Text

Reload Licensing Rept for Browns Ferry Unit 2,Cycle 6. Determination of No Significant Hazards Encl
	ML20095G859
Person / Time
Site:	Browns Ferry
Issue date:	07/31/1984
From:	; TENNESSEE VALLEY AUTHORITY
To:	;
Shared Package
ML18026B181	List: ML18026B180; ML18026B182; ML20095G859;
References
	TVA-RLR-002, TVA-RLR-2, NUDOCS 8408280226
	Download: ML20095G859 (38)
	v • d • e

.

TVA-RLRM BROWNS FERRY NUCLEAR PLANT RELOAD L CENS NG REPORT UNIT 2, CYCLE 6

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Y RELOAD LICENSING REPORT FOR BROWNS FERRY UNIT 2, CYCLE 6 TENNESSEE VALLEY AUTHORITY July 1984

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. I.' -

Introduction l'

,77

.1Mkis reload licensing report presents : the' results of the core design and. safety analyses performed for Browns Ferry unit 2, cycle 6

' operation. The. methodology and technical bases employed in the

o performance.of these analyses are discussed in references 1-6.

Items specifically' addressed here include the nuclear fuel assemblies I

and core loading to be.used in cycle 6, the reload core nuclear design

~

. characteristics, the transient and accident safety analysis results, and the proposed opeasting thermal limits.

The cycle 6 reload core will include four Westinghouse QUAD +'

^

' demonstration assemblies located in nonlimiting core peripheral locations. A complete description of the demonstration assemblies is contained in Westinghouse Report WCAP-10507 (reference 8).

II.

Reload Cycle Information I

A.

Design Basis Exposures 1.

Projected cycle 5 core average exposure at end of cycle:

20.5 GWd/ ST l

2.

Minimum cycle 5 core average exposure at end of cycle from cold shutdown considerations: 19.5 GWd/ ST 3.

Assumed cycle 6 core average exposure at depletion of reactivity (DOR)*:

17.4 GWd/ ST

DOR - End of full power capability l

2 B.

Reload Fuel Assemblies Fuel Tyne Cycle Loadad Number Irradiated P8DRB284L,R2 3

15 NDRB284L,R3 4

201 P8DRB265H,R4 5

80 P8DRB284L,R4 5

168 New P8DRB284L.R5 6

296

' QUAD + Demo 6

4 1DTAL 764 Descriptions of the nuclear and mechanical design of the General Electric irradiated and new fuel assemblies to be loaded in cycle 6 are contained in reference 7.

The nuclear, mechanical, and therms 1-hydraulic design descriptions for the Westinghouse demonstration assemblies are contained in refarence 8.

4-C.

Reference Core Loading Pattern The reference loading pattern is the basis for all reload licensing and operational planning and is comprised of the fuel assemblies designated in item II.B of this report. It is based on the best possible prediction of the core condition at the end of the previous cycle and on the desired core energy capability for the reload cycle. The reference loading pattern is designed with the intent that it will represent, as closely as possible, the actual core loading pattern. Figure 1 shows the reference core loading pattern for cycle 6.

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The' reference-loading pattern includes four Westinghouse QUAD +

demonstration assemblies loaded =in peripheral '1ocations.

Eysluations performed by Westinghouse (reference 8). indicate 1that the results of licensing analyses for the lead P8x8R fuel assembly bound those for the QUAD + demonstration assemblies.

Cycle specific analyses performed by TVA confirm this-conclusion.- The results documented in this report are for the limiting loading pattern.

D.

Special Conditions

'The use of increased core flow (ICF) is pisaned for cycle 6 operation. Safety analyses were performed for both 100 percent and 105 percent of rated core flow with the most conservative results used for determining the operating limits. The i

conclusions regarding LOCA analysis, reactor internals pressure drop, and flow-induced vibration as discussed in reference 9 are applicable to cycle 6.

The flow-biased instrumentation for the rod block monitor will be signal clipped for a setpoint of 106 percent. since flow rates higher than rated would otherwise result in a ACPR higher than reported for the rod withdrawal error.

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III.. Nuclear Desian Characteristics i

A.

Shutdown Margin The reference core is analysed in detallito ensure that adequate shutdown margin exists. This section-discusses the results of

. core calculations for shutdown margin (including the liquid poison system).

1.

Core Effective Multiplication and Control Rod Worth Core effective multiplication and control rod worths were calculated using the TVA BWR simulator code (references 2 and

4) in conjunction with the TVA lattice physics data generation code (references 3 and 4) to determine the core reactivity with all rods withdrawn and with all rods inserted. A tabulation of the results is provided in table 1.

These three eigenvalues (effective multiplication of the core, uncontrolled, fully controlled, and with the strongest rod out) were calculated at the beginning-of-cycle 6 core average exposure corresponding to the minimum expected end-of-cycle 5 core average exposure. The core was assumed to be in a menon-free condition.

Cold keft was calculated with the strongest control rod out at various exposures through the cycle. The valus R is the difference between the strongest rod out k gg at BOC and e

the maximum calculated strongest rod out k gg at any e

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

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' exposure point. The maximum' strongest rod out k,gg at any

-exposure point is equal to or less than:

Maximus k

=k (BOC) + R

-2.

Reactor Shutdown Margin Technical _ Specifications require that the refueled core musc be capable of being made subcritical with 0.38 percent Ak margin in'the most reactive condition throughout the subsequent operating cycle with the most reactive control rod

~

in its full out position and all other rods fully inserted.

