ML20137G607
| ML20137G607 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 03/13/1997 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20137G594 | List: |
| References | |
| NUDOCS 9704010447 | |
| Download: ML20137G607 (11) | |
Text
~
ATTACHMENT 1 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION E SARDING PROPOSED CHANGES TO THE TECHNICAL SPECIFICATIONS AND BASES COMMONWEALTH EDISON COMPANY 00AD CITIES NUCLEAR POWER STATION. UNITS 1 AND 2 DOCKET N05. 50-254 AND 50-265
1.0 INTRODUCTION
By letter dated June 10, 1996, as supplemented by letter dated February 17, 1997, Commonwealth Edison Company, the licensee, requested changes to the Quad Cities Nuclear Power Station, Units I and 2, Technical Specifications (TS).
Quad Cities Units 1 and 2 currently use General Electric (GE) fuel and licensing methodologies.
Siemens Power Corporation (SPC) fuel and licensing methodologies are planned for use at Quad Cities beginning with Unit 2 Cycle 15 and Unit 1 Cycle 16. The Siemens' LOCA methodology and fuel assembly designs are approved for use at other licensed BWR facilities.
Thus, the proposed changes to the Quad Cities Units 1 and 2 TS represent the transition from one NRC-approved methodology to another NRC-approved methodology. Other minor editorial changes are also proposed.
By letter dated February 17, 1997, Comed submitted revisions that were required for the approval of Technical Specification changes for SPC fuel transition for LaSalle County Nuclear Power Station Units 1 and 2.
The revision lists the specific NRC approval date and the revision / supplement for each of the new topical reports, and revises section 5.3.A description of fuel assemblies.
2.0 EVALUATION 2.1 Definitions Linear heat generation rate (LHGR) limits are monitored for GE fuel by the parameters fraction of limiting power density (FLPD) and maximum fraction of limiting power density (MFLPD). The licensee proposed to add "(applicable to GE fuel)" to the end of each of these definitions to distinguish GE parameters from SPC parameters.
SPC uses Fuel Design Limiting Ratio For Centerline Melt (FDLRC) and Fuel Design Limiting Ratio (FDLRX) to monitor LHGR, The licensee has proposed to add the following definitions of FDLRC and FDLRX which are applicable to SPC fuel:
D)El DESIGN LIMITING RATIO FOR CENTERLINE MELT (FDLRC)
The FUEL DES;GN LIMITING RATIO FOR CENTERLINE MELT (FDLRC) shall be 1.2 times the LHGR at a given location divided by the product of the TRANSIENT LINEAR HEAT GENERATION RATE limit and the FRACTION OF RATED THERMAL POWER (applicable to SPC Fuel).
9704010447 970313 DR ADOCK 05000254 PDR
4 2
FUEL DESIGN LIMITING RATIO (FDLRX) l The FUEL DESIGN LIMITING RATION (FDLRX) shall be the limit used to assure that the fuel operates within the end-of-life steady-state design criteria by, among other items, limiting the release of fission gas to the cladding plenum (applicable to SPC Fuel).
The licensee has proposed to delete the definition of Rod Density.
Rod density will be replaced by critical control rod configuration in order to make use of the capability to monitor actual K,, versus predicted K.,,.
The licensee proposed to add the definition of the SPC transient LHGR limit to the Definitions. The proposed definition is as follows:
IRANSIENT LINEAR HEAT GENERATION RATE (TLHGR)
The TRANSIENT LINEAR HEAT GENERATION RATE (TLHGR) limit protects against fuel centerline melting and 1% plastic cladding strain during transient conditions throughout the life of the fuel (applicable to SPC Fuel).
The staff notes that the GE LHGR limits will be applied to the co-resident GE fuel in the core and the SPC LHGR limits will be applied to the SPC fuel in the core. The existing LHGR Technical Specification Bases will be modified to show applicability to both GE and SPC fuel.
2.2 Safety Limits Bases The licensee proposed an editorial change to the section 2.1, third paragraph, of the Bases. The current wording states "the fuel cladding integrity limit is set such that no calculated fuel damage would occur as a result of an A00."
The proposed change would consist of the following:
The fuel cladding integrity limit is set such that no fuel damage is calculated to occur as a result of an A00.
The staff concludes that the change clarifies the meaning of the sentence and is acceptable.
