B s , --- ,,vm ----e,,--w. -en, y - c.,-, e,v-, - - . , - - , . ,

. - ----,m,_., -

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i j Spent Fuel Assembly Storage  !

B 3.7.16 I

l BASES APPLICABLE i Interface recuirements have'been established to ensure k,,, I SAFETY ANALYSIS is maintainec within the appropriate limits. There are ,

3 (continued) interface requirements between Region 1 racks. between Region 1 and-Region 2 racks, between Region 2 racks, and. 1 i

within racks between different checkerboard configurations.  ;

i These requirements are necessary to-account for unique geometries and configurations which exist at'the interfaces.

Interface requirements exist between adjacent racks to

• account for the potential reactivity increase in 3-out-of-4 '

and 2-out-of-4 storage configurations along the interface  :

with non-aligned racks.

The configuration of fuel assemblies in the spent fuel pool satisfies Criterion 2 of 10 CFR 50.36(c)(2)(11).

- g).  !

LC0 The restrictions on the placement of fuel assemblies within D the spent fuel pool in accordance with the requirements in i 0~

the accompanying LCO ensure that the k , of the spent fuel pool will always remain < 1.0 assuming,,the pool is flooded -

with unborated water and s 0.95 assuming the presence of 550 ppm soluble boron in the pool.

7 In LC0 Figures 3.7.16-1 and 3.7.16-2 the Acceptable Burnup Domain lies on, above, and to the left of the decay time  ;

line applicable to the fuel assembly to be stored. The decay time for that assembly is measured from the time since the assembly was last discharged.

In LC0 Figure 3.7,16-3. the Acceptable Burnup Domain and the--

4 Unacceptable Burnup Domain are separated by a single line because decay time is not credited in the 2-out-of-4 Checkerboard storage configuration. The Acceptable Burnup Domain lies on, above, and to the left of the line, s

APPLIClBILITY.- This LC0 applies whenever fuel assemblies are stored in the e spent fuel pool.

4 7

(continued)

, BYRON - UNITS 1 & 2 B 3.7-105c 10/10/97 Revision.B

-,-6-d q- - > < - g--%gg 9 - w rt .yqgwge-c e-* pe -veq s-- w=44*-md--e +w- ny -ywy'- , ='g---r5

Spent Fuel Assembly Storage B 3.7.16 1

BASES (continued) l I

ACTIONS The ACTIONS have been modified by a Note indicating that I LCO 3.0.3 does not apply. j U i When the configuration of fuel assemblies stored in the spent fuel of the LCO. pool is not in accordance with the requirements-immediate action must be taken to make the necessary fuel assembly movement (s) to bring the configuration into compliance, If moving fuel assemblies while in H0DE S or 6. LC0 3.0.3

-would not s)ecify any action. If moving fuel assemblies  !

while in M0)ES 1, 2. 3, and 4, the fuel movement is independent of reactor operations. Therefore, inability to  !

suspend movement of fuel assemblies is not sufficient reason  !

to require a reactor shutdown, i S

SURVEILLANCE SR 3.7.16.1. SR 3.7.16.2. and SR 3.7.16.3 are performed REQUIREMENTS prior. to storing the fuel assembly in the intended spent fuel pool storage location. These frequencies are.

appropriate because compliance with the SR ensures that the relationship between the fuel assembly and its-storage location will meet the requirements of the LC0 and preserve.

the assumptions of the analyses.

SR 3.7.16'1 .

This SR verifies by administrative means that the initial' nominal enrichment of the fuel assembly or a-minimum number of 16 IFBAs-is met to ensure that the assumptions of the safety analyses are preserved, SR 3.7.16.2 This SR verifies by administrative means that the combination of initial enrichment, burnup, and decay time of the fuel assembly is within the Acceptable Burnup Domain of Figure 3.7.16-1, 3.7.16-2. or 3.7.16-3 for the intended storage configuration to ensure that the assumptions of the safety analyses are preserved.

7 (continued)

BYRON.- UNITS 1 & 2 B 3.7-105d 10/10/97 Revision B

8 Spent fuel Assembly Storage B 3.7.16 BASES

-SURVEILLANCE SR 3.7 16.3 .

REQUIREMENTS (continued) This SR verifies by administrative means that the interface i requirements within and between adjacent racks are met to- '

ensure that the assumptions of the safety analyses are i j preserved.  ;

a- 1 4

REFERENCES 1. WCAP-14416-NP-A " Westinghouse Spent Fuel Rack Criticality Analysis Methodology," Rev. 1. dated November,1996.

l 2. CAC-97-162 " Byron and Brai hood Spent Fuel Rack i Criticality Analysis Using Soluble Boron Credit,"

dated May, 1997, ,

(D

3. UFSAR, Section 15.7.4.-

4. " Byron /Braidwood Spent Fuel Pool Oilution Analysis,"

'Rev. 3, dated June 17, 1997.

l'

5. Double contingency principle of ANSI N16.-l - 1975, as B specified in the April 14. 1978 NRC letter (Section 1.2) and implied in the proposed revision to Regulatory Guide _ l.13 (Section 1.4, Appendix A),

i. 6. ANSI /ANS 8.1 - 1983 "American National Standard for-Nuclear Criticality Safety in Operations with i L Fissionable Materials Outside Reactors."

7. Safety Evaluation Report (SER) dated October 25, 1996. .

issued by the Office of Nuclear Reactor Regulation for Topical Report WCAP-14416-NP-A " Westinghouse Spent Fuel Rack Criticality Analysis Methodology."

l s

L

]

.l BYRON'- UNITS 1-& 2 B'3.7-105e 10/10/97 Revision B

e BMI) WOO) ::TS e

Spent Fuel Pool Water Level-3.7.14 3.7 PLANT SYSTEMS-3.7.14 Spent Fuel Pool Water Level LC0 3.7.14 The spent fuel pool water level shall be a-23 ft over the top of irradiated fuel assemblies seated-in the storage-racks.

APPLICABILITYi During movement of irradiated fuel assemblies in the spent

- fuel pool.

j ACTIONS

.....................................N0TE.....................................

[ LCO 3.0.3 is not applicable.

b &:

i --

CONDITION -REQUIRED ACTION COMPLETION TIME ~

A. Spent fuel pool water A.1 Suspend movement of' Immediately-level not within irradiated fuel l limit. assemblies in the

- spent fuel pool, j

s 4

NURVEILLANCE REQUIREMENTS

} SURVEILLANCE _ FREQUENCY i-

.SR 3.7.14.1 Verify the spent fuel pool water level is '7_dayr

' = 23 ft above the top of the irradiated fuel assemblies seated in the storage racks.

j-

,4 -

3 D f 4

J

- l' i

j. BRAIDWOOD - UNITS'1'&,2 3.7 33 10/10/97 Revision B-

+ g- wr -- ,i-3im- y- y pr y g eM1- y 6 W m y 7 w - -W ++-4 m '

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Spent Fuel Pool Boron Concentration 3.7.15 3.7 PLANT SYSTEMS 3.7.15 Spent Fuel Pool Boron Concentration LCO 3.7.15 The spent fuel pool boron concentration shall be = 2000 ppm.

APPLICABILITY: Whenever fuel assemblies are stored in the spent fuel pool.

ACTIONS

.................N0TE ------- - ---- -----------.-..-.....

LCO 3.0.3 is not applicable.

g CONDITION REQUIRED ACTION COMPLETION TIME f- A. Spent fuel pool boron A.1 Suspend movement of Immediately concentration not fuel assemblies in within limit, the spent fuel pool. ,

bhD A.2 Initiate action to Immediately restore spent fuel pool boron concentration to within limit.

I SURVEILLANCE REQUIREMENTS l SURVE!LLANCE FREQUENCY SR 3.7.15.1 Verify the spent fuel pool boron 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> concentration is = 2000 ppm.

l

[ BRAIDWOOD - UNITS 1 & 2 3.7-34 10/10/97 Revision B

Spent Fuel Assembly Storage 3.7,16 3.7 PLANT SYSTEMS 3.7.16 Spent Fuel Assembly Storage LC0 3.7.16 Each spent fuel assembly stored in the spent fuel-pool shall:

a. Region 1-Have an initial nominal enrichment of s 4.7 weight aercent r U 235 or satisfy a minimum number of Integral uel Burnable Absorbers (IFBAs) for higher initial enrichments up to 5.0 weight percent U-235 to permit storage in any cell location,

b. Region 2 '

Have a combination of initial enrichment, burnup, and decay time within the Acceptable Burnup Domain of e Figure 3.7.16-1, 3.7.16-2. or 3.7.16-3. as applicable for that storage configuration.

c.- Interface Requirements Comply with the Interface Requirements within and between adjacent racks as described ~1n the " Byron and-Braidwood Spent Fuel-Rack Criticality Analysis Using Soluble Boron Credit."

APPLICABILITY: Whenever fuel assemblies are stored in the spent fuel-pool, i

I l-H BRAIDWOOD ~ UNITS l'& 2 3.7-35 10/10/97 Revision'B s.

-, -~ . , , - , - - , - , , . - . - - . + . - - - , ..- ,

Spent Fuel Assembly Storage 1 3.7.16 l

ACTIONS

..................................... NOTE ------- - --------...........-

LCO 3.0.3 is not applicable.

CONDITION REQUIRED ACTION COMPLETION TIME A. Requirements of the A.1 Initiate action to Immediately LCO not met, move the noncomplying fuel assembly into a  !

location which restores compliance.

k SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.16.1 Verify by administrative means the initial Prior to nominal enrichment of the fuel assembly is storing the s 4.7 weight percent U-235 or a minimum fuel assembly number of IFBAs is met. In Region 1 SR 3.7.16.2 Verify by administrative means the Prior to combination of initial enrichment. burnup, storing the and decay time of the fuel assembly is fuel assembly within the Acceptable Burnup Domain of in Region 2 Figure 3.7.16-1, 3.7.16 2, or 3.7.16 3.

SR 3.7.16.3 Verify by administrative means the Prior to interface requirements within and between storing the adjacent racks are met, fuel assembly in the spent fuel pool BRAIDWOOD - UNITS 1 & 2 3.7-36 10/10/97 Revision B

3 pent fuel Assembly Storage 3.7.16 00000 DECAT TIM E:

55000  ; O YEARS 5 YEAlts

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Figure 3.7.16-1 (page 1 of 1)

Region 2 All Cell Configuration Burnup Credit Requirements l BRAIDWOOD - UNITS 1 & 2 3.7-37 10/10/97 Revision B

Spent Fuel Assembly Storage 3.7.16 15000 - i

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Region 2 3 out-of-4 Checkerboard Configuration Burnup Credit Requiremnts i 1 l-T

. l -BRAIDWOOD - UNITS 1 & 2 3.7-37a 10/10/97 Revision B 1

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Figure 3.7.16 3 (page 1 of 1)

Region 2 2-out-of 4 Checkerboird Configuration Burnup Credit Requiremnts BRAIDWOOD - UNITS 1 & 2 3.7-37b 10/10/97 Re/ision B

Spent Fuel Pool Water Levei B 3.7.14 B 3.7 PLANT SYSTEMS B 3.7.14 Spent Fuel Pool Water Level BASES i

l BACKGROUND The minimum water level in the spent fuel pool meets the assumations of iodine decontamination factors following a fuel landling accident. The water also provides shielding during the movement of spent fuel.

A general description of the spent fuel pool design is given-in the UFSAR. Section 9.1.2 (Ref.1). A description of the:

S>ent Fuel Pool Cooling and Cleanup System is gi/en in the U:SAR. Section 9.1.3 (Ref. 2). The assum handling accident are given in the UFSAR.ptions Sectionof15.7.4-the fuel (0

(Ref.'3).

k APPLICABLE The minimum water level in the spent fuel pool meets the

' SAFETY ANALYSES assumptions of the fuel handlin Regulatory Guide 1.25 (Ref. 4).g.The accident described resultant 2 hourin thyroid dose per person at the exclusion area boundary is a small fraction of the 10 CFR 100 limits (Ref 5).

According to' Reference 4.-there is 23 ft of water between the top of the damaged fuel bundle and the fuel pool water surface during a fuel handling accident. With 23 ft of water the assumptions of Reference 4 can be used directly..

In practice, this LC0 preserves the assumption for the bulk of the fuel in the storage racks, in the case of a single bundle dropped and lying horizontall,y on top of the spent fuel racks. however, there may be < c3 ft of water above the-width of the bundle. To offset this small nonconservatism, the analysis assumes that all. fuel rods fail, although analysis shows that only the first few rows fail from a hypothetical maximum drop.

The s')ent fuel pool water level satisfies Criterion 2 of 10 CF1 50.36(c)(2)(ii).

i

-i i

1 (continued)

-BRAIDWOOD - UNITS 1-& 2 B 3.7-85 10/10/97 Revision B

Spent Fuel pool dater Level B 3.7.14 BASES (continued)

LCO The spent fuel pool water level is required to be = 23 ft over the to) of irradiated fuel assemblies seated in the +

storage racts. The specified water level preserves the assumptions of the fuel handling accident analysis (Ref. 3). t As such it is the minimum required for fuel storage and movement within the spent fuel pool. i APPLICABILITY This LC0 applies during movement of irradiated fuel assemblies in the spent fuel pool, since the potential for a release of fission products exists, i

ACTIONS Ihe ACTIONS have been modified by a Note indicating that LCO 3,0.3 does not apply.

M When the initlal conditions assumed in the accident analysis cannot be met, steps should be taken to preclude the accident from occurring. When the spent fuel pool water ,

level is lower than the required level, the movement-of irradiated fuel assemblies in the spent fuel pool is immediately suspended to a safe position. This action effectively precludes the occurrence of a fuel handling accident. T11s does not preclude movement of a fuel assembly to a safe position, if moving irradiated fuel assemblies while in MODE 5 or 6.

LC0 3.0.3 would not saecify any action. If moving irradiated fuel assem) lies while in MODES 1, 2. 3. and 4.

the fuel movement is independent of reactor operations.

Therefore inability to suspend movement of irradiated fuel assemblies is not sufficient reason to require a reactor shutdown.

(continued)

BRAIDWOOD - UNITS 1 & 2 B 3.7-86 10/10/97 Revision B

Spent Fuel Pool Water Level B 3.7.14 BASES (continued)

SURVElLLANCE SR 3.7.14.1 4' REQUIREMENTS This SR verifies sufficient spent fuel pool water is available in the event of a fuel handling accident. The water level in the spent fuel pool must be checked periodically. The 7 day Frequency is appropriate because the volune in the pool is normally stable. Water level

hanges are controlled by plant procedures and are ,

acceptable based on operating experience.

During refueling oper6tiens, the level in the spent fuel pool is in equilibrium with the refueling cavity when they are hydraulically toupled, and the level in the refueling cavity is checked daily in accordance with SR 3.9.7.1.

$ REFERENCES 1. UFSAR. Section 9.1.2.

2, UFSAR. Section 9.1.3.

(

I

3. UFSAR. Section 15.< .4.

~

4. Regulatory Guide 1.25. May 1972.

5. 10 CfR 100.11.

m 3 BRAIDWOOD -~ UNITS 1 & 2 B 3.7-87 10/10/97 Revision B

Spent Fuel Pool Boron Concentration B 3.7.15 B 3,7 PLANT SYMEMS B 3.7,15 Spent Fuel Pool Boron Concentration, BASES BACKGROUND The spent fuel pool provides for storage of various Westinghouse Optimized Fuel Assembly (DFA) types of different initial fuel enrichments and exposure histories in two distinct regions. There are 23 separate racks which provide placement locations for a total of 2870 new or used fuel assemblies. Included in this are six spect fic storage locations in one of the racks for placement of failed fuel assemblies.

! fuel storage These cells. locations are identified as the failed Of the 23 racks, four are designated

" Region 1" with the remaining 19 racks designated as

" Region 2". The analytical methodology used to develop the 03 criticality analyses has been reviewed and approved by the NRC (Ref. 1).

k Region 1 racks contain 392 cells which are analyzed for storing Westinghouse 0FAs-in an "All Cells" arrangement (that is, the criticality analysis assumes that spent fuel assemblies reside in all available cell locations, with the exception of the boundary requirements). The stored fuel assemblies may contain an initial nominal enrichment of s 4.7 weight >ercent U-235 (without Integral Fuel Burnable Absorbers (IF3As) installed) up to an initial nominal enrichment of 5 5.0 weight percent U-235, provioed that the requirement for a minimum number of 16 IFBAs is met (Ref 2). The IFBAs are re gired to have, as a minimum, a boronlogdingof1.0X,equaltoanamountof 1.5 mg B / inch. This is the minimum standard poison material loading offered by Westinghouse for 17X17 0FAs.

Region 2 racks contain 2472 cells which are also analyzed for storing Westinghouse 0FAs in a combination of storage configurations. These patterns are:

1) "All Cells" Storage:

2) "3-out-of-4 Checkerboard" Storage; and

3) "2-out-of-4 Checkerboard" Storage.

(continued)

BRAIDWOOD - llNITS 1 & 2 B 3.7-88 10/10/97 Revision B w

2 dhdllibMw Sps . N ei Pool Baron Ccncee ration L e.15 BASES BACKGROUND For the "All Cells" storaganf wuration. ti.$ 'J fuel (continued) assemblies may contain an in 1e1 noare en";nment of 5 1.14 weight percent U-235 ( +bst tau na eedit for fuel burnup or radioactive decay of si etniti' ints) u) to an initial nominal enrichment 3f s " ew.1 percent'J-235 when fuel burnup and radioactive s c; v. (uel constituents are credited.

For the "3-out-of-4 Che c Y n ~ .,ge configuration. the stored fuel assemblies contain an initial nominal enrichment of s 1.64 wt , '+ p'ercent U-235 (without taking credit for fuel burnup e ^ ctive decay of fuel constituents) u) to an in a ;.

inal enrichment of s 5.0 weight percent J-235. .. we' ct aup and radioactive decay of fuel constituents r . cr eeiied in this storage pattern, there can be no more  % Wree c id assemblies in any 2X2 matrix of cell lattice (D

For the "2-out-of-4 Checte. . . Storage configuration. the stored fuel assemblies may contain an initial nominal E enrichment of 5 4.10 wei credit for fuel burnup) ghtup topercent U-235 an initial (without nominal taking enrichment of 5 5.0 weight percent U-235. when fuel burnup is credited.

In this storage pattern, no two fuel assemblies may be stored " face adjacent" (that is, there must be an empty cell _

opposite each face of the fuel assembly).

The water in the spent fuel pool normally contains soluble boron which results in large subcriticality margins under actual operating conditions.

APPLICABLE NRC approved methodologies were used to develop the SAFCTY ANALYSES criticality analyses (Ref. 1). The fuel handling accident analyses are provided in Reference 3. ~

The acciden'. dnalyses for criticality and spent fuel pool dilution are provided in References 2 and 4. respectively.

(continued)

BRAIDWOOD - UNITS 1 & 2 B 3.7-89 10/10/97 Revision B I

Spent Fuel Pool Boron Concentration B 3.7.15 BASES 4

APPLICABLE The criticality analyses for the spent fuel assenbly storage SAFETY ANALYSIS racks confirm that k remain (continued) uncertainties and tol,e,rances)s at<a1.0 95% (including probability with a 95% confidence level (95/95 basis). based on the accident condition of the pool being flooded with unborated water.

Thus, the design of both regions assumes the use o' unborated water while maintaining stored fuel in a subcritical condition.

However, the presence of soluble boron has been credited to provide adequate safety margin to maintain spent fuel assembly storage rack k,,, s 0.95 (also on a 95/95 basis) for all postulated accident scenarios involving dropped or misloaded fuel assemblies and loss of spent fuel pool temperature control, Crediting the presence of soluble boron for mitigation of these scenarios is acceptable based on applying the " double contingency principle' which states

([) that there is no requirement to assume two unlikely.

independent. cordurrent events to ensure protection against d a criticality accident (Refs, 5 and 6),

k The accident analyses address the following five postulated scenarios:

1) fuel assembly drop on top of rack:

2) fuel assembl drop between rack modules:

3) fuel assembi drop between rack modules and spent fuel pool wa 1:

4) change in spent fuel pool water temperature: and

5) fuel assembly loaded contrary to placement rastrictions.

Of these, onl/ the last two have the capacity to increase reactivity beyond the analyzed condition, Cal mlnticas were performed to determine the re6ctivity ct.ange causedo ~ y a change in spent fuel pool water temperature outside the nornal range (50 - 160'F). For the change in spent fuel pool water temperature accident, a temperature range of 32 - 240*F is considered. In all cases, addit 4nal reactivity margin is tvailable to the 0.95 k ,, limit to allow for temperature accidents. The temper,ature change accident can occur et any time during operation of the scent fuel pool, (continued)

BRAIDWOOD - UNiiS 1 & 2 B 3,7-30 10/10/97 Revision B

I Spent Fuel Pool Boron Concentration I B 3.7.15 i BASES 1 j APPLICABLE . For. the fuel assembly misloed accident, calculations were -

SAFETY ANALYSIS performed to show the largest reactivity increase caused by (continued) a Westinghouse 17X17 0FA fuel assembly misplaced into a 1 r

' storage cell for which the restrictions on location. >

enrichment, or burnup are not satisfied. The assembly

• misload accident can only occur during fuel handling '

operations'in the spent fuel pool.

For the above postulated accident conditions. the double contingency principle can be o) plied. Specifically, the i

)resence of soluble boron-in tie spent fuel pool water can '

)e assumed as a realistic initial condition since not assuming its presence would be a second unlikely event. ,

Spent fuel pool soluble boron has been credited in the criticality safety analysis to offset storage rack and fuel assembly tolerances, calculational uncertainties. ,

uncertainty associated with burnup credit and the reactivity 0 increase caused by postulated accident conditions.

Based on the above discussion. should a spent fuel pool E water temperature change accident or a fuel assembly misloat -

- accident occur in the Region 1. Region 2. or failed fuel storage cells, k,,, will be naintained s to 0.95 due to the presence of at least 550 ppm (no fuel handling) or 1650 ppm (during fuel handling) of soluble boron in:the spent fuel pool water. a A spent fuel pool dilution analysis (Ref. 4) has been performed as required by Reference-7. The analysis assumes '

an initial boron concentration of 2000 ppm. The dilution analysis concludes that an unplanned or. inadvertent event

- that-would result in the dilution of the spent-fuel pool boron concentration from 2000 ppm to 550 ppm (minimum non-accident boron concentration)-is.not credible. -

t l

(continued)

BRAIDWOOD - UNITS 1 & 2 B 3.7-91 10/10/97 Revision B we-w w-e-,egna4-w-,w w -,--pp, --v -, ----

at e. w s r-p pgeb p-+p- etsd--9't- "e.---- r ~

• Fr

Spent Fuel Pool Boron Concentration B 3.7.15 BASES APPLICABLE Interface requirements have~been established to t..isure km, SAFETY ANALYSIS is inaintained within the appropriate limits. There are (continued) interface requirements between Region-1 racks, between Region 1 and Region 2 racks, between Region 2 racks, and within racks betwaen different checkerbrar'1 configurations.

These requirements are necessar to acc>unt for unique geometries and configurations W ich ex st at the interfaces.

Interface requirements exist between-adjacent racks to account for the pott.ntial reactivity increase in 3-out-of-4 l and 2-out-of-4 storage configurations along the interface l with non-aligned racks.

-The concentration of dissolved boron in the spent fuel pool satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

LC0

(!) The spent fuel pool boron concentration is required to be

= 2000 ppm. The specified concentration of dissolved boron in the spent fuel pool preserves the assumptions used in the E analyses. This concentration of dissolved boron is the minimum required concentration to permit fuel assenbly-storage within the spent fuel pool.

APPLICABILITY This LCO applies whenever fuel assemblies are stored in the spent fuel pool.

The presence.ot soluble boron (in various concentrations) is-assumed in the criticality analyses and is credited for ensuring that spent fuel pool % , will be maintained s 0,95 at a 95% confidence-level for aN storage configurations.-

The 2000 ppm minimum boron concentration is also an initial condition in the spent fuel pool dilution analysis.

Therefore, the restriction on soluble boron concentration in the spent fuel 2001 water must be maintained at all times when fuel assem> lies are stored in the spent fuel pool.

(continued)

BRAIDWOOD - UNITS 1.& 2 B 3.7-92 10/10/97 Revision B 4

Spent Fuel Pool Boron Concentration B 3.7.15 BASES- (continued)

ACTIONS The t.CTIONS have' been modified by a Note indicating that LC0 3.0.3 does not apply.

\

A.1 and'A.2 I i

When the concentration of boron in the spent fuel pool is-less than required, immediate action must be taken to preclude the occurrence of an accident- or to mitigate the-  ;

i consequences of an accident in progress. This is most ,

efficiently achieved by immediately suspending the movement 1 of fuel assemblies. This does not preclude movetent of a i

fuel assembly to a safe prsition. Immediate actions are i-also taken to restore spent- fuel pool boron concentration to j -. = 2000 ppm. -

If moving fuel assemblies while in MODE 5 or 6. LC0 3.0.3

-would not specify any action. If moving fuel assemblies-00 while in MODES 1, 2, 3 and 4. the fuel movement is

.; . independent of reactor operations. Therefore. inability to i

U suspend movement of fuel assemblies is not sufficient reason D to require a reactor shutdcwn.

.j SURVEILLANCE SR 3.7.15.1 I REQUIREMENTS .

.This SR verifies that the concentration of boron.in the-  :

s)ent fuel pool is = 2000 ppm. As long as this SR is met. i tie analyzed accidents are fully addressed. .

The 48 hour-frequency is uppropriate based on o>erating.

experience-and because significant changes-in-tle boron l concentration in the spent fuel pool are difficult to' produce without detection. considering the large volume of:

water contained in the spent fuel pool. An analysis has concluded.that a spent fuel pool boron dilution event-of sufficient magnitude to reduce boron concentration below the minimum non-accident requirement is not credible (Ref. 4).

i

~

(continued)

BRAIDWOOD'- UNITS 1 & 2' B 3.7 10/10/97 Revision B

' Spent Fuel Pool Boron Concentration

. B 3.7.15  :

BASES (continued)

REFERENCES 1. WCAP-14416-NP-A " Westinghouse Spent Fuel Rack i Criticality Analysis Methodology." Rev.-1. dated  !

November, 1996.  !

1 r

2. CAC-97-162 " Byron and Braidwood Spent Fuel Rack l Criticality Analysis Using Soluble Boron Credit."

dated May, 1997 -l

. 3. UFSAR ~Section 15.7.4. s

4. " Byron /Braidwood Spent Fuel Pool Oilution Analysis,"  ;

Rev. 3. d:ted June 17, 1997. '

5. Double contingency principle of ANSI N16.1 - 1975, as -

03 4

• specified in the April-14. 1978 NRC letter (Section 1.2) and implied in the proposed revision to ,

33 Regulatory Guide 1.13 (Section 1.4, Appendix ~A). '

6. ANSI /ANS 8.1 - 15B3 "American National Standard'for s

Nuclear Cri"icality Safety in Operations with Fissionable Materials Outside Reactors."

!' 7. Safety. Evaluation Report (SER) dated October 25, 1996, i ,

issued by the Office of Nuclear Reactor Regulation for 1

Topical Report WCAP-14416-NP-A " Westinghouse Spent  ;

Fuel Rack Criticality Analysis Methodology."

4 d

L 3

1 k

l? BRAIDWOOD.- UNITS'1 & 2 r,3.7 10/10/97 Revision B

_. .- _ _ ~. - _ _ . _ _ _ _ __ _. . . . _ . . _ . . _ _

Spent Fuel Assembly Storage B 3.7.16 B 3.7 PLANT SYSTEMS )

B 3.7.16 Spent Fuel Assembly Storage BASES f

BACKGROUND The spent fuel pool provides for storage of various Westinghouse Optimized Fuel Assembly (OFA) types of *

-different initial fuel enrichments and exposure histories in two distinct regions. There are 23 separate racks which provide placement locations for a total of 2870 new or used - 4 fuel assemblies, included in this are six specific storage '

locations in one of the racks 'or placement of failed fuel assemblies. These locations are identified as the failed fuel stora

" Region l'ge withcells. Of the 2319racks, the remaining racks four designated are designate as "P.egion 2" The anal tical methodology used to develop the  ;

CD.

criticalit analyses las been reviewed and approved by the NRC (Ref. ).

-k Region 1 racks contain 392 cells which are analyzed for '

storing Westinghouse 0FAs in an "All Cells" arrangement (that is, the criticality anal sis assumes that spent fuel assemblies reside in all avail ble cell locations, with the i

• exception of the boundary requirements). The stored fuel assemblies may contain an initial nominal enrichment'of s 4.7-weig t percent U-235 (without Integral Fuel Burnable-Absorbers IFBAs) installed) up to an initial nominal enrichment of s-3,0 weight percent U-235. provided that the requirement for e minimum number of 16 IFBAs is met (Ref, 2). The IFBAs are required to have. as a minimum, a  !

boron logding of 1.0X. equal to an amount of-1.5 mg B / inch. This is the minimum standard poison  :

material loading offered by Westinghouse for 17X17 0FAs.  ;

Region 2 racks contain 2472 cells which are also analyzed for storing Westinghouse 0FAs in a conbination of storage-configurations. These patterns are:

1) "All Cells

• Storage:

2) "3-out-of-4 Checkerboard" Storage: and

3) "2-out-of-4 Checkerboard" Storage.

(continued)

-BRAIDWOOD - UNITS 1 & 2. B 3.7-95 10/10/97 Revision B 1

_ . ~ _...,,_..___.m

, _ _ , , , , . , , . -,v.,- , y

Spent Fuel Assembly Storage -

t i B 3.7.X6 i

BASES BACKGROUND For the "All Cells storage' configuration, the stored fuel t

(continued) assemblies may contain an initial nominal enrichment of s 1.14 weight percent-U-235 (without taking credit for fuel i

burnup or radioactive decay of fuel constituents) u) to an initial nominal enrichnet of 5 b.0 weight percent lJ-235..

when fuel burnup and radioactive decay of fuel constituents are credited.

For the "3-out-of-4 Checkerboard" storage configuration -the ,

stored fuel assemblies may contain-an initial nominal enrichment of s 1.64 weight percent U-235 (without taking credit for . fuel burnup or radioactive decay of fuel constituents) u) to an initial nominal enrichment of s 5.0 weight percent )-235, when fuel burnup and radioactive decay of fuel constituents are credited, in this storage pattern. 1 there can be no more than three stored assemblies in any 2X2 matrix of cell lattices, D

L ->

For.the "2-out-of-4 Checkerboard" storage configuration, the stored fuel assemblies may contain an initial nominal

@ enrichment of s 4.10 v.ei credit for fael burnup) ght percent U-235 (without taking i

up to an initial nominal enrichment of s 5.0 weight percent U-235, when-iuel burnup is credited.

In this storage pattern, no two fuel assemblies may be stored " face adjacent" (that is, there must be an empty cell opposite each face of the fuel assembly).

The water in the spent fuel pool normally contains soluble-boron which results in large suberiticelity margins under actual operating conditions.

