ML20236B809

Proposed Tech Specs Revising Scram Time Testing in Order to Achieve Consistency W/New NRC Approved GE Transient Analysis Methodology

ML20236B809
Person / Time
Site:	Hatch
Issue date:	10/21/1987
From:	GEORGIA POWER CO.
To:
Shared Package
ML20236B791	List: ML20236B787 ML20236B809 ML20236B835
References
TAC-66524, TAC-66525, NUDOCS 8710260363
Download: ML20236B809 (13)

Text

' ' '

j .w ,

i s , ,

7 ,

l 1

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS ,'

3.10.C. Core Monitorina Durina Core 4.10.C. ? Core Monitorina Durina Core ~

Li Alterations. Alterations' ,

1. . During normal core alterations,' . two. Prior to making 'n'ormal ' alterations  ;

SRMS shall, be operable;-:one iin the to the core the SRMs shall bel . '

1 core-quadrant where fuel or. control functionally. tested'and checked'for? 3 srods 'are being moved and one -in an . . . neutron. response. Thereafteri

. adjacent quadrant, except 'as . specified ' -while: required to be operable' ,- the:

in 2 and 3 below. SRM's'wil_l be checked. daily for response.

For 'an SRM to ,be ' considered l operable.

-it shall be inserted to.the normal Use'of-special-movable," dunking i operating level and shall have a : type detectors during initial fueli  %

minimum of 3 cps with all rods capable . ; loading and major core alterations fi of normal insertion fully inserted. in' place of normal detectors is1 {

permissible as long as the detector '

-j

2. . Prior to spiral: unloading the SRMs l 1s connected to the normal SRM; 11

. shall be proven operable as stated circuit. d above, however,.during spiral. .. .

1

, j unloading the count- rate _may_ drop Prior to spiral unloading orj below 3 cps.

1 reloading the SRMs shall be cl 1

._ _. functionally tested..~ Prior!to l 3. spiral' unloading'the SRMs should

! Priorassemblies fuel to spiral reload,'up to four.(4)_ also be checked.for neutron-will be loaded'into core positions next to: , .

response. '

4 each of the 4 SRMs to obtain the I required 3 cps. These assemblies may be any which have been shown to meet the criteria for storage in the .i spent fuel pool. Until these  ;

assemblies:have been' loaded, the 3' cps requirement is not necessary. [;

.i D. Spent Fuel Pool Water level .D. Soent Fuel Pool Water Level ' )

Whenever irradiated fuel is stored in, Whenever irradiated fuel is stored -

the spent fuel pool,'the pool water in the spent fuel pool, the water level shall be maintained at-or above level shall be checked and recorded 8.5 feet above the top of the active daily.

fuel. 3

, E. Control Rod Drive Maintenance E. Control' Rod' Drive Maintenance l ..

l 1. Requirements' for Withdrawal 1. Requirements ' for Withdrawal ~

of 1 or 2 Control-Rods of 1 or 2 Control Rods A maximum of two control rods _  !

separated by at least two control l cells in all directions may be withdrawn or removed from the core for .

the purpose of performing control rod' .{

drive maintenance provided that:

a. The Mode Switch is locked in the REFUEL' position.The refueling 'a. This surveillance requirement is j interlock which prevents more than the same as given in 4.10.A.

one control rod from being withdrawn:

may be bypassed for one of the control rods on which maintenance is being HATCH - UNIT 1 3.10-2 Proposed TS/0141g/219-135 k

P 1 N

(

8ASES FOR LIMITING CONDITIONS FOR OPERATION i 3.10.A.2. Fuel Grapple Hoist load Settinc Interlocks

\

Fuel handling is normally conducted with the fuel grapple hoist. The total load on this hoist when the interlock is required consists of the weight of the fuel f grapple and the fuel assembly. This total is approximately 1500 lbs, in comparison I to the load setting of 485 1 30 lbs.

3. Auxiliary Hoists load Settina Interlock I

Provisions have also been made to allow fuel hardling with either of the three l' auxiliary hoists and still maintain the refueling interlocks. The 485 1 30 lb load setting of these hoists is adequate to trip the interlock when a fuel bundle is being. handled.