The shutdown margin is determined by using the BWR simulator code to calculate the core multiplication at selected exposure points with the strongest rod fully withdrawn.- The shutdown margin for the reloaded core is obtained by subtracting the maximum k from the critical k,gg of 1.0, resulting in a calculated minimum cold shutdown margin of 1.1 percent Ak for Browns Ferry 2, cycle 6.

Table 1 CALCULATED CORE EFFECTIVE NULTIPLICATICN AND CONTROL ROD WORTHS - NO VOIDS, NO IENON, 20'C Uncontrolled, KUNC (BOC) 1.116 eff i.

Fully Controlled, KCON (BOC) 0.953 off Strongest Control Rod Out, KSRO (BOC) 0.980 off R, Maximum Increase in Cold Core Reactivity 0.009 With Exposure Into Cycle, Ak

e 6:

3.

Standby Ligsid Control System t'

The standby liquid control system (SLCS)' is designed to provide the capability of bringing the reactor, at any time in a cycle, from full power and a minimum control rod

' inventory (which is defined to be at the-peak of the xenon transient) to a subcritical condition with the reactor in the most reactive zenon-free state.

Tie SLCS shutdown margin is determined by using.the BWR simulator code to calculate the core multiplication for the cold, zenon-free, all rods out conditions at the exposure point of maximum cold reactivity with the soluble boron l'~

concentration given in the Technical Specifications. The resulting k-effective is subtracted from-the critical k-effective of 1.0 to obtain the SLCS shutdown margin. The results of the SLCS evaluation are given in table 2.

Table 2

$iANDBY LIQUID CONTROL SYS1EN CAPABILITY Shutdown Margin (Ak)

EEM (20* C. Ionon Free) 600 0.018 4

B.

Reactivity Coefficients The reactivity coefficients associated with the nuclear design of Browns Ferry 2, cycle 6 are implicit in the 1-D cross sections l

e y.

used ror-the' safety analyses.- As such, reactivity coefficients

~

are not separately calculated for input.to th's transient analyses. However, a void coefficient :is generated in the 3-D to

.1-D cross section collapsing process and is used as a verification check. For Browns Ferry 12, cycle '6 the following results - for DOR conditions were obtained:

100% core flow -

-0.0742

%Ak/% void 105% core flow -

-0.0745

%Ak/% void C.

Fuel Performance The Browns Ferry 2,' cycle 6 fuel performance is predicted by l

projecting the fuel burnup to the end of cycle with the 3-D simulator code. The calculated peak pellet exposures for the various fuel. types are less than-the limits specified in references 7 and 8.

Furthermore, peak linear heat rates satisfy the assumptions made in the fuel vendors' thermal-mechanical integrity analyses (references 7 and 8).

All-fuel 2-types loaded in cycle 6 are predicted to operate within these bounding assumptions. Additionally, the QUAD + demonstration assemblies are predicted to have greater than 20 percent margin to the lead P8x8R assembly in steady state bundle power and thermal limits throughout cycle 6.

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IV.

Transient Analyses-A.

Pressurization Events-

=The-RETRAN computer code (reference 10) is used to analyze both the reactor system and hot channel responses during core-wide pressurization transients.- The analytic models used in these analyses are described in reference 5.

A description of the CPR correlation and its application to Browns Ferry is contained in reference 11.

Analyses are performed for the potentially limiting

-events at the most adverse initial conditions expected during the cycle.. Rel'ond unique initial conditions and transient analyses results are samnarized in the following tables.

NSSS Initial Cenditions 4

Steam Flow Core Flow Gap Conductance

.Exnosure

(% Rated)

(5 Rated)

(BTU / f ts-hr *F)

DOR 105 105 650 1

Hot Channel Initial Conditions (Limitina Event)

Fuel Bundle Bundle Gap Conductance Iygg if2R Power (kW)

Flow (K1b/hr)

R-Factor (BTU / f t s-hr

F)

P8X8R 1.301 6.388 123.9 1.051 1287 i

I S

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9 Pressurization Event Analysis Results Peak Power Peak Heat Peak Vessel.

ACPR System Transient

(% Rated)

Finz (% Rated)

Press. (nsis)

P8x8R Resnonse Load 403.2 121.1

'1235.0 0.231 Figures Rejection.

2-5 w/o Bypass Feedwater 234.0 115.3 1214.0 0.150 Figures Cgntroller 6-9 Failure 6

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Nonpressurization Events.

The nonpressurization events analyzed.fo~ reload licansing are either steady state events, or relatively slow transients that ccn be analyzed in a quasi-static manner using a 3-D BWR' simulator

~(reference 2).

The methods used to analyze these events are described in reference 1.

Results are summarized below.

i Nonoressurization Event Analysis Results ACPR Peak LEGR(kW/ f t):

Eveat P8x8R/ QUAD +

P8 x8R/ QUAD +

Loss of 0.18 17.7

~

j Feedwater Heating (100*F)

Rod Withdrawal 0.172 17.2 Error Rotated Bundle 0.158 15.3 Error Nislocated Bundle 0.18 14.6 Error

}

1 For increased core flow based on a signal clipped rod block setpoint of 106 percent.

8 Includes 0.02 penalty required when using the varlble water gap nethod (reference 7).

s Results presented were calculated for the P8x8R fuel type and conservatively bound the results calculated for the QUAD &

demonstration assemblies.

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l 11 C.

Overpressure Protection The main steamline isolation valve closure with failure of direct scram is analyzed to demonstrate sufficient overpressure protection (peak vessel pressure must be less than 110 percent of design pressuro - 1390 psia). The event is analyzed using the models and methods described in reference 5.

Results are summarized below.