In section 2.1.B, Thermal Power, High Pressure and High Flow, the licensee proposed editorial changes to paragraph one. The current wording of the last sentence states that "the fuel cladding integrity Safety Limit is defined as the CPR in the limiting fuel assembly for which more than 99.9% of the fuel rods in the core are expected to avoid boiling transition considering the power distribution within the core and all uncertainties." The editorial change would have the last two sentences of paragraph one consist of the following:
Therefore, the fuel cladding integrity Safety Limit is defined such that, with the limiting fuel assembly operating at the MCPR Safety Limit, more than 99.9% of the fuel rods in the core are expected to avoid boiling transition. This includes consideration of the power distribution within the core and all uncertainties.
The staff notes that this wording is consistent with the bases in Improved
1 3
Standard Technical Specifications, NUREG-1433, Rev.1, and therefore, it is acceptable.
2.3 Limitina Safety System Settinas Bases In section 2.2.A.1, Reactor Protection System Instrumentation Setpoints -
l l
Intermediate Range Monitor, Neutron Flux - High, the licensee proposed an editorial change to the third paragraph. The sentence with the proposed change states that "the results of this analysis show that the reactor is scrammed and peak power is limited to 1% of rated power, thus maintaining MCPR above the fuel cladding integrity Safety Limit." The licensee proposed to i
change 1% to 7.7% to reflect the correct value in the UFSAR and SAR analysis.
l Section 7.6.1 of the UFSAR cites, in graphical form, 7.7% as the power level at which the IRMs terminate the low power RWE transient. With two other editorial changes, the proposed statement will read as follows:
The results of this analysis show that the reactor is scrammed and peak local power is limited to 7.7% of rated bundle power.
Based on this information, the staff finds this editorial change acceptable.
In section 2.2.A.4, Reactor Protection System Instrumentation Setpoints -
Reactor Vessel Water Level - Low, the licensee proposed to add a clarification of the top of active fuel at the end of the last paragraph. The proposed last sentence of the paragraph would read "the top of active fuel is defined to be 360 inches above vessel zero." This statement is consistent with footnotes and other sections of the bases and therefore, is acceptable.
2.4 Instrumentation Bases The licensee proposed a clarification to section 3/4.2, Instrumentation. The licensee proposed to add the following paragraph to section 3/4.2.
Current fuel designs incorporate slight variations in the length of the active fuel, and thus the actual top of active fuel, when compared to the original fuel designs.
Safety Limits, instrument water level setpoints, and associated LCOs refer to the top of active fuel.
In these cases, the top of active fuel is defined as 360 inches above vessel zero.
Licensing analyses, both accident and transient, utilize this definition for the automatic initiation and manual intervention associated with these events.
The proposed additions provide a clear definition and use of the top of active fuel reference point. The staff finds this addition to the bases acceptable.
2.5 Egactivity Control Limitina Conditions For Operation And Bases TS 3.3.B. Reactivity Anomalies, currently requires that the reactivity equivalence of the difference between the actual R0D DENSITY and the predicted R0D DENSITY shall not exceed 1% Ak/k.
In addition, Surveillance Requirement 4.3.B require.s that the reactivity equivalence of the difference between the actual R0D DENSITY and the predicted R0D DENSITY shall be verified to be less than or equal to 1% Ak/k.
This limit ensures that plant operation is
i l
4 i
maintained within the assumptions of the safety analyses.
The licensee has proposed to replace " ROD DENSITY" in both the technical specification and the surveillance requirement with " critical control rod i
configuration." This proposed change is necessitated by the additional l
proposal to include an alternative to monitoring reactivity anomalies in the l
technical specification bases.
Both the SPC core monitoring code, Powerplex, and the GE Core Monitoring Code (CMC) provides the capability to monitor actual K, versus predicted K Rod Densl,ty to critical contrld,. The licensee stated that the change from rod configuration was necessary in order to use this capability. The staff notes that this method is currently used at Dresden to monitor reactivity anomalies. Thus the following will be added to section 3/4.3.B, Reactivity Anomalies Bases:
Alternatively, monitored K calculatedbyanapprovedY#canbecomparedwiththepredictedK,,,as D core simulator code.
When the reactor core is critical or in normal power operation, a reactivity balance exists and the net reactivity is zero. A comparison of predicted and i
measured reactivity is convenient under such a balance, since parameters are i
being maintained relatively stable under steady state power conditions. The staff notes that this proposed change only revises the current method of measuring the difference between predicted and monitored core reactivity and does not change the required limit, therefore, the changes to TS 3/4.3.B and its Bases is acceptable.