8. Fuel Loadina I

-j i

To minimize the possibility of loading fuel into a cell containing no control rod, i it is required that all control rods are fully inserted when fuel is being loaded into the reactor core. This requirement assures that during refueling the ] '

ref ueling interlocks, as designed, will prevent inadvertent criticality.

C. Core Monitorina Durina Core Alterations 1

The SRMs are provided to monitor the core during periods of Unit shutdown and to guide the operator during refueling operations and Unit startup. Requiring two a operable SRMs in or adjacent to any core quadrant where fuel or control rods are  ;

being moved assures adequate monitoring of that quadrant during such alterations, j The requirements of 3 counts per second provides assurance that neutron flux is j being monitored.

i During spiral unloading, it is not necessary to maintain 3 cps because core I alterations will involve only reactivity removal and will not result in criticality.

The loading of up to four fuel bundles around the SRMs before attaining the 3 cps is permissible because these bundles form a subtritical configuration.

D. Spent Fuel pool Water Level The design of the spent fuel storage pool provides a storage location for 3181 fuel assemblies in the reactor building which ensures adequate shielding, cooling, and the reactivity control of irradiated fuel. An 6nalysis has been performed which shows that a water level at or in excess of eight and one-half feet over the top of the active fuel will provide shielding such that the maximum calculated radiological doses do not exceed the limits of 10 CFR 20. The normal water level provides 14-1/2 feet of additional water shielding. All penetrations of the fuel pool have been installed at such a height that their presence does not provide a possible drainage route that could lower the water level to less than 10 feet above the top of the active fuel. Lines extending below this level are equipped with two check valves in series to prevent inadvertent pool drainage. All fuel loaded into the Edwin I. Hatch Nuclear Plaret spent fuel pool shall have an uncontrolled lattice Ka less than or equal to the limit for high-density fuel racks described in the

' General Electric Standard Application for Reactor Fuel" (GESTAR II),

NEDE-24011-P-A-8. Alternatively, fuel not described in GESTAR 11 shall have been analyzed with another NRC-approved methodology to ensure confonnity to the FSAR design basis for fuel in the spent fuel racks.

E. Control Rod Drive Maintenance During certain periods, it is desirable to perform maintenance on two control rod drives at the same time.

HATCH - UNIT 1 3.10-7 Proposed TS/0141g/219

'j

, 's 5 . 0.' MAJOR DESIGN FEATURES-A. Site Edwin I. Hatch Nuclear Plant Unit' No.1 is located on a site of about 2244 acres, .

which is owned by Georgia Power Company, on the south side of the Altamaha River in

~

Appling County near Baxley, Georgia. The Universal Transverse Mercator Coordinates of the center of the reactor building are; Zone 17R LF 372,935.2m E and 3,533,765.2m N.,

B. Reactor Core

1. Fuel Assemblies The core shall consist of not more'than 560 fuel-assemblies and shall be limited to.

those fuel assemblies which have been analyzed with NRC-approved codes and methods-

, and have been shown to comply with all Safety Design Bases in the' Final Safety' Analysis Report (FSAR).

~

2. Control Rods The reactor shall contain 137 cruciform-shaped control rods.

C. Reactor Vessel-The reactor vessel is described in Table 4.2-2 of the FSAR. -The' applicable design '

specifications shall be as listed in Table 4.2-1'of the FSAR.

D. Containment

1. Primary Containment The principal design parameters:are characteristics of the primary containment shall be as given in Table 5.2-1 of the FSAR.

2. Secondary Containment * (See Page 5.0-la) .l The secondary containment shall be as described in Section'5.3.3.1 of ~ the FSAR-and the applicable codes-shall be as given in Section 12;4.4 of the FSAR.

3. Primary Containment Penetrations Penetrations to the primary ~ containment and piping. passing through.such :

penetrations shall be designed.in accordance with standards set forth in Section 5.2.3.4 of the FSAR.

E. Fuel Storace

1. Spent Fuel All arrangemen'ts of fuel in the spent fuel storage racks and in other credible i configurations in the spent fuel pool outside the racks'shall be evaluated and shown to have a eK yf not greater than 0.95.

2. New Fuel The new fuel storage vault shall be such that-the keff dry shall'not be greater than 0.90 and the keff flooded shall not be greater than 0.95.

