ML20058L581
| ML20058L581 | |
| Person / Time | |
|---|---|
| Site: | Clinton  |
| Issue date: | 08/02/1990 |
| From: | Hannon J Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20058L589 | List: |
| References | |
| NUDOCS 9008070309 | |
| Download: ML20058L581 (17) | |
Text
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UNITED STATES i
g.
NUCLEAR REGUL ATORY COMMISSION 4
O E 5*
' WASHINGTON, D. C. 20655
[
ILLIN0IS POWER COMPANY, ET AL..
DOCKET NO. 50-461 CLINTON POWER STATION, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.42 License No. NPF-62 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by Illinois Power Company * (IP) and Soyland Power Cooperative, Inc. (the licensees) dated February 5,1988 complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act). and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the.
Comission; L
C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering;theLhealth and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations;
['
D.
The issuance of this amendment will not be inimical to the comon
{
defense and security or to the health and safety of the public; and l
h E.
The issuance of-this amendment is in accordance with 10 CFR Part 51' of the Commission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifications f
asindicatedin'theattachmenttothislicenseamendment,andparagraph2.C.(2) 4 of Facility Operating License No. NPF-62 is hereby amended to read as follows:
- Illinois Power Company is authorized to act as agent for Soyland Power Cooperative, Inc. and has exclusive responsibility and control over the l-physical construction, operation and maintenance of the facility.
9008070309 900802 PDR ADOCK 05000461
'P PDC
4
' i (2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A'and the 1
Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 42, are hereby incorporated into this license.
Illinois Power Company shall operate the facility in accurdance with the Technical Specifications and the Environmental Protection Plan.
3.
This license amendment is effective as of its date of issuance, r
FOR THE NUCLEAR REGULATORY COMMISSION W-John N. Hannon, Director Project Directorate III-3 Division of Reactor Projects - III, IV, V and special Projects Office of~ Nuclear Reactor Regulation
Attachment:
Changes to the Technical.
Specifications
- Date of issuance: August 2, 1990 l
l l
l l
I
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9 ATTACHMENT -TO LICENSE AMENDMENT NO. 42 FACILITY OPERATING LICENSE NO. NPF-62 DOCKETNO.50-4R
-. Replace the following pages-of the Appendix "A" Technical Specificatiom with-
'the attached pages. The revised pages are_ identified by amendment number and N
contain vertical-lines-indicating.the area of change. Corresponding overleaf ptges are provided to maintain document completeness.
Remove Insert 3/4 3-21
- 3/4 3-21 3/4.3-22 3/4 3-22 3/4 3-23 3/4 3-23 3/4 3-24 3/4 3 3/4 4-2 3/4 4-2
'3/4 4-3 3/4 4-3
- 3/4~4-4 3/4 4-4 3/4 4-6 3/4 4-6 3/4 4-19 3/4 4-19 B 3/4 4-11 B 3/4 4-1
TABLE 3.3.2-2 (Continued) n 5
CRVICS INSTRUMENTATION ~SETPOINTS E
TRIP FUNCTION TRIP SETPOINT
-ALLOWABLE VALUE Q
1.
PRIMARY AND SECONDARY CONTAINMENT ISOLATION (Continued) k.
~ Containment-Pressure High 1 2.62 psid 5 3.00 psid 1.
Main Steam Line Radiation -' High
.s'3.0 x' full power background
$ 3.6 x full power background m.
Fuel Building Exhaust Radiation -
High i 10 mR/hr i 17 mR/hr l
n.
Manual Initiation MA NA f
2.
MAIN STEAM LINE ISOLATION I
a.
R Low Low Low, Level 1
> -145.5 in.*
> -147.7 in.
b.
Main Steam Line Radiation - High 1 3.0 x full power background
$3.6xfullpowerbackgroun/
O c.
Main Steam Line Pressure - Low
> 849 psig
> 837 psig d.
Main Steam Line Flow-- High 5 170 psid i 178 psid e.
Condenser Vacuum Low
> 8.5 in. Hg vacuum
> 7.6 in. Hg vacuum f.
Main Steam Line Tunnel Temp. - High 5 165'F 5 176*F g.
Main Steam Line Tunnel' a Temp. - High 5 4.5*F 1 60*F 5
h.
Main Steam Line Turbine Bldg.
