ML20203F458
| ML20203F458 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 04/15/1986 |
| From: | GEORGIA POWER CO. |
| To: | |
| Shared Package | |
| ML20203F452 | List: |
| References | |
| TAC-61283, TAC-61284, NUDOCS 8604250161 | |
| Download: ML20203F458 (76) | |
Text
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(_IMITING CONDITIONS FOR OPERATION SURVEILLANCE RE0VIREMENTS 3.3.F. Doeration with a limiting Control 4.3.F. Operation with a Limiting Control Rod Pattern (for Rod Withdrawal Rod Pattern (for Rod Withdrawal Error. RWE) Error. RWE)
A Limiting Rod Pattern for RWE exists During operation when a Limiting when: Control Rod Pattern for RWE exists and only one RBM channel is
- 1. Thermal power is below 90% operable, an instrument functional of rated and the MCPR is less test of the RBM shall be performed than 1.70, or prior to withdrawal of the control rod (s). A Limiting Rod Pattern for
- 2. Thermal power is 90% of rated RWE is defined by 3.3.F.
or above and the MCPR is less than 1.40.
During operation with a Limiting Control Rod Pattern for RWE and when core thermal power is > 30%, _
either:
- 1. Both RBM channels shall be oper- G. Limiting the Worth of a Control Rod able, or Below 20% Rated Thernal Power
- 2. If only one RBM channel is oper- 1. Rod Worth Minimizer (RWM) able, control rod withdrawal shall be blocked within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, or Prior to the start of control rod withdrawal at startup, and as soon
- 3. If neither RBM channel is oper- as automatic initiation of the RWM able, control rod withdrawal shall occurs during rod insertion while be blocked. shutting down, the capability of the Rod Worth Minimizer to properly G. Limiting the Worth of a Control Rod fulfill its function shall be veri- '
Below 20% Rated Thermal Power fled by the following checks.
- 1. Rod Worth Minimizer (RWM) a. The correctness of the Banked l Position Withdrawal Sequence l Whenever the reactor is in the Start input to the RWM computer
& Hot Standby or Run Mode below 20%
- ti rated thernal power, the Rod Worth ' shall be verified.
' St Minimizer shall be operable.or a ._ .. b. The RWM computer on line diag-second licensed operator shall nostic test shall be successfully verify that +.he operator at the performed.
reactor cons le is following the .
control rod program.
"- c. Proper annunciation of the selec-tion error of at least one out-of-sequence control rod in each fully inserted group shall be verified,
- d. The rod block function of the RWM shall be verified by withdrawing or inserting an out-of-sequence contrvl rod no more than to the block point.
HATCH - UNIT 1 3.3-5 8604250161 860415 PDR ADOCK 05000321 P PDR
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.3.G.2. Rod Seouence Control System (RSCS) 2. Rod Secuence Control System (RSCS)
- a. Doerability a. Operability When the reactor is in the Start As soon as the group notch mode and Hot Standby or Run Mode below is entered during each reactor 205 rated thermal power and control startup and as soon as automatic rod movement is within the group initiation of the RSCS occurs notch mode af ter 50% of the during rod insertion while cont.ci rods have been withdrawn, shutting down, the capabil-the Rod Sequence Control System ity of the Rod Sequence Control shall be operable except when System to properly fulfill its performing the RWM surveillance function shall be verified by at-tests. tempting to select and move a rod in each of the out-of-sequence groups.
When the control rod movement is within the group notch mode and as soon as automatic initiation of the RSCS occurs during rod insertion while shutting down, the operability of the notching restriction shall be demonstrated by attempting to move a control rod more than one notch in the first programned rod group,
- b. Failed Position Switch b. Failed Position Switch Control rods with a failed " Full- A second licensed operator shall in" or " Full-out' position switch verify the conformance to Spect-may be bypassed in the Rod Se- fication 3.3.G.2.b before a rod quence Control System if the ac- may be bypassed in the Rod Se-tual rod position is known. These quence Control System.
rods shall be moved in sequence to
. their correct positions (full in on insertion or full out on withdrawal).
HATCH - UNIT 1 3.3-6 w-m . --w-. J., ,m .w - - --
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE RE0VIREMENTS 3.3.G.2.c. Shutdown Marcin/ Scram 4.3.G.2.c. Shutdown Marcin/ Scram Time Time Testino Testino In order to perform the Prior to control rod with-required shutdown margin drawal for startup, verify demonstrations subsequent the conformance to Speci-to any fuel loading opera- fication 3.3.G.2.b. before tions, or to perform con- a rod may be bypassed in trol rod drive scram and/or the RSCS. The requirements friction testing as specified to allow use of the indi-in Surveillance Requirement vidual rod position bypass 4.3.C.2 and the initial start- switches within rod groups up test program, the relaxa- A12. A34, 812 or 834 of tion of the following RSCS the RSCS during shutdown restraints is permitted. The margin, scram time or fric-sequence restraints imposed tion testing are:
on control rod groups A12
, A34, 812, or 834 after 50% (1) RWM operable as per Spect-of the control rods have been fication 3.3.G.I.
withdrawn may be removed for the test period by means of the (2) After the bypassing of individual rod position bypass the rods in the RSCS groups switches.
A12. A34' 812. OF B34 for test purposes, it shall be demonstrated that movement
" of the rods in the 50% dens-ity to the preset power level range is blocked or limited to the single notch mode of withdrawal.
(3) A second licensed operator shall verify the conformance to procedures and this Specification.
H. Shutdown Recuirements If Specifications 3.3.A through 3.3.G are not met, an orderly shutdown shall be initiated and the reactor placed in the Cold Shutdown Condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
HATCH - UNIT 1 3.3-7
BASES FOR LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REOUIRENENTS 3.3.6.1. Rod Worth Minimizer (RW) (Continued)
In performing the function described above, the Rm and RSCS are not re-quired to impose any restrictions at core power levels in excess of 205 of rated. Material in the cited references shows that it is impossible to reach 280 calories per gram in the event of a control rod drop occur-ring at power greater than 20%, regardless of the rod pattern. Tisis is true for all normal and abnormal patterns including those which maximize the individual control rod worth.
At power levels below 20% of rated, abnormal control rod patterns could produce rod worths high enough to be of concern relative to the 280 cal-orie per gram rod drop limit. In this range of RWM and the RSCS con-strain the control rod sequences and patterns to those which involve only acceptable rod worths.
The Rod Worth Minimizer and the Rod Sequence Control System provide auto-matic supervision to assure that out of sequence control rods will not be withdrawn or inserted; i.e., it limits operator deviations from plan-ned withdrawal sequences. They serve as a backup to procedural control of control rod sequer.ces, which limit the maximum reactivity worth af control rods. In the event that the Rod Worth Minimizer is out of ser-vice, when required, a second licensed operator or other qualified tech-nical plant employee whose qualifications have been reviewed by the AEC can manually fulfill the control rod pattern conformance functions of this system.
l The functions of the RWM and RSCS 9ake it unnecessary to specify a license limit on rod worth to preclude unacceptable consequences in the event of a control rod drop. At low powers, below 20%, these devices force ad-herence to acceptable rod patterns. Above 20% of rated power, no con-sequences are acceptable. Control rod pattern constraints above 20% of rated power are imposed by power distribution requirements as defined in Section 3.11 and 4.11 of these Technical Specifications. Power level for automatic cutout of the RSCS function in sensed by first stage turbine pressure. Because the instrument has an instrument error of 1,10% of full power the nominal instrument setting is 30% of rated power. Power level for automatic cutout of the RWM function is sensed by feedwater and steam flow and is set nominally at 30% of rated power to be consistent with the RSCS setting. . .-
} Surveillance Requirements:
~
Functional testing of the RWM prior to the start of control rod withdrawal at startup,' and prior to attaining 20% of rated thermal power during rod in-sertion while shutting down, will ensure reliable operation and minimize the probability of the rod drop accident.