HATCH - UNIT 1 ' 5.0-1 Propo' sed TS/01704/266-132

_ _ - =-

m X Ek ysE w P E S L

E wz3 " WO E N

N A

"' S A , 1 H 0

7 C

' 1 . 4 L

e I s

M-

. 0

4. .

8

t. s I

i s

a A F

L 9 NE OR e XITU I

) d 9A6 8

7 DR O P

(

NE AN X

,E E 7

9 I

5 3 SGR E T A

)

L

1. T 8

1 3

)

/

t NA EN A d

D N L A H 0 U 1

8 G P HRE I

1 3 (

6 E 8 R CAG H )s

(

U G

L E MIN 0 I 4 O 8 A 1 1 P LE I 2 F 5 X .

V BL 2 E 0 0 R R9 R A 1 A A DX 7-N D N 89 I A AU P( I 0

L NL S I

2 P E

AP R

- E G 0 E 0

2 1

D.9

(

5 A

R E

V A

9 G )V HAR 4R G I

1 9E H 2V L 3

B R AP 0

2 9 4 -

DM A M I

1 1

0 SU(

P 1

B M DXI TE I

2 NA RA AM l

5 H 4

) 9 2 )

2 O

8

( B s R

< 9 D

" 4 1

. 8 O P

" e ", S E

P E 23 k WO Y T

L E

U F

i EMx . EZ m - W oE8 g .

o t

I

'e lI o t

% 4876 0 s p 8508 7 7788 1 RA 8 6666 1 B FL 0000

=

, X 8 9 W P 9X O R 1 448 L S OD 671 2 e

% % %F T i p F N p 8528 i

s A A I

5 0 0 X T 4443 2 / 21 A ML I

0000 0 0 L U 1 1 1 M gR g G s%

I

)

F pWE

_T

,N -

RHA OL M% %S OD 731 6 1 LN 5385 HP L A p6 1 6 + F E p 6556 s pE P .R O 6666 P

A MDP ARE 78 0000

>W OD  : XX 0 p MDR R F WO 0 CE 1 P W78.

i 8 I C

• pD A MOL LF

( DY OR 8 D E A

(

F R MTE T LOX EF 81 41 9270 T F C

A TANUMOF ROUAL B 8 F 6407 A P F

P S MF 6 MR I UE N E

1. 4443 0000 R A A I N F F V  % M -

N 08 M = M1 0 M NR I

I OE I t

G a

A W 2

D ===O S O p S T

1 M W)

I

) T p T D e L -

R R pT L O I F

(,CA F N U OE 7 1 -

5000 GGC H HA U DRF A p .

A MF TA T I L

2727 E

R 2 bg L L F M P S XER 001 1 O 3 P P P P R D N AR 1 1 1 1 C

A A AA F =N O E MMMM C M O%

MF A C(

0[ 7 R

U 8N G 4yt ,

I l.0 I

F

$ y 9 p _

e I

s I 0 4

- - - - - - 0 3

0 0 a .. 0 9

0 3

0 9

a 0 0 O s. ,

t

]*4* 3 gG*. gO"* $I/j I>t CZG m o A mab r..

c.

y %a g%eN$

i!,

-1 I

POWER DISTRIBUTION LIMITS' l

. 4 3/4.2.3 MINIMUM CRITICAL POWER RATIO 'j i

LIMITING CONDITION FOR OPERATION 3.2.3. ALL ' MINIMUM CRITICAL ~ POWER. RATIOS (MCPRs) for: two-loop operation,-

shall be equal to or greater than the MCPR operating limit (0LMCPR),Lwhich ' j is a fun: tion of average scram' time,- core: flow,' and core' power. For'25% 1 s Power < 30%, the OLMCPR is given in Figure 3.2.3-4. For' Power;2 30%,

the OLMCPR is the_ greater _ of either:

a. The applicable limit determined from Figure 3.2.3-3, or a

b. The appropriate Kp given by Figure 3'.2.'3-4, multiplied by the' appropriate- ]

limit from Figure 3.2.3-1 or 3.2.3-2, where 't is the relative measured- I scram speed with respect _to Option A and Option'B' scram' speeds.* If t is determined'to'be less than zero,'then the OLMCPR.is' evaluated at t = 0.