Temp. - High' y
(1) 1E31 - N559 A, B, C, D 1 131.2*F
$ 138*F g
IE31 - N560 A, B, C, D 1E31 - N561 A, B, C, D 1E31 - N562 A, B, C, D
[
(2) 1E31 - N563 A, B, C, D 1 143.2*F i 150*F
[
1.
Manual Initiation
.NA NA i
P 3.
REACTOR WATER CLEANUP SYSTEM ISOLATION a.
A Flow - High
~$ 59 gpa 5 66.1 pu b.
A Flow Timer
> 45 sec.
4 1 7 sec.
4
,2.
TABLE 3.3.2-2 (Continued) n-CRVICS INSTRUMENTATTON SETPOINTS g TRIP FUNCTION TRIP SETPOINT ALLOWABLE VALUE.
Z 3.
REACTOR WATER CLEANUP SYSTEM ISOLATION (Continued) i c.
Equipment Area Temp. - High 1.
Pump Rooms - A, B, C 2.
Heat Exchanger Rooms - East,
-< 186.5'F
< 197.1'F' West 5 201'F
$ 212*F-d.
Equipment Area A Temp. - High 1.
Pump Rooms - A,-B, C i
2.
Heat-Exchanger Rooms - East,
-< 54.5'F
< 60*F West
_ 54.5'F
_ 60*F e.
T Low Low, Level 2
> -45.5 in.*
> -47.7 in.
f.
Main Steam Line Tunnel Ambient Temp.~- High 5 165'F
$ 176*F g.
Main Steam Line Tunnel A Temp. - High
< 54.5'F 1 60*F h.
SLCS Initiation NA NA
.i i.
Manual Initiation NA NA 4.
REACTOR CORE ISOLATION COOLING SYSTEM ISOLATION Fg a.
RCIC Steam Line Flow - Hign
- 110 in. H O
< 118.5 ia H O l
2 g
2 g
b.
RCIC Steam Line Flow - High Timer
> 3 sec.-
1 13 sec.
,y c.
RCIC Steam Supply Pressure - Low
> 60 psig
> 52 psig O
d.
RCIC Turbine Exhaust Diaphragm Pressure - High
< 10 psig
< 20 psig a
-a y
m,
- - = -
w
t TABLE 3.3.2-2 (Continued) n.
E
'i CRVICS INSTRUMENTATION SETPOINTS e
TRIP FUNCTION TRIP SETPOINT ALLOWABLE VALUE c-5
[
4.
REACTOR CORE ISOLATION COOLING SYSTEM ISOLATION (Continued)
?
e.
RCIC Equipment Room Ambient Temp. - High 5 222.5'F
$ 233.1*F f.
RCIC Equipment Room a Temp. - High i 34.5*F 1 40*F' g.
Main Steam Line Tunnel-Ambient Temp. - High 5 165'F i 176'F h.
Main Steam Line Tunnel ws A Temp. - High
.$ 54.5'F
$ 60'F i.
Main Steam Line Tunnel Temp. Timer
> 25 min.
1 28 min.
j.
Drywell Pressure - High 1 1.68 psig i 1.88 psig k.
Manual-Initiation NA NA 1.
RHR/RCIC Steam Line Flow - High
$ 179.5 in. H O i 188 in. H O l
2 2
m.
RHR Heat Exchanger A, B Ambient Temperature - High 5 138.5'F
$ 149.6'F k
n.
RHR Heat Excha r A, B g
A Temp. - Hig 1 74.2*F
$ 79.6*F k
5.
RHR SYSTEM ISOLATION i
a.
RHR Heat Exchanger Rooms A, B Ambient Temperature - High 5 138.5'F i 149.6*F' b.
RHR Heat Exchanger Rooms A, B A Temperature - High 5 74.2*F
$ 79.6'F
, _a.
TABLE 3.3.2-2 (Continued) nC5 CRVICS INSTRUNENTATION SETPOINTS E
d TRIP FUNCTION TRIP SETPOINT-ALLOWBLE VALUE 5
[
5.
RHR.'fSTEM ISGOTION (Continued) c.
2 3 in.
8 1
9 in.*
8 Low, level 3 d.
Low Low Low, Level 1
> -145.5 in.*
1 -147.7 in.
e.
Reactor Vessel (RHR Cut-in Permissive) Pressure - High 5 135 psig i 150 psig l
5 f.