MSIV Closure (Fluz Scram) Results 4

Peak Vessel Peak Steauline System Pressure (nsis)

Pressure (nsis)

Resnonce 1283.0 1242.0 Figures 10-13 4

1 m

12 V.: NCPR Onoratina Limit Summary c.

The methods used to determine the required OLNCPR values for each event

~

analyzed are described in references 1 and 5.

The application of.

4 Option A and B limits in determining the cycle OLMCPR is described in the. unit Technical Specifications. Results are summarized below and in figure'14.

OMCPR for Pressurization Events (B006-EOC6)

Ontion A1 Oction B1 P8 x8R/ QUAD +

P8x8R/ QUAD +

-Load Rejection Without Bypass (GLRWOB)

-1.35 1.26 Feedwater Controller Failure (FWCF) 1.27 1.23 OMCPR for Nonoressurization Events (BOC6-EOC6)

P8x8R/ QUAD +1 Loss of Feedwater Heaters (LFWH) 1.25 I

f Rod Withdrawal Error (RWE) 1.24 Rotated Bundle Error (RBE) 1.22 i

4 Mislocated Bundle Error (MBE) 1.25 1 Results presented were calculated for P8x8R fuel types. The QUAD +

demonstration assemblies will be loaded into nonliniting core locations and monitored to the sane OLNCPR.

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f 13 VI~.

Accident Analyses

' A.

Loss of Coolant Accident-(LOCA)

.LOCA analysis results for fuel types previously loaded in unit 2 are described in reference 12.

Reference 8 indicates that the MAPLEGR limits for fuel type P8DRB284L can be conservatively

-applied to the QUAD + demonstration assemblies. These limits are presented below.

LOCA Limits for QUAD & Demonstration Assemblies Average Planar MAPLHGR Exnosure (mwd / t)

'(kW/ f t) j 200 11.2 1,000 11.3 5,000 11.8 I

10,000 12.0 15,000 12.0 20,000 11.8 25,000 11.2 30,000 10.8 35,000 10.0 40,000 9.4 B.

Rod Drop Accident (RDA) i The methodology used to analyze the rod drop accident is described in appendix A of reference 6.

Resnits for BI'2, cycle 6 are summarized below.

Results for the Limitina RDA Condition: COLD (68'F), EOC Exposure

[-

Rod Worth: 1.33% Ak Rod Position:

22-07 Peak Fuel Enthalpy: 245 cal /sm Core Response: Figures 15-18

14~

~

EVII. ' Stability Analyses The methodology 'used to analyze core and channel stability is described in appendix B of-reference 6.

The minimum stability margin occurs at the intersection of the natural circulation line and the 105 percent rod line (the flow biased scram line also passes through this point). Resnits for BF2, cycle 6 are summarized below and in figure 19.

Stability Analysis Results at Limitina Initial Conditions Maximum Analysis Decar Ratio a.

Core Stability 0.85 Channel Stability P8x8R/ QUAD +

0.591 1 Results presented are for the P8x8R fuel type and conservatively bound the QUAD + demonstration assemblies.

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References i

l '.

TVA-EG-047, 'TVA Reload Core Design and Analysis Methodology N

Tfor the Browns. Ferry Nuclear Plant,' Tennessee Valley Amt.hority, January 1982.

-2.

1Yh-TR78-03 A, 'Three-Dimensional LWR Core Simulation

- Metlods, ' Tennessee Valley - Authority, January 1979.

-3.

TVA TR78-02A, '!'ethods for the Lattice Physics Analysis of LWRs,' Tennessee Valley Authority, April 1978.

4.

TVA-TR79-01A,- ' Verification of TVA Steady-State BWR Physics Methods,' Tennessee Valley Authority, January 1979.

5.

TVA TR81-01, 'BWR Transient Analysis Model Utilizing the RETRAN Program,' Tennessee Valley Authority, December 1981.

6.

1YA-RLR-001, ' Reload Licensing Report for Browns Ferry Unit 3, Cycle 6,' Tennessee Valley Authority, January 1984.

7.

NEDE-24011-P-A-6, ' General Electric Standard Application for Reactor Fuel,' General Electric, April 1983.

J 8.

WCAP-10507, ' QUAD + Demonstration Assembly Report,' Westinghouse Electric Corporation, March 1984.

9.

NEDO-22245, ' Safety Review of Browns Ferry Nuclear Plant Unit No. 2 at Core Flow Conditions Above Rated Core Flow During Cycle 5,'

General Electric, October 1982.

10.

EPRI NP-1850-COE, 'RETRANO2 - A Program for Transient Thermal-Hydraulic Analysis of Complez Fluid Flow Systems,'

Electric Power Research Institute, May 1981.

11.

NEDE-24273, 'GEXL Correlation Application to TVA Browns Ferry Nuclear Power Station,' General Electric.

12.

NEDO-24088-1 (as amended), ' Loss-of-Coolant Accident Analysis for Browns Ferry Nuclear Plant Unit 2,'

General Electric,

. February 1978.