APPLICABLE- - NRC approved methodologies were used to develop the SAFETY ANALYSES criticality analyses (Ref. 1). The fuel handling accident-

-analyses are provided in Reference 3. The accident analyses for criticality and spent fuel pool dilution are provided in References 2 and 4. respectively.

(continued) i BRAIDWOOD-- UNITS 1 & 2 B 3.7-96 10/10/97 Revision B m

i e qp -mt >w wg--wm -e vr- ge wsu p v y,w, *' y F 4--==-ywrwwv- 9 - <qr't-- qu e eP-g-

Spent Fuel Assembly Storage B 3.7.16 BASES APPLICABLE The criticality analyses for the spent fuel assembly storage SAFETY ANAL (SIS racks confirm that k ,, remains < 1.0 (including (continued) uncertainties and td.erances) at a 95% probability with a 95% confidence level (95/95 basis), based on the accident condition of the pool being flooded with unbarated water, Thus, the design of both regions assumes the use of unborated water while maintaining stored fuel in a subcritical condition.

' However, the presence of soluble boron has been credited to provide adequate safety margin to maintain spent fuel assembly storage rack ke, s 0.95 (also on a 95/95 basis) for all postulateo accident scenarios involving dropped or misloaded fuel assemblies and loss of spent fuel pool temperature control. Crediting the presence of soluble boron for mitigation of these scenarios is acceptable based on applying the " double contingency principle" which states that there is no requirement to assume two unlikely.

9 independent, concurrent events to ensure protection against g a criticality accident (Refs. 5 and 6).

oc The accident analyses address the following five postulated teenarios:

1) fuel assembl drop on top of rack:

2) fuel assembi drop between rack modules:

3) fuel assembl drop between rack modules and spent fuel pool wa 1:

4) change in spent fuel pool water temperature: and

5) fuel assembly loaded contrary to placement restrictions.

Of these, only the last two have the capacity to increase reactivity beyond the analyzed condition.

Calculations were performed to determine the reactivity change caused by a change in spent fuel pool water temperature outside the normal range (50 - 160*F). -For the change in spent fuel pool water temperature accident, a temperature range of 32 - 240'F is considered. In all cases, additional reactivity margin is available to the 0.95 k,,, limit to allow for temperature accidents.

The temperature change accident can occur at any time during operation of the spent fuel pool.

(continueo)

BRAIDWOOD - UNITS 1 & 2 B 3.7-96a 10/10/97 Revision B

Spent Fuel Assembly Stora e B 3.7. 6 BASES APPLICABLE For the fuel assembly t load accident, calculations were SAFETY ANALYSIS performed to show the largest reactivity increase caused by (continued) a Westinghouse 17X17 0FA fuel assembly misplaced into a s'.orage cell for which the restrictions on location, enrichment, or bumup are not satisfied. The assembly misload accident can only occur during fuel handling operations in the spent fuel pool.

For the above postulated accident conditions, the double contingency principle can be a) plied. Specif*- sily, the 3resence of soluble boron in t1e spent fuel pooi water can 3e assumed as a realistic initial condition since not assuming its presence would be a second unlikely event.

Spent fuel pool soluble boron has been credited in the criticality safety analysis to offset storage rack and fuel assembly tolerances, calculational uncertainties, uncertainty associated with burnup credit and the reactivity

, g increase caused by postulated accident conditions.

5 Based on the above discussion should a spent fuel pool y water temperature change accident or a fuel assembly misload accident occur in the Region 1. Region 2. or failed fuel storage cells.'k will be maintained 5 to 0.95 due to the

{

presence of at l,e,a,st 550 ppm (no fuel handling) or 1650 ppm (during fuel handling) of soluble boron in the spent fuel i i

pool water.

A spent fuel pool-dilution analysis (Ref. 4) has been performed as required by Reference 7.

4 The analysis assumes an initial boron concentration of 2000 ppm. The dilution analysis concludes that an unplanned or inadvertent event that uould result in the dilution of the spent fuel pool boron concentration from 2000 ppm to 550 ppm (minimum non-accident boron concentration) is not credible.

(continued)

BRAIDWOOD - UNITS 1 & 2 B 3.7-96b 10/10/97 Revision B

-Spent Fuel Assembly Stora e-B 3.7. 6 BASES APPLICABLE Interface recu'rements have' been established to ensure k,,,

SAFETY ANALYSIS. is maintainec.within the appropriate limits. There are (continued) interface req'.lirements between Region 1 racks between-Region 1 and Region 2 racks, between Region 2 racks._and within racks between different checkerboard configurations.

These requirements are necessary to account for unique 1

' geometries and configurations which exist at- the' interfaces.

Interface requirements exist between adjacent racks to account-for the potential reactivity-increase in 3-out-of-4 and 2-out-of-4 storage configurations along the interface with non-aligned-racks.-

The configuration of fuel. assemblies in the spent fuel pool satisfies Criterion 2 of 10 CFR 50.36(c)(2)(11).

t LCO The restrictions on the placement of fuel. assemblies within the spent fuel, pool in accordance with the requirements in the accompanying LCO ensure that the k , of the spent fuel--

-pool will always remain < 1.0 assuming,,the pool is flooded with unborated water and's 0.95 assuming the presence of 550 ppm soluble boron in the pool.

tin LCO Figures 3.7.16-1 and 3.7.16-2.-the Acceptable-Curnup

-Domain lies on, above, and to the left of the decay time =

-line-applicable to-_the fuel assembly to be-stored. lThe decay. time for that assembly is measured from the time since the assembly was last-discharged.

In LC0 Figure 3.7.16-3. the Acceptable Burnup Domain ed the Unacceptable Burnup ocmain are separated by a single lit e because tecay time is not credited in the 2-oet-of-4 Checkerboard storage configuration. The Acceptable'Burnup

. Domaini l es on, above. and to the left of the-line.

APPLICABILITY This LC0 applies whenever fuel assemblies are stored in the spent fuel pool.

(continued)

BRAIDWOOD - UNITS 1 & 2 B 3.7-96c 10/10/97 Revision B

Spent Fuel Assembly Stora e B 3.7. 6 BASES (continued)

ACTIONS The ACTIONS have been modified by a Note indicating that LC0 3.0.3 does not apply.

Al When the configuration of fuel assemblies stored in the spent fuel pool is not in accordance with the requirements of the LCO. immediate action must be taken to make the necessary fuel assembly movement (s) to bring the configuration into compliance.

If moving fuel assemblies while in MODE 5 or 6. LCO 3.0.3 would not specify any action. If moving fuel assemblies while in MODES 1, 2. 3. and 4. the fuel movement is independent of reactor operations. Therefore. Inability to 00 suspend movement of fuel assemblies is not sufficient reason to require a reactor shutdown.

k SURVEILLANCE SR 3.7.16.1. SR 3.7.16,2 and SR 3.7.16.3 are performed REQUIREMENTS prior to storing the fuel assembly in the intended spent fuel pool storage location. These frequencies are appropriate because compliance with the SR ensures that the relationship between the fuel assembly and its storage location will meet the requirements of the LC0 and preserve the assumptions of the analyses.

SR 3-7.16.1 This SR verifies by administrative means that the initial nominal enrichment of the fuel assembly or a minimum number of 16 IFBAs is met to ensure that the assumptions of the safety analyses are preserved.

SR 3.7.16.2 This SR verifies by administrative means that the combination of initial enrichment, burnup, and decay time of the fuel assembly is within the Acceptable Burnup Domain of s

Figure 3.7.16-1, 3.7.16-2. or 3.7.16-3 for the intended storage configuration to ensure that the assumptions of the safety analyses are preserved.

(continued)

BRAIDWOOD - UNITS 1 & 2 B 3.7-96d 10/10/97 Revision B

Spent Fuel Assembly Stora e B 3.-7. 6 BASES' SURVEILLANCE SR 3.7.16.3 -

REQUIREMENTS (continued) This SR verifies by administrative means that the interface-requirements within and between adjacent racks are met to ensure that the assumptions of the safety analyses are-preserved.

REFERENCES 1. WCAP-14416-NP-A " Westinghouse Spent Fuel Rack

- Criticality Analysis Methodology " Rav.1. dated November, 1996.

2. CAC-97-162 " Byron and Braidwood Spent Fuel Rack Criticality Anal Mated May,1997.ysis Using Soluble Boron Credit." '

QD

3. -UFSAR. Section 15.7.4. 4 cy 4. " Byron /Braidwood Spent Fuel Pool Dilution Analysis."

Rev. 3. dated June 17, 1997.

5. 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).

6.

ANSI /ANS 8.1 - 1983 "American National Stancard for Nuclear Criticality Safety in Operations withL Fissionable Materials Outside Reactors."

7. Safety Evaluation Report (SER) dated October 25, 1996.

issued by the'0ffice of Nuclear-Reactor Regulation for Topical Report WCAP-14410-NP-A " Westinghouse Spent Fuel Rack Criticality Analysis Methodology."

m.

l l BRAIDWOOD - UNITS 1 & 2 B 3.7-96e 10/10/97 Revision B

BYRON C~~S VARKUPS i

_ _ _ _ = _-_ _ ___ - ________ _ _ ____________-_ _ _ -__ _ _ . . ..

l- 3 lC 3 7 IT e

@: e.7 PLANT sammesa. S.YSTE ~@_ MS 3.'7.14 2!'.2.!! WATER LEVEL / BORON CONCENTRATION - STORAGE P00L LIMITING CONDITION FOR OPERATION Lco a.7li

- .0.21 At least 23 feet of water shall be maintained over the-top of

. irradiated fuel assemblies seated in the storage racks. Th; di:::h:d 5:r::

-:th:2 - :-trertien ef the e:t: Y- '" th: :t:r:;; ;;;i :h:11 h reir.tein;d et 9..ete.

p 1 te = ;--

c moveu APPLICABILITY: Whenever irradiated fuel assemblitt areMn the storage pool.

M CO N D A -e . With the water level requirements of tha above h

satisfied, suspend all movement of fuel , assemblies ensamene_

cification not i

--- erie- et"r ess. r ir = :::ric: cr:ae :: r::::r; :: :: :r-

_l: :? i: ^ Ith'

. t: '.;'.t ;f '

th'O O ?2 714tmmectio& eld

y. with the Doro concentration r uirements of t above speTific ion

/ not satisfi , suspend all asent of fuel semblies and cra e (

1 operation with loads in-t fuel storage action oas and immedia1,af y take limit restore s soon the as possi dis.4 ved boron con ntration to wi in its-

-ACT # +: The provisions of Specification 3.0.3 are not applicable.

Ncte { t

. L SURVffLLANCE REOUIREMENTS d e23H olme the % 4 the irrodiated fuel oc embliet re. 4ed n. +he chroqe rocki 4.0.11 The water level in the storage pool shall be determined to be at least c

U_ M n: = int r;u=t re d::tt.)at least once per 7 days t...... .. . _ .. - . -.;

E;;.;.:,1t;; ;r; ;; it; ne; :terece seen '

-4.0.11.& r,erea ;;;;;atr;ti;n in th; :t:r:;; ;;;l :h:115: d:t:=

greet:r th;n er ;g;l t: 2000 ;; :t 1:::t :::: ;' '"' h;;r;.*in;d t: M t 7

' j 4hese-requir.;.;;t: : hell k in-effect-until December.31,-19E C BYRON . UNITS 1 & 2

/

3/4 9-13 AMENDMENTNO.p a~.e .

h 3.7 PLANT SYSTEM G mmmeasma- h Lco 3 7 15' GIO

m. is

!!' a. !! WATER LEVEL / BORON CONCENTRATION - STORAGE P0OL

-LIMITING CONDITION FOR OPERATION .

L4o 3.7.15 3.0. : At 1=:t !! f::t :f =t:r chl' S :1-t:i M :v:r th te; Of irrdi:tM ft:1 rr-ilie: :=tM te th et:r:;: r= h . The dissolved boron

. concentration of the wate the torage pool shall be maintained at greater thanorequalto2000ppeQ.

4 h APPLICABILITY: Whenever G Ga:= fuel assemblies are in the storage pool.

Ellati:

/ With the w er level requi - nts of the satisfied suspend all ve specifica on not enent of fuel ssemblies and ane i r

' operati s with loads.i the fuel stor e areas and r tore the water level within its lidit within 4 ho s.

RA A.) b. - With the boron concentration' requirements of the above specification not satisfied, suspend all movement of fuel assemblies c: : :r :t (

)

aret;=; witt. in ; ta :t.; Tue; ;terese ereesland i diate1y t'ake

' R A A ?- action limit as to soonrestore the diss ved boron concentrat n o within its as possible.

g ACTIONS e. The provisions of Specification 3.0.3 are not appl cable.

Ncde.

{f 4 -

e SURVEILLANCE REOUIREMENTS t.0.11 The a:ter 1 =;l in the-st:r:;: ;::1 chli h d:t:=ind t: h :t le::t its 2.ir. ira r;;eir;d d;;th t in:t :::: ;;r ' d:y: d= f rrdittM f=1 n:;11i;; :.r; it, th f=1 :t:r:; ;nl .

. M. 4.0.11.e Boron concentration in the storage pool shall be determined to be ill5dgreater than or equal to 2000 ppe at least- once per hours. 4 (j h

'Th:: requ+rement+-shall h in-effect-until December.31,-19N C BYRON - UNITS 1 & 2 3/4 9-13

(

AMENDMENTNO.p Rev.B i

- ~

~ 16 5FUEESTORAGE -

ho Lci 3 ~7. IG CRITTCDT{ A M' c

edion 4 0 -

e.

(

r 6.1.1 ThespentfuelstorageracksaredMgnedandsha11'bmaintainediiftY a less than or equal to 0.95 when flooded with unborated water , which / I inc , es a conservative allowance for uncertainties as described in Section'9.1 of the FSAR. This is ensured byy controlling fuel assembly placement in each region a folbws: ,- '

> .. c

/

a. REGION 1

/- N l 1. I' nominal 10.32inchnorth-southand10.42incheast-west, I

- center-to-center distanca is maintained between fuel assemblies placed -in the spent fuel storage racks. ,.J (y

) 2. fuel assemblies may be stored in this r[gion with i -

N '

' a) a maximum nominal initial U-235 enrichment of less than or equal, to 4.2 weight percent; or

/ -

. . . d b) a maximum nominal initial U-235 enrichment of 5.0 weight

)

/ .

percent with sufficient Integral Fuel Burnable Absorbers A

i 4 present in each fuef essembly such that the maximum '

f' reference fuel assemb1'y .k is less than or equal to 1.470  !

at 68 r. Nr /

,t . . / ;3 N

f' ',[

M b. REGION 2 '

6,

\

l

(. l

/

)' 1. A nominal /d3 inch ' center-to-cente 1 stance is maintained \ D.

between fuel assemblies placed 1.n.the spent fuel storage racks. i t; '

/ y:cv

. ; M \. p l

11 2. a) duel assaicblies may be stored in this region with a maximum Cr[

/'/ nominal no burnupinitial U-235 enrichment of 1.6 weight percent 9)",

and up to 5.0 weight discharge burnup as specified in Figure 5.6-1;:.cr ~

N rcent U-235 with a minimum

/, ' . -

'. . .;g,T.y q '

N ,

• b) '

Fuel assemblies with a maximum nominal initial U-215 .

enrichment.of greater than 1.6 and 'less than or equals to

. J.

4.2' weight percent that do'not' meet the minimum burnupN

'\ "

specified in Figure' 5.6-1, shall be" loaded in a \' /

a checkerboard pat, tern for storage"in this region.+ AA 5.6/1.2 spent fuel The storagM k new racks shall W n eDstTore JoMtored d/jri~'tlie n assumed. exceed 0.98w)enaqueousfoan'moderationis y

' .J DRAINAGE ,

5 pr.6.2 Th spent fuel sto age pool is det gned and shall maintained o

^ 1 y- event Miadvertent draiding of the poor below elevati

' ,. ~,.

423 feet 2 i hes.

- l

\ ' AGetretl in %x:hn 4.O <

l.NSc7. T Aso g^3 --

p. m 2, 'co7 the spent fuel storage racks a=" h h uoneo N with a K.,, of less than or ~""2 0." ooded with water containing i

a =w r .,RGev ppm #so:rvble boron. e a L

v f O Q- , . . . ,

BYRON - UNITS 1 & 2 5-5 Amendment No. g Flev. 8

_ _ _ __ _- - nnsutr- 90 5 " 2 22 " '5-*8-" "

t o 3 . ,' 3 5 P 86 Ao se,h o n< . 40 M.

W ~ ' - " -

5.5.1.1 De spent fuel storage racks are designed and shall be matotained with:

-:- Fuel' assemblies beving beinhial U-2 f 4r-

/ '

A L < 1.0 if fully ooded with enrichmant of 5.0 ght percent,N' gg_ '

uncertainties as ' bod in WCAP-1 water, whichinclud an allowance for 16-NP A,"We mwn Rack Criti Analysis with C or Soluble Boron,@" Spent Fuel

. See Doc ' 'on 1. No 1996'-

'DD '

-e- A L 0.95 if fbily flo water borated t 50 ppm, whichinclu es an ce for u s

as describedin W -14416-NP-A, "Weninghouse FuctRack Analysis with Cr 'N ovember 1996; forSolubleBoro # Revision 1, \

d A nominal 10. inch nonh-south 10.42 inch east- center-to-contar distance fuel assesnbliespin the Regiony ek

\ I New or spent assemblies whh sufficient Integral Fuel Burnable Absorbers prese '

tcosa.it .q in each fbel assembly, as desenbed in the " Byron and Braidwood Spent Fuel Rack Criticality Analysis Using Soluble Baron Credit."147.

~17. C ^_C ?? ! /

! which may be allowed unrestricted storage in the Region 1 racks;

-f.-

A nonunal 9.0 Inch camer-to cent A eane* bE e the Region fanamhlies placed in \

k g-

{

s New or spent fuel assemblies with a combination of discharge burnup, initial t

/ enrichene, and decay time in the acceptable region of Figures 5.6-1, 5.6-2, or uce 7 g,4 5.6-3, as applicable, which may be stored in the Region 2 racks in the applicable j checkerboard configuration, as described in the " Byron and Braidwood Spent Fuel j Rack Criticality Analysis Using Sohable Boron Credit," %.00 . CAC-;~-le.; j A

Interface requirements wnhin and betwocn adjacent racks as described in the

/ LC6 M Ik C " Byron and Braidwood Spent Fuel Rack Criticality Analysis Using Soluble Boron

' Credit," L 199^. CAC-9 -161.

l 0

x V

c1ncen c-ca .

ReJ.B

4 CTS INSERT (S)

SECTION 3,7 LCO 3.7.16 INSERT 5 56l (Mn )

LCO 3.7.16 Each s shall: pent fuel assembly stored in the spent fuel pool APPLICABILITY: Whenever fuel assemblies are stored in the spent fuel pool, f

ACTIONS-f -----

NOTE- ----

l.C0 3.0.3 is not applicable.

g E . .

CONDITION REQUIRED ACTION COMPLETION TIME A. Requirements of the A.1- Initiate action to LC0 not met, Immediately-move the noncomplying fuel asserably into a location unich restores compliance, 4

10/10/97 Revision B

CTS INSERT (S)

SECTION 3.7 LC0 3.7.16 INSERT 5 5A.- (continued) (Hn )

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY' SR-3.7.16.1 Verify b Prior-to nominalenrichment y administrative of themeans the initial fuel assembly is storing-the s 4.7 weight percent U-235 or.a minimum fuel assembly number of.IFBAs is met. in Region 1

• SR 3.7;16.2 Verify by administrative means the

.T Prior to-

)

-combination of initial enrichment. burnup, storing the

-and decay time of the fuel assembly is- fuel assembly 4- within the Acceptable Burnup Domain of- in Region 2 Figure 3.7.16-1. 3.7.16-2. or 3.7.16-3.

'SR 3.7.16.3 Verify-by administrative means the Prior to-interface requirements within and between storing the adjacent racks are met, fuel assembly in the spent fuel-pool 10/10/97 Revision B

DESIGN FEATURES .

( 5.6 rUEL STORAGE (continued) -

dPACITY

/

5.6.3 The spe fuel storage p is designed an shall be maintain with a '

storage capydty limited to nfmore than 2870 fye assemblies.

d.7 COMPONENTCYfdCORTRANSIENTLdIT -;

5.7.1 The c >onents identifi Table 5.7-1 a designed and shal maintained it tin the cyclic /r transient limit of Table 5.7-1.

Ackhecced in reebn 4.3 c d f

• c5 'O SccIlon 5.6

'_see Docs in sechon to ae OZs.Ers di60 5.0 s

. r s

._ s t

( ts ..

,. . . < [.

! - ~

Q$f s

..s .-d. .

, ,r.--- 4 -

+.  ; . v.v- .'

a , i a e?.: .. .g: w.

• a-

.,  : n,w--l ,.-

r.. . af*'

,. ^ ,

. ;p

';- .l U.. ,: .--  :: '.:  ;'. . .

, : ' jk: : . .

.- .,.- -. .. .m. ,,a . .

w ,

. a.

. ., .7. .

. g, i . ..

• ,.*3' ? 's ,.j

• % 6'

'"('-

~., , .g ee ....,

? .

9 k

(- BYh0N - UNITS 1 & 2 5-Sa bndment No. 86 Rev.B

6/1 m%;&c s______ c_- i .

c- a amo c 3 l- ,

r.)j

., QQ ,

f- 50,000 ,

4 ,

( -i l l l

l. i a i i

• I l

l 4e 66 i

(

l

$ $ 1 0*$

6 e i e isa 45.0 -

Snriemmen.: summe i i ,

4 ,

li 6 i e ye e M/ \ ., * (wr ol (MWD /Mrcl

, , . , , f. . .

" 1.60 OLL 1 6 , y, e e i

[ 1.80 4,635 M '

I V' ' 'i

' 'i 40'000 ,2.00 8.565 M '

t/t i

. ' - 2.20

/ 6 ' '

11.845 E '

' ' i ' ' ' < i

.- 2.40 14,729 E Accepta.de U ' '

2.50 17,397 -

Region * .,I '

2 G 7 g/

k 35,000 - - 3,co

• 2. 0 \ 20.085 22,742 ""M r; r ,- '

3.20 \

g 25,132 m g ~~

3.40 \27,310M #

/ -

/

< e * *

._ 3.60 MO,179 E I ' '.

g 30,000 --- _ 3.80 32.651 " '

4.00 35,047 m

f f '

U --

4.20 f

"- 37,389 m j ,

E 4.40 39,655 m /

, i e 25'000 .[ 4.40 . 42,024 m r '<

en . 4.80 44,290 W '

/ /

i i e i <

$ ._ 5.00 46,442 M /

g ._ '

f_/_/ 7 __________

g 20,000 , ,, ,

/ 4 /

Dnacceptade s x' _ won

/

f

/' , 15,000

/

3

,, [ '

V ,' ,

l

(/

i ,

/i N

/

' s

-w 6 i fi f' ' 4 i

/! e s i N 1 i .

l J / '

l s / ' \' '

i ,

4 ,

/

5.000 s

\

s 1l 6 a

s i '

~.

/e e i  %

_ /! l I 't f i

\ )

t s /

0 I '

.' ' ' , .' .' . ' 6

.. .' , ... .4 4 , . . ,

1.60 2.00 2.40 2.30 3.20 W

3.60 4.00 j 4.

9

' 4.80 5.20 x Fuel Assembly Initial U-235 Enrichment /0)

\

t *

. ..g j No 's: j The use of linear interpolation between the mM== ups reported above is acceptable.

/

s FIGURE 5.6 - 1 '

\

i l

~\ MmT*M BURNUP VERS"JS INITIAL ENRICHMENT /

FOR REdION 2 STORAGE BYRON - UNITS 1 5 ent

[

^* Rgd re .3, 7 ,1(. - t -

f INSERT C-1 80000 .'

Mt ew, 55000 , e s re r

/

e , sz e r 00C '

', le sone.

s ., ,, ne s

- ~ r e e r no s

--- . ACCEPTABLE

. 4500, 7 s

s 7

1 e

7 r,s x > r , ,,

% 1 1 1 s.-

, 2. i ,e , s

' > e <<

~ 4000C x .

r , ,,,

7 1

s ,s, en . x > r , ,,

a x

.' ,'es77 3500C 8 x

i ,,

1. # #,ri

. ,, ; ,. N V V f A f,V

% f

• f f Ff x, '

',y v

,ll' 30000 -

r -

... ' F

,F > ff 5 / JJ f I

2500C

~

v 's

< r,, , ,

1%

a- E .

~ i ~ es i

• i / /z' ,

e.4

/ /}// es

. b I 20000 J e< esr

> .rJp' s

g

• # IIr q IJ IE g

> 1a ' =

e i e- <

15000-- />i-s

~

r is,

. . 8 ,

2 11, Ia ff NOT CEPTABLE

',; r 20000 -

., . , x ,- r

. g

.e g, '

c. , '

5000 F \

L ~

j \

.1 '

y.

-g_ #1 s ,

x 1.0 2.0 *

- : 3.0 .e 4.0 5.0 ..

2nitial U-235 Enriciument (w/o Note:

The use oflinear interpolation between the minimum burnu is acc.eptable.

,t .. '.*

~

FIGURE 5.6-1 BURNUP VERSUS ENRICHMENTFORREGIO C ALL CELL CONFIGURATION STORAGE ,

4

/

/

Rev B sycs 30

. - -- A

CTS INSERT (S)

SECTION 3.7

< LC0 3.7.16 INSERT 5 5bA (Ag )

60000 ,

_3 i i DECA).

i T151E:

55000 ,

j O YEARS

.f

. /_

i i , ,

i t

4

/

- e if

2 5 YEARS.

50000 ,

/ / .

10 YEARS ACCEPTABLE 15 gars 1

DURNUP DOh1 AIN -

, , l f . 'i f ' f) 20 iEARS 1

4b000 s

, 4 .

f f: f f//

l

, f: /+ // ~

t i .

. i e f ff ,

i , i ; , f

! i 4

e f fe i i .

! t / f f // i e

! i i  !  !/i / / ff

, ; . . - , ,  ? I 40000 ,

, i f :f

_ / / /

f ff 4 ,

ff g  ; i i .

e

, i i

e if v ff!

! / / / />r i i i

3 ,

r  ;

if i f ;f f f i

I

> g 35000 ,

/ / / //

O . i ,

f f , f ff , , ,

k #

t i

, i i i /

f f e ff !

/ / /f t i i i i

> i 1

t i i .f / ffe i  ! i i i 30000 l ,

t i i i /, / , / //4 i '

D o

'I / /h i i  !

2Z ~5000 '

, ,I,/

, fif /// ,

a

/ / /// i '

i I  ! i i

f

- 20000 i e

= ,ffM",

i i

// ff.

/ j Q .

f s

i t i

s 15000 " '

< .,/  ;  ;-

< i

'jy 1 . i e  : ,

t

H  ;

;l P 10000

-- i./ '

UNACCEPTAbLd

-[ .

BURNUP DOM AIN --

5000 -

, i , . 4

-t , i

, i

. i .i i -

O ' -

1.O 2.0 3.0 4.0 5.0-INITIAL U-235 ENRICIlh!ENT (w/o)

Figure 3.7.16-1 (paga 1 of 1)

Region 2 All Cell Configuration Burnup Credit Requirements 10/10/97 Revision B I

Replocc w+ h Insert' E-5b A

~

h' figure 3/7& -2 .-

g INSERT C '

45000

/

/

d meeer 40000 / m

, , e sense

/ /

' / J 2

^ 35 0 ' # # * * ' "

I _..

\

^

\ ,

e /

/

/ / / as samme

/ /aff 39 3ense

) J />ff gygg ) / f///

- E3000*"

w

\

? \

/ fJ//

/ ) //)F t

g ggqqc 4.

\

\

\

y

~

/

/

/ /. f/f/

) / f4

/ M '/

'sr

.> /,en

\ / /,F/>V

/ J ' f, V//

\ > /, %f 2000e -

.o

\ /, % '

g <

/ Vfa r

\

( (~ 3

/ / fag y

[. 8 (1, v  :.

j1500c

/

X ['

jlFM \

7.h 3.

/ 4,lr \

• _, v Ja s 1000C 7

\

, 2 x ,

" _. / e s 1

"'n *

[

'.. 500C NOT ACCMTABLE '-

I \

/ f r T

/

J .

\

/ / N

' \

O __ II '

\ '

1.0/ -2.0 3.0 .

a;4.0 5.0 '

2nitialU-235 Enrichment)(w/o "4 Note.

% ., .e. .

The use oflinear interpolation ber;.a the minimum s is acceptable. .

-. . . ro :': - -

'l '

, 4 ;, ,  ?.

- FIGURE 5.6-2 ,

'.1 .

. .,.. , a:

/

MINIMUM BURNUP VERSUS INITIAL ENRICHMENT FOR REGIO 2 1 3-OUT-OF-4 CHECKERBOARD CONFIGURATION j 5 g.

.is ey c s-sh)

CTS INSERT (S)-

SECTION 3.7 LCO 3.7.16 INSERT 5.FtA (continued) (Ag )

45000 ,

, i . , i 4 , i i l

ll l_  ;- , ll '

l l

, DECAY 40000 i , i . 4 TIME:

l ,

, ,  ;'  ;; , l j 0 YEARS i . -  !  !

1/

4 /, 5 YEARS

^ i , i ,

i i  ; i i i i i ;f if 10 YEARS 335000 ,,,, , , - , , , , , , , , / //

15 YEARS i

E

> ACCEPTABLE

--BURNUP DOMAIN - l lll ij, f)Cff f /ffe 20 YEARS g 30000

.iii i i e i i i i /f.u/7-e = '+I~~ '

, f f fff, i-

= l l' jfW{

/

> D i i - eure !

l g z~ 25000

< i

, , , , , f fffj I/ ffr/ I i ii i if /ff/

! O io i ' 'T ~ ! ii i i

(- O '

i ///// i I i a

! i ii I///// I l M 20000

=

=

l

, . i i ,

'l N [,

// -

, ', l' i , , -

i i ~//f b '

i l l ;,f f Fl,,,; l

' 3 t I

$l15000 l l.

c. , i T i i i e t i i  !

'. ,- , , , , ,/ <- , ,! i g ,

ee i / . i i i 4

i g 10000 '

lfj# l ' l l l,,',

UNACCEPTABLE '

[4li,:

l '

BURNUP- DOMAIN t-5000 ;l L_ 4

! t : i i ll[lL_ .

i t

! ! l O

l.0 2.0 3.0 4.0 5.0 INITIAL U-235 ENRICHMENT (w/o)

Figure 3.7.16-2 (page 1 of 1)

Region 2 3-out-of-4 Checkerboard Configuration Burnup Credit Requirements 10/10/97 Revision B

, ,_ g- ,, . ..

-h Figure. a . 7, l(,- 3 .

~

r. INSERT C-3 5000 ,

/

/ )

j j

,,, / /

/ /

l ,

f g \ / /

- \ ACCEPTABLE / /

390c N / / '

\ / /

$ \ '

\

l j k N l /

j- l

\

\ /

/ /

. 2 acoe -- .' f - -

.= - \/ /

3 /\ /

/ h(

/ /N < -

/

100c / ' \ NOTACCEPTABLE s

.- /  ! \ M

/ / '

\ [.

y i.

/ -

_ / .