- 2. Rod Secuence Control System (RSCS)
- a. Operability Limiting Conditions for Operation:
See bases for Technical Specification 3.3.G.I. Rod Worth Minimizer.
HATCH - UNIT 1 3.3-16
BASES FOR LINITING CONDITIONS FOR OPERATION AND SURVEll(ANCE REQUIRENENTS
-3.3.6.2.a. Doerability Surveillance Requirements:
The RSCS can be functionally tested after 50% of the control rods have been withdrawn, by. demonstrating that the continuous withdrawal mode for the control drives is inhibited.
This demonstration is made by attempting to withdraw a control rod more than one notch in the first programmed rod group subsequent to reaching the 50% rod density point. This restriction to the notching mode of op-eration for control rod withdrawal is automatically removed when the re-actor reaches the automatic initiation setpoint.
During reactor shutdown, similar surveillance checks shall be made with regard to rod group availability as soon as automatic initiation of the RSCS occurs and subsequently at appropriate stages of the control rod insertion.
- b. Failed Position Switch Limiting Conditions for Operation:
In the event that a control rod has a failed " Full-in" or " Full-out' position switch, it may be bypassed in the Rod Sequence Control System if its position is otherwise known. It is a safer and more desirable condition for such rods to occupy their proper positions in the control rod patterns during reactor startup or shutdown.
Surveillance Requirements:
Having a second licensed operator verify the actual rod position prior to bypassing a rod in the Rod Sequence Control System provides assurance that Specification 3.3.G.2.b. is met.
- c. Shutdown Narain/ Scram Time Testino After initial fuel loading and subsequent refuelings when operating above 950 psig all control rods shall be scram tested within the constraints imposed by the RSCS and before the 40%. power level is reached. To mein-tain the required reactor pressure conditions the individually scrammed or inserted rod should be withdrawn to its original position immediately following testing of each rod. In order to select and withdraw the scram-med or inserted insequence control rod (also to select and insert a fully withdrawn insequence rod in case of friction testing) it will be neces-sary to simulate all the insequence withdrawn rods of the succeeding RSCS groups as being at full in position by utilizing the individual rod post-HATCH - UNIT 1 3.3-17 e
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.11. FUEL R005 4.11. FUEL ROOS Applicability Applicability The Limit 1'n g Conditions for Operation The Surveillance Requirements apply associated with the fuel rods apply to to the parameters which monitor the those parameters which monitor the fuel rod operating conditions, fuel rod operating conditions. -
Objective '
Objective The Objective of the Limiting Condi- The Objective of the Surveillance tions for Operation is to assure the Requirements is to specify the type performance of the fuel rods. and frequency of surveillance to be applied to the fuel rods.
Specifications Specifications A. Average Planar Linear Heat Genera- A. Average Planar Linear Heat Genera-tion Rate (APLHrR1 2 tion Rate ( APLHGR)
During power operation, the APLHGR The APLHGR for each type of fuel as for all core locations shall not a function of average planar exceed the appropriate APLHGR limit exposure shall be determined daily for those core locations. The APLHGR during reactor operation at 125%
limit, which is a function of average rated thermal power.
planar exposure and fuel type, is the appropriate value from Figure 3.11-1, sheets 1 through 6. multiplied by the i smaller of the two MAPFAC factors de-termined f rom Figure 3.11-1, sheets 7 and 8. If at any time during oper- l ation it is determined by normal surveillance that the limiting value for APLHGR is being exceeded, action shall be initiated within 15 minutes to restore operation to within the pre- -
scribed limits. If the APLHGR is not returned to within the pre-scribed limits within two (2) hours, ' *
- then reduce reactor power to less than 25% of rated thermal power with-in the next four (4) hours. If the limiting condition for operation is restored prior to expiration of the specified time interval, then further progression to less than 25% of rated thermal power is not required. '
During power operation, the LHGR as The LHGR as function of core a function of core height shall not height shall be checked daily dur-exceed the limiting value shown in ing reactor operation at 125%
Figure 3.11-2 for 7 x 7 fuel or the rated thermal power, limiting value of 13.4 kw/f t for any
- 8 x 8 fuel. If at any time during HATCH - UNIT 1 3.11-1
==
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LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.11.B. Linear Heat Generation Rate (LHGR)
(Continued) operation it is determined by normal surveillance that the limiting value for LHGR is being exceeded, action shall be initiated within 15 minutes to restore operation to within the prescribed limits. If the LHGR is not returned to within the prescribed limits within two (2) hours, then reduce reactor power to less than 25% of rated thermal power within the next four (4) hours. If the limiting condition for operation is restored prior to expiration of the specified time interval, then further progression to less than 25%
of rated thermal power is not required.
C.
Minimum Critical Power Ratio (MCPR)4.11.C.I. Minimum Critical Power Ratio (MCPR)
The minimum critical power ratto (MCPR) MCPR shall be determined to be shall be equal to or greater than the equal to or greater than the operating limit MCPR (OLMCPR), which applicable limit, daily during is a function of scram time, core reactor power operation at t 25%
power, and core flow. For 25% $ rated thermal power and following power < 30%, the OLMCPR is given in any change in power level or dis-Figure 3.11.6. For power 2 30%, tribution that would cause opera-l the OLMCPR is the greater of either: tion with a limiting control rod pattern as described in the bases
- 1. The applicable limit determined for Specification 3.3.F.
f rom Figure 3.11.3, or 4.11.C.2. Minimum Critical Power Ratio Limit
- 2. The applicable limit from either Figures 3.11.4 or 3.11.5 The MCPR limit at rated flow and multiplied by the Kp factor rated power shall be determined for determined from Figure 3.11.6, each fuel type, 8X8R, P8X8R, BP8x8R or where: l 7X7 from figures 3.11.4 and 3.11.5 respectively using:
t = 0 or 'tave TB' , whichever is a. 5-1.0 prior to initial scram greater time measurements for the.
" tA tB -
cycle, performed in accordance TA = 0.90 sec (Specifications 3.3.C.2.a. wlth specifications 4.3.C.2.a.
scram time limit to 20% insertion from fully withdrawn) or t/s b. t as defined in specification rg = 0.710+1.65 N1 (0.053) Otef.103 3.11.C.
I Nj The determination of the limit i=1 must be completed within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />
. of the conclusion of each scram tide surveillance test required by specification 4.3.C.2.
HATCH - UNIT 1 3.11-2
. . ..- . - _ . - - .. _ . . . .___ _ . - _ -._ -..~ - - - .
BASES FOR LINITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIRENENTS 3.11. FUEL ROOS A. Averace Planar Linear Heat Generation Rate (APLHGR)
This specification assures that the peak claddi g temperature following the postulated design basis loss-of-coolant accident will not exceed the limit l specified in the 10 CFR 50, Appendix K, even considering the postulated
+
. effects of fuel pellet densification. ,
The peak cladding temperature following a postulated loss-of-coolant acci-dent is primarily a function of the average heat generation rate of all the rods of a fuel assembly at any axial location and is only dependent second-I arily on the rod to rod power distribution within an assembly. Since ex-pected local variations in power distribution within a fuel assembly affect the calculated peak clad temperature by less than i 20'F relative to the peak temperature for a typical fuel design, the limit on the average linear heat generation rate is sufficient to assure that calculated temperatures
' conform to 10 CFR 50.46. The limiting value for APLHGR at rated conditions '
is shown in Figures 3.11.1. sheets 1 thru 6.
l
- A flow dependent correction factor incorporated in to Figure 3.11-1 (sheet 8) is l applied to the rated conditions APLHGR to assure that the 2200*F PCT limit is complied with during LDCA initiated from less than rated core flow. In addition, other power and . flow dependent corrections given in Figure 3.11-1 (sheets 7 and 8) are applied to the rated conditions APLHGR limits to assure
' that the fuel thermal-mechanical design criteria are met during abnormal l transients initiated from off-rated conditions.