I l

FIGURE 1

REFERENCE CORE LORDING PATTERN BROWNS PERRY UNIT 2

CYCLE 6

60-B B

B B

B A

B B

R B

B R

B B

58 8

B E

E E

E B

B 56 8

8 C

E D

E D ED E

E D

E C

B B

54 A

E E

D E

B B

E D

E E

R 52 B

B E

C E

B E

C E

E C

E B

E C

E B

B 50' B

R B

D D

E D

E C

E D

D E

C E

D E D D

B B

B t

48 8

E E

D B

B B

C 8 D

B D

E E

D 8

D B

C B

B B

D E

E B

46 B CE C

E B

E C

E D

E B

3 E

D E

C E

B E

C E

C B

44 -

B B

E D

B C

B D 8 0

8 D

E E

D 8 D

8 D 8 C BD E

D E

B B

42 -

B E

D E

C E

D E

C C

E D

E C

E D

E R

40 -

B E

E D

E C

B D

E C

B D

E E

D B

C E

D 8 D

B C

E D

E E

B 38 -

B E

D E

B E

D E

C E

D E

B B

E D

E C

E D

E B

E D

E B

36 -

B E

E D

B D

B D 8 0

E B

E E

B E

D B

D 8 0

8 0

E D

E E 8 34 B

E D

E C

E D

E B

E C

C E

B E

D E

C E

D E

B 32 -

B E

E B

E D

E B

E C

E B

E C

B B

C E

B E

C E

B E

D E

B E

E B

30 -

B E

E B

E D

E B

E C

E B

E C

B B

C E

B E

C E

B E

D E

B E

E 8 28 B

E D

E C

E D

E B

E C

C E

B E

D E

C E

D E

R 26 -

B E

E D

B D

E B

E E

B E

D B

D 8

D E

E B

24 --

B E

D E

B E

D E

C E

D E

B B

E D

E C

E D

E 8 E

D E

8 l

22 -

A E

E D

E C

B D 8 D

E C

B D

E E

D B

C E

D B

C E

D E

E B

20 - 8 E

D E

C E

D E

C C

E D

E C

E D

E B

18 -

B B

E D

B C

B D

B D 8 0 E

E D

B D

B C 8 0

E D

E B

B 16 B

C E

B E

C E

D E

B B

E D

E C

E B

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E C 8 14 A

E E

D 8

B B

C B

D B

D E

E D

8 0 B

C B

B B

D E

E B

12 B

R 8 0 0 E

D E

C E

D E

D D

E D

E C

E D

D B

B B

10 B

B E

C E

B E

C E

E C

E B

E C

E B

B B

R E

E DE D

E D

E B

B E

D E

E B

6-B B

C E

D E

E D

E C

B B

4 8

8 E

E E

E B

R 2-B B

B B B!B BlB R

B R

B B

B l

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1 5

9 13 17 21 25 29 33 37 41 45 49 53 57 3

7 11 15 19 23 27 31 35 39 43 47 51 55 59 FUEL TYPES i'

R-P80RB284L,R2 (15)

D-P80RB284L,R4 (168) 8-P80RB284L,R3 (201)

E-P80RB284L,R5 (2961 C-P80RB265H,R4 (80)

Q-DURD+ DEMD,R5 (4)

FIGURE 2 GENERATOR LOAD REJECTION W/O BYPASS

.500 Legend TOTAL POWER (%)

400-AV E, S U,R F_A C E, H E AT, F,L_U X,(M C..O..R..E..I.N..L..E.T...F.L.O..W....(.%..)..........

CORE INLET SUBCOOLING (%)

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-300-l 200-r-

f*.,-.-~.~..'....'..~..~..~..._~~--_-~--.--,

100-

'O, s

s s

s a

n 0

1 2

3 4

5 6

7 TIME (SEC)

+

e FIGURE 3 GENERATOR LOAD REJECTION W/O BYPASS i

200 15 0 --

10 0 -

j i

0 I

C 50-l Legend

,f VESSEL PRESS RISE (PSI) s TOTAL S/R VALVE FLOW (%)

o

/

8.Y?g S S,,ve tv,E,,r gg,vt,ts,1,,,,

0-0 1

2 3

4 5

6 7

TIME (SEC)-

FIGURE 4 GENERATOR LOAD REJECTION W/O BYPASS ISO 10 0 -

ll '9r', es

I I

8 I

I I

50 -- !i aj

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i,

8

I 8

g

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Legend s t l/

l s' LEVEL (INCH-REF-SEP SKIRT)

~6

~

'l nsse.h s.IMenow H-.

.W.R.8J.N,E,,S,T,E3,y, f,(g,W,,(,]6, },,

FEEDWATER FLOW (%)

-100,

0 1

2 3

4 5

6

.7 TIME (SEC)

FIGURE 5 GENERATOR LOAD REJECTION W/O BYPASS 2

Legend TOTAL REACTIVITY ($)

SCRAM REACTIVITY ($)

0-

'~ ~. *e.g'm g

~%

2-N

's I

t

-4,

i 0

05 1

1.5 2

25 3

TIME (SEC)

FIGURE 6 FEEDWATER-CONTROLLER FAILURE i

.1 250

-Legend TOTAL POWER (%)

ME pgR,(,,CE,H, EAT FLQ{%,1, A

200-c.9.M.!.N (E J,,((g,W,,{,%),,,,,,,,,,

CORE INLET SUBCOOLING (%)

s

a..

_ y -..

. r. r. z..c........ q, s,

,gg _

..,***....s.

50 --

O-0 5

10 15 20 25 TIME (SEC)

FIGURE 7 FEEDWATER CONTROLLER FAILURE 15 0 Legend VESSEL PRESS RISE (PSI) 19IA_'_ SLR,,Q(V,E,F,,Lp,W, j$)

B.Y..P..A..S..S..V..A..L.V.. E...F..L.O..W....(.%...)...

r-i

(

50-l i--

l l

8 3

0-0 5

10 15 20 25 TIME (SEC) l I

I

FIGURE 8 FEEDWATER CONTROLLER FAILURE 15 0 --

~

i s,,

g i,

l i

'u

!l i i

s ii i ' !