~

. / / N o / / \

0 4.2 4.4 4.6 4.u. 3 . 0 ,.,

Zaitial u+

U-235 Enrichment (w

)

Note-The use oflinear interpolation beraw the mini:num burnup is acceptable.

FIGUR, E 5.6-3 hENIMUM BURNUP VERSUS INITIAL ENRICHMENT FOR REGIO 2-OUT-OF-4 CHECKERBOARD CONFIGURATION

.6 ,

By(5-sb) Rev.B

CTS INSERT (S)

SECTION 3.7

.- LCO 3.7.16 INSERT 5 5bA (continded) (Ao)s 5000 , i i ; i i 4  ; i 1 i i i i  ! > l

) i ! !

j-l ! 1 i ) i I !l l l  !  ! t I l  !  ; ! i i I I 1l I I l l  ! l ji iiij i l ! I IIi ii 4

i )  ! l ! '! l l i l} i  !

Ii[

i i ii i ; ii ii!I i l If i G } } ! ACCE'PTABLE ' i I II e j j i BURNUP DOh1Alh' l

/

$ l II; I! I  ! l /

a E

> 0 {

i II I I Il l l l/l l Il

$ S i i i ! Ii i i IA I I i

x l

li i i i! l  ! / I x i 1

! i l II I l f

/

E I I!!

I i i I l /

E!2000 i l !

' i j

I l ll I / l l l t  ; l ;  ;  ;  ; ;y  !  ;

@ i !!  !  ! l! i yt I  !

1. ! l t

! i i  ! / I  ! i d i !  !  !  !!: i /l i 1000 I !I UNACCEPTABLN

, i

',fl j l BURNUP DOh1AIN l ll 1

t i ! !i !l!

i  !

! t l l i

/  !  ; I  ! i j l I

g. / 4 ,

! i i !  ! i i i i i I 4.0 4.2 4.4 4.6 4.8 5.0 4

INITIAL U-235 ENRICll:.1ENT (w/o)

Figure 3.7.16-3 (page 1 of 1)

Region 2 2-our.-of-4 Checkerboard Configuration Burnup Credit Requirements l

10/10/97 Revision B e

e BND CTS FARKU3S t

x:

e

< ~ A

. . . . . .1 .

3.TPLANT SYSTEMS ^E 045 [ N ~ 3.7. H t!' a.!! WATER LEVEL!!O"^" C^"CEF""!S - STORAGE POOL f s

LIMITING CONDITION FOR OPERATION L(o 37I.M

!.a.11- At least -23 feet of water shall be maintaineh over the top of irradiated fuel assemblies seated in the storage recks. "- "--'"-' '----

=:.;=tr:ti= Of the e ter i- ika e k a.

ma.

= =, . m 1- en SAA,A nne a ,

ite t; pet! th!' 50 --!-tri :d :t ;rnt:r c ' l' C.-

[d L:

3 APPLICABILITY:

Whenever irradiated fuel assamblies arehn t.he storage pool.

AG110H: .

. IN coND A -4s With the water level requirements of the above spot fication not l ratisfied. susoend all movement of fuel ass 1es e- cr=;T car.

. tis: .; inh

.. .m ' :da

_m_ '.- th:,_ m . _ =,_1 ,m---,

f  :::r:= :rn: W =:tr: th: ;;ts:

Lo S

~'

/

(minedsately L20 t.

t ith th br= :::=:tr:ti= r: ;f th ; hve e

-;pr;ti n-t =t!:fi-d.  !"r; -i !!' -^r::w::ir: fr.t:t Of f=1 =:--ilt= =gcificette,,

=; .;tth ! =d: d;r=;

f=1 :t:= =:= =d tmdi;tely teh k'

.d. u.; . .. te re;; =; th u.

din.in

' th:. ..__ .. -_ _ _te.eitMo-Me-u. .Utd h : b, /:=;:=:t= tie d)

" ACTION 5 g'>

Nofe The provisions of Specification 3.0.3 are not applicable.

t sR 23. l A. lSURVE I L L ANCE REOUIREMENTS 4 a 23 f+ obove the be d ihe trrodscued fuel assen Abec, seated m the thrag

-its ea"i-=d -i-!="- dapth at least once per 7 days " :4.9.11- ' The water , level in

=t:-ilie ire i- the f=1 ste s;e ;=1.

i =ditted f=1 A4 t.:.::..

gr=ter th: er : =1 te 2^^0 ;; :t in:t == p;r " h.,2.nr= =r.=:t=ti= t m

in tu :t=:;; pa

.. c

, g5 .

e anit4 Pamant t Chall ha 4n a f fernt unt41 hage=har ${, } i BRAIDWOOD - UNITS 1 & 2 3/4 9-13

- AMENDMENT NO.

Rev.B (b, . r I wuer dai en Tun e 30,1117

3. 7 PLANT SYSTEMS An L.C 3 7.15 RMBEMM-ePHAHeNS '

3.7. I 5

!!t ".11- "' tea LE"Et!"0RON CONCENTRATION - STORAGE P0OL #

1 s

LIMITING' CONDITION FOR OPERATION LCo 3:1.\5 .

- ,a ,, .. ,.... .. ,__. .,....__ _m ,, .

...d;eted '.;'

< _ _ . , _ . . , _ _ > _ . _ _ g y,; ; g

..n :dlir inted '- the sters eMe . The-dissolved boron concentration of the water,,ipe h t,h,s. storage pool shall be maintained at greater than or equal to 2000 ppe. (

Mi c _

h APPLICABILITY: Whenever "

O; ;Cfuel assemblies are in the storage pool.

..r . - .

AGIlQti: cf ..;  ;.

. .;r; w n ~ N. -

. "*th th: n ter level" equi--- t: ef the de:- :;=if* nti= =t nthf*ed. -":;rd el' ev- ::t ef feel re:-in rd rn: -

'd.

i..?^$$225$

... .. ........ ... }23E. . 95. 3.d..f...IN-F'i'.*1'

, . . . . . 'i" 'i'd " " ".

b.

RA A.I With the boron conce'ntration requirements of the above specification not satisfied, suspend all movement of fuel assemblies c= ar.c -

r::::.: ::= :rr :: =;. =:

Rh.A.2

' ~

action to restore the diss Wed be-nrconcentrati r.

e and immediat ly take k' l

' limit as soon as possible. withi its- /

^(Tio NE rn Nok. The provisions of Specification 3.0.3 are not applicable.  % CD 4-- 0 l t SURVfittANCE REOUIREMENTS-4.".11 Ti.e eler h.el u tM -tr:;; ;;.1 :hl'. h d;te.7;;.ed te h et heet its eiet-- --ae!--d d-ath at 1-est once ;^r ' hy: 9.= 'rr:duted '.el

=ss==h!!es re i r the fe:1 :ter::; ;nl.

La 3.7. is.1 4.a greater than or equal to 2000 ppe at leasthours. once per.11.e- Boron concent t

v.

{7c 3 %j .-

b ans_t 4 eament t thall ha 4n affert sent41 haga h e ${, }^ /

w-

.' Rev.B

)

((' l t . ' < (( .'f(ghk [LilIC 3 $Y 1

~ _ _ _ _ . - -

5.6 FUEL STORAGE - -- ---- Moro n c redit in the d g 3,7 4 h CRITICALITY- '

p 5.6.1.1' k The spent fuel storage rahs art desi ned aMhall bv maintainTd wit / .,

' less than or equal to 0.95 when flooded ith unborated water, which i of Yt! des a conservative allowance for uncerta e UFSAR. This is ensured by controlling fisel assembly placement in regio as follows:- -

ac

.1

\>

a. REGION 1

/

1.

• A nominal 10.32 inch north-south and 10.42 inch east-west,-

center-to-center laced in the spent distance is maintained fuel storage racks.betweerufuel assemblies .

2. Fue assemblies may be stored in this n with a) a equaximum to 4.2nominal initial U-235 nrichment of less than or l

weight percent, <

-: . s, &

b) a maxi . nominal initial -235 enrichment of 5.0 weight l

I percent w h sufficient ntegral Fuel Burnable Absorbers present in ch fuel sembly such that the maximum reference fue asse y ke is less than or equal to 1.470

\ at 68'F.

// A h

b. REGION 2 j

j l. A nominal g.03 rich center- /

between fuel ssemblies place -center distance is maintained in the spent fuel storage racks.

'I g 2.- a) Fu assemblies may be' store in this region with a caximum )- i inal initial U-235 enrichme of 1.6 wei ht percent with no burnup and up to 5.0 weight d discharge burnup as specified in cent U-23 with a minimum ure 5.6-1, or /

b) Fuel assemblies with a maximum nominal

./. '

tal U-235 i enrichment of greater than 1.6 and less tha or equal to

\ 4.2 weight percent that do not meet the mini urnup,

' specified in Figure 5.6-1, shall be loaded in a checker pattern for rage in this ion.

, 5.4.1X %e k,, rm nww' fuel r te G 6t te TcMin st red de -

1 pentfuelstorgeracksshal

< assumed.

not exceed 0. 8 when aque;ou -foam moderat on is g DRAINAGE, -

t j ,. M .2 T spent fuel s orage pool is esigned and all be mainta ed to y , p-rev.R nadvertent dr ining of the ol below ele tion 423 feet inches. j ,

n .i l' ~

\ .

~ u ~ m r r m ,

[w.~ Unt h :MU99Ithe spent fiel storace nb 2:11 W=i ,teihed D a K ,,afof?Cless tnagy --pie -

with a minimo= 7 ppa soluble boron.Qj flooded with ,

water containing BR'A[DWOO 6 UNIT 5'1 C W' '

knefrdme I Mo. [

Can El Mer dam Tune SO MT7

___a

b t..cc E7 l G g

.Gec+ ton 9 0 cs mre r 3 x

tco 3,7,16 L

\

l M 1 The spent fuel storage racks are designed and shall be maintained with:

\

Y -e-Fu 4ssemblieshaving maximum initial %5 enrichment of/0 weight percent-re: red m k.a < 1.0 if fully ded with unbo al water, which ' udes an allowanc or ghggg. hc uncertainties as scribed in WCAP- 4416-NP-A,"W ghouse Spent F -

sec w n ,o Rack Criticali Analysis with C t for Soluble Bo " Revision 1, No mber

• i 1996; ,y /

- : e.:. ,. /

• A k.a5 .95 iffully floode 'th water borat o 550 ppm, whic cludes an '

allow ce for uncertainti as described in W AP-14416-NP-A Westinghouse S Fuel Rack Criti ty Analysis with 't for Soluble ' ron," Revision 1, vember 1996; .

.  :. t .

X. ?E,.s;i , ' ...f.6. . ,

2.'..T.l'Es..tr.64 '

4.- A nominal 10.32 ch north-south

/

10.42 inch east-

~

center-to-cent distance betw fuel assemblies aced in the Regio Iwh;

.:;w. . -:.n,; . '

-e. New or spent assemblies with sufficient Integral Fuel Burnable Absorbers present W 3 ~M 6. O in each fuel assembly, as described in the " Byron and Braidwood Spent Fuel Rack Criticality Analysis Using Soluble BoroiCredit,"e : y 10;7 CAC-;7-162 which [

may be allowed unrestricted storage idtlie Regio ~n' t racks;1.;; . .

l 4 * ' "J<W@AG.70$,9'ebf l*

l

/ f- /A nominal 9.03 in center-to-cente- ce between assemblies plac

[.

/ ' the Region 2  ;

4:&!GF*3NIN W .M ~-

/

! co g,7, f 4,g+

" +14:@.W2QMF

j. t New or spent fuel assemblies awith ' ginbmationif%e burnup, initial i ,

enrichment, and decay time in the acceptable region of Figures 5.6-1, 5.6-2, or i

5.6-3, as applicable, which may beltoie'd in'the RIgloE2 racks in the applicable checkerboard configuration, as desc'ritiedin thei Rack Criticality Analysis Using Soluble e N h"- BoidiCr'yiEn'Tdd Brsidwood S

7 C AC-07-164

.!LCO .,. 7.16. c

.. - . . w;e.w u w. '. '

i .,

-h- Interface requirements within and between adjacent rack. , 2.--s as described in the

" Byron and Braidwood Spent Fuel Rack Criticality Analysis Using Soluble Boron Credit,"c. ay :007 CAc.;7-162 Cf x ,

..n .a. .:,u. . - s.

,/ 4e.9 . .:

- '".39 A;I Gi EC w, W .,.

5 4 d $.. 9 vy i cN.w.~,n.d. v+ ;;rm Op.'

o,,m 3- u  ; 2, ' w,e. -

et * *. * ' * , * '

mn

-::. 3 % g >.. % ,,.{,.,% Q '

m. . .g%

w - ... . 9. w:.} y .-- .

.. 2

-b.$ ,.

s , . ...

jV s ,. . . .

K A T.

Con.EA lefter cinied I'etAen .Le t 25 , i W7 gev. 8 w

CTS INSERT (S)

SECTION 3.7 LCO 3.7.16 INSERT 5 5A (Mu )

LC0- 3.7.16 Each spent fuel assembly-stored in the spent fuel pool shall:

i-APPLICABILITY: Whenever fuel assemblies uc- stored in the spent fuel pool.

ACTIONS-S NOTE

, LC0 3.0.3 is not applicable.

1 CONDITION REQUIRED ACTION COMPLETION TIME A. Requirements of the. A.1 Initiate action to - Imediately -

LCO not met. -move the noncomplying fuel assembly into a location which-restores compliance.

i 10/10/97 Revision B l

~ . _ . . . _-_-_-.._.. -

1 4

4- CTS INSERT (S)

SECTION-3.7-LCO 3.7.16 5

INSERT 5 5A - (continued)- (Mn )

i SURVEILLANCE REOUIREMENTS

• F SURVEILLANCE _ FREQUENCY

'SR 7.16.1

. Verify by administrative meane the initial Prior to L nominal enrichment of the fuei- assembly-is storing the 5 4.7-weight percent U-235 or a minimum- fuel' assembly

number of IFBAs is met, in Region 1 i

!- ;g' .SR 3.7.16.2 Verify by administrative means the Prior to n ,

combination of initial enrichment. burnup, storing the and decay time of the fuel assembly is fuel assembly-p $[L within the Acceptable Burnup Domain of in Region-2 Figure 3.7.16-1. 3.7.16-2. or 3.7.16-3.

1 p'

SR 3.7.16.3- Verify-by administrative means the Prior to

• interface requirements within and between storing the-adjacent racks are met, fuel-assembly 4

' in the spent Lfuel pool 1

i i

4 i

L i

i 10/10/C7 Revision B 4

/

t

-DESIGN FEATURES 5.6 FUEL. STORAGE (continued) -

CA/ACITY

/ ~

M.6.3 The spe t fuel storag pool is desig' d and shall b maintained w ha storage capa ty limited to o more than 2 0 fuel asse ' ies.

~

5.[COMPONENTCY[LICORTRANSI5dTLIMIT 7.1 The com ents identif ed in Table 5 -1 are design d and shall b maintained w lin the cycli or transient mits of Tabl 5.7-1.

(0 .,

Addressed in Sethon 9.3

/dlr.cett in Sec+ ton 5.0 See ocG 6r sechen 5.o s

s BRAIDWOOD - UNITS 1 & 2 5-5 a Amendment No.78 Rev.8

_ _- ______ -_- ------- r n-- c- r ~

(

~

)y l ( \~ 'f )0 'b]

50.000 -Q s

A A A A s i . t t _i ,

t t f *f f .

i 6 6 . i . .

- e i i 4 ,

! . t ie i6 .e v 4 i 45 000 - amrasament aurae , .

e a  ! . I,

=

twiel tiedDntTU)

, . . , ,e j

. . y f.

1.60 f "- O t1_ # , . .

• 6 /

g go . 4,g3$ m i 6 e 4 +' t/ ,

40.000 2.00 2.20 0.555 m 11.545

, , ' ;, (* .,,

f -

h 2.40 14.729 .Acceptaale -

= =- .60 17.397 Regien f(i j# l ,

g, i 35.000 = 3, ,7 3 " ""

's ' 3.2s 25.132 m / '

~

E  : 2 40 21.820 = ' ' '

6

.- 3.60 30.179 m ' ' ' ' ' ' '

30.000 3.80 '

=-

7 4.00, 32.851 ", ,

7 35.047 m j ,

4.20 7.389 m  ;

/

4.40 " 655 m f t . 6

' / i i ei 4.60 g 25.000 (( 4. 00 4. . . 44.024 m r- ' ' ' ' ' ~

I 90 m /

. 5,00.' J " 44, 2M / '

y g -

y--

4 ---

g 20.000

'h oeptable h

m [ \ '

_ Region-2 < /

r N2

' i i g 15.000 ^ ' ' ' ' ' '

e / '

d v'

\

h6

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10.000 f' ' ' '

• f , g. , , ,' , ,

/  !

< s s .

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5.000 ' ' ' ' \ '

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r i v' '.' . .

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• 6 6 i,, a

. i6 ,

. ..4 ... . . . ... ... ...

1.6 2.00 2.40 2.30 3.20 3.60 4.00

' 4.40 4.30 5.20 Fuel Assembly 2nitial U-235 R2ri t (w/o) -

([ , tes: The use of linear interpolation between the =4n4=

reported above is acceptable.

burnups FIGURE 5.6 - 1 .

(

xzxze sammF msus xxzTrAr. nazemam FOR REGrQN 2 STORAGE

% saAztwoon - unzrs i s. 2 s.5b Amenement no.g A

Cm,,Ed lefter dated ~ June 30,W 7 ges.B

i eif ow ta f h 1.u u.cr t- c- b b A wion a wan ii i rii,  ;&

h Figure E.7. jp -l D4 SERT Col n .e 0

.s . ,

0000 --

a 550g ,

)

s s re

/ ,/ ,

' r / /

50000 s 1' 5=

l, 1s s ee

\ -

8

,' --l

• f 7 / /, so saare CCEPTABLE ^ e e << <

45000 8 '- ' '

y \ / f/ f if

\ ~

/ /// //

A / t'fs '/

l40000 tear

\ /

/

/ /

/ // //

/a '

h \

f / s. s

/ A ' ):/, V i

3 \ / / / /, '/

g 35000 's ,' , ' , ', ' ' '

• g \ f f 2 ' fr: F 8 \ / / / IF

\ / / f/ F h 3o000 s

\

e4

% a >< u V / // /

f\,F / // ,/ ~

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/

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l 20000 ',',,'j ,'

Ihs

/ /IF' /

\ -

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/) /2 / \

' i /> 2' /

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15000

,'4f

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\

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10000

' ' NOTACC LE

,/ /j ,

hff

~

s L

w 1

\

Y }

\

5000 [

~

r s f

\

i / s I / \

0 v s s

\

1.0 2.0 30 4.0 5.0 Initial U-235 Enrichament (w/o) te. He use oflinear interpolation between the minimum burnups is a table.

FIGURE 5.6-1 hENIhRTM BURNUP VERSUS INITIAL ENRICHMENT FOR REGION 2 ALL CELL CONFIGURATION STORAGE Czo,Ed \eher dcde3 June 20,I(D j ( 3, g RN b

CTS INSERT (S)

SECTION 3.7 LCO 3.7.16 INSERT 5 5ba (A,,o) ,

00000 -

DECAY

. Tih!E:

55000 j 0 YEAltS

,_p 5 YEAlts

~~ *~

50000 ' 10 YEAlts

- A CC E l'TA131.E '

15 YEAltS

-+ 11 U lt N U I) DOM AIN .

20 T.EAlts 15000 / 7 -- -

~

i <

i/

2 Nw f'[~g~ } ~ -

i -

, ++  ; . .-,_ .- /,e 3 ; :

l. .l , .

Q)  ;

if '-- .

. i . i k35000 ' '

n +

a- b[' _~." j

! 1 5 *=-

' il l E 3OOOc '_'_'

y f^

g i i O ' '

[~_ -

i i

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{~ ~ ' ! '

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t

.a

=

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/~ g,;

, i i

)4

. 20030 y

r 9 -+--

-/ <

+- ~

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4_ __

4+-  ! I p 15000  !

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- 1

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m'< -4 b ,

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H 10000 -

~

-- U N A CC151'THIEE E

__. _ . - _ , - ~ . _ _ .+ _ - -- 13 UltN UI) D O M AI N_&_

. 5000 -.. . . _ _ _ _ _ .

7

~~

, _ _ , _ _ ~~*.___.._._ __ _Z 'Z - i i 0

1.0 2.0 3.0 1.0 5.0 INITI AL U-235 ENIllCllMENT (w/o)

Figure 3.7.16-1 (page 1 of 1)

Region 2 All Cell Configuration Burnup Credit Requirements 10/10/97 Revision B 1

g Figu rc 3.1,16 - 2, INSERT C 2 i

45000 l

l 1

i l

400 7 .

um.,

l i

4 j 'x fj e n ere i 'u . r/

\ '

1

• 3500C R s F / so seen.

\ J > I as v

. E f / rf 29 fomes

]

2

> > fu

, a ,EPTABLE i r/ff .-

g SObOC > / s/r

, r .,

w

\ ,

- i rjss .

> rrir

\ f. r/ f/

j

- . j f M r i 25000 \ t

i. rf f r A s vs> .

i '

> \ - s, rsa r

d. i s ' /, w

\ r, rsi r

2000C 3 2 r/ 2 r i

x f rf; r

. & 3 rs v s =

A FA rm l

I 15000 / rau X  ;

E /z r 3 -  ;

, A1 'rA \

a r> r x

. > ri e .

10000 'A P x 1 i

- = g 4 '

g i

'A \

/ F 5000 / f

NOTACCEPTABk A

< r \

/ J .

/ /

\

\

0- / 1 1.0 2.0 3.0 4.0 *0 Zaitial U-235 Enrichment (w/o) ote: '!he use oflinear interpolatiC&w. b h hpa h ,

FIGURE 5.6 2 MINIMUM BURN'JP VERSUS INITIAL ENRICHMENT FOR REGION 2 3-OUT-OF 4 CHECKERBOARD CONFIGURATION cen. Ed le&r clat ed June. 20,1397 5 Brd ( 8 - 5 0 Rev.B

__ _ _ _ , _ _ _ . . - . _ . _ . - . ~ . _ - _ _ . - _ - _ . _ . . . _ . _ . _ . - -

(

-CTS INSERT (S)

SECTION 3.7

,. LCO 3.7.16 INSFRT 5.ShA (contiriued) (Ag) 45000 4 i i1 1 i .

.

• i i >

! F ll!

' I ' '

a DECAY 40000 , , ,

TIME:

1  ! l l -----!  ! i l j 0. YEAllS '

~!  ! /'

o 35000 U

l , .

! 'l

' ' ' r 5

l lir/h i

Ig 5 YEAllS 10 YEAlts EAltS ACbEPTAbid: l /

15 20 {S EAllS

} "DURNUP DOMAIN--

! l

'~

$ 30000

~

' i ' >

lll l'bI~ ((

x  ? l I h25000

' ' ' ! iYN/ I 5

e i i i l  ; l 1

t 1

4 i

l !// -

i l

j

~

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• i i  ! i 4' i H 20000 #' ' i s , i i O

i ' ' !  ! , it i i X .

4 i -

i! i 4:  ! 1 i i i ,ij  ! ; j il5000 l , l ' ,,

'i

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, 'i

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! ' !  ! i i

/ r-3

! 1 i . i i i i '

0 i 1.0 2,0 3.0 4.0 . 5.0 INITIAL. U-235 ENRICllMENT (w/o)

Figure 3,7.16 2 (page 1 of.1)-

Region:2 3-out-of-4 Checkerboard Configuration Burnup Credit Requirements 10/10/97 Revision B

--__-__m__.___w._a

y- yypygg y -5 u' * ~*

PIb u re S.7. II. - 3 INSERT C-3

,J: .? T'. s . J

an . ..'#Gi:& 'y.~- .2 ' - '

5000 ,

3, ., .; _

.' i <.

I : ,' ,

i, p 1 s.

j')

e- .

/j y . -. .. f

~

400C 3 ~

,/j

\ .

ta b- - f g p ,

s R f $s t j ,

f

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,\  !'

j {

4CCEPTABLE -

/ /

e

.N t:s c 3000

/ / -

)

.: s.',

t) $ r

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. $' - L ki ! .

N / / .

.  ; t' h j f

j v }. y hj j ,

j 200C ,

v. 3 M

r .-

) yf .

f

/\

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't j f

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. i

/ / \

/ / \

j j \

/ / \

0 l l \

4. 4.2 4.4 4.6 4.5 5.0 taitial U-235 Barichment (w/o) ote:

- The use oflinear interpolation ts w. the minimum burnups la le.

FIGURE 5.6-3 MINIMUM BURNUP VERSUS INITIAL ENRICHMENT FOR REGION 2 2-OUT-OF-4 CHECKERBOARD CONFIGURATION Corn Ed -lenet dated June. 30,1397 sed (s - sb) RcV B

CTS INSERT (S)

SECTION 3.7

, LCO 3.7.16 INSER*f 5 5bA (continued) (Ag) 5000 .

I !  !

iii i  !  !  !

! 1 i l l {

! l l l I i l l 'l i I<' _ .;  !}l  ! I ! I l i i  ! j i i i

i I l ll I I il ii', _ + . - .

13d3,

,! l

! , i  ;

i i i i iiI I i 1 -. . I i 4

! ! I I !l!  !,i E 5 !!l 'ACCEPTAOLE I II I

t-- l j j BURNUP DO31AIN_ /~ i b ll} i I

[

5

- 3000 i i II i  ! I

~

l/l 6 7 ll 5

1 l l  !  !  ! l l l /

j I i i  !  ! i  ! I i l l / l D lI i

• I i l I 7 l i I  ; I i j i i I /

i '

I i i  ! i j j

/ l N2000 +

t i  !

I j / il i h.

s.

i I

I l

;3 -

!! l /,

I

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a i i i i i _l W '

i l- ,i' q-t / -:

i ,

! l C'1000 i / UNACCEPTABLE i

~

4

! l l HURNUP DOh!AIN

i l i l

i i i  ! !  !

i .

! I l l

• i l  ! i I j 0

4 l

1.0 1.2 1.1 1.0 1.8 5.0 i INITIAL U-235 ENRICllA!ENT (w/o)

Figure 3.7.16-3 (page 1 of 1)

Region 2 2-out-of-4 Checkerboard Configuration Burnup Credit Requirements l 10/10/97 Revision B

C S DOCS

DISCUSSION OF CHANGES TO CTS ITS SECTION 3.7 PLANT SYSTEMS An CTS LCO 3.9.12 Action b provides information with respect to exiting Action b and returning to Action a if-one of the two trains is returned to operable status (Action A allows fuel movement if the remaining operable train is placed in operation and is capable of being powered by an emergency power source). ITS LC0 3.7.13. Condition C re suspension of fuel movement if both trains are inoperable,quireswhile ITS Condition B-contains similar allowances as CTS. ITS relies on LCO 3.0.2 and Section 1.0 to allow / require entry into appropriate Conditions.

Upon restoration of onr inoperable train, the other train would still be inoperable, therefore ITS Condition B would be applicable and if the operable train was placed in operation and capable of being aowered by an emergency power source, fuel movement could take place (T1e Required Actions-of Condition B would be satisfied). Therefore, this change is administrative. No technical changes (either actual or interpretational) are made to the TS unless identified and justified.

A3 U By letter dated November 5, 1996. Comed requested a change to CTS LC0 3.9.11. Design Features 5.6.1.1 and Administrative Controls E

k 6.9.1.10. NRC letter dated April 2. 1997 issued Amendment 86 for Byron and Amendment 78 for Braidwood for this change.

An CTS SR 4.9.12.d.2 includes a parenthetical statement: (unless already operating). This statement is deleted. Current verification of system start, including the fan. procedures Thereforerequire this deletion is administrative in nature. No technical change (either actual or interpretational are made.

A3 A note has been added to CTS SR 4.9.12,d.3 (ITS SR 3.7.13.5) to indicate that this SR is only required during movement of irradiated fuel assemblies in the fuel handling building with the equipment hatch intact. This note is used to differentiate this SR versus ITS SR 3.7.13.3. which is required when the hatch is not intact. No

. technical changes (either actual or interpretational) are made to the TS.- unless identified and justified.

BYRON /BRAIDWOOD UNITS 1 & 2 37 7 10/10/97 Revision B i

i DISCUSSION OF CHANGES TO CTS ITS SECTION 3.7 PLANT SYSTEMS A., CTS SR 4.9.11 states. *The water level Jin the storage pool shall be determined to be at least its minimum required depth at least ont.e per 7 da s when irradiated fuel assemblies are in the fuel storage pool." The

-IT has revised this SR (SR 3.7.14.1) to read. ." Verify the spent fuel pool water level is = 23 ft above the to

-assemblies seated in the storage racks"atpaofFrequency the irradiated fuel The of 7 days.

ITS SR provides a more descriptive and accurate representation of what  ;

the SR is to accomplish. This change is editorial in nature and does not change any meaning or intent of the NUREG.

Ag NRC letter dated April 2. 1997 issued Amendment 86 for Byron and Amendment 78 for Braidwood for soluble boron in the spent fuel pool (SFP). Since the license amendments were temporary in nature. Comed .

letter dated June 30, 1997 proposed changes to permanently take credit for soluble boron in the SFP. Additionally. Comed responded to the ,

NRC's request for additional information in Comed letter dated -

m September 25, 1997. Although not yet approved by the NRC. Comed has 3 used the June 30. 1997 and the September 25, 1997 submittal revisions as o the CTS markup ) ages for the ITS conversion. The clouded portions M reflect these clanges.

An CTS 3.9.11 is titled. " WATER LEVEL / BORON CONCENTRATION - STORAGE POOL."

This LC0 has been divided and incoraorated into two separate LCOs. ITS' LCO 3.7.14. ' SPENT FUEL POOL WATER _EVEL.* and 3.7,15. " SPENT FUEL POOL BORON CONCENTRATION." This change is editorial in nature and does not change any meaning or intent-of the NUREG.

Au CTS APPLICABILITY 3.9.11 states. "Whenever irradiated fuel assemblies are in the storage pool " The ITS has been revised to state. "Whenever fuel assemblies are stored in the spent fuel pool." This is an administrative change since the intent of the wording of the CTS is the same as the ITS. The intent of the CTS is that whenever irradiated fuel assemblies are in the spent fuel pool, they are in-a storage configuration. Any irradiated fuel assembly brought into the not-just pass through, it will be seated, even if temporarily, pool and will will therefore be considered stored.

l i

I i

BYRON /BRAIDWOOD UNITS 1 & 2 3.7 12a 10/10/97 Revision B L

u. . - ._ . - . - -

DISCUSSION OF CHANGES TO CTS ITS SECTION 3.7 PLANT SYSTEMS M, A note is added to CTS SR 4.9.4.2 clari'fying that ITS SR 3.7.13.3 (FHB negative 1/4 inch, with the equipment hatch not intact) is required during CORE ALTERATIONS or during movement of irradiated fuel assemblies  ;

(whether inside containment or inside the fuel handling building). This SR verifies the integrity of the enclosure: in this case the FHB and the containment. CTS 4.9.4.2 and 4.9.12 do not require this SR to be met if irradiated fuel assemblies are being moved in the FHB with the equipment i D hatch not intact. This represents an additional restriction on plant  !

operation. The CTS words " removed" are modified to "not intact." to dl encompass the postulated scenarios (e.g., both air lock doors opened).

