ML20195H770
| ML20195H770 | |
| Person / Time | |
|---|---|
| Site: | Summer  |
| Issue date: | 06/20/1988 |
| From: | SOUTH CAROLINA ELECTRIC & GAS CO. |
| To: | |
| Shared Package | |
| ML20195H766 | List: |
| References | |
| NUDOCS 8806280377 | |
| Download: ML20195H770 (16) | |
Text
_ _ _ _ _ _ - _ _
ENCLOSURE 1 LICENSING SUBMITTAL CHANGE PAGES l
This enclosure contains change pages for the VCSNS licensing submittal of May 1
20, 1988 which modify the text and Technical Specifications markups to be appropriate for a design Tave of 587.4 F and a thennal design flow of 283500 gpn. All changes are highlighted by dual lines in the right margin. 1he following pages of the May 20th licensing submittal are affected.
ATTACHMENT ITEM 1
Pages 5, 22, 25, 26, 31, 41, and 46 2
Figure 2.1-1, Safety Limits Table 2.2-1, Minimtzn Measured Flow Table 2.2-1, (Notes 1 & 3),
Ove rpowe r-d elt a-T/Ove rt empe ra ture-d el t a-T Trips Figure 3 2-2, RCS Flow Limits Figure 3.2-1, Tave Limit 6
Table 1 DO R
i P
25
2.0
SUMMARY
AND CONCLUSIONS Consistent with the Westinghouse standard reload methodology for analyzing cycle specific reloads, Reference 5, parameters were selected to conservatively bound th6 values for each subsequent reload cycle and to facilitate determination of the applicability of 10CFR50.59.
The objective of subsequent cycle specific reload safety evaluations will be to verify that applicable safety limits are satisfied based on the reference evaluation / analyses established in this report.
The mechanical, thermal and hydraulic, nuclear, and accident evaluations considered the transition core effects described for a VANTAGE 5 mixed core in Reference 1.
The summary of these evaluations for the V. C. Summer core transitions to an all VANTAGE 5 core are given in the following sections of this submittal.
The transition design and safety evaluations consider the following conditions: 2775 MHt core thermal power, 555'F core inlet temperature, 2250 psia system pressure and 263 5'co gpm RCS thermal design flow.
These j
conditions are used in core design and safety evaluations to justify safe operation with the conservative assumptions noted in Section 1.0.
The conditions summarized in the SER for the VANTAGE 5 reference core report, HCAP-10444, have been considered in the V. C. Summer plant-specific safety evaluations.
The results of evaluation / analysis described herein lead to the following conclusions:
1.
The Westinghouse VANTAGE 5 reload fuel assemblies for the V. C. Summer Nuclear Plant are mechanically compatible with the current LOPAR fuel assemblies, control rods, secondary source rods and reactor internals interfaces.
The VANTAGE 5/LOPAR fuel assemblies satisfy the current design bases for the V. C. Summer reactor.
2.
Evaluations / analyses have shown that all or any combination of thimble plugs may be removed from the Cycle 5 core and subsequent reload cores.
i 00681:6/880504 5
the safety analyses and the design DNBR values is broken down as follows. A fraction of the margin is utilized to accommodate the transition core penalty (12.5% for VANTAGE 5 fuel and none for LOPAR fuel) and the appropriate fuel rod bow DNBR penalty, Reference 10, which is less than 1.3%.
The existing 6.3% margin in the LOPAR fuel and 17.5% margin in the VANTAGE 5 fuel between the design and safety analysis DNBR limits also includes a greater thsn 4/5 DNBR margin in the LOPAR fuel and a greater than 2.7% DNBR margin in the VANTAGE 5 fuel reserved for flexibility in the design.
The LOPAR and VANTAGE 5 designs have been shown to be hydraulically compatible in Reference 1.
The major impact of thimble plug removal on the thermal-hydraulic analysis is the increase in bypass flow which is reflected in Table 5.1.
The phenomena of fuel rod bowing, as described in Reference 10, must be accounted for in the DNBR safety analysis of Condition I and Condition II l
events for each plant application.
Internal to the fuel rod, the IFBA and fuel pellet designs are not expected to increase the propensity for fuel rods to bow.
External to the VANTAGE 5 fuel rod, the Inconel non-mixing vane aid Zircaloy mixing vane grids provide fuel rod support.
Additional restraint is provided with the Intermediate Flow Mixer (IFM) grids. Applicable generic credits for margin resulting from retained conservatism in the evaluation of DNBR are used to offset the effect of rod bow.
The safety analysis for the V.
C. Summer Plants maintain sufficient margin between the safety analysis limit DNBRs and the design limit DNBRs to accommodate full-flow and low-flow DNBR penalties.
The Westinghouse transition core DNB methodology is given in References 2 and 17 and has been approved by the NRC via Reference 18.