>,a 50-

i

h:i i

ll !' *

!ll t 0-IE Legend LEVEL UNCH REF SEP SKlRT)

E**EE'*M 09* l"2--

50-TN.8 8.!N,{,,S,J,E,A,y,,F,LO,W (%,1,,,

a FEEDWATER F10W (%)

-10 0,

0 5

10 15 20 25 TIME (SEC)

FIGURE 9 FEEDWATER CONTROLLER FAILURE i

2 0-

%g

% \\ s

\\

s

~4~

Legend

's, TOTAL REACTIVITY ($1 s

Q C,,R,A,,y g A,(Tjy,1,Jy,[},1,,

t 6,

15 16 17 18 19 TIME (SEC)

L

' FIGURE 10 MSIV CLOSURE (FLUX SCRAM) 600 Legend TOTAL POWER (%)

pyE, g g R,F,A,C),H_ EAT FL,UX,(%,)

c.9.R,E, } N,[EJ, f,(g,W,,(%,},,,,,,,,,

400-CORE INLET SUBCOOLING (%)

f 200-

~'

0, 0

2 4

6 8

TIME (SEC)

FIGURE 11 MSIV CLOSURE (FLUX SCRAM) 250 200-160 --

100-l Legend 60-e' VESSEL PRESS RISE (PSI) l 19Mt_S/_R,ygyEytow (x) j

/

.B.1P,A S,S,,y,A,1;y,E,f,(g,W, j,%,),,,,

0, n

s s

0 2

4 6

8 TIME (SEC)

. FIGURE 12 MSIV CLOSURE (FLUX SCRAM)

ISO Legend LEVEL (INCH-REF SEP SKIRT)

VESSEL STEAM FLOW (%)

kkkkb..,55,~,r,ih,,ti,

~

i

~

iU ~

S

/v'----

FEEDWATER FLOW (%)

's \\. j

,s,' s i.

I I

I 4

8 l

l 50-

'l

/

'~

J\\ i

n 6:

..:.*v &,

l 483 If II g ii Q_

11 1 i

1

-50,

O 2

4 6

8 TIME (SEC)

FIGURE 13 MSIV CLOSURE (FLUX SCRAM) 2 Legend TOTAL REACTIVITY (S)

EEEOM.R,E,A,C,TjMy,,(j[

o-

,,'s, i

% \\

's

-4 O

1 2

3 4

5 TIME (SEC)

L l

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

FIGURE 14 OLMCPR FOR P8X8R/ QUAD +

1.36 1.35 i

1.34-1.32-GLRWOB 1.30-at Qo 1.28-2 FWCF 1.27 1.26-MBE & LFWH 1.25

,,v------------------------

RWE 1.24 - ----------,ef-----------------------------------

't.23 RBE 1.22- - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - -

1.20-0 0.2 0.4 0.6 0.8 1

TAU *

' SCRAM SPEED INTERPOLATION PARAMETER AS DEFINED IN THE TECHNICAL SPECIFICATIONS

FIGURE 15 1

REACTOR POWER (MWT) VS TIME (SEC) i 14000 12000-l, 10000-E3 4

y 8000-l d'

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I ENCLOSURE 3 DETERMINATION OF NO SIGNIFICANT HAZARDS BROWNS FERRY NUCLEAR PLANT UNIT.2 (TVA BFNP TS 199) x,.

D p ription ut amendment respes t :

i The amendment 'would revise the. Technical Specifications (T.S.) of the operating l h ense to:

(1) modify the core physics, thermal and hydraulic limits to be consistent with the reanalyses associated with' replacing about 1/3 of the core during the cycle 5 refueling outage for unit 2 and (2) ref1cet plant modifica-t ions performed.during the c.ycle 5 refueling and modification outage.

Specifically, the amendment would result' in changes to the T.S. in the f ollowing twelve areas:

1.

Changes to the license related to the Cycle 6 core reload involving removal of depleted fuel. assemblies in.about one-third of the nuclear reactor core and replacement with new fuel of the same type _previously

' loaded in the core with attendant license' changes in the core protection safety limits and reactor protection system setpoints. The actual changes are slight adjustment (by 0.01 in initial core life) in the Operating' Limit Minimum. Critical Power Ratio (OLHCPR), deleted two of four tables on maximum average planar linear heat generation rate (MAPLliGR) versus average planar exposure that will not be needed due to the fuel change and a change to the ' references in the bases to reflect that TVA perfo'rmed the reload transient analysis.

The loading pattern also includes four Westinghouse QUAD & demonstration-assemblies loaded in peripheral locations. Evaluations performed by Westinghouse indicate that the results of licensing analyses for the y

previous t;ce assembies bound those for the QUAD + assemblies. Cycle specific analyses performed by TVA confirm this conclusion.

2.

Changes in the T.S. to reflect modifications to the torus as part of the Mark I containment program.

This includes revising the tables listing surveillance instrumentation for suppression pool bulk temperature reflecting the installation of 16 sensors for an improved torus temperature monitoring system and a revision to the basis for the existing limits on torus water temperature.

3.

Change to the T.S. to reflect modifications to the scram discharge instrument volume (SDIV); each of the SDIVs now have new, diverse level

-instrumentation. The changes to the T.S. are to add operability, surveillance and calibration requirements on the new level instrumen-tation.

4.

Change 'to T.S. surveillance instrumentation tables to add new instru-mentation for containment high-range radiation monitors and to add new instrumentation; and delete current instrumentation for drywell pressure-wide ra,nge and suppression chamber wide-range water level in response to

-requirements in NUREG-0737; items II.F.1.3, II.F.1.4 and II.F.1.5.

5.