l For single-loop operation, the MCPR limit.is increased by 0.01 over the H comparable two-loop value. Ll APPLICABILITY: CONDITION 1, when THERMAL POWER 2'25% RATED THERMAL P'0WER' l

1 4

l 1

1 1

i

• The specific formula for determining t is provided in plant' procedures. l HATCH - UNIT 2 3/4 2-6~ Proposed TS/0143q/219-77  :

3/4.2.3 MINIMUM CRlTICALCPOWER RATIO (CONTINUED)'

ACTION:

With MCPR less thanlthe' applicable. limit. determined?from Specificati.on 3.2.3.a ,or 3.2.3.b for two-loop orfsingle-loop operation, initiate'correcti'eL v , 1 action within_15 minutes and continue corrective / action so that MCPR is equal' to or greater..than the applicable limit within 1! ho'urstor. reduce:THERMAD POWER -

to less than or equal to 25% of RATED THERMAL. POWER within the:next' 4: hours.

SURVEILLANCE REQUIREMENTS 4.2.3 The MCPR limit at. rated flow:and. rated power shall be'determinedffor: ,

each ' type of fuel- (8X8R, P8X8R,' BP8X8R, 9X9 LFA,: and- 7X7); from Figures :3.2.3 -l . ,

and 3.2.3-2, using:

a. t = 1~0 prior to the initial scram: time measurements for the' cycle'-

performed in accordance with; Specification 4.1 3.2.a, or;

b. t is determined from scram time measurements performed-in accordance with, Specification 4.1.3.2. The. determination off the limit must be completed within 72< hours of the conclusions of each scram time surveillance testirequired by Specification:

4.1.3.2, MCPR shall be determined to be equal to ok greater -than the app'licable limit:

a. At least once per_24 hours,

b. Whenever THERMAL POWER has been increa' sed by at.least 15% of i RATED THERMAL POWER and steady state operating l conditions have. 1 been established, and

c. Initially and at least once per112 hours when the reactor is operating with'a LIMITING. CONTROL ROD PATTERN for MCPR.

3

]

q i

F 1 i HATCH - UNIT 2 3/4 2-7 Proposed TS/0143q/219c-7 7 . <

i 4

1.38 1.37 g 1.36 E 1.36 l 0  !

y 1.34 9

< us

- 1,m -

.l l  ! /

! i t.31

• [

l ,E , .s0

/

1 ,,

g 1.28

)

0.0 0.2 0.4 0.6 0.8 1.0 'I T I ALL 8 X 8R AND 9 X 9 LFA FUEL TYPES l FIGURE 3.2.31  !

1 4 l .

'f, a

l .

i i

1 3

.h j

i 1

HATCH - UNIT 2 3/4 2-7a ProposedTS/0146/g9

y j q

}.

. INSTRUMENTATION.

-SURVEILLANCE REQUIREMENTS' CONTINUED

b. . Performance ~of a~ CHANNEL FUNCTIONAL TEST:

~

1

1. Within.24 ' hours' prior to the start of' CORE ~ ALTERATIONS, and, i r

2. At least 'oncec per: 7 days.

c. Verify that the channel' count' rate'.is at:least 3 cps:atlleast o'nce..

~

'per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> during CORE ALTERATIONS, and at .least 'once' per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, except.

cj The 3 cps'is not required during. core alteration'ls involving _

~

l '. 'i only fue1 u'nloading provided t_he;SRMs were: confirmed to' read. ,

at least 3 cps initially and'were checked 'for-neutron . response. '

~

2. The 3 cps is not required' initially-on a. full. core-re1oad. .

Prior to the reload, up;to _four fuel assemblies will be-loaded into core positions next'to.each of the 4 SRMs.to obtain:the

~

required count rate. These' assemblies'may.be anylwhich'have:

been shown to meet the criteria given in Section 5.6.1 of.these Technical' Specifications -for storage:in.the spent fuel pool,

d. Verifying that the RPS circuitry " shorting 11nks"'have_been removed-1 and that the RPS circuitry-is in a non-coincidence' trip mode:within-8 hours prior to.' starting CORE ALTERATIONS or shutdown margin demonstrations.

l t

N i

i

-HATCH - UNIT 2 3/4 9-4 _ Proposed TS/0144q/219-39.

l

~ .'

l

~ ,

3' ,

'y

..; i y l

-l

-Bases Table'B 3.2.1-1  !