Ha Not used.

Mn Consistent with NUREG-1431. LCO 3.7.3 "Feedwater (FW) Isolation Valves" and LC0 3.7.4. " Steam Generator (SG) Power Operated Relief Valves (PORVs)" are added to the ITS. Changes to CTS containment isolation valve (CIV) requirements are individually annotated and any technical changesarejustifiedseparately.

CTS CIV Actions do not include, nor is it appropriate in the CIV TS.

Required Actions for the condition of two feedwater valves in the flow path inoperable, because the second containment isolation boundary is the secondary side piping. From the standpoint of ITS LC0 3.7.3, a second isolation valve is required. Therefore. LC0 3.7.3 Condition B is

{ provided. These changes represent an additional restriction on plant i operation.

b i

Hu Consistent with NUREG-1431. ITS SR 3.7.13.5 adds an upper flow rate limit to CTS SR 4.9.12.d.3. This SR verifies the ability of the FHB Ventilation System to maintain the enclosure at a negative pressure. If the system were to run at a flow rate greater than design, the negative

]@l a

pressure may be met but the larger flow rate could be indicative of system degradation. This represents an additional restriction on plant operation.

5Mu

' CTS 5.6.1.1.e. g and h describe conditions for new and spent fuel as well as interfacing requirements for their storage in various regions of j the spent fuel pool. Tqese conditions have been relocated to ITS i LCO 3.7.16 a. b. and c respectively, in order to make the ITS LC0 N complete. an introductory LCO sentence. APPLICABILITY ACTIONS and E Surveillance Requirements have been added. The additional Actions and C

Surveillance Requirements constitute a more restrictive change.

BYRON /BRAIDWOOD UNITS 1 & 2 3.7 17 10/10/97 Revision B

l l

E l l

. i DISCUSSION OF CHANGES TO CTS l

! ITS SECTION 3.7 PLANT SYSTEMS 4

Ha CTS LCO 3.9.11 APPLICABILITY states. "Whenever irradiated fuel

assemblies are in the storage sool." ITS LC0 3.7.15 APPLICABILITY -

{ revises the CTS b spent. fuel pool."y stating, The CTS"Wienever fuel assemblies APPLICABIllTY areirradiated was only when stored infuel the . !

D .was stored in the pool. The-ITS is more restrictive since-the f

' P- elimination of the word " irradiated" now requires that the APPLICABILITY '

E is for anytime new or irradiated fuel is in the pool. This change is *

j. consistent with NUREG-1431.

i )

t l

[  :

I i

i i

i a.

?

r

+

BYRON /BRAIDWOOO UNITS 1 & 2 3.7 17a 10/10/97 Revision B

l DISCUSSION OF CHANGES TO CTS ITS SECTION 3.7 PLANT SYSTEMS 14 CTS Actinns 3.9.11.a and 3.9.11.b. reqtlire suspending crane operation with loads in the spent fuel pool area when the water level or boron concentration is below their limits. These details are being relocated-m- to the TRM since this information is not necessary to be included in the

~3 TS to provide adequate protection of public health and safety. The Y relocation of this information maintains the consistency with NUREG-1431. Any changes to this descriptive information will be made in accordance with 10 CFil 50.59.

i

-BYRON /BRAIDWOOD UNITS 1 & 2 3.7 32a 10/10/97 Revision B

DISCUSSION OF CHAISES TO CTS ITS SECTION 3.7 PLANT SYSTEMS L3 CTS SR 4.7.7.d.3)-requires that the Non' accessible Area Exhaust Filter Plenum Ventilation System maintains, with two trains operating, the ECCS equipment rooms at < -0.25. inches with respect to atmosphere with each train operatino witRin a flow rate band (between 55.669 cfm and 66.820 cfm Byron: 66.900cfm110%Braidwood). Consistent with NUREG-1431 SR 3.7.12.4. the requirements associa1:ed with a lower flow rate are deleted from this surveillance. The design bases analyses (offsite and control room dose assessments) assume the-ECCS leakage outside containment is instantaneously released from the volume containing the

-leakage source (i.e., no credit is taken for holdup)- and that the release is filtered. As such, the system flow rate is not a parameter in the dose analyses. As described in the Bases for SR 3.7.12.4. the g puroose of this surveillance is to verify the integrity of the ECCS pump S room areas. If the surveillance can be satisfactorily performed at a -

t flow rate less than the stated flow rate the intent of the surveillance

• is still met. The VFTP Surveillances associated with the Nonaccessible

' Area Exhaust Filter Plenum Ventilation System (SR 3.7.12 2) test the HEPA and charcoal filters in conjunction with individual trains. These 4{ Surveillances include specific flow rate bands for the train.- The remaining LCO 3.7.12 SRs verify that the system will start when called upon.

L3 CTS Action 3.9.11.a states " ... restore the water level to within its

• limit within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.' ITS 3.7.14. Required Action A.1 revises this to state. ' Suspend movement of irradiated fuel assemblies in the_ spent fuel pool immediately." The ITS is less restrictive since it does not require the restoration of the water level in order to place the plant in a safe condition. The purpose of the ITS Specification is to ensure that there is sufficient water level above the top of the stored fuel to U maintain the initial conditions assumed in the design basis fuel a handling accident. This accident assumes an irradiated fuel assemble is

,' e dropped onto irradiated fuel assemblies seated in the storage racks.

E- resulting in ruptured fuel rods. Suspending fuel movement precludes a W-

': fuel handling accident from occurring. With the initiation event for the design basis accident removed, restoration of the initial conditions

}t is not required. This change does not reduce the level of safety as described in the applicable _ Bases. This change is consistent with E, NUREG-1431.

l L

. -BYRON /BRAIDWOOD UNITS 1 & 2 3.7 46a 10/10/97 Revision B L

~

_ . _ _ _ _ _ . _ _ . _ . . ._._.._.~. _ _._._._. _ . _ _ _ _ . . _ _ . _ _ _ . . _ . _ _ _

• DISCUSSION OF CHANGES TO CTS i ITS SECTION 3.7_ PLANT SYSTEMS f Lu CTS 3.9.11 APPLICABILITY states. "Whene'ver irradiated fuel assemblies are in the storage pool." ITS APPLICABILITY 3.7.14 revises this to state. "During movement of irradiated fuel assemblies in the spent fuel pool." This is considered to be a less restrictive change since the

' APPLICABILITY has been changed from whenever fuel assemblies are in the s)ent fuel pool to only during moveirent of irradiated fuel assemblies in tie spent-fuel pool. This is acceptable since the accidentl3f concern is the release of fission products resulting from a fuel handing-accident. Since both Byron and Braidwood have, and will continue to og have, irradiated fuel stored in the fuel pool, the CTS Specification would be applicable at all times. With fuel being stored, there is no 1 action that would exist, other than movement of the fuel.-that would be -t Qt an initiator of a fuel handling accident. Although less restrictive, the ITS only requires the APPLICABILITY for when fuel is being moved, thus a potential for a fuel handling accident. This change. although '

less restrictive than the CTS. is consistent with NUREG-1431.

. )'

r f

BYRON /BRAIDWOOD UNITS 1 & 2 3.7 46b 10/10/97 Revision B

. - _ - ~ - -_ . . .. .

g, , ,_ __e a m,.aon -- -J--nk-"A " " # # A' "' ~^^' ~' ~

O LCO MAR (JPS

~

7 Fuel-St:r : Pool Water Level 3.7.tk

pl 14 3.7 P EMS n ,

3.7. e. Pool Water level '

LCO 3.7. Th 1 steente pool water level shall'be h 23 ft over the top of irradiated fuel assemblies seated in the storage racks.

APPLICABILITY: During movement of irradiated fuel assemblies in t el eters;e pool.

ACTIONS <

i CONDITION REQUIRED ACTION COMPLETION TIME

A. uel ster

g: pool A.1 --------NOTE--------- - F water leve' not within LCO 3.0.3 is not g limit. applicable.

Suspend movement of Immediately irradiated fuel assemblies in the '

t

r:;e pool.

6 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR . 7.1 .1 Verify th 1 <ter:;: pool water level is 7 days h 23 ft above the top of the irradiated fuel assemblies seated in the storage racks.

WOG STS 3.7-35 Rev 1, 04/07/95 Rev.8

. . - . = - _ _ . - _ - _ . _ - . . _ . . - . . _ _ _ . . - - ~ . .

r-Fuel Str:; Pool Boron Concentra81on

. 3. 7.R IS 3.7 PL TEMS

< 5 peci 3.71

• Fuel .itr:;; Pool Boron Concentration ,

15 LCD 3.7.1% The% f g%,.uel .t :traae pool boron concentration shall be l

l 6 7pm. -

Wh[never APPLICABILITY: Wen fuel assemblies ara stored in th risr - nooll d a el sto ge po nce th ast nu(f verificgion I (s not beenderf ed vement of fuel (ssemblies M th fuel ,

storag l 001l, Y

ACTIONS <

CONDITION REQUIRED ACTION COMPLETION TIME (GpenO _

A. +/uel e4+eege pool ----------- NOTE------------

boron concentration LCO 3.0.3 is not applicable--

not within limit. ----------------------------

A.I Suspend movement of Innediately fuel assemblies in the fuel str we -

pool. e eg}

ANp A.2 $ action to .Isnediately Initiatgfuel :ter:;:

restore pool boron concentration to within limit.

/*

A.2.2 Verify y mmediately admin trative means (Re on 2) fuel st age pool rification has een erformed sine the last movement of fuel assemblies the fuel stor e pool.

4 WOG S'. 3.7-36 Rev I, 04/07/95 Rev.8

p Cpant NLV Fuel it:r:;; Pool Baron Concentration

3. 7.R W -

)

SURVEILLANCE REQUIREMENTS i W SURVEILLANCE FREQUENCY

'll 'M 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br /> SR 3.7.M 1 Verify the uel eter:;: pool boron concentration is eithd- ' f ;it,

' day:- h A 2000 mm - @

t t

L h

. t WOG STS- 3,7-37 .

Rev 1, 04/07/95 Rev.8

~

h h Spent Fuel Assembly 5 ra 14 4 3.7 PLANT SYSTEMS i

3.7. Spent Fuel Assembly Storage $,ie .i. , - s.

• c ) ,and decai

b. Region 2 MLVe-LCO 3.7.%g Haeone co'mbination of initial enrichment end 3 f burnup ef :::h

.y n6 iv .1 u..iid,1,7 ;t-. .J ;,. " i

t.;11 Le within the

Acceptab' MBurnup Domain 7'of .I5;;c;r;. .'.'.12 , 1

r '-

' - 3 r :-5.: dith 5 i'!:Mi: ' 3 . 2 .?'. .

!E f-)

Figute. 3.'.5-17 3.7.16-2, or 3 7.16 -3 , oc oppli ca bie for + hat .doroge cc.ofiguration.

l APPLICABILITY: Whenever any fuel assemb1'y@nored in fA:;i:: ?) :f the

, spent fuel e4+eege pool.

ACTIONS I CONDITION , REQUIRED ACTION COMPLETION TIME A. Requirements.of the- A.1 --------NOTE---------

A LCO not met. LCO 3.0.3 is not applicable.

..................... }

Initiate action to lmediately 3 move the noncomplying -

fuel assemen ,,bly+_a__ won. into o location

'"""W which rectorec com phonce.

, , pordelecay

+nne SURVEILLANCE REQUIREMENTS toa ...,

C:1mer t 5 7 - 2ns e ] SURVEILLANi .E FREQUENCY 2 -(combination of)

SR 3.7. a$) . Verify by administra ive means the+ initial Prior to enrichment3 end burnu) of the fuel assembly storing the isj ir, eccorder.;; wit . Figur; ;.7.R ' cc fuel assembly ss 17./6.3 3;;ific;ti:.'.2.!/. l(a in1 Region 2f

( 1.e, t 3 7.z e e 3 earnu --

wdhm 4he. AcceptoLie37.16 - I , 3.7.I L p cemcu n c4 f~igure. 2 a o r 3.~I llo - 3.

WOG STS 3.7-38 Rev 1, 04/07/95 Rev.8

LCO INSERT (S)

SECTION 3.7  !

< LCO 3.7.16 INSERT 3.7 38A (P,) 3 LCO- 3.7.16 Each s shall: pent fuel assembly stored in the spent fuel pool i i

a. Region 1 Have an initial nominal enrichment of s 4.7 weight Q3 3ercent U-235 or satisfy a minimum number of Inte ral r  :

[ uel Bu. jble Absorbers (IFBAs) for higher initia enrichments up to 5.0 weight percent U-235 to permit storage in any cell location.

i l

b. ...

c. Interface Requirements Comply with the Interface Requirements within and between adjacent racks as described in the " Byron and 4 Braidwood Spent Fuel Rack Criticality Analysis Using Soluble Boron Credit." j 1 -

i l_ 10/10/97 Revision B

_._m . _ _ . __ _ . _ - . _ . _ _ .-- _- . . _ _ . _ - _ . - . _ __. ._ - . - - - - - _

LCO INSERT (S)

SECTION 3.7 LCO 3.7.16 INSERT 3.7 38B (Pa )

SURVEILLANCE FREQUENCY SR 3.7.16.1 Verify by administrative means the Prior to initial nominal enrichment of the fuel storing the assembly is s 4.7 weight percent U-235 fuel assemblv or a minimum number of IFBAs is met. in Region 1 Q

J ...

i SR 3.7.16.3 Verify by administrative means the Prior to interface requirements within and storing the between adjacent racks are met, fuel assembly in the spent fuel por l 10/10/97 Revision B

[ h Spent Fuel Assembly 5 age M

40

~

35 /

~ ACCEPTABLE BURNUP DOMAIN g 30 ,

!! ~

R ~

2 25 gn t

E

$ 15

~ \ \

,/  !

/

E  : "

UNACCEPTABLE h

j BURNUP DOMAIN 5 lo /

~

~

/

5 i '

/ '

/ \

....\.ig...

~

o

i. ..,..

/.g.... ....

1.5 - 2.0 2.5 3.0 3.5 4.0 4.5 5.0 INITIAL ENRICHMENT, %U 2 Not to be used for Operation, k, For illustration purpows onir ,

Figure 3.7.N 1 (page 1 of 1)

Fuel A

' ssembly Burnup Umits in Region 2

/

/

/ .-

/ .

WOG STS 3.7-39 Rev 1, 04/07/95 Re.v.B l

1

___ - __m._____ ____ _ ..__. - . . . . . . _ . _ . . _ . _. .__- _ _ _ _ ._ - . _ . _ . _ . . . _ _ . _ _ _ _ _

LCO INSERT (S) i SECTION 3.7 >

i LCO 3,7.16 ,

INWRT 3.7 39A (P.)

00000 DECAY  !

TihlE:

55000 j 0 YEARS

/~

f- 2 5 YEAltS

, f r 50000 , . f ' , ,y ,. 10 YEAltS i

ACCEPTABLE  ; j,/!f f 15 i;EAlts 1 BURNUP DOh1AIN i

i

, i f, 20 T EAllS i / -

45000 . , , , ,  ; . ,

, j f f ,. ,,ff,/

i i i i/ f1 /1 ef !

t t i

! i

,' - 1

/_

/: /

/ / ff !

I /f i I i i I i i / 'I / ff ,

2, - .3 O ,

, , .f, f j ff ,

5 g- i a

! i i i i ' / i/ /;Ff

/ / Nf *

[35000 f f f ff q

a i i , i / via i i

} '

ll I

! i/

' Y' '! i E 30000 l .f g

C ,

, , / Viky i i i i

. t /_-

//

r_ i i ,

i I N

M 25000 j

~

4  ! 'I i  ! I D / / /// ,

O ' '

/ // '  ! i

) j g 20000

-[j ff. , , , ,

= i _/ /

! I i  ! I

-y g / /;

/} _

l ' _i

, t i _i 1 1 M<

15000 , /" ~ , ,

l

< 1 i i f f i .

R u 10000 . >

w-

, . UNACCEPTABLE i i

r

- -DURNUP

, - i DOM AINi it-5000 , ,

i '

g_ , .! i

! i i I

i. _

i 1 ,

9

1.0 2.0 3.0 4.0 5.0 3 j i

INITIAL U-235 ENRICllMENT (w/o)

. Figure 3.7.16-1 (page 1 of 1)

Region 2 All Cell Configuration Burnup Credit Requirements c ,1 10/10/97 Revision B

LC0 INSERT (S)

SECTION 3.7 LCO 3.7.16 INSERT 3.7 39A (continued) (P )

45000 j , DECAY

' ' TIME:

40000 0 YEAllS

,[t i , ,

! 5 YEARS E 35000 i

! / to YEAllS t-- i ' ' '  !  !  !

! ! /. - 15 YEARS E

ACCEPTABLE ll '

!f [t 20 YEARS iil HDURNUP DOMAIN- ,

E30000 ' i ii ' '

l. !l l

'~

f

~

D \

/

C =  ; ^ I' l l ,

IVM /'I cc e-. ,5000 , , , , ,

, , f, ,

i , , j i j ff I E , -

i i '// i

.- 20000 i

'ji J

l,l, i)[

= ~

i,i i j "i i C i ,

'f l I i i G

f 15000 .

7

' ~

i

,/# f/ ',!

ll i < ^'

i/ .  !

1 l  ! I d

g 10000 if l d!l ,

_74 ll ,

l  ; ,

-/

UNACCEPTABLE -

j _ . _ -

BURNUP DOMAIN -

I

,  ; y , ;

a000 .

_t - ! . !  !

.a jql 4_3 j

0 1.0 2.0 3.0 4.0 5.0 INITIAL U-235 ENRICHMENT (w/o)

Figure 3.7.16-2 (page 1 of 1)

Region 2 3-out-of-4 Checkerboard Configuration Burnup Credit Requirements I

l 10/10/97 Revision B

LCO INSERT (S)

SECTION 3.7 iG.'3.7.16 IEFRT 3.7 39A (cont'inued) (P,)

5000 ,

i' l- i

! 4 ! !  !. ! r1 _j t -

I

! I II

- , i i ! -! l i! r

! ,I h.Q'_

'I I I i ! !

! I_  ! a _

I I l-!

1000 I I I

Iii I i l l I j/i

~I I t I ! !1 l l  ! I i I/t t E ACCE'PTABLE I

-i b

BURN,UP DohtAIN

/

OD E t I i I

/

3000 ,

/t 1i I i!!I i If E 55 II l 1 -1 5 i i!I i i

/

1 ii II /

I 5 t-i-I i j I lI

/ l i !ii i / II h2000 ,

! ) ) ; y  ;

i M-  ! I i !I /

I 4 i II l Ii /l I

{ -j t i il i i1 /  !

N 1000 I I ;

UNACCEPTABLE i >a i .+ j j j BURNUP D0hlAIN 1-i i -3 i. .

!  ! i

~

i 1 i i , ii j j j -- I

! (/

4 i ii  !

0  !/ '  ;

I!  !! ,

l

+

1.0 <l .2 1.1 1.6 l' <1. 8 5- 0 t 'NITIAL U-235 ENRICllh!ENT (w/o)

Figure 3.7.16-3 (page 1 of 1)

Region 2 2-out-of-4 Checkerboard Configuration Burnup Credit Requirements I

);

10/10/97 Revision B

JUSTIFICATION FOR DIFFERENCES TO NUREG 1431 LCOS SECTION 3.7 PUWT SYSTEMS PLANT SPECIFIC CHANGES (P)

P During the development certain wording preferences. English language conventions, reformatting, renumbering, or editorial rewording consistent with plant specific nomenc16ture were adopted. As a result, the Technical Specifications (TS) should be more readily readable by, and therefore understandable to plant operators and oth;r users.

During this reformatting, renumbering, and rewording process, no technical changes (either actual or interpretational) were made to the TS unless they were identified and justified.

P, Consistent with the CTS, the LCO and Applicability requirements for AF and CST in MODE 4 (ITS LCOs 3.7.4 and 3.7.6) are deleted. The justification is that in Mode 4. the pressure and tem >erature limitations are such that the priability for design ) asis events (e.g..

SGTR) requiring the operation of the SG PORVs is low. In addition, the RHR system is available to 3rovide for decay heat removal. With the RHR system available for decay leat removal, there exists adequate heat removal capability. Consistent with the CTS Applicabilities. ITS LCO 3.7.5 (which is a new LCO for Byron and Braidwood) MODE 4 requirements are deleted.

P3 gmropriate plant specific values are provided.

P. NUREG LC0 3.7.15 and 3.7.16 have been changed to ITS 3.7.14 and 3.7.15.

Tnis numbering change is due to the deletion of a previous LCO. In addition, the NUREG uses the phrase. " Fuel Storage Pool Water". This has been changed in the ITS to read. " Spent Fuel Pool Water". This change makes the LCO consistent with plant terminology. This change is also consistent with the NUREG philosophy in utilizing specific )lant system and component names. In addition. NUREG Figure 3.7.16-1 las been replaced with three different Figures to be consistent with the revised g LCOs (Reference Pa).

Ps ITS LC0 3.7.14 and ITS LCO 3.7.15. Required Action A contains a Note 5 stating. "LC0 3.0.3 is not applicable." This Note is being relocated in its entirety to directly below the " Actions" statement. Moving the Note is an editorial change since there is only one Action requirement. This change is for consistency with other ITS single Action statement Specifications.

BYRON /BRAIDWOOD UNITS 1 & 2 3.7 6 10/10/97 Revision B

JUSTIFICATION FOR DIFFERENCES TO NUREG 1431 LCOS SECTION 3.7 PLANT SYSTEMS Pa The CTS on Containment Penetrations allows the equipment hatch between containment and the fuel handling building to be removed, during core alterations or movement of irradiated fuel assemblies in the containment, if the FHB Ventilation System is capable of maintaining the fuel handling building at a specified negative pressure. This allowance has been moved to the ITS FHB Ventilation System specification.

Additions have been made to the ITS LCO Applicability. Actions, and Surveillances, as well as the Bases to reflect this allowance.

NUREG Conditions C and D are modified to delete "during movement of irradiated fuel assemblies in the fuel handling building." Given the ITS Applic-bility. this informatson is not correct. A separate SR is added to verify that the FHB Vertilation System can maintain the fuel handling building at a specifieo negative pressure when the equipment I

hatch is not intact, similar to the CTS SR. This SR requires that the test be performed with the hatch not intact versus the NUREG SR which is written assuming the fuel handling building is the entire enclosure.

Notes are added to the "1/4 inch" tests to differentiate the volume tested as we9 as to reflect the different frequency requirements.

Pp Consistent with CTS SR 4.7.3.3.b. ITS SR 3.7.7.2 is added to verify that the Essential Service Water System valves directly serving the CC heat exchanger are in the correct position or can be aligned to the correct position.

.] BYRON /BRAIDWOOD UNITS 1 & 2 3.79 10/10/97 Revision B

JUSTIFICATION FOR DIFFERENCES TO NUREG 1431 LCOS SECTION 3.7 PLANT SYSTEMS Pu NUREG 3.7.17 has been changed to ITS 3.7.16. In addition. NUREG LC0 3.7.17 has been divided into three distinct LCO subparts. The three parts of the ITS LCO are consistent with CTS Specification 5.6.1.1.

They have been moved to this ITS LC0 to enhance their use and not require operators or other users of the LC0 to look in two different Technical Specifications. The three distinct subparts are as follows:

ITS LC0 3.7.16.a contains the requirements for Region 1 fuel storage requiring the following. "An initial nominal enrichment of s 4.7 weight percent U-235 or satisfy a minimum of Integral fuel Burnable Absorbers (IFBAs) for higher initial enrichments up to 5.0 weight percent U-235 to permit storage in any cell location".

ITS LCO 3.7.16.b revises the NUREG LC0 by stating. "A combination of initial enrichment. burnup and decay time within the Acceptable Burnup Domain of Figures 3.7.16-1. 3.7.16-2, and 3.7.16-3. as applicable for -

4 that- storage configuration." This change 1s-basically consistent with the NUREG with the exception that two additional Figures have been added i

e to represent the actual configuration used by the plant.

? ITS LC0 3.7.16.c provides Interface Requirements for the Spent Fuel e Pool. ITS LCO 3.7.16.c.sta+es." Compliance with the Interface Requirements within and between adjacent racks as described in the

" Byron and Braidwood Spent Fuel Rack Criticality Analysis Using Soluble

{, Boron Credit.""

In addition. NUREG LC0 3.7,17 Required Action A.1 has been revised. The NUREG Required Action A.1 states. " Initiate action to move the

, noncomplying fuel assembly from [ Region 2]." This has been revised in ITS 3.7.16. Required Action A.1 to read. " Initiate action to move the

' noncomalying fuel assembly into a location which restores compliance."

This clange is made to be consistent with the ITS LC0 3.7.16.

ITS SR 3.7.16.1 and 3.7.16.3 have been added. SR 3.7.16.1 requires a verification by administrative means that the initial enrichment of the fuel assembly is s 4.7 weight percent U-235 or a minimum number of 16 IFBAs is met prior to storing the fuel assembly in Region 1. ITS SR 3.7.16.3 requires a verification of the interface requirements within and between adjacent racks be met for the intended storage location.

These new SRs are needed to ensure compliance with added LCOs 3.6.17.a -

and 3.7.16.c.

- In addition. ITS SR 3.7.16.2 has been revised to be consistent with LC0 3.7,16.b.

BYRON /BRAIDWOOD UNITS 1 & 2 3.7 9a 10/10/97 Revision B

-JUSTIFICATION FOR DIFFERENCES TO NUREG 1431 LCOS SECTION 3.7 -

PLANT SYSTEMS P,

i ITS LC0 3.7.4 and associated bases are revised to reflect plant specific nomencleture and plant specific analyse's. The Atmospheric Dump Valves are referred to as the Steam Generator (SG) Power Operated Relief Valves

! (PORVs) at Byron and Braidwood.

g.l BYRON /BRAIDWOOD UNITS 1 & 2 3.7 9b 10/10/97 Revision B

JUSTIFICATION FOR DIFFERENCES TO NUREG 1431 LCOS SECTION 3.7 PLANT SYSTEMS P,, (Byron Only) Consistent with CTS L".0 3.7.S f and g, Action f. and SR 4.7.5.c. ITS SR 3.7.9.1 and % 7.9'.4 are added. These SRs confirm river level (and flow. if neces w . The Byron VHS design relies on makeup capability to provide in w.sory to the cooling tower basins. The source of this makeu) is the Essential Service Water makeup pumps which are supplied by the Rock River.

P u (Byron Only) ITS SR 3.7.9.6. related to SX makeup system valve lineup.

is added consistent with CTS SR 4.7.5.e.3).

P,,

The Completion Time for ITS LCO 3.7.4 Condition A is revised from 7 days to 30 days; There are no CTS " Plant System" requirements associated with the SG PORVs. Maintenance /Repaf r of an inoperable SG PORVs has.

sometimes in the past taken longer than the NUREG allowed 7 days.

P,3 The Notes contained in the Required Actions for ITS LCOs 3.7.7 and 3.7.8 are modified by adding the word "if" just prior to the phrase made inoperable by .... This modification clarifies that the Required Actions and Completion Times referenced do not need to be entered if the component is not impacted. Byron and Braidwood have significant flexibility in the CC and Essential Service Water Systems such that it would be rare that a single inoperability would 1mpact the systems listed in the Notes.

P3 CTS LCO 3.9.12 Actions require verification that the remaining Operable ventilationtrainiscapableofbeingpoweredbyanOperableemergency power source.

this requirement. ITS LC0 J.7.13 Re'~iired Action B.1.2 is added to reflect P,,

Consistent with CTS SR 4.6.1.1.a and ITS LC0 3.6.3 Actions Note 1 (Containment Isolation Valves) and the philosophy of the Generic Change described in Bases JFD C 4 . a note is added to LC0 3.7.3 to allow FW lsolation Vsives to be unisolated intermittent'iy under administrative control.

Pa Consistent with CTS LCO 3.7.1.5 and LC0 3.6.3, the Applicability of ITS LCO 3.7.2 c.1d LCO 3.7.3 are revised to MODES l'. 2. and 3. with no-provisions for exclusion if all the valves are closed.

d l BYRON /BRAIDWOOD UNITS 1 & 2 3.7 11 10/10/97 Revision B i

JUSTIFICATION FOR DIFFERENCES TO NUREG 1431 LCOS SECTION 3.7 PLANT SYSTEMS P3 NUREG LCO 3.7.16 A>plicability states in part. "... and a fuel storage pool-verification las not been performed since the last movement of fuel assemblies in the fuel storage pool." This statement was deleted in the ITS. -

In addition. NUREG 3.7.16. Re uired Action A.2.2 has been deleted. '

This Required Action stated. "Verif by administrative means [ Region 2) >

fuel storage pool verification has een performed since the last movement of fuel assemblies in the fuel storage pool." The NUREG

' assumes that the rack k i water, as long as they ,do,not remains have asfuel 0,95assembly when flooded with unborated mispositioned. With the spent 'uel pool still unborated, and a fuel assembly mispositioned, D the rack k,,7 approaches or exceeds the k,,, limit of .> 0.95. In accordance with the Braidwood and Byron analysis, both plants must t

e maintain at least a boron concentration 550 ppm in the spent fuel pool, evenwithallfuelassembliesproperlystored,tomaintainak,bppmis s 0.95. -If an fuel assembly is mispositioned, a minimum of 16$ of required to be in the spent fuel pool. Boron is credited at both plants at all times for keeping the fuel rack k,,, s 0.95, there-is no capacity '

to use or rely upon the boron concentration rt. laxation which results from performing a position verification. Therefore, the subject-Required Action is deleted.

P3 (Byron Only) Consistent with CTS SR 4.7.5.e.2). ITS SR 3.7.9.7 includes verification that the SX makeup pumps will start on t low basin level signal. The frequency is specified as in accordance with the IST Program.

BYRON /BRAIDWOOD UNITS 1 & 2 3.7 11a 10/10/97 Revision B l

l

Spent Fuel 4t: 'g= Poel Wa ter I.av'cl B 3.7.NL

,. M B 3.7 PLANT SYSTEliS B 3.7. t,"uel-St;r;;;PoolWaterLevel BASES BACKGROUND The minimum water level in the fuel :t:r:.ge pool meets the assumptions of iodine decontamination factors following a fuel handling accident.l The gecif t;d w;;;r 1:v::-:M:n p eM riMr?:er ne p erel r:: 6:e "5 the :terige acks

a. , , . . .- . ....ir =i = ;;;:::ity. I The water also provides shielding during th ovement of spent fuel.

A general description of theWuel :t;rege pool design is given in the@SAR, Section'i9.1.27'(Ref.1). A description of theJpent Fuel Pool Cooling and Cleanup System is given g in theUSAR, Section'19.1.37'(Ref. 2). Thegssumptions of

' the fuel handling accident are given in theEfSAR, Section T15.7.47'(Ref 3).

