ML20247K027
| ML20247K027 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 09/13/1989 |
| From: | COMMONWEALTH EDISON CO. |
| To: | |
| Shared Package | |
| ML20247K014 | List: |
| References | |
| 0284T:1, 284T:1, NUDOCS 8909210050 | |
| Download: ML20247K027 (14) | |
Text
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, k, ATTACHMENT A' UNIT 1 (NPF-11)
SUPPLEMENTAL CHANGES PAGE CHANGE B 3/4.2-l' Revised INSERT C to discuss mechanical design.
aspects.
B 3/4.2-5 Added INSERT E to retain appropriate Kf discussion.
6-24 Revised INSERT B to correct typographical error in Provision 6.C.
l' if 8909210050 s90913 fDR ADOCK 05000373 FDC Ol84T:9-
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3/4.2 POWER DISTRIBUTION LIMITS BASES The specifications of this section assure that the peak cladding temperature followin the2200gthepostulateddesignbasisloss-of-coolantaccidentwillnotexceed F limit specified in 10 CFR 50.46.
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 50.46. The' specification also assures that fuel rod mechanical integrity is maintained during normal and transient operations.
The peak cladding 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 a 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 AVERAGE PLANAR LINEAR HEAT GENERATION RATE
/(ZA/5EF (APLHGR) is this LHGR of the highest powereu rod divided by its local peaking C farter.
The calculational procedure used to establish the APLHGR values for.the in 1 cycle and first reload fuel shown,on Figures 3.2.1-1 and 3.2.1-2 ar based loss-of-coolant accident analysis.
The analysis was perforsu using Gene Electric (GE) calculational models which are consisten th the requiremen of Appendix K to 10 CFR 50.- Ac te discussi of each code employed in analysis is presented in Ref; e
O erences in this analysis compare previous analyses on t
forence 1 are:
(1) the analysis assumes el assemb1 p/sionp sistent with 102% of ar pc the MPLHGR shown in Figure 1-1; ct decay is computed assuming an energy release rate o y/f ss
- (3) pool boiling is assumed after nucleate boiling is 1 t duri e f1 stagnation period; and (4) the effects of core spray ent it an co ar-current flow limitation as described in Refer i#fude n ti flooding calculations.
The APLHGR val e
oad fuel shown in ure 3.2.1-3 are based on the fuel thermal-mec ca sign analysis.
The impro SAFER /GESTR-LOCA analysis (Reference 3 formed before the startup of Cyc used bounding M PLHGR values of and 14.0 kw/ft, independent of nodal re. These MPLHGR ~ values igher than the expected " thermal-mechanical M R" for both BP8x8R GE8x8EB fuel. Therefore, SAFER /GESTR established that all BP8x8R an 8x8EB fuel designs, the MAPLHGR values are not expected to be limi y LOCA/ECCS considerations.
However, MAPLHGR values are still requi to sure that the LHGR limits are not compromised; and, consequently, fuel rod chanical integrity is maintained.
LA SALLE UNIT 1 B 3/4 2-1 Amendment No. 58
F.
s t
s; l:
INSERT C i
However, the current General Electric (GE) calculational models (SAFER /GESTR described in Reference 3), which are consistent with the requirements of Appendix.K to 10 CFR 50, have established that APLHGR values are not expected to be limited by LOCA/ECCS considerations. APLHGR limits are still required, however, to assure that fuel. rod mechanical. integrity is maintained.
They are specified for all resident fuel. types in the Core Operating Limit Report based on the fuel thermal-mechanical design analysis.
02841.4
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. a; POWER DISTRIBUTION SYSTEMS B,ASES MINIMUM CRITICAL POWER RATIO ~(Continued)
The value for 1 used in Specification 3.2.3 is 0.687 seconds which is 8
conservative for the following reason:
For simplicity in formulating and implementing the LCO, a conservative n
value for IN9 of 598 was used. This represents one full core data set i=1 at BOC plus one full core data set following a 120 day outage plus twelve 10% of core, 19 rods, data sets.
The 12 data sets are equivalent to 24 operating months of surveillance at the increased surveillance frequency of one set per 60 days required by the action statements of Specifications 3.1.3.2 and 3.1.3.4.
That is, a cycle length was assumed which is longer than any past or contemplated refueling interval and the number of. rods tested was maximized in order to simplify and conservatively reduce the criteria for the scram time at which MCPR penalization is necessary.