SIGNIFICANT INPUT PARAMETERS T0'TH.E s 3

' LOSS-OF-COOLANT ACCIDENT. ANALYSIS FOR HATCH-UNIT 2 q q

Plant Parameters: -

Core Thermal Power ..................... 2531'Mwt which corresponds- j to 105% of: license. core 1 power

• 1 Vessel Steam Output .....................:10.96 x 105Libm/h which .

4 corresponds to 105% of rated-steam flow =

[d J

Vessel Steam Dome Pressure ............. 1055 psia' .j u

Design Basis Recirculation Line j Break Area For: I 1

a. Large Breaks ...................- 4.0,-2.4, 2.0,,2.1 and'l.0 ft ~ 2 ,

b. Small Breaks ................... 1.0, 0.9, 0.4 and 0.'07 ft*- )

o Fuel Parameters: 1 PEAK TECHNICAL SPECIFICATION DESIGN'

' INITIAL -

MINIMUM-1 '

LINEAR' HEAT. AXIALL CRITICAL' FUEL BUNDLE GENERATION RATE' PEAKING ' POWER.-

4 FUEL TYPE GEOMETRY (kW/ft)- ' FACTOR RATIO 1 Initial Core 8x8 .13.4 1.4 1.18 l l

A more detailed list of input to each model and its source is presented in-Section II of Reference 1 and subsection 6.3.3 of the FSAR.

For convenience, the APLHGR limits are reported in the. units-of kW/ft, l which is the bundle planar power normalized to the number of fueled 1 rods. Figure 3.2.1-9 shows that the'9x9 LFAs have the same planar power-limits as the GE P80RB284H fuel; however, on a kW/ft basis, the'APLHGR' limits for the~LFAs are 62/79 times the P80RB284H limits.

• This power level meets the Appendix K requirement of 102%. The core . O heatup calculation assumes a bundle power consistent with operation. of.

the highest powered rod at 102% of its Technical Specification  ;

linear heat generation rate limit.

HATCH - UNIT 2 B 3/4 2-2( > Proposed TS/0145q'/223-j

9 P0t!ER DISTRIBUTION' LIMITS BASES MINIMUM CRITICAL POWER RATIO (Continued)

- As . depicted on figure 3.2.3-1 or 3.2.3-2 the 100% power.100% flow operating . .j limit MCPR (0LMCPR) depends on the average , scram time, t, of the contro1J '

rods, where: j J

t = 0 or - Tave TB, whichever is greater

'A tB )

T where: A=1.096see-(Specification 3.1.3.3, scram ltimelimit-to notch 36). d

• B = y + 1.65 Ng 1/2, ]

n EN

.i"I ;. {

l where: u = 0.822 sec'(mean scram time used in the transient ana!ysis) o = .018 set (s';andard deviation of p) ]i n

intgj

• ave = i=1 .l r1 IN 4 q i=1 where: n = number of surveillance tests performed to date in the cycle N  ;

1 = number of active control rods measured in the ith surveillance test T

i = average scram time to notch 36 of all rods in the ith surveillance test '

N 1 = total number of active rods measured in 4.1.3.2.a. i The purpose of the MCPR f , and the Kp of Figures 3.2.3-3 and 3.2.3-4, respectively is to define operating limits at other than rated cort flow and power conditions. At less than 100% of rated flow and power, the required MCPR I is the larger value of the MCPRf and MCPRp at the existing core flow and )

power state. The MCPR f s are established to protect the core from inadvertent i core flow increases such that the 99.9% MCPR limit requirement can be assured.

The MCFRf s were calculated such that for the maximum core flow rate and the corresponding THERMAL POWER along'the 105% of rated steam flow control line, the limiting bundle's relative power was adjusted until the MCPR was slightly above the Safety Limit. Using this relative bundle power, the MCPRs were calculated at different points along the 105% of rated steam flow control l line corresponding to different core flows. The calculated MCPR at a given point of core flow is defined as MCPRf.

The core power dependent MCPR operating limit, MCPR p , is the rated power, and rated fics MCPR operating' limit multiplied by the K p factor given in Figure 3.2.3-4.