APPLICABLE The minimum water level in the U -

F fuel :ter:g; pool meets SAFETY ANALYSES the assumptions of the fuel handling accident described in Regulatory Guide 1.25 The resultant 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> thyroid dose' per personat (Ref the4).exclusion area boundary is a small fraction of the 10 CFR 100ctRet. 5)?limitsp According to Reference 4, there is 23 ft of water between the top of the damaged fuel bundle and the fuel pool surface during a fuel handling accident. With 23 ft of water, the assumptions of Referer.ce'4 can be used directly. In practice, this LCO preservss this assumption for the bulk of the fuel in the storage racks. In the case of a single bundle dropped and lying horizontally on top of the spent fuel racks, however, there may be < 23 ft of water above the e -- ' W f = 1 br0:e ar,0tu imren. . A .ted Lv UnD width of the bundle. To offut this small nonconservatism, $

the analysis assumes that all fuel rods fail, although analysis shows that only the first few rows fail from a hypothetical maximum drop.

The+ fuel :ter:g: pool water level satisfies Criterion 2 of g ;_.m. _ . _ _ _ _ . .

I vocreto.xca nun 2 (continued) i WOG STS B 3.7-78 Rev 1, 04 '07/95 Rev B

Spent Fuel ite. .;, Pool Water Level i

B 3 7.R 14 BASES (continued)

LCO The fuel ;ter-se pool wa1Ier level is required to be 123 ft over the top of irradiated fuel assemblies seated in the storage racks.

The specified water level preserves the assumptions of the fuel handling accident analysis-(Ref. 3).

As such, it is the minimum required for fuel storage and movement within 1 :t;r;;;-pool.

APPLICABILITY This LCO applies uring movemen't'of irradiated fuel assemblies in the fuel stepage pool, since the potential for

.a release of fission products _ exists.

l ACTIONS A.1 The ACTIONS hoVC btCO .

":eir:d A;th: A.! h modified by a Note indicating that MP LCO 3.0.3 does not apply.

U When the initial conditions C.assumai en 4he

p. - . ..= :.. =anedenf anoivse s

x. ;;;;;n.r cannot be met, steps should be taken to preclude the- -

P3 accident from' occurring. When the+ fuel et: :ge pool water en level is lower than the-required level, the movement of-irradiated fuel assemblies in the+ruel 44eeege pool is issuediately suspended to a safe position. This action effectively precludes the occurrence of a fuel _ handling accident. This does not preclude movement of a fuel assembly to a safe position.-

If moving irradiated fuel assemblies while-'in' MODE 5' or 6, LCO 3.0.3 would not specify any action. If moving irradiated fuel assemblies while' in MODES 1, 2, 3, and 4, the fuel- movement .is inoependent of reactor operations.

Therefore, inability to suspend movement of irradiated fuel assemblies is not sufficient reason to require a reactor

--shutdown .

SURVE'l. LANCE

-REQU6TMENTS-SR 3.7. .1 geg This SR verifies suff' ient5fuel :t:r:;; pool- water -is available in the ev t of-a fuel handling accident. The water 'ievel in the uel :t:r:g; pool must be checked

_ periodically. The 7 day Frequency is appropriate because (continued)

WOG STS B 3.7-79 , Rev 1, 04/07/95 Rev.8

. . . ___j

Speni Fuel ',;er;;; Pool Water Level' B 3.7.h I4 BASES 14 SURVEILLANCE SR 3.7.M.1 (continued).

REQUIREMENTS the volume in the pool is normally stable. Water level changes are controlled by plant procedures and are acceptable based on operating experience.

Duoing refueling operations, the level in th fuel :t;r.ge

! pool is in equilibrium with the refuelin9 6tt!b@, and the level in the refueling is checked daily in accordance withSR3.9.@.1. cavit G ca vs whenThe are.

7 h hydt uticony cc39ted l

REFERENCES 1. @ SAR, Section'T9.1.2 K

'? h SAR, Section9 9.1.3 f

3. h5AR, Sectionil5.7.4[May 1912 4.

5, Regulatory Guide 1.25, T av O f 10 CFR 100.11.

g WOG STS B 3.7-80 Rev 1, 04/07/95 Rev.8

~

.8HurE 3.7 -C1O % .S.7.lMpag)es_-

8 3 7 -84 ,,

repkced by ecces fuel Storage Pool Boron Concentration Cncert 0 3.~/ 81 A B 3. 7.

If B3 PLANT SYSTEMS B 3.7. Fuel Storage Pool Boron Concentration I

BASES BACKGROUND In ,the Maximum Density Rack (MDR) [(Refs. I nd 2)) design, the spent fuel storage pool is divided int two separate and istinct regions which, for the purpose criticality nsiderations, are considered as separ e pools.

[R ion 1), with [336] storage positio s, is designed to acc odate new fuel with a maximum U-235 or spent fuel regardless of arichment of [4.65) wtX discharge fuel burnup. [ Region 2), with [2670 designe to accommodate fuel of] s orage positions, is rious initial enrichments

'which hav accumulated minimum rnups within the acceptable domain acco ding to Figure [3 .17-1),intheaccon.;anying LCO. Fuel a emblies not me ing the criteria of Figure [3.7.1 1) shall be paragraph 4.3.1 cred in accordance with in Sect n 4.3, Fuel Storage.

The water in the s ent uel storage pool normally contains soluble under boron, actual operati whic r suits in large suberiticality margins conditions. However, the NRC guidelines, based u on he accident condition in which all soluble poison is ssume to have been lost, specify that the limiting k of 0.95 evaluated in the absence of soluble boron.,, ence, the esign of both regions is based on the use of nborated wate which maintains each region in a subtrit' cal condition dur'ng normal operation with the regions ful y loaded. The doub contingency principle discussed n ANSI N-16.1-1975 an the April 1978 NRC letter (Ref. 3) allows credit for soluble oron under other abnonn or accident conditions, sin e only a single accid t need be considered at one ti

. For example, the most severe accident scenario is associ ted with the mo ment of fuel from m'sloading of a fuel as[sembly in [RegionRegion . This could I to Regio 2], and otentially increase the criticality of [Re ion 2). To mitigate these postulated criticality relate accidents, boron is dissolved in the pool water. Safe op ation of the MDR with no movement of assemblies may therefor be achieved by controlling the location of each assembly in a ordance with LCO 3.7.17, " Spent Fuel Assembly Storage." P 'or to movement of an assembly, it is necessary to perform SR 3.7.16.1.

(continued)

WOG STS B 3.7-81 Rev 1, 04/07/95 Rev.B

h Fuel Storage Pcol Boron Concen8 ration' B - 3. 7. 4 l( -

BAS (c'ontinued)

/

APPLIC E SAFETY YSES Most.accidentconditionsdonotresultinanincresein the activity of either of the two regions. Exam es of these accident conditiens are the loss of cooli (reactivity increase with decreasing water den ity)--and the dropping of a fuel assembly on the top of th rack.

However, accidents can be postulated.that c Id increase the i

reettivity. ,This increase in reactivity i unacceptable

ith unborated water in the storage pool Thus, for-these a

l ident occurrences,-the presence of s uble boron in the-L st age pool prevents criticality in b h regions. The post 'ated accidents are basically o two types. A fuel asse could be incorrectly trans rred from Region (e.g.,; an unitradiated- f el assembly ro(Region 1 to insuffic %ntly depleted fuel ass ly). The second an

.postulatec type of ccidents is-associa d with a fuel assembly which is dr ped adjacent to e fully loaded-(Region 2) storage rack. This could ha effect on (Reg n 2). Howe r,=a small positive reactivity the: negative reactivity effect of the so uble' bor compensates for the increased reactivity caused y-eit r one of the two postulated accident scenarios. Th accident analyses is-provided in the FSAR,_Section [1 .4) (Ref. - 4) . -

The concentratior -di olved boron in the fuel storage pool satisfies Ce' erion of the NRC Policy Statement..

LCO The fuel sto age pool' borun con ntration is-required to be 2.[2300) p, . The specified con ntration of_ dissolved boron in ie fuel storage: pool pre rves the assumptions:

used in e analyses of the potenti critica1Laccident scenar' s as described in Reference 4 of di olved baron is the minimum requ'This concentration ed concentration for fuel assembly storage and movement withi the fuel storage po .

APPLICABILITY .This LCO applies whenever fuel assemblies are s red in the

- spent fuel storage pool, until a complete spentL el_ storage-pool verification has been performed following the ast movement of fuel assemblies in-the spent fuel stora pool.

This LCO does not apply-following the verification, s ce the fuel verification assemblies. would confirm that there are no mislo ed With no further fuel assembly movement in (continued)

WOG STS-B 3.7-82 Rev 1, 04/07/95 ReN. B

Fuel Storage Pool Boron Concentration 8 3.7.%

If B ES APPLIC ILITY . progress, there is no potential for a misloade' ^"'

(cont ved) assembly or a dropped fuel assembly.

-ACTIONS- - A .J . A. 2.1. and A. 2. 2 The Required Actions _are modified by a No indicating that O 3.0.3 does not apply.

Who the concentration of boron in th fuel storage pool is less han required, immediate_ actio must be taken to preclu the occurrence of an acci nt or to mitigate the consequ ces of an accident in pr ress. This is most

' ' efficient achieved-by immedia ly suspending the movement of fuel as oblies. The conce ration of boron is restored simultaneous with suspendi movement of fuel assemblies.

An acceptable iternative 1 to verify by administrative  ;

means that the el stora pool verification has been performed since -t e last vesent of fuel assemblies in the-fuel storage pool. Ho ver, prior to resuming movement of::

fuel- assemblies, the ncentration of boron must be restored. This does t preclede movement of a fuel assembly to a safe osi ion.

If the LCO is n met whil moving irradiated fuel as::emblies - in DE 5 or 6, L 0 3.0.3 would not be applicable. f moving irradi ed fuel assemblies while in-MODE 1, 2, ', or 4, the fuel mo ement is independent of reactor o ration. Therefore, i bility to suspend movemeat of fuel semblies is not sufficie reason to require a reactor shutdown.

f is SURVEILLANCE SV 3. 7.M.1 REQUIREMENTS This SR verifies that the concentration of b on in the fuel storage pool is withir, the required limit. As long_as this SR is met, the analyzed accidents are ' fully add ssed. The 7 day Frequency is appropriate because no major replenishment of pool water is expected to take pla e over such a short period of time.

/

/ (continued)

WOG STS B 3.7-83 Rev 1, 04/07/95 Rex. 8 a

Fuel Storage Pool Boron Cencentratio'n '

B 3.7.}6 If BASES (continued) j/

REFERENCE "-' 1. Callaway FSAR, Append'ix 9.lA, "The Maximum Densit ~~

Rack (MDR) Design Concept.'

2. Description and Evaluation for Proposed Chan s to Facility Operating Licenses DPR-39 and DPR- (Zion PowerStation). __

3. ' Double contingency principle of ANSI N1 .1-1975, as specified in the April 14, 1978 NRC 1 ter Section 1.2) and implied in the pro sed revision to

. R ulatory Guide 1.13 (Section 1.4, ppendix A).

4. FSA Section[15.7.4].

l t

- WOG STS B 3.7-84 Rev 1, 04/07/95 Rev.B

BASES-INSERT (S)

SECTION 3.7 Bases 3.7.15 INSERT B 3.7 81A (P 3 ) (Page 1 of 7)

Spent Fuel Pool Boron Concentration B 3.7.15 B 3.7 PLANT SYSTEMS" B 3.7.15 Spent Fuel Pool Boron Concentration BASES r =-

BACKGROUND The spent fuel pool provides for storage of various Westinghouse Optimized Fuel Assembly (0FA)-types of different initial fuel enrichments and exposure histories in two distinct regions. There are 23 separate racks which provide placement locations for a total of 2870 new or used fuel assemblies. Included in this are six specific storage l

! locations in one of the racks for placement of failed fuel assemblies.

fuel storage These cells. lecations are identified as the failed

4 Of the 23 racks. frur are ' designated

" Region 1" with the remaining 19 racks asignated as-g ce

" Region 2". The analytical methodology used to develop the criticality analyses has been reviewed and approved by the NRC (Ref. 1). '

l l "- Region 1 racks contain 392 cells which are analyzed for storing Westinghouse OFAs in an "All-Cells" arrangement (that is, the criticality analysis assumes that spent fuel assemblies reside in all available cell-locations, with the exception.of the boundary requirements). The stored fuel assemblies may contain an initial nominal enrichment of s 4.7 weight percent U-235 (without Integral Fuel Burnable .!

Absorbers (IFBAs) installed) up to.an initial nominal enrichment'of s 5.0 weight percent U-235. provided that the requirement for a minimum number of 16 IFBAs is met (Ref. 2). The IFBAs~ are requi d to have, as a minimum, 'a boron loading of 1.0X. equal-to an amount of 1.5 mg B"/ inch. This is the minimum standard poison material loading offered by Westinghouse for 17X17 0FAs.

Region 2 racks contain 2472 cells which are also analyzed for. storing Westinghouse OFAs in a combination of storage configurations. These patterns are:

1) "All Cells" Storage:

2) "3-out-of-4 Checkerboard" Storage; and

3) "2-out-of-4 Checkerboard" Storage.

(continued) 10/10/97 Revision B

BASES INSERT (S)

SECTION 3.7 Bases 3.7.15 INSERT B 3.7 81A (P 3 ) (Page 2 of 7) Spent Fuel Pool Boron Concentration BASES B 3.7.15 BACKGROUND For the "All Cells" storag configuration, the stored fuel (continued) assemblies may contain an initial nominal enrichment of s 1.14 weight percent U-235 (without taking credit for fuel burnup or radioactive decay of fuel constituents) u) to an initial nominal enrichment of s 5.0 weight percent 'J-235, when fuel burnup and radioactive decay of fuel constituents are credited.

For the "3-out-of-4 Checkerboard" storage configuration. the stored fuel assemblies may contain an initial nominal enrichment of s 1.64 weight percent U-235 (without taking credit for fuel burnup or radioactive decay of fuel constituents) weight percent u) to an initial nominal enrichment of s 5.0 J-235, when fuel burnup and radioactive decay of fuel constituents are credited. In this storage pattern, g there can be no more than three stored assemblies in any 2X2 matrix of cell lattices.

For the "2-out-of-4 Checkerboard" storage configuration, the stored fuel assemblies may contain an initial nominal enrichment of 5 4.10 weight percent U-235 (without taking credit for fuel burnup) up to an initial nominal enrichment of s 5.0 weight percent U-235 when fuel burnup is credited.

In this storage pattern, no two fuel assemblies may be stored " face adjacent" (that is, there must be an empty cell opposite each face of the fuel assembly).

The water in the spent fuel pool normally contains soluble boron which results in large subcriticality margins under actual operating conditions.

APPLICABLE NRC approved methodolo SAFETY ANALYSES criticality analyses (gies Ref.were usedfuel 1). The to develop handlingthe accident analyses are provided in Reference 3. The accident analyses

• for criticality and spent fuel pool dilution are c :Mded in References 2 and 4. respectively.

(continued) 10/10/97 Revision B

l BASES-INSERT (S)

SECTION 3.7 Bases 3.7.15 INSERT B 3.7 81A (P n ) (Page 3 of 7) Spent Fuel Pool Boron Concentration BASES B 3.7.15 1

. APPLICABLE The criticality-analy s f r the spent fuel assembly storage SAFETY ANALYSIS- racks confirm that k remain

-(continued)- uncertainties and tol,e,rances)sat<a1.0 95%(including probability with a 95% confidence level-(95/95 basis), based on the accident condition of the pool being flooded with unborated water.

Thus, the design of both regions assumes the use of  ;

unboratetwater while maintaining stored fuel in a

. subcritical condition.

i However, the presence of soluble boron has been credited to f

provide adequate safety margin to maintain spent fuel i

1 assembly storage rack k,,, s 0.95 (also on a 95/95 basis) for all postulated accident scenarios involving dropped or 4

misloaded fuel assembhes and loss of spent fuel pool temperature control. Crediting the presence of soluble 00 boron for mitigation of these scenarios is acceptable based 3 on applying the " double contingency principle" which states

u that therr 's no-requiremerc to assume two unlikely, t

independent, concurrent events to ensure protection against- '

7 a criticality accident (Refs. 5.and 6).

1 The accident analyses address the following five postulated f scenarios:

4 i

1) fuel assembly drop on top of rack:

2) fuel assembly _ drop between rack modules:- ,

3) fuel assembly drop between rack modules:and spent fuel pool wall:.

E 4) change in spent fuel pool water temperature: and

5) fuel. assembly _ loaded contrary to placement j restrictions.

Of these, only the last two have.the capacity to increase reactivity beyond the analyzed condition.

Calculations were performed to determine the reactivity change caused by a change in spent fuel pool-water temperature outside the normal range _(50 - 160 F). For the change in spent fuel pool water temperature accident, a temperature range of 32 - 240*F is considered. In all cases, additional reactivity nargin-is available to the-0.95 k,,, limit to allow for temperature accidents. The temperature change accident can occur-at any time during operation of the spent fuel pool.

(continued) 10/10/97 Revision B

BASES INSERT (S)

SECTION 3.7 Bases 3.7.15 INSERT B 3.7 81A (P u ) (Page 4 of 7) Spent Fuel Pool Boron Concentration B 3.7.15 BASES APPLICABLE For the fuel assembly misl ad accident. calculations were SAFETY ANALYSIS performed to show the largest reactivity increase caused by (continued) a Westinghouse 17X17 0FA fuel assembly misplaced into a storage cell for which the restrictions on location, enrichment, or burnup are not satisfied. The assembly misload accident can only occur during fuel handling operations in the spent fuel pool.

For the above postulated accident conditions. the double contingency principle can be applied. Specifically, the 3resence of soluble boron in t1e spent fuel pool water can

)e assumed as a realistic initial condition since not assuming its presence would be a second unlikely event.

Spent fuel pool soluble boron has been credited in the criticality safety analysis to offset storage rack and fuel assembly tolerances, calculational uncertainties.

4 uncertainty associated with burnup credit and the reactivity 6 increase caused by postulated accident conditions.

O k

Based on the above discussion. should a spent fuel pool water temperature change accident or a fuel assembly misload accident occur in the Region 1. Region 2. or failed fuel storage cells, k will be maintained s to 0.95 due to the presence of at l,,a,st e 550 ppm (no fuel handling) or 1650 ppm (during fuel handling) of soluble boron in the spent fuel pool water.

A spent fuel pool dilution analysis (Ref, 4) has been performed as required by Reference 7. The analysis assumes an initial baron concentration of 2000 ppm. The dilution analysis concludes that an unplanned or inadvertent event that would result in the dilution of the spent fuel pool boron concentration from 2000 ppm to 550 ppm (minimum non-accident boron concentration) is not credible.

(continued) 10/10/97 Revision B

BASES INSERT (S)

SECTION 3.7 Bases 3.7.15 INSERT B 3.7 81A (P 3 ) (Page 5 of 7) Spent Fuel Pool Boron Concentration BASES B 3.7.15 APPLICABLE Interface recuirements have been established to ensure k,n SAFETY ANALYSIS is maintainec within the appropriate limits. There are (continued) interface requirements between Region 1 racks, between Region 1 and Region 2 racks, between Region 2 racks, and within racks between different checkerboard configurations.

These requirements are necessary to account for unique geometries and configurations which exist at the interfaces.

Interface requirements exist between adjacent racks to account for the potential reactivity increase in 3-out-of-4 and 2-out-of-4 storage configurations along the interface with non-aligned racks.

The concentration of dissolved boron in the spent fuel pool satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

4 LCO The spent fuel pool boron concentration is required to be j = 2000 ppm. The specified concentration of dissolved boron s in the spent fuel pool preserves the assumptions used in the analyses. This concentration of dissolved boron is the minimum required concentration to permit fuel assembly storage within the spent fuel pool.

APPLICABILITY This LC0 applies whenever fuel assemblies are stored in the spent fuel pool.

The presence of soluble boron (in various concentrations) is assumed in the criticality analyses and is credited for ensuring that spent fuel pool k , will be maintained 5 0.95 at a 95% confidence level for aN storage configurations.

The 2000 ppm minimum boron concentration is also an initial condition in the spent fuel pool dilution analysis.

Therefore. the restriction on soluble boron concentration in the spent fuel pool water must be maintained at all times when fuel assemblies are stored in the spent fuel pool.

(continued) 10/10/97 Revision B

BASES INSERT (S)

SECTION 3.7- Bases 3.7.15 INSERT B 3.7 81A (P n ) (Page 6 of 7) Spent Fuel Pool Boron Concentration BASES (continued)

B 3.7.15

' ACTIONS The ACTIONS have been modified by a Note indicating that LC0 3.0.3_does not apply.

-A.1 and-A.2-When the concentration of boron in the spent fuel pool is '

less than required, immediate action must be taken to L

preclude the occurrence of an accident or to mitigate the consequences of an accident in progress. This is most l

efficiently achieved by imediately suspending the movement 4 of fuel assemblies. This does not preclude movement of a fuel assembly to a safe position. Imediate actions.are.

also taken to restore spent fuel pool boron concentration.to

= 2000 ppm.

If moving fuel assemblies while in MODE 5 or 6.- LCO 3.t 3 would not specify any action. If moving fuel assemblies -

N> while.in MODES 1. 2. 3, and 4. the fuel movement is

. independent of reactor operations. Therefore, inability to (E .- suspend movement of fuel-assemblies is not sufficient reason to require a reactor-shutdown.

SURVEILLANCE SR 3.7.15.1 REQUIREMENTS This SR verifies that the ccacentration of boron in the s)ent fuel pool: is = 2000 ppm As long as:this SR is met.

tieanalyzedaccidentsarefuilyaddressed.

The 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> frequency is appropriate based ~on operatin9 experience and because significant changes in t1e boron concentration in the spent fuel pool are difficult to -

prode:e without detection considering the.large_ volume of water- contained in tne spent fuel pool. - An analysis has-concluded that a spent fuel-pool boron dilution event of sufficient magnitude to reduce boron concentration below the- '

minimum non-accident requirement is not credible.(Ref. 4).

(continued) 10/10/97 Revision B

BASES INSERT (S)

SECTION 3.7 Bases 3.7.15 INSERT B 3.7 81A (P u ) (Page 7 of 7) Spent Fuel Pool Boron Concentration BASES (continued)- B 3.7.15

~

REFERENCES 1. WCAP-14416-NP-A " Westinghouse Spent Fuel Rack Criticality Anal November, 1996. ysis Methodology." Rev. 1. dated

2. CAC-97-162 " Byron and Braidwood Spent Fuel Rack Criticality Analysis Using Soluble Boron Credit."

dated May. 1997.

3. UFSAR. Section 15.7.4.

l-

4. " Byron /Braidwood Spent Fuel Pool Dilution Analysis." )

Rev' 3. dated June 17, 1997

5. Double contingency principle of ANSI N16.1 - 1975, as g specified in the April 14, 1978 NRC letter (Section 1.2) and implied in the proposed revision to g Regulatory Guide 1.13 (Section 1.4. Appendix A).

6.

ANSI /ANS 8.1 - 1983 "American National Standard for Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors."

7. Safety Evaluation Report (SER) dated October 25. 1996, issued by the Office of Nuclear Reactor Regulation for ,

Topical Report WCAP-14416-NP-A " Westinghouse Spent Fuel Rack Criticality Analysis Methodology."

10/10/97 Revision B i

_-- - w vvw --- ~

8 3.~1- B5b B D.*l-O ) p replaced by Schccc 3s Inceri 0 3.I-85A Spent Fuel Assembly Storag'd B 3.7.N N

3.7 PLANT SYSTEMS B 3. .N Spent Fuel Assembly Storage ,

I(a BASES BACKGROUND-IntheMaximumDensityRack(MDR)((Refs.1a 2)] design, the. spent fuel storage pool is divided into wo separate-and distinct regions which, for the purpose of riticality onsiderations, are considered as separa ' pools, ac egion 1), with [336) storage position , is designed to U-2 odate new fuel with a maximum e ichment of [4.65) wty,

, or spent fuel regardless of t discharge fuel burnu [ Region 2), with [2670] s rage positions is design l to accommodate fuel of v ious initial enr,ichments l

.hich w ha accumulated minimum rnups within the acceptable domain ac rding to Figure 3.7 1, in the accompanying LCO. Fuel usemblies not me ing the criteria of Figure [3.71-1) shall be cred in accordance with paragraph 4.3'."A 3 in Secti n 4.3, Fuel Storage.  !

The water in the ent uel storage pool normally contains' soluble boron, whic suits in lar ,

under actual operati conditions. ge subcriticality However, the NRCmargins guidelines, based on he accident condition in which all the limiting k of 0.95 to have been lost, specify that soluble poison i assume soluble boron., Hence, the evaluated sign of in the absence of both regions is based on the use o unborated wate which maintains each region in a subtri ical condition dur g normal operation with the regions f ly loaded. The doubl discusse in- ANSI N-16.1-1975 and contingency principle he April 1978 NRC letter (Ref. . allows credit for soluble ron under other abno 1 or accident conditions, sin only a single ace' mo ent need be considered at one ti For example, the severe accident.

vesent of fuel from scenario is associa d with the-isloading of a fuel as[ Region 1 to Region ), and accidental sembly in [ Region 2). This could potentially increase the criticality of [Regt n 2). To mitigate these postulated criticality related cidents, boron is dissolved in the pool water. Safe ope tion of the MDR with no movement of assemblies may therefore achieved by controlling the location of each assembly in act dance with the accompanying LCO. Prior to movement of an assembly, it is necessary to perform SR 3.7.16.1.

-/ ,

(continued)

WOG STS B 3.7-85 Rev 1, 04/07/95 Re.t B

Spent Fuel Assembly Storage, B 3.7.g BAS (continued)

APPLICA E ~

The hypothetical accidents can only take place durirl r as SAFETY LYSES a result of the movement of an assembly (Ref 4). for these accident occurrences, the presence of soluble bor 'n in the spent fuel storage pool (controlled by LCO 3.7.1, " Fuel Storage Pool Boron Concentration") prevents er icality in both regions. By closely controlling the mov ent of each assembly and by checking the location of ea assembly after ovement, the time period-for potential ac dents may be mited to a small fraction of the total perating time.

D ing the remaining time period with ng. potential for M

acc ents, the operation may be under he auspices of the acco anying LCO.

The con guration of fuel assembli in the fuel storage pool satt fies Criterion 2 of the RC Policy Statement.

LCO ihe restrictio on the pla ment of fuel assemt, lies within the spent fuel p ol, in at ordance with Figure 3.7.17-1, in the accompanying 0, en res the k ,, of the spent fuel storage pool will a way remain < 0,.95, assuming the pool to be flooded with unbo ed water. Fuel assemblies not meeting the criteria Figure [3.7.W g l] shall be stored in accordance with Spe ifi ation 4.3.1.1 in Section 4.3.

APPLICABILITY This LCO appl _s whenever an fuel assembly is stored in

[ Region 2) o the fuel storage col.

\

ACTIONS A.1 Requ' red Action A.1 is modified by a No e indicating that LC .3.0.3 does not apply, en the configuration of fuel assemblies s ored in

[ Region 2) the spent fuel storage pool is no in accordance withFigure3.7.lgl,orparagraph4.3.1.1,th immediate action is to initiate action to make the necess y fuel assembly movement (s) to bring the configuration 1 to

- compliance with Figure 3.7. N -1 or Specification 4 .l.%.

14,

/

/ -

/

(continu d)

\

WOG STS B 3.7-86 Rev 1, 04/07/95 Rev 8

la Spact Fuel Assembly Storage B 3.7.F7

-B ES ACTIO A_1 (continued) /

If unable to move irradiated fuel assemblies whil in H0DE 5 or 6, LCO 3.0.3 would not be applicable. If una e to move irradiated fuel assemblies while in MODE 1, 2, , or 4, the

' action is independent of reactor operation, erefore, inability to move fuel assemblies is not suf icient reason to require a reactor shutdown.

SURVEILLANCE SR\7. .1 REQUIREMENTS This SR erifies by administrative ans that the initial enrichmen and burnup of the fue assembly is in accordance with Figur (3.7.Mel) in the a ompanying LCO.

assemblies i the Eacceptabl range of Figure 3.7.17-1, For fuel performance o this SR will sure compliance with Specification.4 .1 /.

4 REFERENCES 1. Callaway FSAR, A endix 9.lA, "The Maximum Density Rack (MDR) Desi neept."

2. Description d Evalu ion for Proposed Changes to.

Facility Op ating Lic ses DPR-39 and DPR-48 (Zion

___ Power Sta on). ___

3. Double entingency principi of ANSI N16.1-1975, as specif ed in the April 14, 19 NRC 7ette.r (Sec on 1.2) and implied in t proposed revision to Re atory Guide'l.13 (Section 1. , Appendix A).

4. AR,Section(15.7.4].

a d

WOG STS 3 3.7-87 Rev 1, 04/07/95 Rev. s

BASES INSERT (S)

SECTION 3.7 Bases 3.7.16 INSERT B 3.7 B5A (P n ) (Page 1 of 7)

Spent Fuel Assembly Storage

, B 3.7.16 B 3.7 PLANT SYSTEMS' B 3.7.16 Spent Fuel Assembly Storage BASES BACKGROUND The spent fuel pool provides for storage of various Westinghouse Optimized Fuel Assembly (OFA) types of different initial fuel enrichments and exposure histories in two distinct regions. There are 23 separate racks which provide placement locations for a total of 2870 new or used fuel assemblies. Included in this are six specific storage locations in one of the racks for placement of ' ailed fuel assemblies. These locaticns are identified as the failed fuel storage cells. Of the 23 racks, four are designate O " Region 1" with the rernaining 19 racks designated as

" Region 2". The analytical methodology used to develop the '

E criticality analyses has been reviewed and approved by the t NRC (Ref. 1).

Region 1 racks contain 392 cells which are analyzed for storing Westinghouse 0FAs in an "All Cells" arrangement (that is, the criticality analysis assumes that spent fuel assemblies reside in all available cell locations, with the exception of the boundary requirements). The stored fuel assemblies may contain an initial nominal enrichment of s 4.7 weight percent U-235 (without Integral Fuel Burnable i Absorbers (IFBAs) installed) up to an initial nominal enrichment of s 5.0 weight percent U-235 provided that the requirement for a minimum number of 16 IFBAs is met (Ref. 2). The IFBAs are required to have. as a minimum, a bcron loading of 1.0X. equal to an amount of 1.5 mg B"/ inch. This is the minimum standard poison material loading offered by Westinghouse for 17X17 0FAs.

Region 2 racks contain 2472 cells which are also analyzed for storing Westinghouse OFAs in a combination of storage configurations. These patterns are:

1) "All Cells" Storage:

2) "3-out-of-4 Checkerboard" Storage: and

3) "2-out-of-4 Checkerboard" Storage.

(continued) l 10/10/97 Revision B

BASES INSERT (S)

SECTION 3.7 Bases 3.7.16 INSERT B 3.7 85A (P 3 ) (Page 2 of 7) Spent Fuel Assembly Storage BASES B 3.7.16 BACKGROUND For the "All Cells" storag configuration, the stored fuel (continued) assemblies may contain an initial nominal enrichment of .  !

s 1.14 weight percent U-235 (without taking credit for fuel burnup or radioactive decay of fuel = constituents) u) to an -

initial nominal enrichment of s 5.0 weight percent- J-235, when fuel burnup and radioactive decay of fuel constituents are credited.