. Changes to T.S. RPS instrumentation requirement tables to delete the bypass function if reactor pressure is less than 1055 psig and the mode switch not in the RUN mode.

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2 6.

Changs th T.S. 'siirveillance' instrumentation tables to reflect new

' ins'.rnment' numbers f or t he new upgraded drywell temperature and pressure -

' i ns t : tuaent.it s of t.

7.

Revisions to the table of t estable penet rations to reflect the new

-testable penetrations as a result of moditirat. ions to make the flange

..ide of several isolation valves testable.

.8.

. Revision t o : the T.S. table f or containment isolat. ion valve surveillance

~

.to aihi two iiew isolat ion valves that are part of a newly installed redund. int disrharge'line from the drywell compressor into containment and to delete one isolation valve which was removed from the demineralized water syst em.

's.

Revision to the T.S. table for containment isolation valve surveillance ito. delete two isolation valves for the residual heat removal head spray Iine which is being removed.

10.

Revision of T.S; to' provide limiting conditions for operation and surveillance requirements for electric power monitoring for the reactor protection system power supply.

11.

Modify the T.S.- to' apply to ' the new analog (continuous n:casuring) instru-mentation. The analog instemnentation replaces certain mechanical-type pressure and level switches with a more accurate and more stable electronic. transmitter / electronic switch system and will provide improved performance of trip functions for reactor protection system acttiation,.

and containment isolatica. The changes to the T.S. include:

a.

in the tables on functional test frequencies, calibration frequencies and surveillance requirements, for each switch replaced, add the instrument number and type of sensor beneath the parameter being monitored and/or controlled.

I h.

add notes to-the above tables to specify how the functional and it calibration tests are to be conducted.

in addition to the above administrative changes, the calibration c.

requirements have been changed to incorporate extended calibration intervals. liowever, the required setpoints, functional test frequencies and channel check frequencies for the instrumentation will not be changed. The new calibration requirements, together with the new instrumentation, are expected to provide a more reliable instrumentation system.

12.

Administrative' changes to the T.S. ir.volving changes to the Table of

?

Contents to reflect the above license changes and misec11aneous editorial changes.,

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- Ihhes ~ for 1.roposed no nigni!icant hazards consideration determination:

.The Conunission' his provided xnidance concerning the application.of the standarda

_ hy providing exaisples of a'ct ions that are 1ikely, and are not 1ikely, to involve Lsignificant hazard coiaiidera t ions - (48 FR 14870).

Four examples of actions not.

likely to involvef signit'icant - hazards considerations are:

"(i)

A purely administ rat ive ' change to ~ technical speci fica'tions:

for example, a~rhange to achieve consistency throughout the technica1

- speci t ications, correction of an error, or a change in nomenclature.

hi) i A change that constitutes an additional limitation, restriction, or coritrol not presently included in the technical specifications:

f or example, a more significant surveillance requirement.

(iii) For a nuclear power reactor, a change resulting frmn' a nuclear reactor _~ core reloading, if no fuel assemblies significantly different irom those found previously acceptable to the NRC for a previous core at the faci? i ty in question are involved.. This assumes that no significant changes.are made to the acceptance criteria for the technical specifications, that the-analytical. methods used to demonstrate conformance with the teshnical specifications and regu-lations are not significantly changed, and ' hat NRC has previously found such methods acceptable...

(vi)

A change which either may result in some increase to the probability or consequences of a previously-anal'yzed accident or may reduce in some way a safety margin, but where_the results of the change are clearly within all acceptable criteria with respect to the system or component specified in the Standard Review Plan;'for example, a change resulting from the application of a small refine -

ment of a' previously used calculational model or design method."

Each of the twelve changes to the T.S. described previously is encompassed by one of the above example of actions not likely to involve a significant hazards consideration. The basis for this determination on each of the twelve changes is discussed below.

1.

Core Reload 1.a Fuel Changes The changes to the T.S. associated with removing depleted spent fuel from the reactor and replacing these with new fuel assemblies is -

encompassed by example-(iii) above of those actions not likely to involve a significant hazards consideration.

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The_ proposed reload involves fuel assemb' es which have been shown to be analytically similar or which are of the same type as previously found acceptable by the staff'and loaded in the core in previous cycles. The analytical methods,used by the licensee to demonstrate conformance to the technical specifications have been previously approved by the staff.

In addition,_no changes have_been made to the acceptance criteria for the technical specification changes involved.

Since the replacement fue1' assemblies are analytically similar or of

-the same type previously added to all three Browns Ferry units and other BWRs and.since the codes, models, and analytical techniques used to_ analyze the reload have been generically approved by the NRC, the changes to the technical specifications associated with the reload-are clearly encompassed by example (iii) of the guidance provided by-the Commission for an action not likely to involve a significant hazards-consideration.

l.h

!<e t e reau c:. in the !tases The changes in the T.S. associated with changing the references in the Bases to reflect that the reload transient analysis is now being perf ormed by TVA is encompassed by examples (i) and (iii) above of those actions not likely to involve a significant hazards considera-tion.

The reload analysis, in the past, has been performed by General Electric Company. This reload analysis has been performed by TVA using analytical methods described in TVA-TR81-01-A. The analytical methods have been approved by the staff. Since NRC has previously found these methods acceptable and the T.S~. changes are being made.

to achieve consistency betweed the methods used and the references in the Bases, these changes to the T.S. are clearly encompassed by examples (i) and (iii) of the guidance provided by the Commission for an action not likely to involve a significant hazards considera-tion.

2.