{[7A7SgRT E)
The purpose of the K atotherthanrate$coreflowconditions. factor of Figure 3.2.3-2 is to define operating 1
1 At less than 100% of ra flow the ired MCPR is the product of the MCPR and the K factor, eK(ors factors assu t the Safety Limit MCPR will not be viola ed.
K fac were' derived using L POWER and core f ow corresponds 105%of rated steam flow.
The K factors were Ic ala or the maximum core flow rate g
and the correspond LM ong t 5% of rated steam flow control line, the 1 b
s relative powe adjusted until the MCPR was slightly above y Limit. Using this relat ndle power, the r
MCPRs were calcula at different points along the 105% of steam flow control line sponding to different core flows.
The ratio of PR calcula t a given point of core flow, divided by the operating lia de tes the K.
f At THERMAL POWER levels less than or equal to 25% of RATED THERMAL POWER, the reactor will be operating at minimum recirculation pump speed and the moderator void content will be very small.
For all designated control rod patterns which may be amployed at this point, operating plant experience indicates that the resulting MCPR value is in excess of requirements by a considerable margin.
During initial start-up testing of the plant, a MCPR evaluation will be made at 25% of RATED THERMAL POWER level with minimum LA SALLE UNIT 1 B 3/4 2-5 Amendment No. 58
l f
I h
BSERT E The purpose of the Kf factor specified in the Core Operating Limits Report is to define' operating limits at other than rated core flow conditions.
- At less than 1007. of rated flow,.the required MCPR is the product of the MCPR and the Kf factor.
The Kf factor assures that the Safety Limit MCPR will not-be violated. Methodology for establishing the Kf factor is described'in Reference 4.
0284T:5
9 ADMINISTRATIVE CONTROLS Semiannual Radioactive Effluent Release Report (Continued)
The radioactive effluent release report shall include the following information for each type of solid waste shipped offsite during the report period:
Container volume, a.
b.
Total curie quantity (specify whether determined by measurement or estimate),
Principal radionuclides (specify whether determined by c.
measurement or estimate),
d.
Type of waste (e.g., spent resin, compacted dry waste, i
evaporator bottcas),
Type of container (e.g., LSA, Type A, Type B, Large Quantity),
e.
and
\\
f.
Solidification agent (e.g., cement, urea formaldehyde).
t The radioactive affluent release reports shall include unplanned 2
releases from the site to unrestricted areas of radioactive materials in gaseous and liquid effluents on a quarter.ly basis.
1
)
The radioactive effluent r:1 ease reports shall include a,ny changes to the PROCESS CONTROL PROGRAM (PCP) made during the reporting pedod.
i S.
Monthly Operating Report Routine reports of operating statistics and shutdown experience, including documentation of all challenges to safety / relief valves, shall be submitted on c monthly basis to the Director, Office of Commission, Washington, DC 20555, with a copy of th Regional Office, to arrive no later than the 15th of each month l
following the calendar month covered by the report.
i i
Any changes to the OFFSITE DOSE CALCULATION MANUAL shall be submitted with the Monthly Operating Report within 90 days in which the change (s) was made effective.
In addition, a report of any major changes to the radioactive weste treatmtnt systems shall be submitted with the l
Monthly Operating Report for the period in which the evaluation was i
reviewed and accepted by Onsite Review and Investigative Function.
4 B.
Deleted
~N.
pygy-g }
I i
i LA SALLE UNIT 1 6-24 Amendment No. 66
_____m_ _ _ _ _ _ _. _ _ _ _ _. _ _ _ _
l INSERT B-6.
CORE.0PERATING LIMITS REPORT a.
Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT before each reload cycle or any remaining'part of a-reload cycle for the following:
(1) The Average Planar Linear Heat Generation Rate (APLHGR) for Technical Specification 3.2.1.
(2) The minimum Critical Power Ratio (MCPR) (including 20% scram time, tau (f), dependent MCPR limits, and Kf core flow MCPR adjustment factors) for Technical. Specification 3.2.3.
(3) The Linear Heat Generation' Rate (LHGR) for Technical Specification 3.2.4.
(4) The Rod Block Monitor Upscale Instrumentation Setpoints for Technical Specification Table 3.3.6-2.
'b.
The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by NRC in NEDE-24011-P-A, General Electric Standard Application for Reactor Fuel.(latest approved revision).
c.
The core operating limits shall be determined so that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met.
d.
The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements thereto, shall be provided upon issuance, for each
-reload cycle, to the U.S. Nuclear Regulatory Commission Document Control Desk with copies to the Regional Administrator and Resident Inspector.