-For the "3-out-of-4 Checkerboard" storage configuration, the stored fuel assemblies may contain an initial nominal enrichment of s 1.64 weight percent U-235 (without taking credit for fuel burnup or radioactive decay of fuel constituents) u) to an initial nominal enrichment of s 5.0 weight percent :J-235 when fuel burnup and radioactive decay of fuel constituents are credited. In this storage pattern, there can be no more than three stored assemblies in any 2X2 matrix of cell lattices.

- For the "2-out-of-4 Checkerboard" storage configuration, the

@ stored fuel assemblies may contain an initial nominal enrichment of 5 4.10 weight percent U-235 (without taking credit for fuel burnup) up to an initial nominal enrichment of s 5.0 weight percent U-235, when fuel burnup is credited.

In this stcrage pattern, no two fuel assemblies may be stored " face adjacent" (that is, there must be an empty cell opposite each face of the fuel assembly).

The water. in the spent fuel pool normally contains soluble boron which res"'ts in'large subtriticality margins under actual operatin3 conditions.

APPLICABLE NRC appr'oved methodolo SAFETY ANALYSES criticality analyses-(gies Re were

f. 1) . used to develop The fuel the handling accident-analyses'are provided in Reference 3. The accident analyses for-criticality and spent fuel pool dilution are provided in References 2 and 4. respectively.

1 (continued) 10/10/97 Revision B

BASES INSERT (S)

SECTION 3.7 Bases 3.7.16 INSERT B 3.7 85A (P3 ) (Page 3 of 7) Spent Fuel Assembly Storage B 3.7.16 BASES

-APPLICABLE The criticality analyses f r the spent fuel assembly storage SAFETY ANALYSIS racks confirm that k remain (continued) uncertainties and tol,e,rances)sat<a1.0 95% (including probability with a 95% confidence level (95/95 basis). based on the accident condition of the pool being flooded with unborated water.

Thus, the design of both regions assumes the use of unborated water while maintaining stored fuel in a subcritical condition.

i However, the presence of soluble boron has been credited to provide adequate safety margin to maintain spent fuel assembly storage rack k , s-0.95 (also on a 95/95 basisi for all postulated accident,, scenarios involving dropped or misloaded fuel assemblies and loss of spent fuel pool temperature control. Crediting the presence of soluble 4- boron for mitigation of these scenarios is acceptable based 0 on applying the " double contingency principle" which states that there is no requirement to assume two unlikely,.

k independent, concurrent events to ensure protection against a criticality accident (Refs. 5 and 6).

The accident-analyses address the following five postulated scenarios:

1) fuel assembly drop on top of rack:

2) fuel assembl drop between rack modules:

3) fuel assemb1 drop between rack modules and spent fuel pool wa 1:

4) change in spent fuel pool-water temperature; and

5) fuel assembly loaded contrary to placement restrictions. l ll Of these, only the last two have the capacity to increase reactivity beyond the analyzed condition.

Calculations were performed to determine the reactivity change caused by a change =in-spent fuel pool water temperature outside the normal range (50 - 160 F). For the change in spent fuel pool water temperature accident, a temperature range of.32 - 240'F is considered. In all cases, additional reactivity margin is available to the 0.95 k,,, limit:to allow for temperature accidents.

The temperature change accident can occur at any time during operation of the spent fuel pool.

(continued) 10/10/97 Revision B

BASES INSERT (S) -

-SECTION 3.7 Bases 3.7.16 INSERT B 3.7 85A (P u ) (Page 4 of 7) Spent Fuel Assembly Storage BASES B 3.7.16 APPLICABLE .

For the fuel assembly misload accident, calculations were SAFETY ANALYSIS performed to show the largest reactivity increase caused by

.(continued) a Westinghouse 17X17 0FA fuel assembly misplaced into a storage cell for which the restrictions on location, enrichment, or burnup are not satisfied. The assembly misload accident can only-occur during fuel handling operations in the. spent fuel pool.

For-the above postulated accident conditions. the double contingency principle can be a) plied. Specifically, the  ;

3resence of soluble boron in tie spent fuel pool water can '

3e assumed as-a realistic initial condition since not assuming its presence would be a second unlikely event.

Spent fuel pool soluble boron has been credited in the criticality safety analysis to offset storage rack and fuel 00 assembly tolerances, calculatior.al L scertainties.

g_

uncertainty associated with burnup credit and the reactivity increase caused by postulated accident conditions.

E Based on the above discussion..should a-'s nt fuel pool-water temperature change accident or a fu assembly misload accident occur in the Region 1. Region 2. or failed fuel

' storage. cells, k will be maintained s to 0.95 due to the presence of at l,,a,st e 550 ppm (no fuel handling) or 1650 ppm (during fuel handling) of soluble boron in the spent fuel pool water.

A spent fuel pool dilution analysis (Ref. 4)-has been performed as required by Reference 7. The_ analysis assumes an initial-boron concentration of 2000 ppm. The dilution analysis concludes that an unplanned or inadvertent event that would result in the' dilution of the spent fuel pool boron concentration from 2000 ppm to 550 ppm (minimum non-accident boron concentration) is not credible.

(continued) 10/10/97 Revision B a - _

BASES INSERT (S)

SECTION 3.7 Bases 3.7.16

_LNSERT B 3.7 85A (P u ) (Page 5 of 7) Spent Fuel Assembly Storage BASES B 3.7.16 APPLICABLE Interface recuirements have been established to ensure k,,,

SAFETY ANALYSIS is maintainec within the appropriate limits. There are (continued) interface requirements between Region 1 racks, between Region 1 and Region 2 racks, between Region 2 racks. and within racks between different checkerboard configurations.

These requirements are necessary to account for unique geometries and configurations which exist at the interfaces.

Interface requirements exist between adjacent racks to account for the potential reactivity increase in 3-out-of-4 and 2-out-of-4 storage configurations along the interface with non-aligned racks.

The configuration of fuel assemblies in the spent fuel pool satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

a)

LC0 The restrictions on the placement of fuel assemblies within 3 the spent fuel pool in accordance with the requirements in Ce the accompanying LCO ensure that the k,,, of the spent fuel i

' pool will always remain < 1.0 assuming the pool is flooded with unborated water and s 0.95 assuming the presence of 550 ppm soluble boron in the pool.

In LCO Figures 3.7.16-1 and 3.7.16-2. the Acceptable Burwo Domain lies on. above, and to the left of the decay time line applicable to the fuel assembly to be stored. The decay time for that assembly is measured from the time since the assembly was last discharged.

In LCO Figure 3.7.16-3 the Acceptable Burnup Domain and the Unacceptable Burnup Domain are separated by a sirigle line because decay time is not credited in the 2-out-of-4 Checkerboard storage configuration. The Acceptable Burnup Domain lies on, above, and to the left of the line.

APPLICABILITY This LCO applies whenever fuel assemblies are stored in the spent fuel pool.

(continued) 10/10/97 Revision B

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BASES INSERT (S)

SECTION 3.7 Bases 3.7.16 INSERT B 3.7 85A (P u ) (Page 6 of 7) Spent Fuel Assembly Storage

-B 3.7.16 BASES (continued)

ACTIONS- The ACTIONS have been modified by a Note indicating that LCO 3.0.3 does not apply.

L1 When the configuration of fuel assemblies stored in the spent fuel of the LCO, pool is not -in accordance with the requirements immediate action must be taken to make the necessary fuel assembly movement (s) to bring the configuration into compliance.

If moving fuel assemblies while in MODE 5 or 6. LCO 3.0.3 '

would_not saecify any action. If moving-fuel assemblies while in M0)ES 1. 2. 3. and 4. the fuel movement is independent of reactor operations. Therefore. inability to g suspend movement of fuel assemblies is not sufficient reason  :

g to require a reactor shutdown.

.g '

SURVEILLANCE SR 3.7.16.1. SR 3.7.16.2. and SR 3.7.16.3 are performed REQUIREMENTS prior to storing the fuel assembly in the intended spent fuel pool sto age location. These frequencies are appropriate because compliance with the SR ensures that the relationship between the fuel assembly and its storage location will meet the requirements of the LCO and preserve.

the assumptions of the analyses.

SR 3.7.16.1 This SR verifies by administrative means that the initial nominal enrichment of the fuel assembly or a minimum number of 16 IFBAs is met to' ensure that the assumptions of the safety analyses are preserved SR 3.7.16.2 This SR verifies by administrative means that the combination of initial enrichment, burnup, and decay time of the fuel assembly is within the Acceptable Burnup Domain of Figure 3.7.16-1, 3.7.16-2 or 3.7.16-3 for the intended storage configuration to ensure that the assumptions of the se fety analyses are preserved.

(ctntinued) 10/10/97 Revision B l

1

BASES INSERT (S)

SECTION 3.7 Bases 3.7.16 INSERT B 3.7 85A (Pn )- (Page 7 of 7) Spent Fuel Assembly Storage BASES B 3.7.16 SURVEILLANCE SR' 3.7.16.3 REQUIREMENTS

'(continued) This SR verifies by administrative means that the interface requirements within and between adjacent racks are met to ensure that the assumptions of the safety analyses are preserved.

REFERENCES 1. WCAP-14416-NP-A " Westinghouse Spent Fuel Rack Criticality Analysis Methodology." Rev,1. dated November. 1996.

2. CAC-97-162 " Byron and Braidwwa Spent Fuel-Rack-Criticality Analysis Using Solr5 e Baron Credit."

dated May. 1997.

s 3. UFSAR. Section 15.7.4. i

4. " Byron /Braidwood Spent-Fuel Pool Dilution Analysis."

Rev.-3. dated June 17, 1997. '

5. 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).

6.

ANSI /ANS 8.1 - 1983 "American National Standard for Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors.~"

7. Safety Evaluation Report (SER) dated October 25, 1996, issued by the Office of Nuclear Reactor Regulation for Topical-Report WCAP-14416-NP-A " West'aghouse Spent Fuel Rack Criticality Analysis Methodology."

10/10/97 Revision B D

BASES JFDs l

i JUSTIFICATION FOR DIFFERENCES TO NUREG 1431 BASES  !

SECTION 3.7 . PLANT SYSTEMS  !

PLANT SPECIFIC CHANGES (P) i

. P3 During the development certain wording preferences. English language conventions, reformatting.- renumbering. or editorial rewording ,

consistent with plant specific nomenclature were adopted. As a result, the Technical- Specifications (TS) should be more readily readable by, and therefore understandable to, plant operators and other users. -;

During this reformatting, renumbering, and rewording process, no t technical changes (either actual or interpretational) were made to the TS unless they were identified and justified.

• i i

P, Consistent with the modifications to the NUREG-1431 LC0's, the Bases are modified to delete the LC0 and Applicability requirements for AF SG  :

PORVs and CST.

P3 . (Byron Only) The Byron IST program references ANSI /ASME OM-1-1987 and I applicable addenda, This is added for consistency with existing documentation.

P. NUREG information with respect to a turbine driven AF pump has been deleted. Byron and Braidwood's AF system consists of a motor driven and a diesel driven pump.

P. The MSIV Bases are revised to reflect the B3ron and Braidwood CTS.

design and analyses (e.g., actuation signals..etc..).

P. The ITS Bases indicating an MSIV and Feedwater (FW) Isolation Valves Applicability exception when the MSIVs or FW lsolation Valve are closed and de-activated is deleted. This is consistent with the proposed ITS '

LC0 Applicability.

O, NUREG SR 3.7a2.1 and SR 3.7.3.1 to verify the closure' time of each MSIV.

or FW lsolation Valve on an actual or simulated signal is separated into two separate surveillances, one for the closure time and one for-actuation on an actual or simulated actuation signal. This change maintains consistency with the containment isolation valve specifications. The NUREG fre specified as the IST proyr,n. quency of the closure time test SR isThe IST p '

- actuation test. Appropriate Bases modifications are provided.

P, The Bases for LCO 3.7.S and LC0 3.7.6 have been revised to reflect the Byron and Braidwood AF system and CST design.(analysis and references).

The AF System consists of a motor driven and a diesel ariven pump. All references to a turbine driven AF pum Information with respect to the available AF: sources i.e.. (p are deleted.-

CST and Essential Service Water) is adoc

n..

- $l BYRON!BRAIDWOOD-UNITS 1&2 3.7 7 . 10/10/97 Revision B aae m -,y--, , , .-.

m.-,. - , , ,-, ., _r w - ..v.-- +

JUSTIFICATION FOR DIFFERENCES TO NUREG 1431 BASES SECTION 3.7 PLANT SYSTEMS P, The Bases for ITS LC0 3.7.14 have been revised to change the NUREG W- ,

wording. " Fuel Storage Pool Water" to PS>ent Fuel Pool Water". This

-? change makes the E3ses consi: stent with tie LCO and with plant is _ terminology.

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-BYRON /BRAIDWOOD UNITS 1 & 2 3.7 7a 10/10/97 Revision B a

JUSTIFICATION FOR DIFFERENCES TO NUREG 1431 BASES SECTION 3.7 PLANT SYSTEMS Pn CTS 3.7.6 Actions require verification that the ventilation system placed in operation be capable of being' powered by an Operable emergency power source. LCO 3.7.10 and 3.7.11 Required Action C 1.2 is added to incorporate this requirement. Appropriate Bases descriptions have been added.

CTS 3.7.6 Actions require suspension of positive reactivity additions, j in addition to thc suspension of Core Alterations and movement of '

irradiated fuel. LCO 3.7.10 and 3.7.11 Required Actions C.2.3 and 0.3 reflect this req"irement. Appropriate Bases descriptions have been added. ,

P,4 ITS SR 3.7.10.2. SR 3.7.12.2 and SR 3.7.13.2 Bases are revised to clarify general conformance to the applicable Regulatory Guides. This change is necessary since the UFSAR contains plant specific exception to the Regulatory Guides. This eliminates the potential for 1

misinterpretation of the licensing basis requirements. This is an editorial enhancement only and does not involve a technical change.

P n ITS SR 3.7.10.3. SR 3.7.12.3 and SR 3.7.13.4 Bases are revised to clarify that the scope of the SR includes alignment of the system.

P,,

CTS 4.7.6.a requires verification of the control room temperature every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. SR 3.7.11.1 is added to reflect this requirement and appropriate Bases are provided. .

P,, NUREG Bases 3.7.15 Required Action A.1 contains a statement that the M

Required Action A.1 is modified by a Note indicating that LC0 3.0.3 does j not apply. Since there is only One Required Action (A.1), the Note is being moved.

This is consistent with the NUREG format.

P a NUREG LCO 3.7.12 and associated Bases are revised to reflect the Byron and Braidwood Nonaccessible Area Exhaust Filter Plenum Ventilation System design and requirements. The system is common to both units and has three 50% capacity trains. The nature of the design and electrical and damper requirements potentially affecting more than one train, would necessitate excess detail in the ITS LCO statement. The ITS LCO statement is that three trains shall be OPERABLE. The Background

'd LCO Bases discussions provide the details with respect to system cesign including. train and electrical and damper interrelationships.

P,,

The Bases discussion associated with SR 3.7.12.4 is editorially revised consistent with the other HVAC (VC Filtration System and FHB Ventilation System) Bases discussions.

In addition. the NUREG statement that the SR is conducted concurrently with filter testing is not accurate.

BYRON /BRAIDWOOD UNITS 1 & 2 3.7 10 10/10/97 Revision B

}

i JUSTIFICATION FOR DIFFERENCES TO NUREG 1431 BASES SECTION 3.7 PLANT SYSTEMS P3 The Bases are revised to clarify that, if a particular room is isolated such that there is no potential for post accident fluids to pass through the room, or that room's ECCS equipment is not required. that room can be excluded from meeting the acceptance criteria of the SR. Because the Nonaccessible Area Exhaust Filter Plenum Ventilation System is comon to both units, the potential exists for one unit to be shutdown and extensive maintenance being conducted, such that the specified negative

)ressure in that room could not be met. As long as rooms where post

.0CA recirculation fluids can be postulated are capable of meeting the SR. the function ventilation System of the Nonaccessible Area Exhaust Filter Plenum is satisfied.

Pu NUREG Bases 3.7.16 and 3.7.17 have been replaced in their entire based on NRC letter dated A)ril 2. 1997 which isrued Amendment B6 for ron and Amendment 78 for 3raidwoed )r soluble boron in the spent fue pool O (SFP). Since the license amenc..ents we $ temporary in nature. Comed-letter dated June 30, 1997 3roposed changes to permanently take credit a:

for soluble boron in the SF). Additionally. Comed responded to the NRC's request for additional information in Comed letter dated September 25,-1997. Although not yet approved by the RC Comed has used the June 30, 1997 and the September 25, 1997 submittal revisions as

-the basis for the Bases for ITS LCOs 3.7.15 and 3.7.16.

Pn NUREG LC0 3.7.14 " Penetration Room Exhaust Air Cleanup System (PREACS)."

and associated Bases are not utilized. The Byron and Braidwood analyses only credit the Nonaccessible Area Exhaust Filter Plenum Ventilation i System (LC0 3.7.12) for filtration of post LOCA leakage. See UFSAR l Sections 6.5 and 15.6.5.

P u The Actions section of the Bases-for NUREG LC0 3.7.15 states, "When the initial conditions for prevention of an accident cannot' be met. . . , ."

This statement has been revised to read. "When the initial conditions-assumed in the accitient analysis cannot be met ... ." This change does not change any intent of the original statement, but provides a more accurate description.

4 P3 The Surveillance Requirements section of the Bases for NUREG LCO 3.7.15

( states. "During refueling operations, the level in the fuel storage pool

- 1s in equilibrium with the refueling canal, and ... ." 'This has been revised to read. "During refueling operations. the level in the spent fuel pool is in_ equilibrium with the refueling cavity when they are hydraulically coupled. ... ." This addition provides a more accurate description of the relationship between the spent fuel pool and the refueling cavity. In addition. the word " cavity" re places the NUREG word " canal". This is also a specific plant term.

BYRON /BRAIDWOOD UNITS 1 & 2 -3.7 11 10/10/97 Revision B

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JUSTIFICATION FOR DIFFERENCES TO NUREG 1431 BASES {

SECTION 3.7 PLANT SYSTEMS j P3 ITS LC0 3.7.4 and associated Bases are revised to refl$ct plant s nomenclature, plant "^ ific design and plant specific analyses. pecific The Atmospheric Dutro Valves are referred to as the Steam Generator (SG) '

Power Operated Relief Valves (PORVs) at Byron and Braidwood. The valves  !

are electrohydraulic, powered by the Class 1E Buses, and have handpumps, i which provide for manual operation. The applicable safety analyses-section is revised to reflect the Byron and Braidwood SGTR analyses presented in UFSAR Section 15.6.3.

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a h l : BYRON /BRAIDWOOD - UNITS 18' 2 3.7 11a 10/10/97 Revision B

JUSTIFICATION FOR DIFFERENCES TO NUREG 1431 BASES SECTION 3.7 PLANT SYSTEMS Pu (Byron Only) NUREG Bases 3.7.9 Actions section is revised to reflect the Byron CTS. design and analysis requ'irements. . The Byron VHS design includes two towers (each with four fans)-and a diesel powered makeup pump for each basin which draws on the Rock River. The design basis analysis for the VHS assumes only 6 fans are Operable. Therefore "requ? red" is used when referring to fan requirements. Byron CTS only allow 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> for-restoration of-an inoyerable required fan:-therefore.

Condition A is revised consistent with tie CTS requirement. Byron CTS allow restoration of basin level in 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />s: therefore. Condition B reflects this allowance. Ino>erability of SX makeup capability can have an impact on VHS function. T1e Byron plant-design includes deep well pumps that can provide makeup capability. In addition certain pump inoperabilities (e.g., basin level switch inoperability) would not preclude .nanual initiation of the SX makeup pump. Therefore Conditions i

C and D are provided to address SX makeup pump inoperabilities. '

Pu (Byron Only) NUREG Bases 3.7.9 Surveillance Requirements section is '

revised to reflect the Byron CTS requirements associated with low river level and valve lineup verification consistent with the SRs added in the LCO.

P. The Completion Time for ITS LC0 3.7.4 Condition A is revised from 7 days to 30 days. There are no CTS " Plant. System" requirements associated with the SG PORVs. Maintenance / Repair of an inoperable SG PORVs has sometimes in the past taken longer than the NUREG allowed 7 days.

Po The LC0 Bases of 3.7.5 are revised consistent with other ITS Bases descriptions of required piping, valves...etc.

P. The Background section of the Bases for NUREG LCO 3.1.15 contains the sentence. "The s)ecified water level shields and minimizes the general area dose when tie storage racks are filled to their maxiinum capacity,"

' This sentence has been deleted in the ITS. This statement does not-provide any real- useful or technical information and is not needed.

• Pu The Applicable Safety Analyses section of the Bases for NUREG LC0 3.7.15 i contains the statement. "In the case of a single bundle dropped and

• lying hrrizontally on to) of the spent fuel racks however, there may be

< 23 ft of water above tie top of the fuel bundle and the surface, indicated by the width of the bundle." The statement. "to) of the fuel bundle and the surface. indicated by the" is confusing. T1e deletion of

this statement provides a more accurate and understandable description l- of the condition.

l BYRON /BRAIDWOOD UNITS 1 & 2 3.7 14 10/10/97 Revision B

' JUSTIFICATION FOR DIFFERENCES TO NUREG 1431 BASES SECTION 3.7 PLANT SYSTEMS P.,

Consistent with CTS SR 4.6.1.1.a and ITS LC0 3.6.3 Actions Note 1

-(Containment Isolation Valves) and the philosophy of the Generic Change described in-Bases JF0 C 4 . a note is added to LC0 3.7.3 to allow FW lsolation Valves to be unisolated intermittently under administrative control. Appropriate Bases revisions are provided.

P.,_ Not used.

Pu The Bases for SR 3.7.1.1 are revised to reflect that Braidwood is not currently testing the MSSVs to the OM-1987 code. Appropriate Reference changes are also made.

a)

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dl---BYRON /BRAIDWOODUNITS 1.& 2 3.7 14a 10/10/97 Revision B

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NO SIGNIFICANT HAZARDS EVALUATION ITS SECTION 3.7 PLANT SYSTEMS TECHNICAL CHANGE LESS RESTRICTIVE "Soecific'

(*Lg Labeled Comments / Discussions)

. Commonwealth Edison Company (Comed) has evaluated each of the proposed Technical Specification changes identified as " Technical Change - Less Restrictive (Specific)" in accordance with the criteria set forth in-10 CFR 50.92 and has determined that the proposed changes do not involve a significant hazards consideration.

The bases for the determination that the proposed changes do not involve-a  !

significant hazards consideration is an evaluation of these changes against i each of the criteria in 10 CFR 50.92. The criteria and the conclusions of the j evaluation are presented below. <

1.

Does the change involve a significant increase in the probability or consequeaces of an accident previously evaluated?

Q No, the proposed change does not involve a significant increase in the g probability or consequence of an accident previously evaluated.

CTS Action 3,9.11.a. requires that if the spent fuel pool water level is not within limits.. suspend movement of fuel assemblies in the s

pool and restore water level within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.- NUREG LC0 3.7,14.Required pent fuel Action A.1 only requires that mwement of fuel be suspended immediately.

The NUREG does not require the restoration of the spent fuel pool water level.

The plant safety analysis.provides the minimum water level in the spent fuel pool for a fuel handling accident as described in Regulatory. Guide 1.25. For Byron and Braidwood, this minimum spent ^ fuel water level is 23 ft of water above the top of the damaged fuel assembly and the fuel pool water surface during a fuel handling accident.

In the event the-spent fuel pool water level falls below the 23 ft minimum level, the correct action is to immediately suspend fuel movement. This is acceptable since the fuel handing accident is the analyzed accident. Suspending the movement of fuel prohibits the precursor or initiating event of the fuel handling event. With fuel movement stopped, thus no chance of a fuel handling accident occurring, then it is not necessary to re pool water level as an action. quire the restoration of the spent fuel The restoration of the water is considered secondary since no fuel can be moved until the spent fuel pool water' level is restored.

BYRON /BRAIDWOOD UNITS 1 & 2 3.7 66c 10/10/97 Revision B

1 NO SIGNIFICANT HAZARDS EVALUATION ITS SECTION 3.7 PLANT SYSTEMS

2. Does the change create the possibility /of a new or different kind of accident from any accident previously evaluated?

No. the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

The proposed change does not involve a physical alteration of the plant.

No new equipment is being introduced, and installed equipment is not being operated in a new or different manner. There is no change being made to the parameters within which the plant is operated. There are no setpoints at which protective or mitigative actions are initiated affected by this change. This change will not alter the manner in which equipment operation is initiated, nor will the function demands on credited equipment be changed. No alteration in the procedures which g ensure the plant remains within analyzed-limits is being proposed, and no change is being made to the procedures relied upon to respond to an A}' off-normal event. As such, no new failure modes are being introduced.

The change does not alter assumptions made in the safety analysis and licensing basis. Therefore, the change does not create the possibility of a new or different kind of accident from any accident previously l evaluated.

3.

!. Does this change involve a significant reduction in a margin of safety?

No, the proposed change does not involve a significant reduction in a margin of safety.

The margin of safety is established through equipment design operating parameters, and the setpoints at which automatic actions are initiated.

There is not detrimental impact on any equipment design parameter, and the plant will still be recuired to operate within prescribed limits of the safety analysis. The celetion of the Action to restore the spent fuel pool water level to within limits is not a precursor to the fuel-handling accident. The only action that could initiate the fuel handling accident is if a fuel assembly were dropped. Maintaining the Required Action to suspend fuel handling, in the event the water level is not within its limits. is the only initiation action assumed in the safety analysis. Therefore, the change does not involve a significant reduction in the margin of safety.

BYRON /BRAIDWOOD UNITS 1 & 2 3.7 66d 10/10/97 Revision B

NO SIGNIFICANT HAZARDS EVALUATION ITS SECTION 3.7 PLANT SYSTEMS l

TECHNICAL CHMGE LESS RESTRICTIVE "Soecific'"

("Ln" Labeled Comments / Discussions)

Commonwealth Edison Company (Comed) has evaluated each of the pro Technical Specification changes identified as " Technical Change Lessposed Restrictivo (Specific)" in accordance with the criteria-set forth in 10 CFR 50.92 and has determined tb ; the proposed changes do not involve a significant hazards consideratior The bases for the determination that the proposed changes do not involve a significant hazards consideration is an evaluation of these changes against each of the evaluation criteria are in 10below.

presenteo CFR 50.92. The criteria and the conclusions of the

1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?

No, the proposed change ' es not involve a significant increase in the

$ probability or consequence of an accident previously evaluated.

CTS LC0 3.9.11 requires, in part, that the spent fuel pool water level be maintained a 23 feet over the top of the irradiated fuel assemblies seated in the storage racks whenever irradiated fuel assemblies are in the storage pool. ITS 3.7.14 is essentially the same except, the APPLICABILITY is for when there is movement of irradiated fuel in the spent fuel pon1. This is a less restrictive change since the CTS requirement is for when irradiated fuel is in the spent fuel pool and the ITS is only applicable for when irradiated fuel is being moved in the spent fuel pool.

The plant safety analysis provides the minimum water level in the spent fuel 1.25. pool for a fuel handling accident as described in Regulatory Guide If there is no movement of fuel, in the pool, this would not be an initiator for the fuel handling accident or release of fission products. Since the ITS is still within the boundaries of the safety analysis, and there is no new or additional initiator of a fuel handling accident, the proposed change does not increase the probability or consequence of an accident previously evaluated.

BYRON /BRAIDWOOD UNITS 1 & 2 3.7 66e 10/10/97 Revision B l

NO SIGNIFICANT HAZARDS EVALUATION ITS SECTION 3.7 PLANT SYSTEMS

2. Does the change create the possibility bf a new or different kind of accident from any accident previously evaluated?

No, the proposed change does not create the possibility of a new or different kind of. accident from any accident previously evaluated.

The proposed change does not it.volve a physical alteration of the plant.

No new equipment is being introduced, and installed equipment is not being operated in a new or different manner. There is no change being made to the parameters within which the plant is operated. There are no setpoints at which protective or mitigative aci.lons are initiated affected by this change. This change will not alter the manner in which equipment operation is initiated nor will the function demands-on credited equipment be changed. No alteration in the procedures which ensure the plant remains within analyzed limits is-being proposed, and no change is being made to the procedures relied upon to respond to an O off-normal event. As such, no new failure modes are being introduced.

The change does not alter assumptions made in the safety analysis and licensing basis. Therefore, the change does not create the possibility.

$ of a new or different kind of accident from any accident previously evaluated.

3. Does this change involve a significant reduction in a margin of safety?

No, the proposed change does not involve a significant reduction in.a margin of safety.

The margin of safety is established through equipment design. operating parameters, and the setpoints at which automatic actions are initiated.

There is no detrimental impact on any equipment design parameter, and the plant will still be required to operate within 3rescribed limits of the safety analysis. The revision to change the AP)LICABILITY from essentially all times to only during movement of irradiated fuel

. assemblies.is not a-precursor to-the fuel. handling accident. The only action that could initiate the fuel handling accident is if a fuel assembly were dropped. Therefore, the change does not involve a significant reduction in the margin of safety.

y BYRON /BRAIDWOOO UNITS 1 & 2 3.7 66f 10/10/97 Revision B

ENCLOSURE 2  :

Section 3.9 i 4

nwec.c.co umea wn s, i ~ -- ~~

' ".' 'd '

sutrch h hon 3.7; - '

sechon- 3 9

• 6fn Lco 3.l.14 .

REFUELING OPERATIONS ,

L Co 3 TM., ,

7-3/4.9.11 WATER LEVEL / BORON CONCENTRATION - KTORAGE P0OL LIMITING CONDITION FOR OPERATION . .

3,9.11 At least 23 feet of water shall be ' maintained over the top of irradiated fuel assemblies seated in the storage racks. The dissolved boron

. concentration of the wate -

than or equal to 2000 ppeQ.the storage ,

.. pool shall be maintained at greater

.. f:. -

APPLICABILITY: Whenever irradiated fuel assemblies are in the storage pool.  !

~ . .. ..

ETIQti: '- h. .n. ,

~

c ' . '.. .

w:. s. ,f.r er s < .

.. a. With the watc.r level requirements of the above specification not

. satisfied, suspend all movement of fuel assemblies and crane .

i* '

. operations with loads in the fuel storage areas and restore the water i J.,' . -

. level,to

.. n up3.~.s. within its limit  : ~: within

qqy 4rhours.

. , c; ;y . L' n...

~

b. With the boron ' concentration requirements of the above specification (

i not satisfied suspend all movement of fuel assemblies and crane . )

j operations with loads in the fuel storage areas and inmediately take

' es

' action to restore the diss_plved boron concentration to within its limit as soon as possible.W1 - - '

l ,_

, * ;Xu.p.g ,, y.e.g'.y. ;. ,: p.& . .:m. .. .:N.'.;::ny .

3 y y ~ -.:of :q :.. t .. . ~

c.

• .The provisions of Specification 3.0.3 are not applicable."