Using this methodology, transition cores are analyzed as if they were full cores of one assembly type (full LOPAR or full VANTAGE 5), applying the applicable transition core penalties.
This penalty is included in the safety analysis limit DNBRs such that sufficient margin ovar the desig7 limit DNBR exists to accommodate the transition core penal +y and the appropriate rod bow DNBR penalty.
00681:6/880504 22
-.------_--_--J
TABLE 5.1 (Continued)
V. C. SUMMER THERHAL AND HYDRAULIC DESIGN PARAMETERS Design HFP Nominal Coolant Conditions Parameters Vessel Minimum Heasured Flow +
0 Rate (including Bypass), 10 lbm/hr
/88 /
GPH 261,500 Vessel Thermal Design Flow +
6 Rate (including Bypass), 10 lbm/hr
/05.9 GPH 283,500 Core Flow Rate (excluding Bypass, based on TOF) 6 10 lbm/hr 9 (,.47 GPM 258;270 i
Fuel Assembly Flow Area +'
2 for Heat Transfer, ft (LOPAR) 41.55 (V-5) 44.04 Core Inlet Hass Velocity, 6
10 lbm/hr-ft (Based on TOF)
(LOPAR) 2.32,
.(V-5) 2./9 Includes 157. steam generator tube plugging
{
+
Assumes all LOPAR or VANTAGE 5 core
++
00681:6/880504 25 j
i
TABLE 5.1 (Continued)
V. C. SUMMER THERHAL AND HYORAULIC DESIGN PARAMETERS Design Thermal and Hydraulic Desian Parameters Parameters (Based on Thermal Design Flow)
Nominal Vessel / Core Inlet Temperature, 'F JI55.o Vessel Average Temperature, 'F 5874 Core Average Temperature. *F E92.3 Vessel Outlet Temperature, 'F 6/ 9. g Average Temperature Rise in Vessel, 'F 44.8 Aserage Temperature Rise in Core, 'F 70.4 Heat Transfer Active Heat Transfer Surface Area,++
(LOPAR) 48,598 2
ft (V-5) 46,779 2
Average Heat Flux, BTU /hr-ft (LOPAR) 189,820 (V-5) 197,200 Average Linear Power, kw/ft 5.45
+++
Peak Linear Power for Normal Operation, kw/ft 13.30 1
Assumes all LOPAR or VANTAGE 5 core
++
i
+++ Based on 2.45 F peaking factor g
~
00681:6/880504 26
Reactor Coolant System Flow Reduction All non-LOCA safety analyses reanalyzed for this report have incorporated a reduction in the reactor coolant system flow.
The reduced flow corresponds to a thermal design flow of 2 83500 y m and a minimum measured flow of 269500 ym.
Thimble Pluo Deletion The non-LOCA analyses performed incorporated the impact of thimble plug deletion. Thimble pivg deletion affects core pressure drops and bypass flow.
These effects have been conservatively incorporated into the non-LOCA safety analyses.
Debris Filter Bottom Nozzle The VANTAGE 5 fue! design will also include the Debris Filter Bottom Nozzle (DFBN).
In the DFBN, the relatively large flow holes in the conventional bottom nozzle are replaced with a new pattern of smaller flow holes.
These holes are sized to minimize the passage of debris particles large enough to cause damage while st!'1 providing sufficient flow area, comparable pressure drop, and continued structural integrity of the nozzle.
As such, no parameters important to the non-LOCA safety analyses are impacted.
Increased Overoower/Overtemoerature AT Reactor Trio Response Time The total time delay of the overtemperature AT and overpower AT trips (including RTD time response, trip circuitry and channel electronics delay) assumed in the non-LOCA analyses is 8.5 seconds.
The 8.5 second delay includes a 7 second first order lag incorporated into the determination of the time at which the overtemperature AT and overpower AT trip setpoints are reached.
The remaining 1.5 seconds is the delay from the time at which the trip signal is initiated until the rod cluster control assemb11Es are free to drop into the core.
i 1406v:lD/880517 31
7.0
SUMMARY
OF TECHNICAL SPECIFICATION CHANGES The preposed changes to the Virgil C. Summer Nuclear Station (VCSNS) Technical Specifications are summarized in Table 7.1.
These changes reflect the impact of the design, analytical methodology, and safety analysis assumptions outlined in the SCE&G amendment request and are given in the proposed Technical Specification page changes (see Attachment 2 of this report).
A brief overview of the significant changes follows.
7.1 Core Safety Limits Core safety limits and associated bases for 3-loop operation during modes 1 and 2 are revised to reflect the impact of the transition to VANTAGE 5 with:
1.
The use of ITDP and the WRB-1 and WRB-2 DNB Correlation.
2.
An FaH of 1.62 (see oection 7.11).
3.
