ML20141K505
| ML20141K505 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 05/23/1997 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20141K499 | List: |
| References | |
| NUDOCS 9705290185 | |
| Download: ML20141K505 (14) | |
Text
,
I nnuqk UNITED STATES p
l NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 30866-4001
'+n..... s o
i c
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 177 TO FACILITY OPERATING LICENSE NO. DPR-29 AND AMENDMENT NO. 175 TO FACILITY OPERATING LICENSE NO. OPR-30 l
COMMONWEALTH EDISON COMPANY AliQ MIDAMERICAN ENERGY COMPANY OVAD CITIES NUCLEAR POWER STATION. UNITS 1 AND 2 DOCKET NOS. 50-254 AND 50-265 I
1.0 INTRODUCTION
1 By letter dated June 10, 1996, as supplemented by letter dated February 17, 1997, Commonwealth Edison Company (Comed, the licensee) requested changes to the Quad Cities Nuclear Power Station, Units 1 and 2, Technical Specifications i
(TS).
Quad Cities, Units 1 and 2, currently use General Electric (GE) fuel l
and licensing methodologies.
Siemens Power Corporation (SPC) fuel and licensing methodologies tre planned for use at Quad Cities beginning with Unit 2 Cycle 15 and Unit 1 Cycle 16.
The Siemens' loss-of-coolant accident (LOCA) l methodology and fuel a.sse:mbly designs are approved for use at other licensed l
boiling uter reactor (CMR) 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 also required for the approval of TS changes for SPC fuel transition for LaSalle County Station, Units 1 and 2.
The revision lists the specific NRC approval l
date and the revision / supplement for each of the new topical reports, and revises the Section 5.3.A description of fuel assemblies. This letter provided additional clarifying information that did not change the initial proposed no significant hazards consideration determination.
2.0 EVALUATION 2.1 Mechanical Desion The ATRIUM-98 fuel design is a 9x9 lattice design with an internal water box to enhance neutron moderation. The ATRIUM-9B fuel design was analyzed and assessed by Siemens according to the approved methodology, entitled " Generic Mechanical Design for Advanced Nuclear Fuels 9x9-IX and 9x9-9X BWR Reload j
Fuel," ANF-89-014(P)(A), Revision 1 Supplement 1.
The ATRIUM-9B fuel 9705290185 970523 U
DR ADOCK 050002 4
l s
i I
mechanical design followed the approved methodology and, therefore, is acceptable for Quad Cities Nuclear Power Station, Unit 2 Cycle 15.
2.2 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:
FUEL DESIGN LIMITING RATIO FOR CENTERLINE MELT (FDLRC)
The FUEL DESIGN 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).
FUEL DESIGN LIMITING RATIO (FDLRX) l The FUEL DESIGN LIMITING RATIO (FDLRX) shall be the limit used to assure that the fuel operates within the end-of-life steady-state design l
criteria by, among other items, limiting the release of fission gas to l
the cladding plenum (applicable to SPC fuel).
1 The licensee has proposed to delete the definition of Rod Density.
Rod density will be replaced by critical control rod configuration in order to l
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.
The proposed definition is as follows:
I TRANSIENT LINEAR HEAT GENERATION RATE (TLHGR)
The TRANSIENT LINEAR HEAT GENERATION RATE (TLHGR) limit protects against fuel centerline melting and 1 percent 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 TS Bases will be modified to show applicability to both GE and SPC fuel.
The staff finds these definition changes acceptable.
i 2.3 Safety Limits Bases The licensee proposed an editorial change to 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:
l l
i i
-3_
b The fuel cladding integrity limit is set such that no fuel damage is calculated to occur as a result of an A00.
i The staff concludes that the change clarifies the meaning of the sentence and j
is acceptable.
In Section 2.1.8, Thermal Power, High Pressure and High Flow, the licensee
?
proposed editorial changes to paragraph one. The current wording of the last a
j 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 percent 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 utences of paragraph one consist of the i
following:
1 i
Therefore, the fuel cladding integrity Safety Limit is defined such that, with the limiting fuel assembly operating at the MCPR Safety j
Limit, more than 99.9% rf the fuel rods in the core are expected to avoid boiling transition. This includes consideration of the power l
distribution within the core and all uncertainties.