Changes Related to Torus Modifications One of the changes to the T.S. is to revise the tables that list the surveillance instrumentation associated with the suppression pool bulk temperature. This modification provides an improved torus temperature monitoring system which consists of 16 sensors. This will provide a more accurate indication oi the torus water bulk temperature as required by NUREG-0661 and will replace the suppression chamber water temperature f

inst riuneutr. present ly Iisted in the T.S.

The change to the T.S. are necessary administrative follou up actions essential to the implementation of this improvement. The changes to the T.S. place operability and calibration requirements on the new temperature monitofing system. Since these are new instruments, the surveillance requirements are not presently in the T.S.

Thus, adding those restrictions and controls is encompassed by example (ii) provided by the r.

Commission.

p

's

+

t.

-1 Scram Di ; charge inst rument Volume

't he SUVs and - SDIV: are being modi fied. t o address inadequacies identi fi ed

- by the part ial rod insert son event on llrowns Ferry nuit 3 in June 1980.

One of the modification:, includes adding electronic level switcher, to

-initiate a ver.nu on a high level in the SDIV.

Thus,_the changes to the T.S. are necessary administ rative follos up actions essential to the imliti ment at ion of t his improvement. Adding these new restrictions and controls, which otherwise would not he in the T.S.,

is encompassed by example iii). of t he guidance provided by the Conenission.

4.

Acrident tionitoring lustrumentation i t em J f.F. I ut Nt! REG-0737, "Clari t ication of Ttli Action Plan Requi rements,"

s eipii res all licenrec> to install f ive new monitoring systems and to provide ont,ite sampling / analysis capability f or a specifini range'of sadionuclides.

Fur all six cat egories, Nt1 REC-0737 states:

" Changes to technical r.pceitications will be required." During this refueling outage, the licensee has installed:

(a) a containment high-range moni t oring syst em, (b) a drywell wide-range pressure monitoring system and (c) a suppression chamber wide-range water level monitoring system.

These tbree items were regnired by NUREG-0737,~ items I1.F.1.3, I1.F.1.4 and ll.F.1.5, respectively. The changes to the T.S., which track the model T.S. provided to t 7e licensee by the staff, are to add operability and surveillance requirements on the new monitoring systems to the T.S.

The revisions also delete the present drywell pressure and suppression chamber water level instruments since they are being replaced by items b and e above.

The changes to the technical' specifications are necessary administrative follow up actions required by the Commission. Adding the new surveillance requirements and controls is encompassed by example (ii) of the gnidance provided by the Commission.

5.

Scram Permissive Pressure Switches at 1055 psi 3 Present configuration on unit 2 has a bypass function which allows a scram in the refuel and startup/ hot standby modes of operation by the stram functions main st eamline isolation valve closure and turbine condenser low vacunm when the reactor pressure is greater than 1055 psig.

The reactor high pressure scram is set at 1055 psig and is operable in these two modes of operation.

If react.or pressure exceeds 1055 psig, the reactor scrams due to the reactor high pressure scram function, and the main steamline isolation valve closure and the turbine condenser low vacuum functions become operable. The bypass circuit therefnre serves no real purpose. When the two scram functions become available, the reactor-

. is al ready scrammed. Since the reactor is protected by the high pressure scram f, unction, the proposed change does not result in any reduction in

-the margin of safety. The T.S. changes therefore are encompassed by example (vi) of the guidance provided by the Commission.

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Y D81wiki,[lymperat ure and Pressure

. fhe drywei 1 'tQinperht ure and pressure survei i lance ists t riunen talion i s heing upgraded t hi:. h.nt age - t o pr ovidi epaaliI jed, more ret iable. inst rumenta-

^

. tion.

'the T.S. are being revised to rettert new i ns t rumen t - ninnbe rs The

!L

. surveillance re. psi s ement s ' remain unchanged. The changes to the technical

['

speci firat ions are necessary administ rative _ follow up actic>ns reepsired by

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llieIComnti:. ; ion and are rica rly' encomp.insed bylcxampic (i) of the guid.nire

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pinvided-by..the Cemmission.

T.

festabli pein !

.it ion",

1

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1 Modi ficat ions are b'eing made Lo 't he flange side of fourteen cont.ainment

. isolation valves which cannot be isolated from primary containment. to he' tested. This inoditication will provide two gaskets with a pressure tap.

-between the gaskets to allow the flange to'be leak tested.. Operability

~

of the valve wili' not he af fected by this modification. Fourteen new rtestable penetrations resnited and they were adde'd to the table of testable penetrations withl double o-ring seals.. New surveillance requirements are heihg.added.;The change is encompassed by example (ii) of the guidance-provided by the Commission.

-Several editorial ch'anges were also made te this table. They include

~

revising.t he ~ identi fication name on ~ several penetrations, adding a penetration that was tested but was inadvertently lef t out of the t able and removing penet ration X-213A which no longer exists. These

' changes are purely administrative and are chcompassed by example (i) of the guidance provided by the Commission.

Redund' ant Air Supply to Drvwell 8.

During the current outage, TVA has-installed a second discharge line from the drywell compressor into containment. This line was added t.o provide the capability for isolation of approximately one-half of the drywell suppression equipment in the case of a drywell line leak. This air supply will be used to supply two inboard main steam isolation valves (MSIVs),

approximately one-half of the main steam relief valves (MSRVs), and approximately one-half of all other air-operated equipment in the drywell.. -

This will 'signi f icantly reduce the possibility of any one cont rol air pipe break inside containment from reepairing immediate shutdown and isolation as a result of MSIVs, MSRVs, and drywell coolers being inoperable, Since any line penetrating containment requires two isolatiori valves, the table in the Technical Specifications listing the isolation valves that must be periodically tested is being revised to add these two new isolation valves. TVA has concluded that this modification will increase the margin of safety. The changes to the technical specifications are necessary

/

administrative follow up actions essential to the implementation of'this i

i

-improvement. The two isoln ion valves being added to the T.S. are new i

valves *not presently listed in the T.S.