Reduced RCS flow to accommodate the increased resistance of the VANTAGE 5 fuel assembly and to support SG tube plugging up to 15% (see Section 7.2).
The proposed limits corresponds to those for the LOPAR fuel which are limiting during tne transition period. Less limiting values will be possible with a full core of VANTAGE 5.
7.2 Thermal Design Flow The VCSNS thernial design flow is being decreased from 258600 ypm to 283 roo 3 pen.
This flow reduction accommodates:
The increased resistance of the VANTAGE 5 fuel assembly, a.
b.
Up to 15% SG tube plugging in all three SG's.
l 14c5olo/c51682 41
.n
--v-
~ - - - -
~
7.11 Nuclear Enthaley Rise Hot Channel Factor
~
The following f values (includes uncertainties) are proposed for the aH VANTAGE 5 transition.
F3g = 1.56 [1 + 0.3 (1-P))
where P is the fraction of full power. These higher values allow increased fuel cycle design flexibility and lower leakage core loading patterns.
7.12 DNB Parameters The proposed limits on DNS related parameters (T,ygand Pressurizer Pressure) assure that each are maintained within the normal steady state envelope of operation assumed in the transient and accident analyses.
The proposed revisions are consistent with new accident analyses supplied in the Transient Safety Evaluation which utili4s the ITOP (see Section 5.0) for DNB evaluations.
The T,yg reflects the nominal baseline T,y of 587.Y'F assumed in the VANTAGE 5 analysis in order to support fulk power operation with:
1.
15% uniform SG tube plugging.
s.
.to in,
2.
A thermal design flow conservatively caLWed to support 15% SG tube plugging and the use of VANTAGE 5 fuel.
t 1405v:1o/c51688 46
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{
10Ve2 muet De considered 2005 MTP for this figure.
FJ9ure E.2-2 Reector Core setety L$ msg - Three 20000 in Operation Sunner - Unit i 9-E
E TABLE 2.2-1 49 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS c-Total
}
Functional Unit Allowance (TA)
Z_
S Trip Setpoint Allowable Value 1.
Manual Reactor Trip Not Applicable NA NA NA NA 2.
Power Range, Neutron Flux 7.5 4.56 0
1109% of RTP 1111.2% of RTP High Setpoint low Setpoint 8.3 4.56 0
Power Range, Neutron F1ux
- 1. 6 0.5 0
15% of RTP with 16.3% of RTP *ith High Positive Rate a time constant a time constant 12 seconds 12 seconds 4.
Power Range, Neutron Flux 1.6 0.5 0
15% of RTP with 16.3% of RTP with High Negative Rate a time constant a time constant m
12 seconds 12 seconds 5.
Intermediate Range.
17.0 8.4 0
125% of RTP 131% of RTP Neutron Flux 6.
Source Range, Neutron Flux 17.0 10.0 0
<105
<1.4 x 105 cps cps 9.8 7.29 f.9h1.N*4 7.
Overtemperature AT Jd-W M
See note 1 See not.e 2 S.z 2.2 s 1.9 8.
Overpower AT Adr-X X
See note 3 See note 4 9.
Pressurizer Pressure-Low 3.1 0.71
- 1. 5 11870 psig 11859 psig 10.
Pressurizer Pressure-High 3.1 0.71 1.5 12380 psig
$2391 psig 11.
Pressurizer Water Level-High 5.0 2.18
- 1. 5 192% of instrument
$93.8% of instrument span span 12.
Loss of Flow 2.5 1.0 1.5 190% of loop 189.2% of loop design flow
- design flow
- O Loop design flow =.96:40tr gpm RTP = RATED THERMAL POWER l
C4 IM 7. 5Peo Foe 'DELT4-T (RTDs) Auo I.27. FoA PRE 55me:EE R PeEssu.stE h
w TABLE 2.2-1 (Continued) h NACTOR TRIP $YSTEM INSTRUMENTATION TRIP SETPOINTS m
MOTATION HOTE 1: OVERTEMPERATURE AT
.-4 AT i AT, (R - Re g
[T.- T'] + R (P - P') - f (at)]
3 ifhere:
AT
= Measured AT by RTO Montfold Instrumentatten AT, 6 X Indicated AT at RATED THERMAL POWER R,
L v peen 1.*Lo3 Rs 2 y y 0. 0300lo tj 1
The function generated by the lead-lag controller for T y
=
g dynamic compensatten
==
tg. A ta = Tfse constants utilfred in the lead-1de controller for Tg. t K 28 secs..
t {4 secs.
= g temperature *F T
at RATED THERNRL POWER T'
3 g F Reference Tg Rs.