The staff notes that this wording is consistent with the bases in Improved Standard Technical Specifice.t. ions, NUREG-1433, Revision 1 and, therefore, it is acceptable.
2.4 Limitina Safety System Settinas Bases In Section 2.2.A.1, Reactor Protection System Instrumentation Setpoints -
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 percent of rated power, thus maintaining minimum critical power ratio (MCPR) above the fuel cladding integrity Safety Limit." The licensee proposed to change 1 percent to 7.7
)
percent to reflect the correct value in the Updated Final Safety Analysis Report (UFSAR) and Safety Analysis Report (SAR).
Section 7.6.1 of the UFSAR cites, in graphical form, 7.7 percent 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, thus maintaining MCPR above the fuel cladding integrity Safety Limit.
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
l p
_4_
l l
i 360 inches above vessel zero." This statement is consistent with footnotes and other sections of the bases and, therefore, is acceptable.
1 2.5 Instrumentation Bases The licensee proposed a clarification to Sectico 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 with 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 l
associated with these events.
The proposed additions provide a clear definition and use of the top of active l
fuel reference point. The staff finds this addition to the bases acceptable.
2.6 Reactivity 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 R00 DENSITY shall not exceed 1 percent Ak/k.
In addition, Surveillance Requirement 4.3.B requires that the reactivity equivalence of the difference between the actual R0D DENSITY and the predicted R00 DENSITY shall be verified to be less than or equal to 1 percent Ak/k. This limit ensures that plant operation is maintained within the assumptions of the safety analyses.
l The licensee has proposed to replace " ROD DENSITY" in both the TS and the surveillance requirement with " critical control rod configuration." This proposed change is necessitated by the additional proposal to include an alternative to monitoring reactivity anomalies in the TS basss. Both the SPC core monitoring code, Powerplex, and the GE Core Monitoring Code (CMC) l provides the capability to monitor actual K,,ity to critical conh'o.
versus predicted K The licensee stated that the change from Rod Dens l rod configuration was necessary in order to use this capability.
The staff notes j
that this method is currently used at Dresden to monitor reactivity anomalies.
l Thus, the following will be added to Section 3/4.3.B, Reactivity Anomalies l
Bases:
i Alternatively, monitored K
. calculated by an approved Y# can be compared with the predicted K,, as D core simulator code.
knen 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 j
measured reactivity is convenient under such a balance, since parameters are being maintained relatively stable under steady state power conditions. The staff notes that this proposed change only revises the current method of
~
l 4
measuring the difference between predicted and monitored core reactivity and does not change the required limit; therefore, the change 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 Rod 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 j
data);"
i second paragraph: "In the statistical treatment of the~ limiting transients, a i
statistical distribution of total scram delay is used rather l
than the bounding value described above;"
i 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 i
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 not changed significantly;" and fourth paragraph: "If test results should be determined to fall outside of the
. statistical population defining the scram performance i
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 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 a GE methodology.
Current SPC methods used to evaluate the 5 percent, 20 percent, 50 percent and 90 percent 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 j
determine the operating limit for MCPR.
Following completion of each set of scram testing, the results will be compared with the assumptions L
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 t
l insertion times.
l l
. r
~. _ _
_.m m _
i 1.
l-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 i
i Ratio, requires that the MCPR shall be equal to or greater than the MCPR 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 l
meet TS 3.11.C.
Based on this information, the changes to Section 3/4.3.0, 3.E, and 3.F bases are acceptable.
l 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 i
control rod segments which are with# awn 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 result in a passage that reads as follows:
l 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 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.
l The licensee also proposed editorial changer to the last sentence of the third paragraph.
The last scubnce will be replaced b.y the following:
The methodology used for the control r)d drop accident analysis is NRC-approved and is part of the license bases referen.:ed 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 /gm limit up to the Low Power Set Point (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 editorial changes acceptable.
i 2.7 Primary System Boundary Bases
-In Sections 3/4.6.E and 3/4.6.F, Safety Valves and Relief Valves, the licensee p
proposed to add the following sentence to the middle of the first paragraph:
. ~. _ _
l l
SPC methodology determines the most limiting pressurization transient l
each cycle.