If they are not added to the

' table of valves to be periodically tested, there would be no T.S. require-ment to test these valves. Adding these additional controls is encom-

. passed by example (ii) of the guidance provided by the Commission.

!y l}

a q._

One-inolation valve on the demineratived water nystem was renueve.i t rinn mi t t 2.

The demi nr i il ize.1 w.it er sys t em i:. no longer used.

The i :.o l a t s on valve was timoved.nni the line capped. The T.S. are being revised to remove this valve from the t able ut v.i l ves to be tested.

The changes to the te( hnica l npeci f ica t i ons a re neccana ry administ.rative follow up actions essential to the implementation ut the improvement. The changes are clearly encompasse.l hy ex.nuple (i) provided by the Couunission.

9..

des i.tua i llea L Remova1 llead Spry 1.ine Twu isolation valves on the residual heat removal head spray line were removed trom unit 2.

The head spray line was removed and the in ne ration rapped. The T.S. are being revised to remove these valves from the table of valves to be tested. The changes to the T.S. are necessary whnin-I intiative follow up actions essential to the implementation of the improvement.

The thanges are clearly encompassed by example (i) provided by t he conuni nr. ion.

10 tionitoring of RPS Power Supply lly letter dated August 7, 1978, the Commission advised TVA that during review of llatch unit 2, the staf f had identified certain deficiencies in the design of the voltage regulator system of the motor generator sets

}

which supply power to the reactor protection -system (RPS).

Pursuant to 10 CFP 50.54(f), TVA was required to evaluate the RPS power supply for llrowns Ferry 1, 2, and 3 in light of the information set forth in our letter.

By letter dated Sep ten.ber 24, 1980, the staff informed TVA (and most other BWRs) that "we have determined that modifications should be performed to provide fully redundant Class IE protection at the interface of non-Class IE power supplies and the RPS."

The staff also advised TVA that "we have found that the conceptual design proposed by the General Electric Company and the installed modification on llatch are acceptable solutions to our concern." Hy letter dated December 4, 1980, TVA committed to install the required modifications.

Ily letters dated October 30, 1981 and July 28, 1982, NRC sent TVA model Technical Specifications for electric power monitoring of the RPS design and modification.

During the current outage of unit. 2, the RPS is being modified to provide a fully redundant Class IE protection at the interface of the non-Class IE power supplies and the RPS. This will ensure that failure of a non-Class IE reactor protection power supply will not cause adverse interaction to the class IE reactor protection system.

The Techaical Specifications are being revised simila-to the model T.S.

provided to TVA to refleet the limiting conditions for operation and surveillance requirements associated with the RPS modifications.

Page 42

/

is being modified to add a description of these sections in the bases.

The chgnges to the T.S. are necessary administrative follow up actions essential to the implementation of these improvements. The additional liriitatiens and controls, which are presently not in the T.S., are encompassed by example (ii) of the guidance provided by the Commission.

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8.

y.a II; AnalMri gyst ens i

~The HpS,E the, primary nstainment-isolation system (pCIS), and the core st andby cooling-systeins (CSCS) use ' mechanical-type switches ' in the sensors -

- t hat. moni tor plant - process ; parameters. These mechanical-type swit.ches ' a re y

't

-very-subject:to; drift-in the setpoint.as is evident from the many 1icensee event: reports 1(LENS) that have been submitted reporting ca1ihra-

~

1 inn drii ts in t hese ' switches.

J :

Adv.inces in technol' ugy-make it possible to replace the mechanical-type

^

mitches with almore accurate and more stable electronic! Lransmitterf elect ronic switch system. Tor several years, TVA has_been planning to

replace existing pressure switches that sense drywell and reactor

' pressures with ' analog : loops and modi fy the reactor water level i rrli ca t ion loops to improve the reliability, accuracy and response time of t his i ns t rument a t ion.

The. modification involves' removing one device and sulistituting other devices.to perform the same f unc t. ion.

Changer. in design bases, protective functiori, redundancy, trip point and logic are not. : i nvo lved. Similar modifications have been approved. for other BWRs.

I As described previously, nast of the changes to the T.S. are administra-tive inEnature (i.e., adoing the specific number and types of sensor and adding notes-to describe how testing is conducted). As such, they are

. encompassed by example (i) of the guidance provided by the Commission.

The changes in surveillance requirements relates to example (ii) of the guidance.provided by'the Commission. Some of the surveillance intervals have been decreased as appropriate for each' new instrument, llowever, the overall' ef fect. of the changes. in technical specificat. ions will be to increase the total surveillance requiremens in support of a more reliable instrumentation system.

1'2.' hdministrative Changes i~

Several administrative changes are being.made to the Technical 1

Specific'tions. These' include revising the Table of Contents to reflect the change discussed above, and miscellaneous, editorial changes. The surveillance-requl_rement for the personnel air lock is being changed to be consistent' with the' surveillance for units 1 and 3.

These changes are editorial in nature and have no safety significance. These changes are encompassed by. example (i) cited by the Commission as an action not likely to pose a significant hazards consideration.

. Since all of _ the changes to the T.S. given in the twelve areas above are encompassed by an example in the' guidance provided by the Commission of act. ions-not 'likely to involve a significant hazards consideration, the a'

staf f has made a proposed determination that the application for amendment involves no significant hazards consideration.

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ML20095G859

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