Drywell Pressure - High 4
< 1.68 psig
< 1.88 psig w
- 2) Fuel Pool Cooling
{1.68psig
{1.88psig g.
Manual Initiation NA NA
- See Bases Figure B 3/4 3-1.
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F.
E e
N O
3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REtlRCULATION SYSTEM RECIRCULATION LOOPS
_ LIMITING _ CONDITION FOR OPERATION 3.4.1.1 Two reactor coolant system recirculation loops shall be in operation with:
Total core flow greater than or equal to 45% of rated core flow, or a.
b.
THERMAL POWER within the unrestricted zone of Figure 3.4.1.1-1, or THERMAL POWER within the restricted zonet of Figure 3.4.1.1-1 and APRM c.
or LPRMit noise levels not larger than three times their established baseline noise levels.
APPLICsB]LITY: OPERAT10NAL CONDITIONS la and 2*.
ACTION:
kith one reactor coolant system recirculation loop not in operation:
a.
1.
Within 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />s:
a)
Place the recirculation flow contrni system in the Local Manual (Position Control) mode, and b)
Reduce THERMAL POWER TO $ 70% of RATED THERMAL POWER, and c)
Increase the MINIMUM CRITICAL POWER RATIO (MC.PR) Safety Limit by 0.01 to 1.08 per Specification 2.1.2, and d)
Reduce the Maximum Average Planar Linear Heat Generation Rste (MAPLHGR) limit per Specification 3.2.1 and the CORE OPEr.ATING LIMITS REPORT, and e)
Reduce the Average Power Range Monitor (APRM) Screm and Rod Block Trip Setpoints and Allowable Values to those applicable p
for single-recirculation loop operation per Specifications 2.2.1 and 3.3.6, and i
- See Special Test Exception 3.10.4.
I tThe operating region for which monitoring is required.
See Surveillance Requirement 4.4.1.1.2.
itDetector levels A and C of one LPRM string per core octant plus detectors A and C of one LPRM string in the center of the core should be monitored.
CLINTON - UNIT 1 3/4 4-1 Amendment No. 28
i 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 RECIRCULATION SYSTEM RECIRCULATION LOOPS i
LIMITING CONDIT hN FOR OPERATION f)
Reduce the volumetric flow rate of the operating recirculation loop to 5 31,341 gpm*
l g)
Perform Surveillance Requirement 4.4.1.1.4 if thermal power is
< 30%** of RATED THERMAL POWER or the recirculation loop flow Tn the operating loop is 1 30%** of rated loop flow.
l 2.
The provisions of Specification 3.0.4 are not applicable.
3.
Otherwise, place the unit in HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
b.
With no reactor coolant system recirculation loops in operation, immediately initiate action to reduce THERMAL POWER so that it is in the unrestricted zone of Figure 3.4.1.1-1 within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and initiate measures to place the t
unit in at least STARTUP within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, With one or two reactor coolant system recirculation loops in operation and c.
total core ilow less than 45% but greater than 35.5d of rated core flow l
and THERMAL POWER within the restricted zone of Figure 3.4.1.1-1, and with the APRM or LPRMt neutron ix noise levels greater than three times their established baseline no", avels, immediately initiate corrective action to restore the noise levM s to within the required limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> by increasing core flow or by reducing THERMAL POWER.
d.
With one or two reactor coolant recirculation loops in operation, and total core flow less than or equal to 35.5/ and THERMAL POWER within the l
restricted zone of Figure 3.4.1.1-1, within 15 minutes initiate corrective action to reduce THERMAL POWER to within the unrestricted zone of Figure 3.4.1.1-1, or increase core flow to greater than 35.5d within l
4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
- This value represents the measured volumetric recirculation loop flow which produces 100% core flow at 100% THERMAL POWER.
- The threshold THERMAL POWER and recirculation loop flow which will sweep the cold water from the vessel bottom head preventing stratification.
- Care flow with both recirculation pumps at rated speed and minimum control valve position.
tDetector levels A and C of one LPRM string per core octant plus detectors A and C of one LPRM string in the center of the core should be monitored.