The addition of this statement clarifies the SPC methodology for analyzing the i
overaressurization event and, therefore, is acceptable.
2.8 Powar Distribution Limits - Limitino Conditions For Operation And Bases
?
As stated above, FLPD and MFLPD are LHGR terms that 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 l
f from off rated transients by the application of FDLRC to'the Average Power Range Monitor (APRM) setpoints.
Based on this, the licensee proposed to i
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 proposed change to TS 3/4.11.B is l
identical to Dresden, Units 2 and 3, TS 3/4.11.8 except for the fol. lowing:
l
- 1) footnote (a) is added to the appropriate FDLRC statements. and 2) the current footnote (a) will become footnote (b).
l The proposed footnote (a) will state the following:
For GE fuel, MFLPD/FRTP is substituted for FDLRC. Ad.justments are based i
on the lowest APRM setpoint or highest APRM reading resulting from the l
two limits.
The staff notes that TS 3/4.11.B will be titled Transient Linear Heat l
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 l
the proposed changes above.
Since the TLHGR and FDLRC limits for SPC fuels l
are applied to the APRM setpoints, the staff finds the propose changes to TS 3/4.11.B acceptable, i
l The licensee also proposed changes to the TS Bases Sections 3/4.11.A, 3/4.11.B l
and 3/4.ll.C, Average Planar Linear Heat Generation Rate (APLHGR), APRM i
Setpoints and Minimum Critical Power Ratio, in order to provide clarification of the SPC methodology for the application of thermal limits.
TS 3.11.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.
l For Section 3/4.11.A, the licensee proposed the following changes:
1) relocate the last two paragraphs of Section 3/4.11.A on Bases page 83/4.11-1 to the beginning of Section 3/4.11.A, 2) insert "GE Fuel" in front of the current first paragraph, and 3) add the following paragraphs to describe the i
SPC methodology:
I
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 folicwing 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 3/4.11.A Bases changes described above acceptable.
TS 3.11.B requires, based on the proposed change discussed above, that the TLHGR shall be maintained such that the 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.
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.B.2 and 3.ll.B.3.
Therefore, the addition of the above paragraphs clarifies the SPC methodology and is acceptable.
i l
l In Section 3/4.11.C, the licensee proposed minor editorial changes to the 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, increate in pressure and power, positive j
reactivity insertion, and coolant temperature decrease.
The licensee proposed to replace the fourth paragraph to again clarify the SPC methodology, which uses four scram insertion points to calculats 3PR Operating Limit and hCPR 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 1
insertion times identified in Specification 3.3.E.
The measured scram 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 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 autor cic 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 and does not change the current requirement that MCPR meet the limits specified in the COLR. Therefore, the proposed change is acceptable, i
l l
a
- 2.9 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, Revision 1, and to better reflect the ATRIUM-98 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." Upon further review and discussions with Comed on May 22, 1997, it was determined that to be consistent with Improved Standard Technical Specifications, NUREG-1433, Revision 1, and provide a clarification of what the substitutions will be for, additional words needed to be added. The words " filler rods for fuel rods" had been omitted from the amendment request.
The sentence was changed, with approval from Comed, to read, " Limited substitutions of Zircaloy or ZIRLO filler rods for fuel rods, in accordance with NRC-approved applications of fuel rod configurations, may be used." This change was approved in the NRC safety evaluation (SE) related to Amendment No.173 issued for Unit 2 on May 2, 1997, and in the NRC SE related to Amendment No. 174 issued for Unit 2 on May 22, 1997.
The proposed change accurately describes the SPC fuel design, is consistent with NUREG-1433, Revision 1, includes the cladding material cited in 10 CFR 50.44 and 50.46, and does not affect any current TS requirements. This change will make TS Section 5.3.A appropriate for Unit 1 and 2 and deletes Unit 2 specific page 5-Sa; therefore, the proposed change is acceptable.
2.10 Feactor Coolant System Section 182a of the Atomic Energy Act of 1954, as amended, requires applications for nuclear power plant operating licenses to include TSs as part of the license.
The Commission's regulatory requirements related to the content of TS are set forth in 10 CFR 50.36.