The calculational procedure used to establish the APLHGR shown in Figures i 3.11.1, sheets 1 thru 6, is based on a loss-of-coolant accident analysis. l The analysis was performed using General Electric (GE) calculational models which are consistent with the requirements of Appendix K to 10 CFR 50. A complete discussion of each code employed in the analysis is presented in Reference 1. Differences in this analysis as compared to previous analyses , .
performed with Reference 1 are: (1) The analyses assume a fuel assembly planar power consistent with 102% of the MAPLHGR shown in Figure 3.11.1; (2) Fission product decay is computed assuming an energy release rate of 200 MEV/ Fission; (3) Pool boiling is assumed after nucleate boiling is lost during the flow stagnation period; (4) The effects of core spray entrainment and counter-current flow limiting as described in Reference 2 are included in the reflooding calculations.
A list of the significant plant input parameters to the loss-of-coolant accident analysis is presented in Table 1 of NE00-21187(s). Further discussion of the APLHGR bases is found in NEDC-30474-p(11).
~
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r 1
1 HATCH - UNIT 1 3.11-3 h
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-w -y--,.g,-- >,w-- e, rww-p., -- y--- -- g--- y-- - - -
n-T-W'T9--Pn$---w--rwru---y=ri"'-"-w-"m9'7N-F"mNw"-- -+---f'm'w' T T' w
BASES FOR LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3.11.C. Minimum Critical Power Ratio (MCPR) (Continued)
The purpose of the MCPRg, and the Kp of Figures 3.11.3 and 3.11.6, respectively, is l!
to define operating limits at other than rated core flow and power conditions. At less than 1005 of rated flow and power, the required MCFR is the larger value of the MCPRf and MCPR, at the existing core flow and power state. The MCPRgs are established to protect the core from inadvertent core flow increases such that the 99.95 MCPR limit requirement can be assured.
l The MCPRgs were calculated such that for the maximum core flow rate and the corres-ponding THERMAL POWER along the 1055 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 1055 of rated steam flow control line corresponding to dif forent '
core flows. The calculated MCPR at a given po;nt of core flow is defined as MCPRf .
The core power dependent MCPR operating limit MCPR a ' is the power rated flow MCPR operating limit multiplied by the Kp factor given In Figure 3.11.6. ll The Ky s are established to protect the core from transients other than core flow increases, including the localized event such as rod withdrawal error. The Kys ,
were determined based upon the most limiting transient at the given core power level. (For.further information on MCPR operating limits for of f-rated conditions, reference NEDC-30474-P.(11))
4 l
t MATCH - UNIT 1 3.11 -4a l
' FUEL T PE 7X7 '
IC 1&2 m 80 MIL CHANNELS l
$ 14.5
~[
J 4m 1 14.4 14.1 14h(14.2 k_a E
- a. Z 13 g9 h 12./
12 46 \
30 11.5 2k 11 Ry 10 102
~
0 5 10 15 20 25 30 35 40 AVERAGE PLANAR EXPOSURE (GWd/t) l0 FUEL TYPE 7X7 m IC 3
$; 80 MIL CHANNELS l
14.7
$w I4 14.5 \ 14.3 14p 14.2 k,
_m. k a2 89 13 _
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11.5 yI 11
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m 10.3 0 5 10 15 20 ,
25 30 35 40
' ~
AVERAGE PLANAR EXPOSURE (GWd/t)
FIGURE 3.11-1 (SHEET 1) l HATCH - UNIT 1 .
F
14 m
4 W, FUEL TYPE: 8D250 13 28 }e
.x 12.3 80 MIL CHANNELS 12.2 kW b N E9 \ 11.5
- a. z ,
11.3
$9 11 11.2
<$ h 10.6
<w IC 10 \b NI 2
9.0 8
0 5 10 15 20 25 30 35 40 AVERAGE PLANAR EXPOSURE (GWd/t) 14 i i ,
i FUEL TYPES: 8DRB265H, g P8DRB265H, AND BP8DRB265H
< 13 80 AND 100 MIL CHANNELS
$E 33 12.1 12.1
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$9 11 N 10.7
.$E s 10.2
'o R$ N9.6 sc EY ng S
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'8 0 5 10 15 20 25 30 35 40 AVERAGE PLANAR EXPOSURE (GWd/t)
FIGURE 3.111 (SHEET 2)
HATCH - UNIT 1
14 FUEL TYPES: P8DRB284H E
AND BP8DRB284H 6= 13 80 MIL CHANNELS 8 .x e- 12.0 12.0 j
n<_,0 go 2k x ~
9.9 EE 9 4
2 8
0 5 10 15 20 25 30 35 40 AVERAGE PLANAR EXPOSURE (GWd/t) 14 x
4 FUEL TYPE: 8DR183 (8DRB175)
EE 13 100 MIL CHANNELS 31 k *l 12.0 12.0 11.4
$2 11 m m as 10 %.4
!5 E$
4 2
g
\ 92 8
0 5 10 15 20 25 30 35 40 AVERAGE PLANAR EXPOSURE (GWd/t)
FIGURE 3.11 1 (SHEET 3)
HATCH - UNIT 1
16 i i i i FUEL TYPE: 80R233 (8DRB221) 100 MIL CHANNELS l
- 2 2s 14 3} UNACCEPTABLE OPERATION NN 11.9 12.1 12.0 11,9 w$ AM5 h 11.3 kk EE
.8 10.2 ACCEPTABLE OPERATION 9.6 aa E$
4 8 2
6 e 0 5 10 15 20 25 30 35 40 45 50 AVERAGE PLANAR EXPOSURE (GWd/t) 16 . . . .
FUEL TYPES: P8DRB284LA AND BP8DRB284LA k 100 MIL CHANNELS
$k 14 3j UNACCEPTABLE OPERATION z~
4W
$$ 11.8 12.0 12.1 12.1 12.0 N 2 \ 11.5 y9 Il N .8
$j ' <10.1 Is ACCEPTABLE OPERATION sa pg 8
2 6
0 5 10 15 20 25 30 35 40 45 50 AVERAGE PLANAR EXPOSURE (GWd/t)
FIGURE 3.11 1 (SHEET 4)
M l l l FUELTYPE: P90RB283 l 100 MIL CHANNELS l l l l r UNACCEPTA8LE OPERATIC N .
I~
5Ia 12.1 h ,, 11.9 12.1 g g,g 11.3 w$
11.3 % *
$ ACCEPTA8LE OPERATION 3 9.8 U
lg 9.2
=5 iz I e o
e s to a se as se a se es no AVERAGE PLANAR EXPOSURE (GWd/t) l I I l l FUEL TYPE 5: P90RB283 AND BP90R8283 80 MIL CHANNELS 16 14
" UNACCEPTABLE
{ OPERATION 3BE 13 E
9 fI
. 5 12 ir.o- 12.1 1
11.4 (gE ii \ !
z w i 3 >- ACCEPTABLE Ed OPERATION loJ N
- Io N ... .
. \ ~ .,
8 o s to is no as so as AVERAGE PLANAR EXPOSURE (Gwd/t)
FIGURE 3.11 1 (SHEET 5)
HATCH - UNIT 1 l l
18 , . .