In sections 3/4.3.D, 3/4.3.E, and 3/4.3.F, Control Red Maximum Scram Insertion Times, Control Rod Average Scram Insertion Times, and Four Control Rod Group Scram Insertion Times, of the TS Bases, the licensee proposed to remove the following comments:
first paragraph:
"(as adjusted for statistical variation in the observed data);"
second paragraph:
"In the statistical treatment of the limiting transients, a statistical distribution of total scram delay is used rather than the bounding value described above;"
third paragraph:
" Observed plant data or Technical Specification limits were used to determine the average scram performance used in the transient analyses, and the results of each set of control rod scram tests performed during the current cycle are compared against earlier results to verify that the performance of the control rod insertion system has i
not changed significantly;" and j
fourth paragraph:
"If test results should be determined to fall outside of the statistical population defining the scram performance characteristics used in the transient analyses, a re-determination of thermal margin requirements is undertaken as required by Specification 3.11.C.
A smaller test sample than that required by these specifications is not statistically significant and l
l l
1 5
should not be used in the re-determination of thermal margins."
The licensee stated that the above information is based on past data which is j
a GE methodology.
Current SPC methods used to evaluate the 5%, 20%, 50% and 90% control rod scram insertion times, collected during the performance of the scram timing Surveillance Requirement 4.3.D, will replace the above information as follows:
Transient analyses are performed for both Technical Specification Scram Speed (TSSS) and nominal scram speed (NSS) insertion times. These analyses result in the establishment of the cycle dependent TSSS MCPR limits and NSS MCPR limits presented in the COLR. Results of the control rod scram tests performed during the current cycle are used to determine the operating limit for MCPR.
Following the completion of each set of scram testing, the results will be compared with the assumptions used in the transient analysis to verify the applicability of the MCPR operating limits.
Prior to the initial scram time testing for an operating cycle, the MCPR operating limits will be based on the TSSS insertion times.
The NSS insertion times are typically faster than the TSSS insertion times, thus, the NSS insertion times are used to calculate the NSS MCPR operating limit.
If any of the average scram insertion times do not meet the NSS times, the TSSS MCPR operating limit is used.
TS 3.11.C, Minimum Critical Power Ratio, requires that the MCPR shall be equal to or greater than the MCPR l
operating limit specified in the COLR.
These changes to the bases clarify the SPC methodology that will be used at Quad Cities and how it will be used to meet TS 3.11.C.
Based on this information, the changes to section 3/4.3.D, 3.E, and 3.F bases are acceptable.
j In section 3/4.3.L Rod Worth Minimizer, the licensee proposed editorial changes to the first paragraph of the Bases.
Currently, the first two sentences state that " control rod withdrawal and insertion sequences are established to assure that the maximum insequence individual control rod or control rod segments which are withdrawn at any time during the fuel cycle could not be worth enough to result in a peak fuel enthalpy greater than 280 cal /gm in the event of a control rod drop accident. These sequences are developed prior to initial operation of the unit following any refueling outage and the requirement that an operator follow these sequences is supervised by the RWM or a second technically qualified individual." The editorial changes would have the first two sentences consist of the following:
l Control rod withdrawal and insertion sequences are established to assure that the maximum insequence individual control rod or control red segments which are withdrawn at any time during the fuel cycle could not I
have sufficient reactivity worth to result in a peak fuel enthalpy greater than 280 cal /gm in the event of a control rod drop accident.
These low power (up to the LPSP) sequences are verified during the cycle reload analysis to ensure that the 280 cal /gm limit is not exceeded.
The requirement that an operator follow these sequences is supervised by the RWM or a second technically qualified individual.
6 The licensee also proposed editorial changes to last sentence of the third paragraph. The last sentence will be replaced by the following:
The methodology used for the control rod drop accident analysis is NRC approved and is part of the license bases referenced in Specification 6.9.A.6.
The staff notes that these editorial changes clarify that the control rod sequences used during the cycle are not all written prior to cycle startup, but are verified to meet the 280 cal /gr. limit up to the Low Power Set Point 4
(LPSP). This verification is completed using NRC-approved methodologies which are referenced in TS 6.9.A.6.