That regulation requires that the TS include items in five specific categories, including (1) safety limits, limiting safety system settings, and limiting control settings; (2) limiting condit:ons for operation; (3) surveillance requirements; (4) design features; and (5) administrative controls. However, the regulation does not. specify the particular requirements to be included in a plant's TS.
The Commission has provided guidance for the contents of TS in its " Final Policy Statement on Technical Specification Improvements for Nuclear Power Reactors" 58 FR 39132 (July 22, 1993), in which the Commission indicated that compliance with the Final Policy Statement satisfies Section 182a of the Act.
In particular, the Commission indicated that certain items could be relocated from the TS to licensee-controlled documents, consistent with the standard enunciated in Portland General Electric Co. (Trojan Nuclear Plant), ALAB-531, 9 NRC 263, 273 (1979).
In that case, the Atomic Safety and Licensing Appeal Board indicated that "techni:al specifications are to be reserved for those matters as to which the imposition of rigid conditions or limitations upon reactor operation is deemed oecessary to obviate the possibility of an abnormal situation or event giving rise to an immediate threat to the public health and safety."
i
l l.
The Final Policy Statement identifien four criteria to be used in determining l
whether a particula ~r matter is required to be included in the TS limiting conditions for operation, as follovs:
(1) installed instrumentation that is l
used to detect, and indicate in tha control room, a significant abnormal degradation of the reactor coolant pressure boundary; (2) a process variable, design feature, or operating restriction that is an initial condition of a design-basis accident or trans19 analysis that either assumes the failure of or presents a challenge to the ica grity of a fission product barrier; (3) a l
structure, system, or component that is part of a primary success path and which functions or actuates to mitigate a design-basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier; (4) a structure, system, or component which operating experience or probabilistic safety assessment has shown to be significant to public health and safety. As a result, existing TS j
requirements which fall within or satisfy any of the criteria in the Final Policy Statement must be retained in the TS, while those TS requirements which l
do not fall within or satisfy these criteria may be relocated to other l
licensee-controlled documents. The Commission recently amended 10 CFR 50.36 to codify and incorporate these four criteria (60 FR 36953).
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]."
Design temperatures, pressures, and volumes of the Reactor Coolant System in existing TS Section 5.4 will be detailed in the Updated Final Safety Analysis Report (UFSAR).
Changes to these facility design parameters are controlled by l
the requirements of 10 CFR 50.59.
Furthermore, these design parameters are encompassed by existing TS Limiting Condition for Operation (LCOs) that establish acceptable requirements for ensuring that performance of the reactor coolant system is maintained. Any changes to the LCOs would receive prior NRC l
review and approval.
Since the features with a potential to impact safety are l
sufficiently addressed by LCOs, and since design features, if altered in t
accordance with 10 CFR 50.59, would not result in a significant impact on safety, the criteria of 10 CFR 50.36(c)(4) for including the above design features in the TS are not met.
The above relocated requirements relating to design features are not required to be in the TS under 10 CFR 50.36, and are not required to obviate the possibility of an abnormal situation or event giving rise to an immediate threat to public health and safety.
In addition, the staff finds that sufficient regulatory controls exist under 10 CFR 50.59 to assure continued protection of public health and safety. This proposed change is consistent with Improved Standard Technical Specifications, NUREG-1433, Revision 1, and is acceptable.
The Additional Condition in Appendix C of the license will valuate the acceptability of removal of the contents of Specification 5.4.
Accordingly, the staff has concluded that these requirements may be relocated from the TS to the licensee's UFSAR.
l l
l l
l' 2.11 Reportina Reauirements TS 6.9 requires that, in addition to the applicable reporting requirements of Title 10, Code of Federal Reaulations, the identified reports shall be submitted to the Regional Administrator of the appropriate Regional Office of
. the NRC unless otherwise noted.
TS 6.9. A.6.a(4) describes the MCPR limit in i
.the COLR. The. licensee proposed to delete the 20 percent in the statement
" including 20 percent scram insertion time" to reflect the SPC methodology.