FUEL TYPE: BP8DRB299 80 AND 100 MIL CHANNELS 16 m
$~ 14 UNACCEPTABLE
- zE OPERATION 35 12.1 12.1 12.0
$ , k k 11.0
'Z
" 9 10 10.9 \ A 10.3 ACCEPTABLE , 9,0 Ww OPERATION 35 8 zo DH 24 R$ 6 E
4 2
0 0 5 10 15 20 25 30 35 40 45 50 AVERAGE PLANAR EXPOSURE (Gwd/t)
FIGURE 3.111 (SHEET 6) 4 HATCH - UNIT 1 i
1.0 2
3=
-4 O
3 -
, I C
1 2
- Q 0.9 -
"n N
t j
j OA 5
N g MAPLHGRp = MAPFACp
- MAPLHGR STD j MAPLHGR STD = STANDARD MAPLHGR LIMITS J 0.7 g FOR 25% > P. NO THERMAL LIMITS MONITORING REQUIRED 2 NO LIMITS SPECIFIED e <
~ 50% CORE FLOW 2 _
FOR 25% $P < 30%:
Q MAPF ACp = OA06 + 0.05224 (P - 3014 w
FOR$50% CORE FLOW
$ 0.6 - M APF ACp = 0.433 + 0.05224 (P - 30%)
m FOR > 50% CORE FLOW w
l FOR 30% $P:
g MAPFACp = 1.0 + 0.006224 (P - 100%)
> 50% CORE FLOW l
05 -
l 3 I
1 i
l I 0.4 I I ! I I ! ! ! I I I I I I
, 20 25 30 40 50 00 70 SO 90 100 POWE R (% R ATED)
FIGURE 3.111 (SHEET 7) MAPFACp
Z
?
9
' i.00 C
Z Q g 102A%
w 107A%
P ti
~z -
/ 112.0%
, MAX FLOW = 117A%
0.90 - '
O.06 - -
1 a
} OAO -
MAPLHGR, = MAPFAC, eMAPLHGRSTO M APLHG R STO, = STANDARO MAPLHOR LIMITS 3, MAPFACp = MINIMUM (MFRPOp ,MAPMULYp l e
o MAPMULT = 1 A FOR FLOW > St%
O 4 0.70 - = 0.88 FOR FLOWSS1%
MFRPOp lF) = MINIMUM {1.0, Ap + Sp F) g F = FR ACTION OF RATED CORE FLOW, g
AND A y, B, ARE FUEL TYPE DEPENDENT g CONSTANTS CIVEN SELOW:
4 0.00 -
FOR 7X7, MAMW SXS,SXSR FOR PSMSR CORE FLOW A E AF EF
(% R ATED) P F 102.5 0.4898 0.8467 0.4881 0.0794 107.0 0.4421 0.8633 0.4674 0.8754 112.0 0.4074 0.6581 0.4214 0.0007 0.50 - 117.0 0.3701 0.0866 0.3828 0.0008 t
0.40 8 I I I I I 30 40 80 70 a0 s0 00l 100 110 81 CORE FLOW (% RATEDI j FIGURE 3.111 (SHEET 8) MAPFACp
TOP 12 LIMITING VALUE FOR LHGR AS A FUNCTION OF CORE -
11 HEIGHT
= 18.5 [1 .026 (f2II 10 9
WHERE: Z = HEIGHTIN FEET 8
ABOVE BOTTOM OF THE CORE I '
N \
a 7 5
m Z
8 e
8 5 3 4
PERMISSIBLE REGION 3 OF OPERATION 2
1 BOTTOM 0 17.9 18.0 18.1 18.2 18.3 18.4 18.5 18.6 LHGR (kw/ft) -
FIGURE 3.11-2 LIMITING VALUE FOR LHGR FUEL TYPE 7X7 HATCH - UNIT 1
6 e
1.35 1.34 ,
,_u - /
I.32 1.31 e
/
w 3 f.30 E /
'V
. i .2,
/
1.28 1.27 1.26 1.25 0.0 1.0 T
FIGURE 3.11.4 MCPR LIMIT FOR ALL 8X8 FUELTYPES FOR RATED POWER AND RATED FLOW l
1
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)
HATCH - UNIT 1 w- m--
e 1.31 1.30
.., 1 1.28 1.27 )
t~
s 3 1.26 m
b 2 1.25 1.24 1.23 1.22 0.0 13 FIGURE 3.11-5 MCPR LIMIT FOR 7X7 FUEL !
FOR RATED POWER AND RATED FLOW HATCH - UNIT 1
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%0CT d WO A (dM1 WElldli1nW WJ3W 031VW 1EOC) d5 iMit WOd W3dW10 HATCH - UNIT 1
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 l those fuel assemblies which have been analyzed with NRC approved codes and methods I and have been shown to comply with all Safety Design Bases in the Final Safety Analysis Report (FSAR).
- 2. Control Rods l
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)
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 Storage
- 1. Spent Fuel All arrangement of fuel in the spent fuel storage racks shall be maintained in a subcritical configuration having a kept not greater than 0.95.
- 2. New Fuel !
The new fuel storage vault F 11 be such that the ek rr dry shall not be greater than 0.90 and the ek rr f'% de shall not be greater than 0.95.
HATCH - UNIT 1 .0-1 f 1
l 5.0.F. Seis~ic Desian The reactor building and all enhineered safeguard systems are designed for the design basis earthquake with a horizontal. ground acceleration of 0.15 g.
The operating basis earthquake has a horizontal ground acceleration of 0.08 g.
- 6. References
- 1. FSARSection4.2,ReactorVesselandAphurtenancesMechanicalDesign
- 2. FSAR Section 5.2, Primary Containment System
- 3. FSAR Section 5.3, Secondary Containment System
- 4. FSAR Section 12.4.4, Governing Codes and Regulations
- 5. FSAR Section 10.3, Spent Fuel Storage
- 6. FSAR Section 10.2, New Fuel Storage l
l l
I HATCH - UNIT 1 5.0-2
i REACTIVITY CONTROL SYSTEMS 3/4.1.4 CONTROL ROD PROGRAM CONTROLS ROO WORTH MINIMIZER LIMITING CONDITION FOR OPERATION 3.1.4.1 The Rod Worth Minimizer (RWM) shall be OPERABLE.
APPLICABILITY: CONDITIONS 1 and 2*, when THERMAL POWER is less than 20%
of RATED THERMAL POWER.
ACTION:
With the RWM inoperable, the provisions of Specification 3.0.4 are not applicable, operation may continue and control rod movement is permitted provided that a second licensed operator or other qualified member of the technical staff is present at the reactor control console and verifies compliance with the prescribed control rod pattern.
SURVEILLANCE REQUIREMENTS 4.1.4.1 The RWM shall be demonstrated OPERABLE:
- a. In CONDITION 2 prior to withdrawal of control rods for the purpose of making the reactor critical, and in CONDITION 1 when the RWM is initiated during control rod insertion when reducing THERMAL POWER, by:
- 1. Verifying proper annunciation of the selection error of at least one out of-sequence control rod, and
- 2. Verifying the rod block function of the RWM by moving an out-of-sequence control rod.
- b. By verifying that the Banked Position Withdrawal Sequence input to the RWM computer is correct following any loading of the sequence program l into the computer.
- Entry into OPERATIONAL CONDITION 2 and withdrawal of selected control rods is permitted for the purpose of determining the OPERABILITY of the RWM prior to withdrawal of control rods for the purpose of bringing the reactor to criticality.
HATCH - UNIT 2 3/4 1-14 e
REACTIVITY CONTROL SYSTEMS R00 SEQUENCE CONTROL SYSTEM LIMITING CONDITION FOR OPERATION 3.1.4.2 The Rod Sequence Control System (RSCS) shall be OPERABLE.
APPLICABILITY: CONDITIONS 1* and 2*#, when THERMAL POWER is less than 20%
of RATED THERMAL POWER and control rod movement is within the group notch mode after 50% of the control rods have been withdrawn.