Based on the above, the staff finds the 1
editorial changes acceptable, l
2.6 Primary System Boundary Bases In sections 3/4.6.E and 3/4.6.F, Safety Valves and Relief Valves, the licensee proposed to add the following sentence to the middle of the first paragraph:
SPC methodology determines the most limiting pressurization transient each cycle.
The addition of this statement clarifies the SPC methodology for analyzing the overpressurization event and therefore, is acceptable.
2.7 Power Distribution Limits Limitina Conditions For Operation And Bases As stated above, FLPD and MFLPD are LHGR terminology which are specific to GE fuel. The co-resident GE fuel in the core will be monitored by the GE fuel dependent LHGR limits, FLPD and MFLPD, and the SPC fuel will be monitored by the SPC LHGR limits, FDLRC and FDLRX.
The staff notes that the SPC fuel is protected from off rated transients by the application of FDLRC to the APRM setpoints. Based on this, the licensee proposed to revise TS 3/4.11.B, Average Power Range Monitor Setpoints, to reflect the SPC FDLRC limit and the requirement to modify the APRM setpoints if FDLRC is greater than 1.0 for SPC fuel. The propose change to TS 3/4.11.B is identical to Dresden Unit 2/3 TS 3/4.11.8 except for the following:
- 1) a footnote, (a), is added to the appropriate FDLRC statements, and 2) the current footnote (a) will become footnote (b).
The proposed footnote (a) will state the following:
For GE fuel, MFLPD/FRTP is substituted for FDLRC. Adjustments are based on the lowest APRM setpoint or highest APRM reading resulting from the two limits.
The staff notes that TS 3/4.11.B will be titled Transient Linear Heat Generation Rate instead of the current title, Average Power Range Monitor Setpoints. The staff has reviewed the Dresden TS 3/4.11.B and compared it to the proposed changes above.
Since the TLHGR and FDLRC limits for SPC fuels are applied to the APRM setpoints, the staff finds the propose changes to TS 3/4.11.8 acceptable.
7 The licensee also proposed changes to sections 3/4.11.A, 3/4.11.B, and 3/4.11.C,. Average Planar Linear Heat Generation Rate, APRM Setpoints, and Minimum Critical Power Ratio, of the TS Bases, in order to provide clarification of the SPC methodology for the application of thermal limits.
TS 3.II.A requires that all APLHGR for each type of fuel as a function of bundle average exposure shall not exceed the limits specified in the COLR.
For section 3/4.11.A, the licensee proposed to the following changes:
1) relocate the last two paragraphs of section 3/4.ll.A to the beginning of the section 3/4.ll.A Bases, 2) insert "GE Fuel" in front of the current first paragraph, and 3) add the following paragraphs to describe the SPC methodology:
SPC Fuel This specification assures that the peak cladding temperature of SPC fuel following a postulated design basis loss-of-coolant accident 'will not exceed the Peak Cladding Temperature (PCT) and maximum oxidation limits specified in 10 CFR 50.46.
The calculational procedure used to establish the Average Planar Linear Heat Generation Rate (APLHGR) limits is based on a loss-of-coolant accident analysis.
The 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 not strongly influenced by the rod-to-rod power distribution within the assembly.
The Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) limits for two-loop and single-loop operation are specified in the Core Operating Limits Report (COLR).
The staff finds the section 3/1.ll. A Bases changes described above acceptable.
TS 3.11.B requires, based on the proposed change discussed above, that the Transient Linear Heat Generation Rate shall be maintained such that the Fuel Design Limiting Ratio for Centerline Melt (FDLRC) is less than or equal to 1.0.
The licensee proposed to delete the first sentence of the paragraph since it is no longer applicable.
Furthermore, the licensee proposed to add "or FDLRC" following "MFLPD" in the last sentence of the original paragraph and add the following paragraphs to expand on the SPC methodology:
SPC Fuel The Fuel Design Limiting Ratio for Centerline Melt (FDLRC) is incorporated to protect the above criteria at all power levels considering events which cause the reactor power to increase to 120% of rated thermal power.
The scram settings must be adjusted to ensure that the TRANSIENT LINEAR HEAT GENERATION RATE (TLHGR) is not violated for any power distribution.
This is accomplished using FDLRC. The scram setting is decreased in accordance with the formula in Specification 3.11.B, when FDLRC is greater than 1.0.
i 8
The adjustment may also be accomplished by increasing the gain of the APRM by FDLRC.
This provides the same degree of protection as reducing the trip setting by 1/FDLRC by raising the initial APRM reading closer to the trip setting such that a scram would be received at the same point in a transient as if the trip setting had been reduced.