CLINTON - UNIT 1 3/4 4-2 Amendment No. S 42
i REACTOR COOLANT SYSTEM RECIRCULATION LOOPS SURVEILLANCE REQUIREMENTS 1
4.4.1.1.1 Each reactor coolant system recirculation loop flow control valve shall be demonstrated OPERABLE at least once per 18 months by:
Verifying that the control valve fails "as is" on loss of hydraulic a.
pressure at the hydraulic control unit, and b.
Verifying that the average rate of control valve movement is:
1.
Lsss than or equal to 11% of stroke per second opening, and 2.
Less than or equal to 11% of stroke per s m nd closing, t
4.4.1.1.2 When THERMAL POWER is within the restricted zone of Figure 3.4.1,1-1, and one or two pumps are in operation, establish a baseline APRM and LPRM*
neutron flux noise value within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of entering this operating region unless baselining has previously been performed in the region since the last CORE ALTERATION, and Determine the APRM ano LPRM* noise levels at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, and a.
b.
Determine the APRM and LPRM* noise levels within 30 minutes after the completion of a THERMAL POWER increase of at least 5% of RATED THERMAL POWER.
t 4.4.1.1.3 With one reactor system recirculation loop not in operation, at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> verify that:
a.
Reactor THERMAL POWER is 5 70% of RATED THERMAL POWER.
b.
The recirculation flow contrcl system is in the Local Manual (Position Control) mode, c.
The volumetric flow rate of the operating loop is 5 31,341 gpm**, and l
d.
Core flow is greater than 35.5% when THERMAL POWER is within the l
restricted zone of Figure 3.4.1.1-1.
4.4.1.1.4 With one reactor coolant system recirculation loop not in operation, within no more than 15 minutes prior to either THERMAL POWER increase or recir-culation loop flow increase, verify that the following differential temperature
- Detector levels A and C of one LPRM string per core octant plus detectors A and C of one LPRM string in the center of the core should be monitored.
- This value represents the measured volumetric recirculation loop flow which l
produces 100% core flow at 100% THERMAL POWER.
- Core flow with both recirculation pumps at rated speed and minimum control valve position.
CLINTON - UNIT I 3/4 4-3 Amendment No. 22. 42
+-
REACTOR COOLANT SYSTEM RECIRCULATION LOOPS SURVEILLANCE REQUIREMENTS requirements are met if THERMAL POWER is < 30%* of RATED THERMAL POWER cr the cecirculition loep flow in the operating Toop is 5 30%* of rated loop flow:
l
< 100'F between reactor vessel steam space coolant and bottom head a.
Brain line coolant, b.
5 50'F between the reactor coolant within the loop not in operation and the coolant in the reactor pressure vessel, and 5 50'F between the reactor coolant within the loop not in operation c.
and the operating loop.
The differential temperature requirments of Specification 4.4.1.1.4.b and c do not apply when the loop not in operation is isolated from the reactor pressure vessel.
1 l
- The threshold THERMAL POWER and recirculation loop flow which will sweep the cold wcter from the vessel bottom head preventing stratification.
CLINTON - UNIT 1 3/4 4-4 Amendment No. 42
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REACTOR COOLANT SYSTEM JET PUMPS LIMITING CONDITION FOR OPERATION 3.4.1.2 All jet pumps shall be OPERABLE.
APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2.
ACTION:
i With one or more jet pumps inoperable..be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
SURVEILLANCE RE0VIREMENTS 4.4.1.2 All jet pumps shall be demonstrated OPERABLE as follows:
t Each of the above required jet pumps in an operating loop shall t,e demonstrated OPERABLE at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> whgn THERMAL POWER 13 greater than or equal to 25% of RATED THERMAL POWER by detennining recirculation loop flow, total core flow and diffuser-to-lower plenum differential pressure for each jet pump and verifying that no two of the following conditions occur:
a.
The indicated recirculation loop flew differs by more than 10% from the established flow control valve position-loop flow characteristics.
i b.-
The indicated total _ core flow differs by more than 10% from the established total core flow value derived from recirculation loop flow measurements.
c.
The indicated jet pump diffuser-to-lower plenum differential pressure (or jet pump flow) of any individual jet pump differs from establisted patterns by more than 20% (10% for flow).
- The provisions of Specification 4.0.4 are not applicable.