ACTION:
With the RSCS inoperable control rod movement shall not be permitted, except by a scram.
SURVEILLANCE REQUIREMENTS 4.1.4.2 The RSCS shall be demonstrated OPERABLE by:
- a. Selecting and attempting to move an inhibited control rod:
- 1. As soon as the group notch mode is entered during each reactor startup, and
- 2. As soon as the rod inhibit mode is automatically initiated during control rod insertion.
L 4
- See Special Test Exception 3.10.2.
- Entry into CONDITION 2 and withdrawal of selected control rods is permitted for the purpose of determining the OPERABILITY of the RSCS prior to withdrawal of control rods for the purpose of bringing the
, reactor to criticality.
HATCH - UNIT 2 3/4 1-15
REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
- b. Attempting to move a control rod more than one notch as soon as the group notch mode is automatically initiated during control rod: -
- 1. Withdrawal each reactor startup, and
- 2. Insertion.
- c. Performance of the comparator check of the group notch circuits prior to control rod;
- 1. Movement within the group notch mode during each reactor startup, and
- 2. Insertion to reduce THERMAL POWER to less than 20% of RATED THERMAL POWER.
l l
t HATCH - UNIT 2 3/4 1-16 3
l 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE LIMITING CONDITION FOR OPERATION 3.2.1 ALL AVERAGE PLANAR LINEAR HEAT GENERATION RATES (APLHGRs) shall be equal to or less than the applicable APLHGR limit, which is a function of fuel type and AVERAGE PLANAR EXPOSURE. The APLHGR limit is given by the applicable rated power, rated-flow limit taken from Figures 3.2.1-1 through 3.2.1-11, l multiplied by the smaller of either: *
- a. The factor given by Figure 3.2.1-12, or l
- b. The factor given by Figure 3.2.1-13. l APPLICABILITY: CONDITION 1, when THERMAL POWER 2 25% of RATED THERMAL POWER.
ACTION:
With an APLHGR exceeding the limits of Figures 3.2.1-1 through 3.2.1-11, as adjusted per Figures 3.2.1-12 and 3.2.1-13, initiate corrective action within 15 minutes and continue corrective action so that the APLHGR meets 3.2.1 within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hourt.
SURVEILLANCE REQUIREMENTS 4.2.1 All APLHGRs shall be verified to be equal to or less than the applicable limit determined from Figures 3.2.1-1 through 3.2.1-11, as adjusted per Figure 3.2.1-12 and 3.2.1-13:
- a. At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />,
- b. Whenever THERMAL POWER has been increased by at least 15% of RATED THERMAL POWER and steady state operating conditions have been estabitshed, and
- c. Initially and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor is operating with a LIMITING CONTROL ROD PATTERN for APLHGR.
HATCH - UNIT 2 3/4 2-1
\
5 a
x to E
a ro k UNACCEPTABLE OPERATION
$k 3.!
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<$ ACCEPTABLE OPER ATION IC D >-
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0 5 10 15 20 25 30 35 40 45 50 AVERAGE PLANAR EXPOSURE (GWd/t)
FUEL TYPE 8D1B175 (80RL183) 100 MIL CHANNELS l
MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR) VERSUS AVERAGE PLANAR EXPOSURE FIGURE 3.2.1 1
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J FUEL TYPE 8018221 (8DRL233) 100 MIL CHANNELS l
MAXIMUM' AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR) VERSUS AVERAGE PLANAR EXPOSURE FIGURE 3.2-1-2 a
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FUEL TYPE IE 711-00GD-100 MIL CHANNELS MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION l RATE (MAPLHGR) VERSUS AVERAGE PLANAR EXPOSURE i FIGURE 3.2.1-3 e
4 M
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FUEL TYPES P8DRB284LA AND BP8DRB284LA 100 MIL CHANNELS -
MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION l RATE (MAFLHGR) VERSUS PLANAR EXPOSURE FIGURE 3.2.1-4
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FUEL TYPES P80RB283 AND BP8DRB283100 MIL CHANNELS MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR) VERSUS AVERAGE PLANAR EXPOSURE FIGURE 3.2.1-5 O
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._ FUEL TYPES P8DRB265H AND BP8DRB266H 80 AND 100 MIL CHANNELS MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR) VERSUS AVERAGE PLANAR EXPOSURE FIGURE 3.2.17
E M
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FUEL TYPE 80RB265H 80 MIL CHANNELS l l MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR) VERSUS AVERAGE PLANAR EXPOSURE FIGURE 3.2.1-8
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FUEL TYPES P80RB284H AND BP8DRB284H 80 AND 100 MIL CHANNELS l MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR) VERSUS AVERAGE PLANAR EXPOSURE FIGURE 3.2.1-9
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l FUEL TYPES P90RB283 AND BPSDRB283 80 MIL CHANNELS l MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHOR) VERSUS AVERAGE PLANAR EXPOSURE FlGUR E 3.2.1 10
% 18 a
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- m
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FUEL TYPE BP80RB299 80 AND 100 MIL CHANNELS
- MAXI'10M AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR) VERSUS AVERAGE PLANAR EXPOSURE FIGURE 3.2.1-11
h
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- 2:
s 1A0 C
b 1-AE
~ [ [ [ 197AE t%++ 1, / .
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- 117AE 0.90 -
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l 0.80 -
u MAPLHGR, = MAPFACp eMAPLHGR STD 4
g
< MAPLHGRSTD = 5 A DARD MAPLHGR L.lu ,
M
{ MAPFACp = MINIMUM (MFRPOg ,MAPMULTy l e
) O u
MAPMULT = 1 A FOR FLOW > 41%
N 4 = 0.. FOR FLOWSSt%
0.70 -
E
" {o MFRPOp lFl = MINIMUM (1.0, Ap + SpF) g F = FRACTION OF RATED CORE FLOW, r AND Ay, SpARE FUEL TYPE DEPENDENT C
.a CONSTANTS GIVEN SELOW:
4 2 0.00 -
FOR 7X7, M AXIMUM SMS,8XSR FOR PSXSR CORE FLOW
(% R ATED) AF E A EF F F 102.5 0.4088 0.8867 0.4881 0.8704 107.0 0.4421 0.8633 0.4874 0.8758 112.0 0.4074 0.8531 0.4214 0.0007 050 - 117.0 0.3701 0.8858 0.3828 0.8888 I , i i i
- 0. 0 i 1
- 0 . 20 . .
.I,6 ,. ,,0 CORE FLOW 1% RATED)
, FIGURE 3.2.1-12 MAPFAC p .
l
12 h
M x -
e E
4 0 -
ro 1
W m
> en -
E 5
y -
g MAPLHORp = MAPF ACo
- MAPLHORSTD M
E1 s
g MAPLHORSTO = MANDARO MMHOR LWIM FOR 25%>P: NO THERMAL LIMITS MONITORING REOUIRED A 2 NO LIMITS SPECIFIED
> $ 50% CORE FLOW 7 2 FOR 25% $P <30%:
pc- MAPF ACp = 0385 + 0.m224 (P - 30%)
Q g FOR$50% CORE FLOW MAPFACp = 0.433 + OAe5224 (P - 30%)
o 0.s -
E FOR > 504 CORE FLOW FOR 30% $P:
- MAPFACp = 1.0 + 0.008224 (P - 100%I g
> 50% CORE FLOW l
02 -
I '
I I
l 1
0.4 I I I I I I I I I I I I I I 20 25 30 40 50 90 70 90 90 100 POWER (% R ATED)
FIGURE 3.2.1-13 MAPFACp
POWER DISTRIBUTION LIMITS ;
3/4.2.3 MINIMUM CRITICAL POWER RATIO LIMITING CONDITION FOR OPERATION 3.2.3 ALL MINIMUM CRITICAL POWER RATIOS (MCPRs), shall be equal to or greater than the MCPR operating ifmit (OLMCPR), which is a function of average scram time, core flow, and core power. For 25% 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
- 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 = 0'er
- ave *B , whichever is greater, tg tB.
t g = 1.096 sec (Specification 3.1.3.3 scram time limit to notch 36),
~
N 12 (0.059),
t g = 0.834 + 1.65 1 n ;
EN-i i=1 n
E Njrj t,y, = i=1 n N E
i=1 n= number of surveillance tests performed to date in cycle, th surveillance Ng = number of active control rods measured in the i test, t g = average scram time to notch 36 of all rods measured in the i th surveillance test, and Ny = total number of active rods measured in 4.1.3.2.a.