The proposed change will state " including scram insertion time." This reflects the current SPC methodology and is acceptable, h
TS 6.9.A.6.b lists the analytical methods used to determine the operating i
limits that are previously reviewed and approved by the NRC in the latest approved revision or supplement of topical reports. The licensee proposed to include references to the list of topical reports which are used to determine the core operating limits by adding the following:
(5) Advanced Nuclear Fuels Methodology for Boiling Water Reactors, XN-NF-80-19(P)(A), Volume 1, Supplement 3, Suppiement 3 Appendix F, and Supplement 4, Advanced Nuclear Fuels Corporation, November 1990.
(6) Exxon Nuclear Methodology for Boiling Water Reactors: Application l
of the ENC Methodology to BWR Reloads, XN-NF-80-19(P)(A), Volume 4, Revision 1, Exxon Nuclear Company, June 1986.
i (7) Exxon Nuclear Methodology for Boiling Water Reactors THERMEX:
Thermal Limits Methodology Summary Description, XN-NF-80-19(P)(A),
l Volume 3, Revision 2, Exxon Nuclear Company, January 1987.
iL (8) Exxon Nuclear Methodology for Boiling Water Reactors - Neutronic Methods for Design and Analysis, XN-NF-80-19(P)(A), Volume 1 and Supplements I and 2, Exxon Nuclear Company, March 1983, 1
(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.
l (10) Qualification of Exxon Nuclear Fuel for Extended Burnup l
Supplement 1:
Extended Burnup Qualification of ENC 9x9 BWR Fuel, l
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 9x9-lX and 9x9-9X BWR Reload Fuel, ANF 014(P)(A), Revision 1 and Supplements 1 and 2, Advanced Nuclear Fuels Corporation, October 1991.
{
(12) Generic Mechanical Design Criteria for BWR Fuel Designs, i
ANF-89-98(P)(A), Revision 1 and Revision 1 Supplement 1, Advanced i
Nuclear Fuels Corporation, May 1995.
o
- (13) Exxon Nuclear Plant Transient Methodology for Boiling Water Reactors, XN F 71(P)(A), Revision 2 Supplements 1, 2, and 3, Exxon Nuclear Cc w March 1986.
(14) ANFB Critical Power Correlation, ANF-1125(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 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.
(17) Advanced Nuclear Fuels Corporation Methodology for Boiling Water 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.
t The additional topical reports are those used in SPC methodology and have been approved by the NRC and are appropriate for the Quad. Cities plant design and
)
are acceptable for use.
References (5), (6), (7), and (8) in the current TS l
for Unit 2 on page 6-16a were added in Amendment No. 173.
These references are now items (5), (11), and'(12) on new page 6-16 and (18) on new page 6-16a, for Units I and Unit 2.
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 with applicable design bases and safety limits (e.g., fuel thermal and mechanical limits, core L
thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin and transient and accident analysis limits) and assist safe operation of the facility.
2.12 Conclusion Comed requested changes to the Quad Cities Nuclear Power Station, Units 1 and 2, TS which would incorporate NRC-approved thermal limit licensing methodology in the list of approved methodologies used in establishing the i
cycle specific thermal limits.
Other minor editorial changes were also proposed.
The staff concluded that these TS revisions are compatible with the STS, and SPC methodology.
Based on the above, the staff concluded that i
operation in the proposed manner will not endanger the health and safety of i
o
' proposed. The staff concluded that these TS revisions are compatible with the STS, and SPC 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 comon defense and security or to the health and safety of the public.
3.0 STATE CONSULTATION
In accordance with the Comission's regulations, the Illinois State official was notified of the proposed issuance of the amendments. The State official had no coments.
4.0 ENVIRONMENTAL CONSIDERATION
The amendments change a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and change surveillance requirements.
The NRC staff has determined that the amendments involve no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Comission has previously issued a proposed finding that the amendments involve no significant bazards consideration, and there has been no public coment on such finding (61 FR 44355). Accordingly, the amendments meet the eligibility criteria for l
categorical exclusion set forth in 10 CFR 51.22(c)(9). The amendment also relates to changes in record keeping, reporting or administrative procedures or requirements. Accordingly, with respect to these items, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(10).
Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.
5.0 CONCLUSION
The Comission has concluded, based on the considerations discussed above, that:
(1) there is reasonable assurance that the health and safety of the l
public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendments will not be inimical to the comon defense and security or to the health and safety of the public.
Principal Contributor:
K. Kavanagh Date: May 23, 1997
_