CLINTON - UNIT 1 3/4 4-6 Amendment No. 42
i REACTOR COOLANT SYSTEM 3/4.4.5 SPECIFICACTIV,IH l
LIMITING CONDITION FOR OPERATION t
3.4.5 The specific activity of the primary coolant shall be limited to:
Less than or equal to 0.2 microcuries per gram DOSE EQUIVALENT I-131, and a.
b.
Less than or equal to 100R microcuries'per gram.
APPLICABILITY:
OPERATIONAL CONDITIONS 1, 2, 3, and 4.
ACTION:
In OPERATIONAL CONDITIONS 1, 2, or 3 with the specific activity of the a.
primary coolant:
1.
Greater than 0.2 microcuries per gram DOSE EQUIVALENT I-131 but less than or equal to 4.0 microcuries per gram DOSE EQUIVALENT I-131 for more than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> during one continuous time interval or greater than 4.0 microcuries per gram DOSE EQUIVALENT I-131, be in at least HOT SHUTDOWN with the main steam line isolation valves closed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
2.
Greater than 100A microcuries per gram, be in at least HOT SHUTOOWN with the main steamline isolation valves closed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, b.
In OPERATIONAL CONDITIONS 1, 2, 3, or 4, with the specific activity of the primary coolant greater than 0.2 microcuriss per gram DOSE EQUIVALENT I-131 or greater than 100A microcuries per gram, perform the s.mpling and analysis requirements of Item 4a of Table 4.4.5-1 until the specific activity of the primary coolant is restored to within its limit.
c.
In OPERATIONAL CONDITION 1 or 2, with:
1 1.
THERMAL POWER changed by more than 15% of RATED THERMAL POWER in one hour, or j
2.
The off-gas level, at the off gas recombiner effluent, increased by t
l more than 10,000 microcuries per second in one hour during steady l
state operation at release rates less than 75,000 microcuries rer second, or 3.
The off gas level, at the off gas recombiner effluent, increased by
- 7.cre than 15% in one hour during steady state operation at release rate: greate. than 75,000 microcuries per second, perform the sampling and analysis requirements of Item 4b of Table 4.4.5-1 until the specific activity of the primary coolant is restored to within its limit.
1 CLINTON - UNIT I 3/4 4-19 Amendment No.42
1 J
REACTOR COOLANT SYSTEM SPECIFIC ACTIVITY SURVEILLANCE REQUIREMENTS r
4.4.5 The specific activity of the reactor coolant shall be demonstrated to
~
be within the limits by performance of the sampling and analysis program of Table 4.4.5-1.
l
.CLINTON - UNIT 1 3/4 4-20
3/4.4 REACTOR COOLANT SYSTEM BASES 3/4.4.1 RECIRCULATION SYSTEM The impact of single recirculation loop operation upon plant safety is assessed and shows that single-loop operation is permitted if the MCPR fuel cladding safety limit is increased as noted by Specification 2.1.2, APRM scram and control rod block setpoints are adjusted as noted in Tables 2.2.1-1 and 3.3.6-2, 1
respectively, MAPLHGR limits are decreased in accordance with the values speci-
)
fied in the CORE OPERATING LIMITS REPORT andMCPRoperatinglimitsareadjusted inaccordancewiththevaluesspecifiedIntheCOREOPERATINGLIMITSREPORT.
Additionally, surveillance on the volumetric flow rate of the operating recir-4 culation loop is imposed to exclude the possibility of excessive core internals vibration.
The surveillance on differential temperatures below 30%* THERMAL POWER or 30%* rated recirculation loop flow is to mitigate the undue thermal stress on vessel nozzles, recirculation pump, and vessel bottom head during the
(
extended operation of the single recirculation loop mode.
l An inoperable jet pump is not, in itself, a sufficient reason to declare a re-circulation loop inoperable, but it does, in case of a design-basis-accident, l
increase the blowdown area and reduce the capability of reflooding the core; thus, the requirement for shutdown of the facility with a jet pump inoperable.
Jetpumpfailurecanbedetectedbymonitoringjetpumpperformanceonapre-scribed schedule for significant degradation.
significant degradation is indicated if more than one of three specified surveillances performed confirms l
unacceptable deviations from established patterns or relationships.
The surveillances, including the associated acceptance criteria, are in accordance t
with General Electric Service Information Letter No. 330, the recommendations of which are considered accep' Closeout of IE Bulletin 80-07: table for verifying je z'
according to NUREG/CR-3052, BWR Jet Pum Assembly Failure." Performance of the specified surveillances, however,pis not required when thermal power is less than 25% RATED THERMAL POWER because flowoscillationsandjetnoiseprecludesthecollectionofrepeatable meaningful data during low flow conditions approaching the threshold response of the associated flow instrumentation.