The Kps are established to protect the core from transients other than core

~ flow increases, including'the localized event such as rod withdrawal error. The Kps were determined based upon the most limiting transient at the given core power level. -For further information on MCPR operating limits for off-rated conditions, see NEDC-30474-P (Reference 2).

HATCH - UNIT 2 B 3/4 2-4 Proposed TS/0145q/219-l

. 3/4.9 REFUELING OPERATIONS' BASES 3/4.9.1 REACTOR MODE SWITCH ,

Locking the OPERABLE reactor mode switch'in the refuel position ensures that the restrictions on rod withdrawal and refueling platform movement during the refueling operations. are properly activated. These conditions 2

reinforce the refueling procedures and reduce the probability of inadvertent j criticality, damage the reactor internals or . fuel assemblies, and exposure .

of personnel to excessive radioactivity.

'1 3/4.9.2 INSTRUMENTATION j The OPERABILITY of at least two source rtnge monitors ensures' that redundant monitoring capability is available to detect changes in 'the l reactivity condition of the core. During the unloading, it is not necessary J to maintain 3 cps because core alterations will involve only reactivity removal and will not result in criticality. _The loading of up to four bundles around the SRMs before attaining the 3 cps is permissible because .

these bundles form a subcritical configuration. l ,

3/4.9.3 CONTROL R0D POSITION The requirement that all control rods be inserted during CORE-ALTERATIONS ensures that fuel will not be loaded into a cell without a ,

control rod and prevents two positive reactivity changes from occurring l simultaneously.  ;

3/4.9.4 DECAY TIME ,

The minimum requirement for reactor subcriticality prior to fuel q movement ensures that sufficient time has elapsed to allow the radioactive decay of the short lived fission products. This decay time is consistent with the assumptions used in the accident analyses.  !

k 3/4.9.5 SECONDARY CONTAINMENT ]

Secondary containment is designed to minimize any ground level release of radioactive material which may result from an accident. The reactor building provides secondary containment during normal operation when the .

drywell is sealed and in service. When the reactor is shutdown or during refueling, the drywell may be open and the reactor building then becomes the primary containment. The refueling floor is maintained under the secondary q containment integrity of Hatch-Unit 1.

Establishing and maintaining a vacuum in the building with the standby gas treatment system once per 18 months, along with the surveillance of the - l doors, hatches and dampers, is adequate to ensure that there are no l violations of the integrity of the secondary containment. Only one closed damper in each penetration line is required to maintain the integrity of. the secondary containment. <

HATCH - UNIT 2 B 3/4 9-1 Proposed TS/0145q/219

-a 1

l .

DESIGN FEATURES CONTROL R00' ASSEMBLIES

~

5.3.2 The reactor core shall contain 137 cruciform-shaped control-' rod assemblies.

l 5.4 REACTOR COOLANT SYSTEM  !

I DESIGN PRESSURE AND TEMPERATURE l 5.4.1 The reactor coolant system is designed and shall be maintained:

a. In accordance with the code requirements specified in' Sect' ion 5.2 j of the FSAR, with allowance for normal degradation pursuant to the applicable Surveillance Requirements, )a I

b. For a pressure of 1250 psig, and -

l

c. For a temperature of 575 F VOLUME i

5.4.2 The total water and steam volume of the reactor vessel and recirculation system is approximately 17,050 cubic feet at a nominal Tave l of 540 F.

\

1 5.5 METEOROLOGICAL TOWER LOCATION ,

5.5.1 The meteorological tower shall be located as shown on Figure 5.1.1-1. q 5.6 FUEL STORAGE .

CRITICALITY 1

5.6.1 The new and spent fuel storage racks are designed and shall be l maintained with sufficient center-to-center distance between fuel assemblies placed in the storage racks to ensure a keff equivalent to s 0.95 when flooded with unborated water. The k eff of s 0.95 includes conservative allowances for uncertainties in calculations of both normal and abnormal storage conditions as specified in the FSAR..

l l

l 1

HATCH - UNIT 2 5-3 Proposed TS/0169q/266-39 U_____'_ _ _ _ _ _ _ _ _ _ m - - - - - - - - - - - - - - -