APPLICABILITY: CONDITION 1, when THERMAL POWER 2 25% RATED THERMAL POWER ACTION:
With MCPR less than the applicable limit determined from Specification 3.2.3.a, or 3.2.3.b, initiate corrective action within 15 minutes and continue corrective action so that MCPR is equal to or greater than the applicable
/ limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POWER to less than or equal to 25% of l RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
HATCH - UNIT 2 3/4 2-6 l-
, _ _ _ . . . _ _ _ _ _ _ . . _ _ = _ . _ . _ _ ___ _ _ _ _ _ ._ __
3/4.2.3 MINIMUM CRITICAL POWER RATIO (CONTINUED)
SURVEILLANCE REQUIREMENTS 4.2.3 The MCPR limit at rated flow and rated power shall be determined for each type of fuel (8X8R, P8X8R, BP8X8R, and 7X7) from Figures 3.2.3-1 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 as defined in Specification 3.2.3; the determination of the limit must be completed within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of the conclusion of each scram time surveillance test required by Specification 4.1.3.2.
MCPR shall be determined to be equal to or greater than the applicable limit:
- a. At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />,
- b. Whenever THERMAL POWER has been increased by at least 15% of RATED THERMAL POWER and steady state operating conditions have been established, and
- c. Initially and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor is operating with a LIMITING CONTROL ROD PATTERN for MCPR.
HATCH - UNIT 2 3/4 2-7
1.3 1.37 i 1.30
/
f,
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/
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=
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0
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.I 1.29
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1.2B 0.0 0.2 OA 0.4 03 1.0 T
ALL 8X8 FUEL TYPES FIGURE 3.2.3-1 l HATCH - UNIT 2
/ 2-7a 3'4
b o
e O
1.33 1.32 1.31 ACCEPTABLE OPER ATION 1.30 1.29 E
1.28 1.27 1.26 UN ACCEPTABLE OPE R ATION 1.25 1.24 1.23 0.0 0.2 0.4 0.8 0.8 1.0
. T FIGURE 3.2.3 2 l
MCPR LIMIT FOR 7X7 FUEL AT RATED FLOW AND RATED POWER l
l lIATCH - UNIT 2 3/4 2-7b ,
r-2 8
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HATCH - UNIT 2- - -- --
3/4 2-7c
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1 2.0 -
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i Z I g 2.5 -
OPERATIOGG LlettT RACPR OP) = pK eOPERAT6000 LItstT RACPR {tegl 2
l >50EFLOW
- y* FOR P <39E
- NO THERRAAL LIR84TS Rs000tTORefeG REQUIREO 2.4 -
NO LamelTS SPECIFIED t
j FOR 25%f PC Pgypggg: iPgypagg = M FM N- W t
~
i l K, = (Kgy, + OM (M - MNM (M Kgy, = 2A0 FOR$50% CORE FLOW
- ,, _ l = 2.4o FOR > son CORE FLOW 4 FOR 35Ef P<404
, e. g, E Kg = 1.M + OA134 (40% - PI go2.1 -
l vg FOR 46%$P<e0E:
y lw y
A 2.8 g
- l FOR 80E(P:
Kp = 1.15 + sages 7 (GSE - Pl O ',25g k3 b 3 Kp = 1 A + omas7s (1een - P)
Q- r3 u2 13 -
%l ge l
-b o2 l
,\
. - o l
- 1.4 - l a l I
' s 1.3 -
g i
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g I I i
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1 1 i
1 1.1 -
l l 3
l I l 1
1 8 l
j 1 I I I I I I i I I I i i f 30 25 30 de so 80 7e se se tes FOWER (% RATEOl P. P FIGURE 3.2.34 Kp l
POWER DISTRIBUTION LIMITS 3/4.2.4 LINEAR HEAT GENERATION RATE LIMITING CONDITION FOR OPERATION _
3.2.4 ALL LINEAR HEAT GENERATION RATES (LHGRs) shall not exceed 13.4 Kw/ft for 8X8R/P8X8R/BP8X8R fuel or 18.0 Kw/ft for 7X7 fuel. l APPLICA8ILITY: CONDITION 1, when THERMAL POWER 225% of RATED THERMAL POWER.
ACTION:
With the LHGR of any fuel rod exceeding the limit, initiate corrective action within 15 minutes and continue corrective action so that the LHGR is within the limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
SURVEILLANCE REQUIREMENTS 4.2.4 LHGRs shall be determined to be equal to or less than the limit;
- a. At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />,
- b. When THERMAL POWER has been increased by at least 15% of RATED THERMAL POWER and steady state operating conditions have been established, and
- c. Initially and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor is operating on a LIMITING CONTROL R00 PATTERN FOR LHGR.
l l
HATCH - UNIT 2 3/4 2-8 '
l
?
- _ . _ _ _ - _ . . _ _ _ . _ _ _ _ _ - ~ _ _ . _ - _ .
M i
i REACTIVITY CONTROL SYSTEMS BASES 1
i CONTROL RODS (Continued) i
.than has been analyzed even though control rods with inoperable accumulators may still be inserted with normal drive water pressure. Operability of the
! accumulator ensures that there is a means available to insert the control rods even under the most unfavorable depressurization of the reactors.
Control rod coupling integrity is required to ensure compliance with the analysis of the rod drop accident in the FSAR. The overtravel position
' feature provides the only positive means of determining that a rod is properly coupled and therefore this check must be performed prior to
- achieving criticality after each refueling. The subsequent check is 4
! performed as a backup to the initial demonstration.
- In order to ensure that the control rod patterns can be followed and
- therefore that other parameters are within their limits, the control rod position indication system must be OPERABLE.
The control rod housing support restricts the outward movement of a control rod to less than (3) inches in the event of a housing failure.
.l The amount of rod reactivity which could be added by this small amount of rod withdrawal is less than a normal withdrawal increment and will not i
contribute to any damage to the primary coolant system. The support is not required when there is no pressure to act as a driving force to rapidly eject a drive housing.
i The required surveillance intervals are adequate to determine that the rods are OPERABLE and not so frequent as to cause excessive wear on j the system components. '
3/4.1.4 CONTROL ROD PROGRAM CONTROLS
, Control rod withdrawal and insertion sequences are established to '
l assure that the maximum insequence individual control rod or control rod segments which are withdrawn at any time during the fuel cycle could not L
} be worth enough to cause the peak fuel enthalpy for any postulated control i rod accident to exceed 280 cal /ge. The specified sequences are characterized by homogeneous, scattered patterns of control rod withdrawal. When '
- THERMAL POWER is it 20% of RATED THERMAL POWER, there is no possible rod l worth which, if dropped at the design rate of the velocity limiter, l j could result in a peak enthalpy of 280 cal /ge. Thus, requiring the RWM to i
be OPERABLE below 20% of RATED THERMAL POWER and the RSCS to be OPERABLE from
! 50% control rod density to 20% of RATED THERMAL POWER provides adequate control.
t' I
i i
j HATCH - UNIT 2 8 3/4 1-3
^
3/4.2 POWER DISTRIBUTION LIMITS' i
BASES The specifications of this section assure that the peak cladding temperature following the postulated design basis loss-of-coolant accident will not exceed the 2200*F limit specified in the Final Acceptance Criteria (FAC) issued in June 1971 considering the postulated effects of fuel pellet densification. These specifications also assure that fuel design margins are maintained during abnormal transients.