Recirculation loop flow mismatch limits are in compliance with ECCS LOCA analysis design ctiteria for two recirculation loop operation.
The limits will ensure an adequate core flow coastdown from either recirculation loop following a LOCA.
In the case where the mismatch limits cannot be maintained during two loop operation, continued operation is permitted in a single recirculation loop mode.
In order to prevent undue stress on the vessel nozzles and bottom head region, the recirculation loop temperatures shall be within 50'F of each other p'rior to startup of an idle loop.
The loop tes.perature must also be within 50 F of the reactor pressure vessel coolant temperature to prevent thermal shock to the recirculation pump and recirculation nozzles.
Sudden equilization of a tempera-ture difference > 100'F between the reactor vessel bottom head coolant and the coolant in the upper region of the reactor vessel by increasing core flow rate would cause undue stress in the reactor vessel bottom head.
- The threshold THERMAL POWER and recirculation loop flow which will sweep the cold water from the vessel bottom head preventing stratification.
CLINTON - UNIT 1 B 3/4 4-1 Amendment No. 42
BASES 3/4.4.1 RECIRCULATION SYSTEM (Continued)
The objective of GE BWR plant and fuel design is to provide stable operation i
with margin over the normal operating domain.
However, at the high power / low flow corner of the operating domain, a small probability of neutron flux limit cycle oscillations exists depending on combinations of operating conditions (e.g., rod pattern, power shape).
To provide assurance that neutron flux limit cycle oscillations are detected and suppressed, APRM and LPRM neutron flux noire levels should be monitored while operating in this region.
Stabili',y tests at operating BWRs were reviewed to determine a generic region of the power / flow map in which surveillance of neutron flux noise levels should be performed.
A conservative decay ratio of 0.6 was chosen as the bases for determining the generic region for surveillance to account for the plant to plant variability of decay ratio with core and fuel designs.
This generic region has been determined to correspond to a core flow of less than or equal i
to 45% of rated core flow and a THERMAL POWER greater than that specified in Figure 3.4.1.1-1.
Plant specific calculations can be performed to determine an applicable region for monitoring neutron flux noise levels.
In this case the degree of conserva-tism can be reduced since plant to plant variability would be eliminated.
In this case, adequate margin will be assured by monitoring the region which has a decay ratio greater than or equal to 0.8.
Neutron flux noise limits are also established to ensure early detection of limit cycle neutron flux oscillations.
BWR cores ty flux noise caused by random boiling and flow noise. pically operate with neutron Typical neutron flux noise levels of 1-12% or rated power (peak-to peak) have been reported for the range of low.to high recirculation loop flow during both single and dual recirculation loop operation.
Neutron flux noise levels which significantly bound these values are considered in the thermat/mechar.ical design of GE BWR fuel and are found to be of negligible consequence.
In addition, stability tests at operat-ing BWRs have demonstrated that when stability related neutron flux limit cycle oscillations occur they result in peak-to peak neutron flux limit cycles of 5-10 times the typical values.
Therefore, actions taken to reduce neutron flux noise levels exceeding three (3) times the typical value are sufficient to 1
ensure early detection of limit cycle neutron flux oscillations.
Typically, neutron flux noise levels show a gradual increase in absolute magni-
{
tude as enre flow is increased (constant control rod pattern) with two reactor recirculation loops in operation.
Therefore, the baseline neutron flux noise level obtained at a specific core flow can be applied over a range of core flows.
To maintain a reasonable variation between the low flow and high flow and of the flow range, the range over which a specific baseline is applied should not exceed 20% of rated core flow with two recirculation loops in operation.
Data from tests and operating plants indicate that a range of 20% of rated core flow will result in approximately a 50% increase in neutron flux noise level during operation with two recirculation loops.
Baseline data should be taken near the maximum rod line at which the majority of operation will occur.
However, baseline data taken at low rod lines (i.e. lower power) will result in a conservative value since the neutron flux noise level is proportional to the power level at a given core ficw.
CLINTON - UNIT 1 B 3/4 4-2 Amendment No. 18
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