ML17213B025
| ML17213B025 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| Issue date: | 01/24/1983 |
| From: | FLORIDA POWER & LIGHT CO. |
| To: | |
| Shared Package | |
| ML17213B024 | List: |
| References | |
| NUDOCS 8302010582 | |
| Download: ML17213B025 (8) | |
Text
3/4.1 REACT IVITY CONTROL SYSTENS 3/4.1.1 BORAT ION CONTROL SHUTDOWN NARGIN T
> 200 F
LINITING CONDITION FOR OPERATION 3.1.1.1 The SHUTDOWN MARGIN shall be
> 3.6g ak/k.
APPLICABILITY:
NODES 1,2*,
3 and 4 ACTION:
'With the SHUTDOWN HARGIN ( 3.6$ hk/k, immediately initiate and continue boration at
> 40 gpm of 1720 ppm boron or equivalent until the r'equired SHUTDOWN NAR7ÃIN is restored.
SURVEILLANCF. REQUIRBlENTS 4.1.1.1.1 The SHUTDOWN NARGIN shall be determined to be
> 3.6X, hk/k:
a.
Within one hour after detection of an inoperable CEA(s) and at lease once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter while the CEA(s) is
- .- inoperable.
If the inoperable CEA is immovable or untrippable, the above required SHUTDOWN MARGIN shal 1
oe increased by an amount at least equal,to the withdrawn worth of the immovable or untrippable CEA(s).
b.
lichen in NODES 1 or 2~, at least once -per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by veri fying that CEA group withdrawal is within the Power Dependent Insertion Limits of Specification 3.1.3.6.
C ~
d.
When in NODE 2<~, at least once during CEA withdrawal and at least once per hour thereafter until the reactor is critical.
Prior to initial operation above 55 RATED THERNAL POWER after each fuel loading, by consideration of the factors of e.
below, with the CEA groups at the Power Dependent, Insertion Limits of Specification
- 3. 1.3.6.
- See Special Test Exception 3.10.1
~
With Keff > 1.0.
AP'With K ff g 1,0.
ST.
LUCIE - UNIT 1 3/4 1-1 83020i0582 830i20 PDR ADOCK 05000335 P
0
~
0 REACTIVITY CONTROL SYSTEMS MODERATOR TEMPERATURE COEFFICIENT LIMITING CONDITION FOR OPERATION 4
3.1.1.4 The moderator temperature coefficient (MTC) shall be:
a.
b.
C ~
Less positive than 0.7 x 10 " zk/k/'F whenever THERMAL POWER is
< 70$ of RATED THERMAL POWER, Less positive than 0.2 x 10 " &/k/'F whenever THERMAL POWER IS
> 7(C of RATED THERMAL POWER, and Less negative than -2.8 x 10 4-bk/k/
F at RATED THERMAL POWER APPLICABILITY.:
MODES 1 and 2*g ACTION:
/
With the morley rator temperature coefficient outside any one of the above limits, be in HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
SURVEILLANCE REQUIREMENTS 4.1.1.4.1 The MTC shall be determined to be within its limits by confirmatory measurements.
MTC measured values shall be extrapolated and/or ccmpensated to permit direct canparison with the above limits.
With Keff > 1.0.
g See Special Test Exception 3.10.2.
ST.
LUCIE - UNIT 1 3/4 1-5
~>
kg
'E
3/4.1 BASES REACTIVITY CONTROL SYSTBPS 3/4. 1. 1 BORATION CO NTROL 3/4.1.1.1 and 3/4.1.1.2 SHUTDOMN MARGIN A sufficient SHUTDOWN MARGIN ensures that 1) the reactor can be made subcritical from all operating conditions,
- 2) the reactivity tran'sients associated with postulated accident conditions are controllable within acceptable limits, and 3) the reactor will be maintained sufficiently subcriti'cal to preclude inadvertent criticality in the shutdown condition.
SHUTDOWN MARGIN requirements vary throughout core life as a function of fuel depletion, RCS boron concentration, and RCS Ta The most restrictive-condition occurs at EOL, with Tav at no load opera$ fng temperature, and is associated with a postulated steA line break accident and resulting uncontrol led RCS cooldown.
In the analysis of this accident, a minimum SHUTDOWN tlhRGIN of 3.65 zk/k is required to control the reactivity transient.