/ 4 i

xw:% . ' ..;;n; .  ;; 1

<: ..: + . f 7h[I.'.Wf 3,RV..n;.'

d.M*$N&u:;;.t.rj . ;y ;g.+h:?'"d> .jf',' .j N.%. n d d

. -' /WM:.,?

. . y>;xg: . .o . .w{!.n.&: w..~ . :y.W _.

. + .

' . . ; .; c : y

- .a.

~b; a. .. .

. .. r a

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.,w . z. 4.,.:.-  :. : A e . , ' . ...,,, w :\.;:' u y. y >

' . n7:g. ..

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.w. : .s

u M..

.. . , m y,,e

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n y .~v

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' SURVEILLANCE REQUIREMENTS '"' 5 .'

l  : ~-  :

.%g.p:a; .: y.m m.y .~w;m.l.;n:..,g g,:uw..e m

.: .c .

, , . . = t . ;, ry.

: y.; , y. ,: ; , m. .. .,z.:e.y; . g. .,r: u:::, y , . y: a -

n y ..

~ . . - Y t . .. : . = , :. . . p. .. . 'Q v .gn . : .;. . a, .;.q.$. 4,,twv'; . ;.d: ~gm.

4.9.11 ' The water level in .the ; storage pool shall be",detemined to be at least its minieurs required depth at least once per 7 days when irradiated fuel i assemblies are in the fuel storage pool.;- ..? : pg..,4 . . . , , , , .  ; . ,. ,

.j .

~

,: , ,. .c w. , - ,

i 4.9.ll.a Boron concentration in the storage pool shall be d termined to be 6 greater than or equal to 2000 ppe at least once pe hours.

- ysj

($

~ '

- ,/ .f g 7. ++ ; ,

, c.t . ,

• ' m,

' .a. .

,1

. , 7. + j. = 4

• .f

, kp

..3., ,

, f. .

m, - . i , a,,,

3 c

TM : NQ;ingbts Chall k i -effect until Decembe,r. k,-1007. ~ .;..

(

n .

/

BYRON - UNITS 1 & 2 3/4 9-13 AMENDMENT NO. Adf kV. b .

W BYRON CTS VAR (UPS I

Bw C"S VAR (PS

A.WM in sew, Qcgg 75l Gee Coc for seche>n 3. r $

gm~,' j g' REFUELING OPERATIONS '

Y' 3/4.9.11 WATER LEVEL / BORON CONCENTRATION - STOPAGE P0OL s '

I 4

LIMITING CONDITION FOR OPERATION

I 3.9.11 At least 23 feet of water shall be maintained over the top of irradiated fuel assemblies seated in the storage racks. The dissolved boron <

- concentration than or equal to 2000 ppa of the waterge. storage pool shall be maintained at greater (

'- c_ _

APPLICABILITY:

Whenever irradiated fuel assemblies are in the storage pool.

i

&GilQH: .

a. With the water level requirements of the above specification not satisfied, suspend all movement of fuel assemblies and crane [{i .  ;

operations with loads in the fuel storage areas and restore the water level to within its'.limit , .

within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

b.

With the boron conce;ntration requirements of the above specifica not satisfied, suspend all movement of fuel assemblies and crane operations with loads in the fuel storage areas and immediately take . k' y ,'s action to restore the ciss n concentration to within its limit as soon,as possibls.

.; ( -

c. The provisions of 5'pecification 3.0.3 are not applicable.

t .

. G,.;c*.(

~ ,

4-

.~ ,,

f '. '

3 . '5' ; '

'

• It '#' '

- J

.Q . , .> - ,s .g 1 . ,ct ,

l:y. ae.s: '.

' n'-: C :: ' '

' '.C -

SURVfILLANCE REQUIREMENTS ' ' I ' # '

' y , ; , .. L 1)e;NY.p.yD;:Q

^ i.f.,, y .)*y ' n_

4.9.11 The water level in1the storage pool shall be determined to be at least its minimum required depth at least once assemblies are in the fuel storage pool.~ per r J days when irradiated fuel

• a:t ,

4.9.ll.a

~

Boron concentration to 2000,ppein at theleaststorage pool shall be deg o be E'd greater than or equal once per g) hours.

+

h .

n; .

, s.). ' -

M at4r nte dall km in fFfeeten=*i1 ,

h3}, ;;0/

~

BRAIDWOOD - UNITS 1 & 2 -

3/4 9-13 AMEN 0 MENT NO. ;g!

Rev.6 '

$3 m I'

• 9 ENC _0SJRE 2 Section 4.0

BYRON ::- S

~

Design Features 4.0 DESIGN FEATURES (continued) 4.3 Fuel Storage -

4.3.1 Criticality The spent fuel storage racks are designed and shall be maintained with:

a. Fuel assemblies having a maximum U-235 enrichment of 5.0 weight percent:

b. < 1.0 if fully flooded with unborated water which k,,,ludes inc an allowance for uncertainties as described in WCAP-14416-NP-?., " Westinghouse Spent Fuel Rack Criticality Analysis with Credit for Soluble Boron";

a1

> c, k s 0.95 if fully flooded with water borated to 550 ppm, u

" wfiIch includes an allowance for uncertainties as described in WCAP-14416-NP-A, " Westinghouse Spent Fuel Rack Criticality Analysis with Credit for Soluble Boron":

d. A nominal 10.32 inch north-south and 10.42 inch east-west center to center distance between fuel assemblies placed in Region 1 racks; and

e. A nominal 9.03 inch cuiter to center distance between fuel assemblies placed in Region 2 racks.

4.3.2 Drainaae The spent fuel pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 423 ft, 2 inches. '

4.3.3 Canacity The spent fuel pool is designed and shall be maintained with a storage capacity limited to no more than 2870 fuel assemblies.

l I

BYRON - UNITS 1 & 2 4.0-2 10/10/97 Revision B

_ _ _ , a4-4~- 4%- Ae-~ -- " ' "~

4 BRAIDWOO) ITS

Design Features 4.0 ;

DESIGN FEATURES (continued) 4.3 Fuel Storage- -

4.3.1 Criticality The spent fuel storage racks are designed and shall be maintained with

a. Fuel assemblies having a maximum U-235 enrichment of '

5.0 weight percent:-

b. < 1.0 if fully flooded with unborated water which-k,,,ludes inc an allowance for uncertainties as described in WCAP-14416-NP-A " Westinghouse Spent Fuel Rack Criticality

. q. Analysis with Credit for Soluble Boron":

c. ke s 0.95 if fully flooded with water borated to t0 c ppm.

cE wIlich includes an allowance for uncertainties as described -

in WCAP-14416-NP-A " Westinghouse Spent rJel Rack Criticality Analysis with Credit for SoluDie Boron";

d. A nominal 10.32 inch north-south and 10.42 inch east-west center to center distance between fuel assemblies placed in Region 1 racks: and

e. A nominal 9.03 inch center to center distance between fuel assemblies placed in Region 2 racks.

4.3.2 Drainaae The spent fuel pool is designed and shall be maintained to prevent inadvertent. draining of the pool below elevation 423 ft. O inches.

4.3.3 Caoacity The spent fuel pool is designed and shall be maintained with a storage capacity limited to no more than 2870 fuel assemblies.

g BRAIDWOOD -' UNITS 1 & 2 4.0-2 10/10/97 Revision'B

I l

BYRON CTS VARKU3S

l UtblGN FEATURES 5p;s.We. h6 4. 0

42. m REACTOR CORE FUEL ASSEMBLIES LA zustitT 5-4A h u, g ,, )

4.2.1 SAF:n The core shal contain 193 fuel assembi s with each fuel assembly containing 264 fuel ods clad with Zircalo or ZIRLO, except that limited substitution of fu rods by filler rods consisting of circaw, ;. Imm or l

stainless relond steel o by vacancies may be made if justified by a cycle specific analysis. M: ec e ' n:ve-a-nomna+-astWe f e ' ength er LA F th: .) The ' itialV t e ' re load 9 shalf ave a m ximum en ichment of le han 3.2 weigh percent -235. oad f shall e simila in phy ical pdesign o the itial e e loadi or pr ious cy e loadi The nrich nt I of an reload uel de gn shal .

}

eit r the s ent fue pool or e)enew dete ined to e accep ble for stora in fuel vaul . Such ceptant crit ia I sh 1 be b ed on t e r m.y==== = :esults 70=Eof mc the/ 'RITICALI;

/

i-Y"., OT OY" ll A!!O-7 7 </ t (DNTROL R0D ASSEMBLIES 4.11 G:"A The core shall contain 53ffe:: length :.nd ne prt 1:ngih\ control rod assemblies. gh: fu;: ;ength-contel--reda :::blics shall sntein - m-mi M LAa

}4a-4nc-hes-of-absorbe mateM4M All control rods shall be hafnium silver-indium-cadmium. or at_ mixture of both tvoes. A;; cont e: rods sh d qd ci d h

&r m t + 1r n -te ' t w .:.1

$N REACTOR COOLANT SYSTEM DESIG

/

RESSURE AND TEMPERATURE x /

5.4.1 The Reacto olantSystemisdesigneder.dshallbemain[ained:

a. In ai,* rdance wi the Code requirements,,sp d in Section 5.2 of the UFSAR, with allo artce for normal, degradation pursuant to the applicable Surveillance h irementi. -

b. For a pressure of 2485 psig, an

c. For a temperatuye of 650*F, except for t 680'F. ressurizer which is

/,./

VOLUME

/"g/

5.4.2A6e total water and steam volume of the Reactor Coolant System is t

l2,257 cubic feet at a nominal 3 of 588.4*F.

5.5

"[TEGRO: 06 f GAL--TOWER 40CAH99 f h 5.5.1 The-meteorekg4caMower-05:1' S: loeated-as-shown-on--s9ure - 5. : :. /'

. . . - 1 _ m.

("'e2lao cu cl eraidwcc>d sprM E uel' Rock Criti"Cr Analycin ca)d Ucmg --

7 C ( Ac - r7 -1 L 2, ai ih cali t y '

A ncuble hjcit coron of We e Crechy ron/ BrcodMag IW/,d FreA Fuel Eack s, " June 19 89.

woes -

WO , ,_ A _. / ' ._ _./ ~ , m .- .. A . _ . / _

A ReV.8 BYRON - UNITS 1 & 2 5-4 Amendment No.7E

L6 FUEL STORAG h 3 ,g CRIIICALITY _

N- _

j" 6.1.1 The spent fuel i orage racks Y desi i I. '

a less than or equal to 0.95 whenith flooded unborated hned wateranM1 , which be maintainMt 'l inc , es a conservative allowance for uncertainties as described in Section'9.1 '

f the SAR. This is ensured by, controlling fuel assembly placement in sach gion a follows: . .-

~

a. REGiqN1 '

.y .. , . -

1. A,' nominal 10.32 inch north-south and 10.42 inch ea,st-west /

l

'.- . center-to-center distance is maintained between fuel assem.blies.(

placed in the spent fuel storage racks. ..i f o, . .

, ,o .

,e., -

l t 2. Fuel assembli.es may .p .be' a , stored in this.regi.on with 1

.,. .. o c .. :,. -

a) a maximum nominal initial U-235 enrichment of less than or equal to 4.2 weight percent; or . -

. r ,.

Q.w g ,.. 4.y . n :c;

• _

b) a maximum nomina'l initia U-235 enrichment of 5.'0 weight percent with sufficient Integral Fuel. Burnable Absorbers 4- presentineach' fuel'essemb1 such that the maximum '

reference fuel 'assemb1'y..k i less than or equal to 1.470 N

j'- at 68'F., .

g, #

p:  ;;tg;} . ..

. g.

b* REGION 2 M. 'Clk

'- e4 .4e#'s +' *

. +

/'.,.1. .d.a .t. ,.: .e e.

s. . p:)z *:a; a .: d u n .4' .

..,i... . -

1. A nominal .

(, 3 inch c\ e.eLenter-to-center. 1 stance is maintained

• l A between el assemblies placed in the spent fuel storage racks'. .

\*' *$ K f4. w d,$.t ':.'c ' u. k

2. duel assemblies may be storedli thi Pregion.. with &% .. . :

. a)/ nominal initial ~U-235 enrichment [of,,1.6 wel ht \percent wit a maximum a

..;, / . "e;

/ no burnup and uii'to 5.0 wei ht"" reent U-23 with a minimum discharge burnup'as~'specift i 3; -' /-

/<- p.^ y.Qg;s y yQ yffj. $ gt Fjgure' 5.6-1; GW:8'or Acm. v 3 ::

. / b) Fuel assemblies with a maximum nolmi

. initial U-215 . - .

enrichment.of gfeater than 1.6 ',and ;1ess than or"equa'(to@j'

4 4.2'wei ht percent that do*not" meet'the minimum'burnupN h " s'ecift in Figure 15.6-1 G, e

c eckerboard patterthfor s,Yiha11;belloaded in?a^giW;%toraij

~ s - - ~ , ~ ,

4 4.1.2 The-k ferre:k:

rrr fael

m.t f=1 :te,r,:,;: thl' fer r.ettM4!rstes::f a.** e - 1sdi g 'ste-d d y'ia tM W . e> aeans. fe== =~4eratiaa is assumed. - . w. *;a @ 2 v .c;r.&- 'av .%.._,

DRAkNAGE [~' .~. 2 O~

<. n 5 ..: . x, . ;[x

r. .

643 The spent fuel storage pool is designed and shall be maintained to

^

revent inadvertent draining of the pool below elevation'423 feet 2. inches. _-

4 fg M,' d,2  :

j- 3}.J.. }.@'),, th 9,,1...}gf.'"S I.NSERT ]k ,

^ -- S A

2l ~,,;1.g

_A4 .-

t ..,'m{;o Myv:,_ . pg{} 4:

1 M~:

( . . . .

m ~ e- m m

.[, V

,, ::, 'ont the spent fuel storage rack, a.t, n ,, ; .mneo .4 awith aK _. ". --a ^~

\ =4=4r,, of less than = or aa"= -

riooded with water containing -

J =5;~sa;~ ublejron. -e ,-

,7 iLsv j

~ -w , ,

BYRON - UNITS 1 & 2 5-5 Amendment No. f6' FMv 8

O m e n n r_: _Lss

- -~- ~

g--

rox wo.: sts 234 s441 227s as-to-tr esits Incert 5- SA ~

tco 3, 7. ((,

P.se 4.3.1 Cr d t co ltt y N Eedisn 4,0 43 4NSERT-R \v N M I.] Tbc spent fucI storage racks are designed and aba5 be maintained with' '

a.

Fuel assemblies having a nundmum laitial U-235 enrichment of 5.0 '. weight perc

b. ' \

• A L< l.0 if fully Booded with unbosated watw, which includes an allowance for uncertainties as described in WCAP-14416 NP A, " Westinghouse Spent/Fuel '

Rask Criticality Analysis with Credit for Soluble Boron,"r.r= L Mm ij N @v c. A kas 0.95 If fully Aooded with water borated to 550 ppa, which includes/ an Ai

/

allowance for uncertainties as described in WCAP 14416-NP-A, " Westinghous Spent _Fuel /

M '

=pRack Criticality Analysis with Credit for Soluble Boron,'L"2=% 5, _

\

'k d. \

\ A nominal 10.32 loch north south and 10.42 inch east-west centw-to-ce distance ber;.w. fuel assemblies placed in the Region I racks; N or spent assemb ' with sufficicet Int

/ f each fud ==nene as described in the Fud Burnable rbers passent and B

• Spent Fue1 Rack 0 Caiticality Analysip abg Soluble Bor may be allowed unrestricted stornam

• redit," May 1997, Ranian 1 raA+

C-97-162, which

(

, e . f:-

/ A nominal 9.03 inch center-to center distance between fuel assemblies placed in the Region 2 racks; l s e- N spr.t fuel asseenblies with a cotabinati

\

ofdischarge , initial chment, and decay in the neceptab 'on of Figums 5.

.6 3, as applicable, 1, 5.6-2, or

'ch may be stored the Region 2 racks checkerboard con 5guration, as d the applicable in the" Byron and aidwood Spent Fu Rack Criticality Xnalysis Using Solu Baron Credit," M 1997 CAC-97-16 ,

and h- Interfa I equirements within d between adj cks as described "Byro and Braidwood Sp Fuci Rack Critical Analysis Using Ic Boron t," May 1997. CACf 7-162.

l 7ed sn Cecit bn 3.~1

\ A l'Q.e se DOCr, for GedNO M Rev,8

Sedlun 5.0 DESIGN FEATURES '

( 5.6 FUEL STORAGE (continued)

• CAPACITY 4.33 -

-5*.s The spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than 2870 fuel assemblies.

5.7 COMPONENT 0[CLIC OR TRANSIEd Lftill . .

5.7.1 The m)onents identi ed in Table 5.7 are designed a shall be maintaine itiin the.c.yc1 or transient 1 its of Table S. 1.

. /

Addrecced in Cedton 5.o Gee DOC for Sechen 5.0 .

s

' * ), . ' .

O.

. < i

. .  ?

E

'.'.s; .

,4. .. s, .3 . . -.

- ls p,

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& e. s .. 4- ,J.

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'. .: Y m4.~. -. t ,);r n:- ':: '

+:*

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e. y;> . . .. .e,p.

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( BYRON - UNITS 1 & 2 5-Sa .. Amendment No. 86 e

%&v8 I

n s o

i

-( J p s e ,9, T s C - l , C- Q and C 3 )

c 50.0a

,. Q K ~ ,G. 'M i i 6

, , , , , , , /

i e 6.

i i # , . i a 1 ,

a e e e,i I

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4 45,0 - i 4 { 4

4) 6 znricament s u mis i l 4 i i l . # 4 . pe i Iw/o) (WD/MTU) , ,,, r, , ,

" 1.60 OLL / , , ,

[ 1.50 4.635 E , /#

/. e ie i e i

' i 40,000 . ,2.00 8,565 E i i

/ 6 ' ' 'i

._ 2.20

• 11,845 E ' '

ti - ' '

2.40 14,729 - Accepta21e 1 I ' ' '

1

~ --

2.60 17,397 --

Regica b '

35,000 -

• ~

'* 2.80g 3.00

• 20,085 22,742 ""ggh

' "*~)-g [Y l~ 3.20 'N 25,132 m ~

ggg -

. 3.40 . 27,810 /

/ -

/

e i i i

. 3.60 'J0,179 ,

M 1OG 3 7 30,000 - - 3.80 32,651 0 2

i hGee DOC & -- 4,00 35,0475 ' -' ' '

~Sechco 3, 0 --

4.20 37,389 g j

l

~~" j E 4.40 39,655 m /

f e 25,000 - 4.60 42,024 M r i

i

,_ 4.80 44,290 M / <

i t

5.00 46,442 M

.g g .

q_/ __

7

• 20,000 ' ' ' ' '

Q /

t Unaccipta21e ii i m

h

\

,/ ,' /s - *Rayyian " ]

s,

, j i i / / 1 f-g 15,0M a

f f ' .

/ / s 4 ) >

10,000 - ,' #'

\

~ (

f ,

f .' ,

N .

D -;

J' / 1

\ *

/ '

, , f f

$,ggg *

,I l t A i 3

• 4 d'

\

• J t t vi i

_t ' W ,1 t/ I 6/i I /8 '

\ .

/i i 's .

N 1

g

_f 4 I l/ 4 4 p,

. . . ., ... \- 3 4

...< i<< . i 1 ., . . ., i..

i

,E* *'

• Q* ,

• 1.60 2.00 2.40 t' g, \ .. . J' y

• 2.80 3.20 ,3.60 ,0 94 4, '8 0 5.20

' 4.00w m.

f.. n .. +

Fuel Assembly Initial De?'IS Enrichment /o) i-

Q.

, . 9,; .

'; .2

~'

3

, 3.,' .

No s: The use of linear interpolation between the m4"4== ups reported above is acceptable. .-

9, , ; - . ' "

~q t

. 7 .

FIGURE 5.6 n. -1 w,

• 1. ', t . /. .

• MINDfUM BURNUP VERSUS INITIAL ENRICHMI2iT FOR REGION 2 STORAGE '

I-YRON - UNITS 1 & 2 5-Sb Amendment No.

e- e

_ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ - - - _ - - - - - - - - - - - - - - - - - - - - - ^ ^

.. ,c,

,,c _,,,_-. -.- - -

i -

e 6r # erg 3f _n p . j " ?- '

d go O 3,7 i

l l .- -

. s I 1

c . .

INSERT C-1 .- '

( ,. . '.

s0000 sa .  ;

i 55000 > es . <

f -

s

)

, -,s 1 7  :

i 500uc '

1 s u i

, ,'f a some. '

-

• r s s 7- se s

. ACCEPTABLE i .r r </

45000 1 ' ' ' "

> r e rs s 1 1 1 1:

r i < /s 1 i s is 4000C- 1 >

> s.

< s 1/1-a, i e i 1/

, a 1 -r res .

3500C i < ' ' "

i 1 1 s /1 . , .

r .

< :1- r ...

1 i s er , ..

1 1 ri r .< ..

j m '

a m' 30000 1 1

,',"1 r / s rs

1 s ia ,

1 ,tr> -

a ggggc s

> ser -

g, ,

i 1/13 o- $

• 1 V/ r/ . ., ,,u .  ;

q , ,-1,. -

f.,.ji.l%

4 s isis  : -. -

- ganan J i e ise r FX

. g),a;- " , .

, . b / ,I F o.1,

. c,* ,

ss in . . ,, .:

+

e r, , . . , .

15000

~

137- T,#w'. ~ . .

, is, . , . ~ + ".

> sis Q.M ...r .

"' ' NOT ACCEPTABLE" -  %.4; 'v

, 1000c -

. 1 13 u

.+.

a w a g ._., m.-,,. ,-

R' . ~n  ? $'L*~ 's t

=

5000-

!{A +

. _.7 4 W !"

}f a i -

4. y -

n

.t '1' ui :i, .

1.0 2.0..

.,d/R3.0.4:1 +;;.z(44. 0 , #.,3 Initial U-235 Enrichment lw/o) jbip ppe ' E.0. ..,.,

+ t.? l + y Q L ,.. .i s u h h D O h /; <

Note: The use oflinear interpolati s. bet.veen the minimum burnups is accectable. - .

1-55%y X};QQjyj%fpg _'

.!:Yn '.Q FIGURE

- Vi$7 .*':'Q 5.5-1 ,,

?f,f<f1}fif ^ u-MINIMUM BURNUP VERSUS INITIAL ENRICHMENT FOR REGION 2 ALL CELL CONFIGURATION W.go STORAGE ./ L a'W; . . -

J.. . . -

4 -

.P

. p,4*,

By is ~'5$

\

, y n -_

__ - - . :,- --- - ~

9 Ackliencect in cecbn 3. ; - . 7$1,:

Secb G 3 ,~}

"ec

% Doc. b rech en .3-i

., ' A %. -

..a Tr.. .

(. INSERT C-2 3

45000 i 1 1 s peooy 40000

- t'a**

. > e seere

• /

j , * '****

• 3500C ' ' ** **""*

) ) / ss seene

/ /, rs se seere

% \ ,

) ) f, r/

ACCEPTABLE / UA c R30000

.' w / " ' -

/ fJ /s' .

. & w

-- J ffj r .

l . l\~.;

+

.e

/ rsts m s.. . . .' .c

a. ) y fg . W , s ;' . *

. *. ;p g 2500C * '

> r,"r/>

Sr. . . w,-

- qw.g -

/svnf .c e.1.

m > ~ s, rff 3.,3..

q w

/, rn r U ; ,'.-L 20000 - M # "

, s, rsa -

. . ...' 2. . , ., . .

t !ss -

&. cv. . .

s .r rg , ,

w b r. ;.. L (s

m s., n.

in xv ,

y ."

1500C # "

1. • ~ (' ' {

/z r .' .' .-

!"?'

j/g .. . .

/> V ,

i .

Q ,.

~;

rn , ,. .. ., .

v g.%'} . .

1000C d '

. .* 1 a

F -

NIMk ts p w.%

, i m a t..

ss.ms . .

ep1 e y, $.wr%

l 500C j .f,'NOT ACCEPTABLE *"

.M,,' E M $lF ' t .

.. ,. , ., . +

i

t. g. ;. .:...

~ . ..

r I

s

, .a .,c a i

j i . .

. y j8k,.ii.,,z.fr

-e, .. k .'. g ".

O I a g:=

37/;M "2:.

l 1 .- ,

1.0 2.0 .A .:3.0 g3 Q p. 5.0 3., ,; eg; , --

2nitial y-235 adrictamenP;id,0 .

w M :::, dos.,3.(,,/of.?

. . ,' c. -

0FM

-%. 6 }-

Tr gc.q 4C - e4 '.

Note: s. -

The use of'incar interpolation bnwa?' . WW fthe minimuin burnups .

e. c. R. m,p ..q,.;.:

, e. wq . % y s,.r , .

~- .

. .,x. m.a . e.. ~ , ~ a. 'c. '.

F 5 +:  ! r7 ' .- '

W: ,' N.. .

. FIGURE.p6...2. g%p.:r. ggD.+-R.

e ~ .. 2

.p.~~mgm.;m-- p.v.': ,.

,. ~ g;.r.s . .a -

MINIMUM BURNUP VERSUS INITIAL ENRICHMENT FOR REGION 2. .

3-OUT-OF-4 CHECKERBOA,RD . .: a.

CONFIGURATION ' ' -

h "5 A}-  ;% ^

h -

w

. .p c -

eco e eycs-ss)

-- ggw.cc4g7mreason $Mi "ee Doc. Gr techon .33 3 g ,g 3,1

- - .f. s INSERT C-3 ,

i 5000 ' '

.i  ;

t 7 r*r

_L- ,

s_ q.

i ,) [

y

)

a 't b.

I ., .

r

~

400c- -1 s- >

J f

4

\

l l

w ACCEPTABLE

/

f . __.

a, .

s 300C  !

0 3 )

)

s.

h -

r M

)

r 4

r -:

) 200c-i <

m -

. I .

n, a .

s .,.

/ -

.' f l 9, .

2 .

. ;u~

r.

10cc

/ NOT ACCEPTABLE S f - , . gu:*

t .b

, p.

j 4- ..

ig L 4 -

r

/ .,

a3

/

• 8 ,

,'s:.

.e .>....

u 7 .

. a r-4.0 4.2 , d . 4.4 p e ..- 4.6 .49,f ?.4. 8  ? .5.

ZaAtial U-235 ""!; ^ ,0 .a

.- ' +. . ,.Enrisdument l L ; ut, +:

(w/o)

,s. a.

Note:

1 The use oflinearinterpolation ^ :y w.

W.ew the minimum burnups is d. a% ble. .

i. . . . ' - .-

.

• I '9,  ? ' . . R ,v; x[.Q* '.k..  : ' I T [= f*;**

~

, p s'

, . FIGURE 5.6-3 ~- -

>=*>.~+jG;p:. i f.i" P'-_ f:yl. :y.

MINIMUM BURNUP VERSUS INITIAL ENRICHMENT FOR REGION 2 '

2-OUT-OF-4 CHECKERBOARD CONFIGURATION ' , -

e Mame'

. O

.w Reg 6 By L5 - sb)

BRWD CTS VARKU3S l

gj _____

42 C""* REACTOR CDR_E FUEL ASSEMBl1ES Mar 5-4 g,3 hh M l 41.1(SEE3 containingThe 264 core fuel shal contain 193 fuel assembli with each fuel assembly s clad with Zircaloy substitution of fu rods by filler rods co@sisting n of :- ::::or IIRl.0g except that limited stainless steel o "

. '""'2 or reland man 1wsis. r_hv ' vacancies may be

__: . _.: e_c : r_ ._ ;

made if justified by a cycle specific AG ....: _ R Thfinitial than .20 weig ; perce core loadi shall ha n. e r _; enric a maxi A c ; s ; ;t _ ,r_ 'h U-235. oad fuel nt of le (desintotheinitial of ny relo ore loadi all be s milar in hysical or previ s cycle 1 ading.

fuel di ign shall e enric e her the pent fu pool or datemi d to be captable or stor ei nt l e new fue vault. ch accep nce cri ria

• 4 Q)

_ b. $.". Y'. ... *$. N .

CONTROL RDD ASSEMBLIES 4 21EEED ass"_4sblies.

The coreI Theshall 7.11 -contain 53n:n 1;..;th ;st-! r:d ::ru' ;tt

:: n: :: ;;--t n ;t'iN:ontrol rod 1

f ..e:... J' et...L.;  ;;; hl. 2.:.1' enntair ; asi.;; ;

ndium-c& um. or a 71xtureAllofcontrol bothrods types. shall be P"h- -'-' - ' afnium, silver- @

E ith .tml:::: -" n

'::1 t- W, _ ;

" -' -^ Q h4. REACTOR COOLANT SYSTEM --

DESI SURE AND TEMPERATURE 5.4.1 The r Coolant System is designed and shall be main ned:

a. In accordan the UFSAR, witwith the Code requirements speci d in Section 5.2 of applicable Survei llowance for normal degrad on pursuant to the ce Requirements, l

b. Fcr a pressure of 2485 p and

c. For a temperature of 650*F xc for the pressurizer which is 680*F.

VOLUME 5.4.2 The total w 12,257 cub); fe and steam volume of the Reactor Coolant Syster is at a nominal T , of 588.4*F. {'

5.5 METEOR LOGICAL TOWER LOCATION s

5.

j The meteorological tower shall be located as shown on Figure'N 5.1-1 N j t

• ) - -- - , . . ~~.,_ s, / - x ~ ~~~~

(oc ay con ond t:.r oidwood heat Fue) Rod Crit i coldy A nalyc.d. Um '

(sMN ng . e 4 mec u bic Sc>ron Cr edC Hoy ID~7 CAc q7- tc 2. , oud "G d.

3 I

jj. eyrw / E: raid wocd Frech Rd Rcats," II.n , t. NB9. h .

M. . -t  % .-

.._. ^ -- '

w BRAIDWOOD

u. - UNITS 1 & 2 5-4

... ,-, , % 4 , Ler K 1 M 'l A4 Amendment No. 7n Rev . B

5.6 FUEL STORAGE 4 thron C redit in EC.F P

%sciso n 9 o v N7  % 7 T 7 N- x 5.6.1.1 i

k less than or equal to 0.95 when flooded which with /unborated of Yi! des a conservative, allowance for uncertainties as described in Sec .

a UFSAR.

regio as follows:

This is ensured by controlling fuel assembly placement in ach j

, .3 . '

d:

/. a. REGION 1 . + t n, /

r3'.m ,

j .

\ 1.

• A nominal 10.32 inch north-south and 10.42 inch exst-west,.

f, center-to-center laced in the spent distance is maintained fuel storage racks. betweervfuel ' as!.emblies

g. : ., ,

i

2. Fue asiemblies may be stored in this re n with

' g.n m .

)

^ -

a),[ a 'ximun' nom,inal initial U-235 nrichment of less than or l equa to,4.2 weight percent,

( M% %M y 3

b) a maxi nominal initial -235 enrichment of 5.0 weight )

< percent present,in w h sufficient ntegral Fuel Burnable Absorbers

/ ch fuel sembly such that the maximum reference fue asse y he is less than or equal to 1.470 i at 68'F. c

/ REGION 2 f ' b. MlC .

-( l.