I 0284T:12
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l 1
AllAC11 MENT B UNIT 2 (NPF-18)-
SUPPLEMENTAL CHANGES.
EAGE CHANGE B 3/4.2-l' Revised INSERT C to discuss mechanical. design aspects.
B 3/4.2 Added INSERT E to retain appropriate Kf discussion.
6-24 Revised INSERT B to correct typographical error in Provision 6.C.
0284T:10
3/4.2 POWER DISTRIBUTION LIMITS 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 10 CFR 50.46.
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 50.46.
This specification also assures that fuel rod mechanical integrity is maintained during normal and transient operations.
The peak cladding 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 a LHGR for the highest powered rod which is equal to or less than the design LHGR corrected for densification.
This WGR 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 AVERAGE PLANAR LINEAR HEAT GENERATION RATE (AP is this LHGR of the highest powered rod divided by its loc eg ag.
The calculational procedure used to establish the APLHGR values for the in 1 cycle and first reload fuel shown on Figure 3.2.1-1 and 3.2.1-2 are based a loss of coolant accident analysis.
The analysis was perform sing General tric (GE) calculational models which are consistent with requirements Appendix K to 10 CFR Part 50.
A complete discuss of each e
code employed in e analysis is presented in Reference 1.
Di rences in this analysis compared t evious analyses performed with Refe ce 1 are:
(1) the analysis assumes a fue sembly planar powe iste ith 102% of the MAPLHGR shown in Figure 3.. -1, (2) fissi ecay is computed assuming uc an energy release rate of 200
/fisr#n 3 p boiling is assumed after nucleate boiling is lost during t sta ion period; and (4) the effects of core spray entrainment and
@te r
flow limitation as described in Reference 2, are include n
beflo calculations.
The APLHGR values fo b.
oad fuel sho in Figure 3.2.1-3 are based l
on the fuel thermal-mecha c design analysis.
improved SAFER /GESTR-LOCA 1
analysis (Reference 3) p ormed for Cycle 3 used bou g MAPLHGR values of 13.0 and 14.0 kw/ft, ependent of nodal exposure.
Thes APLHGR values are higher than the e ted " thermal-mechanical MAPLHGR" for bo IP8x8R and GE8x8EB fuel.
erefore, SAFER /GESTR established that for all 8R and GE8x8EB fue esigns the MAPLHGR values are not expected to be limi by LOCA/ECC considerations.
However, MAPLHGR values are still required hat the LHGR limits are not compromised and, consequently, fuel r assu anical integrity is maintained.
m LA SALLE - UNIT 2 B 3/4 2-1 Amendment No.
41 L_- i_--- --
t '-
i i
1 i
INSERT C i
However, the current-General Electric (GE) calculational models (SAFER /GESTR described in Reference 3), which are consistent with the requirements of Appendix K to 10 CFR 50, have established that APLHGR values 4
are not expected to be limited by LOCA/ECCS considerations. APLHGR. limits are still required, however, to assure that fuel rod mechanical integrity is maintained.
They are specified for all resident fuel types in the Core Operating Limit Report based on the fuel thermal-mechanical design analysis.
i i
i i
a
)
0284T:4 1
1 POWER DISTRIBUTION SYSTEMS BASES MINIMUM CRITICAL POWER RATIO (Continued)
The value for T used in Specification 3.2.3 is 0.687 seconds which is g
conservative for the following reason:
For simplicity in formulating and implementing the LCO, a conservative n
value for I N9 of 598 was used.
This represents one full core data set i=1 at 80C plus one full core data set following a 120 day outage plus twelve 10% of core, 19 rods, data sets.
The 12 data sets are equivalent to 24 operating months of surveillance at the increased surveillance frequency of one~ set per 60 days required by the action statements of
~
Specifications 3.1.3.2 and 3.1.3.4.
That'is, a cycle length was assumed which is longer than any past or contemplated refueling interval and the number of rods tested was maximized in order to simplify and conservatively reduce the criteria' for the scram time at which MCPR penalization is necessary.
f-[I)yyggr E }
Ine purpose of the K totherthanrate8coreflowconditions. factor of Figure 3.2.3-2 is to define operating lim At less than 100% of r flow the fred MCPR is the product of the MCPR and the K(e eK,fac(ors factor eK factors assure the Safety Limit MCPR will n t be viola were derived using L POWER and core respon o 105% of rated steam flow.
The K factors were eu at ht r the maximum core flow rate 7
and the corresponds H
leWE g
105% of rated steam flow control line, the 1 ' i s relative p as adjusted until the MCPR was slightly above t Limit.