Accordingly, the SHUTDOWN HARGIN required by Specification 3.1.1.1 is based upon this limiting condition and is consistent with FSAR accident analysis assumptions.
For earlier periods during the fuel cycle, this value is conservative.
With T v
< 200'F, the reactivity transient resulting frcm a boron dilution event wit3 3 partially drained Reactor Coolant System requires a 2/ Ak/k SHUTDOWN MARGIN and restrictions on charging pump operation to provide adequate protection.
A 2g hk/k SHUTDOWN MARGIN is 1.(g hk/k conservative for trode 5 operation with total RCS volume present, however; LCO 3.1.1.2 is written conservatively for simplicity.
3/4.1.1.3 BORON DILUTION AND ADDITION A mimimum flow rate of at least 3000 GPM provides adequate
- mixing, prevents stratification and ensures that reactivi.ty changes <<ill be gradual during boron concentration changes in the Reactor Coolant Systan.
A flow rate of at least 3000 GPH will circulate an equivalent Reactor Coolant System volume of 11,400 cubic feet in approximately 26 minutes.
The reactivity change rate associated with boron concentration changes <<ill be within the capability for operator recognition and control.
3/4. 1. 1.4
%DERATOR TEMPERATURE COEFFICIENT ETC)
The limiting values assumed for the MTC used in the accident and transient analyses were +0.7 x 10-4 hk/k/'F for THERi~iAL POWER levels
< 7(g of RATED TflERHAL POWER,
+ 0.2 x 10-4 ak/k/'F for THERMAL POWER levels
> 70$ of RATED THERMAL and -2.8 x 10-4 hk/k/'F at RATED THEINAL POWER.
Therefore, these l imiting values are included in this specification.
Determination of NTC at the spocified conditions ensures that the maximum positive and/or negative values of the ETC will not exceed the limiting values.
ST.
LUCIE - UNIT 1 B 3/4 1-1
Document Number XH-75-27 9 Supp.
1,2,3
- Supp. 4, XN-NF>>507
- Supplement XN-75-48 Xt)-74-5, Rev.
1 XH-209 XN-75-3 2, Supp.
182 XH-75-41 XH-76-27 XH-NF-78-30 XH-NF-81-2 2 XN-HF-81-58 XH-NF-82-06 XN-HF-82-20
- Supplement Attachment 2
Title Exxon Nuclear Neutronic Design for Pressurized Water Reactors EHC Setpoint methodology for C.E.
Reactors (Incorporates XH-NF-81-22)
Definition and Justification of Exxon Nuclear Company OHB Correlation for PWR's Description of the Exxon Nuclear Plant*
Transient Simulation Model fo r Pressurized Water Reactors (PTSPWR)
Densification Effects on Exxon Huclear Pressurized Water Reactor Fuel Computational Pr ocedures for Evaluating Fuel Rod Bowing Ex%.on Nuclear Company WREN-Based Gener ic P WR ECCS Eva luat i on Model Exxon Nuclear Company WREN-Based Generic PWR ECCS Evaluation Model Update ENC WRY-ll Exxon Nuclear Ccnpany WREN-Based Generic PWR ECCS Evaluation Model Update ENC WREN.-11A Generic Statistical Uncertainty Analysis Nethodology Fuel Rod Thermal-Mechanical
Response
Evaluation i4lodel (RODEX2) qualification of Exxon Nuclear Company Fuel for Extended Burnup Exxon Nuclear Company Evaluation ttodel EXEN/PWR ECCS tlodel Update Example Problem for C.E.
Reactors Submi tted to HRC
'1/83 11/82 6/75 5/81 8/81 2/82 2/82 8/82 Anticipated Appr oval Approved 3/83 Approved for Fort Calhoun 3/83 Approved for Previ ou sly Licensed PWR Approved for Previ ously Licensed PWR Approved 3/83 Approved Approved Approved 3/8 3 9/83 1/83 3/83
- Appr oval needed first for St; Lucie Unit 1; other reports applied to other reactors first.
Document Number Title Submitted to NRC Anticipated Ap roval
+XN-NF-82-09 Generic Mechanical Design for Exxon Nuclear 14x14 Rel oad Fuel for C.E.
Reactors 10/82 3/83 XN-NF-621 XH-NF-82-07 Exxon Nuclear DNB Correlation for PMR Fuel Designs Exxon Nuclear Company ECCS Cladding Swelling and Rupture Model 2/82 1/83 Approved
H
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