A nominal between fuel9 03 rich center-ssemblies plac -center distance is maintained in the spent fuel storage racks.

, g

} 2. a) Fu s . (

Rg assemblies may be store in this region with a maximum inal initial U-235 enrichme of 1.6 weight percent with

! no burnup and up to 5.0 weight p cent U-235 with a minimum discharge burnup as specified in re 5.6-1, or 5- b) Fuel assemblies with a maximum nominal tal U-235

/

/

- enrichment of greater than 1.6 and less tha or equal to 4.2 weight percent that do not meet the minimu rnup.

s specified in Figure 5.6-1, shall be loaded in a

% ~ A checkerboard

^

pattern fcr storage in this region.

A A

, 7.5.73 The k,,, fe- r fe ! fer the "r:t ::r: ::: A m ; ::::::

A A 1

g ] .:1 :t rt; rid da!' "et ^~rri 0 ** "-a =;ea^re Fe = r;; e, : =

. . . - . =da-= t 4 = is g DRAINAGE g

+R Sj 'l j-g The spent fuel storage pool is designed and shall be maintained to pregt-4nadvertent draining of the pool below elevation 423 feet 0 inches.

h [X f ~ b A '

N7 Y j_ ~-

~

,7

~ Y  %

ni with a K cf less an m -y:14 1997, the spnt fuel storaae n+ 2:115:  : W.tsihed *

(

k a minim"=,,,af ?C00 pp 1muble boron.Q flooded with water containing s , /

BRMt%0 ( - UNA:s 1 & 2 U M -5#"

MJ l',, ' Amerrdment go.B C[

E-l f -+ler +1) ed Tun e ?O . I fi'7 m ~ - -

W A Cect ton 4 0 A WWV

4. z .i . t

^ ^ ~ ~ ' ' "

)

5.0.1.1 The spent fuel storage racks are desiped and shall be maintained with:

3

a. Fuel assemblies having a maximum initial U-235 enrichment of 5.0 weight percent; <

b. A ke < l.0 if fully flooded with unborated water, which includes an allowance for uncertainties as described in WCAP-14416-NP-A, " Westinghouse Spent Fuel Rack Criticality Analysis with Credit for Soluble Boron,"hE- - Ncc-M f

c. A ka 50.95 if fully flooded with water borated to 550 ppm, which includes a a allowance for uncertainties as described in WCAP-14416-NP-A, " Westinghouse _

_ Spent Fuel Rack Criticality Analysis with Credit for Soluble Boron,"Aceu 1,

_.m.m...._  ;+ee, (f i

d.

l A nominal 10.32 inch north-south and 10.42 inch east-west center-txenter d distance between fuel assemblies placed in the Region I racks;

-e, or spent assemb 'es with sufficient gral Fuel Burnabl sorbers present each fuel as , as described in " Byron and Braid od Spent FuelRa '

Criticality Analy s Using Soluble Bo n Credit,"May 1 , CAC-97-162, w ch I may be allow unrestricted stora in the Region 1 r= _s; e -f A nominal 9.03 inch center-to-center distance between fuel assemblies placed in the Region 2 racks; .

j

.gr.

Np(v or spent fuel assje dblies with a comb' tion ofdischarge up, initial gehment, and decay time in the ace le region ofFi 5.6 1, 5.6-2, or 5.6-3, as applic , which may be st edin the Region 2 ck;,in the applica checkerboard y guration, as de bed in the "Byro d Braidwood S Fuel Rack Criti ty Analysis Using luble Boron Credi', May 1997, CAC. 162; and '

i .

h- Int ' ce requirernems vpbe iMween a ' cent racks as des '

}

ed in the "B on and Braidway pH bd' 4 ' ticality Analysis U g Soluble Boron j redit," May 1997.AAC4 Mg _ , , ,

f ge ma in sec+mn s.,

see ocCn 6 5;ede 3.7

... Ea leuer daied sep e m Ler 25 , IH 7

y DESIGN FEATURES 5.6 FUEL STORAGE (continued) -

CAPACITY 4.3.3 . ..

5.5.1 The spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than 2870 fuel assemblies.

.7 COMPONENT /CYCLICORTRANsIENTLIMIT 5.7.1 The onents ide ified in Tabl 5.7-1 are de gned and shal 'be maintained ithin-the cy ic or transie limits of T le 5.7-1.

~e

. Add rersetl 10 Sddt on 5.O -

for secAion 5.0 i

l s

r-l BRAIDWOOD - UNITS 1 & 2 5-5a Amendment No.78 Rev.8

.4\

~

6 - r - -- Tu"F' "FMT is e+3 sl0c-l-;L , and L O > 3%

n - N t

, , 6 ,

. 4 i

e e ,

4 e ,

deinnD s6 es a> / .

45 000 - aamtssam.at surne, s < . . , ss y, tw/a) ,. . ,e f liefD/wrgg

, , y y.

1.60 081 6e M5 (O 1.80 (_635 m ' '

e i' i 8 i

V

/*

is 40.000{ 8.565 W ' '

s w b n 3 7. '2.00 Sec DC)C fog =

s; :n

.60 11:'"

17.357

. .'.Regiam d l. " "'

/ 'C '

b(.

-- i- ,i

.Ceckita) 5 . ~/ I 35.000 -* - 3 0 0 20.085 22.142 ~M i i ,

""1"" .

T--"""

'O [ 3. 25.132 m /

g ._ 3.40 27.110 M #

I i

. 3.60 30.179 M ' '

30.000 - - 3.00 32.651 " ' '

4.00 4.20 35.047 m j

7 ,'

f" ll

• "" 7.309 m j j t . ,

4.40 .655 m 4 2

r  ; e 4.60 4 024 m r

. . i a 25.0H - [ 4.80

- I ' ' '

44. So m / '

.- 5p00 46, 2E / '

y ' ' '

g - ----

+ ----------------

• .20.000 '

2 an tal. ,

w I N '

R.gion 15.000  ;

, .- q s _

^

=g g g --_

w ' ' ' '

i e -

s ,

J t v \

4 6

i

/ / h 6 6 10.000 I' ' ' x

, _i 2

I r

\,

i i t 64 .

/ /

N J' / a

1. ' (t . e 5.000 i . <,

,' ' ' ' \ ' '

I / *

. , ix . . . .' . ',

J / .

4 e1e i t t .

i 6 x.

I . .

s +

I '

, t .h t e . t 0 "' ' '

1.6 2.00 2.40 2.50 3.20 3.60 4.00 4.40 4.30 5.20 Fuel Assembly initial U-235 Enri t (w/o)

W

,etes: The use of linear interpolation between the =4"4= burnups reported above is acceptable.

l'IGURE 5.6 - 1 l

MINIMUM BU3GUP VERSUS INITIAL ENRICHMENT FOR REGION 2 STORAGE BRAIDWOOD - UNITS 1 & 2 5-Sb Amendment No.

keil. O 0 I ) . E~d ICNCC dahe$ lunC 30,UY]

eee for Secho 13,7 7-- _ ,g gg ,, , gyg g

h

• 9 INSERT C-1

s. <

60000 a =r Tisse

. 55000 > ' *****

s , ss 1 s

. 5000C ' >

u 2.===

, , ,' ri un

^ ' ' ' '~ ** *****

ACCEPTABLE **

i .r r ss s

E4500C , ,' ,'/s' '

i s s s.-

I 4000C 1

s s

r e i s

', 's'. '

r /s s,r a, < s s. s i e i s.

s i s 7- rs r

~ ' "

G* 35000 ,'- '

s 1 ss F r ' 1. F M f 2 / F1 1 7 ri r

,'1 ,"

n' 30000_ 1 1 F .ri ss s i 12 ,

r s, si 25000 '

s ssrs 1 /7 rr i i se l

$ / /4 AF 20000 a

~'

i r-k i ri r 1i sm

, s. -

- 15000 , 'O A ffA

, ',; " NOT ACCEPTABLE .

10000 ,,,,

L W m,

M 5000 7 i

D

& I.

1.0 2.0 3.0 4.0 5.0 Zaitial U-235 anrictuneat (w/o)

Note:

h use oflinear interpolation between the minimum burnups is acceptable.

FIGURE 5.61 MINIMUM BURNUP VERSUS INITIAL ENRICHMENT FOR REGION 2 ALL CELL CONFIGURATION STORAGE

~;,,,ici lener ciaf ed Tunc ?O>in9 4 gen.B es 3 t:-cd

LAcicfr.-ried in Seth on 5. 7 ' ' "NT "' 'hr -r* ~

($. ce. ' s & Sechen 3.7 . -

q g710 fi 7

. INSERT C-2 45000 t '

, l

.40000 W

' ria..

, o 3.ar.

, /

• 35000 j , su ,

/ / so **a==

1 I

% / > / as v nes

/ / , rf 20 Tears

_\ ~t I

I (

) ) f, rf

-ACCcW ABLE / ////

• 3c000 ) ////

6 / fJ//

8 ) /f)f

/ f/f/

l J / f4 '

es 25000 .' I // '/

> // />

et

> /, F/s f' A , /, '//

/, F/1 f 20000 -

/ W

/F/A '

& F/s f

& /f/A '

f/2 Y I 15000 "d 6 //> V 3

~

>/2

/A V 3 fair 10000 AF "

/2 J F A l ,

' i t 5000 \

'[ NOTACCEPTABLE v

I j 0 '

i I  !

i 1.0 2.0 3.0 4.0 5.0 t Initial U-135 Enrichstent (w/o)

Note:

The use oflinear interpolation between the minimum burnups is accepta r FIGURE 5.6-2 MINIMUM BURN'JP VERSUS INITIAL ENRICHMENT FOR RE 3-OUT-OF-4 CHECKERBOARD CONFIGURATION

'i. <, , F , Ir n er da 4 ed Tu n e 23, l 1/7 5 Rev; 8 g73, g . g y )

INSERT C 4 5000

/

r

./

~

400C A f

i s

.s

(

• ~

{

w ACCEPTABLE /

sk 8 '

)

ls 300C .

) '

}

n r

) .

(

)

{

j 200C 2 M '

f 100C - / NOTACCEPTABLE I

J r

/

l 0 >

4.0 4.2 4.4 4.6 4.8 5.0 Initial U-235 Enrichment (w/o)

Note:

The use oflinear interpolation between the minimum burnups is acceptable.

FIGURE 5.6-3 MINIMUM BURNUP VERSUS INITIAL ENRICHMENT FGR REGION 2 2-OUT4F-4 CHECKERBOARD CON"GURATION coned leHer dof ed . Tune 30,I #7 8 Re V B sed (s - sb)

l CTS DOCS

^ m-

' DISCUSSION OF. CHANGES TO CTS ITS SECTION 4.0 DESIGN FEATURES ADMINISTRATIVE CHANGES (A)- -

A3' All reformatting, renumbering, and editorial rewording is in accordance with the Westinghouse Standard Technical Specifications, NUREG-1431.

During the development certain wording preferences or English language conventions were adopted, As a result, the Technical Specifications-(TS) should be more readily readable, and therefore understandable, by-plant operators and other users. During the reformatting, renumbering, and rewording process, no technical changes (either actual or interpretational) to the TS were made unless they were identified and

-justified.

A2 Consistent with NUREG-1431. ITS Specification 4.1,1 adds a general description of the site. location. This change is editorial in nature and does not involve a technical change (either actual or interpretational) to the TS.

Aa CTS Specification 5.3.1 has been revised to delete details associated with the station criticality analysis. These details were previously added to the TS to allow the provision to change-the fuel enrichment-limits contained in the criticality analysis without NRC approval. This information has since been superseded by Amendment 58/68, Braidwood/ Byron and the limits specified directly in CTS 5.6.1.1. This change deletes information which is no longer applicable. -This change is editorial in nature and does not involve a technical change (either actual or interpretational) to the TS. .This change is consistent with NUREG-1431.

A4 NRC letter dated April 2,1997 issued Amendment 86 for Byron and Amendment 78 for Braidwood for soluble boron in the spent fuel pool (SFP). Since the license amendments were temporary .in' nature, Comed letter dated June 30, 1997-aroposed chanaes.to permanently take credit U for soluble boron in the SF). Additionally, Comed responded to the NRC's request for additional information in Comed letter dated-1 cc

. September 25, 1997. Although not yet approved by the NRC, Comed has used the June'30 1997 and the September 25. 1997 submittal revisions as

'the CTS markup aages for the ITS conversion. The clouded portions reflect these_ clanges.

As CTS 5.6.1.2 contains requirements associated with the first core loading only. This requirement is no longer applicable for Byron and Braidwood, and-is deleted. This change is editorial in nature and does not involve a technical change (either actuai or interpretational) to the TS. This change is consistent with NUREG-1431.

BYRON /BRAIDWOOD UNITS 1 & 2 4.0 1 10/10/97 Revision B

DISCUSSION OF CHANGES TO CTS

.ITS SECTION 4.0 DESIGN FEATURES e A

'. -By letter dated November 5, 1996. Comed' requested a change to CTS i LC0 3.9.11. Design Features 5.6.1.1 and Administrative Controls g- 6.9.1.10. - NRC letter dated April 2,1997 issued Amendment 86 for Byron and-Amendment 78 for Braidwood for this change.

BYR0d/BRAIDWOOD - UNITS 1 & 2 4.0 2 10/10/97 Revision B

JUSTIFICATION FOR DIFFERENCES TO NUREG 1431 f SECTION 4.0 DESIGN FEATURES P5 NUREG 4;3.1.1.b states. "k ,, s 0.95 if- fully flooded with unborated water. . ." and 'has been rev,ised to state. "k ,, < 1.0 if fully flooded with unborated water..." -This is a plant , specific change in k,,, from s- 0.95 to < 1.0. This change. is ~ required so the ITS is consistent with WCAP-14416-NP-A. This WCAP contains specific k , values for both borated and unborated s>ent fuel pool water. TMs change is also-

-consistent with a CTS'clange currently under review by the NRC.

W P. NUREG 4.3.1.1.c has been added. This new specification states.

> "k s 0.95 if fully flooded with water borated to 550 ppm which b in,c,iudes an- allowance for uncertainties as described in WCAP-14416-NP-A.

D " Westinghouse S Soluble Boron." pent Fuel Rack Criticality Analysis with Credit for The add' tion of this specification is considered plant specific since it is consistent with.the CTS and the WCAP.

BYRON /BRAIDWOOD UNITS 1 & 2 4.0 2 10/10/97 Revision B l

_ __J

ENCLOSURE 2 Section 5.0

max <~ w km

1. 1. 'IN]STRATIVE_ CONTROLS M --

- m

, - ~

x 1 M ALITY ANALYSIS OF BYRON AND BRAIDWOOD STATION FUEL STORAG 6.9.1.10 Fue richment limits for storage shall be establish '

d documented in the CALITY ANALYSIS OF BYRON AND BRAID TATION FUEL STORAGE RACKS. The'ana al methods used to dete the maximum fuel enrichments shall be those pr sly reviewed approved by the NRC in  :

" CRITICALITY ANALYSIS OF BYRON AND TATION FUEL STORAGE RACKS." The fuel enrichment limits for storage / $

limits (e.g., suberiticality etermined so that all applicable j 3

he safety an is are met. s g The CRITICALI RACKS report ALYSIS OF BYRON AND BRAIDWOOD STA FUEL STORAGE i be provided upon issuance of any changes, t Documen ntrol Desk, with copies to the Regional Administrator pn NRC ~j

( nt Inspec p . A s~ r w 1x ~ -

A A ,

SPECIAL REPOPTS

9. 2 Sp ial report shall be bmitted to he Regio 1 Adminis ator of he NRC Reg nal Office ithin the ime period pecified or each r ort.

110 RECORD RETENTION /

i

\ n addition to the applicable record retention requirements of Title Code o least th Federal Regulations, the following records shall be retained fo at i

inimum period indicated.

6.10.1 The llowing records shall be retained for at least 5 ye s:

a. Recor power.le and logs of unit operation covering time in val at each 1;

b. Records and gs of principal maintenance act ties, inspections, repair and rep cement of principal items o equipment related to nuclear safety;

c. All REPORTABLE EVE -

d. Records of surveillance ctivities inspections, and calibratipns required by these Technica Spe ications;

e. Records of changes made to rocedures required by Specification 6.8;

f. Records of radioactiv hipments;

g. Records of sealed  ;

urce and fission ector leak tests and results; and

h. Records of a ual physical inventory of all of record. eled source material 6.10.2 The foll Operating Lice e: ng records shall be retained for the dur ion of the unit

a. R ords and drawing changes reflecting unit design modi ' cations ade to systems and equipment described in the Final Safe Analysis Report;

b. Records of new and irradiated fuel inventory, fuel transfers an assembly burnup histories; I ^

BRAIDWOOD - UNITS 1 & 2 h 6-23 7~

,. AMENDMENT NO. 39 g .r, = ; ;. n n .d ..+eni.-r7F :; 7 &

Rev 8 (

~

CTS INSERT (S)

SECTION 5.0 Specification .0

. INSERT 6 '3A_ (A3anjA 3 _ _ , , . . . . ,

Io addi on 'o th up requ nts conhined in t.) currently app' roved I cr icali a alysis, he requir nts of cat-46-222 "B on and Brai ood Spe Fuel c Critic h(y Analysi ith Credit r Solub Boron / wi be satis (ed. T requirements will be n effect 1 Dec , 1997.

s N ,

J ~.~ . - + . ,'

/

,o

/

{) /

Revision A f'ev 6

CTS DOCS i

e 4

4 h

d 4

d a

4 e

4 6

-~. - , , . - - . , . - . , - - - , . - _ . , , , , - - - - - , _ _ . - . _ - - . - , , _ . - _ _ . - - - _ - - , - _ _ - - - - , , . , , . _ _ - _ - _ _ _ _ _ _ _

k E LCO MA1(UPS

?

, s '

Design Features 4.0 4.0 DESIGN FEATURES h 4.1 Site L;;etur  :::: :scristier, ;f : t: icesticr. 1):

[35snT 40-IA/ '

L2 Reactor Core g 4.2.1 Fuel Assemblies 4,q3 [

The reactor shall centain diB fuel ssemblies Each assembly shall consist of a matrix of [Zircal an initial composition of na,tural or slightly enriched uraniumy W or ZIRL ffue dioxide (UO,) as fuel material.

Limited substitutions of_rfor ed*]

zirconium alluy or stainless steel filler rodsuor tuel rods, tn accordance my be used. with approved applications of fuel rod configurations, Fuel assemblies shall be limited to those fuel msicns that have been analyzed with applicable NRC staff approved ccdes ano 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.

'h 4.2.2 N Control Rod Assemblies 53f h The reactor core shall contain  % control rod]Fassemblies. The control material shall be4 silver indium cadmium, be  ::-bid

@ 889 hafniumas# -

=r

=r N m ;;r-1 ( , n mime of bon 6 peg) 4.3 Fuel Storage i

4.3.1 Cr,ticality

• OP,'

9 . 2 .1.1- Thespentluelstorageracksaredesignedandshallbe maintained with:

a. Fuel assemblies having a maximum,U-235 enrichment e weight cent; 8 o (l.O b k w .. h if fu y flood with unborated water, I.

I wNi'ch iricludes an allowance for uncertainties as described iniSectier 9.1 cf the FSfM;'

. P --  %^ ^

wcAr- M4 %- NP- A,"Weshigiouce fe d Fuel Rock Gltimhty kong ri; mil. Cred it foi Eclulle Eoren " Revioen 1, Noven &r i f%

.- (continued)

WOG STS 4.0-1 Rev 1, 04/07/95

• Rev. B

Design Features

+**I' 4:0 1

cxnmtA.o-aA 4.0 DESIGN FEATURES 4.3 Fuel Storage (co itinued) h~ o3? wh certh -couth ord) woc -wr>

d A nominal' io.4 2.9:t9finctScenter to center distance between fuel assemblies placed in9the hi;S d:::ity-feel;L.c racks # P 3.o3 A nomina 19 10.0 finch center to center distance Recf an i h

t (g]

1 between fuel assemblies placed in91:a d:::ity %el &

e+a ::: racks W +he Recyon 2

[e. New or partially spent fuel assemblies with a #

Lco 3. l. % discharge burnup in the " acceptable range" of Figure,{3.7;17.-1] may be allowed unrestri storage in [either]efuel storage rack (s);,cted and]

c New er part ly spent fuel semblies with f /[f discharge rnup in the 'u cceptable range of Figure [ .7.*17-1] will be stored in comp 1 nce with h 'the NRC pproved [speci c document con ining the anal cal methods, t le, date,-or s cific- /

% , conf guration or fi e].] _ //

_ v_ - --

. .l.2 The new fuel storage racks are designed and shall b '

maintained with: -

a. Fuel assemblies having a maximum U- enrichment o 4.5] weight percent; 3 b. k 5 0. if fully flood f

wNichinclu an111 with unborated water,I ce for uncertainties as described in [Se . 9.1 of the FSAR];

c. k s 0.98 moderate aqueous foam, which1 includes allowance for u tainties as descr' ed in [Section 9.1 of th AR]; and

d. nominal (10.95] inch center to center stance '

between fuel assemblies placed in the stora racks.

4.3.2 Drainace The spent fuelIn:m;;lpool is designed and shall be maintained to prevent inadvertent draining of the pool below elevatio fit'3El.

B y.n 23 4 2.in chu .)

(continued)

Brs.'d-o El433C+ oik &

WOG STS 4.0-2 Rev 1, 04/07/95 Fie.V. B l

LCO INSERT (S)

SECTION 4.0 g INSERT 4.0 2A (P.)

7 c. k s 0.95 if fully flooded with water borated to 550 ppm.

E- wNichincludesanallowanceforuncertaintiesasdescribed in WCAP-14416-NP-A; " Westinghouse Spent Fuel Rack Criticality Analysis with Credit for Soluble. Boron":

l L

10/10/97 Revision B

LC0 J )s

e .

JUSTIFICATION FOR DIFFERENCES TO NUREG 1431 SECTION 4.0 DESIGN FEATURES BRACKETED CHANGES (B)

• Br The brackets were removed and plant specific wording was added.

B,- The brackets were removed and the plant specific value was added. 1 B3 The brackets were removed and the optional wording was retained.

GENERIC CHANGES (C)

None.

PLANT SPECIFIC CHANGES (P)

P3 During the development certain wording preferences, English language conventions, reformatting, renumbering, or editorial rewording consistent with plant specific nomenclature were adopted. As a result,

> the Technical Specifications (TS) should be more readily readable by, and therefore understandable to. plant operators and other users.

During this reformatting, renumbering, and rewording process, no technical changes (either-actual.or-interpretational) were mdde to the ,

TS unless they were identified and justified.

P, NUREG 4.3.1.1.f states "New or partially spent fuel assemblies with a-  !'

discharge burnup in the " unacceptable range" of Figure [3.7.17-1] will be stored in compliance with the NRC approved [ specific document containing the analytical methods, title.- date, or specific configuration or figure).]" This is not. incorporated into the.ITS.

Byron and Braidwood plants do not have a NRC approved document for fuel ,

assemblies in the " unacceptable range". Therefore. this Specification is y not included in the.ITS. In addition. ITS LC0 3.7.16 has been expanded 0- to account.for all possible storage configurations such that any fuel enrichment, burnup and decay time used at Byron and Braidwood can be d stored within one of the specified conditions and-still remain within 1

the limits of LCO 3.7.16. Since irradiated fuel cannot be stored outside the acceptable region referencing or using Section 4.0 is not

. required, Pa NUREG Specification 4.3.1.2 was deleted. CTS 5.6 contains no requirements with respect to new fuel storage racks. A reformatting, consistent with NUREG-1431 is performed. ppropriate P,

CTS 5.3.1 includes the allowance for limited substitutions of fuel rods by vacancies. This allowance is reflected in ITS Specification 4.2.1.

BYRON /BRAIDWOOD -UNITS 1 & 2 4.0 1 10/10/97 Revision B if

BYRON C~S mal (U3S .

r I

..e

• d

Tpni9 s ca9s on 5.0 Cpect f t cab n 4 0 DESIGN FEATURES ' - _ _

5.6 FUEL STORAGE (continued)

~

/APACITY 5.6.3 The sp t fuel storage ol is design'ed a shall be maintai d with a storaae cap ity limited to n more than 2870 f el assemblies.

55.5(-W3 COMPONENT CYCLIC OR TRANSIENT LIMIT Q components a'- * ' -

l are designed and shall be h_

maintaine fwithin the cyclic or transient limitsm ut!e t. '-n.

Thic picgtoin provides coo \rol lo + rock the VFSAR , Ced so n 3 33 cychc. Onc\ +tontieth occurrences b enture 4hett

(

)

i rv-AddreLC M in Sed *l 4 0 feeCOCzfoi5*'ctivo46

~

i BYRON - UNITS 1 & 2 5-Sa Amendment No. 86 Rev.B j

4 a

.-.. - - _. - ~. - - . . . - . - - - - - - -

..~

Q -

o pe 6 L h n er.o I ADMINISTRATIVE 40NTt01:5s 7-

~

M rkm.) x ICALITY AHALYSIS OF BYRON AND SRAIDWOOD STATION FUEL STORAGE 6.9.1.30 Fue

,/ '1l t chment limits for storage-shall be establis

'i 4

t nd- t documented in the LITY ANALYSIS OF BYRON AND BRAI TATION FUEL .-

LST0FGF P.ACKS, The anal methods used to de the maximum fuel '.

l enrichments shall be those prov k

I revi approved by the NRC in '

" CRITICALITY ANALYSIS OF SYRON AND STATION FUEL STORAGE RACK $." -The A _' fuel enrichment limits for stera 1 mined so that all applicable -

lA" limits (e.g.,suberiticalit the safety ana are met. /

i

/ The CRITI LYSIS 0F BYRON AND BRAIDWD00 STAT L STORAGE

(. RACKS repn all be provided upon issuance of any channes, to

' RC Doc control Desk, with copies to the Aggional Admin wtrator and ,

\

pas %pr.

7 ,A y A .

A - -

w SP IAL REPORTV j

Ov

.9.2- Speci RC Regional Office wit in the ti reports s 11 be s tied to Regio'ng Administ stor of l periodspecifiedfo/eachreprt.  !

1 10 RECORD RETENTION ~

b n addition to the applicable record retention requirements of Title 10 Code o Federal Regulations, the following records shall be retained for a "

, least th inimum period indicated.

4 6.10.1 The 11owing records shall be retained for at least 5 years- i

a. Reco and logs of unit operation covering time inter L power le  ;

.at each ,

b. Records and s of principal maintenance activi es inspections, rcpair and rep ement of principal itses of ipmen,t related to nuclear safety;

c. All REPORTABLE EVENT ,

E

d. Records of surveillance ivities, spections, and calibrations '

required by these Technical 9ecif cations;

, e,- Records of changes made to th cedures required by i

h f.

Specification 6.8; Records of radioactive poents;- '

g. Records of sealed s rce and fission det r leak tests and results; ,

L .end l

h. Records of an al physical inventory of all se d source material l ofrecord./ ,

6.10.2 The foll ng records shall-be retained for the durati of the unit Operating Lice  :

a.' ords and' drawing changes reflecting unit design modifica ons

' de to systems and equipment described in the Final Safety lysis-Report;

b. Records of new and irr6diated fuel inventory, fuel transfers and l

6 . ,y agrmgr,historias; 6 g 6-23 EgENTNO.50 C

- . . ~ -a--.-m.-~ ....L',.l.-w s ,m._. _ . . . . . ~ . . . . . ..+~.m...m-,,_.m..--.m,, y.-..co.4,.,_..m'y 4-.,-,....-#m,..,,w

& CTS INSERT (S) i SECTION 5.0 l

,. Specificati 5.0 mWT62 JAQndAp a 3 _ , , _ , . . . -_

'y ,

lo addihon to he rnup requ nts conhined in th currently approved

.I cr1 icali ana sis. he requir ntsofCAF46-222."B on and rafdwood Spe Fuel ck itica y Analys1 ithCreditforSolub Bo on." wi be sa. tis ed. T se quirgswillbeneffectuntQDec 31, 1997.

. ... .. .._ .._ T .-.--  % ,,,

j I

[

i Revision-A Rev 8

BhD C"S MARRUPS k

f

(

w su a

a

- N .u

. Spectftecthon 4,0 DESIGN FEATURES L1' FUEL STORAGE (continued)

[APACITY f 5.6.3 The ent fuel stora pool is desi ed and shall b maintained wit a storaae et city limited t no more than 70 fuel assem tes.

5.5.5 0FF) COMPONENT CYCLIC OR TRANSIENT LIMIT . - - - -

@ Gli> components " '- " '

" ' " are dec.wned and shall be maintained within the cyclic or transient limitsCof Ts'ir 5. 7- L This program providec coobd to 4 tach she UFr.AP y Section 3.9, c9che onc\ tranctent occurrenCc5 4e encure +hed-N Cee boC.s for Gethon .O 9

BRAIDWOOD - UNITS 1 & 2 5-5 a Amendment No.78 Re.V . e i

I

DISCUSSION OF CHAN3ES TO CTS ITS SECTION 5.0 ADMINISTRATI"E CONTROLS An CTS Specification 6.9.1.7. footnote ** has been revised to delete information that is not applicable to Byron and Braidwood. The radwaste systems are common to the units. therefore, reporting releases from each unit is not applicable. This change is considered editorial in nature and does not involve a technical change (either actual or interpretational) to the TS, This change is consistent with NUREG-1431.

An NRC lettaa dated April 2, 1997 issued Amendment 86 for Byron and Amendment 78 for Braidwood for soluble boron in the spent fuel pool (SFP). Since the license amendments were temporary in nature. Comed letter dated Junc. 30. 1997 >roposed changes to permanently take credit for soluble boron in the SF). Additionally, Comed responded to the D NRC's request for additional information in Comed letter dated September 25, 1997. Although not yet approved by the NRC Comed has t used the June 30, 1997 and the September 25, 1997 submittal revisions as 02 the CTS markup aages for the ITS conversion. The clouded portions reflect these c1anges.

-Aa CTS Specifications 6.12.1 and 5.14.1 have been revised to incorporate references consistent with 10 CFR Part 20,- Since the plant requirements remain the same, the change is considered to be a change in presentation only. During this reformatting no technical changes (either actual or -

interpretational) to the TS were made unless they were identified and-e

-s justifled, This change is consistent with NUREG-1431.

t-l A a Not used, Ag CTS Specification 4.0.5.a has been deleted. The requirement to perform (

ASME SECTION XI testing is denoted in 10 CFR 50.55a(g). Since conformance to 10 CFR is a condition of the-license, specific identification of this requirement in the TS would be duplicative and is not necessary. Since the plant requirements remain-the same, the change is considered to be a change in presentation only. During this reformatting, no technical changes (either actual or interpretational) were made to the TS unless they were identified and justified. This change is consistent with NUREG-1431.

Aa CTS Specification 4.0.5.b has been revised to add a definition of the biennially frequency for the IST 3rogram. This change provides only a

. clarification of the meaning of-tle term and does not add 'any new requirement. This change is considered editorial in nature and does not involve a technical change (either actual or interpretational) to the TS. -This change is consistent with NUREG-1431.

__- BYRON /BRAIDWOOD UNITS 1 & 2 5.0 3 10/10/97 Revision B I