?.s M <0.ocl+9 l
Pressuriter pressure, psis P
=
P' 1 x 2235 psIg, HasInel RCs operating pressure Laplace transfers operator, sec 8
=
l
~
m i
TAfttE 2.2-1 (Continised)
=
h REACTOR 1 RIP SYSTIN INSTRtlNENTATION TRIP SETPOINTS I
O NOTATION (Continued) e 1
5 NOTE 1: (Continued) j
.?
l and f (al) is a function of the Indicated difference between top and bottom detectors of the powar-rango nuclear ton chambers; with gains to be selected based on seasured instrument response during plant startup tests such that:
14 (1) for E!g g between - K percent and +
percent f (at) = 0 where g and g are percent g
]
RATED THERMAL POWER In the top and bottom halves of the core respectively, and gg + g is j
total THERMRt POWER in percent of RATED TMERMRt POWER.
-2t j
(11) for each percent that the magnitude of gt ~ 'b exceeds -p percent, the AT trip setpoint -
l shall be automatically reduced by dpercent of its value at RATED TIERMRL POWER.
2.7.7
+
I ow (Ill) for each percent that the segnitude of qt -A exceeds *A percent, the AT trip setpoint j
b shall be automatically reduced by Adi percent of Its value at RATED 11ERNAL POWER.
i W 2.13 I
NOTE 2:
The channel's samisme trip setpoint shall not exceed its computed trip point by more than
,0 MP'"t ai span.
l NOTE 3: DVERPOWER af AT < af, IR4 - Rs I;]S
} I ~ "* [T - 1*] ]
5 as defined in Note 1 Where:
of
=
1
= as defined in Note 1 j
AT,
}
Ka SK W }.082[
i I
Rs 1F 0.02/*F for increasing average temperature and 0 for decreasing average temperature
= The function generated by the rate-lag controller for T,,,dynanic compensation l
5 3
I i
TABLE 2.2-1 (Continued)
REACTOR TRIP SYST[M INSTRtM NTAi!ON TRIP SETPolNTS
=
NOTATION (Continued) c5 NOTE 3:
(Continued)
- g. t )( 10 secs.
Time con tant utilfred in the rate-lag controller for T
=
3 ts 0 0015 K.
2 jf 0.pBHTJ/*F for T > 1" and K. = 0 for T i T" T
=
as,,definedinNote1 g*J Reference T,,, at RATED THERMAL F0WER la as deffned in Note 1 S
=
i NOTE 4:
The channel's maximum trip setpoint shall not exceed its computed trip point by more than J # pe tent AT span.
1.0 -
l e
l l
1 u.
s,.
i
MEASUREMENT UNCERTAINTIES OF 2.1% FOR FLOW AND 4.0% FOR INCORE MEASUREMENT OF FNAH ARE INCLUDED IN THIS FIGURE 38 ACCEPTABl.E UMACCEPTA ILE OPERAT104.1EGION OPERATION REGION 36 34 T
5 32 C
W 6
so d
30 a
If
(
(1.00,28.95)
-9 (1.00,28.66) 2 00l2b8 SEE NOTE 28 m ni, (1.00,27.79)
(1.00,27.50) a 26 24
.9 95 1
1.05 1.1 R = FNAH/1.56 [1.0 + 0,3(1.0 P)]
l flGURE 3.2 2 RCS TOTAL FLOW RATE VS. R THREE LOOP OPERATION N0Tt: When operationg la this rege, the restrkud power levels s>All be tosidered to te 100% of rsted thermal power (it (P)for Figure 2.1 '
SUMMErt. UNIT 1 3/4210
TABLE 3.2-1 E3 DN8 PARAMETERS 9
[
LIMITS z
U 3 Loops In 2 Loops in PARAMETER Operation Operation D
Reactor Coolant System T 5 592"F avg Pressurizer Pressure
> N*
2204 psisk z
a
'?
M 4
"Limit not applicable during either a THERMAL POWER ramp in excess of 5% of RATED THERMAL POWER per minute or a THERMAL POWER step in excess of 10% of RATED THERMAL POWER.
i
- These values left blank pending NRC approval of two-loop operation.
1 l
1
1 Table 1 VCSNS ANALYSIS BASELINE l
l Parameter Current Value Va 5 rni on NSS Power, HWt 2785 2787 Core Power, HWt 2775 2775 System Pressure, psia 2250 2250 Thermal Design Flow, gpm 286600 26350o ll Core Bypass Flow, %
6.4 8.9'*
TAVE, 'F 587.4 587.4 ll
~
THOT, 'F 618.7 6/9.6 ll F3H 1.55 1.62 F2H Multipiar 0.2 0.3 LOCA F0 2.25 2.45 SG Tube Plugging, %
16 15 AFD Control CAOC RAOC Peaking Surveillance Fxy (z)
FQ (z)
High Head Safety Injection Recirculation Recirculation Isoleted Hot Isolated Thimble Plugs Yes Optional
- Non-ITOP l
[
-- - -