Text

Proposed Tech Specs Changing Main Steam Line Radiation High Setpoint from Three Times Background to Seven Times Background
	ML20206J947
Person / Time
Site:	Susquehanna
Issue date:	06/23/1986
From:	; PENNSYLVANIA POWER & LIGHT CO.
To:	;
Shared Package
ML20206J924	List: ML20206J919; ML20206J947; PLA-2654, Application for Amend 38 to License NPF-22,changing Main Steam Line Radiation High Setpoint from Three Times Background to Seven Times Background in Tech Specs;
References
	NUDOCS 8606270319
	Download: ML20206J947 (58)
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TABLE 2.2.1-1 REACTOR PROTECTION SYSTEM INSTRUNENTATION SETPOINTS co na M

g ALLOWA8LE l

M c

FUNCTIONAL ISIIT TRIP SETPOINT VALUES Om

$8 1..

Intermediate Range Monitor, Neutron Flux-High 5 120/125 divisions 1 122/125 divisions g

of full scale of full scale

@F 2.

Average Power Range Monitor:

s E

a.

Neutron Flux-Upscale, Setdown

< 15% of RATED THERMAL POWER

< 20E of RATED Q

THENIAt. POWER

~

b.

Flow Blased Simulated Thermal Power-Upscale

1) Flow Riased 50.58W+59/,with 10.58W+62d,with a maximum of a maximum of

2) High Flow Clamped 5 313.5% of RATED

$ 115.55 of RATED THERNAL POWER 1HERMAL POWER c.

Neutron Flax-upscale 1 1185 of RATED THERMAL POWER S 120% of RATED THERMAL POWER 7

d.

Inoperative NA NA

^

3.

Reactor Vessel Steam Dome Pressure - liigh

$ 1037 psig i 1957 psig 4.

Reactor Vessel Water Level - Low, Level 3

> 13.0 inches above

> 11.5 inches above Instrument zero Tastrument z.ro a-5.

Main Steam Line Isolation Valve - Closure

7. 0

< 10E closed 8.9-

< 11% closed

, 6.

Main Steam Line Radiation - High 5-4WD-x full power x full power background background 7.

Drywell Pressure - High 1 1.72 psig i 1.88 psig 8.

Scram Discharge Volume Water Level - liigh a.

Level Transmitter 1 88 gallons 5 88 gallons b.

Float Switch 1 88 gallons S 88 gallons 9.

Turbine Stop Valve - Closure 1 5.5% closed i 7% closed g

10.

Turbine Control Valve Fast closure, g

Trip 011 Pressure - Low

> 500 psig

> 460 psig I

11.

Reactor Mode Switch Shutdown Position NA NA I

12.

Manual Scram NA NA

'See Bases Figure B 3/4 3-1.

.See Specification 3.4.1.1.2.a for single loop op

.tfon requirements.

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t I

i f

r I

t TA8tE 3.3.2-2 Eg ISOLATION ACTUATION INSTRt#ENTATION SETPOINTS ALLOWhELE l

TRIP FUNCTION TRIP SETPOINT VALUE 1.

PRIMARY CONTAlleiENT ISOLATION g

a.

Reactor Vessel Water Level y

1)

Low, Level 3

>13.0 inches *

> 11.5 inches 2)

Low Low, Level 2 E-38.0 inches *

[-45.0 inches y

3)

Low Low Low, Level 1 1 -129 inches" 1 -136 inches b.

Drywell Pressure - High

< 1.72 psig

< 1.88 psig t

c.

Manual Initir. tion ilA lik d.

SGTS Exhaust Radiation - High

$ 23.0 mR/hr 1 31.0 mR/hr e.

Main Steam Line Radiation - High 541( full power background

$ Br& X full power background l

/

2.

SECONDARY CONTAlletENT ISOLATION

7. O g,q a.

Reactor Vessel Water Level -

m

)

Low Low, Level 2 1 -38.0 inches" 1 -45.0 inches y

b.

Drywell Pressure - High 1 1.72 psig

$ 1.88 psig U

c.

Refuel Floor High Exhaust Duct Radiation - High 1 2.5 mR/hr**

$ 4.0 mR/hr**

d.

Railroad Access Shaft Exhaust Duct Radiation - High

$ 2.5 mR/hr**

1 4.0 mR/hr**

e.

Refuel Floor Wa11 Exhaust Duct Radiation - High 5 2.5 mR/hr**

$ 4.0 mR/hr**

f.

Manual Initiation 11 4 HA 3.

MAIN STEAM LINE ISOLATI001 7'

I*

a.

Reactor Vessel Water Level - Low Low Low, Level 1 1 -129 inches

1 -136 inches l

b.

Main Steam Line Radiation - High full power background i

X full power background c.

Main Steam Line Pressure - Low 1 861 psig 1 841 psig d.