3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE This specification assures that the peak cladding temperature following the postulated design basis loss-of-coolant accident will not exceed the limit specified in 10 CFR 50, Appendix K.
The peak i:1 adding temperature (PCT) following a postulated loss-of-coolant accident is primarily a function of the average heat generation rate of all the rods of a fuel assembly at any axial location and is dependent only secondarily on the rod-to rod power distribution within an assembly. The peak clad temperature is calculated assuming an LHGR for the highest powered rod which is equal to or less than the design LHGR corrected for densification.
This LHGR times 1.02 is used in the heatup code along with the exposure dependent steady state gap conductance and rod-to-rod local peaking factor.
The Technical Specification APLHGR is this LHGR of the highest powered rod divided by its local peaking factor. The limiting value for APLHGR is shown in the figures for in Technical Specification 3/4.2.1.
The calculational procedure used to establish the APLHGR shown in the figures in Technical Specification 3/4.2.1 is based on a loss-of-coolant accident analysis. The analysis was performed using General Electric (GE) calculational models which are consistent with the requirements of Appendix K to 10 CFR 50. A complete discussion of each code employed in the analysis is
' presented in Reference 'I. Differences in this analysis compared to previous analyses performed with Reference 1 are: (1) the analysis assumes a fuel assembly planar power consistent with 102% of the MAPLHGR shown in the figures in Technical Specification 3/4.2.1;42) fission product decay is computed assuming an energy release rate of 200 MEV/rission; (3) pool boiling is assumed after nucleate boiling is lost during the flow stagnation period; and (4) the effects of core spray entrainment and counter-current flow limitation as described in Reference 2, are included in the reflooding calculations. )
A flow dependent correction' factor incorporated into Figure 3.2.1-12 is applied to the rated conditions APLHGR to assure that the 2200 F' PCT limit is complied with during a LOCA initiated from less than rated core flow. In addition, other power and flow dependent corrections given in Figures 3.2.1-12 and 3.2.1-13 are applied to the rated conditions to assure that the fuel thermal-mechanical design criteria are preserved during abnormal transients initiated from off rated conditions.
A list of the significant plant input parameters to the loss-of-coolant accident analysis is presented in bases Table B 3.2.1-1. Further discussion of the APLHGR limits is given in Reference 4.
HATCH - UNIT 2 B 3/4 2-1
I .
Bases Table B 3.2.1-1 SIGNIFICANT INPUT PARAMETERS TO THE LOSS-OF-COOLANT ACCIDENT ANALYSIS FOR HATCH-UNIT 2 Plant Parameters: -
Core Thermal Power ..................... 2531 Hwt which corresponds to 105% of license cere power
- Vessel Steam Output .................... 10.96 x 10' lbm/h which l corresponds to 105% of rated steam flow Vessel Steam Dome Pressure . . . . . . . . . . . . . 1055 psia Design Basis Recirculation Line Break Area For:
- a. Large Breaks ................... 4.0, 2.4, 2.0, 2.1 and 1.0 ft 8
- b. Small Breaks ................... 1.0, 0.9, 0.4 and 0.07 ft 8 Fuel Parameters:
PEAK TECHNICAL INITIAL SPECIFICATION DESIGN MINIMUM LINEAR HEAT AXIAL CRITICAL FUEL BUNDLE GENERATION RATE PEAKING POWER FUEL TYPE GEOMETRY (kWf t) FACTOR RATIO Initial Core 8x8 13.4 1.4 1.18 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.
- This power level meets the Appendix K requirement of 102%. The core 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 8 3/4 2-2
(
POWER DISTRIBUTION LIMITS BASES MINIMUM CRITICAL POWER RATIO (Continued)
The evaluation of a given transient begins with the system initial parameters shown in FSAR Table 15.1-6 that are input to a GE-core dynamic behavior transient computer program described in NE00-10802i" Also, the .
void reactivity coefficients that were input to the transient calculational procedure are based on a new method of calculation termed NEV which provides a better agreement between the calculated and plant instrument power distributions. The outputs of this program along with the initial MCPR form the input for further analyses of the thermally limiting bundle with the single channel transient thermal hydraulic SCAT code described in NEDO-20566' " . The principal result of this evaluation is the reduction in MCPR caused by the transient.
The purpose of the MCPR f , and the Kp of Figures 3.2.3-3 and 3.2.3-4, re- l spectively is to define operating limits at other than rated core flow and power conditions. At less than 100% of rated flow and power, the required MCPR 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 core flow increases such that the 99.9% MCPR limit requirement can be assured.
The MCPR 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 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 is the power rated flow MCPR operating limit multiplied by the K p factor given in Figure 3.2.3-4.
l 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 Kp s were determined based upon tra most limiting transient at the given core power level. For further information on MCPR operating limits for off-rated conditions, reference NEDC-30474-P.
HATCH - UNIT 2 B 3/4 2-4
5.0 DESIGN FEATURES 5.1 SITE EXCLUSION AREA 5.1.1 The exclusion area shall be as shown in Figure 5.1.1-1.
LOW POPULATION ZONE 5.1.2 The low population zone coincides with the exclusion area and is also shown in Figure 5.1.1-1.
5.2 CONTAINMENT CONFIGURATION 5.2.1 The primary containment is a steel structure composed of a series of vertical right cylinders and truncated cones which form a drywell. This drywell is attached to a suppression chamber through a series of vents. The suppression chamber is a steel pressure vessel in the shape of a torus. The primary containment has a total minimum free air volume of 255,978 cubic feet.
DESIGN TEMPERATURE AND PRESSURE 5.2.2 The primary containment is designed and shall be maintained for:
- a. Maximum design internal pressure 56 psig.
- b. Maximum allowable internal pressure 62 psig.
- c. Maximum internal temperature 340*F.
I
- d. Maximum external pressure 2 psig.
l 5.3 REACTOR CORE l
l FUEL ASSEMBLIES 5.3.1 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).
HATCH - UNIT 2 5-1
DESIGN FEATURES DRAINAGE 5.6.2 The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 185 feet.
CAPACITY 5.6.3 The spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than 2845 fuel assemblies.
5.7 COMPONENT CYCLIC OR TRANSIENT LIMIT 5.7.1 The components identified in Table L,7.1-1 are designed and shall be maintained within the cyclic or transient limits of Table 5.7.1-1.
HATCH - UNIT 2 5-4 -
O
. Georgia Power [ ENCLOSURE 4 TECHNICAL SPECIFICATIONS REVISIONS ROR RWM AND RSC5 OPERATION, FUEL STORAGE REQUIREMENTS, FUEL ASSEMBLY DESIGN, MAPLHGR LIMIT 5, EDITORIAL CHANGES PROPOSED CHANGE 5 TO TECHNICAL SPECIFICATIONS REFERENCED DOCUMENTS The enclosure provides two letters referenced in Enclosure 1 which have not been previously submitted to the Nuclear Regulatory Commission.
1 0429C 1
' 70077$
r GENERAL $ ELECTRIC NUCLEAR ENERGY BUSINESS OPERATIONS NUCLEAR TECHNOLOGIES & FUEL DIVISION 175 Curtner Avenue San Jose, CA 95125 M/C 174 January 10, 1986 cc: J. S. Charnley CJP:86-005 L. K. Mathews C. J. Paone G. D. Plotycia K. G. Turnage Mr. W. R. Mertz P. VanDiemen Southern Company Services P. O. Box 2625 Birmingham, AL 35202
SUBJECT:
High Density Spent Fuel Storage Racks at Plant Hatch
REFERENCE:
" General Electric Standard Application for Reactor Fuel,"
NEDE-24011-P- A-7, August,1985 (GESTAR)
Dear Mr. Mertz:
The following is in response to your recent request for information.