The added paragraphs provide clarification of LC0 Action Statements 3.11.8.2 and 3.11.B.3.
Therefore the above addition of paragraphs clarifies the SPC methodology and is acceptable.
In section 3/4.11.C, the licensee proposed minor editorial changes to second paragraph.
These changes affect the first two sentences and are as follows:
To assure that the fuel cladding integrity Safety Limit is not exceeded during any anticipated abnormal operational transient, the most limiting transients are analyzed to determine which result in the largest reduction in the CRITICAL POWER RATIO (CPR). The type of transients evaluated are change of flow, increase in pressure and power, positive reactivity insertion, and coolant temperature decrease.
Furthermore, the fourth paragraph is replaced which again clarifies the SPC methodology which uses four scram insertion points to calculate MCPR Operating Limit and MCPR Safety Limit:
MCPR Operating Limits are presented in the CORE OPERATING LIMITS REPORT (COLR) for both Nominal Scram Speed (NSS) and Technical Specification Scram Speed (TSSS) insertion times. The negative reactivity insertion rate resulting from the scram plays a major role in providing the required protection against violating the Safety Limit MCPR during transient events.
Faster scram insertion times provide greater protection and allow for improved MCPR performance. The application of NSS MCPR limits utilizes measured data that is faster than the times required by the Technical Specifications, while the TSSS MCPR limits provide the necessary protection for the slowest allowable average scram insertion times identified in Specification 3.3.E.
The measured scram insertion times are compared with the nominal scram insertion times and the Technical Specification Scram Speeds. The appropriate operating limit is applied, as specified in the COLR.
For core flows less than rated, the MCPR Operating Limit established in the specification is adjusted to provide protection of the Safety Limit MCPR in the event of an uncontrolled recirculation flow increase to the l
physical limit of the pump.
Protection is provided for manual and automatic flow control by applying the appropriate flow dependent MCPR limits presented in the COLR. The MCPR Operating Limit for a given power / flow state is the greater value of MCPR as given by the rated conditions MCPR limit or the flow dependent MCPR limit.
For automatic flow control, in addition to protecting the Safety Limit MCPR during the flow run-up event, protection is provided to prevent exceeding the rated flow MCPR Operating Limit during an automatic flow increase to rated core flow.
The proposed change appropriately reflects the NRC-approved SPC methodology
9 and does not change the current requirement that MCPR meet the limits specified in the COLR.
Therefore, the proposed change is acceptable.
2.8 Reactor Core TS 5.3.A, Fuel Assemblies, provides a description of the fuel assemblies.
The licensee proposed to expand this description to be consistent with Improved Standard Technical Specifications, NUREG-1433 Rev.1, and to better reflect the ATRIUM-9B design. The revised description includes a discussion of the use of water rods or water boxes which is consistent with the SPC fuel design, and replaces " zirconium alloy" with "Zircaloy or Zirlo." The proposed change accurately describes the SPC fuel design, is consistent with NUREG-1433, Rev. 1, and does not affect any current TS requirements. Therefore, the proposed change is acceptable.
2.9 Reactor Coolant System TS 5.4 describes the design pressure, temperature, and volume of the reactor coolant system. The licensee proposed to relocate the contents of specification 5.4 to the UFSAR.
Page 5-6 and Table of Contents page XIV are modified to read, "[ INTENTIONALLY BLANK)." This proposed change is consistent with Improved Standard Technical Specifications, NUREG-1433, Rev.1, and is acceptable.
2.10 Egoortina Reauirements TS 6.9 requires that in addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the identified reports shall be submitted to the Regional Administrator of the appropriate Regional Office of 1
the NRC unless otherwise noted.
TS 6.9.A.6.a(4) describes the MCPR limit in the COLR. The licensee proposed to delete the 20% in the statement " including i
[
20% scram insertion time" to reflect the SPC methodology.
The proposed change will state " including scram insertion time." This reflects the current SPC j
methodology and is acceptable.
TS 6.9.A.6.b lists the analytical methods used to determine the operating limits that are previr'isly reviewed and approved by the NRC in the latest approved revision or Lupplement of topical reports.
The licensee proposed to include references to the following topical reports which are used to determine the core operating limits:
(5) Advanced Nuclear Fuels Methodology for Boiling Water Reactors, XN-NF-80-19(P)(A), Volume 1, Supolement 3, Supplement 3 Appendix F, and Supplement 4, Advanced Nuclear Fuels Corporation, November 1990.