Main Steam Line Flow - High 5 107 psid 1 110 psid i

i

TABLE 1 Off-Site Doses J

NEDO-10174 fFlow Blockage Event)

Inhalation (RemJ Whole Body (Rem)

Exclusion Area (2 hr dose)

.5 0.1 Low Population Zone (30 day dose) 1.1 O.03 1

NED0-10174 Adjusted for SSES y/Q's Inhalation (Rem)

Whole Body (Rem)

Exclusion Area (2 hr)

.47

.085 Low Population Zone (30 day)

.062

.0045 Control Rod Drop Accident :(/SAR 15.4-15)

Inhalation (Rem)

Whole Body (Rem)

Exclusion Area (2 hr) 1.07

.136 Low Population Zone (30 day)

.153

.0062 Loss of Coolant Accident (Design Basis)

(FSAR Table 15.6-18)

Inhalation (Rem)

Whole Body (Rem)

ExclusionArea(2hr) 154 5.12 Low Population Zone (30 day) 33.4

.889

l Calc. No. EA.1-N 4.o o 1

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I The functies t(e.b) is defined as e F(e 6). uhere F(e.b) is the Sistert sesent lategrel. Thus E(4.6) = ek(e.b)
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In the fellowing tables K(0.b) is tabulated for values of a between 28 and 90' and b between 0 and 38. The meshes in both'vertebles are suffielently fine to ellem linser laterpolettee with occuracy of better then 1 percent.
i The laterpelatten ogsetten ist K(eede.bett) = E(e.b) + (M/ae)de + (M/46)eb ukere (0.6) are tabeleted values nearest te (seet.b+46).
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ATTACHMENT 2 CONVERSION OF PP&L CALCULATION EA-1-NA-002 TO SUSQUEHANNA UNIT 2 FOR BOTH EQUILIBRIUM CYCLE 8x8 AND 9x9 CORE LOADINGS Prepared By A
L Reviewed By
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~~References:~~
(1) NF&SE Calculation No. EA-1-NA-002, " Failed Fuel Calculation to Support Technical Specification Change," D. A. Matchick, 2/3/84.
(2) " Element MSL Radiation Detector," Bechtel Drawing No. J-G115-A.
(3) Telephone Call Transcript, R. E. Doebler L. M. Olson, subj ect: " Unit 2 MSL N-16 Background," 5/16/86.
Calculation No. EA-1-NA-002, " Failed Fuel Calculation to Support Technical Specification Change," was developed specifically for Susquehanna Unit 1 only, and assuming an equilibrium cycle 8x8 core loading. Conversion of this calculation to Susquehanna Unit 2 can be made due to main steam line symmetry between Units 1 and 2.
Expansion of the results to a 9x9 equilibrium loading can also be performed since the methodology used assumes a homogeneous core.
1.
Conversion to Susquehanna Unit 2 with 8x8 Fuel Conversion of calculation EA-1-NA-002 from Unit 1 specific to Unit 2 specific can be performed due to symmetry between units and near-identical layouts of the four Main Steam Line Radiation Monitors (MSLRMs), (ref. 2).
The only area of difference between the two units occurs in the Normalization Factor (gj;j= detector A,B,C,D), which adjusts the Calculated Dose Rate to the Actual (i.e., measured) Dose Rate for each MSL radiation monitor. The Normalization Factors are calculated in Appendix D Table D-2 (page 30 of EA-1-NA-002), which has been revised to Unit 2 actual N-16 background, and is attached below.
These Normalization Factors are applied in Appendix F, Table F-1 (page 42) to calculate the " Normalized AD" parameters. These parameters are the normalized increase in detected radiation required to reach the 3X and 7X background trip setpoints for each of the four MSL radiation detectors.
The Unit 2 specific Normalization Factors were used to generate a revision of Table F-1, which is attached.
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2 of 6 The Normalized 4D parameters are used in Appendix F, Table F-2 (page 43) to determine the mixed Equilibrium Rat ofFissjonProductConcentration required to produce 3X and 7X trips ( j,Aci/cm ).
A Unit 2 specific Table F-2 is attached below. Because the Unit 2 Equilibrium Ratios in revised Table F-2 ar so similar to the Unit I results, the bounding high Thatis,a[agxedequilibriumfissionproduct average values of 3X and 7X (cale. EA-1-NA-002, page 41) are the same for both units.
3 concentration of ~13 pei/cm will cause a 3X trip, and ~ 39 p.ci/cm will cause a 7X trip.
us, the conservatism used in calculation EA-1-NA-002 to derive [3X and 7X cancels the differences in No lization Factors (Mj) between Susquehanna Units 1 and 2.
Since 3X and 7X remain unchanged, the final percentages of fuel damage required to initiate 3X and 7X background trips (Table G-1) remain unchanged. The results of calculation EA-1-NA-002 are the same for both Susquehanna Units 1 and 2 using 8x8 fuel.
2.
Conversion to 9x9 Fuel Bundle (Equilibrium Core)
Conversion of calculation EA-1-NA-002 from an equilibrium 8x8 core to an equilibrium 9x9 core can be performed, since the Gap Inventories in Appendix A Table A-2 and Melt Release Inventories in Table A-3 are based on homogeneous core Specific Activities (SA, ci/MW, in Table A-2) for various Fission Product Isotopes. Thus,akultipliIrof No. Fuel Rods in 8x8 Core No. Fuel Rods /8x8 Bundle 62 No. Fuel Rods in 9x9 Core No. Fuel Rods /9x9 Bundle " 79 can be applied to correct the Gap Inventories in Table A-2 for 9x9 fuel.
This ratio carries through calculation EA-1-NA-002 and, to correct the final results of the calculation in Table G-1, " PERCENTAGE OF FUEL DAMAGE REQUIRED FOR 3X AND 7X TRIP ON MSLRM HIgRADIATION," we reduce the mixed inventory per fuel rod (i.e., 4.71 x 10 Aci/ fuel rod for cladding failure release) by multiplying by the ratio 62/79.
(This is the same as increasing the "i fuel rods fail" in Table G-1 by a factor of 79/62).
This determines the number of failed rods for an equilibrium 9x9 core.
It is noteworthy that the percentage of failed fuel needed to initiate 3X and 7X MSLRM Radiation-High trips remains unchanged between the 8x8 and 9x9 cores. The 9x9 revised Table G-1 is attached below.
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