Currently, General Electric has a single high density spent fuel storage rack design. A criticality analysis, with design basis fuel, has produced the fuel bundle k-infinity storage criteria described in GESTAR, which are applicable to all sized modules of the GE-designed high density spent fuel storage rack. Different sized modules of the high density rack are provided for optimizing fuel storage for various pool sizes and configurations. All of '
the modules, however, have the same storage pitch and tube design of neutron absorbing material and, therefore, are bounded by the GESTAR k-infinity limit.
If there are any further questions, please call.
Very truly yours, f h h4Lb J. P. Nibert Fuel Project Engineer Hatch 1, 2 (408)925-5345 JPN:jn 86010PR00001 L _ ___ _______-- _ __ _ _ _ __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
p ,
h .$
GENER AL $ ELECTRIC NUCLEAR ENERGY BUSINESS OPERATIONS NUCLEAR TECHNOLOGIES & FUEL DIVISION 175 Curtner Avenue San Jose, CA 95125 M/C 174 March 31, 1986 cc: J. S. Charnley CJP:86-076 8. E. Hunt W. R. Mertz C. J. Paone G. D. Plotycia Mr. L. K. Mathews D. C. Serell Manager, Nuclear Core Analysis K. G. Turnage Southern Company Services P.O. Box 2625 Birmingham, AL 35202
SUBJECT:
Hatch 2 MAPLHGR Limits for Several Fuel Types
Dear Mr. Mathews:
At the request of Southern Company Services, GE has performed an analysis of the MAPLHGR limits for several additional fuel br:ndle types for Hatch 2. In view of the results, GE's current plan is to provide the following MAPLHGR Tech. Spec. limits in the supplemental reload licensing report for Hatch 2 Cycle 7:
- 1) Add results for the P80RB283 80 mil bundle to supplement the 100 mil results.
- 2) Revise the P8DRB284H MAPLHGR's to bound both 80 and 100 mil results.
- 3) Add clarifying notes to specify that the P80RB265H and P80RB299 results bound 80 and 100 mil results.
Please refer to the attached tables of MAPLHGR limits for these fuel types.
Additionally, MAPLHGR's and PCT results developed for Hatch 2 are applicable to Hatch I since the Hatch 2 LOCA analysis is slightly more conservative than that for Hatch 1.
If there are any questions, please call.
Very truly yours, f.kh4 J. P. Nibert Fuel Project Engineer Hatch 1, 2 (408) 925-5345 JPN:jn Attachment
. Gensral Electric Company Nuclear Energy Business Operations Nuclear Technologies and Fuel Division MAPLHGR TABLE FOR BUNDLE TYPE: P8DRB299*
EXPOSURE MAPLHGR PCT LOCAL OXIDATION (GWD/ST) (GWD/MT) (KW/FT) (DEG-F) (FRACTION) _.
0.20 0.22 10.90 2072. 0.023 1.0 1.1 11.00 2074. 0.023 5.0 5.5 11.50 2119. 0.026
- 10. 11. 12.10 2199. 0.033
- 15. 17. 12.10 2198. 0.033
- 20. 22. 12.00 2197. 0.033
- 25. 28. 11.50 2152. 0.029
- 30. 33. 11.00 2056. 0.021
- 35. 39. 10.30 1972. 0.030
- 40. 44. 9.70 1855. 0.010
- 45. 49. 9.00 1781. 0.007
- 80 mil and 100 mil channel thickness l
(CONTINUED) i D
i
f.,
- Genzral Electric Company Nuclear Energy Business Operations Nuclear Technologies and Fuel Division MAPLHGR TABLE FOR BUNDLE TYPE: P80RB283*
EXPOSURE MAPLHGR PCT LOCAL OXIDATION (GWD/ST) (GWD/MT) (KW/FT) (DEG-F) (FRACTION) .
0.20 0.22 11.30 2133. 0.029 1.0 1.1- 11.40 2134. 0.028 5.0 5.5 11.90 2185. 0.033
- 10. 11. 12.10 2195. 0.033
- 15. 17. 12.10 2199. 0.033 ,
- 20. 22. 11.90 2184. 0.032
-25. 28. 11.30 2112. 0.025
- 30. 33. 11.10 2061. 0.021
- 35. 39. 10.50 1981. 0.030
- 40. 44. 9.80 1808. 0.017
- 45. 49. 9.20 1788. 0.008 100 mil channel thickness t
I 1
(CONTINUED) 3
Gen 2ral Electric Company Nuclear Energy Business Operations Nuclear Technologies and Fuel Division MAPLHGR TABLE FOR BUNDLE TYPE: P80RB265H*
EXPOSURE MAPLHGR PCT LOCAL DXIDATION (GWD/ST) (GWD/MT) (KW/FT) (DEG-F) (FRACTION) .
0.20 0.22 11.50 2148. 0.030 1.0 1.1 11.60 2157. 0.030 5.0 5.5 11.90 2184. 0.032
- 10. 11. 12.10 2198. 0.033
- 15. 17. 12.10 2200. 0.033 20, 22. 11.90 2188. 0.032
- 25. 28. 11.30 2113. 0.025
- 30. 33. 10.70 2027. 0.019
- 35. 39. 10.20 1939. 0.014
- 40. 44, 9.60 1840. 0.009
- 45. 49. 8.90 1756. 0.007
- 80 mil and 100 mil channel thickness (CONTINUED) l l
l l
f5 . .
General Electric Company Nuclear Energy Business Operations Nuclear Technologies and Fuel Division MAPLHGR TABLE FOR BUNDLE TYPE: P8DRB284H*
EXPOSURE MAPLHGR PCT LOCAL OXIDATION (GWD/ST) (GWD/MT) (KW/FT) (DEG-F) (FRACTION) .
0.20 0.22 11.20 2111. 0.026 1.0 1.1 11.20 2106. 0.026 5.0 5.5 11.70 2152. 0.029
- 10. 11. 12.00 2189. 0.032
- 15. 17, 12.00 2195. 0.033
- 20. 22. 11.70 2181. 0.032
- 25. 28. 11.00 2089. 0.024
- 30. 33. 10.30 1987. 0.016 35, 39. 9.70 1887. 0.011
- 40. 44. 9.10 1797. 0.008
- 45. 49. 8.40 1716. 0.006
- 80 mil and 100 mil channel ~ thickness-t (CONTINUED) i i
7
. . . , -,rr- , _ _ . . , , , . ,
c- - -
F 1 '
General Electric Company Nuclear Energy Business Operations Nuclear Technologies and Fuel Division MAPLHGR TABLE FOR BUNDLE TYPE: P80RB283*
EXPOSURE MAPLHGR PCT LOCAL OXIDATION (GWD/ST) (GWD/MT) (KW/FT) (DEG-F) (FRACTION) .
0.20 0.22' 11.20 2116. 0.027 1.0 1.1 11.20 2113. 0.026 5.0 5.5 11.80 2171. 0.031
~10. 11. 12.00 2186. 0.032
- 15. 17. 12.10 2199. 0.033
- 20. 22. 11.90 2190. 0.032
- 25. 28. 11.40 2124. 0.026
- 30. 33. 10.80 2038. 0.019
- 35. 39. 10.30 1942. 0.014
- 40. 44. 9.60 1845. 0.010 45, 49, 8.90 1763. 0.007
- 80 mil channel thickness (END) l
+
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