Using this, rela ~
bundle power, the MCPRs were calcul different points along tFa 105% o d steam flow control line sponding to different core flows. The ratio o MCPR calcul a given point of core flow, divided by the operating lia PR,
..ines the K,.
At THERMAL POWER levels less than or equal to 25% of RATED THERMAL POWER, the reactor will be operating at minimum recirculation pump speed and the moderator void content will be very small.
For all designated control rod patterns which may be employed at this point, operating plant experience indicates that the resulting MCPR value is in excess of requirements by a considerable margin.
During initial start-up testing of the plant, a MCPR evaluation will be made at 25% of RATED THERMAL POWER level with minimum recirculation pump speed. The MCPR margin will thus be demonstrated such that future MCPR evaluation below this power level will be shown to be unnecessary.
The daily requirement for calculating MCPR when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER is sufficient since power distribution shifts are very slow when there have not been significant power or control rod changes.
The requirement for calculating MCPR when a limiting control rod pattern is approached ensures that MCPR will be known following a change in THERMAL POWER or power shape, regardless of magnitude, that could place operation at a thermal limit.
LA SALLE - UNIT 2 B 3/4 2-6 Amendment No. 41
INSERLE The purpose of the Kf factor specified in the Core Operating Limits Report is to define operating limits at other than rated core flow conditions.
At less than 1007. of rated flow, the required MCPR is the product of the MCPR and the Kf factor.
The Kf factor assures that the Safety Limit MCPR will not be violated. Methodology for establishing the Kf factor is described in Reference 4.
0284T:5 l
______-__-__________n
O"".!STDnIO" CONTROLS '
Se-ta.r.a1 Radioactive Effluent Release Report (Continued)
'The radioactive effluent release report shall. include the following information for each type of solid waste shipped offsite during the report period:
Container volume, a.
b.
Total curie quantity (specify whe'ther determined by measurement or estimate),
Principal: radionuclides (specify whether determined by c.
measurement or astimate),
d.
Type'of waste (e.g., spent resin, compacted dry waste, evaporator bottoms),
Type of container (e.g., LSA, Type A, Type B, Large Quantity),
e.
and f.
Solidification agent (e.g., coment, urea formaldehyde). -
The radioactive effluent release reports sha'll include unplanned releases from the site to unrestricted areas of radioactive materials in gaseous and liquid effluents on a quarterly basis.
The radioactive effluent release reports shall include any changes to the PROCESS CONTROL PROGRAM (PCP) made during the reporting period.
5.
Monthly Operating Report Routine reports of operating statistics and shutdown experience,
. including documentation of all challenges to safety / relief valves, shall be submitted on a monthly basis to the Director, Office of Nu lear Reactor Regulation, Mail Station P1-137, US Nuclear Regulatory Commission, Washington, DC 20555, with a copy of the appropriate i.
Regional Office, to arrive no later than the 15th of each month following the calendar month covered by the report.
Any changes to the 0FFSITE DOSE CALCULATION MANUAL iikall be submitted with the Monthly Operating Report within 90 days in which the change (s) was made effective.
In addition, a report of any major changes to 1-the radioactive waste treatment systems shall-be submitted with the Monthly Operating Report for the period in which the evaluation was reviewed and accepted by Onsite Review and Investigative Function.
B.
De ted.
hNSART O LA SALLE - UNIT 2 6-24 Amendment No. 47
,4 w
1ESBLR
- 6. l CORE' OPERATING LIMITS REPORT a.
Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT before each reload cycle or any remaining part of a reload cycle for the following:
(1) The Average Planar Linear Heat Generation Rate (APLHGR) for-Technical Specification 3.2.1.
(2) The minimum Critical Power Ratio (MCPR) (including 20% scram time, tau ( ), dependent MCPR limits, and Kf core flow MCPR adjustment factors) for Technical Specification 3.2.3.
'(3) The Linear Heat Generation Rate (LHGR) for Technical Specification 3.2.4.
(4) The Rod Block Monitor Upscalc. Instrumentation Setpoints for Technical Specification Table 3.3.6-2.
b.
The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by NRC in NEDE-24011-P-A, General Electric Standard Application for-Reactor Fuel (latest approved revision).
c.
The core operating limits shall be determined so that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS 11 nits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of'the safety analysis are met.
d.
The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements thereto, shall be provided upon issuance, for each reload cycle, to the U.S. Nuclear Regulatory Commission Document Control Desk with copies to the Regional Administrator and Resident Inspector.
0284T:11 l.
t