(6) Exxon Nuclear Methodology for Boiling Water Reactors: Application of the ENC Methodology to BWR Reloads, XN-NF-80-19(P)(A), Volume 4, Revision 1, Exxon Nuclear Company, June 1986.
(7) Exxon Nuclear Methodology for Boiling Water Reactors THERMEX:
Thermal Limits Methodoingy Summary Description, XN-NF-80-19(P)(A),
Volume 3, Revision 2, Exxon Nuclear Company, January 1987.
2 10 (8) Exxon Nuclear Methodology for Boiling Water Reactors - Neutronic Methods for Design and Analysis, XN-NF-80-19(P)(A), Volume 1 and Supplements 1 and 2, Exxon Nuclear Company, March 1983.
(9) Generic Mechanical Design for Exxon Nuclear Jet Pump BWR Reload Fuel, XN-NF-85-67(P)(A) Revision 1, Exxon Nuclear Company, September 1986.
(10) Qualification of Exxon Nuclear Fuel for Extended Burnup Supplement 1:
Extended Burnup Qualification of ENC 9x9 BWR Fuel, XN-NF-82-06(P)(A) Supplement 1, Revision 2, Advanced Nuclear Fuels
{
Corporation, May 1988.
(11) Advanced Nuclear Fuels Corporation Generic Mechanical Design for Advanced Nuclear Fuels Corporation 9x9-IX and 9x9-9X BWR Reload Fuel, ANF-89-014(P)(A), Revision I and Supplements 1 and 2, Advanced Nuclear Fuels Corporation, October 1991.
(12) Generic Mechanical Design Criteria for BWR fuel Designs, ANF-89-98(P)(A), Revision 1 and Revision 1 Supplement 1, Advanced Nuclear Fuels Corporation, May 1995.
(13) Exxon Nuclear Plant Transient Methodology for Boiling Water Reactors, XN-NF-79-71(P)(A), Revision 2 Supplements 1, 2, and 3, Exxon Nuclear Company, March 1986.
(14) ANFB Critical Power Correlation, ANF-ll25(P)(A) and Supplements 1 and 2, Advanced Nuclear Fuels Corporation, April 1990.
(15) Advanced Nuclear Fuels Corporation Critical Power Methodology for Boiling Water Reactors / Advanced Nuclear Fuels Corporation Critical Power Methodology for Boiling Water Reactors: Methodology for Analysis of 1
Assembly Channel Bowing Effects /NRC Correspondence, ANF-524(P)(A),
Revision 2, Supplement 1 Revision 2, Supplement 2, Advanced Nuclear Fuels Corporation, November 1990.
(16) COTRANSA 2: A Computer Program for Boiling Water Reactor Transient Analyses, ANF-913(P)(A) Volume 1 Revision 1 and Volume 1 Supplements 2, 3, and 4, Advanced Nuclear Fuels Corporation, August 1990.
i (17) Advanced Nuclear Fuels Corporation Methodology for Boiling Water i
Reactors EXEM BWR Evaluation Model, ANF-91-048(P)(A), Advanced Nuclear Fuels Corporation, January 1993.
(18) Commonwealth Edison Topical Report NFSR-0091, " Benchmark of CASM0/MICR0 BURN BWR Nuclear Design Methods," Revision 0, Supplements 1 and 2, December 1991, March 1992, and May 1992, respectively; SER letter dated March 22, 1993.
The additional topical reports are those used in SPC methodology and have been approved by the NRC for use at Quad Cities. The staff finds this change acceptable because the use of identified NRC-approved methodologies will ensure that the values for cycle-specific parameters are determined consistent
._.m-s s.
i s'
11 f
with applicable design bases and safety limits, and assist safe operation of the facility.
3.0 Conclusion i
Comed requested changes to the Quad Cities Nuclear Power Station, Units 1.and l
2, TS which would incorporate NRC approved thermal' limit licensing methodology in the list of. approved methodologies used in establishing the cycle specific r
thermal limits.
Other minor editorial changes were also proposed.
The staff concluded that these TS revisions are compatible with the STS, and SPC i
methodology.
Based on the above, the-staff concluded that operation in the proposed manner will not endanger the health and safety of the public and the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributor:
K. Kavanach Dated:
3/13/97 i
I I
.